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23rd Symposium on Fusion Technology

20 - 24 September 2004 - Fondazione Cini, Venice, Italy

Book of Abstracts

Consorzio RFX – EURATOM - ENEA

PROGRAM

ORAL SESSIONS

  • O1-A Arazzi Hall Monday, 20th 14:00-15:30

  • O1-B Barbantini Hall Monday, 20th 14:00-15:30

  • O2-A Arazzi Hall Tuesday, 21th 09:00-09:30

  • O2-B Barbantini Hall Tuesday, 21th 09:00-09:30

  • O3-A Arazzi Hall Tuesday, 21th 14:00-15:30

  • O3-B Barbantini Hall Tuesday, 21th 14:00-15:30

  • O4-A Arazzi Hall Thursday, 23th 14:00-15:30

  • O4-B Barbantini Hall Thursday, 23th 14:00-15:30

POSTER SESSIONS

  • P1-C Capriate Hall Monday, 20th 16:00-18:00 H-K-I

  • P1-T Tipografia Hall Monday, 20th 16:00-18:00 I

  • P2-C Capriate Hall Tuesday, 21th 16:00-18:00 D

  • P2-T Tipografia Hall Tuesday, 21th 16:00-18:00 E

  • P3-C Capriate Hall Wednesday, 22th 11:30-13:30 A-C

  • P3-T Tipografia Hall Wednesday, 22th 11:30-13:30 B

  • P4-C Capriate Hall Thursday, 23th 16:00-18:00 F

  • P4-T Tipografia Hall Thursday, 23th 16:00-18:00 G-J

Paper Code

P 4 B – J – 491

ID NUMBER

TOPIC

HALL (A=Arazzi B= Barbantini C =Capriate T= Tipografia

SESSION

CATEGORY (P=Poster O=Oral)

Oral presentations are highlighted in red.

INDEX

PROGRAM 2

- A - Current and Next Step Devices 15

P3C-A-11 SELECTION OF DESIGN SOLUTIONS AND FABRICATION METHODS AND SUPPORTING R&D FOR PROCUREMENT OF ITER VESSEL AND FW/BLANKET 15

P3C-A-16 THE PROTO-SPHERA LOAD ASSEMBLY 16

P3C-A-72 OVERVIEW OF THE DIII–D PROGRAM AND CONSTRUCTION PLANS* 17

P3C-A-90 COMMISSIONING AND PRELIMINARY OPERATION OF THE HL-2A TOKAMAK 18

P3C-A-144 PLASMA PHYSICS BASIS AND OPERATIONS OF FUSION-DRIVEN SUBCRITICAL SYSTEM 19

P3C-A-184 THE WENDELSTEIN 7-X MECHANICAL STRUCTURE SUPPORT ELEMENTS: TESTS AND DESIGN 20

P3C-A-203 LEVITATION EXPERIMENTS OF A HIGH TEMPERATURE SUPERCONDUCTOR COIL IN THE INTERNAL COIL DEVICE MINI-RT 21

P3C-A-245 EXPERIMENTAL STUDY OF WATER FLOW DISTRIBUTION INSIDE TWO-CHANNEL MODEL OF ITER VACUUM VESSEL COOLING SYSTEM 22

P3C-A-262 MAGNUM-PSI, A PLASMA GENERATOR FOR PLASMA-SURFACE INTERACTION RESEARCH IN ITER-LIKE CONDITIONS 23

P3C-A-330 THE LASER MÉGAJOULE (LMJ) PROJECT DEDICATED TO INERTIAL CONFINEMENT FUSION : DEVELOPMENT AND CONSTRUCTION STATUS. 24

P3C-A-449 JET ENGINEERING: PROGRESS AND PLANS 25

P3C-A-468 THE JET-ENHANCED PERFORMANCE PROGRAM: MORE HEATING POWER AND DIAGNOSTIC CAPABILITIES IN PREPARATION FOR ITER 26

P3C-A-518 NUCLEAR ANALYSES OF SOME KEY ASPECTS OF THE ITER DESIGN WITH MONTE CARLO CODES 27

P3C-A-522 TRANSPORT, LOGISTICS AND PACKAGING OF ITER COMPONENTS 28

P3C-A-523 STUDIES FOR SITE PREPARATION FOR ITER CONSTRUCTION 29

P3C-A-531 READINESS OF CADARACHE FOR STARTING ITER CONSTRUCTION 30

- B - Plasma Heating and Current Drive. 31

P3T-B-19 THE ALCATOR C-MOD LOWER HYBRID CURRENT DRIVE EXPERIMENT TRANSMITTER 31

P3T-B-25 DESIGN OF AN ULTRA-BROADBAND SINGLE-DISK OUTPUT WINDOW FOR A FREQUENCY STEP-TUNABLE 1 MW GYROTRON 32

P3T-B-51 EXPERIMENTS ON A 170 GHZ COAXIAL CAVITY GYROTRON 33

P3T-B-78 THE UPGRADE OF THE DIII-D EC SYSTEM USING 120 GHZ ITER GYROTRONS 34

P3T-B-82 THE LHCD LAUNCHER FOR ALCATOR C-MOD – DESIGN, CONSTRUCTION, CALIBRATION AND TESTING* 35

P3T-B-94 DESIGN AND OPERATION OF THE WENDELSTEIN 7-X ECRH HIGH VOLTAGE POWER SUPPLIES 36

P3T-B-113 THERMAL ANALYSIS AND OHMIC LOSS ESTIMATION OF POLARIZER FOR ITER ECCD SYSTEM 37

P3T-B-152 TESTS OF LOAD-TOLERANT EXTERNAL CONJUGATE-T MATCHING SYSTEM FOR A2 ICRF ANTENNA AT JET 38

P3T-B-156 NEUTRONIC ANALYSIS OF ITER NEUTRAL BEAM TEST BED 39

P3T-B-160 A REVIEW OF JET NEUTRAL BEAM SYSTEM PERFORMANCE 1994 TO 2003 40

P3T-B-165 DEVELOPING A FULL SCALE ECRH MM-WAVE LAUNCHING SYSTEM MOCK-UP FOR ITER 41

P3T-B-171 DIGITAL MOCK-UP DESIGN OF THE REMOTE STEERABLE ITER ECRH LAUNCHING SYSTEM 42

P3T-B-172 PRE STUDY RESULTS ON HIGH VOLTAGE SOLID-STATE SWITCHES FOR GYROTRON PROTECTION. 43

P3T-B-173 AN ALTERNATIVE ECRH FRONT STEERING LAUNCHER FOR THE ITER UPPER PORT 44

P3T-B-174 DESIGN OF THE MM-WAVE SYSTEM OF THE ECRH UPPER LAUNCHER FOR ITER 45

P3T-B-175 DEVELOPING THE NEXT LHCD SOURCE FOR TORE SUPRA 46

P3T-B-180 TOWARD AN LHCD SYSTEM FOR ITER 47

P3T-B-181 A N-PORT ERROR MODEL AND CALIBRATION PROCEDURE FOR MEASURING THE SCATTERING MATRICES OF LOWER-HYBRID MULTIJUNCTIONS 48

P3T-B-182 DESIGN AND FABRICATION OF THE "ITER-LIKE" SINGAP D¯ ACCELERATION SYSTEM 49

P3T-B-185 OPERATIONAL EXPERIENCE WITH UPGRADED JET NEUTRAL BEAM INJECTION SYSTEMS 50

P3T-B-187 MAST NEUTRAL BEAM LONG PULSE UPGRADE 51

P3T-B-188 THE ITER NEUTRAL BEAM TEST FACILITY : DESIGNS OF THE GENERAL INFRASTRUCTURE, CRYOSYSTEM AND COOLING PLANT 52

P3T-B-201 PROGRESS OF THE KSTAR ICRF COMPONENTS DEVELOPMENT FOR LONG PULSE OPERATION 53

P3T-B-204 RECENT PROGRESS OF NEGATIVE ION BASED NEUTRAL BEAM INJECTOR FOR JT-60U 54

P3T-B-210 TESTS AND FIRST RESULTS OF A LOAD RESILIENT ICRH ANTENNA ON TEXTOR 55

P3T-B-211 REALISATION OF A TEST FACILITY FOR THE ICRH ITER PLUG-IN BY MEANS OF A MOCK-UP WITH SALTED WATER LOAD 56

P3T-B-218 STUDY OF MUTUAL COUPLING EFFECTS IN THE ANTENNA ARRAY OF THE ICRH PLUG-IN FOR ITER 57

P3T-B-221 STATUS AND PLANS FOR THE DEVELOPMENT OF AN RF NEGATIVE ION SOURCE FOR ITER NBI 58

P3T-B-225 DEVELOPMENT AND CONTRIBUTION OF RF HEATING AND CURRENT DRIVE SYSTEMS TO LONG PULSE, HIGH PERFORMANCE EXPERIMENTS IN JT-60U 59

P3T-B-227 RF-SOURCE DEVELOPMENT FOR ITER: LARGE AREA H- BEAM EXTRACTION, MODIFICATIONS FOR LONG PULSE OPERATION AND DESIGN OF A HALF SIZE ITER SOURCE 60

P3T-B-229 DIAGNOSTICS AND MODELING OF THE PLASMA IN BATMAN RADIO FREQUENCY ION SOURCE 61

P3T-B-246 ECH MW-LEVEL CW TRANSMISSION LINE COMPONENTS SUITABLE FOR ITER 62

P3T-B-267 STATUS OF THE TJ-II ELECTRON BERNSTEIN WAVES HEATING PROJECT 63

P3T-B-283 THE ASDEX UPGRADE ICRF SYSTEM: OPERATIONAL EXPERIENCE AND DEVELOPMENTS 64

P3T-B-300 COOLING CONCEPTS OF THE ECRH LAUNCHER STRUCTURE AND THE TORUS WINDOWS 65

P3T-B-310 THE DESIGN OF THE CONTROL SYSTEM FOR THE NEUTRAL BEAM INJECTION IN HT-7 66

P3T-B-312 EXPERIMENTAL STUDY ON UNIFORMITY OF H- ION BEAM IN A LARGE NEGATIVE ION SOURCE 67

P3T-B-314 DEVELOPMENT OF RELIABLE DIAMOND WINDOW FOR EC LAUNCHER ON FUSION REACTORS 68

P3T-B-320 DESIGN OF HIGH POWER COAXIAL DC BREAK FOR ADITYA TOKAMAK 69

P3T-B-332 W7-X NEUTRAL-BEAM-INJECTION: TRANSMISSION, POWER-LOAD TO THE DUCT AND INNER VESSEL AND CONSEQUENCES OF THE STELLARATOR STRAY FIELD 70

P3T-B-337 DIAGNOSTICS OF THE CESIUM AMOUNT IN A RF NEGATIVE ION SOURCE AND THE CORRELATION WITH THE EXTRACTED CURRENT DENSITY 71

P3T-B-345 LONG PULSE OPERATION ON THE KAMABOKO III ION SOURCE. 72

P3T-B-351 DESIGN AND TEST OF A HV DEVICE FOR PROTECTION AND POWER MODULATION OF 140 GHZ/1MW CW-GYROTRONS USED FOR ECRH ON W7-X 73

P3T-B-353 DEVELOPMENT OF CW AND SHORT-PULSE CALORIMETRIC LOADS FOR HIGH POWER MILLIMETER-WAVE BEAMS 74

P3T-B-356 ITER-LIKE PAM LAUNCHER FOR TORE SUPRA’S LHCD SYSTEM 75

P3T-B-359 LARGE CRYOSORPTION PUMP OF THE TEST STAND FOR THE KSTAR NBI SYSTEM 76

P3T-B-362 DEVELOPMENT OF A RF SOURCE FOR ITER NBI: FIRST RESULTS WITH D- OPERATION 77

P3T-B-364 PERFORMANCE TEST OF THE LH ANTENNA WITH CARBON GRILL IN JT-60U 78

P3T-B-371 FIRST RESULTS OF THE TORE SUPRA ITER LIKE ICRF ANTENNA PROTOTYPE 79

P3T-B-372 TORE SUPRA ITER-LIKE ANTENNA CHARACTERIZATION BY FEM ANALYSIS 80

P3T-B-382 MAINTENANCE SCHEMES FOR THE ITER NEUTRAL BEAM INJECTOR TEST FACILITY 81

P3T-B-385 NEUTRAL BEAM INJECTION OPTIMIZATION AT TJ-II 82

P3T-B-387 STATUS OF THE 140 GHZ / 10 MW CW TRANSMISSION SYSTEM FOR ECRH ON THE STELLARATOR W7-X 83

P3T-B-392 THE TEST OF A PAM LAUNCHER ON FTU: THE FIRST STEP TOWARD THE LHCD LAUNCHER FOR ITER 84

P3T-B-410 MATERIAL PROCESSING AND PROTOTYPE FABRICATION OF HEAT TRANSFER ELEMENTS FOR SST-1 NBI SYSTEM. 85

P3T-B-412 AN ALTERNATIVE SCHEME FOR THE ITER NBI POWER SUPPLY SYSTEM 86

P3T-B-439 NEUTRONICS ANALYSIS OF THE ECW LAUNCHING SYSTEM IN THE ITER UPPER PORT 87

P3T-B-455 THE ITER-LIKE ICRF ANTENNA FOR JET 88

P3T-B-460 EFFECTS OF MUTUAL COUPLING ON ICRF LOAD-TOLERANT ANTENNAS 89

P3T-B-479 140-GHZ HIGH-POWER GYROTRON DEVELOPMENT FOR THE STELLARATOR W7-X 90

P3T-B-501 DEVELOPMENT OF THE 140 GHZ GYROTRON AND ITS SUBSYSTEMS FOR ECH AND ECCD IN TEXTOR 91

P3T-B-507 DESIGN OF CRYOSORPTION PUMPS FOR TESTBEDS OF ITER RELEVANT NEUTRAL BEAM INJECTORS 92

P3T-B-512 STATUS OF THE NEW ECRH SYSTEM FOR ASDEX UPGRADE 93

P3T-B-514 IMPROVED 118 GHZ GYROTRON FOR ECRH EXPERIMENTS ON TORE SUPRA 94

P3T-B-542 MATCHING TO ELMY PLASMAS IN THE ICRF DOMAIN 95

- C - Plasma Engineering and Control. 96

P3C-C-62 OPTIMISED MODELLING OF THE TORE SUPRA TOKAMAK FOR PLASMA EQUILIBRIUM CALCULATIONS WITH THE PROTEUS CODE 96

P3C-C-77 HIGH PERFORMANCE INTEGRATED PLASMA CONTROL IN DIII–D* 97

P3C-C-79 PROGRESS TOWARDS ACHIEVING PROFILE CONTROL IN THE RECENTLY UPGRADED DIII-D PLASMA CONTROL SYSTEM* 98

P3C-C-103 REAL TIME CONTROL OF FULLY NON-INDUCTIVE OPERATION IN TORE SUPRA LEADING TO 1GJ PLASMA DISCHARGES 99

P3C-C-114 FEEDBACK CONTROL FOR PLASMA POSITION IN HL-2A TOKAMAK 100

P3C-C-149 DIII-D INTEGRATED PLASMA CONTROL TOOLS APPLIED TO NEXT GENERATION TOKAMAKS* 101

P3C-C-155 CONFIGURATION AND PERTURBATION DEPENDENCE OF THE NEUTRAL POINT IN JET 102

P3C-C-157 DEVELOPMENT OF THE DINA-CH FULL DISCHARGE TOKAMAK SIMULATOR 103

P3C-C-161 DESIGN, IMPLEMENTATION AND TEST OF THE EXTREME SHAPE CONTROLLER (XSC) IN JET 104

P3C-C-190 CORRECTION POSSIBILITIES OF MAGNETIC FIELD ERRORS IN WENDELSTEIN 7-X 105

P3C-C-207 REAL TIME CONTROL ENVIRONMENT IN THE RFX EXPERIMENT 106

P3C-C-233 A FAST AND VERSATILE INTERLOCK SYSTEM 107

P3C-C-268 NEW VISUALIZATION SYSTEM FOR CONTROLLING AND MONITORING PURPOSES IN THE TJ-II STELLARATOR 108

P3C-C-277 WEB-BASED GROUND LOOP SUPERVISION SYSTEM FOR THE TJ-II STELLARATOR 109

P3C-C-291 A NEW CONTROLLER FOR THE JET ERROR FIELD CORRECTION COILS 110

P3C-C-298 USING REAL TIME WORKSHOP FOR RAPID AND RELIABLE CONTROL IMPLEMENTATION IN THE FRASCATI TOKAMAK UPGRADE FEEDBACK CONTROL SYSTEM RUNNING UNDER RTAI-LINUX 111

P3C-C-301 THE SYSTEM ARCHITECTURE OF THE NEW JET SHAPE CONTROLLER 112

P3C-C-350 AN INTEGRAL APPROACH TO PLASMA SHAPE CONTROL 113

P3C-C-352 LINEARIZED MODELS OF THE PLASMA RESPONSE IN THE NEW RFX LOAD ASSEMBLY 114

P3C-C-354 DESIGN OF THE NEW RFX EQUILIBRIUM ACTIVE CONTROL SYSTEM 115

P3C-C-363 REAL-TIME MEASUREMENT AND CONTROL AT JET- EXPERIMENT CONTROL 116

P3C-C-375 OPEN LOOP CHARACTERIZATION OF AN ACTIVE CONTROL SYSTEM OF MHD MODES 117

P3C-C-377 COMPARISON OF STRATEGIES AND REGULATOR DESIGN FOR ACTIVE CONTROL OF MHD MODES 118

P3C-C-383 ADOPTING MODERN NONLINEAR CONTROL TECHNIQUES FOR THE PLASMA STABILIZATION ON THE NOVEL LINUX-BASED FEEDBACK CONTROLLER OF FTU 119

P3C-C-403 VERTICAL STABILITY OF ITER PLASMAS WITH 3D PASSIVE STRUCTURES AND A DOUBLE LOOP CONTROL SYSTEM 120

P3C-C-408 THE BASIC METHODS FOR UNDERSTANDING OF PLASMA EQUILIBRIUM TOWARD ADVANCED CONTROL 121

P3C-C-457 XSC PLASMA CONTROL: TOOL DEVELOPMENT FOR THE SESSION LEADER 122

P3C-C-463 A FLEXIBLE AND REUSABLE SOFTWARE FOR REAL-TIME CONTROL APPLICATIONS AT JET 123

P3C-C-508 COMMISSIONING TESTS FOR CONTROL PROCESSES IN ASDEX UPGRADE´S NEW CONTROL AND DATA ACQUISITION SYSTEM 124

P3C-C-510 OPTIMIZATION OF THE IGNITOR OPERATING SCENARIO AT 11 MA 125

P3C-C-517 PLASMA FEEDBACK CONTROLLER REORGANISATION FOR ASDEX UPGRADE'S NEW DISCHARGE CONTROL AND DATA ACQUISITION SYSTEM 126

- D - Diagnostics, Data Acquisition and Remote Participation. 127

P2C-D-22 NEUTRON ANALYSIS OF H-ALPHA AND CXRS DIAGNOSTICS OF ITER 127

P2C-D-42 NEW CONSTRAINTS FOR PLASMA DIAGNOSTICS DEVELOPMENT DUE TO THE HARSH ENVIRONMENT OF MJ CLASS LASERS 128

P2C-D-52 THE INTEGRATED VISUALISATION SOFTWARE FOR THE ITER IN VESSEL VIEWING SYSTEM (IVVS) 129

P2C-D-60 USING REMOTE PARTICIPATION TOOLS TO IMPROVE COLLABORATIONS 130

P2C-D-74 CALORIMETRY MEASUREMENTS DURING HIGH ENERGY DISCHARGES AT TORE SUPRA 131

P2C-D-80 REAL-TIME MULTIPLE NETWORKED VIEWER CAPABILITY OF THE DIII-D EC DATA ACQUISITION SYSTEM* 132

P2C-D-91 EXPERIMENTAL STUDY OF RADIATION-INDUCED CURRENTS IN COPPER AND STAINLESS STEEL CORE MINERAL-INSULATED CABLES IN THE BR2 RESEARCH REACTOR 133

P2C-D-98 TORE-SUPRA INFRARED THERMOGRAPHY SYSTEM, A REAL STEADY STATE DIAGNOSTIC. 134

P2C-D-104 NEW INSTRUMENTS FOR ADVANCED NEUTRON EMISSION SPECTROMETRY DIAGNOSIS OF D AND DT PLASMAS AT JET 135

P2C-D-119 SURFACE DIAGNOSTICS WITH APPLICATION OF VIDEOSCOPE ON THE BASIS OF CU-LASER 136

P2C-D-120 NEW MANAGING SYSTEM OF A LARGE AMOUNT OF IMAGES ON TORE SUPRA 137

P2C-D-132 RADIATION RESISTANT BOLOMETERS USING PLATINUM ON AL2O3 AND ALN 138

P2C-D-142 TEMPERATURE DEPENDENCE OF THE TRANSMISSION LOSS IN KU-1 AND KS-4V QUARTZ GLASSES FOR THE ITER DIAGNOSTIC WINDOW 139

P2C-D-159 THERMAL AND NEUTRON TESTS OF MULTILAYERED DIELECTRIC MIRRORS 140

P2C-D-166 LASER DAMAGE INVESTIGATIONS OF CU MIRRORS 141

P2C-D-194 DESIGN OF LOST ALPHA PARTICLE DIAGNOSTICS FOR JET* 142

P2C-D-202 DEVELOPMENT OF THE PHASE COUNTER WITH THE REAL-TIME FRINGE JUMP CORRECTOR FOR INTERFEROMETER ON LHD 143

P2C-D-208 DATA ACQUISITION UPGRADE IN THE RFX EXPERIMENT 144

P2C-D-220 THERMAL DETECTOR FOR THE LOST ALPHA PARTICLE MEASUREMENTS 145

P2C-D-226 ITER RELEVANT DEVELOPMENTS OF NEUTRON DIAGNOSTICS DURING JET TRACE TRITIUM CAMPAIGN 146

P2C-D-230 OPTICAL AND ELECTRICAL DEGRADATION OF HYDROGEN IMPLANTED KS-4V QUARTZ GLASS 147

P2C-D-234 RECONSTRUCTION CAPABILITY OF JET MAGNETIC SENSORS 148

P2C-D-235 QUENCH DETECTION & DATA ACQUISITION SYSTEM FOR SST-1 SUPERCONDUCTING MAGNETS 149

P2C-D-248 LASER DAMAGE OF KU-1 SILICA GLASS COVERED WITH HYDROCARBON FILM 150

P2C-D-250 WIDE AREA DATA REPLICATION IN AN ITER-RELEVANT DATA ENVIRONMENT 151

P2C-D-251 ADVANCES IN REMOTE PARTICIPATION FOR FUSION EXPERIMENTS* 152

P2C-D-258 SOFT COMPUTING AND CHAOS TEORY FOR ANTICIPATION OF DISRUPTION IN TOKAMAK REACTORS 153

P2C-D-263 ADSORPTION IN INSULATOR MATERIALS ENHANCED BY D IMPLANTATION 154

P2C-D-270 RADIATION ENHANCED DEGRADATION OF SIO OVERCOATED ALUMINIUM MIRRORS 155

P2C-D-272 THE NEW MEASUREMENT MONITORING SYSTEM ON FTU 156

P2C-D-281 FIRST RESULTS OF MINIMUM FISHER REGULARISATION AS UNFOLDING METHOD FOR JET NE213 LIQUID SCINTILLATOR NEUTRON SPECTROMETRY 157

P2C-D-282 PRESENT STATUS OF THE TJ-II REMOTE PARTICIPATION SYSTEM 158

P2C-D-347 APD DETECTOR ELECTRONICS AND PXI BASED DATA ACQUISITION SYSTEM FOR SST-1 THOMSON SCATTERING DIAGNOSTICS 159

P2C-D-366 REAL TIME MEASUREMENT AND CONTROL AT JET - DIAGNOSTIC SYSTEMS 160

P2C-D-370 OPTICAL FIBERS FOR PLASMA DIAGNOSTICS UNDER GAMMA-RAY AND UV IRRADIATION 161

P2C-D-384 THE HALO CURRENT SENSOR SYSTEM FOR JET-EP 162

P2C-D-396 DEVELOPMENT OF ACTIVELY COOLED PERISCOPES FOR DIVERTOR OBSERVATION 163

P2C-D-398 DIAGNOSTICS FOR STUDYING DEPOSITION AND EROSION PROCESSES IN JET 164

P2C-D-428 VULNERABILITY OF OPTICAL FIBERS FOR PLASMA DIAGNOSTICS OF LASER MEGAJOULE 165

P2C-D-433 APPLICATION OF ORTHOGONALLY POLARIZED TWO-FREQUENCY LASER TO POLARIMETER FOR MAGNETIC FIELD MEASUREMENTS OF LONG-PULSED FUSION DEVICES 166

P2C-D-437 THE TEXTOR DIAGNOSTIC DATA MANAGEMENT CHAIN 167

P2C-D-446 NEW LOW LOSS TRIAXIAL AND MAGNETICS DIAGNOSTICS FEEDTHROUGH AT JET 168

P2C-D-451 THERMO-STRESS ANALYSIS OF OPTICAL MATERIALS FOR HIGH HEAT FLUX APPLICATIONS 169

P2C-D-458 DESIGN AND MANUFACTURE OF THE UPPER COILS AND OUTER POLOIDAL COILS SUBSYSTEMS FOR THE JET-EP MAGNETIC DIAGNOSTIC 170

P2C-D-459 DESIGN OF EX-VESSEL MAGNETIC PROBES FOR JET-EP 171

P2C-D-466 TRANSDUCERS AND SIGNAL CONDITIONERS OF THE RFX NEW MAGNETIC MEASUREMENT SYSTEM 172

P2C-D-476 WIDE-ANGLE INFRARED THERMOGRAPHY FOR JET-EP 173

P2C-D-477 LITHIUM BEAM DEVELOPMENTS FOR HIGH-ENERGY PLASMA DIAGNOSTICS 174

P2C-D-502 THE NEW TAE - ALFVÉN WAVE ACTIVE EXCITATION SYSTEM AT JET 175

P2C-D-503 NEW MILLIMETER-WAVE ACCESS FOR JET REFLECTOMETRY AND ECE 176

P2C-D-504 CONTROL PROCESS STRUCTURE OF ASDEX UPGRADE´S NEW CONTROL AND DATA ACQUISITION SYSTEM 177

P2C-D-506 MULTI-SUPPORT VECTOR MACHINES FOR DISRUPTION CLASSIFICATION IN TOKAMAK REACTORS 178

P2C-D-513 OPTICAL DESIGN OF THE OBLIQUE ECE ANTENNA SYSTEM FOR JET 179

P2C-D-515 ITER DIAGNOSTICS: MAINTENANCE AND COMMISSIONING IN THE HOT CELL TEST BED 180

P2C-D-519 NEW BOLOMETRY CAMERAS FOR THE JET ENHANCED PERFORMANCE PHASE 181

- E - Magnets and Power Supplies. 182

P2T-E-9 THE BATCH PRODUCTION FOR SUPERCONDUCTING MAGNET COILS OF EAST (HT-7U) 182

P2T-E-20 STUDY ON HIGH-POWER HIGH-FREQUENCY INVERTER FOR FAST PLASMA POSITION CONTROL IN EAST SUPER-CONDUCTING TOKAMAK 183

P2T-E-23 A LOW COST JOINT FOR THE ITER PF COILS, DESIGN AND TEST RESULTS. 184

P2T-E-30 130KV 130A HIGH VOLTAGE SWITCHING MODE POWER SUPPLY FOR NEUTRAL INJECTORS - CONTROL ISSUES AND ALGORITHMS 185

P2T-E-31 FIBERGLASS UNIDIRECTIONAL COMPOSITE TO BE USED FOR ITER PRE-COMPRESSION RINGS 186

P2T-E-34 MEASUREMENT OF CONTACT RESISTANCE DISTRIBUTION IN TYPICAL ITER SIZE CONDUCTOR TERMINATION 187

P2T-E-35 UPDATING THE DESIGN OF THE FEEDER COMPONENTS FOR THE ITER MAGNET SYSTEM 188

P2T-E-36 MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: COMPUTATION OF THE BACKGROUND FIELD AND CONSEQUENCES ON THE DESIGN OF THE ELECTRICAL DISTRIBUTION BOARDS AND CONTROL BOARDS FOR THE ITER TOKAMAK BUILDING 189

P2T-E-37 COMMISSIONING OF THE 10 POWER SUPPLIES OF THE CONTROL COILS OF WENDELSTEIN 7-X EXPERIMENT 190

P2T-E-40 DESIGN AND COMMISSIONING OF THE NEW TOROIDAL FIELD COIL FOR THE NATIONAL SPHERICAL TORUS EXPERIMENT (NSTX) 191

P2T-E-48 ANALYSES AND IMPLICATIONSOF V-I CHARACTERISTIC 192

P2T-E-50 PIONEERING SUPERCONDUCTING MAGNETS IN LARGE TOKAMAKS: EVALUATION AFTER 17 YEARS OF OPERATING EXPERIENCE 193

P2T-E-55 STABILITY, THERMAL EQUILIBRIUM AND DESIGN CRITERIA FOR CABLE-IN-CONDUIT-CONDUCTORS WITH A BROAD TRANSITION TO NORMAL STATE 194

P2T-E-68 DESIGN OPTIMISATION OF THE ITER TF COIL CASE AND STRUCTURES 195

P2T-E-73 FABRICATION OF THE PLANAR COILS FOR WENDELSTEIN 7-X 196

P2T-E-81 OVERVIEW OF THE DIII–D INTERNAL RESISTIVE WALL MODE STABILIZATION POWER SUPPLY SYSTEM* 197

P2T-E-95 THE EUROPEAN DEVELOPMENT OF A FULL SCALE SWITCHING UNIT FOR THE ITER SWITCHING AND DISCHARGING NETWORKS 198

P2T-E-97 MECHANICAL PERFORMANCE OF MAGNET INSULATION MATERIALS FABRICATED BY THE “INSULATE-WIND-AND-REACT “ TECHNIQUE* 199

P2T-E-99 INFLUENCE OF PARAMETER VARIATIONS ON THE FATIGUE BEHAVIOR OF MAGNET INSULATION SYSTEMS 200

P2T-E-101 MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: TESTS ON PROGRAMMABLE LOGICAL CONTROLLERS AND OTHER ELECTRONIC DEVICES 201

P2T-E-105 DESIGN, FABRICATION AND INSTALLATION OF CRYOGENIC TARGET SYSTEM FOR 14 MEV NEUTRON IRRADIATION 202

P2T-E-106 THE EUROPEAN NB3SN ADVANCED STRAND DEVELOPMENT PROGRAMME 203

P2T-E-111 DESIGN AND DEVELOPMENT OF THE POWER SUPPLY SYSTEM FOR HL-2A TOKAMAK 204

P2T-E-112 THE ITER THERMAL SHIELDS FOR THE MAGNET SYSTEM: DESIGN EVOLUTION AND ANALYSIS 205

P2T-E-121 QUALITY ASSURANCE PROCEDURES IN THE EAST MAGNETS MANUFACTURING PROCESS 206

P2T-E-126 THYRISTOR CROWBAR SYSTEM FOR THE HIGH CURRENT POWER SUPPLIES OF ASDEX UPGRADE 207

P2T-E-186 OPTIMIZATION OF THE POWER SUPPLY FOR A HELIAS REACTOR SUPERCONDUCTING COIL SYSTEM 208

P2T-E-198 QUENCH CURRENT MEASUREMENT AND PERFORMANCE EVALUATION OF THE EAST TOROIDAL FIELD COILS 209

P2T-E-209 THERMAL AND STRUCTURAL ANALYSIS OF THE W7-X MAGNET HEAT RADIATION SHIELD 210

P2T-E-212 FILAMENT POWER SUPPLY (AC TO AC CONVERTER) FOR LONG PULSE NEUTRAL BEAM INJECTOR OF SST-1 211

P2T-E-219 TRANSIENT ELECTRICAL BEHAVIOUR OF THE ITER TF COILS DURING FAST DISCHARGE AND TWO FAULT CASES 212

P2T-E-223 STUDIES ON THE BEHAVIOR OF MULTISECONDARY TRANSFORMERS USED FOR REGULATED HV POWER SUPPLIES 213

P2T-E-236 HIGH POWER IGBT BRIGE APPLICATION FOR THE HARMONIC SUPPRESSION IN THE POWER SUPPLY SYSTEM OF THE SPANISH STELLARATOR TJ-II. 214

P2T-E-259 MANUFACTURE AND TEST OF THE NON-PLANAR COILS FOR WENDELSTEIN 7-X 215

P2T-E-266 V-I CHARACTERISTICS WITH BUMPS IN THE MEDIUM SIZE NBTI CICC CABLES. 216

P2T-E-284 HIGH TEMPERATURE SUPERCONDUCTORS FOR THE ITER MAGNET SYSTEM AND BEYOND 217

P2T-E-299 ANALYSIS OF THE RESISTIVE TRANSITION IN NB-TI CABLE-IN-CONDUIT CONDUCTORS VIA AN EXTENDED 1-D MODEL 218

P2T-E-304 EFFECTIVE BENDING STRAIN ESTIMATED FROM IC TEST RESULT OF D SHAPED NB3AL CICC COIL FABRICATED WITH A REACT AND WIND PROCESS FOR THE NATIONAL CENTRALIZED TOKAMAK 219

P2T-E-306 ELIMINATION OF VARIABLE HARMONICS ON MOTOR GENERATOR CIRCUIT FOR EXPERIMENTAL FUSION FACILITY 220

P2T-E-311 FATIGUE ASSESSMENT OF THE ITER TF COIL CASE BASED ON JJ1 FATIGUE TESTS 221

P2T-E-326 EFFECT OF ELECTRICAL CHARACTERISTICS OF SIC POWER DEVICE ON OPERATIONAL EFFICIENCY OF AC/DC CONVERTER 222

P2T-E-338 DESIGN REQUIREMENT, QUALIFICATION TESTS AND INTEGRATION OF A THIN SOLID LUBRICANT FILM OF MOS2 FOR COLD MASS SUPPORT STRUCTURE OF THE STEADY STATE SUPERCONDUCTING TOKAMAK SST-1. 223

P2T-E-344 HOW SHOULD WE TEST THE ITER TF COILS ? 224

P2T-E-379 CYCLIC TESTING OF SHEAR KEYS FOR THE ITER MAGNET SYSTEM 225

P2T-E-390 MODULAR COIL DESIGN DEVELOPMENTS FOR THE NATIONAL COMPACT STELLARATOR EXPERIMENT (NCSX) 226

P2T-E-394 CONCEPTUAL DESIGN OF SPHERICAL TORUS WITH TF-CS HYBRID COILS BASED ON VIRIAL THEOREM 227

P2T-E-405 EMI ON DIAGNOSTICS AND CONTROL CIRCUITS DUE TO SWITCHING POWER SUPPLIES 228

P2T-E-406 THE CONTROL SYSTEM OF THE TOROIDAL POWER SUPPLY OF RFX 229

P2T-E-417 COMPONENTS AND SYSTEM TESTS ON THE RFX TOROIDAL POWER SUPPLY 230

P2T-E-420 COMMISSIONING AND OPERATION OF 130KV/130A SWITCHED-MODE HV POWER SUPPLIES WITH THE UPGRADED JET NEUTRAL BEAM INJECTORS 231

P2T-E-427 MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: TESTS ON LOW VOLTAGE CIRCUIT BREAKERS AND CONTACTORS 232

P2T-E-442 FIRST INTEGRATED TEST OF THE SUPERCONDUCTING MAGNET SYSTEMS FOR THE LEVITATED DIPOLE EXPERIMENT (LDX) 233

P2T-E-462 MODELING AC LOSSES IN THE ITER NBTI FULL SIZE JOINT SAMPLES USING THE THELMA CODE 234

P2T-E-471 POWER DISSIPATION AND ENERGY TRANSFER DURING TESTING OF THE ITER TOROIDAL FIELD MODEL COIL 235

P2T-E-490 DC AND TRANSIENT CURRENT DISTRIBUTION ANALYSIS FROM SELF-FIELD MEASUREMENTS ON ITER PFIS CONDUCTOR 236

P2T-E-491 THE MAGNET SYSTEM OF THE KTM TOKAMAK 237

P2T-E-511 OPTIMISATION OF THE CURRENT DISTRIBUTION IN THE IGNITOR POLOIDAL FIELD COILS AND EVALUATION OF THE COILS TEMPERATURES AND RESISTANCE DURING THE REFERENCE OPERATING SCENARIO 238

P2T-E-528 SAFETY ASSESSMENT OF THE ITER COILS SYSTEM 239

P2T-E-534 WINDING MACHINES FOR THE MANUFACTURING OF SUPERCONDUCTIVE COILS OF THE MAIN EUROPEAN FUSION RESEARCH MACHINES 240

P2T-E-539 A SUCCESS STORY: LHC CABLE PRODUCTION AT ALSTOM MSA 241

- F - Plasma Facing Components. 242

P4C-F-8 TILES CHAMFERING AND POWER HANDLING OF THE MK II HD DIVERTOR 242

P4C-F-12 THERMAL AND MECHANICAL ANALYSIS OF THE EAST PLASMA FACING COMPONENTS 243

P4C-F-13 THE DYNAMIC ERGODIC DIVERTOR IN TEXTOR – A NOVEL TOOL FOR STUDYING MAGNETIC PERTURBATION FIELD EFFECTS 244

P4C-F-24 EAST(HT-7U) IN-VESSEL COMPONENTS DESIGN 245

P4C-F-27 HOT RADIAL PRESSING: AN ALTERNATIVE TECHNIQUE FOR THE MANUFACTURING OF PLASMA-FACING COMPONENTS 246

P4C-F-28 HETS PERFORMANCES IN HE COOLED POWER PLANT DIVERTOR 247

P4C-F-32 THE INFLUENCE OF IRRADIATION REGIMES ON RETENTION HYDROGEN ISOTOPES IN STRUCTURAL MATERIALS 248

P4C-F-33 MANUFACTURING TECHNOLOGY DEVELOPMENT FOR THE VACUUM VESSEL AND PLASMAFACING COMPONENTS 249

P4C-F-41 ENGINEERING AND THERMAL-HYDRAULIC DESIGN OF PFC COOLING FOR SST-1 TOKAMAK 250

P4C-F-49 THE USE OF COPPER ALLOY CUCRZR AS A STRUCTURAL MATERIAL FOR ACTIVELY COOLED PLASMA FACING AND IN VESSEL COMPONENTS 251

P4C-F-54 MANUFACTURING OF THE W7-X DIVERTOR AND WALL PROTECTION 252

P4C-F-59 STUDIES ON GRAPHITE SURFACES DETRITIATION BY PULSED REPETITION RATE NANOSECOND LASERS 253

P4C-F-66 STEADY STATE AND TRANSIENT THERMAL-HYDRAULIC ANALYSES ON ITER DIVERTOR MODULE 254

P4C-F-69 APPLIED TECHNOLOGIES AND INSPECTIONS FOR THE W7-X PRE-SERIES TARGET ELEMENTS 255

P4C-F-76 OVERVIEW OF THE ENGINEERING DESIGN OF THE ITER DIVERTOR 256

P4C-F-92 TOWARDS THE DEVELOPMENT OF WORKABLE ACCEPTANCE CRITERIA FOR THE DIVERTOR CFC MONOBLOCK ARMOUR. 257

P4C-F-100 RESULTS AND ANALYSIS OF HIGH HEAT FLUX TESTS ON A FULL SCALE VERTICAL TARGET PROTOTYPE OF ITER DIVERTOR 258

P4C-F-115 STRUCTURAL AND FRACTURE MECHANICS ANALYSIS OF ITER TOROIDAL FIELD COIL 259

P4C-F-116 CRACK PROPAGATION BEHAVIOR AROUND DSCU/SS316 HIP BONDED INTERFACE BY THERMAL FATIGUE 260

P4C-F-133 SIMULATION OF MANY-ATOMIC INTERACTIONS IN W-O-H SYSTEM WITH THE MD CODE CADAC 261

P4C-F-135 DESIGN OF A LIMITER FOR THE JET EP ICRH ANTENNA 262

P4C-F-145 EROSION OF TUNGSTEN MACROBRUSH ARMOR AFTER MULTIPLE INTENSE TRANSIENT EVENTS IN ITER 263

P4C-F-146 DEVELOPMENT OF AN ORIGINAL ACTIVE THERMOGRAPHY METHOD ADAPTED TO ITER PLASMA FACING COMPONENTS CONTROL 264

P4C-F-162 PLASMA SPRAYED TUNGSTEN-BASED COATINGS AND THEIR PERFORMANCE UNDER FUSION RELEVANT CONDITIONS 265

P4C-F-167 HIGH TEMPERATURE STRESSES IN ITER RELEVANT BRAZED GLIDCOP/W MODEL STRUCTURES 266

P4C-F-176 A MATURE INDUSTRIAL SOLUTION FOR ITER DIVERTOR PLASMA FACING COMPONENTS: HYPERVAPOTRON COOLING CONCEPT ADAPTED TO TORE SUPRA FLAT TILE TECHNOLOGY 267

P4C-F-192 CONCEPTUAL DESIGN OD A HIGH-TEMPERATURE WATER-COOLED DIVERTOR FOR A FUSION POWER REACTOR 268

P4C-F-195 DEVELOPMENT OF A COPPER ALLOY TO BERYLLIUM HIP BONDING TECHNOLOGY FOR THE ITER FIRST WALL 269

P4C-F-228 AN ADVANCED HE-COOLED DIVERTOR CONCEPT: DESIGN, COOLING TECHNOLOGY, AND THERMOHYDRAULIC ANALYSES WITH CFD 270

P4C-F-239 THE NEW ELECTRON BEAM TEST FACILITY JUDITH II FOR HIGH HEAT FLUX EXPERIMENTS ON PLASMA FACING COMPONENTS. 271

P4C-F-253 FORMATION OF CRYSTALLINE NANOSTRUCTURES DURING DEUTERIUM PLASMA INTERACTION WITH TUNGSTEN-BASED MATERIALS IN SIMULATED GAS DIVERTOR CONDITIONS. 272

P4C-F-265 ACTIVITY OF THE EUROPEAN HIGH HEAT FLUX TEST FACILITY: FE200 273

P4C-F-274 PROPOSAL OF LITIZATION OF FTU VACUUM VESSEL BY USING A LITHIUM LIMITER 274

P4C-F-278 DESIGN, PERFORMANCE AND CONSTRUCTION OF A 2 MW ION BEAM TEST FACILITY FOR PLASMA FACING COMPONENTS 275

P4C-F-279 SPECTROSCOPIC STUDIES OF HOMOGENEOUS CARBON FLAKES WITH A HIGH DEUTERIUM CONTENT FORMED IN TOKAMAK T-10 276

P4C-F-280 VACUUM PLASMA-SPRAYED TUNGSTEN ON EUROFER AND 316L - RESULTS OF CHARACTERISATION AND THERMAL LOADING TESTS - 277

P4C-F-285 CAN TOKAMAK DEVICES SURVIVE ELMS DURING NORMAL OPERATION? A SIMULATION STUDY 278

P4C-F-294 EU R&D ON DIVERTOR COMPONENTS 279

P4C-F-305 THERMAL MODELING OF W ROD ARMOR SUBJECTED TO ELMS 280

P4C-F-315 CRITICAL HEAT FLUX TESTING ON SCREW COOLING TUBE MADE OF RAFM-STEEL F82H FOR DIVERTOR APPLICATION 281

P4C-F-328 STATUS OF HE-COOLED DIVERTOR DEVELOPMENT FOR DEMO 282

P4C-F-333 NUMERICAL AND EXPERIMENTAL STUDY OF DEMO HE-COOLED DIVERTOR TARGET MOCK-UPS 283

P4C-F-343 MEASUREMENTS OF H/D DIFFUSIVITY IN AND SOLUBILITY THROUGH TUNGSTEN IN THE TEMPERATURE RANGE OF 600 C TO 800 C 284

P4C-F-367 TESTING OF ACTIVELY COOLED MOCK-UPS IN SEVERAL HIGH HEAT FLUX FACILITIES – AN INTERNATIONAL ROUND ROBIN TEST 285

P4C-F-376 STUDY OF TECHNOLOGICAL AND MATERIAL ASPECTS OF HE-COOLED DIVERTOR FOR DEMO REACTOR 286

P4C-F-386 DESIGN AND THERMAL PERFORMANCE OF SURFACE-BOLTLESS MECHANICALLY ATTACHED MODULE FOR DIVERTOR PLATE OF LHD 287

P4C-F-413 MANUFACTURE OF BLANKET SHIELD MODULES FOR ITER 288

P4C-F-426 THE MAST IMPROVED DIVERTOR 289

P4C-F-429 OXYGEN IMPURITY EFFECTS ON HYDROGEN ISOTOPE RELEASE FROM PLASMA CHEMICAL VAPOR DEPOSITION BORON COATING 290

P4C-F-431 IMPLANTATION TEMPERATURE DEPENDENCE ON DEUTERIUM BEHAVIOR IN HIGHLY ORIENTED PYROLITIC GRAPHITE 291

P4C-F-445 MANUFACTURING AND TESTING IN REACTOR RELEVANT CONDITIONS OF BRAZED PLASMA FACING COMPONENTS OF THE ITER DIVERTOR 292

P4C-F-447 DEVELOPMENT OF THE PLASMA FACING COMPONENTS FOR THE DOME-LINER COMPONENT OF THE ITER DIVERTOR 293

P4C-F-474 THERMAL PROPERTY CHANGES OF ERODED AND REPETITIVELY LOADED CFC 294

P4C-F-475 STUDIES OF HEAT CONDUCTION IN LIQUID LITHIUM CAPILLARY POROUS SYSTEM 295

- G - Vessel-in vessel Engineering and Remote Handling. 296

P4T-G-4 DESIGN AND DEVELOPMENT TOWARDS A PARALLEL WATER HYDRAULIC WELD/CUT ROBOT FOR MACHINING PROCESSES IN ITER VACUUM VESSEL 296

P4T-G-14 BAKING SYSTEM FOR EAST VACUUM VESSEL 297

P4T-G-18 LASER MEGAJOULES CRYOGENIC TARGET DEVICES 298

P4T-G-29 IRRADIATION TESTS ON WATER HYDRAULIC COMPONENTS 299

P4T-G-53 EXPERIMENTAL RESULT OF THE LASER IN VESSEL VIEWING AND RANGING SYSTEM (IVVS) FOR ITER 300

P4T-G-56 ITER VACUUM VESSEL SECTOR MANUFACTURING DEVELOPMENT IN EUROPE 301

P4T-G-71 STRUCTURAL UPGRADE OF IN-VESSEL CONTROL COIL ON DIII D* 302

P4T-G-89 MANUFACTURING OF CRYOSTAT FOR EAST SUPERCONDUCTING TOKAMAK 303

P4T-G-117 SPECIAL BLANKET DESIGN IN THE NB REGION OF ITER 304

P4T-G-137 USE OF ELECTRONIC AND OPTOELECTRONIC INDUSTRIAL SYSTEMS FOR MAINTENANCE TOOLS OF ITER FUSION EXPERIMENTAL REACTOR 305

P4T-G-199 NON-DESTRUCTIVE TESTING OF BONDED STRUCTURES FOR PLASMA FACING COMPONENTS 306

P4T-G-240 VERTICAL DIPLACEMENT EVENTS SIMULATIONS FOR TOKAMAK PLASMAS 307

P4T-G-249 DESIGN OF THE ITER HOT CELL BUILDING 308

P4T-G-264 RECENT DEVELOPMENTS TOWARDS ITER 2001 DIVERTOR MAINTENANCE 309

P4T-G-293 MANAGEMENT OF A WATER LEAK ON ACTIVELY COOLED FUSION DEVICES 310

P4T-G-296 DESIGN PROGRESS OF THE ITER VACUUM VESSEL AND PORTS 311

P4T-G-348 1200 MM BORE VOLTAGE BREAK OF THE NB DUCT FOR KSTAR 312

P4T-G-355 MANUFACTURE OF THE PLASMA VESSEL AND THE PORTS FOR WENDELSTEIN 7-X 313

P4T-G-360 DYNAMIC IDENTIFICATION OF THE HYDRAULIC ITER MAESTRO MANIPULATOR - RELEVANCE FOR MONITORING 314

P4T-G-361 GENERIC CONTROL SYSTEM DESIGN FOR THE CASSETTE MULTIFUNCTION MOVER AND OTHER ITER REMOTE HANDLING EQUIPMENT 315

P4T-G-374 ANALYSES OF THE ITER VACUUM VESSEL WITH THE USE OF A NEW MODELLING TECHNIQUE 316

P4T-G-389 ITER ARTICULATED INSPECTION ARM (AIA): GEOMETRIC CALIBRATION ISSUES OF A LONG-REACH FLEXIBLE ROBOT. 317

P4T-G-393 ITER ARTICULATED INSPECTION ARM (AIA) : R&D PROGRESS ON VACUUM AND TEMPERATURE TECHNOLOGY FOR REMOTE HANDLING. 318

P4T-G-404 ASSESSMENT OF A COOPERATIVE MAINTENANCE SCHEME FOR ITER DIVERTOR COOLING PIPE 319

P4T-G-422 RF TESTS OF THE ELECTRICAL INSULATIONS FOR THE TOROIDAL STRUCTURES OF RFX 320

P4T-G-435 OPERATIONAL EXPERIENCE FEEDBACK IN JET REMOTE HANDLING 321

P4T-G-509 IGNITOR PLASMA CHAMBER STRUCTURAL DESIGN WITH DYNAMIC LOADS DUE TO PLASMA DISRUPTION EVENT 322

- H - Fuel Cycle. 323

P1C-H-17 ADVANCED PROCEDURES FOR TWO-STAGE REPETITIVE PELLET INJECTOR. 323

P1C-H-38 STUDIES OF PELLET DELIVERY AND SURVIVABILITY THROUGH CURVED GUIDE TUBES FOR FUSION FUELING AND IMPLICATIONS FOR ITER 324

P1C-H-93 PELLET INJECTORS FOR STEADY STATE FUELLING 325

P1C-H-107 JET CONTRIBUTIONS TO THE ITER FUEL CYCLE ISSUES. 326

P1C-H-153 COMPARISON OF MODELLING OF TRITIUM RELEASE FROM CERAMIC BREEDER MATERIALS 327

P1C-H-217 ASSESSMENT OF THE ITER DWELL EVACUATION 328

P1C-H-247 STRATEGY FOR DETERMINATION OF ITER IN-VESSEL TRITIUM INVENTORY 329

P1C-H-275 REQUIREMENTS AND SELECTION CRITERIA FOR THE MECHANICAL PUMPS FOR THE ITER TRITIUM PLANT 330

P1C-H-303 HIGH-POWER PULSED FLASHLAMP CLEANING OF CO-DEPOSITED HYDROCARBON FILMS FROM PLASMA FACING COMPONENTS 331

P1C-H-358 GAS PUFFING BY MOLECULAR BEAM INJECTION IN ADITYA TOKAMAK 332

P1C-H-438 INFLUENCE OF DEUTERIUM ON THE DESIGN OF THE JET WATER DETRITIATION SYSTEM 333

P1C-H-441 EXPERIMENTAL VALIDATION OF A METHOD FOR PERFORMANCE MONITORING OF THE FRONT-END PERMEATORS IN THE TEP SYSTEM OF ITER 334

P1C-H-461 PROTECTION OF THE PRIMARY CIRCUITS AND EFFECT ON THE DESIGN OF THE INNER DEUTERIUM / TRITIUM FUEL CYCLE OF ITER 335

P1C-H-467 EVALUATION OF SUPER CRITICAL HELIUM AS A COOLANT FOR DIII-D TYPE CRYOCONDENSATION 336

- I - Materials Technology and Breeding Blankets. 337

P1C-I-1 USE OF THE SPIRAL 2 FACILITY FOR MATERIAL IRRADIATIONS WITH 14 MEV ENERGY NEUTRONS 337

P1C-I-10 SCIENTIFIC AND TECHNICAL FOUNDATIONS AND TECHNOLOGIES OF REDUCTION OF MHD-RESISTANCE OF DUCTS WITH HEAVY LIQUID METAL COOLANTS IN MAGNETIC FIELD OF BLANKET AND DIVERTER OF TOKAMAK 338

P1C-I-26 EFFECT OF UNDERSIZED SOLUTE ATOMS ON MICROSTRUCTURE CHANGE 339

P1C-I-39 RADIATION INDUCED CONDUCTIVITY AND SURFACE ELECTRICAL DEGRADATION OF PLASMA SPRAYED SPINEL FOR NBI SYSTEMS 340

P1C-I-43 BLANKET MANUFACTORING TECHNOLOGIES : THERMOMECHANICAL TESTS ON HCLL BLANKET MOCKS UP 341

P1C-I-58 HIGH ENERGY PROTON DEGRADATION IN KU1 QUARTZ GLASS 342

P1C-I-85 EXPERIMENTAL STUDY OF LITHIUM MHD FLOW IN SLOTTED CHANNEL FROM V-4TI-4CR ALLOY 343

P1C-I-88 A NEUTRONIC INVESTIGATION OF HE-COOLED LI-BREEDER BLANKETS FOR FUSION POWER REACTOR 344

P1C-I-96 MICROSTRUCTURAL CHARACTERISATION OF EUROFER-ODS RAFM STEEL IN THE NORMALIZED AND TEMPERED CONDITION AND AFTER THERMAL AGING IN SIMULATED FUSION CONDITIONS 345

P1C-I-102 NON-DESTRUCTIVE ANALYSIS OF MINIATURIZED FUSION MATERIALS SAMPLES AND IRRADIATION CAPSULES BY X RAY MICRO-TOMOGRAPHY 346

P1C-I-108 INNER STRUCTURES OF COMPRESSED PEBBLE BEDS DETERMINED BY X-RAY TOMOGRAPHY 347

P1C-I-109 THERMAL CREEP OF BERYLLIUM PEBBLE BEDS 348

P1C-I-110 THERMAL CREEP BEHAVIOR OF THE EUROFER97 RAFM STEEL AND TWO EUROPEAN ODS-EUROFER97 STEELS 349

P1C-I-122 SEGREGATED VOID SWELLING IN A HETEROGENEOUS MATERIAL: IMPLICATIONS FOR FUSION MATERIALS 350

P1C-I-128 THERMOCHEMISTRY OF LI-TITANATES CERAMICS IN REDUCING ENVIRONMENTS 351

P1C-I-129 MOLECULAR DYNAMICS SIMULATIONS OF DEFECT PRODUCTION DURING IRRADIATION IN SILICA GLASS 352

P1C-I-130 KINETICS OF LI DEPLETED LI2TIO3 REACTION WITH H3 ADDED TO AR PURGE GAS 353

P1C-I-134 VITAMIN-J/COVA/EFF-3 CROSS-SECTION COVARIANCE MATRIX LIBRARY AND ITS USE TO ANALYSE BENCHMARK EXPERIMENTS IN SINBAD DATABASE 354

P1C-I-140 IN-SITU FORMATION AND CHEMICAL STABILITY OF ER2O3 COATING ON V-4CR-4TI IN LIQUID LITHIUM 355

P1C-I-141 PHYSICO-CHEMICAL PROPERTIES OF AND HYDROGEN ISOTOPE BEHAVIORS IN LITHIUM-TIN ALLOY AS A LIQUID BREEDER FOR FUSION REACTOR 356

P1C-I-143 INTEGRAL EXPERIMENT ON BERYLLIUM WITH D-T NEUTRONS FOR VERIFICATION OF TRITIUM BREEDING 357

P1C-I-147 CREEP STRENGTH OF REDUCED ACTIVATION FERRITIC/MARTENSITIC STEEL EUROFER'97 358

P1C-I-150 REACTION OF TITANIUM BERYLLIDE 359

P1C-I-158 INTEGRAL BENCHMARK EXPERIMENTS ON VANADIUM SPHERES WITH A CENTRAL 14-MEV NEUTRON SOURCE AND INSIDE A SPHERICAL CRITICAL ASSEMBLY 360

P1C-I-163 PRESENT DEVELOPMENT STATUS OF EUROFER AND ODS FOR APPLICATION IN BLANKET CONCEPTS 361

P1C-I-164 MICROSTRUCTURAL INVESTIGATION, USING SMALL ANGLE NEUTRON SCATTERING, OF NEUTRON IRRADIATED EUROFER 97 STEEL 362

P1C-I-168 EFFECT ON IMPACT TOUGHNESS OF REDUCED OXYGEN CONTENT IN 316 STEEL POWDER JOINED TO 316 STEEL BY LOW TEMPERATURE HIP 363

P1C-I-178 ENVIRONMENTAL ASSISTED CRACKING OF EUROFER 97 IN WATER AND PB-LI 364

P1C-I-179 MEASUREMENT AND ANALYSIS OF RADIOACTIVITY INDUCED IN YTTRIUM AND LEAD IN FUSION PEAK NEUTRON FIELD 365

P1C-I-183 EVALUATION OF NUCLEAR HEATING, TRITIUM BREEDING AND SHIELDING EFFICIENCY OF THE DEMO HCLL BREEDER BLANKET 366

P1C-I-189 HYDROGEN EFFECTS ON THE TENSILE AND FATIGUE PROPERTIES OF EUROFER 97 367

P1C-I-191 THE HELIUM COOLED LITHIUM LEAD BLANKET TEST PROPOSAL IN ITER AND REQUIREMENTS ON TEST BLANKET MODULES INSTRUMENTATION 368

P1C-I-196 NUMERICAL AND EXPERIMENTAL STUDY ON TIME-DEPENDENT THERMOMCHANIC DEFORMATION OF CERAMIC BREEDER PEBBLE BEDS 369

P1C-I-197 HYDROGEN ISOTOPE DISTRIBUTIONS AND RETENTION IN THE INNER DIVERTOR TILE OF JT-60U 370

P1C-I-200 CRYSTAL STRUCTURE OF LI2TIO3 WITH SOME DIFFERENT OXIDE ADDITIVES 371

P1C-I-205 EVALUATION OF INSULATING PROPERTY OF CERAMIC MATERIALS FOR V/LI BLANKET SYSTEM UNDER FISSION REACTOR IRRADIATION 372

P1C-I-214 EVALUATION OF HYDROGEN ISOTOPE RETENTION IN BE12TI AS NEUTRON MULTIPLIER OF FUSION REACTOR 373

P1C-I-215 MECHANICAL PROPERTIES OF WELDAMENT USING IRRADIATED STAINLESS STEEL FOR BLANKET 374

P1C-I-224 AB-INITIO VALUES OF THE HE SIEVERT´S CONSTANT IN LIQUID LI 375

P1C-I-237 OUT-OF-PILE TRITIUM RELEASE PROPERTY CORRELATIONS FOR LI-DEPLETED LI2TIO3 AND LI4TI5O12 CERAMICS. EFFECTS OF REDUCTION-ANNEALING TREATMENTS 376

P1T-I-238 MAGNETOHYDRODYNAMIC PRESSURE-DRIVEN FLOWS IN THE HCLL BLANKET 377

P1T-I-242 LIQUID LITHIUM AS THE COOLANT OF THE IFMIF LOOP 378

P1T-I-252 HCLL TBM FOR ITER – DESIGN STUDIES 379

P1T-I-254 INTERNATIONAL COMPARISON OF MEASURING TECHNIQUES OF TRITIUM PRODUCTION FOR FUSION NEUTRONICS EXPERIMENTS 380

P1T-I-257 PEBBLE BED THERMAL-MECHANICAL THEORETICAL MODEL: APPLICATION AT THE GEOMETRY OF TEST BLANKET MODULE OF ITER-FEAT NUCLEAR FUSION REACTOR 381

P1T-I-269 BEHAVIOUR OF TRITIUM IN BREEDING BLANKET MATERIALS 382

P1T-I-276 MUTUAL CORROSION OF EUROFER97 AND THE BLANKET CERAMIC MATERIALS 383

P1T-I-287 AUTOMATIC GENERATION OF A JET 3D NEUTRONICS MODEL FROM CAD GEOMETRY DATA FOR MONTE CARLO CALCULATIONS 384

P1T-I-289 PERFORMANCE OF A HYDROGEN SENSOR IN PB-16LI 385

P1T-I-292 THE CHARACTERIZATION AND STRESS ANALYSIS ON VACUUM PLASMA SPRAYING TUNGSTEN COATINGS 386

P1T-I-295 SOME FEATURES OF BERYLLIUM CORROSION BEHAVIOUR IN BE-LIQUID LI-V4 TI 4 CR ALLOY SYSTEM 387

P1C-I-302 BERYLLIUM AS BLANKET MATERIAL: PECULIARITIES OF RADIATION DAMAGE UNDER HIGH DOSE NEUTRON IRRADIATION 388

P1T-I-307 TRITIUM BREEDING EXPERIMENTS WITH BLANKET MOCK-UPS CONTAINING 6LI-ENRICHED LITHIUM TITANATE AND BERYLLIUM IRRADIATED WITH DT NEUTRONS 389

P1T-I-308 EFFECTS OF GELATION AND SINTERING CONDITIONS ON GRANULATION OF LI2TIO3 PEBBLES FROM LI-TI COMPLEX SOLUTION 390

P1T-I-309 FUSION-DRIVEN HYBRID SYSTEM WITH ITER MODEL 391

P1T-I-313 SURFACE WAVE ON HIGH SPEED LIQUID LITHIUM FLOW FOR IFMIF 392

P1T-I-316 MEASUREMENT OF ENERGETIC CHARGED PARTICLES PRODUCED IN FUSION MATERIALS WITH 14 MEV NEUTRON IRRADIATION 393

P1T-I-318 STRUCTURAL ANALYSIS FOR THE GAS-COOLED HIGH FLUX TEST MODULE OF IFMIF 394

P1T-I-319 THERMAL AND THERMAL-STRESS ANALYSES OF IFMIF LIQUID LITHIUM TARGET ASSEMBLY 395

P1T-I-321 THERMAL HYDRAULIC ANALYSIS OF FDS-II LIPB BREEDER BLANKET 396

P1T-I-323 THERMAL DESORPTION BEHAVIOR OF HYDROGEN ISOTOPES INTERACTING WITH RADIATION DEFECTS IN LI2O 397

P1T-I-325 PRESENT STATUS OF BERYLLIDE STUDY FOR FUSION AND APPLICATION DEVELOPMENT IN JAPAN 398

P1T-I-327 EFFECTS OF IRRADIATION ON MECHANICAL PROPERTIES OF HIP-BONDED F82H STEEL 399

P1T-I-329 ACTIVATION OF EUROFER IN AN IFMIF-LIKE NEUTRON FIELD 400

P1T-I-331 JOINING OF CFC TO COPPER FOR ITER DIVERTOR 401

P1T-I-334 DEVELOPMENT OF RF-INPUT COUPLER WITH A MULTI-LOOP ANTENNA FOR RFQ LINAC IN IFMIF PROJECT 402

P1T-I-335 IN-SITU IN-REACTOR TESTING OF FUSION MATERIALS AND COMPONENTS 403

P1T-I-339 NEUTRONICS AND ACTIVATION CHARACTERISTICS OF THE INTERNATIONAL FUSION MATERIAL IRRADIATION FACILITY 404

P1T-I-340 DESIGN, MANUFACTURING AND TESTING OF THE IFMIF LITHIUM TARGET BAYONET CONCEPT BACKPLATE. 405

P1T-I-365 EFFECTIVE THERMAL CONDUCTIVITY OF A COMPRESSED LI2TIO3 PEBBLE BED 406

P1T-I-368 CONCEPTUAL DESIGN OF THE BLANKET MECHANICAL ATTACHMENT FOR THE HELIUM-COOLED LITHIUM-LEAD REACTOR 407

P1T-I-378 DEVELOPMENT OF EXPERIMANTAL DEVICES FOR IN-REACTOR MECHANICAL TESTS 408

P1T-I-388 NEW MODULAR CONCEPT FOR THE HELIUM COOLED PEBBLE BED TEST BLANKET MODULE FOR ITER 409

P1T-I-391 THE TEMPERATURE DEPENDENCE OF STRAIN-RATE EFFECT ON TENSILE STRENGTH OF MO-ALLOYS 410

P1T-I-397 MECHANICAL AND THERMAL PROPERTIES OF SIC/SIC COMPOSITES IRRADIATED WITH NEUTRONS AT HIGH TEMPERATURES 411

P1T-I-402 THERMAL-HYDRAULIC ANALYSIS AND OPTIMISATION OF THE BREEDER UNIT FOR THE EU HELIUM COOLED PEBBLE BED BLANKET 412

P1T-I-407 INFLUENCE OF NEUTRON IRRADIATION ON TOUGHNESS AND R-CURVE BEHAVIOUR OF SIC/SIC 413

P1T-I-409 IN-VESSEL INTEGRATION OF THE MODULAR EU HELIUM COOLED PEBBLE BED BLANKET IN A DEMO-RELEVANT TOKAMAK GEOMETRY 414

P1T-I-411 DEVELOPMENT AND FABRICATION ASPECTS REGARDING TUNGSTEN COMPONENTS FOR A HE-COOLED DIVERTOR 415

P1T-I-415 SLIP INFLILTRATION AND DENSIFICATION OF POROUS SICF/SIC PREFORMS USING SIC NANOPOWDERS 416

P1T-I-416 DESIGN OF FDS DEMO BLANKETS AND TEST BLANKET MODULE PROPOSED FOR ITER 417

P1T-I-418 STATUS OF THE HFR PETTEN HIGH DOSE IRRADIATION 418

P1T-I-419 HYDROGEN ISOTOPES BEHAVIOR ON LI2TIO3 UNDER VARIED SURFACE CONDITION 419

P1T-I-425 DEFORMATION BEHAVIOUR OF COPPER UNDER IN-REACTOR UNIAXIAL TENSILE TESTS 420

P1T-I-430 A HIGH FLUENCE IRRADIATION OF CERAMIC BREEDER MATERIALS IN HFR PETTEN, MATERIALS CHARACTERISATION AND TEST MATRIX. 421

P1T-I-432 CHARACTERIZATION AND STABILITY STUDIES OF TITANIUM BERYLLIDES 422

P1T-I-434 IN-SITU BONDING OF SIC/SIC BY CONTROLLED SHS COMBUSTION 423

P1T-I-440 NEUTRONIC DESIGN OPTIMISATION OF MODULAR HCPB BLANKETS FOR FUSION POWER REACTORS 424

P1C-I-444 THE EUROPEAN BREEDING BLANKETS DEVELOPMENT AND THE TEST STRATEGY IN ITER 425

P1T-I-450 FABRICATION OF YTTRIUM OXIDE AND ERBIUM OXIDE COATINGS BY PVD METHODS 426

P1T-I-454 ON THE HYPERPOROUS NON-LINEAR ELASTICITY MODEL FOR FUSION-RELEVANT PEBBLE BEDS 427

P1T-I-464 INFLUENCE OF HEATING TREATMENT AND MICROSTRUCTURE ON TRITIUM DESORPTION KINETIC 428

P1T-I-465 TRANSMUTATION AND ACTIVATION OF RUSSIAN STRUCTURAL MATERIALS FOR FUSION REACTORS IN NEUTRON SPECTRA OF FISSION AND FUSION REACTORS 429

P1T-I-470 ON THE NUCLEAR RESPONSE OF THE HELIUM-COOLED LITHIUM LEAD TEST BLANKET MODULE IN ITER 430

P1C-I-472 IN-PILE PERFORMANCE OF THE CERAMIC BREEDER PEBBLE-BED ASSEMBLIES FOR THE HCPB BLANKET CONCEPT 431

P1T-I-493 PRODUCTION OF LOW ACTIVATION V-(4-5)TI-(4-5)CR ALLOYS FOR FUSION REACTOR APPLICATIONS. 432

P1T-I-494 HEAT RESISTANT RAFMS RUSFER-EK-181 FOR FUSION AND FAST BREEDER REACTORS APPLICATIONS 433

P1T-I-496 GETTERING OF NITROGEN IN LIQUID LITHIUM 434

P1C-I-497 PROSPECTIVE TESTING PROGRAMME FOR IFMIF 435

P1T-I-498 NEUTRONIC OPTIMIZATION ANALYSIS OF FDS-‡U LIPB BREEDER BLANKET 436

P1T-I-516 PRODUCTION AND THERMAL STABILITY OF BERYLLIUM WITH FINE GRAIN STRUCTURE TO IMPROVE TRITIUM RELEASE DURING NEUTRON IRRADIATION 437

P1T-I-520 ITER MATERIALS PROPERTIES DATA 438

P1T-I-535 MIXED MHD CONVECTION AND TRITIUM TRANSPORT IN FUSION-RELEVANT CONFIGURATIONS 439

P1T-I-536 OPTIMIZATION OF REDUCED ACTIVATION MARTENSITIC STEEL F82H FOR DEMO BREEDING BLANKET 440

P1T-I-537 EFFECT OF TEMPERATURE CHANGE ON THE IRRADIATION HARDENING OF MARTENSITIC AND AUSTENITIC STEELS IRRADIATED TO 1.5 DPA IN JMTR 441

P1T-I-538 OXIDE DISPERSION STRENGTHENING STEELS R&D FOR WATER-COOLING FUSION BLANKET SYSTEM 442

- J - Power Plants, Safety and Environment, Socio-economics. 443

P4T-J-21 LOW LEVEL CLEANING OF A FUSION TARGET CHAMBER 443

P4T-J-44 THE EVITA PROGRAMME: EXPERIMENTAL AND NUMERICAL SIMULATION OF A FLUID INGRESS IN THE CRYOSTAT OF A WATER-COOLED FUSION REACTOR 444

P4T-J-45 CORROSION OF FUSION-SPECIFIC WASTE MATERIALS 445

P4T-J-63 MATERIALS ACTIVATION INDUCED BY HIGH ENERGY NEUTRONS: A COMPARISON OF ANITA-IEAF CALCULATION WITH MEASUREMENTS FROM THE KARLSRUHE ISOCHRONOUS CYCLOTRON 446

P4T-J-65 EXPERIMENTAL FUSION MATERIAL PHOTON AND ELECTRON DECAY HEAT MEASUREMENTS: ITS USE FOR ACTIVATION CODES VALIDATION 447

P4T-J-70 SAFETY ANALYSIS FOR ITER LICENSING 448

P4T-J-83 VALIDATION OF THE ECART CODE FOR THE SAFETY ANALYSIS OF FUSION REACTORS 449

P4T-J-86 RADIOACTIVE WASTE MANAGEMENT FOR THE IGNITOR FUSION EXPERIMENT 450

P4T-J-124 3D-ANALYSIS OF AN ITER ACCIDENT SCENARIO 451

P4T-J-127 CATEGORISATION OF ACTIVATED MATERIAL FROM FUSION POWER REACTORS AND ACCEPTABILITY FOR FINAL DISPOSAL 452

P4T-J-138 DUST IN ITER: R&D NEEDS FOR SAFETY COMPLIANCE 453

P4T-J-139 RADIOACTIVE WASTE FROM A D-HE3 REACTOR 454

P4T-J-148 FIBER OPTIC SENSORS NETWORKS FOR ENVIRONMENTAL AND SAFETY MONITORING OF FUSION REACTORS 455

P4T-J-151 THE ECONOMIC VIABILITY OF FUSION POWER 456

P4T-J-170 ITER DIVERTOR EX-VESSEL PIPE BREAK 457

P4T-J-177 ENVIRONMENTAL RELEASE TARGETS FOR FUSION POWER PLANTS 458

P4T-J-193 DYNAMIC ASSESSMENTS OF CHAMBER AND WALL RESPONSE TO TARGET IMPLOSION IN INERTIAL FUSION REACTORS 459

P4T-J-216 CONSEQUENCE CALCULATIONS FOR PPCS BOUNDING ACCIDENTS 460

P4T-J-222 COMPONENT FAILURE DATA COLLECTION AND ANALYSIS FROM JET AND TLK OPERATING EXPERIENCE 461

P4T-J-232 COLLECTION AND ANALYSIS OF OCCUPATIONAL RADIATION EXPOSURE DATA RELATED TO JET OPERATIONS 462

P4T-J-244 TFTR OCCUPATIONAL RADIATION EXPOSURE DATA COLLECTION AND ANALYSIS 463

P4T-J-260 FACTORS AFFECTING THE INHALATION DOSE FROM TRITIATED DUST AND FLAKES 464

P4T-J-271 THE EUROPEAN POWER PLANT CONCEPTUAL STUDY 465

P4T-J-273 INTRA ANALYSIS OF WET BYPASS TRANSIENTS INCLUDING TRITIUM 466

P4T-J-317 ACCESSIBILITY EVALUATION OF THE IFMIF LIQUID LITHIUM LOOP CONSIDERING ACTIVATED EROSION/CORROSION MATERIALS DEPOSITION 467

P4T-J-336 AVAILABILITY OF LITHIUM IN THE CONTEXT OF FUTURE D-T FUSION REACTORS 468

P4T-J-342 EFFECT OF ACTIVATION CROSS-SECTION UNCERTAINTIES IN SELECTING STEELS FOR THE HYLIFE-II CHAMBER TO SUCCESSFUL WASTE MANAGEMENT 469

P4T-J-357 EVALUATION OF FUSION STUDY FROM SOCIO-ECONOMIC ASPECTS 470

P4T-J-373 PROGRESS IN THE DEVELOPMENT OF A PIE-PIT FOR THE ITER TOKAMAK 471

P4T-J-380 GLOBAL ENERGY MODEL WITH FUSION 472

P4T-J-400 DUST EXPLOSION HAZARD IN ITER: EXPLOSION INDICES OF FINE GRAPHITE AND TUNGSTEN DUSTS AND THEIR MIXTURES 473

P4T-J-401 ECONOMIC ANALYSIS OF FDS FUSION POWER REACTORS 474

P4T-J-414 RELIABILITY ANALYSIS OF BLANKET MODULES OF FDS 475

P4T-J-421 FUSION SAFETY STUDIES IN RUSSIA IN 2003. 476

P4T-J-423 CORE CONCEPTUAL DESIGN OF FDS FUSION POWER REACTORS 477

P4T-J-452 NEUTRON ACTIVATION AND DOSE RATES MINIMIZATION ON LASER MÉGAJOULE (LMJ) FACILITY 478

P4T-J-524 DESIGN EARTHQUAKES FOR ITER AT CADARACHE 479

P4T-J-525 METHODOLOGY FOR REFERENCE ACCIDENTS DEFINITION FOR ITER 480

P4T-J-527 FIRE RISK ANALYSIS IN ITER TRITIUM BUILDING 481

P4T-J-529 CHEMICAL RISK STUDIES INCLUDING BERYLLIUM AND CHEMICAL ZONING 482

P4T-J-530 PROGRESS IN LICENSING ITER IN CADARACHE 483

P4T-J-532 ALARA APPLIED TO ITER DESIGN. RADIOPROTECTION AND ZONING APPROACH 484

- K - Transfer of Technology. 485

P1C-K-131 100 KV SOLID-STATE SWITCH FOR FUSION HEATING SYSTEMS 485

P1C-K-443 OVERVIEW OF CRYOGENIC REFRIGERATION SYSTEMS FOR THE THERMONUCLEAR FUSION 486

P1C-K-499 DEVELOPMENT & APPLICATION OF MCNP AUTO-MODELING TOOL : MCAM 3.0 487

- A - Current and Next Step Devices

P3C-A-11 SELECTION OF DESIGN SOLUTIONS AND FABRICATION METHODS AND SUPPORTING R&D FOR PROCUREMENT OF ITER VESSEL AND FW/BLANKET

Ioki, Kimihiro, the ITER International Team and Participant Teams

ITER Garching JWS, Boltzmannstrabe 2, 85748 Garching, Germany

The ITER project has started preparation of Procurement Specification Documents for key components. The design of the ITER vacuum vessel (VV) and first wall (FW)/blanket has progressed by selecting design solutions, and R&D results are providing the basis for selection of design solutions and fabrication methods. The VV design has progressed in many aspects, such as an independent cooling configuration in the VV field joint regions, 9 lower ports instead of 18, a single wall structure for the upper and equatorial ports except the NB ports, and the vacuum vessel gravity support located below the lower ports. Double curvature pressing is now selected instead of facet shape welding for inner and outer shells in the upper and lower inboard regions to improve the fabrication and NDT process. By this selection, very short distances between neighbouring welds can be avoided. A challenging UT R&D program is also going on to achieve acceptable S/N ratio for small-angle launching waves (20-30 deg.). Another approach is a combination of progressive PT and conventional UT. Selection of the NDT method in critical areas will be made based on R&D results. Regarding the FW/blanket system, the plasma facing surface of the FW has been defined to avoid protruding the leading edges, especially in the inboard area. Separate FW panels are supported with a central beam, and selection of a race-track shape cross-section for the central beam provides a more robust structure against halo current EM loads and also leads to a new cooling configuration in the shield block, where the pressure drop is significantly reduced to ~0.05 MPa. Detailed EM analysis has been performed by using a newly defined plasma current quench scenario (40 ms linear decay and 25 ms exponential decay), and EM loads due to eddy currents are reduced in the current design with deeper slits and lower steel/water ratio in the shield block. The welding/cutting method of the FW central beam in the hot cell will be selected between YAG laser and TIG welding/mechanical cutting, based on R&D results. For future higher performance operation, the possibility of long pulse operation (3000/1000 s burn time in non-inductive/hybrid operation) and high fusion power operation (700MW) have been assessed. Helium purge gas lines for the ITER breeding blanket have been designed and analysed as a parameter of the tritium partial pressure in the range 1-50 Pa, and further testing is proposed to select the parameter.


Corresponding Author:

Ioki, Kimihiro

Ioki, Kimihiro

- A - Current and Next Step Devices

P3C-A-16 THE PROTO-SPHERA LOAD ASSEMBLY

PAPASTERGIOU Stamos, ALLADIO Franco MICOZZI Paolo MANCUSO Alessandro

c/0 ENEA, VIA E FERMI 45, FRASCATI 00044,ROMA

THE PROTO-SPHERA LOAD ASSEMBLY S Papastergiou, F Alladio, A Mancuso, P Micozzi Absract PROTO-SPHERA is a proposed spherical torus where a hydrogen plasma arc, in a form of a screw pinch field fed by electrodes , replaces the central conductor. This simply connected magnetic configuration, if fusion relevant, might strongly simplify the design of a fusion reactor. The machine design philosophy, basic geometry and operating conditions together with the major components like the vacuum vessel, water cooled coils, electrodes, protection components, divertor etc will be analysed. The thermal and electromagnetic behavior, the duty cycle as well as the predicted and permitted key stresses will be discussed in order to prove that the design, construction and reliable operation of the machine are feasible as demonstrated in an international workshop at ENEA-Frascati in March 2002. Finally reference should be made to the proposed Multi-Pinch experiment, using the START vacuum vessel, to demonstrate the feasibility and stability of the Proto-sfera configuration.


Corresponding Author:

PAPASTERGIOU Stamos

c/o ENEA ,VIA E FERMI 45 ,FRASCATI 00044, ROMA

- A - Current and Next Step Devices

P3C-A-72 OVERVIEW OF THE DIII–D PROGRAM AND CONSTRUCTION PLANS*

Petersen, P.I., the DIII-D Team

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

Selected also for oral presentation O3B-A-72

The DIII-D tokamak is a mid size tokamak operating at reactor relevant parameters. Because of its size it is relatively easy to modify the machine as required to test new ideas or theories. During the last few years several new hardware items have been added to the DIII-D tokamak and improvements have been made to others. The main additions in the last two years were the installation of the I-coil system and upgrades to the electron cyclotron heating (ECH) system. In addition the fast wave system is being brought back into operation after having been idle for three years. The I-coil system, which consists of 12 coils installed inside the DIII-D vessel, is used to stabilize the resistive wall modes and to produce a stochastic edge, which has suppressed edge localized modes (ELMs). ELMs can be detrimental to ITER, since they can erode the plasma facing surfaces. The I-coils are powered by three switching power amplifying units, which together with a flexible patch panel allow the I-coils to be operated in many different configurations. The ECH system has been upgraded to six gyrotrons, which have been used to heat the plasma, modify the current profile and stabilize the neoclassical tearing 3/2 and 2/1 modes. Three ECH launchers built by Princeton Plasma Physics Laboratory are installed on the DIII-D tokamak and have the capability of changing the beam direction in both toroidal and poloidal directions. Three additional gyrotrons have been ordered for the DIII-D program. They are required for current profile control and stabilization of the NTMs. The gyrotrons are scheduled to be installed during a 10–12 month facility enhancement period, which spans 2005–2006. At the same time a modification is scheduled to be made to the lower divertor to make it pump double-null high triangularity plasmas, which are important for studying advanced tokamak plasmas. One of the four neutral beam lines will be rotated for counter injection, which will allow study of the quiescent double barrier mode with central co-rotation of the plasma and of the resistive wall mode with low rotation. *Work was supported by the U.S. Department of Energy under DE-FC02-04ER54698.


Corresponding Author:

Petersen, P.I.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- A - Current and Next Step Devices

P3C-A-90 COMMISSIONING AND PRELIMINARY OPERATION OF THE HL-2A TOKAMAK

LIU Dequan, LIU Yong , YAN Jianchen , CAO Zeng , YANG Qingwei , ZHOU Caipin , LI Xiaodong ,and the HL-2A Team

Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan,610041, P.R. China

HL-2A is a new operating tokamak in SWIP of China, it can be operated in double-null and single-null divertor with closed configurations. The effect of the divertor on impurity behaviors, MHD instabilities, transport, wall conditioning, divertor physics are key issues to study during the first step operation on HL-2A. The construction of the HL-2A project had been finished in the fall of 2002, the first plasma was obtained in the end of 2002. The improving of the vacuum system and other subsystems such as vessel inner pumping, control system and power supply system had been carried out in 2003,the feedback control of the plasma current and plasma position were used on both in limiter and divertor operations. Preliminary experiment with limiter and single null divertor configurations were achieved in 2003. Primarily results of 168KA plasma current , 920ms duration time and plasma linear average density of 1.7*10+13 cm-3 were obtained, impurity especial the low Z impurity was clearly decreased during the divertor operation. During the operation,the vacuum vessel was baked up to 115 C by hot water, glow discharge clearing was applied for approximate 120 hours with four electrodes , Ti metallic getters worked for about 10 hours in all.So far, the best limit vacuum obtained is 4.6*10-6Pa on HL-2A.Cryogetter pump is very useful tool to absorb H3O in fusion device, but the effect of a small capability pump used on HL-2A is not obvious,two great Cryogetter pumps will be used on HL-2A in 2004, better vacuum will be gotten. Higher plasma parameters will be expected with a enhaned power supply system in 2004.


Corresponding Author:

LIU Dequan

Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan,610041, P.R. China

- A - Current and Next Step Devices

P3C-A-144 PLASMA PHYSICS BASIS AND OPERATIONS OF FUSION-DRIVEN SUBCRITICAL SYSTEM

Bin Wu,

The Fusion Driven Sub-critical System (FDS) is a sub-critical nuclear energy system drive by fusion neutron source, which provides a feasible, safe, economic and highly efficient potential of disposing High Level Waste (HLW) and produce fission nuclear fuel as a early application of fusion technology. The system includes a tokamak as fusion neutron driver, a nuclear power system as blanket. Parameters of such kind reactor are following. major radius 4m, minor radius 1m, plasma current 5.7MA, toroidal field 5.2T, Bootstrap current fraction 0.90, Fusion power 143MW, Neutron wall loading 0.5MW/m2 . In this paper, an advanced plasma configuration for FDS system has been proposed, which aims at high bata, high bootstrap current and good confinement. The JSOLVER code has been used to getting equilibrium. Several different advance equilibrium configurations have been proposed. Among these modes, the reverse shear mode is most attractive. In order to determine the feasibility of tokamak operation, a transient simulation has been made which includes the equilibrium, transport and plasma position shape control in FDS. A 1.5D equilibrium evolution code has been used to make this simulation. The code is two-dimensional time dependent free boundary simulation code that advances the MHD equations describing the transport time-scale evolution of axisymmetric tokamak plasma. A detail plasma configuration evolution is obtained by this calculation. The simulation results confirm and constrain the system projections.


Corresponding Author:

Bin Wu

Institute of Plasma Physics, Chinese Academy of Sciences,P.O. Box 1126, Hefei, Anhui, 230031, China

- A - Current and Next Step Devices

P3C-A-184 THE WENDELSTEIN 7-X MECHANICAL STRUCTURE SUPPORT ELEMENTS: TESTS AND DESIGN

Gasparotto, Maurizio, Mr. Simon-Weidner and the W7-X team

Max-Planck-Institut für Plasmaphysik Wendelsteinstraße 1 17491 Greifswald Germany

The WENDELSTEIN 7-X Mechanical Structure Support Elements: Design and Tests M. Gasparotto, J. Simon-Weidner and the W 7-X team Max-Planck-Institut für Plasmaphysik, Euratom Association, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491Greifswald, Germany The stellarator WENDELSTEIN 7-X is in the construction phase at IPP Geifswald, Germany. The main parameters are: average major radius 5.5 m, average plasma radius 0.53 m, maximum magnetic field on the plasma axis 3.0 T, total weight 725t. The magnetic system of the machine consists of 50 superconducting Non-Planar Coils (NPC), 20 superconducting Planar Coils (PC), the Coil Support Structure and the Intercoil Support Structure (ISS). Each PC and NPC is supported by the Coil Support Structure through two Coil Connection Elements (CCE) that must transmit loads and moments up to 3.3 MN and 400 MNmm respectively. All components of the coil system are kept at 4K by liquid helium. The ISS consists of: (i) The Narrow Support Elements connecting adjacent NPC casings in the inner region by sliding joints; (ii) The Lateral Supports connecting adjacent NPC casings at the outer region by welded joints; (iii) The Planar Supports connecting the PC to the NPC by sliding joints. The CCE and the ISS are critical components that should satisfy the following requirements: operate in high vacuum and at cryogenic temperature, withstand high loads and moments and allow the assembly of the machine with high accuracy while minimising the distortion due to the welded connections. A large R&D programme is in progress to qualify the adopted solutions and to check the components which are most critical under operating conditions. The CCE based on inconel bolted connections, is tested in scale 1:1 at 77 K applying the maximum load and moment; the Narrow Support is tested at R.T. at the maximum load (150t) simulating tilting and sliding movements while samples of the material are tested under vacuum and at cryogenic temperature applying a sliding movement and a pressure of about 500 MPa. An R&D programme based on analytical simulation and experimental tests is in progress to optimize the weld sequence during the assembly of the magnetic system of W7-X. The paper will report the main design features of the CCE and ISS and results of tests carried out to qualify materials and critical components.


Corresponding Author:

Gasparotto, Maurizio

Wendelsteinstraße 1,D- 17491 Greifswald, Germany

- A - Current and Next Step Devices

P3C-A-203 LEVITATION EXPERIMENTS OF A HIGH TEMPERATURE SUPERCONDUCTOR COIL IN THE INTERNAL COIL DEVICE MINI-RT

Morikawa Junji, Ogawa Yuichi (1) Ohkuni Kotaro (1) Yamakoshi Shigeo (2) Goto Takuya (2) Mito Toshiyuki (3) Yanagi Nagato (3) Iwakuma Masataka (4) Uede Toshio (5)

(1)High Temperature Plasma Center, Univ. of Tokyo, (2)Graduate School of Frontier Science, the University of Tokyo, (3)National Institute for Fusion Science,(4)Research Institute of Superconductivity Kyushu University, (5)Fuji Electric Systems Co., Ltd.,

An Internal coil device would be expected for exploring high beta plasmas based on plasma relaxation process. Prof. A Hasegawa proposed an advanced fusion reactor with a dipole configuration, and Mahajan-Yoshida developed a new high beta state based on two-fluid relaxation theory. To study these high beta plasmas, we have constructed an internal coil device with a high temperature superconductor. The major radius of the internal coil is 15 cm, and the coil current is 50 kA. Three different types of Ag-sheathed Bi-2223 tapes are employed; i.e., a high critical current(Ic=108A at 77K, s.f., 1 micro-V/cm) tape with a low silver ratio for the main HTS coil, a 0.3wt%Mn-doped Bi-2223 tape for the persistent current switch and 3at%Au-doped Bi-2223 tape for the current lead. The coil is cooled with cold helium gas provided by a GM refrigerator and supplied to the coil through a check valve. The coil current is directly excited with the external power supply through removable electrode. It took about 11 hours to cool the coil down to 21K from the room temperature, and the nominal cable current of 118 A (overall coil current: 50kA) has been achieved. A decay time constant of the persistent current is a few tens of hours. Weight of the HTS internal coil is 16.8kg. Time constant of motion for the internal coil is about 70 ms in the center of vacuum vessel at a normal floating position. The position of the internal coil is monitored with 5 laser sensors which can be detected 5 freedoms (vertical, tilt-X-Y and sliding-X-Y) of the coil. The resolution of the laser sensor is 10 micrometers. A levitating coil is installed on top of a vacuum vessel that is made of copper coil. Rating of the levitation coil is 20kA. The vertical position of the internal coil is feedback-controlled with the regulation of the levitation coil current. The HTS internal coil is successfully levitated in the vacuum vessel during one hour or more. The accuracy of the internal coil position is 20 micrometers. Plasma experiments in a dipole configuration have been initiated. The plasma is produced with 2.45 GHz ECH system. At present, the plasma temperature and density are ~10 eV and 5x1016m-3, respectively.


Corresponding Author:

Morikawa Junji

Graduate School of Engineering, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku Tokyo 113-8656, Japan

- A - Current and Next Step Devices

P3C-A-245 EXPERIMENTAL STUDY OF WATER FLOW DISTRIBUTION INSIDE TWO-CHANNEL MODEL OF ITER VACUUM VESSEL COOLING SYSTEM

Tanchuk Victor, Babykin A., Balunov B., Chtcheglov A., Grigoriev S., Krylov V.

Joint-Stock Company I.I. Polzunov Scientific & Development Association of Research and Design Power Equipment, 3/6 ul. Atamanskay, St. Petersburg 191167, Russia

VV double wall and neutron shielding plates, poloidal and toroidal ribs with holes for water passing form a complex system of parallel-series channels for ITER VV cooling. The cooling channels formed in such manner are characterized by different cross sections (channel heights from 5mm to 90mm), heat loads and orientation in the gravitational field. At extremely low bulk flow velocities (10?200 mm/s) dimensions and position of water passage holes in the VV ribs, other VV design and loading conditions could significantly effect on the water flow distribution inside the parallel channels. This could impact on the VV temperature state. To investigate the flow distribution in the parallel channels and to prove interchannel flow stability a two-channel model of a VV test element has been developed. The VV test element is a rectangular box 0.2m wide, 3m long, with a 2.48 m heated length. The box is divided by an intermediate plate into two channels: upper 50mm in height and lower 12 mm in height. 3 heaters located at the upper, lower and intermediate walls produce heat loads separately for each channel. A total of 241 experiments were performed. The obtained results prove that: (1) design of the channel inlet unit is a decisive factor in water distribution between the parallel channels at extremely low flow rates; (2) hydraulic friction, inclination angle, water inlet temperature are not dominant in the mechanism of flow distribution at a significant influence of thermal gravitational forces; (3) heating of one of the channels has the principal effect on the flow splitting. This effect is especially drastic at low flow rates (Gtotal ? 0.3-0.4 kg/s), when practically the entire flow comes through a heated channel (vertical or inclined channels). No reverse circulation has been observed during the tests for all range of studied flow velocities (10?200 mm/s), so a stable flow distribution is expected for the VV cooling system.


Corresponding Author:

Tanchuk Victor

The D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA), 3 Doroga na Metallostroy, Metallostroy, St.Petersburg 196641, Russia

- A - Current and Next Step Devices.

P3C-A-262 MAGNUM-PSI, A PLASMA GENERATOR FOR PLASMA-SURFACE INTERACTION RESEARCH IN ITER-LIKE CONDITIONS

Groot de, Bart, G.J. van Rooij(1),V. Veremiyenko(1),M.G. von Hellermann(1),C.J. Barth(1),G.L. Kruijtzer(1),J.C. Wolff(1),H.J.N. van Eck(1),W.J. Goedheer(1),N.J. Lopes Cardozo(1),A.W. Kleyn(1),S. Brezinsek(2),A. Pospieszczyk(2),R.A.H. Engeln(3)

(1)FOM-Institute for Plasma Physics Rijnhuizen, Assoc. EURATOM-FOM,The Netherlands, www.rijnh.nl(†) (2)IPP, FZ Jülich GmbH, EURATOM Assoc.,Germany(†) (3)Eindhoven University of Technology,The Netherlands (†)Partners in the Trilateral Euregio Cluster

Introduction In collaboration with its TEC partners, the FOM-Institute for Plasma Physics is preparing the construction of Magnum-psi, a magnetized (3 T), steady-state, large area (100 cm^2) high-flux (up to 10^24 H+ ions m^-2s^-1) plasma generator. Magnum-psi is being developed to study plasma-surface interaction in conditions similar to those in the divertor of ITER and fusion reactors beyond ITER. Magnum-psi will be embedded in an integrated plasma-surface laboratory including in situ and ex situ, in vacuo surface analysis. The scientific program includes a strong modeling effort. A pilot experiment (Pilot-psi) has been constructed to explore the techniques to be applied in Magnum-psi. This contribution addresses the optimization of the cascaded arc plasma source and the effect of the magnetic field on the expanding plasma beam. Experimental results achieved on Pilot-psi will be presented to demonstrating that the required hydrogen plasma flux can be generated with a high-pressure plasma source (cascaded arc) and a longitudinal B-field of 1.6 T. Results of Pilot-psi: In order to obtain a detailed picture of the plasma fluxes for different cascaded arc plasma source geometries and magnetic field strengths, we employed electron density measurements by means of Thomson scattering and the analysis of Stark broadening in atomic emission spectroscopy. Thomson scattering data yielded radial profiles of the electron density and temperature with a spatial resolution of 1 mm and are in agreement with high-resolution spectroscopy results. Typical results in hydrogen are: Ne ranging from 10^20 to 1.5*10^21 m^-3 for B=0.4-1.6 T. Te=0.4 eV, only weakly varying with B. Using additional Ohmic heating of the expanding plasma, the temperature can be increased to a few eV. The flow velocity in the plasma jet was derived from time of flight analysis of plasma perturbations induced by modulation of the arc current and found to be subsonic (~250 m/s). Multiplication of the densities and propagation speeds yields hydrogen ion flux densities well above 10^23 H+ ions m^-2s^-1, proving that even in our pilot experiment ITER relevant flux densities of hydrogen plasma can be reached, albeit not in steady state (due to the pulsed magnetic field) and over a small cross section (1 cm^2).


Corresponding Author:

Groot de, Bart

FOM-Institute for Plasma Physics Rijnhuizen, P.O. Box 1207, 3430 BE Nieuwegein, The Netherlands

- A - Current and Next Step Devices

P3C-A-330 THE LASER MÉGAJOULE (LMJ) PROJECT DEDICATED TO INERTIAL CONFINEMENT FUSION : DEVELOPMENT AND CONSTRUCTION STATUS.

FLEUROT Noël, CAVAILLER Claude BOURGADE Jean-Luc

Commissariat à l'Energie Atomique - Centre DAM Ile de France - BP 12 - 91680 Bruyères le Châtel - France

Selected also for oral presentation O3B-A-330

The Laser Megajoule (LMJ) facility is a 240 beam facility dedicated to our Inertial Confinement Fusion program. Its construction was started in 2003 at the French Atomic Energy Commission CESTA center located near Bordeaux. LMJ is a frequency tripled Nd:glass laser able to focus up to 1.8 MJ – 600 TW of ultraviolet light (0.35 µm) on targets dedicated to laser matter interaction experiments and to achieve ignition and ultimately combustion of DT targets in the laboratory. Typical quadruplet focus spot size on target is in the 600 - 700 µm range in diameter and it can be adapted, by using optical phase plates, to obtain elliptical focal spots. LIL ("Ligne d'Intégration Laser" : the LMJ prototype) has been the first laser in the world to produce 9.5 kJ of UV light in less than 9 ns in 2003 with a single beam. The commissioning of the quadruplet (4 beams) at 0.35 µm is now achieved. We will also present the current LMJ design with its four laser bays (a total of 30 bundles x 8 beams) which produce the infrared light (typically 18 kJ at 1.05 µm per beam at the output of the amplifier section). Each bundle of 8 beams is then separated in 2 quadruplets in the target bay ; the 60 quadruplets of IR light are frequency tripled at 0.35 µm and focused by large optical gratings through 60 ports in the 10 m diameter target chamber onto the target. Plasma diagnostics (X-rays, neutrons …) will require resolutions in the 10 to 100 ps temporal range and 10 µm spatial range to diagnose laser fusion of DT cryogenic targets in the so called "indirect drive" configuration. LMJ will be able to achieve an energy output yield of up to 20 MJ. The first contracts concerning both the laser and target chamber area have already been procured to the French industry.


Corresponding Author:

FLEUROT Noël

Commissariat à l'Energie Atomique - Centre DAM Ile de France - BP 12 - 91680 Bruyères le Châtel - France

- A - Current and Next Step Devices

P3C-A-449 JET ENGINEERING: PROGRESS AND PLANS

Todd, Thomas, Kaye, Alan Pamela, Jerome Murari, Andrea Rolfe, Alan Riccardo, Valeria Brennan, Damian

EFDA-JET, Culham Science Centre, Abingdon, OX14 3DB, UK

Selected also for oral presentation O3B-A-449

The UKAEA-Euratom Association has now operated JET for EFDA for over four years, providing a sophisticated large tokamak facility for experiments run by the Contract of Association institutes. JET continues to offer a state-of-the-art capability strongly relevant to ITER physics and technology issues, including beryllium in the torus and a full tritium fuel cycle system. The original 20g inventory of tritium was re-injected five times for the first DT campaign and 5g of the now remaining 10g was injected and recovered in the recent Trace Tritium Experiment. Diagnostic and control system developments to track the tritium and minimise retention in the machine structure continue, with further new systems now being installed. Tritium operation mandates a remote handling system for work on major in-vessel components while the radiation field is high, eg 8mSv/hr falling to the “ALARP” target of 350microSv/hr for man entry. JET remote handling developments continue both in technical aspects such as load transfer control (260kgs at 10m for the poloidal limiter beams) and in training and rehearsals, ~80% in Virtual Reality since early 2003, and ~20% in the full-scale torus simulator. The VR environment is based on 3D design files, necessitating rigorous design configuration management for all machine modifications. Facility operation requires machine protection systems based on sophisticated stress analyses, to constrain operation within boundaries consistent with the desired plant life. This is especially demanding for Vertical Displacement Events in the high-delta divertor plasma shapes foreseen for ITER, which generate vertical forces around twice those of the pre-2000 plasmas in JET. Analyses show that life consumption of the key plant is ~10% at present, with operational limits of 4T, 5MA and 850t VDE force. Collaboration with the Associations has yielded many valuable improvements to the diagnostic and control systems, eg. real-time control and control of exotic plasma shapes to centimetric precision. The presently ongoing "Enhanced Performance" shutdown will add a range of capabilities to the machine including an ITER-like ICRH antenna and improved plasma diagnostic systems. This paper will detail the principle technical and analytical systems required to meet the challenge of providing an engineering environment for the JET-EP work programme.


Corresponding Author:

Todd, Thomas

Euratom-UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK

- A - Current and Next Step Devices.

P3C-A-468 THE JET-ENHANCED PERFORMANCE PROGRAM: MORE HEATING POWER AND DIAGNOSTIC CAPABILITIES IN PREPARATION FOR ITER

LIOURE Alain, Alan Kaye (2) Andrea Murari (3) Joaquin Sanchez (4) Tom Todd (2) Carlo Damiani (5) Jerome Pamela (1)

(1) EFDA JET, Culham, Abingdon, OX14 3EA UK (2) UKAEA JET, Culham, Abingdon OX14 3DB, UK (3) Consorzio RFX, Corso Stati Uniti, 4, I-35127 Padova Italy (4) CIEMAT, 28040 Madrid, Spain (5) ENEA Brasimone, 40032 Camugnano (BO), Italy

Since early 2000 the JET-EP program has been aiming at optimising JET for ITER-relevant plasma operations, from 2005 onwards. The overall heating capability of JET will be increased to 40 MW. The neutral beam system was up-graded but the major technical challenge is to build an ITER-like ICRH antenna, with particularly stringent specifications (8 MW/m2 for 10s, compatible with Type-I ELMy H-modes, coupling at 12cm distance to the plasma). The new divertor configuration will be able to absorb more than 300 MJ per shot. The physics of the power handling will be monitored by sophisticated new diagnostics, e.g. a high-signal to noise bolometric measurement system and an ambitious IR viewing system using a state of the art camera, looking at the antenna and the divertor. New halo sensors will be installed to better understand disruption phenomena. JET will operate in a wider range of plasma conditions. The divertor’s geometry allows high-triangularity ITER-like scenarios (deltaU~0.44, deltaL~0.56) with a greater flexibility with respect to different plasma configurations. The control of extreme plasma shapes will be re-enforced and a new disruption mitigation system using a very fast gas valve will be provided. The diagnostic capability will be enhanced by several new systems designed to address a number of crucial physical phenomena for ITER. To study Tritium retention further, new technologically challenging erosion-redeposition diagnostics will be installed, particularly in the divertor region, both real time and integrating. New neutron detectors using the latest advances in scintillators and data recording techniques will produce much higher count rates and signal to noise. Detectors for fast á particles with high pitch angle and energy resolution will be installed the closest ever to the plasma in JET. High-resolution Thomson Scattering, with 20 Hz repetition rate, will provide temperature and density profiles with a spatial accuracy close to two centimeters. An improved microwave access will enable broad band reflectometry for density profile and oblique ECE measurement for the first time on JET. High bandwidth coils and high-n Alfven mode-dedicated diagnostics will allow more emphasis on MHD regimes. The new diagnostics will be integrated in the JET real-time system. This paper presents an overview of this program, emphasising the main objectives and pointing out the various technological challenges and innovations.


Corresponding Author:

LIOURE Alain

EFDA, Culham Science Centre, Abingdon, Oxfordshire OX14 3EA (UK

- A - Current and Next Step Devices

P3C-A-518 NUCLEAR ANALYSES OF SOME KEY ASPECTS OF THE ITER DESIGN WITH MONTE CARLO CODES

Iida Hiromasa, L.Petrizzi(2) V. Khripunov(3) G. Federici(1) E. Polunovskiy(1)

(1)ITER Garching Joint Work Site Boltzmannstr. 2 D-85748 Garching Germany (2)Nuclear Fusion Institute, Russian Research Center "Kurchatov Institute", Moscow, Russia (3)Via E. Fermi 45 00044 Frascati ITALY (Rome)

The design of the ITER machine was presented in 2001 . Radiation transport calculations have been very important in the assessment of the ITER design, particularly with regard to operational constraints, access for reactor maintenance and activated waste. A nuclear analysis has been performed on ITER by means of the most detailed models and the best assessed nuclear data and codes. Calculations have been carried out in a progression which began with 1D studies for scoping, taking into account the reactor operating conditions, followed by 2D and 3D calculations taking into account streaming through penetrations, as well as the complexity of the geometry and the different material thicknesses and compositions. As the construction phase of ITER is approaching, the design of the main components has been optimsed/finalised and several minor design changes/optimisations have been made, which required refined calculations to confirm that nuclear design requirements are met. These have included assessment of nuclear heating in various components during various phases of the reactor operation, surface heat load on the in-vessel components due to bremsstrahlung and line radiation from the plasma, nuclear heating and damage of electric insulators due to N-16 in the blanket and divertor cooling water, and decay gamma-ray dose rate distribution around the machine after shutdown. This paper reviews some of the most recent neutronic work with emphasis on (i) critical neutronics responses in the TF coil inboard legs related to design modifications made to the blanket modules and vacuum vessel; (ii) accurate dose rate calculations after reactor shutdown, to confirm that the shielding around the torus is sufficient to allow personnel access for machine maintenance. All these detailed Monte Carlo analyses inevitably require very precise geometry modeling, which demand significant amount of manpower. Some of the ITER participant teams (in particular, Europe and China) are developing specific tools to facilitate conversion of CAD drawing information into MCNP models. A brief mention of this activity will be made, together with anticipated further developments to meet challenges ahead.


Corresponding Author:

Iida Hiromasa

ITER Naka Joint Work Site,c/o JAERI,Naka-machi, Naka-gun,Ibaraki-ken,Japan

- A - Current and Next Step Devices

P3C-A-522 TRANSPORT, LOGISTICS AND PACKAGING OF ITER COMPONENTS

Guérin Olivier, B. Couturier (1) A. Maas (1) and EISS Team

(1) Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France)

The construction of ITER will be an important challenge over the coming years. Components for the machine will be manufactured by all ITER partners, in factories around the world. These components, some of them very large and heavy, will have to be transported to the ITER construction site. In the case of the European site for ITER, at Cadarache in the South-East of France, the transport will have to be ensured over an itinerary of around 100 km, from the nearest industrial harbour to the site. Extensive studies have been undertaken in various fields, including the choice of an itinerary and its optimisation, the use of barges, ships, trucks, trailers and handling tools, kinematics and logistics of transports, packaging of different ITER components. Detailed logistics studies have been performed with world-leading companies in this field. An important feedback from a similar technical challenge, the successful completion in time and budget of an itinerary between Bordeaux and Toulouse for the transport of the future Airbus A380 parts, has also been used. The feasibility of these transports has been demonstrated and the different aspects of the studies retained solutions will be described in the paper.


Corresponding Author:

Guérin Olivier

Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

- A - Current and Next Step Devices

P3C-A-523 STUDIES FOR SITE PREPARATION FOR ITER CONSTRUCTION

Fardeau Agnes, F. Blanc (1) J.-D. Cardettini (1) J.-R. Mandine (1) R. Guérin (1) L. Patisson (1) P. Bergégère (1) A. Santagiustina (2) P. Garin (2) and EISS Team*

(1) Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France (2) Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

The implantation of a nuclear facility as ITER (surface of 40 hectares) requires many preparatory studies and works, particularly with respect to: Underground characterisation (geological survey) Impact of seismic hazard on design Topography, layout Climate data (mainly for the design of buildings and systems) Deforestation, excavations Networks, fences and roads Definition of an area on or close to the site for the companies during the construction phase The aim of this paper is to present the main results of the studies and works carried out within the European ITER Site Studies framework. To perform these studies, all needs of ITER have been taken into account (ITER requirements and design assumptions), but the proximity of the CEA centre has also been valorised. To choose the site for ITER implantation, detailed geological and geophysical investigations have been carried out (60 drillings, 4 km of seismic refraction lines, several tests on samples). Then, taking into account the meteorological data available since 1960 (particularly the main wind direction) and the topography (based on an aerial photos and topographic surveys), buildings and roads have been implemented, on 4 platforms (in order to minimize excavation work). Similarly, detailed studies have been carried out to implement all and satisfy ITER needs in terms of: cooling water supply (6,700 m3/day), potable water supply (400 m3/day), sanitary sewage (200 m3/day), industrial sewage (200 m3/day), cooling sewage (blow down: 3000 m3/day) treatment and exhaust, rainfall network, electrical supply (120 MW of continuous electrical power). Concerning the above items, existing infrastructures of CEA centre could be used, leading to substantial savings. Finally, ITER buildings, as defined in the generic site, have been estimated insufficient, with regard to the character of the project and other buildings, offered by Europe to the ITER partners, have been identified for the construction phase, but also for the exploitation phase. Preliminary studies have been carried out to define: a Welcome Centre (for visitors or workers families), a restaurant, a medical building, and an access control building.


Corresponding Author:

Fardeau Agnes

Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

- A - Current and Next Step Devices

P3C-A-531 READINESS OF CADARACHE FOR STARTING ITER CONSTRUCTION

Lyraud Charles, J.-M. Bottereau (1) A. Fardeau (2) O. Guérin (1) A. Maas (1) S. Mattei (2) P. Garin (1) and EISS Team

(1) Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France (2) Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

Since the beginning of the European ITER Site Studies in 2001, specific attention has been paid to the readiness and preparedness of the European site. These aspects include technical preparation of the site and its surrounding, as well as the welcome of the first international team members and their families in Provence in the best possible conditions. The purpose of this paper is to present the work already carried out and to be performed to ensure the successful construction of ITER in Europe, within time and budget. This will cover technical and socioeconomic aspects, such as: the licensing process, the increase of the industrial environment awareness, the heavy load itinerary road modification programme, the site preparation, the annex buildings offered by the host, the large poloidal field coil manufacturing facilities, erected on site to minimise the manufacturing and handling risks, the international school development (Japanese, Chinese, Russian, Korean and European languages) the housing for several hundred foreign families, the set up of a communication and welcome organisation. Phase 1: Starting on the site decision date, an ITER temporary facility to welcome ITER team members will be set up on the Cadarache site, where all services are already available for 5,000 people. Until the creation of the ITER Legal Entity and the European Legal Entity, the temporary International Team will complete the ITER construction filesThe European team will deal with public enquiries, site works, deforestation and site levelling, annex buildings final studies and start of works, enterprise yard for construction on site, heavy load itinerary works, industrial environment awareness of the thousands of companies located around the site. The same activity will be performed at the European level and World level by the Partners. A close follow-up will be carried out to supervise the availability of the International School soon after ILE creation in close collaboration with educational authorities of the ITER Parties. The licensing process will lead to the authorisation of construction of ITER facilities (French governmental decree). Phase 2: Starting at ILE creation. The annex buildings (Welcome Centre, restaurant, first aid facilities, offices…) as soon as they become available will be offered to the ITER organisation, as they are not linked to the licensing process. ITER building construction programme can be launched.


Corresponding Author:

Lyraud Charles

Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

- B - Plasma Heating and Current Drive.

P3T-B-19 THE ALCATOR C-MOD LOWER HYBRID CURRENT DRIVE EXPERIMENT TRANSMITTER

Montgomery Grimes, David Terry Ron Parker Dexter Beals

Same as corresponding address

Alcator C-Mod, is a high-field, high-density, diverted, compact tokamak, which, in its present form uses inductive current drive and is heated with 5 MW of ICRF auxiliary power. C-Mod is in the process of being upgraded with a 4.6 GHz Lower Hybrid heating and current drive system. The purpose of the experiment is to develop and explore the potential of “Advanced Tokamak Regimes” under quasi-steady-state conditions. In this paper, an overview of the RF transmitter and the controls and protection systems for the Lower Hybrid Project is given. The transmitter will use twelve 250 kW klystrons operating simultaneously which will result in a total nominal power at the klystrons of nearly 3 MW for a planned pulse width of 5 seconds. Active control system vector modulators provide phase and amplitude drive for each klystron, and I-Q detectors are used to monitor phase and amplitude. These feedback signals are used in digital controllers for closed-loop control of klystron phase and amplitude to preset values. An expected upgrade of four additional klystrons will result in a total nominal power of 4 MW. The transmitters have been tested to full power, and installation of the Lower Hybrid Current Drive experiment on the C-Mod Tokamak is expected in 2004.


Corresponding Author:

Montgomery Grimes

MIT Plasma Science and Fusion Center, 190 Albany St., Cambridge, MA 02139 USA

- B - Plasma Heating and Current Drive.

P3T-B-25 DESIGN OF AN ULTRA-BROADBAND SINGLE-DISK OUTPUT WINDOW FOR A FREQUENCY STEP-TUNABLE 1 MW GYROTRON

Xiaokang, Yang, Guenter Dammertz (1a) Roland Heidinger (1b) Kai Koppenburg (1a) Fritz Leuterer (3) Bernhard Piosczyk(1a) Dietmar Wagner (3) Manfred Thumm (1a),(2)

(1) Forschungszentrum Karlsruhe, Association EURATOM-FZK, (a) IHM, (b) IMF-1 76021 Karlsruhe, Germany. (2) Universitaet Karlsruhe, IHE, 76128 Karlsruhe, Germany. (3) Max-Planck-Institut fuer Plasmaphysik, Association EURATOM-IPP,85748 Garching, Germany

For plasma stabilization in the ASDEX-Upgrade tokamak, there is interest in step-tunable gyrotrons operating at frequencies between 105 GHz and 140 GHz. For this purpose a multifrequency gyrotron is under construction at Forschungszentrum Karlsruhe (FZK) in a cooperative parallel development with the Institute of Applied Physics in Nizhny Novgorod, Russia. Output window design is one of the key issues to realize broadband output of a multi-frequency gyrotron. Corresponding to the development of such frequency step-tunable 1 MW gyrotrons at FZK, this paper summaries recent development of broadband single-disk output windows, in particular the Brewster window with a CVD-diamond disk. The thickness of the disk has to be optimized to get low power reflection over a broadband incident angle range around the Brewster angle. Detailed calculations of the transmission characteristics for the CVD-diamond disk Brewster window have been performed for the all considerd 9 modes from TE17,6 at 105 GHz up to TE23,8 at 143 GHz, and for thickness of the disk from 1.5 mm up to 2.0 mm. Calculations show that it is difficult to choose the disk thickness of a CVD-diamond Brewster window for this frequency step-tunable gyrotron, since the choice depends on both the most important frequencies and the availability of the disks. If one prefers to place the low reflection area in the middle of the discussed frequency range, such as 120-130 GHz, the thickness of 1.6 mm is near optimum and its -20 dB bandwidth angle is more than 30 degrees. For operation near 105 GHz and 140 GHz, a 1.9 mm disk is preferable. Its -20 dB bandwidth angle is around 30 degrees, but for other central frequencies, the situation is not so good. Further calculation results also show that the -20 dB bandwidth angle decreases with increasing disk thickness from 1.5 mm to 2.0 mm. However, thin CVD-diamond disks will add mechanical problems to the window construction. Another important factor to be considered is the analysis of the bow and maximum tensile stresses in brazed windows arising from differential pressure uniformly applied over the surface of the disks, when they are not operated in evacuated transmission systems.


Corresponding Author:

Xiaokang, Yang

Forschungszentrum Karlsruhe, Association EURATOM-FZK, IHM, D-76021 Karlsruhe, Germany

- B - Plasma Heating and Current Drive.

P3T-B-51 EXPERIMENTS ON A 170 GHZ COAXIAL CAVITY GYROTRON

Piosczyk, Bernhard, Andreas Arnold (2), Herbert Budig (1a), Guenter Dammertz (1a), Olgierd Dumbrajs (3), Roland Heidinger (1b), Stefan Illy (1a), Jiambo Jin(1a), Georg Michel (4), Tomasz Rzesnicki (1a), Manfred Thumm (1a,2), Xiaokang Yang (1a)

(1a,b)FZK Karlsruhe, (a) IHM, (b) (IMF I), D-76021 Karlsruhe, Germany (2)Universitaet Karlsruhe, IHE, D-76128 Karlsruhe, Germany (3) Helsinki University of Technology, FIN-02150 Espoo, Finland (4) MPI fuer Plasmaphysik, D-17491 Greifswald, Germany

Within a development program performed as an ITER task at the Forschungszentrum Karlsruhe (FZK) the feasibility of manufacturing a multi-megawatt coaxial gyrotron operated in continuous wave (CW) has been investigated and information necessary for a technical design and industrial manufacturing has been obtained. Based on these results the development of a coaxial cavity gyrotron with an RF output power of 2 MW, CW at 170 GHz as could be used for ITER is in progress in cooperation between EURATOM Associations (CRPP Lausanne, FZK Karlsruhe and HUT Helsinki) together with European tube industry (Thales Electron Devices, Velizy, France). In parallel to that work on a first industrial prototype tube, the previously used short pulse 165 GHz, TE31,17 coaxial cavity gyrotron at FZK has been modified for operation at 170 GHz in the TE34,19 cavity mode. The modified experimental gyrotron operates in the same mode as foreseen for the industrial prototype and uses a cavity with same dimensions. In addition, the gyrotron is equipped with an improved quasi-optical RF output system same as designed for the prototype. The experimental operation is planned to start within the next weeks. The investigations have two main goals: (1) to verify experimentally the design of the main components of the industrial prototype by studying both the efficiency of RF generation and mode competition and the properties of the quasi-optical RF output system, (2) to provide a high power, short pulse (~5-10 ms) test possibility for studying a prototype of the remotely steerable launcher of the upper ITER port plug for neoclassical tearing mode stabilization. Results concerning as well the gyrotron operation and the conditions for the launcher test are expected and will be reported.


Corresponding Author:

Piosczyk, Bernhard

Forschungszentrum Karlsruhe, Association EURATOM-FZK, D-76021 Karlsruhe, Germany

- B - Plasma Heating and Current Drive.

P3T-B-78 THE UPGRADE OF THE DIII-D EC SYSTEM USING 120 GHZ ITER GYROTRONS

Callis, R.W., J. Lohr (1), Y.A. Gorelov (1), D. Ponce (1), K. Kajiwara (2), and J.F. Tooker (1)

(1) General Atomics, P.O. Box 85608, San Diego, California, 92186-5608 (2) Oak Ridge Institute for Science Education, Oak Ridge, Tennessee

The planned growth in the EC system on DIII-D over the next few years requires the installation of two depressed collector gyrotrons, a high voltage power supply, two low loss transmission lines, and the required support equipment. Although the original system is based on a frequency of 110 GHz, there is a benefit to the US Gyrotron development program, and the US ITER EC hardware manufacturer, if the next generation of EC equipment for the DIII-D program adopts the 120 GHz ITER startup frequency. This new DIII-D EC equipment could then be considered as a prototype of the ITER EC Startup System. By building the DIII-D hardware to the ITER specifications it would allow the US ITER program to gain beneficial prototyping experience on a working tokamak, prior to committing to building the hardware for delivery to ITER. *Work was supported by the U.S. Department of Energy under DE-FC02-04ER54698 and DE-AC05-76OR00033.


Corresponding Author:

Callis, R.W.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- B - Plasma Heating and Current Drive.

P3T-B-82 THE LHCD LAUNCHER FOR ALCATOR C-MOD – DESIGN, CONSTRUCTION, CALIBRATION AND TESTING*

J. Hosea (1), W. Beck (2) S. Bernabei (1) R. Childs (2) R. Ellis (1) E. Fredd (1) N. Greenough (1) M. Grimes (2) D. Gwinn (2) J. Irby (2) P. Koert (2) C. C. Kung (1) G. D. Loesser (1) R. Parker (2) D. Terry (2) R. Vieira (2) J. R. Wilson (1) J. Zaks (2)

(1) Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ, USA (2) Plasma Science and Fusion Center, MIT, Cambridge, MA, USA

MIT and PPPL have joined together to fabricate a high power lower hybrid current drive (LHCD) system for the Alcator C-MOD device to help support quasi steady-state AT regimes. A 3 MW source and a single launcher system have been provided for initial experiments. The launcher consists of a 24-column by 4-row waveguide array and has independent phasing control for each of the columns to maximize spectral control [1]. It was designed and constructed to support the application of 1.5 MW for up to 5 sec to the plasma, based on previous experimental power limits, and possibly 2 MW with sufficient conditioning. Some of the launcher design was based on previous experience with other devices: e.g., brazing of alumina windows into titanium guides is used to provide isolation of the coupler arrays at the plasma from the power feed guide system -- thereby facilitating the spectral control for the power launched into the plasma. However, much of the design uses new concepts for maximizing the number of guides in the relatively narrow C-MOD port while also maximizing the total power handling capability. Stacked waveguides incorporating a two-hole sidewall splitter design are used to deliver the power to the couplers [2]. All gaskets (microwave seals) are located outside the vacuum, and the alumina windows are “tuned” to the system frequency of 4.6 GHz [3]. Construction, calibration and testing techniques and results used in the carrying out of the design will be discussed. In particular, the bolt/gasket design for attaching the coupler to the stacked waveguide, the brazing of the alumina windows into the titanium couplers, and the power splitter design required considerable analysis and prototyping to achieve the desired performance. In addition, the results of high power tests for each of the component sections of the launcher assembly will be presented. These tests have been successfully conducted to power levels (in the range of 100 kW) representative of the maximum voltage/current conditions that will be experienced on C-MOD. *Work supported by US DOE Contracts No. DE-AC02-76CH03073 and DE-FC02-99ER54512 1. S. Bernabei et al., Fusion Science and Tech., 43, 145 (2003) 2. C. Kung et al., Proceedings of the 20th IEEE/NPSS SOFE Conf., P3-21 (San Diego, 2003) 3. J.R. Wilson et al., 15th Top. Conf. On RF Power in Plasmas, AIP Proc. Vol. 694, 283 (2003)


Corresponding Author:

J. Hosea (1)

Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ, USA

- B - Plasma Heating and Current Drive.

P3T-B-94 DESIGN AND OPERATION OF THE WENDELSTEIN 7-X ECRH HIGH VOLTAGE POWER SUPPLIES

Jürgen Alex, Michael Bader (1) Harald Braune (2) Dr. Volker Erckmann (2) Rüdiger Krampitz (2) Georg Michel (2) Marc Müller (1) Frank Noke (2) Dr. Günter Pfeiffer (2) Frank Purps (2) Edgar Sachs (3) Mario Winkler (2)

(1) Thales Broadcast & Multimedia, Bahnhofstr. 34, 5300 Turgi, Switzerland (2) Max-Planck Institut für Plasmaphysik (IPP), Wendelsteinstr. 1, 17491 Greifswald, Germany (3) FEAG, A Siemens Company, Günther-Scharowsky-Str. 2, 91058 Erlangen, Germany

The high voltage power supplies for the heating systems of Wendelstein 7-X are universal systems to be used on either ECRH or NBI heating. All power supplies are connected to a switching system, allowing to supply any load from any power supply. The power supplies are of the pulse-step-modulator type and rated for up to 130 kV / 130 A. The complete system has been delivered by a consortium between Thales Broadcast & Multimedia and Siemens. The tests on the first system were finished in November 2003. Since then the first power supply has been in operation for the tests on the first gyrotron on site. The paper gives an overview on the results of the power supply testing and the operation on the gyrotron. It shows the performance under normal operation as well as the short-circuit switching-off behaviour.


Corresponding Author:

Jürgen Alex

Thales Broadcast & Multimedia, Bahnhofstr. 34, 5300 Turgi, Switzerland

- B - Plasma Heating and Current Drive.

P3T-B-113 THERMAL ANALYSIS AND OHMIC LOSS ESTIMATION OF POLARIZER FOR ITER ECCD SYSTEM

Saigusa Mikio, K. Takahashi(2), Y. Kashiwa(1), S. Oishi(1), Y. Hoshi(1), T. Nakano(1), A. Kasugai(2), K.Sakamoto(2), T. Imai(2)

(1)Ibaraki University, Nakanarusawa 4-12-1, Hitachi-shi, Ibaraki-ken, Japan, (2)Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken, Japan.

An electron cyclotron current driving (ECCD) method is useful for suppressing the neoclassical tearing modes which degrade the energy confinement of tokamak plasmas. ECCD system in International Thermonuclear Experimental Reactor (ITER) needs to optimize polarization for exciting pure ordinary wave at an oblique injection into the tokamak plasmas. The specification of ECCD system in ITER demand severe operational conditions for transmission lines and polarizers, that is 1MW per one wave guide. Therefore it is important to evaluate ohmic loss of the rectangular grooved mirror installed in a miter bend type polarizer. The several polarizers were made of chromium copper alloy, installed in miter bends and tested at 170 GHz, 441kW during 6 seconds. The increase in temperature on the back plate of the grooved mirror has been measured with thermo couplers. The predicted dependences of ohmic loss of grooved mirrors on mirror rotation angle and the rotation angle of the polarization plane of the incident waves agree with the experimental results, qualitatively. The thermal analysis of grooved mirror has been performed with the 3D FEM code: FEVA, so that the behavior of the grooved mirror temperature could be explained.


Corresponding Author:

Saigusa Mikio

Ibaraki University, Nakanarusawa 4-12-1, Hitachi-shi, Ibaraki-ken, Japan

- B - Plasma Heating and Current Drive.

P3T-B-152 TESTS OF LOAD-TOLERANT EXTERNAL CONJUGATE-T MATCHING SYSTEM FOR A2 ICRF ANTENNA AT JET

Igor Monakhov, A.Walden (1) T.Blackman (1) D.Child (1) M.Graham (1) W.Hardiman (1) P.U.Lamalle(2) M.Nightingale (1) A.Whitehurst (1) JET EFDA contributors (3)

(1) Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK (2) LPP-EPM/KMS, Association Euratom-Belgian State, Brussels, B-1000, Belgium (3) Appendix of J.Pamela, et al., Fusion Energy 2002, IAEA, Vienna (2002)

Antenna matching during strong and fast loading perturbations introduced by ELMs is one of the major challenges of high power ICRF operations in H-mode plasmas both on present-day tokamaks and on ITER. The principle of conjugate-T matching offers a promising general approach to the problem and the methods for its implementation on tokamaks have been the focus of attention lately. A new 'ITER-like' antenna based on in-vessel conjugate-T matching by vacuum capacitors is being developed on JET [1]. A complementary technique to improve the ELM-tolerance of the existing JET A2 antennas by using the external conjugate-T circuit tuned by coaxial phase-shifters ('trombones') was proposed recently [2]. The latter approach relies entirely on well-established coaxial line technology; it also opens an opportunity to conjugate the straps belonging to different antenna arrays and, thus, to ensure an arbitrary array phasing. An upgrade of RF plant incorporating the external conjugate-T matching of two A2 four-strap antenna arrays into the existing JET RF system is now under consideration. In order to make a 'proof-of-principle' assessment of the proposal in realistic conditions a prototype system was installed and tested at JET. The set-up involved one pair of adjacent straps of the same antenna array powered by a single RF amplifier. The experimental program consisted of network analyser circuit characterisation, high voltage tests in vacuum and plasma operations in a range of scenarios including Type I ELMy H-mode. The tests confirmed the feasibility of the proposed matching scheme both for vacuum and plasma loading. Clear indications of high load tolerance during sawteeth and ELMs were observed in agreement with circuit simulations. Reliable trip-free performance was demonstrated in the 32-51 MHz frequency band at <1MW power levels. The paper provides a summary of the recent external conjugate-T matching activities at JET with an emphasis on the prototype test results. The work was performed under the European Fusion Development Agreement, and jointly funded by the UK Engineering and Physical Sciences Council and by EURATOM. [1] F. Durodie, et al., Proc. 15th Top. Conf. RF Power in Plasmas, Moran, 2003, AIP 694, 98 [2] I. Monakhov, et al., Proc. 15th Top. Conf. RF Power in Plasmas, Moran, 2003, AIP 694, 150


Corresponding Author:

Igor Monakhov

Euratom/UKAEA Fusion Association, J20/1/3, Culham Science Centre, Abingdon, OX14 3DB, UK

- B - Plasma Heating and Current Drive.

P3T-B-156 NEUTRONIC ANALYSIS OF ITER NEUTRAL BEAM TEST BED

Michael Loughlin,

It is proposed that ITER have at least two (and possibly three) heating neutral beam injectors. These will inject 1MeV deuterons in to the plasma and are expected to operate for periods of up to one hour. This represents a major technological step forward. It is therefore necessary to operate a test during the construction phase of ITER so that a working reliable system is available for the early operational phase. During the testing of the injector, deuterons will be fired in a calorimeter. This will result in the build up of deuterium and then the production of 2.5MeV neutrons via the D(d,n)3He reaction, A second branch of the d-d reaction (D(d,p)T) produces tritium and then the reaction D(t,n)4He produces 14MeV neutrons. Calculations indicate that the 14MeV neutron production is less than 1% of the total neutron yield. The total neutron production of the test bed facility is estimated to be 1022 neutrons. Neutron transport calculations are therefore important for the determination of activation of machine structures, dose to workers during maintenance, and the design of shielding around the device. This paper describes the results of these calculations. The neutron transport was modelled using the Monte-Carlo particle transport code MCNP. This was used to determine the neutron fluxes and spectra throughout the injector and in ancillary apparatus around it. The activation of the components was calculated using the inventory code FISPACT. Dose estimates were then made using further gamma transport calculations, again using MCNP. It is found that substantial shielding will be needed around the device and no access will be possible within this area during operations. At the end of operations some components will be sufficiently activated that special work practices will be required to allow maintenance while keeping the dose to workers below regulatory limits. It is recommended that steps be taken to minimise the build up of deuterium within the calorimeter to reduce the neutron production and that the use of low activation steels be considered for some items close to the neutron source. The 14MeV neutrons, although produced at a lower level, are shown to produce significant additional activation via threshold reactions. This work was funded jointly by the United Kingdom Engineering and Physical Sciences Research Council and by Euratom.


Corresponding Author:

Michael Loughlin

UKAEA Fusion, Culham Science Centre, Abingdon, Oxfordshire, OX14 1PR, UK

- B - Plasma Heating and Current Drive.

P3T-B-160 A REVIEW OF JET NEUTRAL BEAM SYSTEM PERFORMANCE 1994 TO 2003

Robert King, Clive Challis, Dragoslav Ciric

UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK

The operational performance of the JET Neutral Beam Injector (NBI) system during 2003 is presented and compared with NBI operation from 1994 to 2002. The paper also addresses different demands imposed on NBI operation during the JET Joint Undertaking (until the end of 1999) and the European Fusion Development Agreement (EFDA) JET Operating Contract (from 2000). The JET experimental programme in 2003 consisted of six experimental campaigns including high power, trace tritium and reverse field. The NBI system was used either for auxiliary plasma heating, or in support of various plasma diagnostics, on each of the 141 campaign days. In addition, the NBI system was operated for a further 68 commissioning days. Octant 4 Neutral Injector Box (NIB4) was operated using six 80kV/52A Positive Ion Neutral Injectors (PINIs), one 130kV/60A PINI and one 140kV/30A PINI. Octant 8 NIB was equipped with eight 130kV/60A PINIs, but due to installation and commissioning of two new 130kV/130A power supplies, only four were available before August 2003. Two more were brought into operation in August 2003 and the remaining two in November 2003. The performance figures for 1994 to 2001 were achieved with 16 PINIs. The material presented in the paper shows new operational performance records achieved in 2003, derived from data focused on average and maximum pulse lengths pulse power and injected pulse energy. During the 2003 JET experimental programme the NBI system was used to inject energy into ~2900 JET plasma pulses. The total energy injected was 163GJ, with total beam injection time in excess of 19500s. Over the last ten years the issue of JET NBI PINI reliability and availability has been of significant interest. A discussion is presented where terminology is defined, specific technical systems causing unreliability and non-availability are analysed and operational practices are reviewed. The performance analysis shows that during the period of JET operation under the EFDA contract, the NBI facility has successfully changed from high power - short pulse to high power - long pulse (10s) operation. It also shows that the sources of unreliability and non-availability have largely remained constant during this change. In particular, it is noted that the new Power Supplies have very rapidly achieved reliable operation. Conclusions are drawn on the importance of structured Commissioning procedures. The work is funded by EURATOM through the EFDA JET Operating Contract.


Corresponding Author:

Robert King

UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK

- B - Plasma Heating and Current Drive.

P3T-B-165 DEVELOPING A FULL SCALE ECRH MM-WAVE LAUNCHING SYSTEM MOCK-UP FOR ITER

Elzendoorn Bartholomeus Quirinus Sebastianes, M.P.A. van Asselen (1), W.A. Bongers (1), J.W. Genuit (1), M.F. Graswinckel (1), R. Heidinger (2), B. Piosczyk (2), T.C. Plomp (1), D.M.S. Ronden (1), A.G.A. Verhoeven (1).

(1) FOM Rijnhuizen (2) Forchungs Zentrum Karlsruhe

An ECRH (electron-cyclotron resonance heating) launching system for the ITER upper ports is being designed. The aim of the system is to inject Electron Cyclotron Waves (ECW) in the ITER plasma in order to stabilize neoclassical tearing modes (NTM). Each upper-port launcher consists of eight mm-wave lines each capable of transmitting high power up to 2 MW at 170 GHz. To avoid movable mirrors at the plasma-facing end of the launcher, the concept of remote mm-wave beam steering (RS) is used. The mock-up consists of a full-scale mm-wave system placed in a vacuum environment. The mock-up foresees in two separate vacuum systems, which simulates primary or torus vacuum and secondary vacuum. Secondary vacuum is required for the partly quasi-optical mm-wave beam trajectory and to provide a second tritium boundary. A diamond window will provide the first tritium confinement. A phased testing plan is made in order to end in Lausanne in 2006 for CW full power tests. CW high power tests require cooling on each mm-wave component. The mock-up requires two separate cooling systems. The cooling system for the square corrugated and the fixed mirror will also be used to simulate ITER baking conditions with a coolant temperature of 240 ºC and with a pressure of 4.4 MPa. The second cooling system provides cooling for the mm-wave components placed in secondary vacuum. The operational temperature for the transmission line is 150 ºC the estimated coolant pressure 1 MPa. The operation temperature in the secondary vacuum containment at ITER is 100 ºC, this temperature will be provided by heating blankets. The mm-wave system will be tested under full power CW operation; these tests will provide information about surface temperatures of mirrors and wall thermal loading of the square corrugated waveguide. The systems efficiency will be established by calorimetric measurements, and measuring antenna patterns. The lifecycle tests under influence of temperature variations, realistic coolant pressures and in a vacuum atmosphere will give the last information before the detailed final design of the ECRH launching systems can start. ‘This work, supported by the European Communities under the contract of Association between EURATOM/FOM, was carried out within the framework of the European Fusion Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the European Commission.’


Corresponding Author:

Elzendoorn Bartholomeus Quirinus Sebastianes

FOM Rijnhuizen, Edisonbaan 14, 3430 BE, Nieuwegein, The Netherlands.

- B - Plasma Heating and Current Drive.

P3T-B-171 DIGITAL MOCK-UP DESIGN OF THE REMOTE STEERABLE ITER ECRH LAUNCHING SYSTEM

D.M.S. Ronden, W.A. Bongers (a), A. Bruschi (c), I. Danilov (b), B.S.Q. Elzendoorn (a), J.W. Genuit (a), M.F. Graswinckel (a), G. Hailfinger (b), R. Heidinger (b),T.C. Plomp (a) and A.G.A. Verhoeven (a)

(a) FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Edisonbaan 14, 3439MN Nieuwegein, The Netherlands (b) Forschungszentrum Karlsruhe GmbH (c) CNR Institute for Plasma Physics, Milan

The design of a digital mock-up of the remote steerable ECRH (Electron Cyclotron Resonance Heating) top launcher is a vital part of the entire development process when working in a complex environment such as ITER. The aim is to have a digital model available that at all times represents the latest developments of the overall design. The ECRH top launcher will consist of up to 8 beam lines per upper port, each capable of delivering up to 2 MW, 170 GHz of ECW (Electron Cyclotron Wave) power to the plasma, primarily for the stabilization of NTM’s (Neoclassical Tearing Modes). For this, the need arises for the beams to be steerable. The steerable mirror mechanism is considered a critical component for the ECH&CD system since it is subject to ECW, neutron, magnetic and thermal loads and hence must be cooled during plasma operation. To avoid the need of placing the mirror’s steering mechanisms at the plasma-facing end of the launcher, the concept of remote mm-wave beam steering (RS) is used, having a corrugated square waveguide within the port-plug and the steerable optic is then placed outside of the first confinement boundary (provided by water-cooled diamond windows) of the vacuum vessel. Also a fixed mirror is placed at the end of each waveguide - inside the front shielding blanket - to steer the beam in the right direction. Careful placement of these mirrors is essential to limit the size of the shielding blanket penetration for the ECW-beams to pass through, which has to be kept to an absolute minimum. Since many of these design parameters are still converging to an optimum, a conceptual 3d-model has been created that requires little time to update. This has been accomplished to such a high degree that individual parameter values derived from the physics of millimetre wave beam propagation can be modified inside the model and will result in a direct visual feedback on the dimensional impacts that modifications have on the launcher’s total structure. Another accomplishment of the digital mock-up has been its capability to make highly accurate and adaptive beam tracings and to create complex curved mirror surfaces. This work is being carried out under the EFDA technology research programme activities, EFDA technology task TW3-TPHE-ECHULA and B1, with financial support from NWO.


Corresponding Author:

D.M.S. Ronden

FOM Institute of Plasma physics Rijnhuizen, P.O.Box 1207, 3430BE Nieuwegein, The Netherlands

- B - Plasma Heating and Current Drive.

P3T-B-172 PRE STUDY RESULTS ON HIGH VOLTAGE SOLID-STATE SWITCHES FOR GYROTRON PROTECTION.

A.B. Sterk, A.G.A. Verhoeven(1), T. Bonicelli(2), D.Fasel(3), A.Welleman(4), S. Gekenidis(4)

(1)FOM institute, Nieuwegein, The Netherlands (2)EFDA-CSU, Max Planck Institute, Garching, Germany. (3)CRPP,EPFL, Lausanne, Switzerland. (4)ABB Switzerland Ltd Semiconductors, Lenzburg, Switzerland.

Introduction A pre study to develop a concept for a high voltage semiconductor switch as a protection for gyrotrons or klystrons has been launched by the FOM institute in close co-operation with the industrial partner, ABB semiconductors, Switzerland. Solid-state switches are included in the latest ITER reference designs both for the Electron Cyclotron Heating and Current Drive and the Lower Hybrid H&CD systems. The first (prototype) switch is specified for the European Gyrotron test stand that will be built at CRPP in Lausanne. Description and Design parameters The main function of the solid-state switch is to protect the gyrotron in case of an electric arc in the cavity by disconnecting the gyrotron from the main power supply. The operating time must be no longer than 10 microsec. The total energy deposit by the arc in the gyrotron must be limited to 10 Joule. An additional function is to achieve modulation of the gyrotron power by modulating the main power supply voltage. On-Off modulation frequencies up to 5 kHz are possible in CW operation. Design parameters Technology Solid State (IGBT) Rated load voltage 60 kVDC Isolation voltage to ground 120 kVDC (10 min.) Rated load current 80 A Trip current level < 100 A Peak current limitation < 1 kA Recovery time after short-circuit < 200 ms Fast switch-off time < 10 microsec. Modulation frequency 5 kHz Current wave form Square wave, duty cycle range: 10-90% ON From the two candidates considered, IGBT and IGCT, the IGBT technology is chosen as the most favourable because of their much lower switching losses at high repetition rate (5 kHz) and the controllability through the gate at low power levels. For the IGBT solution two voltage levels are investigated (2.5 and 5.2 kV). Based on the simulation results the 2.5 kV press-pack IGBT is chosen. A 10 kV prototype assembly has been build and extensive measurements are executed. The switch-on and switch-off time, the voltage distribution and the behavior under short-circuit conditions are investigated. All possible fault conditions in the system are analysed and incorporated in the switch specifications. The paper will describe the measurements on device level and on the 10 kV assembly.The thermal management of the whole switch and a mechanical layout will be presented. This work is being carried out under the EFDA technology research programme activities, EFDA technology task TW3-THHE-CCGDS1, with financial support from NWO


Corresponding Author:

A.B. Sterk

FOM Institute for Plasma Physics p.o. box 1207 3430BE Nieuwegein THE NETHERLANDS

- B - Plasma Heating and Current Drive.

P3T-B-173 AN ALTERNATIVE ECRH FRONT STEERING LAUNCHER FOR THE ITER UPPER PORT

Rene CHAVAN, Mark HENDERSON (1) Francisco SANCHEZ (2)

(1)(2) Centre de Recherche en Physique des Plasmas, Association EURATOM - Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne, CH-1015 Lausanne, Switzerland

The purpose of the ITER electron cyclotron resonance heating (ECRH) upper port launcher will be to drive current locally inside a q=3/2 or 2 island in order to stabilize the neoclassical tearing mode (NTM). Unfortunately, the uncertainties due to our limited experience using ECCD for NTM stabilization magnified by extrapolation to ITER, result in a relatively large range of current drive densities and injection angles that may be needed on ITER. Although the remote steering (RS) launcher design offers the advantage of not requiring moving parts within the vessel vacuum boundary (far from the thermal and nuclear radiation of the plasma), it has a limited angular range and a relatively broad deposition at the resonance surface. A front steering (FS) launcher offers an extended angular range and an increased current drive density relative to the RS launcher. A FS launcher is already being planned for the equatorial port where thermal and radiation fluxes are, in fact, higher than at the upper port. In light of this, an alternative FS launcher for application on the ITER upper port is proposed, offering a wider steering angle (?±12?) and a higher ECRH power density than the planned RS launcher. Neutron streaming calculations indicate that miter bends within the plug structure are not required, and so launching systems can use straight waveguides with cross sections of ~16 cm2, which simplifies the optical layout and reduces the space requirements for the internal components. The launcher is capable of injecting over 8MW per port using a two mirror system (1 focusing and 1 steering) for focusing and redirecting the beam towards the q=3/2 or 2 flux surfaces. The steering mechanism is bearing-free with flexure pivots, in a compact cartridge capable of ±10? rotation (corresponding to ±20? for the microwave beam), with cooling tubes coiled around the body for reducing stresses to levels corresponding to ITER design requirements. A pneumatic seal-less actuator using helium integrated into the rotating mirror assembly offers a fast and precise steering response (rotation control) avoiding push-pull rods and remote actuators. The result is a complete self-contained frictionless kinematic assembly. The proposed design takes into account the specific ITER requirements on operational reliability, remote assembly and handling. The complete design concept will be presented along with a detailed comparison with the RS design.


Corresponding Author:

Rene CHAVAN

Centre de Recherche en Physique des Plasmas, Association EURATOM - Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne, CH-1015 Lausanne, Switzerland

- B - Plasma Heating and Current Drive.

P3T-B-174 DESIGN OF THE MM-WAVE SYSTEM OF THE ECRH UPPER LAUNCHER FOR ITER

Verhoeven A.G.A. (Toon), W.A. Bongers, A. Bruschi**, S. Cirant**, I. Danilov*, B.S.Q. Elzendoorn, J.W. Genuit, M.F. Graswinckel, R. Heidinger*, Kasparek***, K. Kleefeldt*, O.G. Kruijt, S. Nowak**, B. Piosczyk*, B. Plaum***, T.C. Plomp, D.M.S. Ronden and H. Zohm****

FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Nieuwegein, The Netherlands, *FZK, Karlsruhe, **CNR, Milan, ***Univ Stuttgart, ****Max-Planck, Garching

The coordination of the design of the mm-wave system to be installed in the ITER Upper Ports is being carried out at the FOM institute. The aim of the system is to inject Electron Cyclotron Waves (ECW) in the ITER plasma in order to stabilize neoclassical tearing modes (NTM). Each upper-port launcher consists of eight mm-wave lines capable of transmitting high power up to 2 MW at 170 GHz. In order to exploit the capability of ECW for localized heating and current drive over a range of plasma radii in ITER, the ECH&CD upper port launcher must have a beam steering capability. The steerable optic is considered a critical component for the ECH&CD system and to avoid movable mirrors at the plasma-facing end of the launcher, the concept of remote mm-wave beam steering (RS) is used, having a corrugated square waveguide within the launcher and the steerable optic is then placed outside of the first confinement boundary of the vacuum vessel. Starting from the gyrotrons mm-wave power will be transmitted towards the tokamak by circular evacu-ated waveguides. Steering of the beam over a range of +/- 12 will be achieved by a mirror system consisting of a combination of curved and rotating mirrors. Via the mirror system the beam will be directed into a square corrugated waveguide. A single diamond-disk window and an isolation valve will provide the tritium boundary between the pri-mary and secondary vacuum. At the end of the square waveguide, mm-wave beams will be guided through penetrations in the front-shield blanket module by a fixed mirror towards the ITER plasma. This mirror will have focusing properties in both directions. The resulting, effective steering range in the plasma is still under study but will be around +/- 8 . The design analysis has demonstrated the feasibility of the remote-steering approach in the ITER envi-ronment. Now, the detailed design of the mm-wave layout has started incorporating the remote-steering con-cept for the upper-port launcher and the aim is to come to a consistent integration into the ITER environment. Furthermore, a full-scale mock-up line is being designed and built at the appropriate ITER frequency, 170 GHz. Testing at the appropriate power level will start early 2005 at the 1.5 to 2 MW coaxial, short pulse gyro-tron at FZK, Karlsruhe. This work is being carried out under the EFDA technology research programme activities, EFDA technology task TW3-TPHE-ECHULA and B1, with financial support from NWO


Corresponding Author:

Verhoeven A.G.A. (Toon)

P.O. Box 1207, 3430 BE Nieuwegein, the Netherlands

- B - Plasma Heating and Current Drive.

P3T-B-175 DEVELOPING THE NEXT LHCD SOURCE FOR TORE SUPRA

KAZARIAN Fabienne, B. Beaumont (1) E. Bertrand (1) L. Delpech (1) S. Dutheil (1) C. Goletto (1) M. Prou. (1) A. Beunas (2) F. Peauger (2) Ph. Thouvenin (2)

(1) Association EURATOM-CEA, CEA/DSM/DRFC, CE Cadarache, 13108 St Paul lez Durance, France (2) THALES ELECTRON DEVICES, 2 rue Latécoère. BP 23. 78141 Velizy cedex France

One of the main Tore Supra objectives is to produce long and performing discharges which studies are crucial for the next step. A few years ago, the CIEL project has raised the power exhaust capability of the machine and last year, a 6 minutes fully non inductive plasma has been sustained by 3 MW of LH power. The performances of the pulses are now limited by the power injection level, and the CIMES project is targeting to improve this point [1]. Associated to the manufacture of an ITER relevant PAM launcher [2], a new klystron is under development at THALES ELECTRON DEVICES [3]. The upgrade will lead to an installed power of 12 MW in the lower hybrid transmitter. Each of the 16 tubes will work at 3.7 GHz, 76 kV, 22 A with an efficiency up to 45 %. Several performances corresponding to different operating modes must be achieved: - 700 kW CW mode on plasma (Value of Stationary Wave Ratio<=1.4), - 750 kW CW mode on matched load, - pulsed mode on vacuum during antennas conditioning phase, - diode operation at full beam parameters. Each mode presents its specific technological difficulties. On the other hand, the new klystrons will take place in our existing installation which requires high compatibility with today equipment and induces fixed parameters. A first breadboard has been manufactured and tested on the THALES test bed. It is now installed on Tore Supra Lower Hybrid test bed. A second one is under finalization. Up to now, 757 kW peak at 76 kW, 22A (50 % duty cycle) on breadboard 1 and 633 kW CW at 72.2 kV, 22A on breadboard 2 have been achieved on matched load. A prototype, derived from this 2 models, will reach full performances in June. The tube’s parameters are described in this paper. They are compared to those of the TH3103 presently settled in the transmitter, their choices are detailed and explained. The developing phases, the results obtained with the breadboards and the prototype as well as the technological difficulties are presented and analysed. [1] B. Beaumont et al. Tore Supra Steady State Power and Particule Injection : the CIMES Project Fusion Engineering and Design (56-57) (2001) [2] Ph. Bibet et al. ITER-Like PAM Launcher For TORE SUPRA LHCD This conference [3] Ph. Thouvenin et al., High Power CW Klystron for Fusion Experiments IVEC Conference (2004)


Corresponding Author:

KAZARIAN Fabienne

Association EURATOM-CEA, CEA/DSM/DRFC, CE Cadarache, 13108 St Paul Lez Durance Cedex

- B - Plasma Heating and Current Drive.

P3T-B-180 TOWARD AN LHCD SYSTEM FOR ITER

Bibet Philippe, B. Beaumont, J. H. Belo, L. Delpech, A. Ekedahl, G. Granucci, F. Kazarian, X. Litaudon, J. Mailloux, F. Mirizzi, V. Pericoli, M. Prou, K. Rantamäki, A. Tuccillo

On ITER, the LH system aims at supplying 20 MW CW for controlling the q-profiles that govern MHD stability and confinement in partially (the so-called ‘hybrid’ regime at 12MA) and fully non-inductive steady-state operation (9MA). The designed LH system relies on a transmitter made of 24 (respectively 48) 5 GHz 1MW (500 kW) klystrons. These tubes are linked to one antenna based on the PAM (Passive Active Multijunction) concept, via a 60 meters long oversized circular transmission line. The antenna geometry has been chosen to radiate the wave with a spectrum having a N// index main value of 2 at a power density of 33 MW/m2. A common European effort including several associations (CEA, ENEA, UKAEA, IST, TEKES, IPP-CZ) has been made in order to solve outstanding problems. The coupling in ITER scenario and environment matters. Experiments performed on JET have shown the possibility to couple the LH wave in environment similar to ITER. In order to verify the PAM concept, an antenna has been tested with success in close collaboration between CEA and ENEA on the plasma of FTU at the end of 2003. The reliability of LH for long pulse operation has been ascertained on Tore Supra where 370 s, 500 kA, one GJ fully non-inductive discharges have been successfully obtained thanks to 3 MW of LH power. The transmission line components and the antenna for ITER have been extensively studied within EFDA tasks. As a next step in demonstrating ITER steady state scenarios and in order to routinely realise long pulse operation (1000 s), the Tore Supra LH system is being refurbished within the CIMES project framework. The transmitter will be equipped with 16 klystrons 2103C from Thales. Their output power will be 700 kW for pulse length of 1000 s on VSWR smaller than 1.4 at a frequency of 3.7GHz. Their development is a good milestone towards the design and the realisation of a 500 kW CW 5 GHz tube. A new LH launcher based upon the PAM concept has been studied and designed. It is made of 6 rows of eight 270 degrees bi junctions fed by TE10 to TE30 mode converters that rely on the same concept than the one chosen for ITER launcher. The antenna is efficiently water-cooled, in order to allow injecting 2.7 MW CW for a power density of 25 MW/m2. The chosen technology is similar to the one envisaged for ITER. The new Tore Supra LH system will be available at the end of 2006. Its achievement is an important step to confirm the implementation of a LH system on ITER.


Corresponding Author:

Bibet Philippe

Association Euratom-CEA, CE Cadarache,13108, St Paul lez Durance, France

- B - Plasma Heating and Current Drive.

P3T-B-181 A N-PORT ERROR MODEL AND CALIBRATION PROCEDURE FOR MEASURING THE SCATTERING MATRICES OF LOWER-HYBRID MULTIJUNCTIONS

João P. S. Bizarro,

In order to routinely test and measure the scattering (S) matrices of the multijunctions that build up lower-hybrid (LH) antennae, a crucial step to ensure that the launched spectrum is well defined and has a high directivity for LH current drive, a method has been developed based on a generalization of the well-known two-port error model and calibration procedure employed by comercially available network analysers. The model presented takes into account the systematic errors inherent to microwave measurements (i.e. source and load matching, reflection and transmission tracking, directivity, and isolation), which appear as error coefficients determined via calibration standards. Furthermore, it allows for devices with any number of ports and for the use of adaptors between the measuring system and the device under test, most probably needed in the case of LH multijunctions, whose waveguides cannot be connected directly to coaxial cables.Once the calibration has been completed, the whole S-matrix of a multijunction, considered as a multiple-port microwave device, can be measured without having to carry out a long and monotonous series of operations, making thus possible to reliably test the S-matrix of any arbitrary multijunction, at every stage of the manufacturing process and as often as necessary.


Corresponding Author:

João P. S. Bizarro

Centro de Fusão Nuclear, Associação Euratom-IST, Instituto Superior Técnico, 1049-001 Lisboa, Portugal

- B - Plasma Heating and Current Drive.

P3T-B-182 DESIGN AND FABRICATION OF THE "ITER-LIKE" SINGAP D¯ ACCELERATION SYSTEM

P Massmann, HPL de Esch, R S Hemsworth and L Svensson

The SINGAP (SINGle APerture - SINgle GAP) acceleration concept is the simplified European alternative to the Japanese Multi-Aperture, Multi-Grid (MAMuG) accelerator of the ITER Neutral Beam Injector (NBI) reference design. To demonstrate ITER NBI (1 MV, 40 A) relevant beam optics in the Cadarache 1 MV, 100 mA test bed a maximum of the ITER SINGAP key parameters have been retained in the design of a new “ITER-like” prototype accelerator, i.e. optimum D­­¯ current density at 1 MeV of 200 A/m², extraction, pre-acceleration and post-acceleration gaps as per the design for ITER. Because of their complexity the extraction and pre-acceleration grid have been manufactured by electrolytic deposition of copper. The system is designed to demonstrate also SINGAP "on to off-axis" beam steering by the simple transverse displacement of the post- acceleration (SINGAP) electrode. Obtaining the current density level of >=200 A/m² is considered a crucial part of the R&D. To maximize the probability of reaching this level two negative ion sources have been developed. The first is a substantially revised, properly water-cooled, version of the prototype “Drift” Ion Source [1]. Like the extraction and pre-acceleration grids, the side walls of this source, which feature different size concentric rectangles of permanent magnet grooves and water channels, are fabricated by copper deposition. The second source, the so-called “Alternative Source”, is a completely new design trying to combine performance with ease of manufacture and low cost. Like the Drift Source, the Alternative Source is immersed in vacuum, so that there are no vacuum tight seals on the source body. The side walls are made of explosion bonded copper – stainless sandwich sheet material. Cooling channels are deep-drilled in the copper layer, and the sheet is bent into an “L”-shape perpendicular to the water channels with the copper layer inside the “L”. Two such “L’s” are put together to form the rectangular source body. The system is presently passing its acceptance tests. In the paper we will present the details of the design, the fabrication methods and the predicted performance. First results, which should be available by the time of the conference, will also be given. REFERENCES [1] A Simonin, G Delogu, C Desgranges, M Fumelli, RSI 70 (1999) 4542


Corresponding Author:

P Massmann

Association EURATOM - CEA CADARACHE, DRFC / SCCP, 13108 STYLE="PAUL LEZ DURANCE Cedex, France

- B - Plasma Heating and Current Drive.

P3T-B-185 OPERATIONAL EXPERIENCE WITH UPGRADED JET NEUTRAL BEAM INJECTION SYSTEMS

Ciric Dragoslav, Clive Challis Stephen Cox Lee Hackett David Homfray David Keeling Robert King Ian Jenkins Timothy Jones Elizabeth Surrey Adrian Whitehead David Young

UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK

In the period 2001¡V2003, the JET Neutral Beam Injection (NBI) System has been upgraded with the design goal of delivering 25 MW of deuterium beam power into the JET plasma. This major project involved the following modification of the JET NBI System: „h Modification, pre-conditioning and installation of nine 130kV/60A Positive Ion Neutral Injectors (PINIs). Eight PINIs were installed on Octant 8 Neutral Injector Box (NIB) and one on Octant 4 NIB in 2001 and 2002. „h Design, manufacturing and installation of new Box Scrapers (in both NIBs) capable of handling higher power load. „h Re-configuration and commissioning of the existing High Voltage Power Supplies (HVPS) to enable 130kV/60A operation for five upgraded PINIs. „h Procurement, installation and commissioning of two new 130kV/130A HVPS units and corresponding control systems to enable operation of four upgraded PINIs. All upgraded PINIs were conditioned without major problems, with HVPS alarms being the most frequent fault condition. Five upgraded PINIs were commissioned in 2002 and four PINIs, powered by the new HVPS modules, were brought into JET operation in summer and autumn 2003. From November 2003 JET NBI System was operated using 16 PINIs for the first time since the beginning of 2001. This lead to a record of 22.7MW of deuterium beam power injected into JET plasma in January 2004. Although a new record in NBI heating power was established, the design value of 25MW could not be accomplished due to following reasons: „h Measurements of the neutral beam power revealed that only 1.4MW (instead of 1.7MW) was delivered by one upgraded PINI. This power deficit could be attributed to the reduction in the neutralisation target caused by the neutraliser gas overheating. „h The operating voltage had to be limited to ~120kV (instead of 130kV) to prevent possible damage of the ion source back-plates caused by back-streaming electrons ¡V one such event occurred in September 2002. Technical improvements that are being carried out in the present JET shutdown (modification of the first stage neutraliser and increase of the deceleration voltage) should enable the JET NBI System to operate at the 25 MW power level in 2005. These improvements will be discussed in the paper, as well as some other issues related to full power operation of the JET NBI System (pulse duration limits, beamline and torus protection, etc.). The work is funded by EURATOM through the EFDA JET Operating Contract.


Corresponding Author:

Ciric Dragoslav

UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK

- B - Plasma Heating and Current Drive.

P3T-B-187 MAST NEUTRAL BEAM LONG PULSE UPGRADE

Gee Stephen, Andrew Borthwick Dragoslav Ciric George Crawford Lee Hackett David Homfray David Martin Joseph Milnes Tim Mutters Martin Simmonds Richard Smith Paul Stevenson Chris Waldon Simon Warder Adrian Whitehead David Young

UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK

Neutral beam heating is the main auxiliary plasma heating system on the Mega Amp Spherical Tokamak (MAST) at Culham. Until summer 2003, experiments on MAST were carried out using relatively short (200-400 ms) plasma pulses. Two Neutral Beam Injectors (NBI), each equipped with one duopigatron ion source (on loan from Oak Ridge National Laboratory), were delivering up to 3 MW of deuterium neutral beam power into the MAST plasma for the duration of up to 300 ms. During the recent shutdown, new components (central solenoid, divertor, etc.) were installed to enable long pulse operation (up to 5s) of the MAST machine. To accommodate the long pulse operation requirement, the NBI system is also being upgraded to deliver up to 5 MW of deuterium neutral beam power into the MAST plasma, for the duration of up to 5 seconds. Two duopigatron ion sources are being replaced with the JET type Positive Ion Neutral Injectors (PINIs). The MAST PINI design is a modification of the JET high current tetrode injector, with nominal deuterium beam voltage and current of 75kV and 65A, respectively. Each injector will deliver up to 2.5 MW of deuterium neutral beam power for up to 5 seconds. In addition to the replacement of the two injectors, the majority of the components of the MAST NBI system are being replaced or modified. Each beamline is now equipped with new, hypervapotron based calorimeters and residual ion dumps capable of handling long pulse/high power beams. They are instrumented with ~100 thermocouples (per beamline) to enable beam characterisation. Most of the high voltage power supplies and controls are being modified or replaced to allow long pulse operation. Some of the new features are high voltage regulation, re-application and modulation. The new beam interlock system is being installed to protect both beamline and MAST vessel components from the excessive beam power loading during fault conditions (over-pressure, magnet current mismatch, low plasma density, etc.). Data acquisition system and timer controls are also being upgraded to allow fast collection and storage of increased number of signals for considerably longer duration. The first MAST PINI will be brought into operation in summer 2004 and the second one at the beginning of 2005. The paper will address various design issues and the initial operational experience with upgraded MAST NBI system. Work partly funded by EURATOM and the UK Engineering and Physical Sciences Research Council


Corresponding Author:

Gee Stephen

UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK

- B - Plasma Heating and Current Drive.

P3T-B-188 THE ITER NEUTRAL BEAM TEST FACILITY : DESIGNS OF THE GENERAL INFRASTRUCTURE, CRYOSYSTEM AND COOLING PLANT

Cordier Jean-Jacques, R. Hemsworth (1), M. Chantant (1), B. Gravil (1), D. Henry (1), F. Sabathier (1), L. Doceul (1), E. Thomas (1), D. van Houtte (1), P. Zaccaria (2), V. Antoni (2), S. Dal Bello (2), A. Masiello (2), D. Marcuzzi (2), M. Dremel (3), C. Day (3)

(1) CEA DSM / Département Recherche Fusion Contrôlée, CEA/Cadarache, 13108 Saint Paul Lez Durance Cedex, France (2) CONSORZIO RFX, Corso Stati Uniti 4, 35127 Padova Italy (3) FZK, Institut für Technische Physik, Karlsruhe 76021, Germany

In the frame an EFDA contract (task ref. TW3-THHN-IITF1) the CEA, in close collaboration with the Consorzio RFX, Padua, and FZK, Karlsruhe, is carrying out a design of the ITER Neutral Beam Test Facility (NBTF). The main objective is to demonstrate its reliability and to optimise the performances of the main beam line components during operation, i.e. the beam source, the neutraliser, the residual ion dump, and the calorimeter. The proposed design of the Neutral Beam Test Facility general infrastructure layout is described in the paper, with taking into account the associated safety requirements (Neutrons and X-ray production). The infrastructure includes integration studies of the cooling plant, the cryosystem and the forepumping system. The ITER neutral beam heating and current drive system is equipped with a cryosorption (activated charcoal) cryopump made up of 12 panels, refrigerated in parallel by 4.5 K, 0.4 MPa supercritical helium. The pump is submitted to a non homogeneous flux of H3 or D2 gas and the absorbed flows vary from 3 Pa.m-3.s-1 to 35 Pa.m-3.s-1. The NBTF also requires a cryosystem to supply the necessary cryogens to the cryopump. The 4.5 K cryopanels must be periodically regenerated at 90 K and, occasionally, at 470 K. The cool-down times from room temperature and after regeneration depend strongly on the refrigeration capacity. Regeneration and cool-down phases of the cryopanels are evaluated for the test facility operation. The consequences of an optimised 4.5 K cold power and 80 K helium gas refrigerators on the operation plan have been analysed and will be discussed. A total power of about 50 MW will have to be removed in steady state during the two stages short and long pulse operation of the NBTF. The cooling plant and the associated pressurised water loops that are required for cooling down the high voltage components (beam source, accelerator grid, transmission line, and HV bushing) and the low voltage components (neutraliser, residual ion dump, calorimeter) are designed for both the short (20 s), and long operating pulses (3600 s) that are to be demonstrated on the test facility. The paper describes the design and the characteristics of both the optimised Primary Heat Transfer System (PHTS) and the associated Heat Removal System (HRS). A comparison is made between the cryosystem and water cooling systems proposed for the NBTF and the corresponding ITER NBI heating system reference design.


Corresponding Author:

Cordier Jean-Jacques

Association EURATOM-CEA, DSM / Département Recherche Fusion Contrôlée, 13108 Saint Paul Lez Durance Cedex France

- B - Plasma Heating and Current Drive.

P3T-B-201 PROGRESS OF THE KSTAR ICRF COMPONENTS DEVELOPMENT FOR LONG PULSE OPERATION

B.G. Hong, Y.D. Bae, C.K. Hwang, J.G. Kwak, S.J. Wang and J.S. Yoon

The ICRF system for the KSTAR tokamak [1] is being developed to support long pulse, high beta, advanced tokamak physics experiments. The system will provide a function of pressure and current density profile control by providing heating and on-axis/off-axis current drive over a range of magnetic fields with the frequency range of 25-60 MHz. And it will deliver 6 MW of RF power to plasma from 2009 with long pulse lengths operation capability up to 300 second. To transmit MW level of RF power for a long pulse, ICRF components such as antenna, vacuum feedthrough, and tuning components should have the high stand-off voltage and current without breakdown, and operational reliability. A high power density (~ 1 kW/cm2) ICRF antenna and a vacuum feedthrough which has two alumina (Al2O3, 97%) ceramic cylinders and O-ring seal have been developed. High power RF tests were performed with the antenna installed in the RF test stand. The peak voltages over 35 kVp for 300 second were found. Tuning components which use silicon oil (relative dielectric constant, 2.74) as insulating medium were developed for long pulse operation. They have a high stand-off voltage (> 40 kV) and can be used for matching during a shot by changing the level of silicon oil. Feasibility study for a coaxial fast ferrite tuner where the space between the conductors is partially filled with coaxial ferro-magnetic materials is also under investigation for matching large and fast changes.of the load, and the results are reported. The results of the development will be applicable for the long pulse, high power operation of the KSTAR ICRF system. [1] G.S. Lee et. al., “The KSTAR Project: Advanced Steady-State Superconducting Tokamak Experiment”, Nuclear Fusion 40 (2000) 575-582.


Corresponding Author:

B.G. Hong

P.O. Box 105, Yusong, Daejeon, 350-600, Korea

- B - Plasma Heating and Current Drive.

P3T-B-204 RECENT PROGRESS OF NEGATIVE ION BASED NEUTRAL BEAM INJECTOR FOR JT-60U

Umeda Naotaka, Yamamoto Takumi Larry Grisham(1) Kawai Mikito Ohga Tokumichi Akino Noboru Mogaki Kazuhiko Yamazaki Haruyuki Kikuchi Kastumi JT-60 NBI Team

Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukouyama Naka-machi Naka-gun Ibarakiken, 311-0193 Japan (1)Princeton Plasma Physics Laboratory, Po Box 451, Princeton, N.J. USA 08543

The 500keV negative ion based neutral beam injection (N-NBI) system for JT-60U was constructed in 1996, and thereafter has been in operation for study of core plasma heating and non-inductive current drive. Some modifications of negative ion source have been recently conducted so as to expand pulse duration to 30 sec from 10 sec, which is design value. Heat load on the grounded grid in the ion source was higher than the design value by three times and the temperature of water cooling for the grounded grid increased up to 90 degree. It was difficult to inject beam for the long time which the acceleration grids reached to thermal steady state. On the other hand beam limiters at NBI port are not cooled forcedly and then those temperatures increase with beam pulse duration. It is also important to reduce the heat load on the limiters in order to expand beam pulse. From the calculation of the beam trajectory, beams extracted from edge grid segments deposit largely on the limiters than inside segments. In order to lessen the heat load on the grounded grid and on the beamline limiters, outermost segments of the plasma grid extracting negative ions were masked and all the acceleration grid segments of the down stream of the masked segments were altered to the grids which had large hole to exhaust gas. By these modifications gas conductances of extractor have decreased and those of accelerator have increased. The gas pressure of the arc chamber was kept around 0.3Pa and the pressure of the extractor and the accelerator diminished about 30%. As a result striping loss of negative ion was simulated to diminish by 20%. From the measurement of the heat load of acceleration grids and beam line components, the ratio of grounded grid heat load to beam power diminished from 0.08 to 0.06 in the optimum condition and the maximum acceleration efficiency increase 0.71 to 0.79. Temperature rise of beam limiter has decreased by 35%. As a result the long pulse injection for 17 sec with 1.6MW power has been achieved at 366keV beam energy.


Corresponding Author:

Umeda Naotaka

Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1, Mukouyama, Naka-machi, Naka-gun, Ibarakiken, 311-0193, Japan

- B - Plasma Heating and Current Drive.

P3T-B-210 TESTS AND FIRST RESULTS OF A LOAD RESILIENT ICRH ANTENNA ON TEXTOR

VERVIER Michel, P. Dumortier S. Grine A. Messiaen G. Van Wassenhove

Trilateral Euregio Cluster Laboratoire de Physique des Plasmas / Laboratorium voor Plasmafysica, EURATOM Association, B-1000 Brussels, Belgium

Due to rearrangement of the diagnostic positions resulting from DED installation on TEXTOR, a new antenna system has been installed to be compatible with the inlet of the diagnostic beam between its two radiating straps. As described in [1] this antenna has been designed to test the “conjugate-T” mode of operation which is foreseen to solve the problem of generator tripping occurring during the ICRF heating of Elmy H-mode plasmas. But this antenna is also able to operate in the conventional way with pi or 0 phasing. The paper describes the installation of the antenna system. It consists of a toroidal pair of resonant straps, each strap being ended by a variable vacuum capacitor and fed by means of a tap. The "conjugate-T" mode of operation is obtained by an appropriate de-tuning of each resonating circuit strap-condenser and by means of the adjustment of the feeding line length between the tap and the “T”. The paper deals with the calibration of the antenna and line system at low power in order to allow detailed measurement of the coupling characteristics and to ensure the protection of the condensers against over-current. It describes also the analysis of the tuning procedure of the conjugate T and of the deduced practical method to optimize its performances. The mutual coupling between the two straps can reduce the performances of the conjugate-T. This problem is also analyzed. Diagnostics by means of current and voltage probes and directional couplers have been installed on the antenna system and on its feeding line. The change of phase difference between the straps which enables the load resilience is also measured. The paper will present the first results on plasma using these data and from the modeling of the complete antenna system. [1] F. Durodié et al., “Development of a load-insensitive ICRH antenna system on TEXTOR”, proceedings of 22nd SOFT, Helsinki 2002, pp. 509.


Corresponding Author:

VERVIER Michel

30 av. de la Renaissance, B-1000 Bruxelles, Belgium

- B - Plasma Heating and Current Drive.

P3T-B-211 REALISATION OF A TEST FACILITY FOR THE ICRH ITER PLUG-IN BY MEANS OF A MOCK-UP WITH SALTED WATER LOAD

MESSIAEN André, P.Dumortier R.Koch P.Lamalle F.Louche J.L.Martini M.Vervier

Trilateral Euregio Cluster Laboratoire de Physique des Plasmas / Laboratorium voor Plasmafysica, EURATOM Association, B-1000 Brussels, Belgium

A conceptual design of a 20MW ICRH plug-in for ITER in the frequency band 40-55MHz with external matching has been developed [1]. The main advantages of this design are the absence of in-vessel remotely operated components to achieve the matching and the use of passive junctions which minimises the number of matching circuits. Indeed the 24 straps of the radiating array are grouped in 4 conjugate T circuits in order to provide the highly load resilient matching needed in presence of ELMy discharges. The straps will unavoidably be coupled to each other as they are radiating in the same medium. The load resilience and the theoretical expectations of the effects of such coupling have to be checked before the installation of the antenna array on ITER with a good simulation of the plasma load. Tests in absence of plasma are useful but will not at all simulate the electromagnetic properties in presence of plasma nor allow testing the tuning algorithm. The first part of the paper shows that: (i) test with realistic plasma-like load conditions can be obtained with a large dielectric constant medium facing the strap array, (ii) when decreasing the length and increasing the frequency by the same scale factor the impedance matrix of the array remains identical, (iii) salted water can advantageously be used as a load. The second part of the paper describes the construction of the mock-up of the complete antenna array (with a scale-down factor of 5), of its feeding by passive 4-port junctions and of its water load. The addition of salt in the water avoids the use of a large tank and allows adjusting the loading properties. Measurements in the frequency range 200-275MHz provides identical impedance matrix as the full scale system and can be directly compared with modelling obtained from the CST Microwave Studio (MWS) software. Resilience of the full-scale system to ELMs can be checked on the mock-up by varying the distance array-water load. The mock-up also allows testing tuning algorithms in presence of mutual coupling and the use of polychromatic heating to decrease the effects of this coupling. [1] P.Dumortier et al., Final report on Task FU05-CT 2002-00094 (EFDA/02-675), LPP-ERM/KMS Int. Rep. 121, A.Messiaen et al. “Radio-frequency power in plasmas” ( Proc. 15th Top. Conf. On Radio-Frequency Power in Plasmas, Moran, Wyoming, May2003) AIP conf proceedings volume 694 p.142.


Corresponding Author:

MESSIAEN André

30, av. de la Renaissance, B-1000 Brussels, Belgium

- B - Plasma Heating and Current Drive.

P3T-B-218 STUDY OF MUTUAL COUPLING EFFECTS IN THE ANTENNA ARRAY OF THE ICRH PLUG-IN FOR ITER

P. U. Lamalle, A.Messiaen P.Dumortier F.Louche

Trilateral Euregio Cluster Laboratoire de Physique des Plasmas / Laboratorium voor Plasmafysica, EURATOM Association, B-1000 Brussels, Belgium

The ICRH launcher proposed for ITER is constituted by a large number (presently 24 for [1] and [2]) of short, closely packed radiating straps in order to decrease the antenna voltage. To insure the compatibility with the Elmy H-mode operation of ITER the antenna system must be insensitive to large load variations. This is achieved by grouping the straps in several “conjugate-T” (CT) matching systems. The tuning is performed by means of vacuum capacitors [1] or line stretchers [2]. The conceptual design has been made without considering the mutual coupling between the radiating straps. However, since they are radiating in the same medium, the latter are unavoidably coupled. This coupling influences the load resilience performance, considerably increases the complexity of the simultaneous tuning of various CT’s, creates significant voltage imbalances between straps, and can even result in power transfer between the different power sources feeding the array. In order to provide realistic loading conditions for the study of matching, the first part of the paper describes the properties of the impedance matrix of the array. In particular, it is shown that the mutual impedances have an important resistive component. The second part describes the effect of the mutual impedance on one CT circuit. It shows that a large load resilience can still be obtained, but that the matching conditions are more critical and that the reactive part of the mutual coupling can lead to large unbalance and phase variation between the radiated power by the two parts of the CT. Remedies and a first practical tuning method are proposed. The third part deals with the problem of the coupling between the different CT’s and their power sources. It underlines the high complexity of the simultaneous tuning because the degeneracy of the tuning values of the capacitors or line stretchers of the different CT’s is lifted by the mutual coupling. Practical tuning algorithms and the possible use of ‘polychromatic’ heating (i.e. the operation of different parts of the array at slightly different frequencies) to alleviate the adverse effects of mutual coupling are discussed. [1] Detailed Design Description Ion Cyclotron Heating and Current Drive System WBS 5.1 (DDD). [2] P.Dumortier et al., Final report on Task FU05-CT 2002-00094 (EFDA/02-675), LPP-ERM/KMS Int. Rep. 121


Corresponding Author:

P. U. Lamalle

30, av. de la Renaissance, B-1000, Brussels, Belgium

- B - Plasma Heating and Current Drive.

P3T-B-221 STATUS AND PLANS FOR THE DEVELOPMENT OF AN RF NEGATIVE ION SOURCE FOR ITER NBI

Franzen, Peter, H. D. Falter, M. Bandyopadhyay, U. Fantz, B. Heinemann, W. Kraus, P. McNeely, R. Riedl, E. Speth, A. Tanga, R. Wilhelm

Selected also for oral presentation O3A-B-221

The reference design for the neutral beam injection system of ITER is based on arc sources rated for 40 A of D- ions extracted from a 1.5 x 0.6 m2 source with a net extraction area of 0.2 m2. The main problem of the arc source is the limited lifetime of the filaments. Furthermore it is suspected that the arc current is responsible for the source non uniformity observed in large arc sources for negative ion production. Therefore RF sources, developed successfully at IPP for neutral beam heating based on H+ and D+ ions, offer substantial advantages for ITER neutral beam heating. The development of an RF ion source for negative ions has been carried on at IPP since December 2002 within the framework of an EFDA contract. So far current densities of 260 A/m2 for hydrogen and 170 A/m2 for deuterium have been achieved for an extraction area of 0.007 m2 at a source pressure of <0.5 Pa. Caesium evaporation is necessary for these high negative ion yields. The electron/ion ratio can be kept below 1 for both hydrogen and deuterium by biasing the plasma grid against the source body with 10-20 V if the filter field, i.e. a magnetic field above the plasma grid which suppresses the electrons, is sufficiently strong. Deuterium requires a stronger filter field than hydrogen. However, the useful RF power is limited by the strong filter field with the present set-up. Modifications to overcome this limitation are being prepared. An extension of the extraction area from 0.007 m2 to 0.015 m2 has already been demonstrated without loss of current density. Parallel to the source development the design and manufacturing of a test facility for pulses of up to 1 hour duration is proceeding, scheduled for commissioning towards the end of 2004. A scaled up ion source with the same width and half the length of the ITER reference source will become available for commissioning early in 2005. The paper will present as a summary an overview of the latest results of the source development, of the design of the half size ITER source and of the status of the long pulse development. The details will be presented in several other papers.


Corresponding Author:

Franzen, Peter

Max-Planck-Institut für Plasmaphysik, Postfach 1533, D-85740 Garching, Germany

- B - Plasma Heating and Current Drive.

P3T-B-225 DEVELOPMENT AND CONTRIBUTION OF RF HEATING AND CURRENT DRIVE SYSTEMS TO LONG PULSE, HIGH PERFORMANCE EXPERIMENTS IN JT-60U

Shinichi Moriyama, Masami Seki, Shunsuke Ide, Akihiko Isayama, Takahiro Suzuki, Tsuneyuki Fujii and JT-60 Team

Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukohyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan

Selected also for oral presentation O3A-B-225

The recent experiment campaign of JT-60U was started in November 2003 with emphasis on long sustainment of high performance plasmas. The maximum duration of the plasma, which was 15 sec, has been extended to 65 sec by means of modification of control systems and saving volt-sec consumption by RF and NB heating and current drive. The major purposes of this experiment campaign are; 1) long sustainment of high boot-strap current fraction, 2) long sustainment of high-beta plasma, 3) improvement of quasi steady state beta by suppression of the neoclassical tearing mode (NTM). These are important issues to the reactor. The electron cyclotron (EC) and lower hybrid (LH) heating and current drive systems play important roles in these challenges. For improvement of confinement by current profile control or by NTM suppression, EC system is effective. Movable antennas can steer beam angle to put current drive location at the mode island by real-time feedback control. The target of the EC operation in long pulse is 0.6 MW for 30 sec with 4 gyrotrons, though 10 MJ (2.8MW, 3.6sec) was recorded in high power operation before 2003. One of the critical issues for the long pulse operation is detuning due to decay in collector current of the gyrotron. The decay comes from the heater cooling by continuous electron emission. As a countermeasure for this issue, active adjustments for the heater current and anode voltage during or just before the pulse have successfully extended the duration of a good oscillation condition for the gyrotron. A "waveguide dummy load" for steady state 1MW absorption is used in these trials. Improvement in cooling of the transmission components and efforts in noise suppression have enabled the long pulse operation. As a result, 0.4 MW for 16 sec with 1 gyrotron has been achieved in March 2004. LH system is effective for current drive and is a key to extend pulse duration of reversed shear plasmas in this experiment campaign. In the LH system, the klystron was adjusted for long pulse, and the antenna mouth was newly implemented with carbon grill. The original metal antenna mouth had been partially deformed by heat load from the plasma and the RF arcing for 10 years' operation. The power handling capability and the durability for heat loads are expected to be improved by the carbon-grill-antenna. Conditioning of the antenna is under going and injection of 0.9 MW (5.1 MJ) has been achieved by March 2004.


Corresponding Author:

Shinichi Moriyama

RF Facilities Division, Department of Fusion Facilities, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukohyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan

- B - Plasma Heating and Current Drive.

P3T-B-227 RF-SOURCE DEVELOPMENT FOR ITER: LARGE AREA H- BEAM EXTRACTION, MODIFICATIONS FOR LONG PULSE OPERATION AND DESIGN OF A HALF SIZE ITER SOURCE

Kraus, Werner, B. Heinemann, H. D. Falter, U. Fantz, T. Franke, P. Franzen, D. Holtum, Ch. Martens, P. McNeely, R. Riedl, E. Speth, R. Wilhelm

At IPP RF ion sources are developed for the ITER neutral beam heating since 2002 through an EFDA contract. While most of the physical experiments are carried out with a net extraction area of 74 cm2 on the “Batman” testbed, on a second test facility (multi ampere negative ion test unit “Manitu”) the experiments are focussed on large area extraction and long pulses. In a first step the extraction area has been extended to 152 cm2 (300 apertures, Ø 8 mm). For the HV power supply a novel HV circuit has been commissioned, which utilizes two switching tubes for the generation of the extraction and the acceleration voltage. After a stable surface production of negative ions has been achieved in the cesiated source, it delivers very reproducible high H- current densities, being almost independent on the filling pressure. At 0.45 Pa with 85 kW RF power a calorimetrically measured H- current density of 20 mA/cm2 has been reached, which is consistent with the results obtained with the small extraction area. The addition of argon reduces the ion current considerably. In a second step the extraction area has been enlarged to 300 cm2 which is about the area supplied by one RF driver in the ITER size source. To demonstrate the current density and plasma homogeneity over the whole ITER extraction area, a half size ITER source has been designed and is under construction. It will have the total width and half of the length of the ITER source (800 x 900 mm2). Two 180 kW RF power supplies, and a dummy plasma grid simulating the ITER gas conductivity are foreseen. Single hole extraction and Faraday cup measurements are planned on about 20 apertures. Furthermore one testbed will be upgraded to demonstrate cw operation in D- with several 3600 s pulses in spring 2005. This requires replacing the existing titanium evaporation pumps by cryo pumps developed by FZK and installing a new calorimeter suitable for 360 kW total power and a maximum power density of 0.6 kW/cm2. New power supplies for beam extraction (15kV/35A), acceleration (35kV/15A) and RF (180 kW) are necessary as well as an upgrade of the data acquisition and cooling system. This paper will describe the results of the beam extraction experiments, the design of the half size ITER source and the modifications of the main components for long pulse operation.


Corresponding Author:

Kraus, Werner

Max-Planck-Institut für Plasmaphysik, 85748 Garching, Germany

- B - Plasma Heating and Current Drive.

P3T-B-229 DIAGNOSTICS AND MODELING OF THE PLASMA IN BATMAN RADIO FREQUENCY ION SOURCE

Tanga Arturo, M.Bandyopadhyay, H. Falter, U. Fantz, P. Franzen, B. Heinemann, W. Kraus, P. McNeely, R. Riedl, E. Speth and R. Wilhelm

Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmann str. 2, D-85748, Garching, Germany.

This paper describes the development of the diagnostics and computational activities for the negative hydrogen ion source for the neutral beam system for ITER done at IPP, Garching. Radio frequency (RF) sources have advantages of low maintenance and more operational time availability compared to the arc sources. Diagnostic measurements on the other hand have to face the difficulties of RF pick up and the modulation of the electron population. Plasma potential, density and temperature profiles are routinely obtained using a Langmuir probe, while a Mach probe has been used to provide the Mach number as well as the pattern of the plasma flow in the ion source. Measurements in the region of the plasma grid show the effect of bias as well as the pattern of the fields which determine the initial orbits of the extracted particles. The measured fluid motion is amenable to fluid dynamic analysis which has been done along the axis of symmetry. From the results of the analysis it is shown that the addition of a transverse magnetic field reduces strongly the plasma flow velocity. Modulation of plasma parameters have been used to produce accurate measurements using phase sensitive techniques. The combination of such experimental data with a Monte-carlo code for the treatment of neutrals, molecules and individual ions will help further to predict the performance in the development of a full size ITER source.


Corresponding Author:

Tanga Arturo

Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmann str. 2, D-85748, Garching, Germany.

- B - Plasma Heating and Current Drive.

P3T-B-246 ECH MW-LEVEL CW TRANSMISSION LINE COMPONENTS SUITABLE FOR ITER

Olstad, R.A., J.L. Doane (1), C.M. Moeller (1)

(1) General Atomics, P.O. Box 85608, San Diego, California 92186-5608

The ECH transmission lines for ITER will require performance parameters not yet entirely demonstrated in ECH systems on present magnetic fusion energy machines. The key performance requirements for the main ITER transmission lines are operation at 1 MW for pulse lengths of 400 s up to 3600 s (essentially cw) at a frequency of 170 GHz. An additional consideration for transmission line performance is the possibility that ITER will use 2 MW coaxial cavity gyrotrons currently under development by Forschungszentrum Karlsruhe (FZK) and other European Associations and European tube industry. This paper addresses the progress made by General Atomics in the various transmission line components suitable for use on ITER at 170 GHz, as well as at 120 GHz for plasma startup. ITER design documents call for a corrugated waveguide inner diameter of 63.5 mm; many components have already been fabricated in this diameter, and those that have been made in other diameters (namely 31.75 mm and 88.9 mm) can readily be modified to a 63.5 mm i.d. design. In some cases, water cooling must be added to present designs to remove heat deposited during cw operation of the components. In addition to the main transmission lines, there are corrugated waveguide components incorporated into the ECH launcher systems (equatorial and upper launchers). The status of the development of these components, including remotely steerable launcher components, is also presented. This paper focuses on those components needing design modifications to meet ITER requirements (i.e. frequency, power level, pulse length, diameter). The heat loads and resultant temperature increases for critical components are estimated. For those components whose temperatures will exceed safe limits, design changes already underway or planned will be addressed. Components being designed for ITER and other cw applications include Matching Optics Units (MOUs), aluminum waveguide sections adjacent to miter bends, compact dummy loads, dc breaks, waveguide bellows, stainless steel waveguides, and remote steering launcher waveguides.


Corresponding Author:

Olstad, R.A.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- B - Plasma Heating and Current Drive.

P3T-B-267 STATUS OF THE TJ-II ELECTRON BERNSTEIN WAVES HEATING PROJECT

Fernández Ángela (1), Karen Sarksyan (2) Álvaro Cappa (1) Francisco Castejón (1) Nicolai Matveev (3) Ángela García (1) Mercedes Medrano (1) John Doane (4) Charles Moeller (4) José Doncel (1) Antonio Pardo (1) Maxim Tereshchenko (2) Nicolai Kharchev (2) Alexander Tolkachev (1)

(1) EURATOM-CIEMAT Association. Madrid, Spain (2) General Physics Institute, Moscow, Russia (3) State Unitary Enterprise, Moscow, Russia (4) General Atomics. San Diego, California, USA

The present status of the main components of the TJ-II Electron Bernstein Waves (EBW) heating system and the theoretical calculations performed to determine the precise launching and beam structure conditions are presented.The O-X-B scenario has been chosen for first harmonic (28 GHz) heating of an overdense plasma. One 300kW-gyrotron (cathode voltage: 60-70kV, current: 13-25 A, pulse length: 100ms), which was used for ECR heating in TJ-IU torsatron, has been checked and is ready for installation in its cryomagnet. The design of a new high voltage power supply unit, which provides the formation of a stabilized negative voltage pulse up to 70 kV and a maximum current of 25 A, is finished. The assembly and installation should be completed at the beginning of 2005. The microwave power will be transmitted by an oversized corrugated waveguide (length: 7 m, two continuous curvature bends with an estimated overall transmission loss of about 2 to 3%, inner diameter: 45 mm, operation at atmospheric pressure). Two ellipsoidal mirrors are necessary to optimise the Gaussian beam parameters at the input of the waveguide to achieve minimal matching losses. Two corrugated mirrors are used to get the optimal polarization, so that the highest EBW absorption efficiency can be achieved. A movable internal mirror is needed in order to focus the beam and to accomplish the restrictive launching angle conditions. The support and its handling is being design and will be finished when the theoretical calculations confirm the optimal position and beam shape. The present cooling system of the two 53.2 GHz-gyrotrons of the ECRH system is being upgraded to cool the 28 GHz-complex. The extension of the system will include the water supply for the gyrotron, the HV power supply and the calorimetric system. On the primary circuit, an additional pump will be installed to supply the different components in parallel circuits, meanwhile the cooling power of the current plate heat exchanger will be increased suitably. To measure the power, a calorimeter with teflon pipes will be installed in front of the gyrotron window. A power monitor will be installed in the waveguide. This element is important to perform power modulation experiments to obtain the EBW power deposition profile. The start of the experiments is schedule for 2005.


Corresponding Author:

Fernández Ángela (1)

Association EURATOM-CIEMAT. Avda. Complutense, 22. 28040 Madrid.Spain

- B - Plasma Heating and Current Drive.

P3T-B-283 THE ASDEX UPGRADE ICRF SYSTEM: OPERATIONAL EXPERIENCE AND DEVELOPMENTS

Faugel Helmut, P. Angene, W. Becker, F. Braun, B. Eckert, F. Fischer, G. Heilmaier, J. Kneidl, J.-M. Noterdaeme, G. Siegl, E. Wuersching

The ICRF system on ASDEX Upgrade (AUG) consists of four generators with 2 MW each from 30 to 80 MHz, declining to 1 MW at 120 MHz, four two stub matching systems and four two strap antennas. The length of the antenna feeding lines allows matching at four frequencies: 30, 36.5 and 40.7 MHz, used for H minority in the 2 to 2.5 T range, and 61.7 MHz for second harmonic near 2 T. The phasing of the antenna straps is set to 0, pi. At 30 MHz, the system can be switched to asymmetric phasing for two antennas in the co-current and two antennas in the counter-current direction. ICRF has been operational on AUG since 1992. First tests in 1996 using 3 dB hybrids on two generators led to there installation on all four generators in 1998. This made operation with type I ELMs possible. ICRF has since become a reliable and powerful heating system on AUG under all conditions. The increased reliability of the ICRF further comes from: - intensive conditioning after each vent - a new system to switch off the generators - repeated use on plasma. The standard use of 4 x 0.8 MW on the first plasma shot of each day provides a renewed on-plasma conditioning - a matching program to calculate the matching ICRF delivered pulses with up to 7 s length, a maximum RF power of 7.2 MW (90% of the installed generator power) and an energy of 38 MJ. Present developments aim at using the ICRF heating increasingly in feed-back control of the discharge parameters, e.g. to keep the plasma energy constant. The huge variation of the generator output power does however raise technical problems, such as a high power dissipation of the final stage tube or a too high screen grid current. This problem can be avoided by controlling the anode voltage. In a first test on a dummy load, the anode voltage of the final stage was set to 14 kV for zero output power, increasing to 23 kV at 2 MW. The overall performance of these tests were much better than with a fixed anode voltage of 23 kV. In this case the anode power dissipation exceeded the 1250 kW limit at about 700 kW RF resulting in a switch off of the generator. Experiments will show if the anode voltage control can be implemented on plasma discharges. Longer term development to use ICRF beyond heating would benefit from an increased flexibility in the choice of frequency and phasing and from an improved antenna spectrum (using 4 straps).


Corresponding Author:

Faugel Helmut

Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr.2, D-85748 Garching, Germany

- B - Plasma Heating and Current Drive.

P3T-B-300 COOLING CONCEPTS OF THE ECRH LAUNCHER STRUCTURE AND THE TORUS WINDOWS

Roland Heidinger, Igor Danilov(1) Guenther Hailfinger(2) Klaus Kleefeld(2) Andreas Meier(1) A.G.A. Verhoeven(3)

(1) Forschungszentrum Karlsruhe, Inst. for Materials Research, P.O.Box, 76021 Karlsruhe (2) Forschungszentrum Karlsruhe, Inst. for Reactor Safety, P.O.Box, 76021 Karlsruhe (3) FOM Institute for Plasma Physics “Rijnhuizen”, Nieuwegein, The Netherlands

The upper port positions for the EC wave launching system on ITER are reserved to stabilise the Neoclassical Tearing Modes (NTM) at the q=3/2 and q=2/1 surfaces by inducing off-axis current drive. The actual mm-wave system design has defined a reference beam line based on the remote steering with focusing in the steering (poloidal) and orthogonal (toroidal) plane. The waveguide system has to be integrated into the frame of the plug (‘main structure’) and the blanket shield module. The boundary for in-vessel components in the port plug is set by a closure plate with CVD diamond ‘torus’ windows forming the primary tritium confinement to the mm-wave system.The in-vessel components including the corrugated waveguides are cooled by regular ITER blanket water from the Primary First Wall/ Blanket heat transfer system. For the cooling of ex-vessel components, a secondary cooling system is admissible, which can be the base for cooling of the torus windows. The cooling for the in-vessel parts is designed to provide single inlet and outlet pipe connections with a forced sequential flow through the walls of the main structure, the blanket shield module and the internal shields. For the waveguides an option is foreseen for lines branching off from the cooling of the internal shield. The piping includes dog legs for thermal expansion but no double containment even outside the closure plate. Only three joints are required for dismantling the structure by remote handling in the hot cells. The integrated cooling concept for the launcher with details on thermal-hydraulic performance will be presented. The CVD diamond window is exposed to non-axially symmetric thermal loads, as there is an input steering range of up +/- 12 projected at the corrugated waveguide. Accordingly the beam center is shifted by up to 27 mm off the window axis. The window structure is formed by copper cuffs which are brazed to the CVD diamond disk (aperture: 95 mm) and connected to a stainless steel flange forming the outer housing. Thermal-hydraulic and thermo-mechanical analysis was performed to show that critical stress occurs in the OFHC copper structure. The stress levels occurring for different steering angles are discussed with respect to their tolerance in relation to available yield strength in soft copper grades. This work is being carried out under the EFDA technology research programme activities (Task TW3-TPHE-ECHULA and B2).


Corresponding Author:

Roland Heidinger

Forschungszentrum Karlsruhe, Association FZK-Euratom, Institute for Materials Research I, P.O. Box 3640, D-76021 Karlsruhe

- B - Plasma Heating and Current Drive.

P3T-B-310 THE DESIGN OF THE CONTROL SYSTEM FOR THE NEUTRAL BEAM INJECTION IN HT-7

Z.M.Liu, Xiaoning Liu Sheng Liu Shihua Song Daoye Yang Yongjun Wang Liqun Hu Chundong Hu

Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031,China

The project for constructing the neutral beam injector at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) was started in 2002 that is based on single injector with one arc discharge source, which can deliver 700KW of neutral beam power at the Princeton Large Torus (PLT), USA. Initial testing during Dec. 2003 to Feb. 2004 has produced arc current up to 100 A rate for 400 msec here. The paper consists of two parts. In the first part the distributed control system, which is the latest procedure control system that can achieve the concentrate synthetically management in NBI are described. The second part detailed introduces the design of each constituent part of total control system in NBI, which can complete accurate sequence control system of the power supplies and the vacuum valves, the data acquisition and data processing. The interlock protection system on the site is based on the programmable logical controller (PLC) system.


Corresponding Author:

Z.M.Liu

Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031,China

- B - Plasma Heating and Current Drive.

P3T-B-312 EXPERIMENTAL STUDY ON UNIFORMITY OF H- ION BEAM IN A LARGE NEGATIVE ION SOURCE

Hanada Masaya, T.Seki, oue, T.Morishita, T.Mizuno1), A.Hatayama1), T.Imai, M.Kashiwagi, M.Taniguchi, K.Watanabe

Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken,311-0193, Japan 1)Faculty of Science and Technology, Keio University, Hiyoshi, Yokohama 223-8522, Japan

The origin giving the non-uniformity of the negative ion density in the JT-60U negative ion source was experimentally studied in the JAERI 10A negative ion source that is the same type as the JT-60U negative ion source. Namely, the negative ion source has two different electron temperature regions divided by a transverse magnetic field (filter field) forming uniform magnetic field along the longitudinal direction. Negative ions are produced in the plasma with low electron temperature (<1eV) where a plasma grid (PG) is situated. The negative ions are extracted from an ion extraction area of 15 cm x 35.6 cm. The longitudinal length is one-third of the JT-60U negative ion source. Correlation between beam profiles and the plasma parameters such as an electron temperature was examined. The spatial beam intensity along the longitudinal direction was relatively low in the upper half region as observed in the JT-60U negative ion source. The electron temperature near PG was also non-uniform even for the uniform filter filed, i.e., 1~3.5eV in the upper half region and < 1eV in the lower half region. This high electron temperature in the upper half region was also observed in the computer simulation. Some high-energy electrons emitted from the cathode leak to the plasma grid beyond the filter field through the channel of a weak magnetic field of <10Gauss near the top of the negative ion source. Since the cross-section for the destructive reaction of the H- ions via a collision with electrons rapidly increases above 1 eV, it is predicted that non-uniformity of the H- ion density in the longitudinal direction is caused by leakage of high-energy electrons to the plasma grid. To confirm this prediction, a 50 mm x 50 mm plate for intercepting the high-energy electrons leaked to the plasma grid was placed on the electron path predicted from the simulation, i.e., at 7 cm from the plasma grid and 2.5 cm from the top wall of the negative ion source. The electron temperature in the upper half region was cooled to 1 eV. This electron cooling dramatically improved beam profile, in particularly, in the upper half region of the longitudinal direction. This resulted in a 20% gain of the beam current. From this result, it was clarified that the leakage of the high-energy electrons to the plasma grid is one of origins for non-uniformity of the negative ion density in the large negative ion source.


Corresponding Author:

Hanada Masaya

Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken,311-0193, Japan

- B - Plasma Heating and Current Drive.

P3T-B-314 DEVELOPMENT OF RELIABLE DIAMOND WINDOW FOR EC LAUNCHER ON FUSION REACTORS

TAKAHASHI Koji, S. Illy*, A. Kasugai, K. Sakamoto, R. Heidinger*, M. Thumm*, R. Minami and T. Imai

Japan Atomic Energy Research Institute, 801-1, Mukoyama, Naka, Ibaraki 311-0193, JAPAN * Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe, Germany

A diamond window is one of important components in an electron cyclotron (EC) launcher (antenna), which controls the injection of millimeter wave power into plasma for electron cyclotron heating and current drive(EC H&CD). The window must have two important functions. One is the capability of high power millimeter wave(RF) transmission. In ITER, for example, a 1MW transmission is required and has been confirmed. Another is that to provide the vacuum and tritium barrier window, which must be the reliable structure, between the EC launcher (vacuum vessel) and transmission lines of EC H&CD system, whose study is reported here. When high power RF transmits through the window, it is heated up due to dielectric loss. Therefore, the diamond window is designed to cool its disk edge to eliminate the heat deposition. Then, if it is assumed that a crack is generated toward the window edge, for instance, by arcing or unexpected mechanical stresses on it, the cooling medium could leak into the launcher attached to a vacuum vessel and the transmission line. In order to avoid this possible event, the new diamond window with the copper-coated edge has been developed. In addition, water can be used for the cooling without corrosion of aluminum blaze between the diamond disk and the Inconel cuffs since the blaze is completely covered by Cu. To form the Cu layer on the edge, a Ti alloy is, at first, metalized on the edge surface. Then, copper is electroformed on its edge, entirely. The thickness of the layer is 0.5mm. A 170GHz, RF transmission experiment of the new diamond window, which is equivalent to a MW-level transmission, was carried out to investigate that the Cu coated window is capable of the edge cooling. The RF power of 55kW and 120kW with the pulse length up to 3sec was transmitted through the window. Temperature increases of 45 deg and 100 deg were obtained at each RF power and they almost became constant. Thermal calculation with loss tangent of 4.4E-4 and thermal conductivity of 1.9±0.1kW/m/K was also carried out and the result agrees with the experiment. Since the loss tangent of the diamond used for the experiment is 4.4E-4, much higher than the actual diamond disk (loss tangent=2.0E-5), the temperature increases correspond to those of the 1MW and 2MW transmission, respectively. It concludes that the Cu coating on the edge dose not degrade the edge cooling capability of the diamond window and improves the reliability of the diamond window.


Corresponding Author:

TAKAHASHI Koji

Japan Atomic Energy Research Institute, 801-1, Mukoyama, Naka, Ibaraki 311-0193, JAPAN

- B - Plasma Heating and Current Drive.

P3T-B-320 DESIGN OF HIGH POWER COAXIAL DC BREAK FOR ADITYA TOKAMAK

Mukherjee Aparajita, D.Bora (1) Raghuraj Singh(1) H. M. Jadav(1) B. Kadia (1) R.A. Yogi (1) Bhattacharya D.S.(2) RF Group (1)

(1) Institute for Plasma Research, Bhat, Gandhinagar – 382428. (2) Variable Electron Cyclotron Centre, Kolkata. (India).

ADITYA tokamak has been upgraded with the inclusion of Ion Cyclotron Resonance Heating (ICRH) system. A 20 – 40 MHz, 200 kW ICRH system has been integrated to increase the plasma energy content. The complete ICRH system has been indigenously designed, fabricated in-house including the RF generator. ICRH system consists of rf generator, Tx-line, matching network (consists of stub tuner and phase shifter), vacuum Tx-line and antenna. Outboard Fast Wave antenna is used as radiating element into the plasma. The return path of the antenna will be directly connected to the vacuum vessel to avoid any unwanted backside radiation. DC break in the transmission line is required to isolate the vacuum vessel from the HV power supply ground, to which the transmitter is connected. A high power coaxial dc break is designed, fabricated and tested for wide band frequency operation (20 MHz – 40 MHz) for blocking of dc on both inner and outer conductors. Design of dc break, which is essentially a l/4 system for both the inner and outer conductors, is being done using ANSOFT software. Current density is kept less than 10 Amp/cm2. Separation between the conductors is kept in such a way so that it can withstand high voltages during mismatch. While designing, VSWR and insertion loss are kept below 1.05 & 0.1 dB respectively for central frequency operation. Low power tests using HP8753E VNA shows that dc break can be used from 22 MHz to 40 MHz with minimum attenuation and 1.25 (Max. at 22 MHz) VSWR. To test the performance away from center frequency, a high power test at 65 kW is conducted at 24 MHz on test bench, which is in good agreement with low power test. In this paper, detailed design and testing of the high power dc break will be presented.


Corresponding Author:

Mukherjee Aparajita

Institute for Plasma Research, Bhat, Gandhinagar – 382428

- B - Plasma Heating and Current Drive.

P3T-B-332 W7-X NEUTRAL-BEAM-INJECTION: TRANSMISSION, POWER-LOAD TO THE DUCT AND INNER VESSEL AND CONSEQUENCES OF THE STELLARATOR STRAY FIELD

N. Rust, M. Kick, E. Speth

The new stellarator W7-X will be equipped with two ASDEX-Upgrade (AUG) like Neutral-Beam-Injector-Boxes for balanced injection. Each of them has the capacity to be equipped with 4 PINI-sized sources of 2.5 MW heating power per source. Because of the good confinement for fast ions of W7-X it is possible to use a nearly radial injection geometry with an injection angle of 7.5 in the duct. This work will describe the result of calculations with the neutral-beam-transmission code DENSB for W7-X. The result is not only the transmission through the duct but also the power load on the duct and the W7-X inner vessel by the NBI. The W7-X NBI duct has approximately the same size as AUG. Since however some W7-X coils are in direct proximity to the duct there are some bottlenecks that limit the transmission. The total transmission for four sources per box is 94% for a beamlet divergence of 1 . The source with the best transmission 96.7 % has nearly no limitation by the coils. But even the source with the greatest limitations by the coils has still an acceptable transmission of 89.7%. The narrow parts of the duct have special requirements for protection and cooling. The Neutral-Beam also hits the inner wall of the W7-X plasma vessel. In case of no plasma the full Neutral-Beam power will reach the inner wall. In this case the Neutral-Beams have to be switched off quickly, because during this time the maximum power deposition on the plasma vessel by the NBI is 47 MW/m2. During the normal plasma operation only the shine trough will hit the inner wall. The power load for the inner wall depends strongly on the plasma density. As a conclusion the NBI pulse length is limited for low plasma densities. For example the target modules will allow a NBI pulse length of 10 s for a central density larger than 5E19 m-3. The geometry of the inner vessel in the region of the NBI power load is quite complex. It consists of normal wall elements. But the W7-X divertor, baffle and one port are also affected. All these in vessel elements have to stand the NBI power load. Because W7-X has superconducting coils, the B-field is not switched off between neutral beam pulses. The stray field of W7-X has a maximum of 300 Gauss near to the NBI-Boxes. This is too much for the AUG like titanium getter pumps. A sufficient shielding with huge iron masses is impossible. So the W7-X NBI may have to operate with fast cryo pumps.


Corresponding Author:

N. Rust

Max-Planck-Institut für Plasmaphysik, Association EURATOM-IPP, D-85748 Garching

- B - Plasma Heating and Current Drive.

P3T-B-337 DIAGNOSTICS OF THE CESIUM AMOUNT IN A RF NEGATIVE ION SOURCE AND THE CORRELATION WITH THE EXTRACTED CURRENT DENSITY

Fantz Ursel, M. Bandyopadhyay, H. D. Falter, P. Franzen, B. Heinemann, W. Kraus, P. McNeely, R. Riedl, E. Speth, A. Tanga, R. Wilhelm

Neutral Beam Injection based on negative ion sources will be a major heating system of ITER. Besides arc sources RF driven ion sources are promising candidates. Negative hydrogen ions (H- and D-) can be formed by plasma volume and surface processes. Due to a short survival length of negative ions in the plasma the surface process at the extraction grid is favoured. The optimisation of the surface process requires a high atomic hydrogen density and a surface with a low work function for which cesium is commonly used: H + Cs -> H-. One of the main tasks is to achieve a thin and homogeneous Cs coverage of the extraction grid by Cs evaporation in the discharge volume with the boundary condition of minimising the consumption of cesium and maintaining a constant Cs coverage at the grid. In order to quantify the amount of cesium in an RF discharge optical emission spectroscopy is used as diagnostic tool. Suitable diagnostic lines are identified. Their emission is observed using a line of sight parallel to the extraction grid at a distance of approximately 4 cm. Therefore, the evaluation of line emission refers to the amount of cesium in this plasma volume. The volume density of cesium is obtained from a simplified population model using the corresponding rate coefficients for electron impact excitation. The observation of lines from Cs ions allows an estimation of the ion density of cesium. A method to deduce particle fluxes from the grid into the plasma is introduced as well. Results will be presented for Cs containing hydrogen and deuterium discharges with and without additional Cs evaporation. The influence of admixtures of rare gases on cesium will be shown. A correlation of line emission with extracted current density of negative ions will be discussed. However, one has to keep in mind that emission spectroscopy refers to the Cs amount in the observed plasma volume whereas the extracted current density reflects the efficiency of the surface process. The obtained data will be used as a basis for a transport model.


Corresponding Author:

Fantz Ursel

Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr.2, D-85748 Garching, Germany

- B - Plasma Heating and Current Drive.

P3T-B-345 LONG PULSE OPERATION ON THE KAMABOKO III ION SOURCE.

Deirdre Boilson(2), A R Ellingboe(2), H P L de Esch(1), R Faulkner(2), , R S Hemsworth(1), , A Krylov(1), P Massmann(1) and L Svensson(1)

(1)Association EURATOM-CEA, CEA/DSM/DRFC, CEA-Cadarache, 13108 STYLE="PAUL-LEZ-DURANCE (France) (2)Association EURATOM-DCU , PRL/NCPST, Glasnevin, Dublin 13, Ireland

Development of negative ion sources is being carried out at the DRFC, Cadarache on the KAMABOKO negative ion source in collaboration with JAERI, Japan. The target performance is to accelerate a D- beam, with a current density of 200 A/m2 with <1 electron extracted per accelerated D- ion, at a discharge power of <2 kW per litre of source volume, at a pressure of 0.3 Pa. For ITER a continuous ion beam must be assured for pulse lengths of £3,600 s. Beam pulses of 1000 s have been demonstrated, but the current density at the expected arc power and pressure was found to be to be low in comparison to the anticipated ³200 A/m2. Accelerated currents of 320 A/cm2 have been accelerated for long pulse operation, but the transmission to the calorimeter is only »50 % s. The loss of accelerated power is being investigated using additional electrical and thermal diagnostics. During long pulse operation increasing the temperature of the plasma grid increases the negative ion yield by £40%, substantially below that expected (100%). It has been shown that if the ion source walls are kept cold (<36 C) the increase in negative ion yield with increased plasma grid temperature can be >60% In an effort to understand the effect of Cs on the source behaviour and the reason for the low H- yield a model of the dynamic behaviour of the Cs in the source was proposed and investigated. A cold Cs trap was installed into the source onto which the Cs which would condense, and then the rate that Cs enters the plasma could be controlled by increasing the temperature of the trap. Additionally, recent experiments suggest that the Cs injected into the source is rendered unusable due to its “burial” under tungsten evaporated from the filaments. A Cavity Ringdown Spectroscopy system (CRDS) has been installed on MANTIS which will allow the quantitative determination of the D- (or H-) density »10 mm in front of the plasma grid. This will allow a quantitative determination of the negative ion line density. If successful these data will be presented. This paper will outline the aforementioned experiments and discuss the poor performance of the source in long pulse operation.


Corresponding Author:

Deirdre Boilson(2)

(2)Association EURATOM-DCU , PRL/NCPST, Glasnevin, Dublin 13, Ireland

- B - Plasma Heating and Current Drive.

P3T-B-351 DESIGN AND TEST OF A HV DEVICE FOR PROTECTION AND POWER MODULATION OF 140 GHZ/1MW CW-GYROTRONS USED FOR ECRH ON W7-X

Brand Peter, Braune Harald (1) Müller G. A. (2)

(1) Max-Planck-Institut für Plasmaphysik, Wendelsteinstr. 1, D-17491 Greifswald. (2) Institut für Plasmaforschung, Universität Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart

The improvement of energetic efficiency of ECRH of fusion plasmas could be realized due to the development of gyrotrons with beam energy recovery by a voltage depressed collector. Gyrotrons of this type for pulsed operation were applied successfully at the W7-AS stellarator. Control of the gyrotron output power was realized by modification of a HV-amplifier used for feeding the gun- anode of a triode type gyrotron used before. For plasma heating by ECR in the stellarator W7-X under construction, 140 GHz gyrotrons with depressed collector and 1MW cw output power have been developed. These gyrotrons are fed by two high voltage sources: a high power supply for driving the electron beam and a precision low power supply for beam acceleration. In addition a protection system with a thyratron crowbar for fast power removal in case of gyrotron arcing is installed. The low-power high-voltage source for beam acceleration is realized by a HV servo amplifier driving the depressing voltage, which can be modulated by feeding an adequate modulation signal to the reference port of the servo amplifier. This new amplifier, designed for cw-operation, contains two high voltage tetrodes working in push-pull giving an acceleration voltage swing up to 15 kVpp at a rise time of 750 V/ìs on a capacitive load of 1 nF. Furthermore the influence of the voltage noise of the main high power supply on the acceleration voltage is suppressed by feedback control. The current of the gyrotron electron beam is controlled by the cathode temperature. Therefore a precision ac/dc source is part of the crowbar desk. In connection with an internal PLC (Siemens SPS) linked by Profibus optical fiber transceiver to the remote system control, monitoring and setting of all relevant parameters is possible on the time scale of the data aqusition of the PLC. For monitoring and control of signals up to a frequency of 100 kHz ADC- and DAC-front ends linked by optical fibers have been developed. In the paper a description of the different modules of the system is given. The results of the operation of the prototype device in conjunction with a gyrotron are presented.


Corresponding Author:

Brand Peter

Institut für Plasmaforschung, Universität Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart

- B - Plasma Heating and Current Drive.

P3T-B-353 DEVELOPMENT OF CW AND SHORT-PULSE CALORIMETRIC LOADS FOR HIGH POWER MILLIMETER-WAVE BEAMS

Bruschi Alessandro, Sante Cirant (1) Franco Gandini (1) Giuseppe Gittini (1) Gustavo Granucci (1) Vittoria Mellera (1) Valerio Muzzini (1) Antonio Nardone (1) Alessandro Simonetto (1) Carlo Sozzi (1) Nicolò Spinicchia (1) Giuliano Angella (2) Enrico Signorelli (2)

(1) Istituto di Fisica del Plasma CNR-EURATOM-ENEA, v.Cozzi 53, 20125 Milano, Italy. (2) Istituto per l'Energetica e le Interfasi, CNR, v.Cozzi 53, 20125 Milano, Italy.

With the development of high power gyrotrons for fusion research, increased power handling of beam dumps is required during the test phase of mm-wave systems. The design of the optics and the techniques suitable for building a compact matched load for high vacuum operation, was developed, leading to two designs: one is capable of 1 MW CW with proper cooling, the second is convenient for precise measurements of short pulses (2MW, 0.1s.). Tests of the first version at 140 GHz, more than 0.5 MW and several seconds of pulse lengths are envisaged at the ECRH plant built for the W7-X stellarator (in Greifswald), during the remote steering antenna tests for ITER ECRH upper launcher. For both loads the spherical internal geometry is the same used in the previous ones installed in the Frascati Tokamak Upgrade (FTU) ECRH Plant. The first CW sphere shell is cast with a cooling pipe with inner diameter of 25 mm directly inserted in the wall thickness: it allows heat removal with high efficiency with a water velocity of around 10 m/s. The short-pulse load has a thin copper shell with a tube electroformed on the outside. Tube length, width and water flow rate were optimised to give a good sensitivity for the instantaneous power and integrated energy measurements, derived by water temperature jump and flow rate. Vacuum tests and X-ray analysis on the first cast shell showed problems of tube adhesion to the casting, whose effects have been evaluated with thermal simulations. Problems were solved in the second shell with a different casting procedure. The validity of the design was evaluated by thermal and structural FEM analysis, both with uniform and realistic wall loading, obtained with analytic and ray tracing modelling of the power distribution in the sphere interior; indication for improvements in the cooling arrangement and power deposition were obtained. New components were designed: a cooled, vacuum compatible vibrating mirror; a back-reflecting pre-load with a dedicated section for pumping, and a diagnostic flange for monitoring the inner coating temperatures. The effect of the pre-load was evaluated with the same ray tracing model used for the power distribution. New millimeter-wave measurements at low power and heavy duty tests on coating materials show a margin for improvements in coating which could be exploited in combination with a new mirror geometry, aiming at a higher power capability.


Corresponding Author:

Bruschi Alessandro

Istituto di Fisica del Plasma CNR-EURATOM-ENEA, v.Cozzi 53, 20125 Milano, Italy.

- B - Plasma Heating and Current Drive.

P3T-B-356 ITER-LIKE PAM LAUNCHER FOR TORE SUPRA’S LHCD SYSTEM

J. H. Belo (1), Ph. Bibet, J. Achard, B. Beaumont, B. Bertrand, M. Chantant, Ph. Chappuis, L. Doceul, A. Durocher, L. Gargiulo, M. Missirlian, A. Saille, F. Samaille, E. Villedieu

(1) Centro de Fusão Nuclear, Associação Euratom-IST, Instituto Superior Técnico, 1049-001 Lisboa, Portugal

Advanced scenarios such as those envisaged for ITER require the development of a novel generation of LHCD systems to achieve a very efficient cooling of the launcher, an essential necessity to remove the heat induced by the neutron flux, the plasma radiated power and the RF losses. To meet these demanding goals a new and innovative antenna based on the PAM concept (Passive-Active Multijunction) [Bibet, P., Litaudon, X., Moreau, D., Nucl. Fusion, vol. 35, 1213 (1995)] already proposed for ITER has been designed to be tested in Tore Supra. It will launch 2.7 MW CW at 3.7 GHz with a power density of 25 MW/m2, radiating a power spectrum peaked at N//=1.7 with a maximum power directivity near the electron cut-off density and with very good coupling properties. This work has a threefold purpose. 1) To give a description of the antenna and of its manufacturing and assembling processes: it uses eight klystrons to power sixteen TE10-TE30 mode converters, each feeding its own three H-plane poloidal junction in turn connected to three E-plane bi-junctions with 270 phasing, the antenna’s front part being made by linking plates of OFHC copper to stainless-steel sheets via explosive diffusion bonding. 2) To study and optimise its RF components: the mode converter in terms of conversion efficiency, overall SWR and balanced power splitting capabilities, the first RF measurements of its prototype being presented; the main waveguide for an optimal transmission at the fundamental mode TE10 and dampening of higher modes, while avoiding reflection to the klystrons; the bijunction length to enhance the plasma coupling; the impact of the mouth profile in the poloidal and toroidal directions will be considered as will be the losses induced by the use of copper in the whole antenna; the necessary measuring devices and their deployment will be defined in particular the waveguide-coaxial transition used for measuring the S parameters; a study of the stability of the launcher to changes in the reflection coefficients at the output ports will be undertaken to better ascertain its behaviour with varying plasma properties. 3) To analyse its thermo-mechanical behaviour: thermal and mechanical stress analysis taking into account the plasma radiated flux at the mouth and the RF losses; additional mechanical stresses due to the eddy current induced in the launcher by disruptions combined with the residual toroidal magnetic field have been computed as well.


Corresponding Author:

J. H. Belo (1)

Association Euratom-CEA, CE Cadarache,13108, St Paul lez Durance, France

- B - Plasma Heating and Current Drive.

P3T-B-359 LARGE CRYOSORPTION PUMP OF THE TEST STAND FOR THE KSTAR NBI SYSTEM

IN SANG RYUL, W. S. Song, T. S. Kim, B. H. Oh

same as above

A test stand was built for developing and examining the ion sources and beam line components to be installed in the KSTAR NBI system. The test stand is equipped with a 60 m3 vacuum chamber, an ion source, and one set of beam line components. In the test stand, the hydrogen ion beam of maximum 2.8 MW (80 keV, 35 A) will be produced with one ion source. Considering the ionization efficiency of 40~50%, the ion source must be supplied with the hydrogen gas at a rate of up to 700 sccm to attain the beam current of maximum 35 A ion beam. Moreover, the gas supply rate to the neutralizer should be at least 2000 sccm to keep the average pressure higher than 3×10-3 mbar. In spite of such a large gas load, the chamber pressure should be low enough not to diminish the neutral beam generated in the neutralizer. The key point in designing the vacuum pumping system for the NBI test stand is how to evacuate the NBI chamber to the pressure of less than 10-4 mbar when the gas throughput is a few thousands sccm. The vacuum pump to fulfill such a requirement should have a pumping speed of around 500,000 L/s. The only reasonable solution to this problem is to use an in-chamber cryopump that can utilize the maximum pumping area available in the chamber. The cryo-pumping system of the NBI test stand is composed of four cryosorption pump bodies, four G-M helium refrigerators and four LN2 bottles of 150 L each. The main component of the pump body is a 20 K cryosorption panel cooled by a G-M refrigerator. The cryopanel consists of 4 identical AC-coated rectangular plates of 145 mm×1000 mm brazed to a center rod at intervals of 90 . The baffle and the lower thermal shield are cooled by liquid nitrogen. The baffle consists of 50 chevron blades of 120 bending angle, each has a LN2 hole of 5 mm diameter along the center axis of the blade. The chevron blades form as a whole a circular ring of 550 mm O.D and 356 mm I.D. The liquid nitrogen level in the baffle blade is controlled by the weight and the vapor pressure of liquid nitrogen in the bottle. The cooling down time of the cryopanel to 20 K was about 6 hours with a liquid nitrogen consumption rate of about 35 L/hr. The maximum pumping speed of the cryosorption pump for the hydrogen gas measured by the steady pressure method was about 90,000 L/s.


Corresponding Author:

IN SANG RYUL

Korea Atomic Energy Research Institute, Dukjin-dong 150, Yuseong-gu, Daejeon, 305-353, Korea

- B - Plasma Heating and Current Drive.

P3T-B-362 DEVELOPMENT OF A RF SOURCE FOR ITER NBI: FIRST RESULTS WITH D- OPERATION

Speth, Eckehart, H.D. Falter, P. Franzen, B. Heinemann, M. Bandyopadhyay, U. Fantz, W. Kraus, P. McNeely, R. Riedl, A. Tanga, R. Wilhelm

Max-Planck-Institut für Plasmaphysik, D- 85748 Garching, Germany, EURATOM-Association

ITER NBI requires among other elements a powerful beam source that delivers 40 Amperes of D- accelerated to 1 MeV. Because of the low current densities accessible with negative ions, a large source of 1.5 x 0.6 m2 cross section is required with a net extraction area of about 0.2 m2. So far the reference design is based on an arc discharge source, which however suffers from reduced availability and increased maintenance effort due to the limited life of the filaments. As an alternative a RF source is being developed at IPP Garching within the frame of an EFDA contract. The aim of this development is to demonstrate the required D- current density of 200 A/m2, accelerated to about 30 KeV and from a reduced extraction area. The target current density is subject to the additional requirement of low source pressure (< 0.3 Pa) and low co-extracted electron fraction (< 1). Till the end of 2003 the experiments had been restricted to hydrogen operation due to the neutron radiation implications of deuterium. After having implemented remote operation of the BATMAN testbed, deuterium operation started recently utilising a small extraction system of 0.007 m2. This paper reports the first results with deuterium. After some optimisation concerning Cs operation current densities of 260 A/m2 for hydrogen and 170 A/m2 for deuterium have been achieved in the right pressure range. In both cases the electron/ion ratio can be kept below 1 in a cesiated source. However, this requires biasing the plasma grid against the source body with 10-20 V on the one hand and a sufficiently strong magnetic filter on the other hand. It appears that deuterium requires a stronger filter field than hydrogen. In the present filter configuration (external permanent magnets) the useful RF power seems to be limited, possibly due to the large source volume filled by the magnetic field. The paper will describe the modifications to overcome this limitation by studying different filter concepts. An interesting side effect is the fact, that the neutron production rate is about a factor 40 lower than expected from positive ions. The paper will discuss possible reasons for this.


Corresponding Author:

Speth, Eckehart

Max-Planck-Institut für Plasmaphysik, D-85748 Garching, Germany, EURATOM-Association

- B - Plasma Heating and Current Drive.

P3T-B-364 PERFORMANCE TEST OF THE LH ANTENNA WITH CARBON GRILL IN JT-60U

SEKI Masami, MAEBARA Sunao, MORIYAMA Shinichi and FUJII Tsuneyuki

Japan Atomic Energy Institute, Naka Fusion Research Establishment 801-1 Mukoyama Naka-machi Naka-gun Ibaraki-ken 311-0193, JAPAN

Current profile control using lower hybrid (LH) wave is remarkably useful, for example, to obtain reversed shear plasmas with higher confinement property. LH wave injection through a multijunction-type antenna (LH antenna) has been contributing to various experiments in JT-60U during 10-year operation. This LH antenna, however, was damaged due to excessive heat loads such as plasma bombardments and rf break downs around its mouth. Then the injection power gradually decreased year by year. To recover the injection power, a carbon grill is installed on the LH antenna mouth in JT-60U. It is a reason of adoption of carbon why a kind of carbon material has high resisting power against heat load and low Z number. The carbon grill consists of a base frame, an rf conductor and a carbon mouth. The base frame is welded with the original LH antenna made of stainless steel. The rf conductor of thin copper plate is used to improve electrical conductivity between the base frame and the carbon mouth. The carbon mouth is made of Graphite and/or CFC, and is held on the base frame by 22 bolts. Therefore if the carbon mouth is damaged, it will be changed. It is possible to compare between status of 6-Graphte type grills and that of 2-CFC type ones after experiment campaign. After construction of the newly developed LH antenna with the carbon grill, conditioning of the LH antenna has favorably done with and without plasmas. Through 9-day conditioning operation with plasmas, rf energy of ~5 MJ (~0.9 MW x 7.1 sec with duty cycle of 84 %) was injected into plasma. In this shot, good coupling property was obtained such as reflection coefficient of ~5 % by controlling plasma-antenna distance. It is found by dropping in one-turn voltage that about 60 % of plasma current of 1 MA was driven by LH injection. Thus performance test of the LH antenna with the carbon grill is under going successfully.


Corresponding Author:

SEKI Masami

801-1 Mukoyama Naka-machi Naka-gun Ibaraki-ken 311-0193, JAPAN

- B - Plasma Heating and Current Drive.

P3T-B-371 FIRST RESULTS OF THE TORE SUPRA ITER LIKE ICRF ANTENNA PROTOTYPE

K. Vulliez, S. Brémond, G. Agarici, B. Beaumont, G. Bosia, B. Cantone, J.M Delaplanche, L. Doceul, G. Lombard, L. Millon, P. Mollard, E. Pignoly, S. Poli, B. Saoutic, E. Villedieu, D. Volpe

Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 St Paul lez Durance (France)

Reliable coupling of Ion Cyclotron Range of Frequency power to plasma in high confinement ELMy regimes is an essential target for ICRF systems. With the present ICRF systems, the fast (typically of the order of 100 ms) and big rise of the antenna loading (typically by a factor of 5) due to the ELMs results either in the tripping of the RF generators or in reducing the coupled power if hybrid junction are used to isolate the RF generator from the antenna, in any case reducing the mean power brought to the plasma. A new antenna RF configuration, now know as the conjugate -T scheme matched on low impedance, was recently proposed and adopted as reference design for ITER ICRF system [1]. The Tore Supra ITER-like ICRF antenna prototype project was initiated in mid 2002 in Cadarache in order to get as quickly as possible some results on this new scheme at reduced cost, as it was developed by modifying the existing ORNL Tore Supra antenna [2]. It aims to be a tool to harvest experience and understandings in the operation of this new type of antenna, in particular valuable for the ITER-like JET-EP antenna. The prototype antenna was first tested on test-bed, then assembled on TS, and the first successful ICRF power coupling on plasma was obtained in February 2004. After a review of the mechanical design, first RF results and lessons learned will be discussed in the paper, including sensitivity of the matching due to the effects of mutual coupling between straps on this low impedance matching scheme, load tolerance performance, antenna loading, power handling. [1]G. Bosia, ITER Joint Central Team, Garching, Germany,High-power density ion cyclotron antennas for next step applications, Fusion Science and Technology, 43(2003) 153-160. [2] K. Vulliez, G. Bosia, G. Agarici, B. Beaumont, S. Bremond, P. Mollard, Tore Supra ICRH antenna prototype for next step devices, Fusion Engineering and Design 66-68 (2003) 531-535.


Corresponding Author:

K. Vulliez

Association Euratom-CEA, CEA/DSM/DRFC, CE Cadarache, F-13108 St Paul lez Durance (France)

- B - Plasma Heating and Current Drive.

P3T-B-372 TORE SUPRA ITER-LIKE ANTENNA CHARACTERIZATION BY FEM ANALYSIS

Testoni Pietro, Giuseppe Bosia (1) Piergiorgio Sonato (2)

(1) CEA-DRFC/SCCP/GSAC Cadarache 13108 Saint Paul lez Durance (France) (2) Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)

The Ion Cyclotron (IC) technique for plasma heating and current drive is widely applied to the TORE SUPRA experiment to develop an efficient, reliable and stationary system for the next step application like ITER. A new four-elements (2poloidalx2 toroidal) IC array designed to launch up to 4 MW in the plasma in the frequency range 40-60 MHz and in pulses up to for 30 s long has been installed in TORE SUPRA vacuum vessel at the beginning of 2004. The array features with the same electric scheme adopted in the ITER reference design, for which a significant tolerance to resistive load variations (such as those induced by ELMs is predicted1). This paper describes the high-frequency electromagnetic (EM) analysis by 3D Finite Elements Modeling (FEM) of the array and of its associated tuning system, consisting of a set of variable capacitive reactances, connected in series with each element and paired poloidally in parallel. The finite element analysis is based on a full-wave formulation of Maxwell's equations in terms of the time-harmonic electric field E implemented in the ANSYS commercial code. The array/tuning system is first decomposed sections suitable to establish boundary conditions and each section is accurately modeled to deduce RF electric fields and currents distribution, so as to assess its voltage standoff capability and ohmic losses. Global (S) parameters are then computed as function of frequency (10 to 80 MHz) for each component in standard matching conditions. The array is finally re-synthesized by a computer code, which uses the S-parameter description to assess field and current distributions at the appropriate tuning conditions. This method is a new approach to the design of high frequency systems, when its physical dimensions are a non negligible fraction of the wavelength, and the effects of local modes cannot be accounted for by a pure transversal electromagnetic mode analysis. 1) G. Bosia “ High power density Ion Cyclotron antennas for next step applications” Fusion Science and Technology 43,153 (2003)


Corresponding Author:

Testoni Pietro

Electrical and Electronics Engineering Dept. - University of Cagliari

- B - Plasma Heating and Current Drive.

P3T-B-382 MAINTENANCE SCHEMES FOR THE ITER NEUTRAL BEAM INJECTOR TEST FACILITY

Zaccaria Pierluigi, A. Antipenkov (3), V. Antoni (1), A. Coniglio (1), S. Dal Bello (1), C. Day (3), M. Dremel (3), R. Hemsworth (2), T. Jones (6), A. Mack (3), D. Marcuzzi (1), A. Masiello (1), M. Pillon (4), S. Sandri (4), E. Speth (5), A. Tanga (5), PL. Mondino (7)

(1) CONSORZIO RFX, Padova, Italy (2) CEA, Cadarache, France (3) FZK, Karlsruhe, Germany (4) ENEA, Frascati, Italy (5) IPP, Garching, Germany (6) UKAEA, Oxford, United Kingdom (7) EFDA CSU, Garching, Germany

The aim of the ITER Neutral Beam Injector (NBI) Test Facility is to build and test the first 16 MW NBI for ITER and to demonstrate its reliability at the maximum operation parameters foreseen for ITER: power delivered to the plasma 16 MW, beam energy 1 MeV, D- ion current 40A, pulse length 3600 s. ENEA-RFX (I), CEA (F), FZK (D), IPP (D) and UKAEA (UK), are involved in an EFDA contract for the ITER NBI Test Facility design. On the basis of the present experience on existing test facilities and of the preliminary experimental program to be carried out on the ITER NBI Test Facility, several interventions for maintenance and modifications are foreseen in order to optimize the beam generation and steering. The maintenance scheme is therefore very important for the Test Facility design in order to maximize the time devoted to the test programme. The paper describes consistently the many interrelated aspects that have been considered during the design phase, such as: the interfaces with auxiliary systems, the need of special handling tools, equipments and cranes, the diagnostic and monitoring systems and remote handling capabilities. Further design requirements derived from the need of testing in advance the remote handling operations and tools foreseen for the ITER NBI. Lifting from the top and/or running from front and rear accesses were considered for the assembly/disassembly of the in-vessel components. Side and top access were designed, together with equipments and fixtures that facilitate personnel access and operations in the most critical zones. Self centering alignment systems were foreseen to speed up all the assembly and disassembly operations. The paper describes the hydraulic, electrical, gas and mechanical connections of all the in-vessel components designed to minimize the need of personnel access into the vessel. Optical lines of sight are located on the beam line vessel to get optimal diagnostic and monitoring information during the operations. CCD and IR cameras will look at the areas undergoing the most intense heat fluxes: leading edges of the neutralizer, entrance/exit of the residual ion dump, V-shaped panels of the calorimeter. Finally the paper presents a design of cryogenic panels compatible with the abovementioned maintenance and monitoring requirements.


Corresponding Author:

Zaccaria Pierluigi

Consorzio RFX - Corso Stati Uniti,4 - 35127 Padova, Italy

- B - Plasma Heating and Current Drive.

P3T-B-385 NEUTRAL BEAM INJECTION OPTIMIZATION AT TJ-II

Fuentes Candida, M. Liniers (1), G. Wolfers (1), J. Alonso (1), G. Marcon (1), R. Carrasco (1), J. Guasp (1), M. Acedo (1), E. Sánchez (1), M. Medrano (1), A. García (1), J. Doncel (1), C. Alejaldre (1), C.C. Tsai (2), G. Barber (2), D. Sparks (2)

(1) Laboratorio Nacional de Fusión, Asociación EURATOM-CIEMAT, Av. Complutense 22, 28040 Madrid, Spain (2) Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6169, USA

The TJ-II stellarator is beginning experiments with neutral beam injection (NBI) after an experimental phase with ECRH only. The first of two tangential injectors is now fully operative, the ion source has been conditioned up to 30 kV accel voltage, 50 Amps extraction current. The H0 beams with power in excess of 300 kW during 200 msec, encounter a target ECH plasma of average line density 1.0 1019 m-3 and 1.5 keV electron temperature. TJ-II is the first heliac experiment that makes use of NBI. The task is particularly challenging in this machine because of the extremely wide magnetic axis excursion (15 % of the major radius) and the relatively small size of the device. For this reason, optimisation of the injected beam power must be carefully performed. Beam transmission and thermal loads in the duct, first toroidal field coil, and Circular Coil groove inside TJ-II have been shown to depend critically on beam orientation. 3D beam simulation studies give 50 % variations of maximum power density in the duct for a horizontal beam deviation of 0.5º. The thermal load due to beam impact on the first toroidal field coil can reach 1100 W/cm2 for a beam tilted 0.5º in the opposite direction. Graphite protection plates have been installed at several locations inside TJ-II, and an infrared camera surveys the hot spots along the beam from a window located on the beam duct. Beam alignment is monitored by means of two sets of symmetrically located thermocouples. The power on the V-calorimeter and the duct diaphragm is measured as a function of beam orientation. The reionization of the neutral beam in the beam box and duct may account for a considerable amount of power loss. Reionization depends strongly on the residual gas pressure in the beam box, and therefore, on the gas inventory of the discharge. Computer simulation studies show that in order to maintain reionization losses below 10%, the residual gas pressure must be kept below 10-4 mbar during the beam pulse. Our efforts have been aimed to optimize gas use in the ion source and neutralizer. Fast ion gauges have been installed on the ion source and beam box that allow us to characterize gas flow and monitor the pressure in the injector during the pulse. The Halfa signal from a monitor located in the beam duct is related to the reionization losses. Calorimetric measurements of beam power and neutralization fraction are compared with Halfa measurements to determine the optimum gas injection scenario.


Corresponding Author:

Fuentes Candida

Laboratorio Nacional de Fusión, Asociación EURATOM-CIEMAT , Av. Complutense 22, 28040 Madrid, Spain

- B - Plasma Heating and Current Drive.

P3T-B-387 STATUS OF THE 140 GHZ / 10 MW CW TRANSMISSION SYSTEM FOR ECRH ON THE STELLARATOR W7-X

Kasparek, Walter, H.Braune(2), G.Dammertz(3), V.Erckmann(2), G.Gantenbein(1), F.Hollmann(2), M.Grünert(1), H.Kumric(1), L.Jonitz(2), H.P.Laqua(2), W.Leonhardt(3), G.Michel(2), F.Noke(2), B.Plaum(1), M.Schmid(3), T.Schulz(2), K.Schwörer(1), M.Thumm(3), M.Weissgerber(2)

(1) IPF, Universität Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart, Germany. (2) MPI für Plasmaphysik, EURATOM-Association, Greifswald, Germany. (3) FZ Karlsruhe, IHM, Association Euratom-FZK, Karlsruhe, Germany.

The stellarator W7-X which is currently under construction at IPP-Greifswald, Germany, will be equipped with a 10 MW ECRH system, working at 140 GHz in cw regime. The microwave power will be generated by 10 gyrotrons delivering 1 MW each and will be transmitted from the gyrotron hall to the W7-X stellarator ports via a fully optical system. The transmission system consists of 10 short single-beam mirror sections including matching optics and polarizers for each gyrotron, and two multi-beam mirror sections (appr. 44 m) which transmit 5 beams each to the torus hall. Near to the stellarator ports, the beams are separated again and launched by individual antennas to the plasma. The launchers allow for arbitrary toroidal (EC-current drive) and poloidal (on/off axis heating) launch angle of each beam. All mirrors (more than 160) are water-cooled and can be adjusted remotely. The status of the construction of the transmission lines and the design of the launchers is reported. Low-power tests of a prototype system at IPF Stuttgart are reviewed, showing high transmission performance (efficiency 90 %, mode purity 98 %). The first gyrotron is operating at IPP Greifswald, and high-power long-pulse tests have started. Measurements on transmission performance, behaviour of the water-cooled mirrors under thermal and microwave loads as well as alignment issues, characteristics of directional couplers, calorimetric loads and other diagnostics are discussed. The system is presently being prepared for high-power tests of a mock-up for the remote steering antenna as planned for ECRH/ECCD on ITER. First results of these experiments are presented.


Corresponding Author:

Kasparek, Walter

Institut fuer Plasmaforschung, Universitaet Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart

- B - Plasma Heating and Current Drive.

P3T-B-392 THE TEST OF A PAM LAUNCHER ON FTU: THE FIRST STEP TOWARD THE LHCD LAUNCHER FOR ITER

Mirizzi Francesco, Maria Laura Apicella(1), Philippe Bibet(2), Giuseppe Calabrò(1), Luigi Panaccione(1), Vincenzo Pericoli Ridolfini(1), Salvatore Podda(1), Angelo Antonio Tuccillo(1)

(1)Associazione EURATOM ENEA sulla Fusione, C.R. Frascati, via Enrico Fermi 45, 00044 Frascati, (Rome) Italy (2)Association CEA-EURATOM sur la Fusion, Centre d'Etude de Cadarache, France

Selected also for oral presentation O3A-B-392

A scaled prototype of the Passive Active Multijunction (PAM) launcher actually proposed for the LHCD system of ITER, has been realised and successfully tested on FTU in the frame of a collaboration between ENEA Frascati and CEA Cadarache. A power density of about 80 MW/m2 at the launcher mouth (corresponding to 50 MW/m2 in ITER at 5 GHz) has been routinely achieved with a power reflection coefficient r ? 2% at the launcher input. A very good coupling has been obtained also with plasma density, in front of the launcher, close to the cut-off value. Direct comparisons with the performances of a conventional grill in a different FTU port, thus in the same operative conditions, are available. The PAM is characterised by thick vertical walls between adjacent columns of active (transmitting) waveguides that assure a good mechanical stiffness to the structure, while ducts drilled in these walls, allow an effective water cooling. These thermo-mechanical characteristics make the PAM launcher very attractive for the use in the harsh plasma environment of ITER. The periodicity of a conventional multijunction launcher is restored by interposing columns of passive (reflecting) waveguides between the active ones at the mouth of the PAM. The directivity of the launcher is improved at low plasma density, where the cross coupling between active and passive waveguides is increased by strong RF power reflection conditions. This allows safe but efficient operations with the launcher positioned far from the plasma scrape-off layer where also the thermal loads are smaller. A full-scale test of the technological aspects of the PAM is now in preparation on Tore Supra. The proposed LHCD launcher for ITER will couple to the plasma a total power of 20 MW at 5 GHz. The launcher is composed of four independent units, each one including 12 PAM modules arranged in 4 poloidal rows and 3 toroidal columns. Each module is made of three rows with 8 active waveguides and 8 passive ones per row. The intrinsic phase shift between adjacent columns of active waveguides in the same module is set to 270 to have an N|| (peak) = 2 and an N|| = 1.9÷2.1 by changing the feeding phase between modules on the same row in the range ±90 . The maximum RF power density at the mouth is limited to 33 MW/m2, a value that has been largely demonstrated as safe during the test on FTU. The paper reports the main results of this test and their fall-out on the main features of the proposed launcher for ITER.


Corresponding Author:

Mirizzi Francesco

Associazione EURATOM-ENEA sulla Fusione, CR Frascati Via Enrico Fermi 45, 00044 Frascati (Rome), Italy

- B - Plasma Heating and Current Drive.

P3T-B-410 MATERIAL PROCESSING AND PROTOTYPE FABRICATION OF HEAT TRANSFER ELEMENTS FOR SST-1 NBI SYSTEM.

C Rotti, P K Jaykumar, K Balasubramanian, A K Chakraborty, S K Mattoo and NBI team

Institute for Plasma Research, Bhat, Gandhinagar-382 428, Gujarat, India Non-Ferrous Materials Technology Development Centre, Kanchan bagh, Hyderabad-500 058, Andhra Pradesh, India

Heat Transfer Elements (HTEs) based upon Cu-Cr-Zr alloy are used for thermal management of the beamline of the neutral beam injector for SST-1. It requires a thermo mechanical treatment during its fabrication which consists of EB welding and milling. Following the recognized material production process of Cu-Cr-Zr alloy with solution heat treatment, quenching and subsequent aging for ~ 4 hrs at 470 C has yielded mechanical properties of the material which are in the low end of the published database. In this paper we show that introduction of a significant percentage of cold work on the alloy yields remarkable enhancement of mechanical properties. A cold work of 60 – 90% on solution heat treated alloy (980 C for 20 minutes) of Cu-Cr (0.8%)-Zr(0.08%) leads to optimum properties (UTS >400 MPa, YS> 300 MPa and % elongation~22%) of the alloy suitable for fabrication of HTE’s. Suitability for fabrication was benchmarked by detailed characterization of EB weld joints for joint strength, micro hardness across the joint and metallographic properties and subsequently a full scale prototype of HTE. These results shall be discussed in this paper, with a recommendation a large database may be built for this route of material production.


Corresponding Author:

C Rotti

NBI Group, Institute for Plasma Research, Bhat, Gandhinagar-382 428, Gujarat, India

- B - Plasma Heating and Current Drive.

P3T-B-412 AN ALTERNATIVE SCHEME FOR THE ITER NBI POWER SUPPLY SYSTEM

Toigo Vanni, A. De Lorenzi, E. Gaio, F. Milani, L. Zanotto

Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova, Italy

This paper describes an alternative scheme developed for the ITER Neutral Beam Injector (NBI) Power Supply System. The main modification proposed regards the Ion Source Power Supply (ISPS), which presents quite high current levels and very low voltages. In the Reference Scheme, this power supply is divided in two sections: in the first, the required power is obtained from ground referenced power regulators and then raised to the high voltage level (-1 MV) through eight individual insulating transformers. Then the power is transmitted via an SF6 insulated HV Transmission Line to a large tank, named High Voltage Deck (HVD), insulated in high pressure SF6 as well. The HVD contains the final step-down transformers and diode rectifiers, which are connected to the Ion Source by means of a second Transmission Line. The inspection of the devices placed inside the HVD is not an easy task, due to the SF6 environment and the closeness to the neutronic area. For this reason, and to allow easier tuning and setting up of the whole system, maintenance, trouble shooting, fault inspections and implementation of further improvements, the possibility to eliminate the HVD and thus guarantee full accessibility to all the ISPS devices has been analyzed. As a result, in this alternative scheme all the ISPS components are installed inside an air insulated Faraday Cage (–1MV to ground), the HVD is removed and the power is transmitted to the ion source via a unique SF6 insulated HV Transmission Line. Only one main insulating transformer is required, placed upstream the whole system. An interesting aspect of this solution is the possibility to operate for tuning the Ion Source at lower voltage and also without the acceleration power supply, with a very easy connection between the Faraday Cage and the Ion Source; also voltage tests and system conditioning will be greatly simplified, using a test generator directly connected to the Faraday Cage. The main drawback of this solution is the fact that the high current conductors are present not only in the last part of the Transmission Line, but in the whole HV line. The paper will describe in detail the alternative power supply scheme, will discuss advantages and disadvantages with respect to the Reference Scheme, and will present the motivations behind the design choices and their implications. In particular, the impact on the HV Transmission Line structure and the general power supply system layout will be presented and discussed.

WITHDRAWN


Corresponding Author:

Toigo Vanni

Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova, Italy

- B - Plasma Heating and Current Drive.

P3T-B-439 NEUTRONICS ANALYSIS OF THE ECW LAUNCHING SYSTEM IN THE ITER UPPER PORT

A. Serikov, U. Fischer(1), Y. Chen(1), K. Lang(1), R. Heidinger(1), Y. Luo(2), E. Stratmanns(1), H. Tsige-Tamirat(1)

(1)Association FZK-Euratom, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, P.O. Box 3640, D-76021 Karlsruhe, Germany. (2)Computer and Information College, Hefei University of Technology, Hefei, Anhui 230009, PR China

The design of an Electron Cyclotron Wave (ECW) launching system for the ITER port is currently under development by a working group from various Euratom associations. The development work includes the ECW launcher with the waveguides, the main structural components such as the port plug, shielding and frame, and the torus window serving as vacuum closure in the waveguides. The major neutronics tasks are (i) to assess the neutron streaming in the waveguide channels for proofing the design limit for the radiation load to the CVD diamond window can be met and (ii) to assess and optimize the shielding of the launching system to ensure the radiation loads to adjacent components such as the vacuum vessel and the super-conducting magnet coils are tolerable. In addition, it must be assured that the radiation dose levels during shut down periods are tolerable to allow maintenance personnel access to the area inside the cryostat surrounding the ECW launcher port. This paper presents results of neutronics analyses conducted for the ECW launching system in the ITER upper port based on the launcher design of FOM Rijnhuizen with a twisted arrangement of 8 straight waveguides. A dedicated two-step approach is used for the neutron streaming calculations with the Monte Carlo code MCNP in the ITER 3D geometry. In the first step, a surface source is calculated at the region of the ECW launcher front using the standard ITER plasma volume source. In the second step, the surface source is used for calculating the neutron flux profiles along the waveguide channels by using point detector estimators. Shielding calculations are performed with MCNP using the importance sampling technique to assess the required dimensions of the shield material around the waveguides. The radiation dose levels during shut-down periods are calculated on the basis of the rigorous 2-step (R2S) computational scheme for MCNP based shut-down dose rate calculations. With all the calculations, use is made of a standard MCNP model of ITER (20 torus sector) to which the ECW launching system was integrated. The MCNP model of the launcher was generated from a suitably modified CATIA model by conversion into the geometry representation of the MCNP code using a newly developed interface programme.


Corresponding Author:

A. Serikov

Association FZK-Euratom, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, P.O. Box 3640, D-76021 Karlsruhe, Germany

- B - Plasma Heating and Current Drive.

P3T-B-455 THE ITER-LIKE ICRF ANTENNA FOR JET

Frederic Durodie, Ph. Chappuis(1), J.Fanthome(2), R.H.Goulding(3), J.Hosea(4), P. U. Lamalle(5), A.Lorenz(6), M.Nightingale(2), L.Semeraro(7), F. Wesner(8)

(1)CEA, (2)UKAEA, (3)ORNL, (4)PPPL, (5)LPP-ERM/KMS, (6)CSU-JET, (7)ENEA, (8)IPP-MPG

The aim of the ITER-like ICRF Antenna for JET[1] is to demonstrate novel antenna design principles in conditions as relevant as possible to ITER in order to validate them for an ICRF heating system for ITER. The power density for a given maximum voltage in the circuit is maximized using poloidally short straps and the resilience to fast varying RF loads by matching pairs of straps by a so-called conjugate-T. The main challenges during the design phase have been (i) to make the launcher - and in particular the in-vessel matching capacitors - resilient against disruption-induced mechanical loads, and (ii) the integration of in-vessel actuators for the matching capacitors with an overall accuracy better than 0.1mm. The antenna strap feeders were optimized to reduce as much as possible the electrical field in the regions of highest voltage near the matching capacitors. The tests on the High Power Prototype (ORNL and PPPL) have confirmed the voltage standoff as well as the overall RF design but have uncovered a number of issues, such as the thermal stress in the antenna straps and antenna box sidewalls. The results have allowed correcting the design in time for the start of the manufacturing phases for the different components. The design phase is completed and the procurement phase of 14 packages comprising the whole of the required hardware is now under way. Given the delivery times currently expected for the main antenna components, the launcher is expected to be installed on the JET torus by early 2005, following its assembly and testing on the testbed in autumn 2004. The experimental campaigns of 2005 are proposed to take into account the challenging task of matching the 4 feed lines of the whole antenna array simultaneously, concentrating first on delivering ICRH power for a limited number of frequencies in RF-optimized plasma conditions, and gradually broadening the operational domain towards achieving a power density of 8 MW/m2 in ITER-relevant plasma conditions and scenarios. As part of this effort, feeding two A2 antenna arrays in a load tolerant way from 3dB couplers, similarly to ASDEX-U, will further increase the total ICRF power available to JET. The paper will report the main characteristics of the final design as well as the main challenges encountered during the manufacturing phase of the key components. 1. Durodié, F., et al., in Radio Frequency Power in Plasmas, AIP Conference Proc., 595, NY: Melville, 2001, pp. 122-125.


Corresponding Author:

Frederic Durodie

LPP-ERM/KMS - Avenue de la Renaissance 30, B-1000 Brussels

- B - Plasma Heating and Current Drive.

P3T-B-460 EFFECTS OF MUTUAL COUPLING ON ICRF LOAD-TOLERANT ANTENNAS

SWAIN, D.(1), Goulding, R.(1) Lin, Y.(2) Parisot, A.(2) Wukitch, S.(2)

1 - Oak Ridge National Laboratory 2 - Massachusetts Institute of Technology

The ability to efficiently couple radio-frequency power in the ion cyclotron range of frequencies (ICRF) can be significantly improved, in principle, by the use of so-called load-tolerant antenna designs. The central concept utilized is the “conjugate-tee” arrangement, in which two nearly identical current straps are symmetrically fed through a resonant loop, producing a resistive impedance at the feed point which varies slowly as the plasma resistance increases. Through the use of this concept, the amount of power reflected back to the transmitters during plasma load transients can be substantially reduced. This improves the ability of the ICRF system to continue delivering rf power in the presence of ELMs and L- to H-mode transitions. Tests of load tolerant antenna feed configurations have been undertaken on JET, Tore Supra, and Alcator C-Mod, and a new load-tolerant antenna is presently being built for installation on JET. Initial work on load-tolerant matching was done using relatively simple models that neglected mutual coupling between neighboring straps, but most present and proposed future antenna designs have multiple straps, usually with significant mutual inductive coupling. The load-tolerant properties of the uncoupled circuit can be significantly modified, and in some cases destroyed, by the inclusion of inter-loop coupling. Initial results from a load-tolerant matching experiment on C-Mod demonstrate that the load-tolerant characteristic can be destroyed by coupling between antenna elements. This paper will examine the changes that mutual coupling cause on several load-tolerant antenna concepts, and the possible application of these concepts to an antenna for ITER will be examined.


Corresponding Author:

SWAIN, D.(1)

PO Box 2008, Oak Ridge National Lab, Oak Ridge, TN 37831-6169

- B - Plasma Heating and Current Drive.

P3T-B-479 140-GHZ HIGH-POWER GYROTRON DEVELOPMENT FOR THE STELLARATOR W7-X

Dammertz, Guenter, D. Bariou(5), P. Brand(4), H. Braune(3), V. Erckmann(3), G. Gantenbein(4), E. Giguet (5), W. Kasparek(4), H.P. Laqua(3), C. Lievin(5), W. Leonhardt(1), G. Michel(3), G. Mueller(4), G. Neffe(1), B. Piosczyk(1), M. Schmid(1), M. Thumm(1,2)

(1)FZK Karlsruhe,IHM,EURATOM-FZK,Postfach 3640,D-76021 Karlsruhe,Germany (2)Universitaet Karlsruhe,IHE,Karlsruhe, Germany (3)MPI fuer Plasmaphysik,Greifswald,Germany (4)Universitaet Stuttgart,IPF,Stuttgart Germany (5)Thales ED,Vélizy,France

Electron cyclotron resonance heating (ECRH) has proven to be one of the most attractive heating schemes for stellarators, as it provides net current free plasma start up and heating. Extensive measurements on stellarators at IPP Garching yield a solid physical and technological basis for ECRH systems. Therefore, ECRH was decided to be the main heating method for the Wendelstein 7-X stellarator (W7-X) now under construction at IPP Greifswald/Germany. A 10 MW ECRH system with continuous wave possibilities operating at 140 GHz will be built up to meet the scientific goals of the stellarator with inherent steady-state capability at reactor relevant plasma parameters. Two prototype gyrotrons with an output power of 1 MW were developed in collaboration between European research laboratories and European industry (Thales Electron Devices, France). The gyrotrons are equipped with a single-stage depressed collector (SDC), an optimized quasi-optical mode converter and a CVD-diamond window. The prototypes have been successfully tested at FZK. With the second one an output power of 0.89 MW with an efficiency of 42% at a pulse duration of 3 minutes and an output power of 0.54 MW for about 15 minutes has been obtained. The first prototype has been installed at IPP Greifswald and has been tested there successfully up to a pulse-length of 10 s. The development has been finished; the series gyrotrons and the superconducting magnet systems have been ordered. In a parallel development at CPI (Communications and Power Industries, Palo Alto, California) another tube had been tested at high power (920kW, 5 ms) and delivered a power of about 500 kW with an efficiency of 35% (SDC) in pulse-lengths of 700 s.


Corresponding Author:

Dammertz, Guenter

Forschungszentrum Karlsruhe, IHM, Postfach 3640, D-76021 Karlsruhe, Germany

- B - Plasma Heating and Current Drive.

P3T-B-501 DEVELOPMENT OF THE 140 GHZ GYROTRON AND ITS SUBSYSTEMS FOR ECH AND ECCD IN TEXTOR

J. Scholten(1), J.W. Oosterbeek(2) A.F. van der Grift(1) J.A. Hoekzema (2) O.G. Kruijt (1) A.J. Poelman (1) P.R. Prins (1) E. Westerhof (1) C.J. Tito (1) W.A. Bongers (1) M. R. de Baar (1)

Partners in TEC: (1) FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, The Netherlands, www.rijn.nl (2) Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, EURATOM Association, D-52425, Jülich, Germany

A 1 MW, 140 GHz gyrotron has been mounted on TEXTOR. First results on heating, current drive and manipulation of islands have been obtained during the 2003 campaigns. Arcs in the transmission line, and the restricted pulse length due to the limited power handling of the window and the launching mirror, hampered the gyrotron operation. Measurements of power absorption in critical components in the quasi-optical transmission line, in particular the quartz tokamak window and launching mirror, are presented. Experience shows that, when the transmitted power through the plasma is high, the protection tiles for the DED on the high field side of the tokamak can be damaged. Sniffer probes are mounted to monitor the transmitted power levels. To avoid gyrotron operation in a wrong mode, a protection is set-up, based on a notch filter for detecting power at frequencies outside a narrow band around 140 GHz. The following measures are being taken to solve the various problems: A CVD diamond tokamak window will be mounted to limit absorption and reflection from the tokamak window. Full shielding of the transmission line has been mounted. To avoid arcs, cooling water in Teflon hoses absorbs stray radiation at critical positions in the transmission line. A new launcher, which can couple 1 MW of EC-waves for 10 s will be installed. It features fast and accurate poloidal and toroidal steering, for fast tracking of magnetic islands.


Corresponding Author:

J. Scholten(1)

(1) FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, P.O.Box 1207, 3430 BE Nieuwegein, The Netherlands, www.rijn.nl

- B - Plasma Heating and Current Drive.

P3T-B-507 DESIGN OF CRYOSORPTION PUMPS FOR TESTBEDS OF ITER RELEVANT NEUTRAL BEAM INJECTORS

Dremel Matthias, A. Mack(1), C. Day(1), H.Jensen(1), E. Speth(2), H.D.Falter(2), R.Riedl(2), JJ. Cordier(3), B.Gravil(3)

(1) FZK, Karlsruhe, Germany (2) IPP, Garching, Germany (3) CEA, Cadarache, France

Special cryosorption pumps based on the adsorption with activated charcoal, coated onto stainless steel panels are being developed at Forschungszentrum Karlsruhe in Germany. A 1:2 scaled ITER torus cryopump has been manufactured and is under testing in the TIMO (Test Facility for ITER Model Pump) testbed. The results of the experimental data from TIMO are used to establish and develop the design of huge cryosorption pumps, which will be installed in the testbeds of Neutral Beam Injectors. This paper presents the results of the design investigations and the manufacturing of two cryosorption pumps for the Neutral Beam Testbed L6 at Max Planck Institute for Plasma Physics (IPP) and the cryopump design investigations for an EFDA contract for the ITER NB Test Facility design. The cryopumps for IPP are foreseen to pump a hydrogen-flow of 3Pam3/s from the beam line with a pumping speed of 350m3/s per pump. The pressure conditions must be maintained over 4 hours pumping without regeneration of the cryopanels. For cooling liquid helium at saturation pressure is used and therefore a two-phase flow in the cryopanel system must be controlled. The hydrodynamic calculations about the cooling with a forced flow of liquid helium are presented and the calculations of the heatloads as well as the required mass flows for the cooling are contents of the paper. For the ITER Neutral Beam Test Facility , design calculations to assess heatloads, pressure drops and pumping parameters of the cryopumps based on experimental experience will be briefly described. We assess and discuss herein the different cryogenic needs during the given operation scenarios of the Neutral Beam Injector.


Corresponding Author:

Dremel Matthias

P.O. Box 3640 76021 Karlsruhe, Germany

- B - Plasma Heating and Current Drive.

P3T-B-512 STATUS OF THE NEW ECRH SYSTEM FOR ASDEX UPGRADE

Leuterer Fritz, G.Gruenwald,F.Monaco,M.Muenich,H.Schuetz,F.Ryter,D.Wagner,H.Zohm,T.Franke W.Kasparek1),G.Gantenbein1),H.Hailer1), G.Dammertz2),H.Heidinger2),K.Koppenburg2),M.Thumm2),X.Yang2) G.Denisov3),V.Nichporenko3),V.Miasnikov3),V.Zapevalov3)

Institut fuer Plasmaphysik,85741 Garching 1) Institut fuer Plasmaforschung, Universitaet Stuttgart, 70569 Stuttgart 2) Forschungszentrum Karlsruhe, 76021 Karlsruhe 3) Institute of Applied Physics, 603600 Nizhny Novgorod, Russia

A new ECRH system with up to 4 MW/10 sec is in construction at ASDEX Upgrade.Each of the 4 depressed collector gyrotrons can operate at different frequencies. The first one, expected in summer 2004, can work at 105 GHz (mode TE17.6) and at 140 GHz (mode TE22.8). It has asingle disc diamond window which is resonant at both frequencies. A second gyrotron, expected later in 2004, con operate at many frequencies within the same interval. It has atunable double disc window. In our installation we will use 4 discrete frequencies. Since the output beam leaves the gyrotron in slightly different directions for each frequency, the matching optics comprises a switchable pair of phase correcting mirrors for each frequency to provide a proper Gaussian beam. The polariser mirrors are designed for the centre frequency of 122.5 GHz, but allows to set the required polarisation for ECRH or ECCD in the whole frquency band. Thr waveguide transmission line (70 m) has a broadband corrugation. At the torus there is another vacuum window, either single disc or tunable double disc, which allows transmission at any polarisation. The launching mirrors, made of graphite with a copper coating, are fast steerable in poloidal direction. A beam scan of 10 degrees in 100 msec has been achieved. This is intended to realise a feedback controlled power deposition, in particular for experiments on suppression of neoclassical tearing modes, where the deposition should remain within the island even when it moves due to the Shafranov shift. The toroidal beam steering is slow, but can be changed between pulses. The first two of these launcher units have been installed into the torus.


Corresponding Author:

Leuterer Fritz

Max Planck Institut fuer Plasmaphysik, Postfach 1322, D-85741 Garching, Germany

- B - Plasma Heating and Current Drive.

P3T-B-514 IMPROVED 118 GHZ GYROTRON FOR ECRH EXPERIMENTS ON TORE SUPRA

LIEVIN Christophe, C. Darbos (2), S. Alberti (3), A. Arnold (4), D. Bariou (1), F. Bouquey (2), J. Clary (2),J.P. Hogge (3), M. Lennholm (2), F. Legrand (1), R.Magne (2), M. Thumm (4)

(1)Thales Electron Devices, 2 rue Latécoère, 78141 Vélizy-Villacoublay, France (2) Association Euratom-CEA, CEA/DSM/DRFC, CEA-Cadarache, (3) Association Euratom-CRPP,1015 Lausanne, Switzerland (4) Association Euratom-FZK, Karlsruhe, Germany

A Gyrotron operating at the frequency of 118 GHz and producing 500 kW output power has been developed thanks to a collaboration between TED (Thales Electron Devices), Association Euratom-Confédération Suisse, Association Euratom-FZK and Association Euratom-CEA, for ECRH (Electron Cyclotron Resonance Heating) and current drive experiments held on the tokamak Tore Supra. Tests performed on the first series gyrotron in Cadarache [1] have shown limitations on pulse length (about 110 s), which have been explained by the overheating of internal components of the tube, mainly due to its inadequate cooling. Moreover, the geometry of the launcher seemed to be at the origin of spurious frequencies oscillating inside and at the output of the gyrotron then leading to possible degraded performances. To improve the gyrotron, new studies between the same partners have been performed and a new tube has been built by TED, mainly with a new cooling system and a different geometry for the launcher. The factory tests up to 500 kW 5s pulses have been completed in January 2004 and showed a significant improvement in the gyrotron conditioning, which was rather fast compared to previous tubes.The gyrotron is now being installed at Cadarache and will be tested in long pulse operation up to 600s. The presentation will describe the design modifications implemented on the new gyrotron and will show main experimental results obtained, with a special focus on long pulse tests. References [1] Very long pulse operation of the Tore Supra ECRH system, C. Darbos et al., SOFT 2002, Helsinki.


Corresponding Author:

LIEVIN Christophe

THALES ELECTRON DEVICES, 2 rue Latécoère, 78141 Vélizy-Villacoublay, France

- B - Plasma Heating and Current Drive.

P3T-B-542 MATCHING TO ELMY PLASMAS IN THE ICRF DOMAIN

Noterdaeme J-M, J-M Noterdaeme (2) B. Beaumont (3) Ph. Lamalle (4) F. Durodié (4) M. Nightingale (5) I. Monakhov (5) the ASDEX Upgrade Team (1) the JET-EFDA contributors and the Tore Supra Team (3)

(2) Gent University, EESA Department, Belgium (3) Association EURATOM-CEA, CEA-Cadarache, France. (4) Association EURATOM - Belgian State, LPP-ERM/KMS, TEC (5) EURATOM/UKAEA Fusion Association, Abingdon, U. K.

Selected also for oral presentation O3A-B-542

The RF generators in the ICRF domain are meant to operate into a constant, matched impedance. The coupling of the antennas to the plasma presents a load impedance typically much lower than the generator needs for optimal power transfer. This mismatch is overcome by a matching system that transforms the antenna impedance to that required by the generator. However, the antenna impedance is dependent on the plasma density in front of the antenna. This necessitates provision of a dynamic matching system, or passive load isolation, that maintains a matched impedance despite rapid antenna impedance variations. Different approaches can be taken with different timescales. The choice can have a substantial impact on the system efficiency and operational reliability. These different methods are presented and discussed. For slow variations, the mechanical change of length of a coaxial line is sufficient. For faster variations during a discharge, capacitors are used on Tore Supra and frequency scanning on JET. Ferrite systems, utilising the change in magnetic properties of a ferritic material with an applied magnetic field, are being developed. The most critical area is that of very fast variations during ELMs. ASDEX Upgrade has implemented load isolation using 3 dB couplers, which are completely passive and very reliable, maintain good current phasing, but lead to a power loss during an ELM. This has strongly improved the performance of the ASDEX system, with up to 90 % of the installed generator power being reliably transmitted to the antenna. JET has tested frequency variation with the addition of appropriate line components. Whereas it was shown to work in principle, the implementation turned out to be presently not practical. JET and Tore Supra have recently tested conjugate-T matching on plasma, the former using an external junction and the latter a new antenna again utilising internal capacitors. These have the advantage that there is no power reduction during an ELM, but the disadvantages that the phasing is load dependent and the matching may demand very precise setting of the components. On JET-EP, both conjugate matching and 3 dB couplers will be implemented. The paper also indicates where more work is still needed for the extrapolation to ITER.


Corresponding Author:

Noterdaeme J-M

Max-Planck Institute for Plasmaphysik, Boltzmannstraße 2, D-85748 Garching, Germany

- C - Plasma Engineering and Control.

P3C-C-62 OPTIMISED MODELLING OF THE TORE SUPRA TOKAMAK FOR PLASMA EQUILIBRIUM CALCULATIONS WITH THE PROTEUS CODE

HERTOUT Patrick, François Saint-Laurent (1) Fernanda Rimini (1) Thomas Pelletier (1)

(1) Euratom-CEA Association, CEA/DSM/DRFC, CEA-Cadarache, 13108 Saint Paul Lez Durance, France

Magnetic configuration calculations with plasma equilibrium finite element codes are based upon numerical solving of Grad-Shafranov and Maxwell equations, respectively inside and outside the plasma. These plasma equilibrium codes require a two dimensional modelling of the tokamak, assuming a global axial symmetry to compute the poloidal magnetic field everywhere in a meridian plane of the machine. This modelling must be carried out very carefully in the case of machines with iron core, where the return circuits violate the axial symmetry: for example the vertical arms are assumed to be equivalent to a vertical axis cylinder with the same total horizontal cross section, and modelled by a thin vertical rectangle in the meridian plane meshing. The iron magnetic permeability must be also accurately determined, and the plasma facing components must be carefully modelled, to provide consistent results in direct or reverse plasma equilibrium computations, respectively plasma boundary determination with a given current configuration in the poloidal field coils, and calculation of the current configuration for a given plasma boundary: in JET plasma equilibrium calculations with the PROTEUS code, this allows a determination of the X point position with a centimetre accuracy. For a better prediction of optimised pre-magnetization configurations and long discharges analysis with the PROTEUS code, the modelling of the Tore Supra tokamak iron circuit and CIEL plasma facing components has been refined: a new meshing has been built, respecting the limiter positions and the inner surfaces of the iron return arms. The magnetic permeability has been adjusted to minimize errors between the code results and magnetic measurements during special shots with significant horizontal or vertical field at different levels of iron saturation. With this new modelling, magnetic configurations with field lower than 1 mT over the whole vacuum vessel have been predicted by the PROTEUS code and successfully applied during experiments on the Tore Supra tokamak. Plasma slow derive during long pulses in preparation to the Gigajoule discharges has been also confirmed by PROTEUS, which proves to be a serious candidate code for ITER plasma scenario predictions.


Corresponding Author:

HERTOUT Patrick

Euratom-CEA Association, CEA/DSM/DRFC, CEA-Cadarache, 13108 Saint Paul Lez Durance, France

- C - Plasma Engineering and Control.

P3C-C-77 HIGH PERFORMANCE INTEGRATED PLASMA CONTROL IN DIII–D*

Humphreys, D.A., R.D. Deranian (1), J.R. Ferron (1), R.D. Johnson (1), R.R. Khayrutdinov (2), R.J. La Haye (1), J.A. Leuer (1), B.G. Penaflor (1), M.L. Walker (1), and A.S. Welander (1)

(1) General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (2) TRINITI Laboratory, Troitsk, Russia

The DIII-D mission to explore the advanced tokamak (AT) regime places significant demands on the DIII-D plasma control system (PCS) [1], including simultaneous and highly accurate regulation of plasma shape, stored energy, density, and divertor characteristics, as well as coordinated suppression of MHD instabilities. To satisfy the control demands of AT operation, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize. Present work describes selected new solutions which address key control problems in DIII-D, and the approach, benefits, and progress made in integrated plasma control at DIII-D. One element of DIII-D AT control which has been successfully addressed is the problem of high accuracy plasma boundary control. The problem is complicated in DIII-D by the need to produce good performance in a wide range of shapes and configurations, as well as by a uniquely constrained PF coil circuit and power supplies routinely operated near current limits. We describe the development and implementation of a complete predictive solution to this problem including multivariable controllers based on novel linear nonrigid, resistive plasma models, and nonlinear algorithms to avoid saturation and windup effects. Integrated plasma control was essential in the successful development of the DIII-D NTM control system, which has achieved full and sustained suppression of 3/2 and 2/1 NTM modes (separately) using the 3 MW ECCD/ECH gyrotron system to replace missing island bootstrap current. Validated models of island response to ECCD were used to design nonlinear controllers, which vary plasma position or toroidal field to achieve alignment of island and ECCD deposition location. We report on design and experimental use of novel algorithms using direct feedback on the q-profile, reconstructed with a realtime Grad-Shafranov calculation [2] including MSE measurements. [1] B.G. Penaflor, et al., Proc. of 4th IAEA Tech. Mtg. on Control and Data Acq., San Diego, 2003. [2] J.R. Ferron, et al., Nucl. Fusion 38, 1055 (1998). *Work was supported by the U.S. Department of Energy under DE-FC02-04ER54698.


Corresponding Author:

Humphreys, D.A.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- C - Plasma Engineering and Control.

P3C-C-79 PROGRESS TOWARDS ACHIEVING PROFILE CONTROL IN THE RECENTLY UPGRADED DIII-D PLASMA CONTROL SYSTEM*

Penaflor, B.G., J.R. Ferron, C.C. Makariou, B.D. Bray, D. Piglowski, and R.D. Johnson

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

This paper describes the improvements being made in the capabilities of the DIII-D plasma control system (PCS) towards achieving optimization of pressure and current profiles in advanced tokamak discharges. Key improvements have been increased processing power and the ability to include profile diagnostic data. The recently completed upgrade of the PCS to Linux based Intel computers connected with 2 Gigabit/s Myrinet networking technology has been successful in achieving its goals of increasing the overall performance and flexibility of the system. The new Intel computing system has increased processing power by a factor 30 over the older i860 based systems. The Myrinet fiber based network has opened the doors to the inclusion of data in real-time from DIII-D diagnostics situated in remote locations within the DIII-D research facility. The PCS now collects 32 channels of motional Stark effect data and uses these data for real-time computation of the safety factor (q) profile. Electron temperature and density profile data from the Thomson scattering diagnostic and electron temperature profile data from the electron cyclotron emission diagnostic are in the midst of being added. Addition of ion temperature and toroidal rotation profile data from the charge exchange recombination diagnostic is planned. Feedback control from the PCS of the electron temperature at a single off-axis point has been demonstrated using either electron cyclotron heating (ECH) or neutral beam power. This has been used to modify current profile evolution during plasma current ramp up. Specifics of the latest improvements to the DIII-D PCS are detailed. *Work was supported by the U.S. Department of Energy under DE-FC02-04ER54698.


Corresponding Author:

Penaflor, B.G.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- C - Plasma Engineering and Control.

P3C-C-103 REAL TIME CONTROL OF FULLY NON-INDUCTIVE OPERATION IN TORE SUPRA LEADING TO 1GJ PLASMA DISCHARGES

van Houtte Didier, G. Martin, A. Bécoulet, B. Saoutic and the Tore Supra team

Association EURATOM-CEA, CEA-DSM-DRFC, CEA Cadarache,13108 STYLE="PAUL-LEZ-DURANCE (France)

Selected also for oral presentation O1B-C-103

With the goal of addressing the critical issue of the long pulse steady-state operation of next fusion devices, the experimental programme of the Tore Supra super-conducting tokamak has been devoted in 2003 to study simultaneously heat removal capability and particle exhaust in a steady-state fully non-inductive current drive discharge. After a major upgrade of all the actively water-cooled in-vessel components and associated cooling loops, partly inductively driven discharges of more than 4 minutes were rapidly obtained at multi megawatts level in 2002. In 2003, a better understanding of several tokamak sub-system limitations, an improvment of the plasma position within a few millimetres range, and new real time cross controls between RF Power and various actuators built around a shared memory network, have allowed Tore Supra to access a powerful steady-state regime with an improved safety level. In addition to the usual real time controls, two primary feedback loops were involved: the plasma current is controlled by the lower hybrid power level and the primary flux consumption is adjusted to zero using the main PF power amplifier voltage. As result of these improvements, feedback controlled fully non-inductive plasma discharges have been sustained in a steady-state regime up to 6 minutes with a new world record of injected-extracted energy exceeding 1 GJ. On the safety aspects level, a real time control of the power deposition on plasma facing components is performed through IR monitoring of the surface temperature of the Toroidal Pumped Limiter and the supervision of water flows and temperatures of the cooling loops. In these discharges, the injected energy is removed steadily from the plasma facing components by high temperature pressurized water loops. The analysis of water calorimetry and surface temperature of the actively cooled in-vessel components indicates that the thermal equilibrium is obtained after a few seconds and shows a very satisfactory energy balance. With regard to the particle balance, half of the injected particles are recovered in the pumps and the other half is implanted into the wall without any indication of saturation, even after two successive 6-minute discharges.


Corresponding Author:

van Houtte Didier

Association EURATOM-CEA, CEA-DSM-DRFC, CEA Cadarache,13108 STYLE="PAUL-LEZ-DURANCE (France)

- C - Plasma Engineering and Control.

P3C-C-114 FEEDBACK CONTROL FOR PLASMA POSITION IN HL-2A TOKAMAK

Li Bo, Song Xianming Li Li Liu Li Wang Minghong Fan Mingjie Chen Liaoyuan Yan Qingwei

Southwestern Institute of Physics, P.O. Box 432,Chengdu,Sichuan,610041, P.R.China

This paper presents the horizontal plasma position feedback control system (FBCS) for HL-2A tokamak device. It describes the hardware configuration, the program, the control algorithm, and the experimental results of FBCS in details. It also introduces a new model of plasma resistance for computing the required voltage waveforms of poloidal field coils for preprogrammed plasma current. HL-2A is a tokamak device with closed divertors. It was put into operation at the end of 2002. But Divertor configuration discharges were achieved only after the successful development and operation of FBCS. From the engineering point of view, controlling horizontal plasma position is to control the vertical magnetic field produced by vertical magnetic field coil (VF), i.e. to control the power supply of VF. In FBCS, a simplified proportional-differential controller was adopted to control the power supply of VF. To guarantee the operational safety of the power supply, some efficient measures were taken into account. It is noticed that the industrial personal computer (IPC) used to acquire, calculate and control was programmed to be an intelligent controller. Its function was to receive commands and control parameters from experimental management computer (EMC) through ETHERNET. All the commands, parameters, and discharge waveforms were set or edited on EMC according to experimental requirements. The operational interface was carefully designed so that the operator only needs to drag and drop by using a mouse to get desired waveforms. The waveforms were calculated with a plasma resistance model, in which plasma resistance was determined by the following two factors: 1) plasma current stage, 2) plasma impurity and equilibrium status. For given plasma configuration, the ratios of plasma current to VF coil current and multiple field coils current are constant. With FBCS, the plasma was confined in the assigned area and the good repeatability of divertor configuration plasma discharges was obtained.


Corresponding Author:

Li Bo

Southwestern Institute of Physics, P.O. Box 432,Chengdu,Sichuan,610041, P.R.China

- C - Plasma Engineering and Control.

P3C-C-149 DIII-D INTEGRATED PLASMA CONTROL TOOLS APPLIED TO NEXT GENERATION TOKAMAKS*

Leuer, J.A, R.D.Deranian(1),J.R.Ferron(1),D.A.Humphreys(1),R.D.Johnson(1),B.G.Penaflor(1),M.L. Walker(1),A.S.Welander(1),D.Gates(2),J.Menard(2),D.Mueller(2),G.McArdle(3),B. Wan(4), M.Kwon(5),R.R.Khayrutdinov(6), A. Kavin(7)

(1)General Atomics,San Diego,CA 92186-5608 (2)PPPL, Princeton,NJ (3)EURATOM/UKAEA,Abingdon, UK (4)ASIPP,Anhui,China (5)KAIST,Daejon,Republic of Korea (6)TRINITI Lab.,Troitsk, Russia (7)ITER Naka, Japan

Current and next generation fusion experiments are increasingly expected to operate in advanced tokamak (AT) regimes and require highly integrated, complex plasma control. The DIII-D program is dedicated to the AT mission and has developed an extensive set of modeling, simulation, and design tools for real-time control development to enable integrated high performance regulation of plasma shape, internal profiles, fueling, pumping, current drive and heating[1]. The highly flexible DIII-D machine allows validation of this software suite over a wide range of plasma parameters and configurations. The system is tightly integrated with our state-of-the-art digital Plasma Control System (PCS)[2] enabling rapid development and testing of algorithms prior to device implementation. This paper provides an overview of this software suite and its application to next generation tokamaks. Modeling environment elements have been used to design controllers for devices that use, or plan to use, the DIII-D PCS, including NSTX, MAST, KSTAR and EAST. DIII-D integrated plasma control tools have been applied to analysis and control simulation of ITER-FEAT using a demonstration PCS. Results of applications to these devices will be presented. The software suite consists of detailed linear/nonlinear models of plasma and system components, simulators, control design tools, and PCS interface/testing modules. Both rigid and non-rigid linear equilibrium models of the plasma shape are used in device simulation and controller development. A nonlinear plasma model based on DINA[3] is used to simulate coupled plasma shape and profile evolution. System simulation is performed using a plant model including power supplies, PF coils, passive elements, plasma models, diagnostics and data filtering/conditioning. The simulator connects to the actual PCS hardware (or a software version of the PCS) to perform hardware-in-the-loop tokamak/PCS simulation. The integrated plasma control suite provides a comprehensive environment for development and testing of complex plasma control algorithms. Applications of the suite have identified power supply characteristics and gains required to satisfy machine design constraints. [1] D.A.Humphreys, in Proc. 20th IEEE/NPSS, San Diego, CA (2003)-[2] B.G. Penaflor, in Proc. 4th IAEA Tech Mtg on Ctrl Data Acq, San Diego, CA (2003)-[3] R.R. Khayrutdinov, J. Comp. Phys. 109 (1993) *Work supported by U.S. Dept of Energy under DE-FC02-04ER54698 & DE-AC02-76CH03073.


Corresponding Author:

Leuer, J.A

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- C - Plasma Engineering and Control.

P3C-C-155 CONFIGURATION AND PERTURBATION DEPENDENCE OF THE NEUTRAL POINT IN JET

Fabio Villone, V. Riccardo (1), F. Sartori (1), A. Cenedese (2), B. Alper (1), P. Beaumont (2) and contributors to the EFDA-JET Workprogramme

(1) EURATOM/UKAEA Fusion Assoc., Culham Science Centre, Abingdon OX14 3DB, UK (2) Consorzio RFX, Assoc. Euratom-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127, Padova, Italy

In the past, several neutral point analyses have been carried out on JT60-U [1-3] and ASDEX-U [4]. These works have pointed out that the direction of vertical plasma movement consequent to a radiative collapse depends on the initial vertical position of the plasma. Indeed, there exists a vertical position (called the Neutral Point, NP) that separates a region of upwards moving configurations from a region of downwards moving ones. If one could place the plasma at the neutral point, it will experience ideally a zero (practically very small) vertical motion. On JET, a number of dedicated experiments have been successfully carried out in the past [5], confirming the existence of a NP also for this device. Moreover, a modelling activity through the CREATE_L model [6] has also justified some unexpected experimental features. This paper presents some recent experimental results on JET aimed at extending the investigation of operation near the NP to a different plasma perturbation: ELMs (Edge Localised Modes) due to their significance in high performance plasma operation. The plasma configuration studied had to be changed with respect to [5], to allow an efficient plasma heating via NBI (Neutral Beam Injection), necessary to reach H-mode and hence have ELMs. The Vertical Stabilization System was switched off at the H_\alpha spike, triggering the subsequent vertical motion. A dependence of the actual excitation of the unstable vertical motion on the initial vertical position was indeed found, although no sign change (and hence no NP) was detected in the explored range. Nevertheless, the results obtained, together with a suitable comparison with simulations obtained with the CREATE_L code, provide extremely useful information about the correct characterization of ELMs in terms of simplified modelling in view of plasma control. [1] Y. Nakamura, R. Yoshino, Y. Neyatani, T. Tsunematsu, M. Azumi, N. Pomphrey, S.C. Jardin, Nucl. Fusion 36 (5) (1996) 643-656. [2] Y. Nakamura, R. Yoshino, N. Pomphrey, S.C. Jardin, Plasma Phys. Control. Fusion 38 (1996) 1791-1804. [3] R. Yoshino, Y. Nakamura, Y. Neyatani, Nucl. Fusion 36 (3) (1996) 295-307. [4] Y. Nakamura, G. Pautasso, O. Gruber, S.C. Jardin, Plasma Phys. Control. Fusion 44 (2002) 1471-1481 [5] F. Villone, V. Riccardo, F. Sartori, A. Cenedese, Fusion Eng. Des., Vol. 66-68 (2003) 709-714 [6] R. Albanese, F. Villone, Nucl. Fusion 38 (5) (1998) 723-738


Corresponding Author:

Fabio Villone

Ass. EURATOM/ENEA/CREATE, DAEIMI, Univ. di Cassino, Via Di Biasio 43, I-03043, Cassino (FR), Italy

- C - Plasma Engineering and Control.

P3C-C-157 DEVELOPMENT OF THE DINA-CH FULL DISCHARGE TOKAMAK SIMULATOR

LISTER Jonathan, V.N. Dokuka(1) R.R. Khayrutdinov(1) B.P. Duval J-Y. Favez V.E. Lukash(2) J-M. Moret H. Weisen

CRPP-EPFL, Association EURATOM-Confederation Suisse, 1015 Lausanne, Switzerland (1)TRINITI, Moscow Region, Russia (2)RRC Kurchatov Institute, Moscow, Russia

We have continued work on developing a well benchmarked full discharge simulator based on a Matlab-Simulink version of the widely known 1.5D DINA code, referred to as DINA-CH. DINA-CH had been previously used to simulate the effects of TCV Poloidal Field voltage signals on the equilibrium modifications, and to follow VDE’s in the highly structured vacuum field pattern of TCV. Work has progressed on simulating the effect of very localised Electron Cyclotron Heating and Current Drive in TCV. We have now reconciled the different versions of DINA developed for ITER, TCV and MAST into a single version which includes the possibility of defining plasma transport in external blocks. External heating and current drive can now be specified both in [R,Z] coordinates, developed for ECH and ECCD on TCV, but also in flux-surface coordinates more suitable for NBI, alpha-heating and radiation losses. We can now save the simulation data as MDS+ data structures, allowing more effective remote access to the results using the standard interfaces to MDS+ data. Work is currently underway to incorporate detailed up to date transport models and results should be presented. We are presently incorporating an equilibrium to diagnostics mapping toolbox which was originally developed for the TCV tokamak and which has been exhaustively tested and refined on the large variety of TCV plasma shapes. An example of real-time simulation of a representative diagnostic is presented for the case of ITER using this toolbox. The nominal ITER magnetic diagnostics are now also included in the simulations of ITER. We are integrating an estimate of the AC losses during operation of superconducting Poloidal Field coil magnets, previously developed and benchmarked against a full evaluation code as an ITER design task. The estimate of the deposited heat is available during the simulation, rather than estimated after the pulse. A full ITER simulation with all these recently implemented features will be presented to illustrate the advances described.


Corresponding Author:

LISTER Jonathan

CRPP-EPFL, 1015 Lausanne, Switzerland

- C - Plasma Engineering and Control.

P3C-C-161 DESIGN, IMPLEMENTATION AND TEST OF THE EXTREME SHAPE CONTROLLER (XSC) IN JET

ALBANESE Raffaele, G. Ambrosino(1) M. Ariola(1) A. Cenedese(2) F. Crisanti(3) G. De Tommasi(1) M. Mattei F. Piccolo(4) A. Pironti(1) F. Sartori(4) F. Villone(5) and JET-EFDA Contributors

(1)Ass. Euratom-ENEA-CREATE, Univ. Napoli Federico II, Italy (2)Consorzio RFX, Ass. Euratom-ENEA sulla Fus., Italy (3)Ass. EURATOM-ENEA sulla Fus., Frascati, Italy (4)Euratom/UKAEA Fusion Assoc., UK (5)Ass. Euratom-ENEA-CREATE, Univ. Cassino, Italy

Since in ITER the reference scenarios are planned to work at extreme plasma shape, JET operation will be progressively focused on the study of this kind of plasmas. The old JET Shape Controller (SC) can only control a few plasma-wall gaps at the same time. This, for strongly shaped plasmas, can lead to large deformations of the shape, mainly in case of large variations of poloidal beta bp and/or internal inductance li. A new JET eXtreme Plasma Shaping controller (XSC) was designed to achieve extremely shaped plasmas with the existing active circuits and control hardware. A linearized plasma model approach was used in designing the XSC. This allowed linking the applied active coils with any of the geometrical descriptors characterising the plasma boundary. Several (up to 36) gaps were used to describe the plasma shape accurately. However only a limited set of actuators is available (only 8 poloidal circuits in JET). The problem was tackled by using a singular value decomposition (SVD) to identify the principal directions of the algebraic mapping between coil currents and geometrical descriptors. These principal directions are then assumed as controller inputs/outputs so that the original multivariable control problem can be solved using a set of separate PID controllers. To take in account the limits of the actuators, the SVD orders the principal directions as a function of the current to shape sensitivity and the XSC normally uses only the first 5 or 6 directions (out of 8). This new system was successfully installed on the JET machine during 2003 without causing any interference to the plasma operation, or requiring long commissioning time. Eventually the new controller was used on really extremely shaped Internal Transport Barrier (ITB) experiments at high poloidal beta and in the presence of quite large variations of bp (Dbp up to 1.5) and/or li (Dli up to 0.5). The quality of the model based controller design approach was also been verified by a large sequence of plasma scenarios where extremely elongated shapes were achieved for the first time by using the controller without requiring any kind of tuning. The XSC controller architecture and philosophy also offer new interesting opportunities, e.g., the separatrix sweeping on the divertor plates without significantly affecting the overall plasma shape, and the possibility of improving the overall tokamak performance via combined control of plasma shape, current and profile.


Corresponding Author:

ALBANESE Raffaele

Assoc. Euratom-ENEA-CREATE, DIMET, Univ. Mediterranea RC, Loc. Feo di Vito, I-89060, RC, Italy

- C - Plasma Engineering and Control.

P3C-C-190 CORRECTION POSSIBILITIES OF MAGNETIC FIELD ERRORS IN WENDELSTEIN 7-X

Kißlinger Johann, Andreeva Tamara

Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald, Germany

WENDELSTEIN 7-X (W7-X) is an advanced stellarator with an optimised magnetic field configuration with respect of plasma confinement and stability. The super-conducting magnet system of W7-X has a modular structure and consists of five identical field periods. In order to reach the high quality of the magnetic field and to preserve the symmetry of the machine the magnet system need to be constructed and assembled with very high precision. Already small statistic deviations from ideal coil shapes or small not symmetric misalignments of the coils cause error field components which lead to additional magnetic islands and asymmetric thermal loads on the divertor targets. The requirement of good plasma confinement and reliable divertor operation limits the relative amplitude of error field component to be below 2 10-4. In order to reach this ambitious goal, correction possibilities are foreseen during the assembly and, if required, by the help of correction coils. Fabrication and assembly of the machine components are continuously accompanied with a precise geometrical survey. The perturbing impact of measured deviations from the ideal geometry is then analysed by magnetic field calculations and becomes the basis for adjustments. Thus, by a combination of distinct shifts and inclinations of some coils or modules it is possible to reduce or to cancel out perturbing field components. A numerical optimisation code was written to predict the best alignment combination of the coils on the basis of the geometrical analysis of the manufactured coils. Despite all measurements are done with high precision, some uncertainties of the geometrical shape of the magnet system and also adjusting tolerances will remain. In the case, that a significant field perturbation exist in the magnetic field, correction coils are required. The 10 control coils in W7-X, planed for island shaping and sweeping, are also usable for field error correction and are able to generate low order Fourier components up to 0.4mT. If larger correction fields are needed additional coils must be provided. As examples for such coils the capability and technical realisation possibility of normal-conducting coils inside the vacuum vessel, super-conducting helical windings within the cryostat and normal-conducting saddle shaped coils on top of the outer shell are investigated.


Corresponding Author:

Kißlinger Johann

Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, D-85748, Garching, Germany

- C - Plasma Engineering and Control.

P3C-C-207 REAL TIME CONTROL ENVIRONMENT IN THE RFX EXPERIMENT

Luchetta Adriano, Barana Oliviero, Cavinato Mario, Manduchi Gabriele, Taliercio Cesare

A complex, distributed, digital system has been implemented to realize a comprehensive set of control schemes that can be applied to the RFX experiment by means of the recently enhanced power supply system and the load assembly. Two basic feedback control schemes, the equilibrium control and the Resistive Wall Mode stabilisation, aim at extending the pulse duration beyond the time constant of the conducting shell for the penetration of radial field (50 ms), while other control schemes, such as the MHD Mode Control and the ‘Intelligent Shell’ aim at achieving a better understanding of the underlying physics. The distributed digital system consists of a set of seven computing nodes where each of which can be either a pre-processing station, in charge of acquiring raw data and processing intermediate control parameters, or a control station, in charge of driving the actuating power amplifiers. The paper provides an overview of the implemented system architecture with particular reference to the control schemes realized. Emphasis is given in the paper to the software real-time environment, which is providing not only basic functions, e.g.: data read-out, real-time communication and data collection, but also useful tools for programming and integrating control algorithms, performing simulating control scenarios and finally commissioning the system. Time-critical algorithms are coded directly in C fully exploiting the specific processor architecture only where necessary. This approach produces code that runs faster but is hardly portable among processor architectures. Non time-critical control algorithms are developed under Matlab and executed on the real-time processor. The advantage is that the coding process needs no software specialists and control engineers can focus on the algorithms. Portability of code depends in this case on the availability of a Matlab real-time toolbox for the candidate real-time operating systems. To perform simulations of various control schemes the real-time software environment is being extended to include a so called “simulation mode”. In this mode, the real-time controllers exchange their input/output signals not with the real system, but with a station running a suitable model of the system, for instance MAXFEA in the case of equilibrium control. In this way the control algorithm can be tested offline and the time needed for commissioning of algorithms can be reduced.


Corresponding Author:

Luchetta Adriano

corso Stati Uniti, 4 35127 Padova - Italy

- C - Plasma Engineering and Control.

P3C-C-233 A FAST AND VERSATILE INTERLOCK SYSTEM

Michel, Georg,

In todays large fusion experiments it is a common task to shut down a component quickly under certain conditions. These conditions can imply a complex logic and timing as well as topologically distributed inputs. In addition, the shutdown should be as instantaneous as possible. The shutdown mechanism should be flexible enough to allow for the changing requirements in the daily experimental work. The paper presents an interlock system which meets these reqirements. The system has a modular design consisting of an arbitrary number of identical distributed units which are connected to the general purpose computer network and to a dedicated interlock bus. The units consist of an FPGA (Field Programmable Gate Array) which implements the fast logic (nanosecond scale) and a microcontroller which is connected to the FPGA via its data and address bus. The microcontroller implements the network connectivity for the FPGA providing configuration, identification and reset functionality (millisecond scale). The units observe analog signals for the overstepping or understepping of a programmable threshold in a programmable time interval. Their main component is a commercial embedded microcomputer engine with FPGA (DragonEngine II) running µClinux. The interlock bus transmits the actual interlock and timing signals with the mere transmission line delay while the computer network is used for all other information. Due to this combination the interlock system can be made fast and versatile at the same time. A prototype of the interlock system has been built at IPP Greifswald and tested at the ECRH plant of W7-X. It consists of four units with ten inputs each.


Corresponding Author:

Michel, Georg

MPI fuer Plasmaphysik, Wendelsteinstr. 1, 17491 Greifswald, Germany

- C - Plasma Engineering and Control.

P3C-C-268 NEW VISUALIZATION SYSTEM FOR CONTROLLING AND MONITORING PURPOSES IN THE TJ-II STELLARATOR

Luis Pacios, Angel De La Peña, Ricardo Carrasco, Fernando Lapayese.

The Spanish Stellarator TJ-II is a highly flexible medium size fusion device of the heliac type in operation since the end of 1997. From its inception the TJ-II control system has performed satisfactorily over more than 10,000 plasma shots. The hardware and software, chosen ten years ago, for the control system architecture is based on a distributed network of embedded computers in VME crates running the OS9 real time operating system. For the realisation of the graphical user interface (GUI) of each sub-system the proprietary software named G-Windows and a dedicated video card were employed. In accordance with the experience acquired during the last 6 years of its operation, and bearing in mind new technological trends, the control system has been upgraded. For this purpose web-based software tools have been applied. The new visualization system based on Java applications frees us from the dependence on G-Windows. In addition, it provides new GUI features such as audio and video capabilities. Systems such as gas puffing, trigger programming, ground loop supervision, neutral beam injection control, and diagnostics can be fully configured and monitored in real time using any web browser or Java programs. Interactions between systems are based on Client-Server architecture and remote method invocation protocols. Two dedicated sockets in each system provide data and control flow between the control application and either client browser or java program. This paper will describe the new visualization system and shows the way it has been applied to real-time control applications.


Corresponding Author:

Luis Pacios

Asociacion EURATOM-CIEMAT para Fusion, Avda. Complutense 22, E-28040 Madrid, Spain

- C - Plasma Engineering and Control.

P3C-C-277 WEB-BASED GROUND LOOP SUPERVISION SYSTEM FOR THE TJ-II STELLARATOR

Angel De La Peña, Fernando Lapayese Luis Pacios Ricardo Carrasco

To minimize electromagnetic interferences in diagnostic and control signals and to guarantee safe operation of TJ-II, ground loops must be avoided. In order to meet this goal the whole grounding system of the TJ-II was split into multiple single branches that are connected at a single grounding point situated near the TJ-II structure in the torus hall. Each cabling rack and all the main components and structure must be grounded to the main grounding bar via separate ground wires. A real-time Ground Loop Supervision System (GLSS) has been designed, manufactured and tested by the TJ-II control group for detecting unintentional short circuits between separately grounded parts. The system measures loop impedance without breaking grounding connections in order to identify the short circuit. In accordance with TJ-II control system requirements, the GLSS hardware selected consists of VME analog and digital I/O boards, sets of emitter and sensor coils embracing the ground wires to be controlled and modular signal conditioning equipment. The measurement technique is based on the current induction method already employed in other fusion laboratories, e.g. in CRPP, Lausanne. For this a pulsed 1700 Hz sinusoidal current is continuously induced by a toroidal emitter coil in the ground wire for periods of 40 ms. Hence, if a ground loop is present this induced current can be detected by a current probe. With this set-up, the GLSS measurement range is between 10 ohm and 10 Kohm. At present, eight ground branches are controlled but this number could be increased easily if needed. The GLSS Software was developed using client-server architecture. A web server running on the Real Time Operating System OS-9 provides the interface with application control. Java graphical user interfaces allow remote access to the real-time ground loop measurement. Multiple mode operations and alarm thresholds can be configured via any web browser. This paper gives the detailed design of the whole TJ-II ground loop supervision system and its results during operation.


Corresponding Author:

Angel De La Peña

Laboratorio Nacional de Fusión, Asociación Euratom-Ciemat, Avda Complutense 22, 28040 Madrid (Spain)

- C - Plasma Engineering and Control.

P3C-C-291 A NEW CONTROLLER FOR THE JET ERROR FIELD CORRECTION COILS

Loris Zanotto, Marco Bigi(1), Fabio Piccolo(1), Filippo Sartori(1), Massimo De Benedetti(2)

(1) EURATOM/UKAEA Fusion Ass. Culham Science Centre, Abingdon, OX14 3EA, United Kingdom (2) Associazione Euratom-ENEA sulla Fusione, Frascati, Italy

Experiments about the stability of Resistive Wall Mode, error field induced locked modes and (neo-classical) tearing modes are routinely run on JET. In some of these experiments the MHD instability is excited by means of an external error field. In the past, such studies were performed by means of the Disruption Feedback Amplifier System (DFAS), which was devoted to supply a set of four Saddle Coils placed inside the vessel; the Saddle Coils are going to be completely dismissed during the 2004 shutdown. In order to continue experiments on error fields and tearing modes, a new system of Error Field Correction Coils has been recently installed outside the vessel aiming at producing a quasi-static error field. This new coils are supplied by the same DFAS. In order to optimise the performances of the DFAS with the new load and to improve the flexibility of the system, a completely new controller has been designed and implemented using a VME hardware platform. Its main tasks are: - to provide references for the DFAS up to 10 kHz - to provide a flexible trigger logic system - to provide reference components for the compensation of the natural error field - to provide the possibility to perform Resistive Wall Mode stabilisation and excitation. The hardware part has been simplified with respect to the previous solution, based on a cluster of C-40 Digital Signal Processors. All analogue input signals are processed by a conditioning stage and then acquired by the VME crate through a data acquisition board. Furthermore, an Asynchronous Transfer Mode interface board provides inputs from the Real Time Data Network (RTDN) of JET. The software part of the controller is based on a 100 kHz interrupt-driven algorithm and features a flexible trigger logic interface based on strings including RTDN signals, logical and mathematical operators. Therefore, reference to the amplifiers can be enabled and disabled using signals from the magnetic sensors or disruption precursor. Thanks to these new features and to the possibility to easily upgrade the system, due to the modularity of the software structure, the new controller will be a useful tool to perform error field and tearing modes studies on JET in the future. This paper will present the hardware and software structure of the new controller, along with the new possibilities offered by its very flexible trigger logic structure. Some results of the commissioning and the first operational experience will also be reported.


Corresponding Author:

Loris Zanotto

Consorzio RFX - Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti, 4 35127 Padova (Italy)

- C - Plasma Engineering and Control.

P3C-C-298 USING REAL TIME WORKSHOP FOR RAPID AND RELIABLE CONTROL IMPLEMENTATION IN THE FRASCATI TOKAMAK UPGRADE FEEDBACK CONTROL SYSTEM RUNNING UNDER RTAI-LINUX

Vitale Vincenzo, Cristina Centioli (1) Franco Iannone (1) Maurizio Panella (1) Luigi Pangione (2) Salvatore Podda (1) Luca zaccarian (2)

(1) Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy (2) Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Via del Politecnico 1 - 00133 Roma, Italy

The feedback control system running at FTU has been recently ported from a commercial platform (based on LynxOS) into an open-source Linux-based RTAI-LXRT platform, thereby obtaining significant performance and cost improvements. Based on the new open-source platform, it is now possible to experiment novel control strategies aimed at improving the robustness and accuracy of the feedback control task. Nevertheless, the experimentation of control ideas still requires a great deal of coding of the control algorithms that, if carried out manually, may be prone to coding errors, therefore time consuming both in the development phase and in the subsequent validation tests consisting of dedicated experiments carried out on FTU. In this paper we report on recent developments based on the Mathworks’ Simulink and Real Time Workshop (RTW) packages to obtain a user friendly environment where the real-time code required to implement novel control algorithms can be easily generated, tested and validated. Thanks to this new tool, the control designer only needs to specify the block diagram of the control task (namely, a high level and functional description of the novel algorithm under consideration) and the corresponding real-time code generation and testing is completely automated without any need of dedicated experiments. In the paper, the necessary work carried out to adapt the Real Time Workshop (RTW) to our RTAI-LXRT context will be illustrated. A necessary re-organization of the previous real-time software, aimed to incorporating the code coming from the adapted RTW, will also be discussed. Moreover, we will report on a performance comparison between the code obtained using the automated RTW-based procedure and the hand-written C code, appropriately optimized: at the moment, a preliminary performance comparison consisting of dummy non-trivial algorithms has shown that the code automatically generated from RTW is faster (about 30% up) than the manually written one. This preliminary result, combined with the fact that the use of RTW eliminates the coding workload leads to the conclusion that this approach to control implementation grants significant. According to the FTU experiments program, the actual deployment of the new tool is foreseen to be scheduled in the next experimental campaign.


Corresponding Author:

Vitale Vincenzo

Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy

- C - Plasma Engineering and Control.

P3C-C-301 THE SYSTEM ARCHITECTURE OF THE NEW JET SHAPE CONTROLLER

Filippo Sartori, G.Ambrosino(1) Marco Ariola(1) Gianmaria De Tommasi(1) Alfredo Pironti(1) Angelo Cenedese(2) Flavio Crisanti(3) Paul Mc Cullen(4) Fabio Piccolo(4)

(1)EURATOM-ENEA-CREATE, V. Claudio 21, 80125 Napoli, Italy (2)Consorzio RFX, EURATOM-ENEA, C.so Stati Uniti 4, I-35127 PD, Italy (3)EURATOM-ENEA, Frascati, C.P. 65, 00044-Frascati, Italy (4)EURATOM-UKAEA Fusion Ass.,Culham Science Centre,OX143DB,UK

The DSP based JET plasma Shape Controller (SC) has been replaced by a Power-PC system. The trigger for the work was the desire to experimentally validate the results of the Extreme Shape Controller (XSC) enhancement. The actual implementation went much further, producing an advanced and comprehensive solution that allows in the same pulse the free mixing of the traditional gap and current based control schemes with the full boundary control introduced by XSC. Changes to both the software architecture of the real-time controller and to the user interface were introduced, while still maintaining backward compatibility. The XSC control scheme was first adapted to JET operational needs in order to allow the imposition of current limits, and then generalised into a two-loop MIMO. The inner loop, SC, takes care of the de-coupling current control problem. The outer, XSC, is the multi-variable shape controller. 16 inputs and 8 outputs are reserved for external system connections to allow integration with the JET real-time control, feature that was successfully employed in the TAE experiment. The user interface was enhanced to allow, in different time windows, the programming of completely different control scenarios, both in term of matrices and set points. The XSC scenarios are just one subset of these, but can also be visually inspected and interactively tuned by the final user in order to achieve the desired objective. A major design objective was the minimisation of the commissioning effort of the final product. Since the most complex and safety critical part of the controller is the current and voltage limit software, the choice of interfacing XSC to SC via the existing limit avoidance logic, allowed the minimisation of the testing time. Once the original control and protection algorithms had been re-commissioned, it was then possible to rely on them while testing the most advanced schemes. The combination of this system architecture, with a model based design technique allowed a speedy introduction of the new controller, which has since been successfully operating catering for all the requirements of the experimental program. The new JET-EP divertor design will require further evolution within the current architecture. The sweeping logic will need to able to follow different sweeping trajectories. The termination scenarios will have to be made more flexible and a scheme will have to be developed in order to reduce the XSC saturation problems.


Corresponding Author:

Filippo Sartori

EURATOM-UKAEA Fusion Ass., Culham Science Centre, OX14 3DB, UK

- C - Plasma Engineering and Control.

P3C-C-350 AN INTEGRAL APPROACH TO PLASMA SHAPE CONTROL

Beghi Alessandro, M.Cavinato(2), A.Cenedese(2), D.Ciscato(1), S.Simionato(1)

(1)D.E.I. and Centro Ricerche Fusione, Università di Padova, Via Gradenigo, 6/A, I-35131 Padova, Italy. (2)Consorzio RFX Associazione EURATOM ENEA sulla Fusione, Corso Stati Uniti, 4, I-35127 Padov

In the future generation of Tokamaks the reactor performances are strictly related to the capability of sustaining and controlling strongly shaped plasmas. Currently the shape control is based on a pointwise description of the plasma boundary usually in terms of a set of plasma-wall distances (gaps). In this paper an alternative approach is presented, based on an integral description of the plasma boundary using the parametric curve model given by the B-Splines [1]. This model permits to construct a piecewise polynomial curve - globally Cn-2 if n is the B-Spline order - by linearly combining piecewise polynomial basis functions. With such a tool it is possible to describe a wide range of curves with a relatively low number of parameters, giving a compact characterization of the whole shape. The procedure consists in the extraction of the plasma boundary curve from the flux map, which is then fitted by means of planar B-Splines. Two approaches to the extraction of the boundary curve have been tested. The first is based on the active contour approach and involves the deformation of the curve according to a functional minimisation, while the second operates directly the fitting of the boundary curve with the parametric model. The attention was then focused on the second approach since it offers an easier way to obtain the surjectivity in the relation shape-parameters and presents a reduced computational cost. In order to ease the control design we studied two techniques to convert the two-dimensional description of the curve in a one-dimensional signal with no or little loss of information. These techniques permit to almost halve the number of integral descriptors while maintaining the same precision. With reference to the ITER machine, the number of shape parameters was reduced to ten allowing the design of a controller based on the decoupling technique similar to the gap controller. The comparison of the performances given by the two control approaches is presently underway and focuses on the control of global shape parameters such as triangularity and elongation. The integral approach is expected to give a smoother control of these quantities which have a strong relationship to the overall plasma performances. [1] A. Blake and M. Isard, “Active Contours” Springer, 1998.


Corresponding Author:

Beghi Alessandro

D.E.I. and Centro Ricerche Fusione, Università di Padova, Via Gradenigo, 6/A, I-35131 Padova, Italy

- C - Plasma Engineering and Control.

P3C-C-352 LINEARIZED MODELS OF THE PLASMA RESPONSE IN THE NEW RFX LOAD ASSEMBLY

Bettini Paolo, (1)M. Cavinato, (1)G. Marchiori, (2)F. Villone

(1) Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova, Italy (2) Ass. EURATOM/ENEA/CREATE, DAEIMI, Università di Cassino, Via Di Biasio 43, I-03043, Cassino (FR), Italy

The RFX load assembly now includes a new 3 mm thick copper shell, close fitting to the vacuum vessel, which will allow to perform active control experiments by means of a set of 192 saddle coils surrounding the torus. The much shorter time constant of the shell also requires an active control system of the plasma equilibrium. At present, the design of the control system in fusion devices is generally based on linearized models of the plasma response at different equilibria. In the past, the CREATE_L plasma response model, derived by linearizing the MHD equilibrium equation and Ohm’s law in the active and passive conductors and in the plasma, was successfully applied to a RFP plasma in the presence of the old RFX magnetic configuration. One of the aims of this paper is to apply this procedure also to study in details the effects of the structural modifications of RFX on the open loop plasma response. Since experimental data are not yet available, a two dimensional FEM code, solving the ideal MHD free boundary problem in axisymmetric geometry, has been used to provide a set of equilibrium reference data, also in terms of coil currents and virtual measurements from pick-up and flux loop probes. As a first step in the validation procedure, in particular to assess the accuracy of the electromagnetic model of the passive structures implemented in the codes, a cross-check has been carried out on a shot without plasma, programmed to apply a vertical field step. The agreement between linear and non-linear model is very good. As a second step, complete comparisons have been performed in the case of pulses with plasma. The currents in the active coils were imposed, while the time behaviour of the currents in the passive structures was self-consistently calculated. This was possible thanks to the fact that the plasma is open-loop stable allowing us to focus only on the error introduced by the open loop model. A further step in view of plasma control has been the inclusion of active coil voltages as inputs to the model. The preliminary results are encouraging in both cases. Cross-comparisons with another available linearized model of the plasma response, derived according to a different perturbation method, are also presented.


Corresponding Author:

Bettini Paolo

DIEGM, Università di Udine, Via delle Scienze, 208, I-33100 Udine, Italy

- C - Plasma Engineering and Control.

P3C-C-354 DESIGN OF THE NEW RFX EQUILIBRIUM ACTIVE CONTROL SYSTEM

Cavinato Mario, G.Marchiori(1)

(1)Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione,Corso Stati Uniti 4, I-35127 Padova,

The RFX new load assembly is characterized by a thin copper shell with a 50 ms time constant for the penetration of the vertical magnetic field, shorter than the pulse length. Thus an active control system is necessary to assure the plasma horizontal equilibrium. The design has been carried out by developing for the RFP the classical method used in the case of ITER or for recent upgrades of presently operating Tokamak devices. First of all the geometry of the RFX new load assembly was implemented in a FE MHD plasma equilibrium code (MAXFEA), then equilibria at different plasma currents (750 kA and 2 MA) were calculated and corresponding linearized models of the plasma response were derived by means of a perturbation method. Finally, the design of the controllers was accomplished on the basis of the available linear models. A nested loop structure was kept, following the scheme already adopted in the old RFX axisymmetric control system. In the outer loop the position regulator produces a correction of the feedforward vertical magnetic field calculated according to Shafranov’s equilibrium theory on the basis of the instantaneous value of plasma current and other equilibrium quantities such as desired major radius, minor radius, coefficient of poloidal field asymmetry. The reference field is then transformed into 8 Field Shaping winding current references for the inner loop. Unlike the old system, where each current was independently regulated by the corresponding amplifier, a decoupling current control system has now been envisaged which should assure the same dynamic behaviour of the 8 current tracking errors. The output of this loop are 8 voltage references for the amplifiers. The whole control system has then been implemented in the non-linear FE model of RFX along with the full poloidal field electrical circuit in order to assess its performances and robustness. Tuning of the regulator parameters was also performed on the basis of the simulations results.


Corresponding Author:

Cavinato Mario

Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione,Corso Stati Uniti 4, I-35127 Padova

- C - Plasma Engineering and Control.

P3C-C-363 REAL-TIME MEASUREMENT AND CONTROL AT JET- EXPERIMENT CONTROL

Felton, Robert,

Selected also for oral presentation O1B-C 363

Over the past few years, the preparation of ITER-relevant plasma scenarios has been the main focus experimental activity on tokamaks. The development of integrated, simultaneous, real-time controls of plasma shape, current, pressure, temperature, radiation, and neutron profiles, and also impurities, ELMs and MHD are now seen to be essential for further development of quasi-steady state conditions with feedback, or the stabilisation of transient phenomena with event-driven actions. For this thrust, the EFDA JET Real Time Project has developed a set of real-time plasma measurements, experiment control, and communication facilities - currently, the largest real-time experiment control facility on a Tokamak. The Plasma Diagnostics used for real-time experiments are Far Infra Red interferometry, polarimetry, visible, UV and X-ray spectroscopy, LIDAR, bolometry, neutron and magnetics. Further analysis systems produce integrated results such as temperature profiles on geometry derived from MHD equilibrium solutions. The signal processing algorithms were validated on many recorded pulses, and the systems have real-time data checks. The Actuators include toroidal, poloidal and divertor magnets, gas and pellet fuelling, neutral beam injection and RF and microwave beam injection. The Heating/Fuelling Operators can either define a power or gas request waveform or select the real-time instantaneous power/gas request from the Real Time Experiment Control (RTCC) system. The Real Time Experiment Control system provides both a high-level, control-programming environment and interlocks with the actuators. It is capable of single-input, single-output and multiple-input, multiple-output controls. A MATLAB facility is being developed for the development of more complex controllers. The communications network has been critical to the successful operation of the facility. It uses ATM (the core technology of ISDN) to multicast more than 500 signals (total), in 30 different data sets, every few milliseconds, amongst 30 systems. The network can be extended easily and quickly, and is reliable and robust. The EFDA Real Time project is essential groundwork for future reactors such as ITER. It involves many staff from several institutions. The facility is now frequently used in experiments. This work has been conducted under the European Fusion Development Agreement and is partly funded by Euratom and the UK Engineering and Physical Sciences Research Council.


Corresponding Author:

Felton, Robert

Euratom/UKAEA Fusion Association, Culham Science Centre Abingdon OX14 3DB UK

- C - Plasma Engineering and Control.

P3C-C-375 OPEN LOOP CHARACTERIZATION OF AN ACTIVE CONTROL SYSTEM OF MHD MODES

Masiello Antonio, Per Brunsell(1) Giuseppe Marchiori(2) Dimitriy Yadikin(1)

(1)Division of Fusion Plasma Physics - Association EURATOM-VR, Alfvén Laboratory, Royal Institute of Technology, S-100 44, Stockholm Sweden (2)Consorzio RFX - Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova Italy

In the framework of the collaboration between the research groups of RFX and T2R on the active control of MHD modes, a digital system was installed in T2R, able to control up to 32 active coils. A state space full electromagnetic model, already developed in view of MHD mode control in RFX, was adapted to the T2R load assembly and saddle coils layout. It consists of a lumped parameter electromagnetic model of the saddle coil system and a linear model of the evolution of RWMs in Reversed Field Pinch plasmas. Electrical parameters are calculated by a FE electromagnetic model, which includes the conductive structures surrounding the plasma column. An experimental campaign both without and with plasma has now been carried out on T2R aimed at the electromagnetic characterization of the open loop system and to the validation of the various parts of the model. The first series of tests was performed without plasma to evaluate the coil self-inductance and mutual inductance with the underlying and adjacent sensors. In the presence of a passive conductor (the two-layered thin shell in T2R case), both these parameters depend on the frequency. By properly programming the control system, tests with sinusoidal reference voltage at different frequencies were performed. In order to analyse steady-state behaviour a voltage reference step was also provided for a time longer than the shell time constant for the penetration of the vertical field. A satisfactory agreement was observed between the electromagnetic parameters calculated by the FE model and those derived by processing the experimental data. Static and rotating perturbation of different harmonic order were then applied with constant and sinusoidally varying reference voltages. In particular, the quality of the harmonic content as a function of the number of active coils has been analysed. A clear knowledge of the sidebands amplitude of each mode in the real machine is mandatory for the design of the mode control system. Finally, the third group consisted of shots with plasma. Static perturbations of different toroidal order n with a step reference voltage were applied to study their effect on the evolution of plasma MHD modes. The purpose was to compare theoretical and experimental growth rates and to understand how low is the threshold to start an unstable mode. This represents a basic information to determine how clean the harmonic content must be to allow a selective mode control.


Corresponding Author:

Masiello Antonio

Consorzio RFX, Corso Stati Uniti 4, 35127 Padova, Italy

- C - Plasma Engineering and Control.

P3C-C-377 COMPARISON OF STRATEGIES AND REGULATOR DESIGN FOR ACTIVE CONTROL OF MHD MODES

Marchiori Giuseppe, Per Brunsell(1), Mario Cavinato(2), Demetrio Gregoratto(2), Roberto Paccagnella(2), Dimitriy Yadikin(1)

(1)Division of Fusion Plasma Physics - Association EURATOM-VR, Alfv¨¦n Laboratory, Royal Institute of Technology, S-100 44, Stockholm Sweden (2)Consorzio RFX - Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova Italy

The active control of MHD modes is essential for the steady-state operation of fusion magnetic confinement devices. Reversed field pinch experiments are characterized by an intrinsic richness of MHD modes. The replacement of old thick shells with much thinner ones, whose time constants are typically shorter than the pulse length, made it possible to design active control systems of RWM's in T2R and RFX. In T2R a set of saddle coils and probes has already been installed. An experimental campaign of mode active control is now under way whose results will be useful for improving the models and the future activity on RFX. Two main control strategies have been considered: intelligent shell and mode control. The former is aimed at reproducing a virtual shell by zeroing the flux through each saddle probe, the latter at investigating the possibility of a feedback stabilization of selected MHD modes. According to the desired scheme, the real control system can handle the magnetic field measurements of the single probes or the harmonic components produced by on-line FFT. Since the output variables of the control model are the magnetic field harmonics, a direct analysis of the mode control scheme is straightforward, while an inverse DFT is needed to reconstruct the field at the sensors. Both strategies can be then implemented in the model and in the real system. The design of intelligent shell controller is carried out taking into account the electromagnetic parameters of a coil and its coupling with the surrounding ones. A more complex approach is needed in the case of mode control, the design being strongly affected by the number of available actuators. In T2R some unstable modes are not within the spectrum directly producible by the set of coils (|n|<=8), even if they can be detected by the set of probes. Two controllers are analysed to cope with this problem: the former aims at controlling the unstable modes not directly reachable by using the knowledge of the sidebands produced by the actuators. The latter is designed on the basis of the developed linear model by means of the pole placement technique. In RFX, since the number of saddle coils allows to counteract directly all the unstable modes, the sidebands are not involved in the design of the controller. Closed loop tests on the T2R machine, for all control strategies, are presented in the paper highlighting their different performances in terms of mode stabilization and modification of plasma spectrum.


Corresponding Author:

Marchiori Giuseppe

Consorzio RFX, Corso Stati Uniti 4, 35127 Padova, Italy

- C - Plasma Engineering and Control.

P3C-C-383 ADOPTING MODERN NONLINEAR CONTROL TECHNIQUES FOR THE PLASMA STABILIZATION ON THE NOVEL LINUX-BASED FEEDBACK CONTROLLER OF FTU

Zaccarian Luca, Cristina Centioli (1) Franco Iannone (1) Maurizio Panella (1) Luigi Pangione (2) Salvatore Podda (1) Vincenzo Vitale (1)

(1) Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy (2) Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Via del Politecnico 1 - 00133 Roma, Italy

In this paper we will report on the experimental results arising from the implementation of modern control techniques for the optimization of the RF power coupling vs the plasma conditions in the FTU experimental facility. These experiments are carried out by employing the open-source Linux-RTAI control system currently running on the FTU digital feedback loop. The RF power source under consideration is a Lower Hybrid System (LH) based on 6 gyrotrons with a nominal power output capability of 1.1 MW each. The optimization of the coupling level between the plasma and the emitting antenna reduces the reflected power thus maximizing the heating effects in addition to avoiding danger to the emitter (equivalently, annoying safety shutdowns of the system). To this aim, the plasma displacement is modified by suitably adjusting the reference input to the stabilizing feedback, according to “extremum seeking” techniques based on a modified gradient algorithm that recently appeared in the control literature. It will be shown in the paper how this algorithm achieves a satisfactory level of robustness with respect to measurement errors and well performs in experimental tests, thus leading to an improved effectiveness of the RF heating system.


Corresponding Author:

Zaccarian Luca

Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Via del Politecnico 1 - 00133 Roma, Italy

- C - Plasma Engineering and Control.

P3C-C-403 VERTICAL STABILITY OF ITER PLASMAS WITH 3D PASSIVE STRUCTURES AND A DOUBLE LOOP CONTROL SYSTEM

Portone Alfredo(1), R. Albanese(1), R. Fresa(2), M. Mattei(1), G. Rubinacci(3), F. Villone(3)

(1) Assoc. Euratom-ENEA-CREATE, Univ. Mediterranea, Feo di Vito, I-89060 RC, Italy (2) DIFA, Univ. della Basilicata, Contrada Macchia Romana, I-85100 PZ, Italy (3)Assoc. Euratom-ENEA-CREATE, Univ. Cassino, Via Di Biasio 43, I-03043 Cassino (FR), Italy

The main focus of this study is the analysis of the stabilizing effect on the plasma vertical motion of the 3D eddy currents induced in the ITER Vacuum Vessel and Outer Triangular Support and the comparison with the predictions obtained by means of simplified 2D models. This study includes – firstly - the derivation of linear models describing the dynamics of the n=0 plasma displacements around the main ITER equilibrium configurations. These models are derived by linearising the Grad-Shafranov equation about few key equilibrium configurations, thus granting the consistency of the derived linear models with the MHD equilibrium constraint. Secondly, an assessment of the effects of the 3D geometry of the Vacuum Vessel, Blanket and Outer Triangular Support is carried out, with particular emphasis on the effects of the upper and lower ports on plasma stability margin, growth time, minimum stabilization voltage and control loop phase margin. At last, the performances of the present ITER control system (single loop) are assessed and compared to those of an upgraded system (double-loop) that is here proposed to improve the stability domain of the ITER plasmas forecasted.


Corresponding Author:

Portone Alfredo(1)

EFDA-CSU, Boltzmannstrasse 2, D-85748 Garching, Germany

- C - Plasma Engineering and Control.

P3C-C-408 THE BASIC METHODS FOR UNDERSTANDING OF PLASMA EQUILIBRIUM TOWARD ADVANCED CONTROL

KURIHARA Kenichi, KAWAMATA Youichi, SUEOKA Michiharu, HOSOYAMA Hiromi, YONEKAWA Izuru, SUZUKI Takahiro, OIKAWA Toshihiro, IDE Shunsuke, JT-60 Team

Japan Atomic Energy Research Institute Naka Fusion Research Establishment 801-1 Mukoyama Naka-machi Naka-gun Ibaraki-ken, 311-01, JAPAN

Since tokamak magnetic fusion research has just made a step forward to an international collaborative project ITER, the existing tokamaks including JT-60 are expected to explore more advanced operation scenarios. To test those scenarios in the JT-60 experiment, the basic methods for understanding of plasma equilibrium applicable to the ITER have been developed. Some of them have been accomplished, and the other are being conducted as follows: (1) A complete plasma shape is precisely reproduced in real time. Plasma entire current and the weight center (near magnetic axis) of the current are provided also in real time. (2) Eddy current effects are considered for shape reproduction. (3) A plasma current profile in the poloidal cross-section is reproduced in real time through a new advanced algorithm. (4) For long-pulse DT operation in ITER, pick-up coil sensors are equipped near a plasma, while absolute magnetic field sensors are placed in a distance from a plasma to avoid high neutron irradiation. A method is developed to correct the drifted signal of the integrator for a pick-up coil by employing distant sensor signals. In the symposium, those methods will be explained in detail with the experimental results at JT-60, and the calculation in the ITER geometry, including the remaining problems. On the basis of such discussion, we would like to envisage a future of plasma equilibrium control toward ITER and a fusion power plant.


Corresponding Author:

KURIHARA Kenichi

Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, 801-1 Mukoyama Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan

- C - Plasma Engineering and Control.

P3C-C-457 XSC PLASMA CONTROL: TOOL DEVELOPMENT FOR THE SESSION LEADER

Ambrosino Giuseppe, R. Albanese(2) M. Ariola A. Cenedese(3) F. Crisanti(4) G. De Tommasi M. Mattei(2) F. Piccolo(5) A. Pironti F. Sartori(5) F. Villone(6) and JET-EFDA Contributors

(2)Assoc. Euratom-ENEA-CREATE, Univ. Mediterranea Reggio Calabria, IT (3)Assoc. Euratom-ENEA-Consorzio RFX, IT (4)Assoc. Euratom-ENEA-Frascati, IT (5)EURATOM-UKAEA Fusion Association, UK (6)Assoc. Euratom-ENEA-CREATE, Univ. Cassino, IT

A JET Enhancement was aimed at designing and implementing a new model-based shape controller (XSC, i.e., eXtreme Shape Controller) able to operate with high elongation and triangularity plasmas. In 2003 the XSC has been implemented on a new hardware architecture and successfully tested in a various experiments. The use of XSC implies a number of steps, which at present are not automated and therefore imply the involvement of several experts. The first step is plasma modelling, based on linear (CREATE-L) and nonlinear (CREATE-NL) tools, which call for: - definition of target plasma configuration in terms of plasma current, shape and main plasma current profile parameters (poloidal beta and internal inductance); - determination of a set of poloidal field currents; - production of a linearized model. The second step is the controller design, based on the CREATE-XSC-GEN tool, which uses SVD (singular value decomposition) to find the best combination of currents to obtain specific changes in the shape, requiring: - selection of weight for plasma shape parameters (plasma-wall gaps) and actuators (circuit currents); - selection of SVD tolerance; - determination of controller gains; - test of controller robustness and definition of operational space. The third step is the creation of the configuration file. The CREATE-EGENE tool creates a set of similar configurations each with a related set of equilibrium currents using as input both the linear model and the controller design and produces the configuration file, containing the information needed for the Level 1 XSC session leader interface. The fourth step is the use of Level 1 XSC session leader interface, offering the possibility of picking up a baseline shape and adjusting it using visual tools, showing the nominal currents needed for each value of poloidal beta and internal inductance.


Corresponding Author:

Ambrosino Giuseppe

Associazione Euratom-ENEA-CREATE, Dip. di Informatica e Sistemistica, Univ. Napoli Federico II, Via Claudio 21, I-80125 Napoli, Italy

- C - Plasma Engineering and Control.

P3C-C-463 A FLEXIBLE AND REUSABLE SOFTWARE FOR REAL-TIME CONTROL APPLICATIONS AT JET

Gianmaria De Tommasi, Fabio Piccolo(1) Filippo Sartori(1) JET-EFDA contributors(2)

(1)EURATOM-UKAEA Fusion Ass. Culham Science Centre, Abingdon, OX143EA, UK (2)Work performed under EFDA and partly funded by EURATOM and the UK Engineering and Physical Sciences Research Council.

At the heart of the JET machine, several real-time control, measurement and protection systems collaborate in order to satisfy the most sophisticated experimental needs. A group of these systems shares the same software architecture: the plasma shaping system “Extreme Shape Controller”(XSC), parts of the “Vertical Stabilisation” control and acquisition, the “Error Field Coil Controller”. Also many can be found among the “Real Time Data Network” (RTDN) processing nodes: the one measuring plasma internal parameters, the density and q profile estimator, two different real-time equilibrium codes, a neural network based disruption prediction code and a Matlab script real-time executor. The fast growth of the JET real-time control network and the resulting need of shorter development cycle were the triggers that started the development of the “JETRT” software. Initially just designed to help producing RTDN control node PCs, because of the XSC project it finally developed in an all-purpose multi-platform real-time environment. This new architecture is designed for maximum reuse. On average two third of the software is the same in all applications independently from the platform. The varying part is the project specific algorithm, which is also compiled into a separate software component, in order to achieve a separation from the plant interface code. This scheme maximises reliability while at the same time reducing development costs. In addition it helps collaboration with external development groups, since it allows focusing only on the novel and interesting parts of the project without the need of knowledge about JET specific details. Finally it enables non-specialist programmers to contribute to the real-time project. A very important feature of the architecture is that it provides an integrated set of debugging and testing tools. The compiled task specific component can be validated on “Matlab” by comparing it to its prototype. It can also be tested by simulating any old JET pulse, either in a standalone fashion or real-time and integrated in a mock-up real-time network. Because of these features it was possible to debug “XSC” on a PC Windows platform where better tools were available. Any bug encountered while testing on the machine were successfully repeated in the laboratory in a controlled simulation. This was the key to reducing the commissioning of Shape Controller to a few dedicated pulses compared to the many days needed in 1994.


Corresponding Author:

Gianmaria De Tommasi

Associaz. EURATOM/ENEA/CREATE, Università di Napoli Federico II, Napoli, Italy

- C - Plasma Engineering and Control.

P3C-C-508 COMMISSIONING TESTS FOR CONTROL PROCESSES IN ASDEX UPGRADE´S NEW CONTROL AND DATA ACQUISITION SYSTEM

Thomas Zehetbauer, D. Zasche(1) T. Vijverberg(2) R. Cole(2) K. Lüddecke(2) G. Neu(1) G. Raupp(1) W. Treutterer(1)

(1)Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, D-85748 Garching (2)Unlimited Computer Systems, Seeshaupterstrasse 15, D-82393 Iffeldorf, Germany

ASDEX Upgrade is being equipped with a new CODAC system. About 20 real-time control applications process about 250 input signals and another 250 reference signals to perform a large number of functions for actuator feedforward and plasma feedback control, for monitoring of machine and plasma states, and for alarm handling, IO and reference value generation. If a control system is commissioned together with a new machine then functions can be added consecutively and be tested iteratively while the machine´s operation range is explored. In our case the control system must be replaced on a running machine. The complete set of functions must be provided instantaneously at full machine parameters. To ensure that algorithmic function and real-time performance of all application processes can be tested sufficiently prior to operation we implemented an in-situ simulation method. The new CODAC system permits to download complete sets of trajectories of simulation data into the processes for the measurement input and the reference value generation. Then an open-loop simulation discharge can be started where the simulation data drives the control application processes. The real-time process output and performance can be protocolled and analysed. Simulation data can be generated from discharges executed and protcolled by the old control system, be computed by other sources of simulation data, or be edited manually, to tune the simulation run to specific states of interest. In future, the method can be enhanced for closed-loop operation with model-driven processes computing the response of specific plasma quantities or machine components. Currently the system is used to verify application process performance upon implementation or whenever modified, and to test control system reaction to specific failure states.


Corresponding Author:

Thomas Zehetbauer

Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, D-85748 Garching

- C - Plasma Engineering and Control.

P3C-C-510 OPTIMIZATION OF THE IGNITOR OPERATING SCENARIO AT 11 MA

Ramogida Giuseppe, Cucchiaro Antonio (1), Galasso Giuseppe (2), Pizzuto Aldo (1), Rita Camillo (1), Roccella Massimo (1), Prof. Coppi Bruno (3)

(1) Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy (2) Ansaldo Ricerche, Corso Perrone 25, 16152 Genova (GE), Italy (3) MIT, 02139 Cambridge (Ma), USA

Several endeavours have been made in order to minimize technological troubles in the IGNITOR design, reducing electromagnetic loads, power supply requests and use of not very known materials. The present design alterations comprise both the operating scenario for the poloidal field (PF) currents and the PF coils geometry. The EM loads and the temperature gradient into the PF coils have been reduced, changing the current distribution between the coils and introducing a more efficient grading in the most solicited ones, avoiding the use of Dispersion Strengthened Copper foreseen in the previous design. The new reference operational scenario for IGNITOR with a maximum plasma current of 11 MA has been modified according to the engineering and power supply constraints. The edge safety factor was kept above 3.5 during the whole plasma evolution, reducing the risk of low-q disruption events. This analysis and a relevant simulation of a typical fast vertical disruption, in the most dangerous plasma conditions in IGNITOR, have been simulated with the MAXFEA code. The obtained values are within the engineering constraints during the whole operating scenario and the typical plasma vertical disruption. This scenario provides poloidal flux capability enough to balance the plasma flux requirement even without relying on the effect of the bootstrap current. In order to reduce the EM loads on plasma chamber and the related stresses due to plasma disruptions, it has been investigated a new approach based on the possibility of mitigating these loads by using copper toroidal layers added to the plasma chamber in proper positions. It has been found that this expedient could be quite effective not only in increasing the time constant of the plasma displacement but even in reducing the vertical force and its combined effect with hoop force on the vessel. This study has been carried on using MAXFEA simulations with different location and extension of the layers, and has shown that the reduction of the vertical force on the vessel, due mainly to the reduction of the halo current component of this force, is maximum when the copper layers are located on the higher and lower end of the plasma chamber. In this configuration the copper layers result to be very effective in increasing the stabilizing component of the eddy current (due to the plasma displacement) without increasing the destabilizing one (due to the plasma current quench) and then in reducing the forecast halo current.


Corresponding Author:

Ramogida Giuseppe

Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy

- C - Plasma Engineering and Control.

P3C-C-517 PLASMA FEEDBACK CONTROLLER REORGANISATION FOR ASDEX UPGRADE'S NEW DISCHARGE CONTROL AND DATA ACQUISITION SYSTEM

W. Treutterer, T. Zehetbauer, V. Mertens, G.Neu, G. Raupp, D. Zasche

Accompanying the introduction of the a new discharge control and data acquisition system ASDEX Upgrade's feedback control scheme is reworked. Since all the algorithms must be ported from the previous programming language OCCAM into C++ and must be embedded into a completely new infrastructure environment, a good chance to restructure also the control architecture emerged. It is not only that the diverse, historically grown algorithmic implementations are unified and equipped with a common interface. Even more important is the trend of clustering groups of up-to-now independent single-input-single-output (SISO) controllers into a few multiple-input-multiple-output (MIMO) controllers. This is necessary to account better for the dependencies and cross-couplings in the plasma process to be controlled and it provides a means for future optimisation once the dependencies are better understood. Also the previous discharge control system had a primitive method to adress this aspect: the control recipies. A recipe contained a valid combination of active SISO feedback controllers, deactivating any competing controllers using the same actuators or tracking the same or dependent control variables with different means. The new approach tries to identify such competing controllers and then group them into MIMO controllers based on common actuators.A control mode for a whole group, switchable in real-time, is the successor of the recipe.With this powerful tool it is not only possible to dynamically choose the active control variables but also to change control algoritm and control parameters. In addition configurable options for actuator specific behaviour like output limitation are provided. Additional flexibility is offered with pre- and postprocessors plugged into the real-time signal exchange architecture in which the controllers are embedded. These may filter measurements or pulse-modulate controller output commands e.g. for NBI heating. Commissioning of the new control system kernel has already started. The restructured feedback control scheme will be implemented during the second part of the commissioning phase. After the summer maintenance break ASDEX Upgrade's new control system will be ready for operation.


Corresponding Author:

W. Treutterer

Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching, Germany

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-22 NEUTRON ANALYSIS OF H-ALPHA AND CXRS DIAGNOSTICS OF ITER

S.V. Sheludiakov, G.E. Shatalov, K.Yu. Vukolov

123182, Moscow, Russia, Kurchatov sq. 1

The neutron analysis was carried out for two optical diagnostic systems of ITER, developed in RF: H-alpha spectroscopy and Charge Exchange Recombination Spectroscopy (CXRS). Both systems are similar from the neutron analysis point of view– they include extended cavities, along which the neutrons can penetrate up to cryostat. H-alpha diagnostic location in the upper port of ITER sector 10 was chosen as a base option for the detailed calculations, because its channel is most dangerous from point of view of neutron flux on window and optical fiber waveguide. On its basis comparative analysis was done for channels of diagnostics in sectors 2 and 9. The obtained data is very important for material selection (mirror, window, fiber waveguide, fastening assemblies), and also for definition of irradiation test conditions such as radiation fluxes and spectra. For example, the neutron flux on the primary quartz window does not exceed 4x108 n/cm2/s. Results of irradiation tests of quartz glass in nuclear reactor are taking into account. It is possible to use windows (lens) from KU-1 quartz glass (practically without change of light transmission in visible range) during all D-T ITER phase. Moreover, there will be possible also to apply fiber waveguide from quartz glass if neutron flux will not be significantly enlarged by background radiation. Neutron flux on cryostat makes 7x107 n/cm2/s even on axis of the diagnostic channel together with background (neutron flux without any diagnostic at port), which is permissible for requirement to access the personnel for 10 days after reactor shutdown. Anyway, the increase of neutron flux carries local character, and the neutron flux promptly decreases from axis of the channel. Effect of several systems integration in one port is difficult for estimating, but the carried out investigation shows the way for construction improvement. It will allow to pick up acceptable design satisfying to the safety requirements. For example, the introduction of the diagnostic channel additional bend allows essentially to decrease the dose rate on cryostat. Besides a port shield plug with H-alpha diagnostic can be enlarged if necessary up to port extension.It allows to decrease neutron flux on window two times.


Corresponding Author:

S.V. Sheludiakov

123182, Moscow, Russia, Kurchatov sq. 1

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-42 NEW CONSTRAINTS FOR PLASMA DIAGNOSTICS DEVELOPMENT DUE TO THE HARSH ENVIRONMENT OF MJ CLASS LASERS

BOURGADE Jean-Luc, V. Allouche, J. Baggio, C. Bayer, J. Beullier, F. Bonneau, C. Chollet, S. Darbon, L. Disdier,D. Gontier, M. Houry, H.P. Jacquet, J.P. Jadaud, J.P. Le Breton, J.L. Leray, P. Leclerc, I. Masclet-Gobin, J.P. Negre, J. Raimbourg, J. Vierne

CEA/DIF, BP n 12, 91680 Bruyères le Châtel, France

Diagnostics developed for laser-produced plasmas have usually been designed without taking into account the direct effects of radiated or nuclear energies emitted by the plasma itself. The future MJ class lasers (LMJ in France and NIF in USA), now under construction, are designed to demonstrate ignition and gain of an ICF target. A 10x gain target will emit up to 1019 DT fusion neutrons. At these neutron production levels, and significantly below them (e.g. for pre-ignition studies), the designs of plasma diagnostics for will be dramatically affected by these new environmental effects. These facilities will be able to focus up to 1.8 MJ UV laser light in few nanoseconds into a mass of a less than 1/3 g. This very large power density will drive a large amount of x-rays, debris, shrapnel. Unfortunately this colossal energy increase with respect to the previous laser facilities (Phébus, NOVA or Omega that have or produce at maximum 30 kJ of UV laser light) cannot be compensated for increasing the size of the emitting region, which remains nearly unchanged. The spatial resolution of primary diagnostics must be preserved and some of the diagnostic parts will need to be nearly as close to the plasma as in the past. The harsh environmental background induced on the diagnostic’s active components must be taken into account not only for successful gain conditions but for many other experiments where neutron yields are within 1014 up to 1016 n/4??? The solution to achieving the goal of diagnostic survivability is not obvious and many new studies must be conducted to verify the diagnostic performance under these environmental conditions. Simulations and experiments conducted at CEA/DAM related to the preliminary studies of this harsh environment effects on diagnostic components will be presented. Recent experiments performed at the OMEGA laser facility at the University of Rochester (USA) within the framework of the CEA/NNSA-LLNL collaboration, including neutron effects on CCD readout, coaxial cable, streak or framing cameras and EMP measurements, will be fruitful for the design of diagnostics for the MJ class laser facilities diagnostic design. New measurement strategies that have been developed to overcome these environmental difficulties will be presented. Synergies with similar environment studies conducted for magnetic fusion diagnostic design for next ITER facility must be considered and enhanced.


Corresponding Author:

BOURGADE Jean-Luc

CEA/DIF - BP12 - 91680 Bruyères le Châtel - France

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-52 THE INTEGRATED VISUALISATION SOFTWARE FOR THE ITER IN VESSEL VIEWING SYSTEM (IVVS)

Riva Marco, Carlo Neri(1), Fabio Bonaccorso(2), Federico Massaioli(2)

(1) Associazione EURATOM-ENEA sulla Fusione, Centro Ricerche Frascati, C.P. 65, 00044, Frascati, via E. Fermi 45 (2) Consorzio CASPUR, Roma, Italy

The Amplitude-modulated In vessel viewing system (IVVS), developed by ENEA for the ITER machine, is a versatile apparatus capable to scan the internal of a vessel and acquire amplitude and range data of a quasi-spherical view. By coupling the two images we obtain a true 3D image of the scene, with a sub-millimetric range precision for each point scanned in the range 2-10 meters. In this article we present the integrated software system we have developed in Frascati. This system allows to drive the image acquisition, show the forming image during the acquisition, generate the intermediate structures for the visualisation and render the acquired image in a 2D and/or 3D framework, allowing pan/tilt/zoom on any image portion, applying a series of graphical filters for image enhancement, colour-coded visualisation, saving image portions in standard formats and eventually compare portions of image with a sample data to highlight differences. Moreover the system allows comparing a reference acquired image with any other acquired image. This would be useful studying changes between the 3D image taken as reference respect to the current acquisition, allowing an eventual maintenance task to be carried out.


Corresponding Author:

Riva Marco

Associazione EURATOM-ENEA sulla Fusione, Centro Ricerche Frascati, C.P. 65, 00044, Frascati, via E. Fermi 45

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-60 USING REMOTE PARTICIPATION TOOLS TO IMPROVE COLLABORATIONS

BALME Stéphane, John How, Philippe Lebourg, Jean-Marie Theis, Nadine Utzel

CEA Cadarache DRFC/STEP 13108 Saint Paul Lez Durance Cedex FRANCE

The Tore Supra project, by its very nature, relies on worldwide collaboration for its physics programme. In the past this has been carried out by site visits for experiments, backed up by use of email and telephone. However today, in the “Département de Recherche sur la Fusion Contrôlée” (DRFC), we make substantial use of modern remote participation (RP) tools such that our distant collaborators have a “virtual presence” in our laboratory. This includes audio/video conferencing technologies, that can be used to participate remotely in experiments, applications code development and meetings (H.323 standards, VRVS and multicast protocols, conference-telephones, text messaging, VNC screen sharing coupled with interactive board, etc.), secure remote access to local computers (Citrix solution, SSH, VPN), including code sharing (CVS), remote Matlab access to Tore Supra data via MDS+, remote control of diagnostic instrumentation by experimentalists, broadcast of experiment states and machine parameters screens (TSTV channels), remote monitoring with synchronisation to TS shot events via MDS+ Last year we equipped a meeting room to enable multi-institutional teams of experts to efficiently collaborate in the fusion development effort. This room concentrates many communication and collaboration technologies using low cost but reliable products and open source software. This effort is constantly evolving, following the real needs of our users and has given us valuable information to maximise participation, interaction and collaboration between the researchers for the future generation of Tokamaks. The objective of this current report is to clearly present our installations and experience for working Tore Supra collaborations such that interested persons can learn and profit for their own collaborations. We shall therefore emphasise the practical applications of remote participation, rather than the principles, which are now well known. We expect a very strong participation for Cadarache with ITER and in this perspective we wish to be able to share our practical experience in RP techniques.


Corresponding Author:

BALME Stéphane

CEA Cadarache DRFC/STEP 13108 Saint Paul Lez Durance Cedex FRANCE

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-74 CALORIMETRY MEASUREMENTS DURING HIGH ENERGY DISCHARGES AT TORE SUPRA

CHANTANT Michel, B.Beaumont, P.Bibet, A.Ekedahl, A.Martinez, JC.Vallet

Association Euratom-CEA DRFC CEA/Cadarache F 13108 St Paul Lez Durance

One particularity of Tore Supra is that the Plasma Facing Components (PFCs) are actively cooled, allowing research on long duration and high energy discharges. During a discharge, a large part of the energy is dissipated in the heating generators (ICRH, ECRH and LHH), auxiliaries and transferred to the primary cooling loops (B50 and B60). The energy which is injected in the plasma is totally recovered by the PFCs and transferred to an other primary loop (B30). The primary loops are cooled via heat exchangers by a secondary loop. The previously existing limitations of the complete Tore Supra cooling system were due to the heat exchange performance of the secondary loop cooling towers. Schematically, during a long duration and high energy discharge, only the PFCs primary loop is cooled by the secondary loop, while the hot water from the heating generators primary loops is stored in tanks which were inserted at the outlet leg of the generators. At the end of the discharge, the hot water stored in the tank is cooled when the PFCs primary loop has recovered its initial temperature. In early 2003, the modifications mentioned above were completed and the loops were fully operational. During the 2003 experimental campaign significant results were obtained, particularly during non inductive discharges with an injected energy by the Lower Hybrid Current Drive (LHCD) system of up to 1.1 GJ. During the discharge, the circulation of one part of the B30 flow is monitored in a heat exchanger to achieve a good stability of the water temperature at the tokamak inlet. This condition is particularly important to measure accurately the energy balance of the tokamak by calorimetry. For the long and high energy discharges, the agreement between the energy measured by the calorimetry on the PFCs cooling loop and the HF energy is very good (90 to 95%). The energy radiated by the plasma can be measured from the calorimetry of the panels protecting the vacuum vessel. Furthermore, calorimetry measurements on the B60 cooling loop allow to assess the global LHCD system efficiency (20%), which was not requested to operate at its maximum efficiency in these experiments. The paper will present the main characteristics and operation modes of the Tore Supra cooling loops. It will also give the description of the calorimetry diagnostic and a detailed analysis of the PFCs calorimetry measurements for discharges with injected energies in the range 400 MJ to 1.1GJ.


Corresponding Author:

CHANTANT Michel

Association Euratom-CEA DRFC CEA/Cadarache F13108 St Paul Lez Durance

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-80 REAL-TIME MULTIPLE NETWORKED VIEWER CAPABILITY OF THE DIII-D EC DATA ACQUISITION SYSTEM*

Ponce, D., Y.A. Gorelov

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

A data acquisition system (DAS) which permits real-time viewing by multiple locally networked operators is being implemented for the electron cyclotron (EC) heating and current drive system at DIII-D. The DAS is expected to demonstrate performance equivalent to standalone oscilloscopes. Participation by remote viewers, including throughout the greater DIII-D facility, can also be incorporated. The EC system at DIII-D consists of five 1 MW class gyrotrons. Operational performance is monitored by collecting beam voltage and current, generated rf, and tube pressure waveforms, among other signals. These are currently disseminated in three ways. First, operators can view a subset of the available signals on traditional oscilloscopes. Second, legacy CAMAC digitizers are used to capture the signals and download them to the DIII-D DAS between DIII-D shots. Third, operators can view signals as they are acquired using compact PXI based digitizers. This last set of digitizers is used in the real-time system. The real-time system uses 1 computer controlled DAS per gyrotron. Each computer acquires 8 channels simultaneously sampled at up to 70 kHz per channel. Additionally, the computer acquires and buffers 4 channels of fault signals simultaneously sampled at 15 MHz and 16 channels of coolant temperature and flow used for calorimetry. All signals can be viewed on the acquiring computer. The 8 real-time channels per system will be distributed to multiple remote viewers. Each DAS computer sends its data to a central data server using individual and dedicated 100 Mbps fully duplexed Ethernet connections. The server has a dedicated 10,000 rpm hard drive for each gyrotron DAS. Selected channels can then be reprocessed and distributed to viewers over a standard local area network (LAN). They can also be bridged from the LAN to the internet. Calculations indicate that the hardware will support real-time writing of each channel at full resolution to the server hard drives. The data will be re-sampled for distribution to multiple viewers over the LAN in real-time. Hard drives will hold about a weeks worth of data. Archives will be kept on DIII-D servers and locally on a terabyte DVD/CD changer. The hardware for this system is in place. The software is under development. This paper will present the design details and up-to-date performance metrics of the system. *Work was supported by the U.S. Department of Energy under DE-FC02-04ER54698.


Corresponding Author:

Ponce, D.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-91 EXPERIMENTAL STUDY OF RADIATION-INDUCED CURRENTS IN COPPER AND STAINLESS STEEL CORE MINERAL-INSULATED CABLES IN THE BR2 RESEARCH REACTOR

Vermeeren Ludo,

To monitor the shape and position of the plasma boundary in ITER, in-vessel magnetic coils made of mineral insulated (MI) cables will play a crucial role. In view of the high radiation levels in ITER and the projected long pulse duration, radiation-induced electromotive force (RIEMF) effects (possibly in combination with thermo-electric effects) might jeopardize a proper functioning of magnetic diagnostic devices: currents will be generated between the core wires and the sheaths of the MI cables and spurious voltage differences can arise along the core wires. An experimental study of the RIEMF effect on MI cables has been performed in the framework of the development of magnetic diagnostics for ITER. In this study, six 1 mm diameter MgO insulated cables (with copper and stainless steel cores) were exposed to the combined neutron-gamma field of the BR2 reactor (thermal neutron fluxes up to 3E14 n/(cm²s), fast neutron fluxes up to 2E14 n/(cm²s) and gamma heating rates up to 3.5 W/g). The currents between the inner conductors and the sheaths were recorded as a function of the position of the irradiation rig and compared to theoretical predictions based on a Monte Carlo code. Promptly induced currents and delayed currents due to radioactive decay of various activation products were analyzed separately. Special attention was paid to the dependence of the induced currents upon the environment and even upon the relative position of the cables and the surrounding materials with respect to the radiation source. A systematic trend in this dependence was observed. In accordance with the theoretical predictions, the copper-core cables show significantly higher induced currents than the stainless steel-core cables due to the beta decay of 66Cu formed after neutron capture by 65Cu. Total equilibrium currents for straight 1 mm diameter cables passing once through the 76 cm high BR2 core ranged from 0.06 to 0.1 µA for copper-core cables and from 0 to 0.02 µA for stainless steel-core cables. Future work in this field will concentrate on the RIEMF voltages developing along the MI cables and on the influence of thermal gradients on RIEMF effects.


Corresponding Author:

Vermeeren Ludo

SCK-CEN, Boeretang 200, B-2400 Mol, Belgium

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-98 TORE-SUPRA INFRARED THERMOGRAPHY SYSTEM, A REAL STEADY STATE DIAGNOSTIC.

GUILHEM Dominique, B.Bertrand, C.Desgranges, M.Lipa, P.Messina, M.Missirlian, C.Portafaix, R.Reichle, H.Roche, A.Saille

Association Euratom-CEA, CEA/DSM/DRFC CEA-Cadarache 13108 Saint Paul Lez Durance (France)

Within the framework of the CIEL and CIMES projects, a flat (7m²) toroidal pumped limiter (TPL) and 5 radio-frequency (RF) antennae, all actively cooled, have been implemented in Tore Supra tokamak. The TPL is designed to extract up 15 MW of conducted power with a maximum power density of 10MW/m². From our past experience with such an entirely actively cooled machine [1] we can conclude that the main limitations may come from 1) defects present on elements due to pre-existing cracks or bonding defects, 2) suprathermal particles impacts such as fast ions/electrons trapped in magnetic field ripple, 3) unexpected events like “superbrillance” [2]. Such events may also happen in ITER [3]. We have developed an infrared (4.4 to 4.6 µm) thermographic system (20ms time resolution, 9mm spatial resolution). It is made of a set of 7 endoscope bodies, each equipped with 3 viewing lines : 2 x 30 sectors of the TPL (merged outside of the port and viewed by one IR camera) and 1 viewing line for 1 antenna (viewed by another camera). Each optical line is made of 28 lenses (Ge, ZnS, ZnSe), 4 mirrors and 2 sapphire windows. The lenses are embedded in an actively cooled (25 C) stainless steel tube. The outer jacket of the endoscope body is at the same controlled temperature as the tokamak (~150 C) to prevent it from becoming a cold point under vacuum. This actively cooled jacket is designed to withstand disruptions as well as radiated power deposition and conducted power from electron ripple losses, thanks to an actively cooled CuCrZr protection plate. The sapphire window closest to the plasma is joined via a OFHC copper ferrule to the actively cooled CuCrZr endoscope head, able to withstand in steady state, the radiated plasma power loss (up to 0,2MW/m²) [4]. These sapphire windows (3 per endoscope body) are protected during the conditioning of the machine (glow discharges or boronisation) by a rotating radiative CFC shutter (860 C max for 10MW of radiated plasma power loss). The inner actively cooled tube supporting the lenses can be removed without breaking the inner vessel vacuum. The successful operation since 2002 has shown that the dimensioning had been correctly done. [1] Equipe TORE-SUPRA. Fusion Technology, vol 29, pp. 417-448, 1996 [2] D.Guilhem, A.Seigneur ; JNM 196-198 (1992) 759-764 [3] F.C.Schuller and A.A.Oomens; Fusion Engineering and Design 22 (1993) 35-55 [4] M. Lipa and al , Fusion engineering and design 61-62 (2002) 801-806.


Corresponding Author:

GUILHEM Dominique

Association Euratom-CEA, CEA/DSM/DRFC/SIPP/CEC, CEA-Cadarache, 13108 St PAUL lez DURANCE, FRANCE

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-104 NEW INSTRUMENTS FOR ADVANCED NEUTRON EMISSION SPECTROMETRY DIAGNOSIS OF D AND DT PLASMAS AT JET

Jan Källne, S.Conroy, G.Ericsson, M.Gatu Johnson, L.Giacomelli, W.Glasser, G.Gorini(1), H.Henriksson, A.Hjalmarsson, M.Johansson, J.Källne, S.Popovichev(2), E Ronchi, H.Sjöstrand, J.Sousa(3), ndén Andersson, M.Tardocchi(1), M.Weiszflog, JET-EFDA contributors

(1)INFM, Univ. Milano-Bicocca and Inst. di Fisica del Plasma, ENEA-CNR Assoc., Milan, Italy (2)JET, Culham Science Centre, ABINGDON, UK, UKAEA Assoc. (3) Assoc. IST, Centro de Fusão Nuclear, Inst. Superior Técnico, 1049-001 Lisboa, Portugal.

The magnetic proton recoil (MPR) spectrometer was developed for use on the deuterium-tritium experiments at JET. This demonstrated new capabilities of neutron emission spectroscopy (NES) diagnosis of plasmas far beyond expectations, especially, with regard to aspects of importance for ITER. This success gave the impetus to including two NES projects in the enhancement program at JET, which is reported on here. One NES project involves an upgrade of the MPR (MPRu) to facilitate measurements with higher precision and enhanced suppression of background radiation. The latter will allow diagnostic utilization of the weakest spectral components set by the statistics of the measurement. Similarly, the MPRu operates over the entire fusion neutron energy range (0-20MeV), which allows diagnosing D plasmas; this was not possible with the MPR. The MPRu uses a focal plane detector array based on the phoswich technique, where laminated scintillators of different time response characteristics are used. The other project entails the development of a time-of-flight spectrometer optimized for rate (TOFOR), which, statistics-wise, should match the performance of the MPR for DT plasmas; i.e., operation in the count rate range of hundreds of kHz which is two orders of magnitude faster than previously attained for D plasmas. This is achieved by careful design of the geometries of the scintillation detectors making up the TOFOR instrument; the design is derived from extensive simulations of the neutron response of the detectors and their light emission and transport characteristics. Both spectrometers will be fully calibrated before installation with reference to certain working points. A rudimentary control and monitoring system was used on the MPR and this has now been further developed. This together with a state of the art communication and data acquisition system permits experiments with MPRu and TOFOR to be operated fully electronically and monitored for stability over both short (transient) and long time periods, allowing remote examination. Time digitizers and transient recorders based on PC cards are being developed for the first time for NES diagnostic applications. This contribution will report on the principles of the most advanced neutron spectrometer diagnostics built for JET and their use to test reliability, machine interface issues and diagnostic capabilities in fusion experiments that mimics ITER as close as todays’ generation of tokamaks permits.


Corresponding Author:

Jan Källne

Dept. Neutron Research, Uppsala University, Box 525, SE-75120 UPPSALA, SWEDEN

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-119 SURFACE DIAGNOSTICS WITH APPLICATION OF VIDEOSCOPE ON THE BASIS OF CU-LASER

Buzhinskij Oleg, Otroshchenko Vladimir (1)

(1) State Scientific Center Troitsk Institute for Innovation and Thermonuclear Researches, 142190 Troitsk Moscow reg. Russia

Research of materials surface in extremely stressed conditions under effect of plasma, welding, melting, radiative radiation represents the great scientific interest, but has a series of formidable experimental restrictions. Use of light amplification effect in a visible range in active optical medium allows successfully to overcome them, in particular, in conditions of intensive plasma and radiative influence. For this goal it is possible to apply a laser with high amplification coefficient,working practically in a single-pass mode. At the illumination of a researched surface area the laser peak intensity should to exceed significantly a background radiation in the given solid angle. For obtaining of the observed surface image it is necessary to collect a laser radiation scattered by a surface area, selectively to amplify it and to focus on the receiving part of a recording device. If laser works in a pulsed operation mode it is possible to receive in real time a spatially temporal surface image. In the work a spatially temporal diagnostics for surface research using videoscope on the basis of Cu-laser is presented. The typical sizes of a researched surface area essential to the given method are determined. The optical scheme of diagnostic device is represented. The description of device block diagram is given. The limiting spatially temporal resolution is determined. Taking into account of optical properties of researched medium the application of chosen observation method is substantiated. The results of experimental researches of the videoscoping system are submitted. In experiments the copper vapor laser with characteristics was applied: -visible radiation at green (510.6 nm) and yellow (578.2 nm) wavelengths; -pulse repetition rate - 12 kHz; -average radiation power - 4 W; -generation pulse duration - 15 ns; -peak radiation power - 80 kW; -laser beam diameter - 12 mm; -laser tube length - 500 mm; -laser radiation divergence - 1 mrad.


Corresponding Author:

Buzhinskij Oleg

State Scientific Center Troitsk Institute for Innovation and Thermonuclear Researches, 142190 Troitsk Moscow. reg. Russia

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-120 NEW MANAGING SYSTEM OF A LARGE AMOUNT OF IMAGES ON TORE SUPRA

BURAVAND Yves, L. Ducobu, D.Guilhem, H.Roche, J.M.Travere, N.Utzel

CEA Cadarache DRFC-STEP BP n 1 13108 Saint-Paul Lez Durance France

Poster presentation After a major replacement of all internal components (terminated in 2002), Tore Supra has re-started operation with a new set of actively cooled components. The safety of these components relies on infrared (IR) thermography. Up to twelve IR cameras will be used to survey nearly all the high flux target zones which are essentially the 360 toroidal pump limiter and the 5 power injection HF antennas. For dealing with the large amount of data produced (~ 8Mbytes/s/camera) a two stages system has been implemented. At the first stage, IR data are acquired and stored in separate units (Industrial PCs) that can serve two cameras. Between shots, a few PCs, running a dedicated software (ShotPlayer), have a fast access to these data for analyses purposes. All these workstations are linked by a dedicated gigabit network. Each night, data stored on the acquisition units are compressed using a free JPEG library. Compression parameters are adjusted in order to have an acceptable degree of loss of information and a rather good compression ratio (~10). These compressed images are then moved to a dedicated IR data server but the original films are kept here for 2 weeks before being deleted, except if required by a physicist to keep the raw data. The IR server has an extensible 700 Go mass storage space and runs two software servers, one dedicated to the ShotPlayer clients, and another one for in-depth analysis via the Matlab® program. Physicists can access raw compressed images through a multithreading, overlapped I/O server, decompress and convert them in absolute temperature on their workstations using functions we have developed and integrated into the Matlab® environment. Acquisition units, network, and the ShotPlayer workstations clients were tested during the 2001 Tore Supra campaign with only one IR camera. Because of its modularity, the acquisition system has easily evolved during the 2002-2003 experiments; 8 cameras will be in operation in 2004. Also, the whole long term storage and retrieval system is now validated. Techniques and methods used to realize this imaging diagnostic give us invaluable information that will also be valuable for defining future generation of imaging data acquisition systems.


Corresponding Author:

BURAVAND Yves

CEA Cadarache DRFC-STEP BP n 1 13108 Saint-Paul Lez Durance France

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-132 RADIATION RESISTANT BOLOMETERS USING PLATINUM ON AL2O3 AND ALN

M. Gonzalez, R. Vila, and E.R. Hodgson

Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain

Present day JET type bolometers using gold meanders on a mica substrate have been considered for use in ITER. However in-reactor tests at JMTR (Japan) demonstrated that they suffer radiation damage due to transmutation of gold into mercury, and detachment of the meander from the substrate possibly related to this alloy change and/or substrate swelling. The formation of a gold / mercury alloy caused a dose dependent increase in the bolometer meander resistance, and the detachment of the meanders was accompanied by loss of electrical continuity. As a result work was undertaken to examine alternative more radiation resistant substrates, together with the substitution of evaporated gold by sputtered platinum. From these initial studies already reported, commercially available sheets of alumina and AlN were selected for further testing. The radiation resistance of these prototype bolometers with sputtered platinum meanders is now being characterised. As a first step their behaviour as a function of temperature during electron irradiation has been examined, before neutron irradiation tests are performed. The results for the meander resistance as a function of thermal cycling and radiation dose are considered to be satisfactory. In parallel a detailed study is being carried out to compare evaporated versus sputter deposited platinum on the ceramic substrates to minimise detachment problems. An analysis on the basis of the quality of the platinum sensor adhesion and the metal-substrate interface features is performed to optimise the bonding in an attempt to obtain bolometers of improved radiation resistance. The role of glow discharge atmosphere, platinum source, and other relevant process parameters will be presented and discussed. In addition to compare with the original gold on mica bolometers, platinum on mica has been prepared and tested for comparison.


Corresponding Author:

M. Gonzalez

Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-142 TEMPERATURE DEPENDENCE OF THE TRANSMISSION LOSS IN KU-1 AND KS-4V QUARTZ GLASSES FOR THE ITER DIAGNOSTIC WINDOW

NISHITANI Takeo, SUGIE Tasuo, MORISHITA Norio, YOKOO Noriko

In the ITER diagnostics, optical windows will be installed at the end of diagnostic ports, where the ionizing dose is expected to be less than several MGy for the ITER lifetime. Quartz glasses of KU-1 and KS-4V are candidate window materials for the ITER optical diagnostics in UV and visible range. KU-1 and KS-4V have been developed as radiation resistant quartz glasses in Russia. KU-1 is characterized by a high OH content. On the other hand, KS-4V is low OH and low chlorine contents. Quartz glasses have relatively large sensitivity in UV range for radiation. The temperature dependence of the radiation induced transmission losses in UV range has been investigated under gamma-ray irradiation for the KU-1 and KS-4V quartz glasses up to 10 MGy. The Co-60 irradiation facility of JAERI/Takasaki was used for this irradiation test. The KU-1 and KS-4V quartz glass samples with the size of 16 mm in diameter and 8 mm in thickness were irradiated at room temperature, 100, 200 and 300 C. Dose rate was 0.46 Gy/s until 1.1 MGy, and 2.6 Gy/s from 1.1 MGy to 10.1 MGy. The transmission spectra of the windows were measured with a spectrometer once a week during irradiation pause. It was confirmed that there are no significant loss in wavelength range longer than 350 nm for both window materials. Transmission loss in KU-1 is larger than that in KS-4V under temperature below 200 C. KU-1 has large temperature dependence. On the other hand, temperature dependence is not clear in KS-4V above 100 C. At 300 C, transmission loss in KU-1 is smaller than that in KS-4V. Prominent recover of the transmission loss was not observed after irradiation for each sample. In the transmission spectrum of KU-1, two absorption peaks were identified; one was E’-center at 215 nm and the other is from the non-bridging oxygen hole center (NBOHC) at 260 nm, which is common feature of the quartz glass for the gamma-ray irradiation. In KS-4V, absorption of NBOHC is not clear, but the absorption peak from oxygen deficient center was observed at 245 nm, which suggests that the defect production mechanism is different in KU-1 and KS-4V. From the transmission loss point of view, KS-4V is better window material than KU-1 at temperature below 200 C. KU-1 is available at 300 C.


Corresponding Author:

NISHITANI Takeo

JAERI, 2-4 Shirakata-shirane,Tokai-mura, Naka-gun, Ibaraki 319-1114, JAPAN

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-159 THERMAL AND NEUTRON TESTS OF MULTILAYERED DIELECTRIC MIRRORS

Ilia Orlovskiy, Konstantin Vukolov

Optical diagnostics for ITER will require mirrors for transmission of radiation from plasma to detectors. Apparently, plasma-facing (first) mirrors should be metallic because they will be exposed to significant fluxes of fast neutrons and charge exchange atoms (CXA). At the same time, secondary mirrors are expected to take less neutron fluxes and no CXA so possibly they can be made of material with higher reflectance. One way to improve the transmittance is to use multilayered dielectric mirrors as the secondary mirrors. Multilayered dielectric mirrors are made up of alternating layers of materials of different refraction and their reflectance reaches 100% in a certain range of wavelengths. Then there is an open question of capability of such mirrors for operating under ITER conditions which are neutron fluxes up to 10E12 n/cm2s and temperatures of 150C to 200C. The goal of our experiment is to examine some samples of dielectric mirrors for their resistance to noted temperatures and neutron fluences to be accumulated in ITER for several years and to draw a conclusion of a possibility of applying of dielectric mirrors to optical diagnostics of ITER. Visible-range mirrors made from TiO2/SiO2 and ZrO2/SiO2 layers applied onto SiO2 substrates by two manufacturers were irradiated to fluences of 10E17 n/cm2 at about 50C in nuclear reactor. The number of layers varied from 13 to 23. Neutron fluence corresponds to a fluence to be accumulated by the fourth mirror of H-alpha spectroscopy diagnostics for 1 year of ITER operation. The neutron test was preceded by thermal tests in which the mirrors were heated up to 280C. Some mirrors retained their reflectance and structure integrity under heating while others did not that can be caused by insufficient layers adhesion to mirrors substrate. All the samples did not change significantly their reflectance under irradiation. The results obtained confirm resistance of dielectric mirrors to neutron irradiation to high fluence. Although the resistance of mirrors to thermal load is limited by the strength of adhesion, it is clear that sufficient adhesion can be provided by manufacturers. In addition, in spring 2004 we are going to perform the second neutron test where the same samples will be irradiated to a neutron fluence of 10E19 n/cm2. The results will be included in the report.


Corresponding Author:

Ilia Orlovskiy

Nuclear Fusion Institute, Russian Research Center "Kurchatov Institute", 123182 Kurchatov Sq. 1, Moscow, Russia

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-166 LASER DAMAGE INVESTIGATIONS OF CU MIRRORS

Alexey Gorshkov, Igor Bel’bas Mikhail Maslov Vladimir Sannikov Konstantin Vukolov

RRC "Kurchatov Institute", Kurchatov sq.1, 123182 Moscow, Russia

Laser tests were performed on Cu mirrors as “first mirror“ prototypes for laser diagnostics for ITER. Data on laser damage thresholds under the influence of frequency-operated pulsed YAG laser (1064 nm wavelength, 12,5 Hz repetition rate, 10-30 mJ per pulse energy, 26 ns pulse duration) were obtained for single laser shot and after about 1.5x10^5 laser shots. The output beam of the laser operating in the TEMoo mode has a Gaussian profile. The experiments were carried out with using three types of the mirrors: one – diamond-turned Cu mirror, second – diamond-turned substrate with Cu coating and third - reflection grating on the Cu coating mirror. The single shot damage thresholds were measured and were equal to 27±5.5 J/cm^2 for diamond-turned mirror, 18.7 ± 3.7 J/cm^2 for Cu coated mirror and 12.3 ± 2.5 J/cm^2 for the Cu grating. The lifetime of the Cu mirror under multiple pulse laser irradiation was studied. Diffusion scattering was used as a monitor of mirror surface condition. The degradation of copper mirrors under multiple pulse laser irradiation is described with satisfactory accuracy by a predictive model for multipulse laser damage of metal mirrors up to 1.5x10^5 laser pulses. This model relates multiple-pulse damage that accumulated on metal surfaces to the thermal stress field induced by the laser pulse. Keywords: Optical diagnostic of the plasma; Cu mirror; Laser; Laser damage threshold; Diffusion scattering.


Corresponding Author:

Alexey Gorshkov

RRC

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-194 DESIGN OF LOST ALPHA PARTICLE DIAGNOSTICS FOR JET*

Baeumel Stefan, A. Werner(1), R. Semler(1), S. Mukherjee(1), D.S. Darrow(2), R. Ellis(2), F.E. Cecil(3), V. Kiptily(4), L. Pedrick(4), J. Gafert(4) and JET-EFDA contributors(4)

(2)Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543-0451, USA (3)Colorado School of Mines, 1500 Illinois St., Golden, CO 80401, USA (4)EFDA-JET, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK

Two diagnostics for the measurement of lost alpha particles are being fabricated for the Joint European Torus (JET). These diagnostics consist of a scintillator based probe [1,2] near the midplane and a poloidal array of five sets of thin foil Faraday collectors, which replaces the existing Kalpha-1 diagnostic [3,4]. Both systems are capable of measuring fast ions, fusion products and ICH tail ions. The Faraday cup array will measure the particle loss at multiple locations at a rate of 1 kHz, while the scintillator probe will be capable of measuring the pitch angle and energy distribution of the lost ions with a time resolution of 2 kHz. The dynamic range of the Faraday cups, which consist of a stack of 2.5 microm thin Ni foils, will be from 1 nA/cm2 to 100 microA/cm2. The dynamic range for current measurements with the scintillator probe will be from 10pA/cm2 to 1 microA/cm2. Not all Faraday cups will have the same energy resolution. It will range from 15-25% for 3.5 MeV alpha particles. The scintillator probe is designed to detect fast ions with pitch angles up to 86 and provides a pitch angle resolution of about 5 degrees. The gyroradius resolution of the scintillator probe will be about 15%. The scintillator material will be P56 (Y2O3:Eu) will be used, which is luminous up to about 400 C. In order to maximize the signal to noise ratio, the detectors of both diagnostics are located as close to the face of the poloidal limiter structure as feasible (5 mm). Therefore both diagnostics protrude quite far into the vessel. Due to Halo- and Eddy currents that lead by the interaction with the background magnetic field to enormous forces onto the diagnostics, considerations of structural stability were a major concern of the design. Since the probe is located beside a neutral beam injector that deposits a significant heat load of 13 MW/m2 on the side of the probe, it will be actively cooled. Experience in operating both diagnostics in a high temperature and high radiation environment will be valuable in preparation for the design of similar diagnostics for future fusion devices. 1. S.J. Zweben et al. Nucl. Fusion 30, 1551 (1990). 2. A. Werner et al. Rev. Sci. Instruments 72, 780 (2001) 3. F.E. Cecil et al. Rev. Sci. Instruments 70, 1149(1999). 4. O.N. Jarvis et al. Fusion Technology 39, 84 (2001) *This work is supported by U.S. Department of Energy Contracts DE-AC02-76CH03073 and DE-FG03-95ER54303 and conducted under EFDA by IPP, PPPL and CSM.


Corresponding Author:

Baeumel Stefan

Max-Planck-Institut für Plasmaphysik, Wendelsteinstr. 1, 17491 Greifswald, Germany

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-202 DEVELOPMENT OF THE PHASE COUNTER WITH THE REAL-TIME FRINGE JUMP CORRECTOR FOR INTERFEROMETER ON LHD

Yasuhiko. Ito, Kenji. Tanaka, Tokihiko. Tokuzawa, Tsuyoshi. Akiyama, Shigeki. Okajima and Kazuo. Kawahata

Dept. of Eng. & Tech. Servises National Institute for Fusion Science, Oroshi-cho, Toki, GIFU, 509-5292, Japan

In a laser heterodyne interferometer system, a phase counter is used to measure plasma electron density from the beat signals of interferometer outputs. One of the important problems in multi-fringe phase detection is the fringe jump error. The error is caused by decreased signal to noise ratio of the beat signal, when the electric noise is increased or the signal amplitude is decreased due to refraction of the probe beam. The fringe jump occurs also due to the spike-like stray pick-up when the discharge of ion source of neutral beam injector (NBI) breaks. The fringe jump causes severe problem for the data analysis and operation of density feedback. We developed the circuit, which called the AFJC (Automatic Fringe Jump Corrector), to compensate the fringe jump automatically. The AFJC circuit was developed to improve fringe jump correction work. The functions of the circuit are fringe jump detection and change incorrect fringe number of the phase counter to the correct fringe number stored before the fringe jump. The circuit can process fringe jump correction within 3us. The circuit is integrated on a CPLD (Complex Programmable Logic Device), which added to the conventional phase detection circuit with the following specifications: 1MHz input beat frequency, 31 fringes phase detection range and 10us phase resolution. The circuit was installed on the FIR laser interferometer and was tested in the electron density measurements of LHD plasma. As a result, the fringe jumps caused by the H3 pellet injection and the radiation collapse can be corrected, the probabilities of correction were approximately 70% and 20% respectively. However, the fringe jump caused by high-density plasma measurement, still can not be compensated. In order to correct the fringe jump due to the discharge of NBI, the AFJC circuit is improving to use the high phase resolution (1/1000 fringe) phase counter for the CO2 laser interferometer on LHD as well.


Corresponding Author:

Yasuhiko. Ito

Oroshi-cho, Toki, GIFU, 509-5292, Japan

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-208 DATA ACQUISITION UPGRADE IN THE RFX EXPERIMENT

Manduchi Gabriele, Barana Oliviero, Luchetta Adriano, Taliercio Cesare

The Control and Data Acquisition system of RFX has been completely renewed and is currently under re-commissioning. Most data acquisition is now performed by means of CompactPCI devices supervised by Pentium PCs running Linux. Real-time control systems, implemented using VME and PowerPCs running VxWorks, produce also data that are then acquired by the data acquisition system. The older CAMAC systems have only been retained for existing diagnostics. New diagnostic systems will employ either CompactPCI data acquisition or custom solutions, usually running under Windows, due to the fact that the drivers for the used devices are normally available for this platform. Despite the variety of hardware and software platforms involved in data acquisition, the same software package is used for all components, thus providing a uniform view of the system. Such functionality is provided by the MDSplus software. MDSplus is now available for a variety of platforms, and includes several Java components that are platform-independent. While data organization is mostly centralized in RFX, i.e. the pulse database components are located on at most two machines, the control and data acquisition tasks are distributed, being carried out by all the supervisory CPUs of the CompactPCI crates, as well as by the CPUs involved in the PC-based diagnostics. This means that during the shot sequence tens of different tasks running on different machines need to synchronize. This is achieved by a dedicated Java component of MDSplus, which takes the required dispatching information from the pulse file. Communication among system components is achieved using mdsip, a protocol built over TCP/IP, which provides the “glue” that is tying MDSplus components together. A typical operating scenario for RFX involves 15 CompactPCI crates (with 14 slots), 7 to 9 VME crates, and 2 to 5 Windows PC for diagnostic supervision. Moreover, 15 to 20 Windows PCs are used for waveform display and graphical user interface. Despite the large number of components, the system proved very reliable. The new configuration has in fact been used in a set of tests for an ITER component, producing more than one thousand of shots. This is mainly due to the robustness of the communication layer and, more generally, to the high quality of the MDSPlus tools, a consequence of the fact that the MDSplus architecture relies on the experience gained in years of operation on several different experiments.


Corresponding Author:

Manduchi Gabriele

corso stati Uniti, 4 35127 Padova - Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-220 THERMAL DETECTOR FOR THE LOST ALPHA PARTICLE MEASUREMENTS

Andrey Alekseyev, I.N. Rastjagaev 1 L.N. Butvina 2 V.A. Dravin 3

1- TRINITI, Troitsk, Moscow reg. 142190 Russia 2- General Physics Institute of the Russian Academy of Science, 38 Vavilov st. Moscow, 117942 Russia 3- Lebedev Physical Institute of the Russian Academy of Science, 53 Leninsky pr. Moscow 119991 Russia

Recently a multi-foil thermal detector (MFTD) has been proposed for the lost alpha particle spectra measurements1. The general idea is to measure the power deposited by a particle flux in a number of successive ultra-thin foils. Detailed analysis shows that approximately 21 foils in the stack are required to obtain the desirable 250 keV resolution in the total 0…3.5 MeV energy range2. In the current work we have tested more reliable and simplified design, which comprises a number of thermal bolometers with proper front-end absorbing foil windows providing different low-energy cut-off limits for the incident particles. The approach is similar to the well-known soft X-ray filter technique for the measurement of plasma electron temperature. Following it, an effective temperature or a “color” of the lost alpha particles could be measured even with a limited number of the detectors. An infrared fiber optic bolometer was developed for the thermal sensing of the incident power. Free-standing diamond-like carbon (DLC) and chemical vapor deposited (CVD) diamond films were used for the front-end filters. The prototype device was tested and calibrated at the He ion accelerator in 30…700 keV energy range. 1. A.G. Alekseyev, D.S. Darrow, et al., "Nanoscale Thickness CVD Diamond Membrane Detector for Energetic Particles Spectra and Profile Measurements", Proc. of German-Polish Conf. on Plasma Diagnostics for Fusion & Applications, Greifswald, Germany, Sept. 4-6, 2002, B11. 2. A.G. Alekseyev, D.V. Portnov, F.E. Cecil “Multi-foil Stack Thermal Detector for Lost Alpha Particle Measurements”, 30th EPS Conf. on Controlled Fusion & Plasma Phys., July 7-11, 2003, St.Petersburg, Russia.


Corresponding Author:

Andrey Alekseyev

TRINITI, Troitsk, Moscow reg. 142190 Russia

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-226 ITER RELEVANT DEVELOPMENTS OF NEUTRON DIAGNOSTICS DURING JET TRACE TRITIUM CAMPAIGN

BERTALOT Luciano, J.M. Adams2, M. Angelone1, S. Conroy3, B. Esposito1, Y. Kaschuck4, D. Marocco1, A. Murari5, M. Pillon1, S. Popovichev2, M. Reginatto6, M.Riva1, H. Schuhmacher6, D. Stork2,A. Zimbal6 and the JET EFDA contributors**

1Ass. Euratom-ENEA, Frascati, Italy 2Ass. Euratom-UKAEA ,Culham, UK 3EURATOM-VR Ass., Uppsala, Sweden 4TRINITI, Troitsk, Russian Federation 5Consorzio RFX – Ass. Euratom, Padova, Italy 6Physikalisch-Technische Bundesanstalt,Braunschweig, Germany

During the JET Trace Tritium campaign a few new neutron diagnostic systems were deployed under different plasma scenarios to provide information on the total neutron emission and its spatial and energy distribution. The 14 MeV neutron yield was measured with one Chemical Vapour Deposited (CVD) diamond detector, which is more resilient to neutron damage. Comparison with the JET 14 MeV monitors (Si diodes) illustrates the good performance of the CVD device and provides an excellent correlation between the two 14 MeV yield measurement systems. Key information on tritium transport and the behaviour of fast particles in the plasma were obtained from the spatially and temporally measurements of neutron emission by means of the Upgraded Neutron Profile Cameras which also provide an independent measure of the total yield. Particular attention was paid to the operational stability of this diagnostic system. Spatial asymmetries in the neutron emission were observed which is evidence for the influence of fast particles on the plasma. With regard to the energy distribution of the neutron emission, a neutron spectrometer was installed based on a liquid organic scintillator and n-g pulse shape discrimination. The results show that such systems can operate in real fusion experiments as compact broadband neutron (from 1.5 MeV up to 20 MeV) and gamma ( from 0.3 MeV up to 10 MeV) spectrometers with good energy resolution. Application of the digital pulse shape discrimination (DPSD) technique with fast transient recorder cards has been successfully carried out. The main advantages of DPSD are in enabling high count rate operation into the MHz range, and in the potential for post-experiment data (re)processing. This new technique can be used for neutron emission counting as well for simultaneous neutron and gamma spectroscopy The experience gained at JET indicates that these neutron measurement systems are suitable for large fusion devices such as JET-EP and ITER where fusion neutron diagnostics will play an increasingly important role. **See annex of J. Pamela et al., Fusion Energy 2002 (Proc. 19th Int. Conf. Lyon, 2002), IAEA, Vienna


Corresponding Author:

BERTALOT Luciano

Associazione Euratom-ENEA sulla Fusione, C.R. Frascati, C.P. 65, Frascati, I-00044, Roma, Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-230 OPTICAL AND ELECTRICAL DEGRADATION OF HYDROGEN IMPLANTED KS-4V QUARTZ GLASS

S.M. González, A. Moroño, and E.R. Hodgson

Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain

The light detection and ranging (LIDAR) diagnostic systems for ITER will employ high power lasers pulses which must pass through highly transparent windows. If the laser intensity is too intense the window material will suffer dielectric breakdown and as a consequence the window may break. The window material will be subjected to neutron and gamma irradiation, and additionally the vacuum face of the window may be subjected to a bombardment by low energy (eV to keV) ions and neutral particles. These low energy particles will deposit most of their energy at or very near the surface and hence the local damage and/or degradation of the physical properties at the vacuum surface could be very high. In particular degradation of the optical transmission and electrical resistivity of windows are important issues for LIDAR diagnostic systems. One candidate material for windows is KS-4V quartz glass. In the work to be presented the optical and electrical degradation of this material implanted with protons has been addressed. KS-4V 16 mm diameter, 1 mm thick disc samples, were implanted at 25 C with hydrogen ions (protons) of energies between 30 and 55 keV, 1 microamp/cm2, up to a dose of 10+16 ions/cm2. After implantation the optical absorption from 195 to 3000 nm was measured, and then the samples were mounted in a system which permitted one to measure the surface and volume electrical conductivities in high vacuum (10-6 torr) at temperatures between 20 and 450 C. Proton implantation of KS-4V quartz glass modified the optical transmission, producing a monotonic increase in the absorption extending from the IR to the UV region. The measured absorption at 800 nm implies that 30 % of the laser power would be absorbed in a surface region of about 1 µm. In addition the electrical conductivity of the material over the implanted area severely increases by many orders of magnitude. These two effects dramatically enhance the risk of laser breakdown if the material is to be used as a LIDAR window. The enhanced surface electrical conductivity increases with increasing temperature, and the observed behaviour suggests the formation of a semiconducting material layer. This type of surface electrical degradation should be studied in other insulators to be used in ITER not only in diagnostics but also in heating systems where electrical insulation is of concern.


Corresponding Author:

S.M. González

Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-234 RECONSTRUCTION CAPABILITY OF JET MAGNETIC SENSORS

Cenedese Angelo, Raffaele Albanese (1) Giovanni Artaserse (1) Massimiliano Mattei (1) Filippo Sartori (2)

(1) Assoc. EURATOM-ENEA-CREATE, DIMET, Univ. Mediterranea di Reggio Calabria, Via Graziella, Loc. Feo di Vito, I-89060 Reggio Calabria, Italy (2) Euratom/UKAEA Fusion Assoc., Culham Science Centre, Abingdon, Oxon, OX14 3DB, UK

During the 2004 shutdown, a new set of magnetic sensors will be installed in JET, designed so as to upgrade and in part to replace the existing diagnostic system. For this reason, as far as the reconstruction capability is involved, it has been fundamental for the sensor definition to quantify beforehand how much the magnetic enhancement will -in principle- augment the measurability of the plasma shape, and therefore extend the JET operating space. To discern between an increased redundancy in the measure and new information brought in by the sensors, a model based statistical analysis resorting to the correlation function among the magnetic measurements has been carried out. To perform this study, experimental and simulated databases have been constructed, spanning the whole variety of already achievable plasmas and the designed new JET-EP configurations for 2005. In addition, the field reconstruction error in proximity to the plasma boundary has been assessed using the present and the augmented sensor sets, which gives indication on the achievable accuracy of the plasma boundary itself. As a matter of fact, besides representing a cross-validating and consistent study on JET magnetics, both these analyses seem to confirm that an enhanced reconstruction can be obtained, with a noise amplitude reduction in localised parts of the boundary (up to 50% on the outboard). The methodology sets also the guidelines for the development of software tools useful for the experimental commissioning and measure validation of the sensors.


Corresponding Author:

Cenedese Angelo

Consorzio RFX Assoc. EURATOM ENEA sulla Fusione, Corso Stati Uniti, 4, I-35127 Padova Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-235 QUENCH DETECTION & DATA ACQUISITION SYSTEM FOR SST-1 SUPERCONDUCTING MAGNETS

A. N. SHARMA, C.J. Hansalia, Y. Yeole, G. Bansal, S. Pradhan, & Y.C. Saxena

INSTITUTE FOR PLASMA RESEARCH, GANDHINAGAR

Superconducting magnet system of Steady State Superconducting Tokamak (SST-1) shall be operating in a very noisy environment. Presence of high inductive voltages in the magnets during off-normal events like VDE, plasma disruption, and PF magnet ramp ups etc, has made quench detection and data acquisition a challenging task. A hybrid of analog electronic circuits and software controlled data acquisition system has been developed and tested to safeguard the magnets. This paper will describe the electronic hardware circuits developed for signal conditioning, high voltage suppression, fail proof quench detection and for noise elimination algorithms and their testing. The SST-1 Superconducting magnets will have large number of sensors like voltage taps, Venturi flow meter, strain gages, hall probes, pressure sensors, temperature sensors, and displacement transducers. A real time data acquisition system has been designed using VMEbus for monitoring and storing signals from all these sensors and initiating control action in case of off-normal events. The paper will also describe the configuration of the data acquisition system with emphasis on hardware used and the software developed for it.


Corresponding Author:

A. N. SHARMA

INSTITUTE FOR PLASMA RESEARCH, BHAT, GANDHINAGAR - 382428, (GUJARAT) INDIA

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-248 LASER DAMAGE OF KU-1 SILICA GLASS COVERED WITH HYDROCARBON FILM

Konstantin Vukolov, Dorian Orlinski (1) Alexander Gorbunov (2) Nikolay Klassen (2)

(1)RRC Kurchatov Institute, Kurchatov Sq.46, 123182 Moscow, Russia (2)Institute of Solide State Physics, 142432 Chernogolovka, Moscow region, Russia

Results are given of the experimental investigation of a laser damage of diagnostic windows covered with hydrocarbon films under the pulse laser irradiation (YAG laser, t=10 ns, f=7-33 Hz and an energy up to 0.4 J/pulse). The optically polished samples (diameters of 10 and 16 mm, thickness 10 mm) of KU-1 silica glass which is planned to be used in ITER were tested. One side of the samples was covered with thin hydrocarbon films (CH and CD with different content of deuterium) of different thickness (10, 50 and 100 nm). Laser beam was introduced through the clean side of widow sample and was focused on the filmed surface with a spot of 0.2 mm in diameter for laser damage threshold measurements and of ~1 mm in diameter for measurements of hydrocarbon films ablation (cleaning threshold). The experiments were done in atmosphere. The cleaning threshold was determined by monitoring of film surface in scattered or reflected light and laser damage threshold – on characteristic click and blue light appearance. Measured cleaning thresholds of hydrocarbon films were 0.1-0.3 J/cm2 depending on the deuterium content and to a lesser degree on the film thickness. The laser damage thresholds were about of 50-60 J/cm2 for KU-1 windows with films on the surface and about of 80 J/cm2 for clean windows. An error of these measurements was 15-20%. Probably same laser damage thresholds for the samples with and without films mean that the film evaporation takes place during the first laser pulse. The expected interaction of film material and glass was not observed. However graphite erosion products will be always presented into ITER vessel and the window surface may be contaminated with hydrocarbons penetration into the glass under laser radiation. This year it is supposed to repeat the experiments in a vacuum chamber with simulation of the ITER vacuum conditions.


Corresponding Author:

Konstantin Vukolov

RRC Kurchatov Institute, Kurchatov Sq.46, 123182 Moscow, Russia

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-250 WIDE AREA DATA REPLICATION IN AN ITER-RELEVANT DATA ENVIRONMENT

Iannone Francesco, Giovanni Bracco (1) Cristina Centioli (2) Silvia Eccher (3) Maurizio Panella (2) Maurizio Steffe' (4) Vincenzo Vitale (2)

(1) ENEA-INFO, Via Enrico Fermi 45, 00044 Frascati (RM) Italy (2) same as main author (3) CASPUR Via dei Tizii, 6b 00185 Roma Italy (4) ENEA - Centro ricerche della Casaccia Via Anguillarese, 301 00060 S. Maria di Galeria (RM) Italy

The next generation of tokamak experiments will require a new way of approaching data sharing issues among fusion organizations. In the fusion community, many researchers at the different sites worldwide could analyze data produced by ITER wherever it will be built. Usually in such large size experiments an approach is the efficient availability of the data near to the location where the computational resources are available. This new approach should go beyond the site-centric model mainly devoted to granting access exclusively to experimental data retained in the device sites. To this aim, we propose a new data replication architecture relying on a wide area network, based on master/slave model and synchronization techniques producing mirrored data sites. In this architecture, data replication will affect large databases (TB) as well as large unix-like file systems, using open source based software components, namely MySQL as database management system, and rsync and bbftp for data transfer. A testbed has been set up to evaluate the performance of the software components underlying the proposed architecture. The testbed hardware layout deploies a cluster of four Dual-Xeon Supermicro each with a raid array of 1 Terabytes. High performance network lines (1 Gbit over 400 Km) will provide to test the components on wide area network. The results obtained will be thoroughly discussed.


Corresponding Author:

Iannone Francesco

ENEA - Centro Ricerche Frascati, Via Enrico Fermi, 45 I-00044 Frascati (RM) Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-251 ADVANCES IN REMOTE PARTICIPATION FOR FUSION EXPERIMENTS*

Schissel, D.P., V. Schmidt (1), J.W. Farthing (2)

(1) Consorzio RFX, Associazione Euratom-ENEA sulla fusione, Padova, Italy (2) Association Euratom-UKAEA, Culham Science Centre, Abingdon, United Kingdom

Selected also for oral presentation O2B-D-251

Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in collaborations between experimental sites and laboratories worldwide. In the US, fusion experimental research is centered on three large facilities involving over 1000 researchers covering 37 states. In the EU, fusion research is coordinated by EFDA, encompassing some 25 laboratories and several major facilities, including JET. Collaborative research within each group, combined with collaboration between the two groups is presenting new and unique challenges in the field of remote participation technology. These challenges are being addressed by the creation and deployment of advanced collaborative software and hardware tools. Grid computing, the secure integration of computer systems over high-speed networks to provide on-demand access to data analysis capabilities and related functions, is being deployed as an alternative to traditional resource sharing among institutions. Utilizing public-key based security that is recognized between the EU and US, the TRANSP analysis code is running on one cluster yet is securely available worldwide. This analysis service includes secure remote data access as well as advanced web-based monitoring capabilities. Traditional audio teleconferencing is being augmented by more advanced capabilities including videoconferencing, instant messaging, presentation sharing, applications sharing, tiled display walls, and the virtual-presence capabilities of the AccessGrid, with its potential for remote control room presence. With these advances, remote real-time experimental participation has begun including several remotely led JET experiment sessions where the lead scientists were in other laboratories in the EU or the US. Work continues to focus on reducing the variety of remote participation methods, on improving interoperability between the different approaches, on ease of use, and on improved security. The collaborative technology being deployed is scalable to fusion research beyond the present programs, in particular to the ITER experiment that will require extensive collaboration capability worldwide. This paper will compare approaches, review the present state-of-the-art in remote participation capability, and identify areas of work required for the success of future large-scale experiments. *Work supported by U.S. Department of Energy under DE-FC02-01ER25455 and by the European Fusion Development Agreement (EFDA)


Corresponding Author:

Schissel, D.P.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-258 SOFT COMPUTING AND CHAOS TEORY FOR ANTICIPATION OF DISRUPTION IN TOKAMAK REACTORS

Francesco Carlo Morabito, Domenico Costantino Mario Versaci

Università di Reggio Calabria. Facoltà di Ingegneria - DIMET

Disruption represents a transfer of energy of the plasma to the surrounding mechanical structures. During the sudden loss of confinement, the energy content collapse in an uncontrollable way, generating mechanical forces and heat loads which threaten the structural integrity of surrounding structures and vacuum vessel components. It is thus of primary importance to design an alarm system for detecting he onset of a disruption. Neural Network models have been proposed in the recent literature as forecasting systems, with the aim of predicting the occurrence of disruptions sufficiently far in advance for protecting procedures to be switched on. Then, it is necessary to check the incoming of disruption by means of a suitable time window of prediction to take into account control actions. The system is subjected to the fusion reactions and it is really complex according to Chaos Theory. In this paper, chaotic analysis and soft computing approach are exploited to predict the incoming of disruptions. In particular, we propose a neuro-fuzzy approach cooperated with chaos theory for the prediction problem above mentioned. By means of chaos theory, it is possible to compute a suitable time window for prediction problem. The aim is to establish the presence of chaos and its degree in the system under study in order to extract a time window for prediction. Power Spectrum Density is computed to determine the presence of chaos in data. Then, states space determination is carried out in order to extract the emdedding dimension (d), that establishes the number of coordinates in which each point of states space is represented, and the time lag (t), that represents an integer multiple of sampling period exploited to reconstruct the time series. The goal of this process is to reconstruct the strange attractor that conserves the topological properties of the system. By means of d and t it is possible to compute Large Lyapunov Exponent (LLE) which measures the divergence of nearby trajectories. LLE can be exploited to compute the Horizon of Prediction (HOP) which represents the short-term predictability. In addition, d and t have been exploited to make a non linear predictive model for our problem. In particular, we propose the use of Neural Networks and Fuzzy Inference Systems to predict the incoming of disruption starting from chaotic parameters. The obtained results have shown a good prediction with respect to time window computed by HOP.


Corresponding Author:

Francesco Carlo Morabito

Università di Reggio Calabria. Facoltà di Ingegneria - DIMET

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-263 ADSORPTION IN INSULATOR MATERIALS ENHANCED BY D IMPLANTATION

A. Ibarra, A. Climent-Font (1.2) A. Muñoz-Martín (2)

(1) Dpto. Fisica Aplicada, Uni. Autónoma Madrid, 28049 Madrid, Spain (2) CMAM, Uni. Autónoma Madrid 28049 Madrid, Spain

Insulator materials are critical components of different heating and current drive systems as well as for many diagnostics to be used in the ITER machine. Insulator materials are used as optical and dielectric windows and as electric insulator in cables, feedthroughs and connectors. In many of these applications, the insulator surface is exposed to a “dirty” gas phase composed of DT gas and plasma with hydrocarbons, stainless steel, W and Be particles, produced by the erosion of the first wall materials by the fusion plasma, combined with the presence of an intense neutron and gamma radiation field and high electric and magnetic fields. These particles can be deposited on the surface of the insulator giving rise to degradation of their properties, like, for example, the presence of hot spots in radiofrequency windows or the increase of electrical conductivity. No clear information is actually available on the characteristics and deposition rate that can be actually expected for ITER applications, neither on the possible effect of the radiation or electromagnetic fields on these properties. In this work, a number of different insulator candidate materials (Al2O3, SiO2, diamond) are implanted at room temperature with low energy (several keV) D and H ions in order to qualitatively simulate some of the DT gas effects. The implantation effects are characterized using optical absorption and Elastic Recoil Detection (ERD) techniques. It is observed an increase of the optical absorption due to the implantation that can be related to an increase of the surface scattering. In parallel it can be observed an increase in the C and H adsorbed at the surface. These effects are observed for all the studied materials and suggest that the implantation degrades the surface characteristics increasing its adsorption capability. This can induce important effects on the long term behaviour of insulator materials for fusion. The implantation energy and dose dependence of the measured behaviour will be discussed.


Corresponding Author:

A. Ibarra

Euratom / CIEMAT Fusion Association, 28040 Madrid, Spain

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-270 RADIATION ENHANCED DEGRADATION OF SIO OVERCOATED ALUMINIUM MIRRORS

E.R. Hodgson, T. Hernandez, and A. Moroño

Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain

High quality mirrors for the optical UV - visible - IR range will be required in ITER for both remote handling applications and diagnostic systems. The reflectivity of the mirror surface can be degraded by many different mechanisms ranging from simple contamination, to sputtering erosion due to particle bombardment, and more complex enhanced radiation effects which may modify the surface structure. The commercially available high quality mirrors being considered for ITER applications consist of a thin evaporated aluminium layer on a solid glass substrate. To protect the delicate aluminium surface the mirrors are covered (overcoated) with a controlled layer of transparent protective material such as SiO, of adequate thickness to obtain optimum optical constructive interference in a given wavelength range, usually the visible region. The work to be presented describes experiments carried out in order to study radiation enhanced degradation of such high quality mirrors, in particularly the importance of the surrounding operational environment. Tests were made on Coherent research quality SiO overcoated Al on Pyrex glass mirrors. The irradiations were performed in a sample chamber mounted in the beam line of a 2 MeV Van de Graaff accelerator. The chamber permits irradiations perpendicular to the reflecting surface to be performed in different environments. In this way mirror samples were irradiated at 25 C in dry air, nitrogen, and high vacuum (10-6 mbar) with a 1.8 MeV electrons at 150 Gy/s up to a dose of 10 MGy. Reflectivity measurements from 250 to 2500 nm were made at different irradiation doses and compared with the reflectivity before irradiation. The mirrors irradiated in nitrogen and vacuum did not show any measurable changes in reflectivity. However mirrors irradiated in dry air exhibited a marked reflectivity decrease in the visible (violet/blue region) and particularly in the UV range. This is believed to be due to radiation enhanced oxidation of the SiO protective layer. As the SiO layer oxidizes both thickness and refractive index of the layer are modified hence degrading the optimised optical interference process.


Corresponding Author:

E.R. Hodgson

Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-272 THE NEW MEASUREMENT MONITORING SYSTEM ON FTU

Bertocchi Alfredo, Salvatore Podda (1) Vincenzo Vitale (1)

(1) Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy

Needs concerning the possibility for customers to easily plot a wide range of FTU subplants measurements, particularly the named “slow measurements”, have lead to redesign the related acquisition system. Opto22 modules – which use Ethernet as fieldbus – were adopted to substitute old PLC devices. An appropriate MySQL database is updated by a C++ Opto22 driver and a CORBA server, running on the same machine hosting the MySQL server, allows the database access to any remote CORBA client by means of some specific methods. In the previous architecture the new signal addition meant a lot of configuration work based on the use of proprietary and dedicated tools. Moreover the monitoring system constrained to use a non user-friendly graphical interface based on a commercial software package with a strong platform dependence (Digital Unix). Vice versa the new situation allows to overcome these limitations: it makes possible to easily configure Opto22 modules and MySQL database within a browser while data management and visualization are achieved by using a graphical interface developed in Java, which yields these operations completely platform independent. In addition the CORBA server introduces the following advantages: 1. a hardware independence, giving maximum flexibility about the choice of platforms and devices as system components, 2. both network and programming language transparency. A remarkable aspect looks out for the use of packages totally free software. This paper will present the new system architecture, last results and future developments.


Corresponding Author:

Bertocchi Alfredo

Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-281 FIRST RESULTS OF MINIMUM FISHER REGULARISATION AS UNFOLDING METHOD FOR JET NE213 LIQUID SCINTILLATOR NEUTRON SPECTROMETRY

Jan Mlynar, John M Adams (1) Luciano Bertalot (2) Sean Conroy (3)

(1)Association EURATOM-UKAEA Fusion, Culham Science Centre, Abingdon, Oxfordshire, OX143DB, UK (2)Association EURATOM-ENEA sulla Fusione, Frascati, C.P. 65, 00044-Frascati, Italy (3)Association EURATOM-VR, Uppsala University, SE-751 05 Uppsala, Sweden

At JET, the NE213 liquid scintillator is being validated as a diagnostic tool for spectral measurements of neutrons emitted from the plasma. The main goals of the diagnostic are to measure characteristics of 2.5 MeV (D-D fusion) and 14 MeV (D-T fusion) spectra, and to evaluate the neutron fraction due to ICRH fast ions. The relation between acquired pulse height spectrum and source neutron spectrum is not straightforward. The response function matrix of the scintillator is based on theoretical prediction and accelerator calibration. This matrix, when multiplied by the (unknown) neutron spectrum, gives the (measured) pulse height spectrum. The unfolding process thus consists of finding an inverted solution. This is an ill-conditioned problem which, without further constraints, leads to unrealistic amplifications of systematic errors and noise. The Minimum Fisher Regularisation (MFR) has been applied in two-dimensional tomography of SXR and bolometric measurements at the TCV tokamak. It is a direct inversion method, which constrains the object function smoothness, providing robust fits rapidly. At JET, the Maximum Entropy Method has been applied to neutron spectra unfolding, with occasional ambiguous results. The MFR, which presents a completely different inversion algorithm, has been implemented as an independent and transparent tool to validate the JET neutron spectra measured with the NE213 liquid scintillators. The adapted MFR algorithm was first thoroughly tested on phantom spectra. Following the encouraging results, real pulse height spectra from the JET NE213 neutron detector were successfully analysed from D-D, Trace Tritium and ICRH fast ions experiments. These first applications of MFR tend to confirm independent JET results, e.g. presence of high energy (approx. 4 MeV) alpha particles in experiments studying their acceleration by 3rd harmonics ICRH waves.


Corresponding Author:

Jan Mlynar

EFDA-JET CSU, Culham Science Centre, OX14 3DB Abingdon, UK / Association EURATOM-IPP.CR, Institute of Plasma Physics AS CR, CZ-182 21 Prague 8, Czech Republic

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-282 PRESENT STATUS OF THE TJ-II REMOTE PARTICIPATION SYSTEM

Vega, Jesus, Sánchez E. (1) López A. (1) Portas A. (1) Ochando M. (1) Ascasíbar E. (1) Mollinedo A. (2) Muñoz J. A. (2) Sánchez A. (2) Ruiz M. (3) Barrera E. (3) López S. (3) Castro R. (4) López D. (4)

(1) Asociación EURATOM/CIEMAT para Fusión. Madrid. Spain. (2) CIEMAT. Computing Center. Madrid. Spain. (3) UPM. Dpto. de Sistemas Electrónicos y de Control. Madrid. Spain. (4) Red.es. Madrid. Spain

The TJ-II remote participation system (RPS) has been designed to extend to INTERNET the present working capabilities provided in the TJ-II local environment, i.e., tracking the TJ-II operation, monitoring/programming data acquisition and control systems, and accessing databases. Critical aspects considered in this development arose as a consequence of three factors: - Growth capabilities. The system architecture had to be flexible enough to permit new requirements to be incorporated at any point in time. - Open system. Local TJ-II group platforms cannot be imposed on remote participants. - Working environment. Firstly, security is “a must” for systems exposed to the INTERNET. Secondly, software distribution and version control can be a major problem in local area networks with the result that any negative effects could be amplified in INTERNET. Thirdly, the unavoidable administration tasks of the system cannot be forgotten. After taking into account the above constraints and the capabilities to be provided, the TJ-II remote participation system was based on web technologies. A web server acts as a communication front-end between remote participants and local TJ-II elements. The remote participation system is based on Java Technology because of its open character, security properties and technological maturity. From the server side, network services are provided by means of resources supplied by JSP pages. The client part makes use of web browsers and ad-hoc Java applications. Regarding software deployment, use was made of solutions based on the Java Network Launching Protocol. The operation requires the use of a distributed authentication and authorization system for filtering user queries. This development employs the PAPI System. At present, approximately 1000 digitization channels can be managed from the TJ-II RPS. Furthermore, processing software based on a 4GL language (LabView) can be downloaded to multiprocessor data acquisition systems. Also, 15 control systems are available from the RPS. Databases and the TJ-II operation logbook are available (the system even allows for the physicist in charge of operation to be in a remote location). Audio/video resources allow on-line access to the TJ-II control room and at present four Spanish universities make use of the TJ-II remote participation system capabilities for joint collaborations: these are the UPM, UNED, UCM and UPC.


Corresponding Author:

Vega, Jesus

ASOCIACION EURATOM/CIEMAT PARA FUSION. Avda. Complutense, 22. 28040 Madrid (SPAIN)

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-347 APD DETECTOR ELECTRONICS AND PXI BASED DATA ACQUISITION SYSTEM FOR SST-1 THOMSON SCATTERING DIAGNOSTICS

Chavda Chhaya, T.Aruna,K.Patel,Y.C. Saxena,Ajaikumar

Institute for Plasma Reserach, Near Indira bridge, Bhat,Gandhinagar, Gujarat,382 428 India

An electronic system with optimum signal conditioning has been designed and tested for a thermoelectrically cooled Si-avalanche photodiode (APD) to measure the Thomson scattered spectrum from the SST-1 tokamak plasma. The electronic system consists of a current-feedback preamplifier and the read-out unit. The read-out system has preamplifiers with a fast (50 MHz) and slow (1 MHz) output. The fast output is first delayed by 100ns and then the delayed and non-delayed output are processed by a differential amplifier to subtract low frequency background light component. The slow output signal is further amplified with a computer control variable gain amplifier for the purpose of error measurement and calibration. To avoid ground loops and other pickups, fast signal output is digitized by in-house developed charge to time converter in the same electronics read out system. The fast time data from detector electronics system is converted to digital data by time to digital converter using TDC from LeCroy. The prototype charge to time converter is tested using CAMAC based precision charge/time generator module with charge varies from 10pC to 600pC with a gate of 20nsec. PXI based system transfer the digital data to PC for further analysis. Multipoint Thomson scattering system needs large number of charge coupled digitizer channels. For reducing the number of charge ADCs, a scheme is worked out to multiplex the charge signals. To test the concept a prototype of 4-channel multipexer system is developed and tested for simultaneous sampling with standard CAMAC based modules.


Corresponding Author:

Chavda Chhaya

Institute for Plasma Research

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-366 REAL TIME MEASUREMENT AND CONTROL AT JET - DIAGNOSTIC SYSTEMS

Murari, Andrea (3), Robert Felton(1) Joffrin, Emmanuel(2)

(1) Euratom/UKAEA Fusion Association, Culham Science Centre Abingdon OX14 3DB UK (2) Association EURATOM-CEA, CEA Cadarache 13108 Saint-Paul-lez-Durance France

To meet the requirements of the scientific programme, the EFDA JET Real Time Measurement and Control Project has developed an integrated set of real-time plasma measurements, experiment control, and communication facilities. Traditional experiments collected instrument data during the plasma pulse and calculated physics data after the pulse. The challenge for continuous tokamak operation is to calculate the physics data in real-time, keeping up with the evolution of the plasma. In JET, many Plasma Diagnostics have been augmented with extra data acquisition and signal-processing systems so that they can both capture instrument data for conventional post-pulse analysis and calculate calibrated, validated physics results in real-time. During the pulse, the systems send sampled data sets into a network, which distributes the data to several destinations. These may do further analysis integrating data from several measurements or may control the plasma scenario by heating or fuelling. The simplest real-time Diagnostic systems apply scale factors to the signals, as with the Electron Cyclotron Emission Diagnostic’s 96 tuned radiometer channels, giving the electron temperature profile. In various Spectroscopy Diagnostics, spectral features are least-squares-fitted to measured spectra from several lines of site, within 50 ms. Ion temperatures and rotation speed can be calculated from the line widths and shifts. For Diagnostics using modulation techniques, the systems implement digital-signal processing phase trackers, lock-in amplifiers and filters. The interferometer samples 15 channels at 400 kHz for 30 s, i.e. 6 million samples/sec ! Diagnostics have specific lines of sight, spatial channels and various sampling rates. The Heating/Fuelling systems have relatively coarse spatial localisation. Analysis systems have been developed to integrate the basic physics data into smaller sets of controllable parameters on a common geometry e.g. temperature, density and safety factor profiles with values at 10 points of normalised radius. The EFDA Real Time project is essential groundwork for future reactors such as ITER, and has successfully involved many scientific and technical staff from several institutions. The facility is now frequently used in experiments. This work has been conducted under the European Fusion Development Agreement and is partly funded by Euratom and the UK Engineering and Physical Sciences Research Council.


Corresponding Author:

Murari, Andrea (3)

(3) Euratom/UKAEA Fusion Association, Culham Science Centre Abingdon OX14 3DB UK

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-370 OPTICAL FIBERS FOR PLASMA DIAGNOSTICS UNDER GAMMA-RAY AND UV IRRADIATION

Sporea Dan, Adelina Sporea(1) Bogdan Constantinescu(2)

(1)National Institute for Lasers, Plasma and Radiation Physics, Atomistilor St.409, 76900 Magurele Romania. (2) National Institute for Physics and Nuclear Engineering, Atomistilor St.409,76900 Magurele Romania

Selected also for oral presentation O2B-D-370

The use of optical fibers under irradiation conditions (gamma, UV, neutron) is of interest, if we consider the role they can play in diagnostics, remote handling and control, in future fusion installations (i.e. ITER, DEMO). The purpose of our investigation was to evaluate the combined effects of various factors (UV radiation, gamma irradiation, temperature stress) on the optical transmission of optical fibers, which can be candidates for the development of light guides. We used large core diameter optical fibers (400 microns), having an enhanced UV transmission, and resisting to high temperatures. The optical fibers employed are commercially available products. They were subjected to different treatments: - gamma irradiation and temperature cycling; - UV exposure and temperature cycling; UV, - gamma and temperature cycling. The UV exposure was done with a stabilized CW operation, continuum spectrum deuterium lamp, which was also used for the measurement of the spectral absorption of the optical fiber. Gamma irradiation was done in 250 krad steps, up to a total dose of 1500 krad. All the measurements were carried out off-line, before and after each treatment, with the set-up based on a multi-channel optical fiber spectrometer. The measurements covered the spectral range from 200 nm to 1000 nm. For the case we consider the base-line form the spectrum graphs as the absorption corresponding to the non-irradiated optical fibers (a value of 0.7 units in a logarithmic scale), our investigations indicate: - the wavelength corresponding to the UV induced peak absorption is at about 220 nm; - the UV induced absorption seems to saturate (i.e. a 2 h UV exposure induces an absorption peak of 0.9 units, a 6 h exposure one of 1.05 units, while an exposure of 10 h produces a peak of 1.08 units); - temperature stress applied after UV exposure reduces the optical absorption in a limited range (about 0.4 - 0.6 units); - the peak for gamma induced optical absorption is located between 230 nm and 235 nm, depending on the gamma total dose and the temperature stress applied; - in the case of pure gamma irradiation (no UV exposure) a temperature stress reduces the absorption peak by 0.3 unit; - any temperature treatment following a gamma irradiation produces a decrease of the optical absorption, but the radiation effect can not be reduced completely.


Corresponding Author:

Sporea Dan

Department of Lasers, National Institute for Lasers, Plasma and Radiation Physics, Atomistilor St. 409, Magurele 76900 Romania

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-384 THE HALO CURRENT SENSOR SYSTEM FOR JET-EP

Marcuzzi Diego, P.Sonato(1) W.Baker(1) P.Beaumont(2) T.Bolzonella(1) C.Damiani(3) P.Fiorentin(1) A.Guigon(3) K.Fullard(2) A.Goodyear(2) L.Grando(1) S.Huntley(2) N.Lam(2) A.Lioure(3) A.Loving(2) S.Mills(2) S.Peruzzo(1) N.Pomaro(1) V.Riccardo(2) M.Way(2)

(1)Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, 35127 Padova Italy (2)UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 DB United Kingdom (3)EFDA-JET-CSU Enhancement Department,Culham Science Centre

Vertical instability of elongated plasma is a problem theoretically predicted and experimentally observed on many different devices. During Vertical Displacement Events (VDEs), currents flowing through plasma and vacuum vessel are present. These currents, called Halo Currents (HCs), induce severe mechanical stresses on the plasma facing components and in the vessel, and are a major concern for present and future fusion experiments The need to better understand the origin, the distribution, and the scaling of HCs is one of the critical points for any next step device like the ITER project. The new system of Halo Current Sensors (HCS) designed for JET-EP should help evaluating HC density distribution, localization and rotation as well as toroidal and poloidal current asymmetries, their nature and correlation with other plasma parameters. The system will be integrated with the sensors (toroidal pick up coils, Rogowski coils and mushroom tiles instrumented with shunts) already in operation. The new system consists of Rogowski coils and toroidal field pick-up coils. The Rogowski coils will measure directly the current flowing through some of the tiles of the upper dump plate. They are housed in a groove machined in the CFC tiles and are designed to collect the current flowing through one single tile. The toroidal field pick-up coils will estimate the total poloidal HC. The HCS system will include 4 identical mechanical structures each including: 8 Rogowski coils and 2 toroidal pick up coils. The coil assemblies are installed at the top of the vessel close to secondary X point in 4 octants equally spaced along the toroidal coordinate. Both sensors are coils wound around a ceramic core. They have to withstand temperature up to 400 C, therefore the windings have to be made of mineral insulated cables; the cable section is only 0.65 mm to have a sufficiently large effective area. The mineral insulated cables have to be UHV sealed and must withstand a voltage up to 1000 V in DC and therefore a special termination has been qualified. A special procedure for the UHV leak tightness qualification of the terminations has been implemented. Special tools and procedure have been implemented and tested for the remote installation of the four assemblies. In the paper the final design will be presented in detail, as well as the results of the tests performed for the qualification of components and the main features of the manufacturing phases of the complete system.


Corresponding Author:

Marcuzzi Diego

Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, 35127 Padova, Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-396 DEVELOPMENT OF ACTIVELY COOLED PERISCOPES FOR DIVERTOR OBSERVATION

Koenig, Ralf, K. Grosser(1), D. Hildebrandt(1), O. Ogorodnikova(2), C. von Sehren(1) and T. Klinger(1)

(1)Max-Planck-Institut für Plasmaphysik, Wendelsteinstr. 1, 17491 Greifswald Germany. (2)IWV-II, FZ-Jülich GmbH, D-52425 Jülich Germany

With the stellarator W7-X a step to quasi-continuous plasma operation will be made. The cooling system of the machine is designed such that two 30 min discharges can be run per day. Right from the start of operation 10 MW of ECRH heating power will be available for quasi-continuous operation. A working group “Plasma Facing Optical Components” has been formed which presently concentrates on the development of water cooled shutters and windows for UV/Vis./IR periscopes which can withstand the expected maximum heat loads of up to 50 kW/m2 which due to the predominantly short wavelength nature of the radiation emitted by the plasma will be absorbed within the first millimeter of any window. We will report on the detailed ANSYS calculations of the heat and stress distribution across the shutter, the bottom of the immersion tube and in particular the windows. In the case of the windows calculations have been undertaken for a large number of different materials which are required for the various spectral regions covered by the miscellaneous diagnostics, so that the most suitable material for each application can easily be identified. Also the dependence of the cooling rate on the window diameter and thickness has been studied. The calculations suggest that CVD-diamond which might be required as a very expensive last resort for a few large windows (dia. >150 mm) can most likely be avoided by using sapphire but that for many of the other materials like ZnSe, ZnS, CaF2, MgF2 and quartz one will be limited to considerably smaller sizes. A vacuum test chamber, equipped with a vacuum compatible IR heater has been build. In this chamber a low cost, easily exchangeable window design using Helicoflex gaskets on either side of Sapphire and Quartz windows has been successfully tested. The design was water tight and the window materials behaved roughly as predicted by the ANSYS calculations, with sapphire, as expected, showing excellent heat removal properties. The test windows are being blackened to ensure effective absorption of the IR radiation at the surface of the substrates on the vacuum side of the windows and IR cameras for different wavelength regions as well as other test equipment were installed at the periphery of the test chamber for detailed investigations of the time evolution of the radial heat distribution across the different window materials at different power loads. These results will be compared with the ANSYS calculations.


Corresponding Author:

Koenig, Ralf

Max-Planck-Institut für Plasmaphysik, Wendelsteinstr. 1, 17491 Greifswald, Germany

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-398 DIAGNOSTICS FOR STUDYING DEPOSITION AND EROSION PROCESSES IN JET

COAD, JOSEPH PAUL, H-G Esser(1), J. Likonen(2), M Mayer(3), G Neill, V Philipps(1), M. Rubel(4), J Vince and EFDA-JET Contributors

(1)Forschungszentrum Jülich, D-52425, Jülich, Germany (2)VTT Processes, P.O. Box 1608, 02044 VTT, Finland (3)Max-Planck Institut für Plasmaphysik, D-85748 Garching, Germany (4)Alfvén Laboratory, Royal Institute of Technology, 100 44 Stockholm, Sweden

Information on how and where hydrogen isotopes are trapped in JET is important in understanding the erosion/deposition process well enough to predict performance in ITER. In the past this information has been largely limited to the analysis of first wall components removed after complete operational campaigns. New complementary diagnostics are being developed for installation in JET in 2004 under the JET Enhancement Programme and Task Force Fusion Technology. A prototype quartz micro-balance (QMB) was developed for JET to measure deposition on-line in a time-resolved manner, and has worked within the JET divertor for two years [1]. New QMBs are being developed from this prototype for installation in 2004. These QMBs are designed to measure deposition in the septum and outer divertor as well as the inner divertor, explore the effect of temperature on deposition from room temperature to ~300?C, and monitor the amount of beryllium evaporated in JET. New collector units are also being developed that have a rotation mechanism powered by the magnetic field. Erosion/deposition will be monitored over ~3000 pulses with a time resolution of about 50 pulses. Since the units need no electrical connections, they will be fitted to the outer vessel wall in addition to the divertor. However, the data on the collector can only be evaluated after the unit is retrieved from the vessel. The transport processes involved in deposition in shadowed areas of JET are not understood. First information that most of the hydrocarbon radicals involved have high sticking coefficients has come from deposition monitors in JET [2]. More deposition monitors will be mounted in 2004 in the new JET divertor. Techniques are being developed to quantify deposition throughout the torus, and to establish the areas of erosion. The surfaces of about 30 new tiles in a poloidal section of the divertor and main chamber are being mapped to an accuracy of ~2 micron in the laboratory, and a similar number are being coated with markers. Re-measurement after exposure in JET will determine the erosion/deposition pattern and lead to a much improved knowledge of erosion/deposition and fuel retention in JET. [1] H-G Esser et al, Proceedings of the Carbon Workshop, Jülich, Germany, Sept 2003 [2] M Mayer et al, ibid This work has been conducted under the European Fusion Development Agreement and is partly funded by EURATOM and the UK Engineering and Physical Sciences Research Council


Corresponding Author:

COAD, JOSEPH PAUL

EURATOM / UKAEA Fusion Association, Culham Science Centre, Abingdon, OXON OX14 3DB, U.K.

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-428 VULNERABILITY OF OPTICAL FIBERS FOR PLASMA DIAGNOSTICS OF LASER MEGAJOULE

S. Girard (1), B. Brichard (2) J. Baggio (1) J-L. Bourgade (1) M. Decréton (2) F. Berghmans (2)

(1) CEA/DIF, BP n 12, 91680 Bruyères-le-Châtel, France (2) SCK-CEN Belgian Nuclear Research Center, B-2400 Mol, Belgium

The Laser Megajoule (LMJ) project is a major component of the French simulation program to study nuclear fusion by inertial confinement. Optical plasma diagnostic systems of the LMJ have to resist to the harsh environment of this facility. The LMJ is able to focalise up to 1.8 MJ ultraviolet laser light into a volume of few mm3 and this high density will drive a large amount of X-rays and nuclear particles (neutron and gamma rays) when DT filled glass microballon implosion experiments are performed. Two different effects limit the fiber integration in the diagnostic systems: the radiation-induced attenuation (RIA) and radiation-induced luminescence. After many years of research, radiation-resistant multimode optical fibers have been developed for steady state radiative environment representative of other facilities (e.g. ITER). Optical fibers with pure silica cores based on SSU, STU or KU silica glasses and fluorine-doped claddings, exhibit low RIA levels. Furthermore, their responses could also be improved by appropriate pre-treatments, like hydrogen-loading or pre-irradiation. These fibers seem to be interesting candidates for the LMJ plasma diagnostics. However, previous studies established that pure silica core fibers, with radiation-hardened properties under gamma-ray irradiation, could present strong RIA levels for the short times after a transient exposure (duration of few nanoseconds as in the case of LMJ). We will characterize the radiation behaviors of these “rad-hard” optical fibers after a X-ray pulse (dose <1 kGy, dose rate > 1 MGy/s) and evaluate their vulnerability for the LMJ facility. Comparisons with similar studies conducted for magnetic fusion diagnostic design for ITER will be presented. This work was possible thanks the support of E.R. Hodgson from Euratom/CIEMAT


Corresponding Author:

S. Girard (1)

CEA/DIF, BP n 12, F91680 Bruyères-le-Châtel, France

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-433 APPLICATION OF ORTHOGONALLY POLARIZED TWO-FREQUENCY LASER TO POLARIMETER FOR MAGNETIC FIELD MEASUREMENTS OF LONG-PULSED FUSION DEVICES

TSUJI-IIO Shunji, MIYAZAKI Takeshi, HAYAKAWA Kazuhiro, AKIYAMA Tsuyoshi*, TSUTSUI Hiroaki, SHIMADA Ryuichi

Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550, Japan *National Institute for Fusion Science, 322-6, Oroshi-cho, Toki-shi, Gifu 509-5292, Japan

We have developed a heterodyne magneto-optic polarimeter for magnetic fields measurements on long-pulsed fusion devices to avoid the problems of zero-points drifts of long integration with commonly used pick-up loops. When frequency shifters such as acousto-optic modulators are used to apply heterodyne techniques, the alignment accuracy of recombined twoÅ@beams limits the precision of long-time field measurements. Hence a transverse Zeeman laser, which emits two orthogonally polarized beams with slightly different frequencies, was adopted as a light source. Since the two beams are perfectly collinear without any adjustments, the accuracy of long-time measurements was significantly improved. However the instability of the laser frequency induced by optical feedback from a fiber coupler to the laser source degrades the polarimetric measurement when a polarization-maintaining fiber was used to freely propagate the orthogonally polarized beams to sensor locations. To minimize the optical feedback, an optical isolator for the orthogonally polarized two-frequency laser was tested. Since the key element of the optical isolator, a Faraday rotator, is sensitive to temperature, the control of the rotator temperature was found to be crucial to keep good isolation. The achieved accuracy of the Faraday rotation angle was less than 0.01 degrees, which corresponds to a resolution of magnetic field measurement of 2 G in the case that a 40-mm long flint glass (OHARA PBH71) is used as a sensor. Magnetic field measurement up to 5 T was tested using a superconducting magnet and nonlinearity of measurement due to elliptic polarization of the beams will be discussed.

WITHDRAWN


Corresponding Author:

TSUJI-IIO Shunji

Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550, Japan

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-437 THE TEXTOR DIAGNOSTIC DATA MANAGEMENT CHAIN

KROM, J.G., KORTEN, M.K. KOSLOWSKI, H.R. KRAEMER-FLECKEN, A.

During the 2002-2003 shutdown of the Textor device many components of the Textor diagnostic data management system were upgraded. Following the summer of 2003 Textor successfully resumed operation with these new and upgraded components in place. This paper intents to give an overview of these components and how they co-operate to form one integrated Data Management system. The components in question are: = JDAQ: The Java (or Juelich) Data Acquisition System. A successful proof of this new Textor data acquisition system has been achieved during the commissioning of Textor with the Dynamic Ergodic Divertor (DED). This new system replaced the greater part of legacy RT2 systems used so far. Embarking from the well established design principles of RT2, JDAQ aimed at an open, distributed and scalable system. It has been completely written in the JAVA object oriented programming language supporting a homogeneous and strictly modular software design, using native interfacing only to connect operating system specific hardware drivers. JDAQ is designed as a multi-tier layered system, which can be run on a single node or distributed over a network. = CSF & TPD: The TEC Common Storage Facility and the Textor Physics Database. A new data storage facility was brought into operation, designed to contain the "raw" data obtained from JDAQ and other acquisition systems. A separate part of this store contains the more more physics orientated data. = TWU: The TEC Web-Umbrella All data stored in CSF, TPD and other Textor related databases is now accessible via one common data access scheme. This "TEC Web-Umbrella" is based on the widely used HTTP protocol. = Chain1: A chain of automatically run analysis software. The data access via the TWU scheme, allowed us to build a framework of analysis programs that get executed whenever a particular data set has been collected by JDAQ. Diagnosticians and/or plasma physicist, can provide their own analysis code, in their preferred programming language, and plug this program into the Chain1 framework to provide data for the TPD. We also intent to describe our experiences with this integrated system during this year of operation.


Corresponding Author:

KROM, J.G.

Institute for Plasmaphysics, Forschungszentrum Juelich, D-52425, Jelich, Germany

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-446 NEW LOW LOSS TRIAXIAL AND MAGNETICS DIAGNOSTICS FEEDTHROUGH AT JET

Dirken, Peter, Lam, Norman Altmann, Henk

Euratom-UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK

During 2003 the UKAEA designed a novel low loss tri-axial and magnetics feedthrough to be installed during the 2004/2005 shutdown. The new low loss tri-axial and magnetics feedthrough incorporates two interacting systems in one assembly. The two diagnostic systems supported by this feedthrough are: -Toroidal Alfvén Eigenmodes [TAE] antennae (high frequency at high voltage and current). -Magnetic pick-up sensors (milliVolts and milliAmps). Primary design considerations were: -Each TAE antenna will operate at up to 500kHz, 600V and 30A maximum. -RF interference from the TAE should not affect the signals from the magnetic pick-up sensors to the extent that the signal-to-noise ration is significantly degraded, particularly those used for MHD measurement. -For this purpose a new flexible tri-axial cable had to be designed, prototyped, developed, manufactured and tested. The cable was required to operate reliably up to 450 degrees C in UHV. -The design had to use materials compatible with the anticipated gamma and neutron radiation levels in JET as well as providing a double tritium boundary. -All in-vessel sockets, plugs, conduits and cabling have to be installed by Remote Handling installation to minimise the manual intervention time and activities. This required remote operation of photogrammetry, welding, alignment and positioning systems and processes. The paper will cover the constraints, design, manufacture, assembly and Remote Handling mock-up trials of the feedthrough in preparation for in-vessel installation, ready for the intended use of the JET facilities beyond 2004. Enhancements take into account the likely International Thermonuclear Experimental Reactor (ITER) operating scenarios, which will be a considerably larger tokamak device. Additionally, design choices are being made for ITER-relevant subsystems, which could be finalised even after the start of construction of ITER. Diagnostic systems for ITER will necessarily be designed to take into account the requirement to minimise manned intervention by installing systems by Remote Handling.


Corresponding Author:

Dirken, Peter

Euratom-UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-451 THERMO-STRESS ANALYSIS OF OPTICAL MATERIALS FOR HIGH HEAT FLUX APPLICATIONS

Ogorodnikova Olga, R. König2), J. Linke1),.G. Pintsuk1)

1)IWV-2, FZ-Juelich GmbH, EURATOM ASSOCIATION, D-52425 Juelich, Germany 2)Max-Planck-Institut fuer Plasmaphysik, Greifswald, EURATOM ASSOCIATION, D-17491 Greifswald, Germany

The ideal optical material for high heat flux applications would have negligible absorption, scattering and birefringence throughout the 0.15-5 micron band even at elevated temperatures. A low index of refraction is also desirable to minimize transmission losses from surface reflections. The materials should maintain acceptably low level of radiation emission and low thermally induced stresses to prevent failure by power load. So, the ideal material should have high strength, high thermal conductivity, low elastic modulus and low thermal expansion even at elevated temperatures. All of these optical and mechanical characteristics are desirable for a reasonable price. In the present work, the optical materials investigated for ultraviolet/visible/infrared window under high heat load are Al2O3, SiO2, ZnSe, ZnS, CaF2, MgF2, BaF2 and CVD-diamond. All of them have a high transmission range. Finite element method has been used for calculations of temperature and stress distributions. Influence of (1) power load, (2) pulse duration, (3) cooling conditions, (4) atmospheric pressure on unloaded side, (5) surface and bulk heating and (6) size of the window has been studied in detail by numerical calculations. The result of this study allows to choose the most suitable window for different diagnostics in the next step fusion devices such as the stellarator W7-X. This procedure is also qualified for the selection of windows materials for aerospace applications and high power continuous-wave CW lasers. The investigations show in which operating regimes with regard to pulse duration, power load and sizes every material can be used until it fails. For example, actively cooled sapphire window is suitable for large window diameters (about 13 cm) which could cope with the expected maximum radiation power loads for W7-X of 50 kW/m2 for more than 20 minutes. Other materials like fused silica, MgF2, ZnS and ZnSe can only be used for smaller diameter windows (less than 5 cm). CaF2 is unacceptable for the protective window for W7-X because of strong in-plane distortions during the long heat load. CaF2 can be used only at low power load.


Corresponding Author:

Ogorodnikova Olga

IWV-2, Forschungzentrum Juelich, 52425 Juelich, Germany

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-458 DESIGN AND MANUFACTURE OF THE UPPER COILS AND OUTER POLOIDAL COILS SUBSYSTEMS FOR THE JET-EP MAGNETIC DIAGNOSTIC

Peruzzo Simone, W.Baker(1), V.Coccorese(2), T.Edlington(3), S.Gerasimov(3), S.Huntley(3), N.Lam(3), A.Loving(3), N.Pomaro(1), and JET-EFDA contributors

(1) Consorzio RFX - Association EURATOM-ENEA, C.so Stati Uniti 4, I-35127 Padova. (2) Consorzio CREATE - Association Euratom-ENEA, Via Claudio 21, I-80125 Napoli. (3) UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK

The enhancement project of the magnetic diagnostics aims at the design, procurement, installation and commissioning of new sets of magnetic transducers to be installed inside and outside the vessel, in order to substantially improve the current capabilities of the JET magnetics. All in-vessel coils have been designed considering the necessity of installing them by means of the Remote Handling system available at JET. The in-vessel sensors are grouped in 3 different sub-systems of two field component pick-up coils, to be located as near as possible to the plasma, assembled on rails in order to ease Remote Handling installation. The 3 sub-systems (Upper Coils, Outer Poloidal Limiter Coils, Divertor Coils) are attached to different structures of the first wall and replicated for redundancy in 2 Octants, for a total of 76 new pick-up coils. This paper will focus on the detailed design and manufacture of the first two subsystems (UC and OPLC). The final design of the two sub-systems has been carried out in order to fulfil all the operation and fault condition design criteria for JET in-vessel components. Preliminary tests on mock-ups have been performed in order to guarantee the compatibility with the features of the Remote Handling system. Most of the coils are made of a Mineral Insulated Cable (MIC) wound around Inconel formers, which give a good reliability in the JET environment. The relatively low frequency response (about 50 kHz) is more than adequate for plasma control and equilibrium reconstruction, which require a maximum frequency of 10 kHz. In addition a sub-set of 14 “tangential coils”, located on the Outer Poloidal Limiter, is designed as titanium bare wire wound onto an alumina ceramic former, so that they can be used for high frequency applications (e.g. MHD studies). For low frequency coils special design attention has been dedicated to the coils terminations, connecting the MIC to standard shielded cables inserted into conduits. The signals are finally driven out of the vessel by means of new feedthroughs, which have been designed in order to minimize interference with other systems. To improve reliability all coils are subjected to severe vacuum and temperature tests. In addition a calibration procedure is applied to all coils, in order to minimise systematic measurement errors.


Corresponding Author:

Peruzzo Simone

Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova Italy.

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-459 DESIGN OF EX-VESSEL MAGNETIC PROBES FOR JET-EP

Chitarin Giuseppe, F. Basso (1), V. Coccorese (2), S. Peruzzo (1), N. Pomaro (1), T. Edlington(3), C.Sowden (3), S.Cramp (3), K.Fullard (3) and JET-EFDA contributors

(1) Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, Padova, Italy (2) Consorzio CREATE, Association Euratom-ENEA, via Claudio 21, Napoli, Italy (3) UKAEA-Euratom Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK

The enhancement of JET magnetic diagnostics will include the installation of new ex-vessel sensors, which are mainly intended to provide information useful for the characterization of the iron transformer during the discharge. The numerical codes used for equilibrium reconstruction need an equivalent, axisymmetric iron model and inaccuracies in the iron core modelling may affect the accuracy of the equilibrium reconstruction, especially during some critical phases of the discharge. The present numerical models of the iron core are based on the field measurements obtained by time-integrated pick-up coil and flux-loop signals, together with simplified information on the geometry and magnetic properties of the iron structure. However, no information is presently available on the magnetic field configuration in proximity of the iron core structure and on the contribution of the residual magnetization of the iron. It is expected that the equilibrium reconstruction should benefit from the absolute measurements of the iron stray field before and during the pulse. The ex-vessel sensor system to be installed consists of a number of Hall sensors, complemented by “local” pick-up coils and “octant average” flux loops. All these sensors are supported by rails to be attached to the iron core structure. There are in total 26 new sensors grouped in 2 subsystems, named Collar probes and Limb probes, respectively. The design activity is conducted in collaboration between the Association EURATOM-ENEA and the JET Operator. For these purposes new Hall probes are being introduced in the magnetic diagnostics. Some prototype Hall probes have already been installed on the Limb surface and tested before the start of the 2004 shutdown, mainly in order to verify the reliability of the absolute field measurements over a time-scale of several weeks and to compare the behaviour of different kinds of Hall sensors under real operating conditions. The paper will describe the preliminary results obtained with Hall probe prototypes, and will discuss the choice of type and location of Collar and Limb probes, together with other design and manufacturing issues.


Corresponding Author:

Chitarin Giuseppe

Consorzio RFX, C.so Stati Uniti 4, 35127 Padova, Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-466 TRANSDUCERS AND SIGNAL CONDITIONERS OF THE RFX NEW MAGNETIC MEASUREMENT SYSTEM

Pomaro Nicola, Basso Francesco

Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova, Italy.

The paper presents the out of vessel magnetic measurement system recently installed on RFX. The system has been completely redesigned in order to meet the electromagnetic, mechanical and thermal requirements of the new toroidal assembly of RFX, with the thinner shell and the new system of coils for MHD modes control. The system allows to measure the main integral plasma quantities: toroidal current, loop-voltage, toroidal flux. Special partial poloidal voltage measurements are foreseen to allow the study of halo currents poloidal and toroidal spectra. Furthermore, 224 two axis probes and 240 saddle loops were installed for local field measurement. In total, 744 independent measurements are available; 712 of them are placed in between the vacuum vessel and the new thin shell, where an air gap only 6 mm thick is available, temperature can rise up to 200 C, and a voltage up to 2000 volt can develop between shell and vessel during machine operation. To comply with such extreme requirements, specific techniques were developed for probes and connections, which make use of advanced polymers for protection and insulation. Particularly challenging was the design and realisation of Rogowski probes, which were placed into existing grooves in the vessel, only 3 mm thick. To avoid mechanical interferences, a detailed study of cable paths was carried out, with the aid of 3D computer modelling. To obtain the best precision in probes and connections installation, a special tool was designed and realised, which allowed to draw directly on the vessel surface the position of each probe and cable, with a precision better than 1 mm. Conditioning electronics was completely redesigned to improve accuracy and interferences immunity. Integrators drift was lowered by an order of magnitude with respect to previous system, and a new Front-End was designed to cope with extended signals bandwidth and the presence of high frequency noise due to new powerful switching power supplies. Also mechanical characteristics of conditioning channels were changed to improve modularity and reliability. This paper describes the realisation of all integral probes, the tools and procedures adopted for installation and tests of the whole probe system, and the design and realisation of the conditioning electronics.


Corresponding Author:

Pomaro Nicola

Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova, Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-476 WIDE-ANGLE INFRARED THERMOGRAPHY FOR JET-EP

Eric Gauthier, E. Thomas, B. Bertrand, P. Chappuis, L. Doceul, D. Guilhem, M. Missirlian, P. Andrew2, P. Coad2, T. Tiscornia2, C. Antonnucci3, C. Damiani3, J. Gafert3, A. Lioure3 and contributors to the EFDA-JET work program

2Euratom/UKAEA Fusion association, Culham Science centre, Abingdon OX14 3DB, UK 3EFDA-CSU Culham, Culham Science centre, Abingdon OX14 3DB, UK

The surface temperature of the plasma facing components need to be measured to operate the tokamak in a safe manner and to calculate the power flux impinging on the different parts of the machine. In the frame of JET-EP (Enhancement Phase), a new infrared thermography diagnostic is being developed. The objective is to provide a wide-angle view in the infrared range (3 to 5 µm) for thermography in the main chamber and divertor aiming at real time machine protection and for analysis of the power flux deposition during normal operation and transient events such as disruptions and ELMs. The diagnostic will be able to measure temperature with a large dynamic range from operating temperature of 200 C up to a maximum temperature of 2000 C. The enhanced dynamic range is achieved by using a multi-exposure time: acquisition with three different exposure times is performed and the corresponding frames are combined in a single thermal image. In order to measure accurately the power and energy deposition during ELMs, a time resolution of the order of 100µs is achieved by reducing the image size to 128x8 pixels, and by using a 40 MHz pixel clock. In order to image a large section of the tokamak in both poloidal and toroidal directions, dedicated optics have been designed. The optics have a field of view of 70 degrees, viewing the divertor, the inner wall, the outer poloidal limiters, the ITER-like ICRH antenna and the top limiter. The main feature of the optics design is to be ITER-relevant. To this end, the optical components are based mainly on reflective optics: the only kind which can sustain high neutron radiation. The optical system consists of an endoscope installed in a lower limiter guide tube, a Cassegrain telescope and a relay group of lenses, the latter being connected to the camera body. The design uses a concave aspheric mirror located behind a plan mirror equipped with a small aperture. Both mirrors are made of stainless steel coated with gold. The Cassegrain telescope is composed by elliptic and hyperbolic mirrors, both made of Zerodur® glass (gold coated). Finally, the image is magnified and transmitted to the detector with 4 Silicon and Germanium relay lenses. Strong requirements in positioning the optical elements, in addition to the vacuum and thermal constraints, imposed a complex and challenging mechanical design. The paper will present the main characteristics of the optical and mechanical designs.


Corresponding Author:

Eric Gauthier

Association EURATOM-CEA, CEA/DSM/DRFC, CEA Cadarache, 13108 Saint Paul Lez Durance (France)

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-477 LITHIUM BEAM DEVELOPMENTS FOR HIGH-ENERGY PLASMA DIAGNOSTICS

Anda Gabor, S. Bató G. Petravic S. Zoletnik

Lithium beam developments for high-energy plasma diagnostics G. Anda, S. Bató, G. Petravic, S. Zoletnik KFKI-RMKI, Association EURATOM, P. O. Box 49, H-1525 Budapest, Hungary The injection of 10-100 keV Li neutral beam into magnetically confined fusion plasmas causes collisionally induced Li line emission at 670.8 nm. Observing the intensity and the fluctuations of this Li resonance line along the beam, it is possible to reconstruct the density profile and the 2 dimensional correlation of the electron density fluctuation. Measurement of the line shape gives information on the magnetic field in the observation volume. Charge exchange with plasma impurity ions gives the possibility of the determination of impurity ion concentrations. In this paper the JET Li-gun and a new ion optic geometry are investigated experimentally and with the CPO simulation code. The beam has been tested in the laboratory at IPP-Garching to find optimal operation conditions and limits. Intersecting the beam path with a Titanium plate the beam ions get neutralised on the surface and some of them become excited. The beam profile was measured by observing the radiation resulting from this excitation. Although the photon yield from one ion (or atom) is unknown, it can be assumed that the individual ions radiate independently and the observed light profile is proportional to the beam current profile. These test results have been used to validate the code calculations and to explore possible beam source upgrades. A new ion optic geometry is developed to determine the relation between the emitter temperature and the drawn current and to test a new type of emitter and emitter material. It turned out that the drawn current highly depends on the temperature of the emitter. The recently developed new emitter material was found to be capable of delivering substantially higher ion current than the conventionally used B-eucriptit and spodumen sources. These developments promise much better signal to noise ratios and potentially new application areas for high-energy Lithium beam diagnostics.


Corresponding Author:

Anda Gabor

H-1121 Budapest Konkoly Thege Miklos Street 29-33.

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-502 THE NEW TAE - ALFVÉN WAVE ACTIVE EXCITATION SYSTEM AT JET

Duccio Testa, A.Fasoli1,2, P.Beaumont3, R.Bertizzolo1, M.Bigi3, C.Boswell2, R.Chavan1, S.Huntley3, N.Lam3, A.Loving3, S.Mills3, V.Riccardo3, S.G.Sanders3, J.A.Snipes2, J.Thomas3, P.Titus2, L.Villard1, M.Vincent3, R.Walton3, M.Way3, and JET-EFDA contributors

(1) CRPP, Association EURATOM-Confédération Suisse, EPFL, 1015 Lausanne, Switzerland (2) Plasma Science and Fusion Center, MIT, Cambridge, MA 02139, USA (3) EURATOM-UKAEA Fusion Association, Culham Science Centre, OX14 3EA, Abingdon

After many years of successful operation, the JET saddle coil system will be dismantled during the next shutdown. A new antenna system has been designed to replace it and excite MHD modes in the Alfvén frequency range (10-500kHz), keeping similar operational capabilities (I~30A, V~1kV, P~5kW). Due to their geometry, the saddle coils could drive only low toroidal mode numbers, n=0-2. Conversely, the n’s that can be driven unstable in ITER by fusion generated alphas or other fast particles are expected to be in the range n~5-20. The mismatch between the modes driven by the saddle coils and those that are made unstable by the fast particles is already observed on JET, which sometimes makes it difficult to extrapolate the present results to large burning plasma experiments. The new antenna system is designed to overcome this limitation. It comprises two assemblies of four toroidally spaced coils each, situated at opposite toroidal locations. Each coil is made using 18 turns of 4mm Inconel 718 wire, covers a toroidal and poloidal extent of ~25cm, and is individually insulated from the supporting frame with Shapal-M spacers. The first turn sits approximately 40mm behind the poloidal limiters. The coils are mounted on a 3mm-thick Inconel 625 open structure, to avoid a closed path for disruption-induced currents. This structure is attached to the poloidal limiters and the remains of the saddle coil brackets with four attachment points, so as to optimise the load distribution, and it is further protected by CFC tiles. Any combination of 4 out of the 8 antennas can be chosen to excite different n-spectra, up to n~20. Code calculations show that this arrangement provides a coupling to the plasma for a n=5 TAE that is of the same order as that achieved with the present, much bigger, saddle coils for an n=2 TAE for the same JET equilibrium. Thus, it is foreseen that the real-time tracking capabilities of the old saddle coil system will be maintained and, possibly, control of the marginal stability limit could be envisaged for the intermediate n-modes driven by the new antennas. In addition to the constraints imposed by halo current and disruption-induced voltages and currents, the design must comply with the requirements of a remote handling installation. Design principles and constraints will be presented along with the results of the coupling and engineering analysis, and a discussion of the possible extrapolation of such a system to ITER


Corresponding Author:

Duccio Testa

CRPP, Association EURATOM-Confédération Suisse, EPFL, 1015 Lausanne, Switzerland

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-503 NEW MILLIMETER-WAVE ACCESS FOR JET REFLECTOMETRY AND ECE

Cupido Luis, L. Cupido(1), E. de la Luna(2), et al from CSU(3), FOM(4), CNR(5), IPP(6), CFN-IST(1), JET-EFDA(7)

1-CFN, IST, Lisboa, Portugal/ 2-CIEMAT, Madrid, Spain/ 3-CSU Culham Science Centre, Abingdon UK/ 4-FOM-Rijnhuizen, The Netherlands/ 5-IFP-CNR Milano Italy/ 6-IPP, Max-Planck-Institut, Garching, Germany/ 7-UKAEA Abingdon UK.

Millimeter-wave ECE and Reflectometry at JET employ state of the art electronics but are limited in performance by the existing waveguide and antenna system. The use of long runs of waveguides with high losses and non optimized antennas lead to difficult measurement conditions for both ECE and reflectometry. An access system has been designed to improve the performance of reflectometry and enable the installation of antennas for ECE oblique view. The new antennas will allow the ECE radiation to be collected at different angles with respect to the magnetic field, which is extremely useful to improve the interpretation of ECE temperature measurements in JET. For reflectometry there is an urgent need to improve the edge density measurements as both the lithium beam and Thomson scattering exhibit limitations of resolution at lower densities. The project aims at the installation of a millimeter-wave access system consisting of six antennas/waveguides for probing the mid-plane of the JET plasma, covering a frequency range of 60-190GHz. Two of those antennas and transmission lines are dedicated to the oblique view ECE. Access to the plasma will be done using a port with direct line of sight to the plasma, a limiter guide tube. This port allows a complete bundle of antennas and waveguides to be inserted from the outside of the vessel. The paper presents the analysis and design of the antennas, corrugated waveguides, vacuum windows and instrument interface. Four antenna apertures take advantage of the excellent coupling of the propagating HE11 waveguide mode to the free-space Gaussian beam which is also inherently broadband. A vacuum boundary with double dielectric windows and inter-stage vacuum was designed to operate in the 60-190GHz range. The corrugated waveguides were designed for the 60-190GHz range. Oblique ECE measurements, with a wider frequency range, 100-400GHz, use smooth circular waveguides. The coupling to the fundamental waveguides of the instruments is provided by quasi-optical coupling to fundamental waveguides using matching mirrors and horn antennas. Quasi-optical units are built in such a way to allow for flexible configuration of input/output ports along with the capability of adjusting polarization and frequency separation. The overall setup results in an improvement of about 30dB for reflectometry that will enable broad band reflectometry for density profile measurement and ECE oblique experiments to be performed for the first time on JET.


Corresponding Author:

Cupido Luis

IST, Centro de Fusão Nuclear, Instituto Superior Técnico, 1049-001 Lisboa, Portugal

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-504 CONTROL PROCESS STRUCTURE OF ASDEX UPGRADE´S NEW CONTROL AND DATA ACQUISITION SYSTEM

Raupp, Gerhard, G. Neu, W. Treutterer, V. Mertens, D. Zasche, Th. Zehetbauer

Selected also for oral presentation O2B-D-504

ASDEX Upgrade´s new real-time CODAC was designed to demonstrate state-of-the art plasma control and operation methods required for advanced fusion machines. It is an open distributed system of controller and data acquisition nodes exchanging process data via a common real-time network. The software has two layers with underlying universal infrastructure functions and superior task specific application processes. The infrastructure provides signal exchange methods, alarm and log mechanisms (required on each node), and specific central processes for time and cycle management, system self- monitoring, and protocol extraction. The application layer consists of various application processes freely allocated onto controller nodes. The initial control implementation breaks down into: - plasma FF&FB with two processes, one for the feedback of position and shape via PF coils, and one for performance control with gas and heating systems - monitoring of machine and plasma with specific processes to check plasma position & shape, plasma performance & instability, machine protection systems, power supplies, coil currents, coil stress, and a supervisor process to take top-level decisions about the discharge sequence - data generation with processes to exchange signals with actuators or sensors, to compute equilibria and other , to evaluate complex input data, and to generate real-time reference values We will present the structure of the control processes, and show how these cooperate to form feedback control loops, to monitor machine and plasma, to take intelligent decisions about the discharge evolution, and to interact with protection systems to terminate the discharge in case of normal and abnormal operation.


Corresponding Author:

Raupp, Gerhard

Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, D-85748 Garching, Germany

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-506 MULTI-SUPPORT VECTOR MACHINES FOR DISRUPTION CLASSIFICATION IN TOKAMAK REACTORS

Mario Versaci, Antonino Greco Francesco Carlo Morabito

Facoltà Ingegneria Università Reggio Calabria Via Graziella Feo di Vito - I-89100 Reggio Calabria Italy

Disruption is a sudden loss of magnetic confinement that can cause damage to the machine walls and support structures. For this reason, it is of practical interest to be able to detect the onset of such an event early. The prediction problem can be expressed in terms of classification. Particularly, when a shot starts, it is imperative to know the kind of disruption. In this paper, a novel technique of classification of plasma disruption in tokamak reactor (JET) which use Support and Multi-Support Vector Machines (SVMs, M-SVMs) with Multi Layer Perceptron Neural Networks (MLPNNs) and Learning Vector Quantization (LVQ) is presented. Actually, in scientific literature, there are two methods for Multi-class Support Vector Machines. The first one is made by combination of different binary classifiers. The second one considers just one optimised relation. M-SVMs can be considered as a natural extension of SVMs for multi-class classification; in fact, SVMs classify by means a binary approach (two classes only) while M-SVMs solve problems in which data can belong to several classes. Training and testing data sets have been made choosing time sampling and particular kinds of shots for classifying disruptions. In addition, to improve the quality of classification, special signals have been take into account for our purpose (plasma current, mode lock,…). The obtained results show the goodness of the proposed approach, with respect to MLPNNs, in terms of percentage of false and missing allarms. Moreover, the reduced computational complexity of M-SVMs is useful for on-line applications especially in the case in which the classification of shot is a step of control task.


Corresponding Author:

Mario Versaci

Facoltà Ingegneria Università Reggio Calabria Via Graziella Feo di Vito - I-89100 Reggio Calabria Italy

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-513 OPTICAL DESIGN OF THE OBLIQUE ECE ANTENNA SYSTEM FOR JET

Carlo Sozzi (1), Alessandro Bruschi (1) Alessandro Simonetto (1) Elena DeLaLuna (2) John Fessey (3) Valeria Riccardo (3) and JET-EFDA Contributors

(1)Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, Milano, Italy (2)Asociación EURATOM-CIEMAT, CIEMAT, Madrid, Spain (3)EURATOM-UKAEA Association, Culham Science Centre, Abingdon, UK

The correct measurement of the plasma temperature has been an important issue since the beginning of thermonuclear fusion research. The introduction of Electron Cyclotron Emission (ECE) diagnostics as routinely available electron temperature measurement in tokamak and stellarators provided a major burst in the understanding of the plasma behavior in fusion relevant conditions. The availability of effective and powerful additional heating systems opened very high temperature scenarios bringing to light new phenomena related to the electron temperature measurements, actually arising the question of what exactly the diagnostic itself is measuring. In particular some systematic disagreements between ECE and Thomson Scattering diagnostics have been observed in the presence of NBI and ICRF heating (TFTR and JET), and ECR heating (FTU), probably due to deviation of the electron population from Maxwellian-bulk distribution. These observations have substantiated the proposal of the so called Oblique ECE diagnostics on JET, in which the ECE radiation is detected along lines of sight outside the poloidal plane. This layout allows the study of the electron distribution function at low energies revealing any non Maxwellian shape. This paper is devoted to the design of the quasi optical antenna for the Oblique ECE diagnostic. The physics requirements imply two lines of sight at about 10 and 20 degrees respectively in the toroidal direction. Severe geometrical constraints are imposed by the mounting method of the antenna, inserted in the vacuum vessel together with the group of six oversized waveguides devoted to Reflectometry and Oblique ECE itself and their surrounding structure. The two Oblique ECE waveguides share the same horizontal plane with one reflectometer waveguide, and are at the opposite side of the structure with respect to this. The antenna will be built with three flat mirrors and an ellipsoidal one, the last being shared by the two lines of sight. The mirror arrangement was optimized using electromagnetic calculations performed at several frequencies in the range of work foreseen for the diagnostic, extending from 100 to 400 GHz.


Corresponding Author:

Carlo Sozzi (1)

Istituto di Fisica del Plasma del CNR - Via R.Cozzi, 53 - 20125 Milano - ITALIA

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-515 ITER DIAGNOSTICS: MAINTENANCE AND COMMISSIONING IN THE HOT CELL TEST BED

Walker Christopher I., A.E.Costley, R.Gottfried(1), B.Haist(2), K.Itami, T.Kondoh, G.D.Loesser(3), J.Palmer(4), gie, A.Tesini, G.Vayakis

ITER International Team, (1) Framatome, (2) Oxford Technologies, (3) PPPL, (4)EFDA

ITER diagnostic equipment is integrated in 6 equatorial and 12 upper ports, 16 divertor cassettes and 5 lower ports. Diagnostic equipment in these locations is designed to be removed and then repaired, tested and commissioned in the hot cell area. In this paper the requirements and methods of repair and testing on these components are described. Design features that facilitate repair are included in diagnostic port plugs etc. Appropriate reception testing allows a repair strategy to be formulated to minimize hot cell time. All equipment to be reinstalled is checked as acceptable before embarking on the complex remote handling transport and installation procedure. At the hot cell a dummy port is provided for tests of mechanical and vacuum integrity as well as commissioning of the diagnostic equipment. The scope of the hot cell maintenance and commissioning activities is summarised and an overview of the integration of the diagnostic equipment is given.


Corresponding Author:

Walker Christopher I.

ITER International Team, Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2 , 85748 GARCHING,Germany

- D - Diagnostics, Data Acquisition and Remote Participation.

P2C-D-519 NEW BOLOMETRY CAMERAS FOR THE JET ENHANCED PERFORMANCE PHASE

McCormick Kent, A. Huber(1) C. Fuchs(2) C. Ingesson(3) J. Fink(2) W. Zeidner(2) A. Guigon(4) S. Sanders(1)

(1) EURATOM/UKAEA, Culham, Abingdon, UK (2) Max-Planck-Institut für Plasmaphysik, Garching, Germany (3) FOM Institute for Plasma Physics, Nieuwegein, The Netherlands (4) Close Support Unit - EURATOM, Culham, Abingdon, UK

JET is an experimental fusion device, the largest in the world, now undergoing a major upgrade. This enables replacement of the bolometer cameras – vertical and horizontal - used to register the temporal evolution and spatial distribution of radiation emanating from the plasma. The need is based on inadequate spatial coverage/resolution over the cross section. Namely, the plasma configuration has progressed from a state where radiation was distributed around the circumference or at the inside wall (limiter plasma) to that where it is concentrated within a poloidal region near the bottom of the vacuum vessel (diverted plasma). The current vertical camera, in operation since 1984, has become increasingly inadequate for the task. In particular, it is often impossible to produce satisfactory tomographic reconstructions of the divertor radiation pattern, although of primary interest for divertor scenarios. The new vertical camera has 24 channels, 16 covering the entire cross section and another set of 8 probing the divertor region - in contrast to 9-10 working channels of the old vertical camera (initially 14 channels). The new horizontal camera has 24 channels (vs. 20 for the old), including a subset of 8 for the divertor region and another 4 for the upper boundary. In addition to the increased number of viewing cones optimized for contemporary JET plasmas, new bolometer detectors and electronics are being employed, similar to those on ASDEX-Up, Tore-Supra and RFX. The new(old) detector is an 8(4)m gold absorbing layer placed on a 20(7.5)m mica(kapton) foil with a gold resistance meander on the backside. The thicker layer effects detection of higher energy quanta over 5eV-8keV. The 1.2kohm meander permits a higher operation voltage (40Vp-p vs. 10V). This, together with use of phase-locked techniques at a carrier frequency of 50kHz (dc for old cameras) and an improved grounding/shielding concept, will lead to an improved (>>10) signal-to-noise ratio and time resolution (<20ms). The paper will discuss salient aspects of the new cameras - to be mounted on JET in Sept./Oct. - showing examples of the enhanced tomographic capabilities. Design criteria demanded of in-vessel diagnostics on JET will be addressed: a) the ability to withstand large forces associated with abrupt termination of the plasma current and b) the nearly reactor-level quality control during construction - dictated by the tritium-phase of JET operation.


Corresponding Author:

McCormick Kent

Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching, Germany

- E - Magnets and Power Supplies.

P2T-E-9 THE BATCH PRODUCTION FOR SUPERCONDUCTING MAGNET COILS OF EAST (HT-7U)

Gao Daming, Chen Siyue, Yu Jie, Wu Jiefeng, Pan Yinnian, Wu Weiyue, Li Baozeng, Wen Jun, Zhang Ping, Zhu Wenhua, Tao Yuming, Pan wanjiang, Wu Yu

Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126 Hefei 230031, P.R.China

Abstract: The construction of Experimental Advanced Superconducting Tokamak ( EAST ), approved by Chinese government as one of Chinese national mega project of science research is smoothly underway according to its schedule. The core of EAST Project is superconducting magnet coils (SMC), composed of 16 toroidal field coils (TFCs) and 14 poloidal field coils (PFCs). The prototypes for SMC had successfully been completed and passed the cryogenic tests early of 2003. Since then the batch production for SMC have begun at ASIPP. Up to now, 44 CIC conductors have been completed among the 58 CIC conductors in total; 26 half TFCs and 8 PFCs have been completed separately among the 34 half TFCs and 16 PFCs including the model and spare coils; other correlative procedures are also made at same time. Nearly 2/3 workloads for SMC have been performed, and whole will be finished late of 2004. This paper emphasizing on the various technology issues that must be faced and solved for 4 R&D lines of SMC after translating to batch production. To describe the optimization of welding program for conduit joint, decreasing protrusion height inside, limiting insertion gap, setting real NDT methods on CICC line. Also to describe construction of 2 new addition winding lines, fabrication of U-shape current joint between 2 half TFCs, welding helium stub on conductor in site, protecting cable over-heat and damage; sealing TF case, making twice VPI (vacuum pressure impregnation) treatments for each TFC in special oven and once VPI treatment in site for each of four big diameter PFCs based on a series of experiments; Quality control methods during batch production process for SMC are also discussed.


Corresponding Author:

Gao Daming

Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126 Hefei 230031, P.R.China

- E - Magnets and Power Supplies.

P2T-E-20 STUDY ON HIGH-POWER HIGH-FREQUENCY INVERTER FOR FAST PLASMA POSITION CONTROL IN EAST SUPER-CONDUCTING TOKAMAK

LIU Zheng-Zhi, Y. Yu(1), X. Zhang(2), R. C. Cheng(1), S. S. Lu(1), Z. Wei(1), C. W. Zhang(2)

(1) Institute of Plasma Physics, Chinese Academy of Sciences, P. O. Box 1126, Hefei 230031, P. R. China (2) Hefei University of Technology, Hefei 230009, P. R. China

In Experimental Advanced Super-conducting Tokamak (EAST*), the fast plasma position control is of fundamental importance for suppression of inherent vertical instability of elongated plasma. An active control system must be incorporated to compensate for resistive decay of the eddy currents of Passive Stabilizing Plate (PSP) and to maintain plasma at a reference vertical position. The Power Supply for Fast plasma position Control (FCPS) is to energize the active control system and to realize the fast current tracking under the command from central control for plasma vertical position control in real time. The technical requirement of FCPS and the parameters of the Internal Vertical Coils (IVC) in preliminary design will be introduced. The preliminary design, digital simulation and principle experiment of FCPS will be presented. R & D on AC/DC/AC converter topology and its control has been made progress. The current-source, three-phase PWM rectifier (AC/DC) with current vector control, and the multi-parallel and phase-shifting PWM H-bridge inverter (DC/AC) topology have been studied and developed. The principle experiment has been carried out and it shows good agreement with the design and simulation. It has been shown that the high-power high-frequency current-source PWM converter (AC/DC/AC) will be satisfied with all the requirements of FCPS and has some unique advantages as the followings: 1. Four-quadrant operation (Bi-directional current) 2. Fast current tracking----up to 100HZ and even higher 3. Accurate regulation----Error of current tracking less than 5% 4. Easy for capacity enlargement and redundancy design----up to 10 MW 5. Multi-parallel and phase-shifting PWM to realize high frequency modulation with low switching frequency of devices 6. Sine wave and Unit PF in AC input 7. High feasibility, high reliability and high flexibility *The former name of EAST was HT-7U super-conducting Tokamak in CAS-IPP


Corresponding Author:

LIU Zheng-Zhi

P. O. Box 1126, Hefei 230031, P. R. China

- E - Magnets and Power Supplies.

P2T-E-23 A LOW COST JOINT FOR THE ITER PF COILS, DESIGN AND TEST RESULTS.

Stepanov Boris, Bruzzone Pierluigi, Vogel Martin

CRPP, Villigen-PSI, Switzerland

The Poloidal Field (PF) coils of the International Thermonuclear Experimental Reactor (ITER) are designed with use of NbTi cable-in-conduit conductors carrying a current up to 45 kA. Each PF coil consists of several modules wound in double pancakes, joints connect that double pancakes electrically in series at the coil periphery. In order to minimize the manufacturing cost maintaining a high reliability, a low cost joint design is developed at CRPP and tested in SULTAN test facility in frame of ITER task for PF coils. The concept of the joint is a single fully welded stainless steel box and one-pitch joint through sectioned copper saddle blocks. The joint sample includes two full-scale joints, one in the hairpin configuration at the bottom of the sample and one as an overlap joint at mid length of the sample. Lifting the sample in the SULTAN facility, either the hairpin or the overlap joint can be placed in the center of the facility and tested under ITER-relevant operating conditions. The low cost joint design for the ITER PF coils is described and the test results (DC and AC loss) are presented in this paper.


Corresponding Author:

Stepanov Boris

CH-5232 Villigen PSI, WMHA/C23

- E - Magnets and Power Supplies.

P2T-E-30 130KV 130A HIGH VOLTAGE SWITCHING MODE POWER SUPPLY FOR NEUTRAL INJECTORS - CONTROL ISSUES AND ALGORITHMS

Ganuza, Daniel, García, Francisco (JEMA) Zulaika, Mikel (JEMA) Perez, Albert (JEMA) Jones, Timothy (EURATOM/UKAEA Fusion Association)

JEMA: Paseo del Circuito 10. E-20160 Lasarte - Oria, Spain EURATOM/UKAEA Fusion Association: Culham Science Centre, Abingdon, OX14 3DB UK

The company JEMA has delivered to the Joint European Torus (JET facility in Culham) two High Voltage Switching Mode Power Supplies (HVSMPS), each rated at 130kVDC and 130A, which will feed the grids of two PINI loads. This paper describes the main control issues and the algorithms developed for the project. The most demanding requirements, from the point of view of the control are an absolute accuracy of +/- 1300V and the possibility of performing up to 255 re-applications of the high voltage during a 20 second pulse. Keeping the output voltage ripple to the specified tolerance has been a major achievement of the control system. Since the output stage is formed of several modules (120) connected in series, their stray capacity to ground significantly influences the individual contribution of each single module to the global output voltage. Two complementary techniques have been used to balance the effects of the stray capacities. On the one hand, a study has been carried out in order to find the optimum firing sequence of the 120 modules of the output stage, considering the distribution of capacitances. On the other hand, an active ripple compensation algorithm has been implemented. The fast re-applications requirement has a significant impact on the intermediate DC Link stage. Such section is composed of a capacitance of 0.83 Farads at 650V, which feeds the 120 output stage modules The DC Link is fed by a 12 pulse SCR rectifier, whose matching transformers are connected to the 36kV Grid. Every re-application and every voltage shutdown supposes a quasi-instantaneous power step from zero to 17 MWatt load and vice-versa. Fast open loop algorithms have been implemented in order to keep the DC Link voltage inside acceptable margins. Moreover, the HVSMPS output characteristics have to be maintained during the rapid and important voltage fluctuations of the 36kV mains (28kV – 37kV). The general control system is based on a Simatic S7 PLC, and a SCADA user interface. Up to 1000 signals are considered. The control system has demonstrated to allow for a rapid and accurate identification of faults and their origin.


Corresponding Author:

Ganuza, Daniel

JEMA. Paseo del Circuito 10. E-20160 Lasarte - Oria (Spain)

- E - Magnets and Power Supplies.

P2T-E-31 FIBERGLASS UNIDIRECTIONAL COMPOSITE TO BE USED FOR ITER PRE-COMPRESSION RINGS

Nardi Claudio, Livio Bettinali (*) Aldo Pizzuto (*)

(*) ENEA - Via E. Fermi 45 - 00044 Frascati (Roma)

The ITER magnet system will be kept in place by a system of pre-compression rings. These rings have stringent requirements as far as the material requirements are concerned. They must assure the mechanical strength and stiffness characteristics at both the cryogenic temperature (4¢XK) and room temperature. The best solution, at present knowledge, is to make rings in composite material, using unidirectional glass fibers and a resin matrix. ENEA performed a series of impregnations in order to study the characteristics of such a material both at room temperature and at liquid nitrogen (77¢XK) temperature. The used glass was an S-2 glass, because of his good mechanical characteristics, and the matrix was epoxy resin. The glass, in form of fibers having 14 ƒÝm diameter and tex 725, have been wounded on the mould, and vacuum impregnated with resin and afterwards cured at a temperature of 140¢XC. From the moulds standard specimens, according to ASTM D 3039, and non-standard specimens have been obtained. The values of the mechanical strength have been as high as 2200 MPa, but, most relevant result, the linearity of the behaviour was kept practically until the failure. This last issue is very relevant, at it assures a behaviour independent from the previous load history of the component. In order to evaluate the time-dependent characteristics of the material, a set of specimens will be tested at room temperature with the 70-80% of the collapse load, in order to verify the creep behaviour of the material. A facility is in course of manufacturing, in order to test, in operating conditions and up to stresses leading to the rupture, rings in scale 1/10 compared to the true dimensions of the ITER rings. These tests, once the facility will be available, will give information about the behaviour of massive structures in unidirectional fibreglass composite in order to supply information needed for the design of the ITER rings.


Corresponding Author:

Nardi Claudio

ENEA - Via E. Fermi 45 - 00044 Frascati (Roma)

- E - Magnets and Power Supplies.

P2T-E-34 MEASUREMENT OF CONTACT RESISTANCE DISTRIBUTION IN TYPICAL ITER SIZE CONDUCTOR TERMINATION

Anghel Alexander, Bruzzone Pierluigi

EPFL/CRPP Fusion Technology, CH-5232 Villigen, Switzerland

A new test facility, JORDI, dedicated to the measurement of contact resistance distribution in full-size conductor termination at liquid helium temperature, has been designed and manufactured at CRPP Fusion Technology in Villigen, Switzerland. The new facility is composed of a medium size liquid helium cryostat which host a full-size termination sample, a cryogenic, parallel resistor array for providing balanced current injection in up to 100 channels from a 10kA DC power supply (100 channels of each 100A), a pair of conventional current leads rated at 10kA and a data acquisition system with 200 analogue input channels (100 voltage and 100 current readouts). Recently, the facility was used to characterize and qualify a first prototype of a typical ITER joint. As opposed to the earlier attempts where the contact resistance was measured under an unbalanced current distribution i.e. only one strand or a small group of strands were connected to the current source while all other strands were not charged by a current, in the new facility the contact resistance is measured under imposed balanced current distribution among all current carrying elements of the termination. Also in the new facility all voltage drops are sensed simultaneously as opposed to the earlier measurements where the voltage drops were measured sequentially. The results of the measurements and interpretation based on a simple statistical, electrical network termination model are presented.


Corresponding Author:

Anghel Alexander

EPFL/CRPP Fusion Technology, CH-5232 Villigen, Switzerland

- E - Magnets and Power Supplies.

P2T-E-35 UPDATING THE DESIGN OF THE FEEDER COMPONENTS FOR THE ITER MAGNET SYSTEM

Yoshida Kiyoshi, Takahashi Yoshikazu (1), Isono Takaaki (2), Mitchell Neil(1)

(1) ITER, Naka Joint Work Site (2) Dept. of Fusion Engineering Research, Japan Atomic Energy Research Institute, Japan

The ITER superconducting magnet system stores energy from 40 to 50 GJ during plasma operation, and generates an average heat load of 23.4 kW at 4.3 K to cryoplant. This heat load is removed by a primary supercritical helium circuit with a flow of 6.0 kg/s of helium. The helium is distributed to the coil through a complex system of 30 separate feeder lines. The feeders also contain the electrical supplies to the coil and are integrated into the current lead transition to room temperature. The interface components between the coils and the service facilities (power supply and cryogenic plant) consist of the in-cryostat feeders, the cryostat feedthroughs, and the coil terminal boxes (CTBs). The layout of the in-cryostat feeders takes into consideration routing restrictions in the cryostat and initial assembly with other Tokamak components. The electrical break boxes on the coil surface are designed for accessibility and manual repair handling in case of failure. The cryostat feedthroughs with S-bend boxes allow thermal contraction of the magnet system. The forced-flow-cooled current leads in the CTBs are adapted to fit in the limited space in the building. A conventional copper heat exchanger is used for the current lead base design although high temperature superconducting leads could also be used if the performance would be improved relative to the base design. Ground faults and short circuits are the main potential accidents in the superconducting feeders. A double insulation and monitoring system together with robust mechanical containment will help prevent and /or reduce the impact of such accidents. This paper presents the latest design concept and parameters of the feeder components.


Corresponding Author:

Yoshida Kiyoshi

ITER, Naka Joint Work Site, 801 Mukaiyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193, Japan

- E - Magnets and Power Supplies.

P2T-E-36 MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: COMPUTATION OF THE BACKGROUND FIELD AND CONSEQUENCES ON THE DESIGN OF THE ELECTRICAL DISTRIBUTION BOARDS AND CONTROL BOARDS FOR THE ITER TOKAMAK BUILDING

Benfatto Ivone, P.Bettini (1) M.Cavinato (2) A. De Lorenzi (2) D. Desideri (3) P. Fejoz (4) L. Grando (2) P. Hertout (4) J. Hourtoule (4) D. van Houtte (4)

(1) University of Udine, Itay (2) Consorzio RFX - Association EURATOM ENEA, Padova, Italy. (3) University of Padova, Italy (4) Association EURATOM-CEA, St Paul Lez Durance, France.

The electrical distribution boards and control boards located inside the ITER Tokamak building, are subjected to constant, or slowly variable, magnetic field up to 70 mT, 10 mT/s. This is a very unusual environmental condition for the components of the electrical installations, therefore very limited data are available on the magnetic field compatibility of standard electromechanical and electronic components for low voltage distribution boards and control boards. Being this information a necessary input for the design of the electrical installation inside the ITER Tokamak building, EFDA placed two specific contracts dedicated to the execution of experimental campaigns addressed to collect data on the magnetic field compatibility of standard industrial components for electrical distribution boards and control boards. Several components of different manufacturers have been tested and a large amount of data have been collected. The test procedures and the results are reported in dedicated parallel papers presented at this conference by CEA and Consorzio RFX (ENEA): the two Euratom Associations in charge of this experimental investigation. In parallel to the test campaigns, EFDA promoted other activities, which are the subject of this paper and have been dedicated to the following theoretical investigations: 1. to update the magnetic field map taking into account the latest ITER reference layout of the Tokamak building; 2. to assess whether the steel embedded in building structure produces significant effects on the magnetic field map in the areas dedicated to the installation of the electrical distribution boards and control boards; 3. to assess the feasibility of magnetic shields, to mitigate strong constraints in the design, manufacturing and installation of the boards. The paper reports on the computation methods and the results obtained from the above theoretical activities. The consequences on the layout of the electrical installations inside the ITER Tokamak building are also discussed in the paper together of the recommendations for the design of the electrical distribution boards and control boards of the ITER Tokamak building.


Corresponding Author:

Benfatto Ivone

EFDA-CSU Garching, Boltzmannstr. 2, D-85748 Garching, Germany

- E - Magnets and Power Supplies.

P2T-E-37 COMMISSIONING OF THE 10 POWER SUPPLIES OF THE CONTROL COILS OF WENDELSTEIN 7-X EXPERIMENT

JAUREGI EDUARDO, T. Rummel, F. Füllenbach

MAX-PLANCK-INSTITUT FÜR PLASMAPHYSIK, EURATOM ASSOCIATION D-17491 Greifswald, Wendelsteinstr. 1. Germany

COMMISSIONING OF THE 10 POWER SUPPLIES OF THE CONTROL COILS OF WENDELSTEIN 7-X EXPERIMENT E. Jauregi, D. Ganuza, I. García, J.M. Del Río, J. Lucas JEMA GJ 20160 Lasarte-Oria, Spain T. Rummel, F. Füllenbach MAX-PLANCK-INSTITUT FÜR PLASMAPHYSIK, EURATOM ASSOCIATION D-17491 Greifswald, Wendelsteinstr. 1. Germany In the region of Greifswald, north-east of Germany, the so-called Wendelstein 7-X Experiment is progressing forward at the Max-Planck Institute for Plasma Physics, IPP, to start operation into next years and become the Europe´s biggest project on “Advanced Stellarators”. For the confinement of plasma some superconducting main field coils are placed, while for the relative positioning 10 smaller coils, control coils, will be used. Each coil must be independently supplied, by a high and precise direct current power supply, superposed to ac current frequency modulated. The contract for the turn-key power supply for the control coils was awarded to the Spanish company, JEMA, which designed, manufactured and tested the ten power supplies as well as the de-mineralised water cooling plant, main overall control station and distribution centre. Actually the ten power supplies are already tested and commissioned onsite, using some dummy loads as final load, and waiting for the last combine test joined to the IPP general control system. Present abstract and future paper will refer to the results got at the partial acceptance tests as well as mains problems occurred during commissioning, and proposed solutions. Each of the ten power supplies should provide a controlled current compounded dc and 0-20 Hz bandwidth ac current in a range of almost 3 kA at low voltage, 30 V, in four quadrants. Stability, precision and very low output ripple are required. One of the most critical test was to check the output ripple measured at load for the output voltage, specified at 1 Vpp, and for current, 1 App specified. Such a long cable distance from power converter up to dummy load, maximum 30 meters, and the extremely low current ripple requirement, lower than 0,05%, made very hard the test. Several testing devices and solutions were discussed to finally get a promising result of 0,02% peak to peak current ripple at full load. E


Corresponding Author:

JAUREGI EDUARDO

Pº DEL CIRCUITO, 10 20160 LASARTE-ORIA (SPAIN)

- E - Magnets and Power Supplies.

P2T-E-40 DESIGN AND COMMISSIONING OF THE NEW TOROIDAL FIELD COIL FOR THE NATIONAL SPHERICAL TORUS EXPERIMENT (NSTX)

Neumeyer, Charles, E. Baker, A. Brooks, J. Chrzanowski, L. Dudek, P. Heitzenroeder, C. Jun, M. Kalish, T. Kozub, R. Marsala, R. Parsells, B. Paul, H. Schneider, M. Williams, I. Zatz

One of the key features, but also one of the most challenging aspects of the spherical torus (ST), is the demountable toroidal field (TF) assembly which permits removal of the entire inner leg and center stack assembly for maintenance. On February 14, 2003, following the morning test shots, the NSTX TF Inner Leg Assembly failed at the lower Inner Leg-to-Flag joint. Analysis of the event identified shortcomings in the structural design of the joint which led to failure after some 7200 machine pulses, with a limited number at the full rating of 6kG. The stiffness of the structural assembly was not adequate, and repeated application of the electromagnetic loads led to unanticipated loads in the bolts, and high local current density which eventually led to failure. Due to the extensive damage it was not possible to recover the original TF inner leg assembly. Furthermore, it was clear that an improved design was needed. Therefore a recovery effort was initiated, beginning with the development of a new design. Extensive engineering resources were applied to the re-design effort to ensure a successful outcome while minimizing the time duration of the recovery period. Extensive finite element analysis was performed to develop an understanding of the structural and thermal behavior of the joint and to guide the development of the new design. Tests were performed to characterize the electrical resistivity of the joint vs. pressure, the friction coefficient of the joint, the pull-out strength of the fasteners, and other features. In addition, a mechanical prototype was exercised at the rated number of cycles of full mechanical loads at elevated temperature, and an electrical prototype was tested at full current for the full time duration at full mechanical loads. Based on the new design, the successful prototype testing, and the improved instrumentation which includes a new fiber optic strain, temperature, and displacement monitoring system, reliable operation at full rated parameters is fully anticipated. Operations at 4.5kG has been re-established, and extensive measurements have been taken, which are presently under review. Although the failure was unfortunate, it has led to an improved understanding of the TF joint behavior which is directly applicable to the design of next step STYLE=" devices. This paper describes the new design, and the commissioning of the new coil.


Corresponding Author:

Neumeyer, Charles

Princeton University Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey, 08543, USA

- E - Magnets and Power Supplies.

P2T-E-48 ANALYSES AND IMPLICATIONSOF V-I CHARACTERISTIC

Rainer Wesche (1), Alexander Anghel (1) Pierluigi Bruzzone (1) Paola Gislon (2) Luigi Muzzi (2)

(1)CRPP-FT, CH-5232 Villigen-PSI, Switzerland (2)ENEA, Centro Ricerche Frascati, Via E. Fermi 45, 00044 Frascati, Rome, Italy

Two short lengths of the NbTi cable-in-conduit conductor used to fabricate the poloidal field coil insert will be tested in the SULTAN facility. The two conductor lengths to be investigated are distinguished by the presence or absence of the subcable wraps. An aspect to be discussed in more detail is the voltage-current characteristic of the two NbTi cable-in-conduit conductors. Previous results obtained for three NbTi subsize cable-in-conduit conductors with one or two out of four petals disconnected indicate that the n factor, defined by the power law for the electric field, is closely related to the current distribution and current transfer effects in the cable. The presence of the subcable wraps can considerably hinder the current transfer between the individual strands in the cable. This effect may lead to changes in the voltage-current characteristic. The implications of the results on the poloidal field coil conductor design will be considered. Finally, the effect of the self-field on the dc performance and the quench behaviour will be addressed. Due to the variation of the magnetic field within the conductor cross-section the peak electric field occurring on the high field side of the conductor is much larger than the average electric field. As a consequence the conductor quenches in the peak field region. The measured quench currents and take-off electric fields will be simulated assuming peak-field-induced quenches. The dc performance of the poloidal field coil NbTi conductors will be compared to previous results obtained for five NbTi subsize cable-in-conduit conductors with parametric variations in the conductor layout.


Corresponding Author:

Rainer Wesche (1)

CRPP-FT, CH-5232 Villigen-PSI, Switzerland

- E - Magnets and Power Supplies.

P2T-E-50 PIONEERING SUPERCONDUCTING MAGNETS IN LARGE TOKAMAKS: EVALUATION AFTER 17 YEARS OF OPERATING EXPERIENCE

Duchateau Jean-Luc, B. Gravil, M. Tena, D. Henry, D. Van Houtte

CEA Cadarache F-13108 Saint Paul Lez Durance Cedex FRANCE

Selected also for oral presentation O2A-E-50

As a part of the Euratom program, it was decided at the beginning of the eighties that Europe will build a large size tokamak Tore Supra, using a superconducting toroidal field magnet to demonstrate the applicability of superconducting magnets to future fusion reactors. Part of the Physicist community was at that time reluctant to consider this possibility, being afraid of possible difficulties in exploitation of the machine. Especially the operation of a large refrigerator with thousands litres of 1.8 K helium was considered as completely unrealistic and industrially impracticable. As a matter of fact, 17 years after the first plasmas, the fusion Community is happy to have accumulated this first experience with superconducting magnets, especially at the time where ITER is on the verge of being launched. Moreover no large fusion magnet is now considered in the world without superconducting magnets. Far from being a burden in the exploitation, the availability of the TF system all the day long for Plasma Physics is on the contrary of great help for the implementation of long shots. The absence of large mechanical cycling, is also a guarantee for the good operation of the coils on the long run. After these 17 years, the specific impact on a Tokamak operation of such a large system at low temperature will be analysed and detailed: - Influence of daily current increase on coils temperature - Influence of plasma shot on coils temperature - Influence of plasma disruption on coils temperature - Influence of plasma disruption on voltage quench detection. - Specific impact of long runs (500 s) on superconducting system - Behaviour of the coils during fast safety discharges (FSD) - Duration of coil recooling after night for a typical operation day - Warming up and cooling down the coil for cryogenic system maintenance Quantitative data will be given, of the TF for the cryogenic system and for the magnet system as well, concerning the number of plasmas shots and the availability of the machine. The origin and the number of breakdowns or incidents will be described, with emphasis on cyogenics to document repairs and changes on the system components. Overall, despite the differences in design and size, the accumulated experience of 17 years of operation is a useful tool to prepare the manufacture and the operation of the ITER magnets.


Corresponding Author:

Duchateau Jean-Luc

CEA Cadarache F-13108 Saint Paul Lez Durance Cedex FRANCE

- E - Magnets and Power Supplies.

P2T-E-55 STABILITY, THERMAL EQUILIBRIUM AND DESIGN CRITERIA FOR CABLE-IN-CONDUIT-CONDUCTORS WITH A BROAD TRANSITION TO NORMAL STATE

Nicolai Martovetsky,

Stability in CICC (cable-in-conduit conductors) against perturbations is often associated with transient heat removal of heat generated in the normal zone, which appeared in CICC as a result of a strong perturbation. In such a transient condition, a simplified approach to stability calls for a sufficient amount of copper in the strands, with sufficiently small diameter, such that the heat removal is higher than the heat generation. This criterion is often used for design of the fusion magnets, like ITER, KSTAR and others. We show that this criterion is not a mandatory requirement for serviceability of CICC and that CICC may work reliably at higher current densities. In conditions of limited perturbations, a sufficient stability is provided not by a large amount of copper and high transient heat transfer, but by a smooth transition to the normal state and ease of current redistribution. A strand parameter space for CICC stability is proposed and discussed. The theory predictions are compared with known experimental data on CICC that meet and do not meet this design criterion.


Corresponding Author:

Nicolai Martovetsky

LLNL, 7000 East Ave, L-641, Livermore, CA, 94550, USA

- E - Magnets and Power Supplies.

P2T-E-68 DESIGN OPTIMISATION OF THE ITER TF COIL CASE AND STRUCTURES

Marco FERRARI (1), Pietro BARABASCHI (1), Cornelis T.J. JONG (1), Reinhard K. MAIX (2), Neil MITCHELL (3)

(1) ITER International Team, Boltzmannstr. 2, D-85748 Garching, Germany (2) ITER VHTP, ATI Atominstitut Wien, Stadionallee 2, A-1020 Vienna, Austria (3) ITER International Team, 801-1 Mukouyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan

The basic ITER toroidal field (TF) coil case design was defined in 2001 and since then has been undergoing a process of refinement and optimisation. The performance and major geometry of the TF coil case, which encloses the winding pack, has remained unchanged, but significant improvements have been made, first to the structural support, relaxing the non-destructive testing inspection levels to achieve the fatigue life, secondly to the material options and main fabrication steps, reducing the fabrication time and cost, and thirdly to design definition of auxiliary systems. On the structural design, the system of poloidal keys, linking the coils at top and bottom of the central vault, has been optimised, virtually eliminating stress concentrations in the keyways. A choice has been made between the two options for the intermediate outer intercoil structures (friction-joint design is preferred over box), eliminating also the consideration of castings for some coil sections. The design of the friction-joint panels and the mechanical connection to the adjacent TF coils have been optimised, providing access to perform two-sided welding of the panels during machine assembly. The reference system of rings for pre-compressing the inner legs of the TF coils has been confirmed as uniaxial glass fibre based on recent R&D results and the design has been modified to allow re-tightening of the rings without the need to remove the central solenoid. For the fabrication, the main components (the TF coil case sub-assemblies before winding pack insertion and subsequent closure welding) have been zoned into three material grades of austenitic stainless steel, minimising the quantity of the top grade that requires electro-slag refining to achieve the specified performance. The final assembly of the winding pack into the case has been redefined to reduce distortion and the risk of insulation damage due to the closure welds while minimising case-winding pack gaps in the peak stress regions. Design of auxiliary systems has included the case cooling inside and outside, including a thermal screen to protect the peak field region of the conductor from the maximum nuclear heating. The paper reports the final design, including the reasons for the design choices that have been made. Structural assessments of the critical components are also shown.


Corresponding Author:

Marco FERRARI (1)

ITER International Team, Boltzmannstr. 2, D-85748 Garching, Germany

- E - Magnets and Power Supplies.

P2T-E-73 FABRICATION OF THE PLANAR COILS FOR WENDELSTEIN 7-X

Viebke, Holger, Th. Rummel (1) K. Riße (1) R. Schroeder (1) R. Winter (2)

(1) Max-Planck-Institut für Plasmaphysik, Greifswald Branch, Euratom Association, Wendelsteinstraße 1, D-17491 Greifswald (2) Tesla Engineering Ltd., Water Lane, Storrington, Sussex RH30 3EA, England

WENDELSTEIN 7-X (W7-X) is a superconducting stellarator, which uses 50 non-planar coils for the main field and 20 planar coils to modify the magnetic configuration. The coils are arranged in five modules requiring five differently shaped non-planar and two differently shaped planar coils. The magnet system is designed for 3 T on the plasma axis. Nominal currents of the non-planar coils are 17.6 kA against 16 kA for the planar coils. One planar coil has an outer diameter of around 4 metres. The main elements of planar coils are the winding packages made of a cable-in-conduit superconductor, a coil case made of stainless steel plates, the embedding filler material, two interlayer joints to connect the double layers and a case cooling using copper plates and stainless steel pipes. Connection of the coil to the coil support structure is performed through forged blocks welded to the casing and bolts. Quench detection, temperature sensors and strain gauges are installed to control operation. Manufacturing of the planar coils is contracted to the company Tesla and has to be performed with a high accuracy to maintain the required symmetry of the magnetic configuration of W7-X. A tolerance of 0.2 mm is allowed for the machined surfaces as compared to the CAD-model. The accuracy of the coils is surveyed by photogrammetry. All steps of production are rigorously controlled by quality assurance. Prior to dispatch the coils will pass a works acceptance test at Tesla thereby demonstrating helium leak tightness, resistance against high voltage and the specified flow-resistance of the cooling channels. Prior to delivery to Greifswald, all coils will be subject to a functional test at cryogenic temperatures at the Low Temperature Lab of CEA. By March three coils have been delivered and one has passed successful the test at nominal current and have shown sufficient margin. The presentation will give an overview about the status of production and address major technical problems, which had to be solved.


Corresponding Author:

Viebke, Holger

Max-Planck-Institut für Plasmaphysik, Greifswald Branch, Euratom Association, Wendelsteinstraße 1, D-17491 Greifswald

- E - Magnets and Power Supplies.

P2T-E-81 OVERVIEW OF THE DIII–D INTERNAL RESISTIVE WALL MODE STABILIZATION POWER SUPPLY SYSTEM*

Szymanski, D.D., G.L. Campbell (1), W.P. Cary (1), R. Hatcher (2), G.L. Jackson (1), A.G. Kellman (1), A. Nagy (2), and C.J. Pawley (1)

(1) General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (2) Princeton Plasma Physics Laboratory, Princeton, New Jersey

With the recent installation in the DIII-D Tokamak of internal resistive wall mode (RWM) stabilization coils (I-Coils), upgrades to the existing RWM and error field correcting power supply systems were necessary. The new I-Coil system is comprised of 12 individual low inductance magnetic field coils that can be rearranged in multiple configurations with the main purpose of providing feedback stabilization up to the ideal wall beta limit without the need for strong plasma rotation. This paper will discuss the power supply system upgrades needed to drive up to 5 kA in these low inductance coils. The power supply system is now comprised of (5) 300 Vdc, 5-7 kA pulse rated power supplies which can either be connected directly to magnetic coils or else provide input power to (4) 300 Vdc, 5 kA pulse rated switching power amplifiers (SPAs). The SPA actuators, when connected to the I-Coils provide maximum current from dc to 300 Hz and can operate up to 2 kHz at reduced current, limited by the inductance of the I-Coils and their cable feeds. In some experimental scenarios faster response with lower phase shift is required than can be provided by the SPAs. In this case, high power audio amplifiers will be installed. We will present the details of the upgraded power system including instrumentation, data acquisition, multiple SPAs powered by a single dc supply, a versatile patch panel, and low inductance cabling. In addition, the design of audio amplifiers will also be discussed. *Work was supported by the U.S. Department of Energy under DE-FC02-04ER54698 and DE-AC02-76CH03073.


Corresponding Author:

Szymanski, D.D.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- E - Magnets and Power Supplies.

P2T-E-95 THE EUROPEAN DEVELOPMENT OF A FULL SCALE SWITCHING UNIT FOR THE ITER SWITCHING AND DISCHARGING NETWORKS

Bonicelli Tullio, A. De Lorenzi (1) E. Gaio (1) D. Hrabal (2) F. Milani(1) R. Piovan (1) E. Sachs (2) E. Salpietro (4) S. Shaw (3) V. Toigo(1) L. Zanotto (1)

(1) Consorzio ENEA-RFX, Italy (2) FEAG, Germany (3) UKAEA, UK (4) EFDA-CSU Garching

Selected also for oral presentation O2A-E-95

The ITER coil power supply systems are provided with Discharging Networks whose main purpose is to dissipate in resistors the magnetic energy stored in the super conductive coils when a quench is detected. Similarly, Switching Networks are series connected to some of the poloidal coils and to the central solenoid to produce the required loop voltage during plasma start-up. These actions are performed diverting the current flowing in a switch into a resistor connected in parallel. The European Fusion Programme included since the mid-90’s the development of a full scale, full rating switching unit to be used as centre-piece of the ITER Commutating Units. The main rating of such units are: steady state and breaking current of 70 kA, peak withstand current 250 kA, rated voltage 17.5 kV rms, recovery voltage 24 kV. Since the current carrying capability in steady state of a vacuum circuit breaker (VCB) is far from been sufficient for the steady state operation, the combined operation of a mechanical by-pass switch (BPS), rated for the continuous current, and a vacuum circuit breaker has been proposed and developed. During the first phase of the activity (years 1995-1999), the VCB and the BPS were individually characterised and tested at their rated performances, including the execution of full scale life testing. An important limitation on the maximum I2t in the VCB before opening was identified. The second phase, concluded at the beginning of 2004, was devoted to check for the first time the combined operation of the by-pass switch and the vacuum circuit breaker up to the full performances. Some minor improvements of the switches were also tested, coming from the results of the first phase. In the paper the successful results of the type and life tests will be presented, including the operation in absence of the saturable reactor series connected in the ITER reference design, which could yield some cost and space savings. Interruption tests at low current and full counterpulse capacitor voltage were also successfully performed. The testing included an extensive and novel characterisation of the interrupting capability of the vacuum circuit breaker in the presence of an external magnetic field as in the actual location of installation in ITER, where a stray magnetic field of up to 25 mT will be present.


Corresponding Author:

Bonicelli Tullio

EFDA-CSU Garching, Boltzmannstr. 2, D-85748 Garching, Germany

- E - Magnets and Power Supplies.

P2T-E-97 MECHANICAL PERFORMANCE OF MAGNET INSULATION MATERIALS FABRICATED BY THE “INSULATE-WIND-AND-REACT “ TECHNIQUE*

Dr. Humer Karl, Karin Bittner-Rohrhofer (1) Karl Humer (1) Harald Fillunger (1) Reinhard K. Maix (1) Harald W. Weber (1)

(1) Atomic Institute of the Austrian Univ. Stadionallee 2 1020 Vienna Austria

Usually, superconducting magnet coils are fabricated according to the standard “Wind-React-Insulate-and-Transfer” technique, where the superconductor is wound and heat treated first, before the insulation is applied and the coil vacuum impregnated with epoxy. In order to simplify the manufacturing process of such coils and to lower the costs, an alternative procedure, the “Insulate-Wind-and-React” technique can be used. In this case, the superconductor is insulated first, followed by winding, heat treatment and impregnation of the coil, i.e. the “transfer”-step can be avoided. Such an insulation system fabricated by European industry (Ansaldo, Italy) has been investigated. It consists of a two-dimensional R-glass-fiber reinforcement heat treated at 650 C and impregnated afterwards with epoxy. In order to characterize the mechanical material performance, both tensile and short-beam shear (SBS) tests were carried out at 77 K. The ultimate tensile strength is about 500 and 250 MPa parallel and perpendicular to the glass fibers, respectively. Furthermore, tension-tension fatigue tests were done to simulate dynamic load conditions caused by the Lorentz forces. In addition, a set of SBS samples, irradiated to a fast neutron fluence of 1x1022 m-2 (E>0.1 MeV), is in order to check for material degradation induced by radiation. *This work has been carried out within the association EURATOM-OEAW.


Corresponding Author:

Dr. Humer Karl

Atomic Institute of the Austrian Universities, Stadionallee 2, 1020 Vienna, Austria

- E - Magnets and Power Supplies.

P2T-E-99 INFLUENCE OF PARAMETER VARIATIONS ON THE FATIGUE BEHAVIOR OF MAGNET INSULATION SYSTEMS

Prof. Weber Harald. W., Humer Karl (1) Weber Harald W.(1)

(1)= TU-Wien/Atominstitut der Österreichischen Universitäten Stadionallee 2 A-1020 Wien AUSTRIA

The reliable application of glass-fiber reinforced plastics as insulation materials for fusion magnet coils (e.g. the Toroidal Field Coils of ITER) requires the full characterization of their mechanical performance under ITER-relevant conditions. One of the common methods to test the material’s response under dynamic load is the tension-tension fatigue procedure. This test can be used to simulate the pulsed tokamak operation of the ITER coils over a lifetime of more than 20 years. Furthermore, it provids information on the maximum tensile or shear stress in the ITER-relevant region of 104-105 cycles. In order to simulate the operation conditions of ITER as closely as possible, several fatigue parameters can be set in the test programme, e.g., the minimum-to-peak stress ratio R and the frequency n of the sinusoidal load function. Further, the fatigue process can run under load or displacement control. All of these parameters may influence the mechanical response of an insulation system under cyclic load. Therefore, it is highly desirable to investigate the influence of test parameter variations on the measured stress-lifetime diagrams. The investigations were performed at 77 K using an industrial glass-fiber reinforced composite impregnated with epoxy resin. For both the load and the displacement controlled mode, R-values of 0.1-0.5 and frequencies of 5-20 Hz were chosen. The results will be discussed and compared with respect to ITER-relevant operation conditions. *This work has been carried out within the association EURATOM-OEAW.


Corresponding Author:

Prof. Weber Harald. W.

TU-Wien/Atominstitut der Österreichischen Universitäten, Stadionallee 2, A-1020 Wien, AUSTRIA

- E - Magnets and Power Supplies.

P2T-E-101 MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: TESTS ON PROGRAMMABLE LOGICAL CONTROLLERS AND OTHER ELECTRONIC DEVICES

HOURTOULE JOEL, D. van Houtte P. Fejoz P. Hertout

The electric switchboards installed in the ITER Tokamak building can be subjected to a static or slightly variable magnetic field induced by the ITER coils, of a value that can reach 70 mT. This environment is really particular and doesn’t find itself in any industrial facility. There are no experiments on this subject and the components manufacturing standards don’t take into account this aspect of magnetic field compatibility. An experimental test campaign has been launched by EFDA, in collaboration with CEA and Consorzio RFX, in order to find the operational limits of the components employed. In this collaboration, CEA took the responsibility for the tests of the electronic components, of the signals conditioning and command control units. For these tests, a test bench has been developed in CADARACHE, composed by a solenoid and a remote control power supply. The choice of the components has been carried out in collaboration with manufacturers, by choosing in middle range material and having a lifespan of at least five years. Particular tests procedures were applied, strongly inspired by the standards in force for the tests in the presence of alternate magnetic field. The tests showed that all the components are more or less sensitive to this type of environment. The observed effects vary from the simple temporary dysfunction until the total destruction of internal electronic component. As expected, the most sensitive components were those presenting a ferromagnetic part, such as the relays or galvanic transformers. Moreover, it was shown the importance of the direction of field. The results record the limits in each position, but retain only the most unfavourable position limit. For the signal conditioning units, a significant increase in consumption was observed. The limits of such components are at about 30 mT. For the command control systems (PLC and peripheral) the limits were found around 40 mT. The most sensitive components are relays, which show operational limits according to their position in the field, below of 20 mT. These first results, that need to be refined, shall to be taken into account, not only in the design of the electric distribution boards, but also for all the sets of measurements that will be installed in the TOKAMAK building and will be subjected to a significant magnetic field.


Corresponding Author:

HOURTOULE JOEL

DRFC/ STEP BT502 13115 SAINT PAUL LEZ DURANCE

- E - Magnets and Power Supplies.

P2T-E-105 DESIGN, FABRICATION AND INSTALLATION OF CRYOGENIC TARGET SYSTEM FOR 14 MEV NEUTRON IRRADIATION

Nishimura Arata, Hishinuma Yoshimitsu (1) Tanaka Teruya (1) Muroga Takeo (1) Nishijima Shigehiro (2) Shindo Yasuhide (3) Takeuchi Takao (4) Ochiai Kentarou (5) Nishitani Takeo (5) Okuno Kiyoshi (6)

(1) NIFS, Gifu 509-5292 Japan (2) Osaka Univ., Osaka 565-0871 Japan (3) Tohoku Univ., Miyagi 980-8579 Japan (4) NIMS, Tsukuba, Ibaraki 305-0047 Japan (5) JAERI, Tokaimura, Ibaraki 319-1195 Japan (6) JAERI, Nakamachi, Ibaraki 311-0193 Japan

The design of the International Thermonuclear Experimental Reactor (ITER) has been progressed and the neutron streaming is clarified analytically. The hard streaming is expected around NBI ports and it will cause irradiation on superconducting magnets. Since the irradiation spectrum is different depending on the location, the effect of pure 14 MeV neutrons on materials is planed to be investigated to clarify the change of mechanical and electrical properties of the magnet materials. A cryogenic target system has been installed in Fusion Neutronics Source (FNS) at Japan Atomic Energy Research Institute (JAERI) under collaboration between Universities, National Institutes and JAERI and makes it possible to perform the electrical measurement at cryogenic temperature without warming up the samples. Deuterium is accelerated to around 350 keV and collides with tritium absorbed in rotating target plate, resulting in D-T reaction which generates 14 MeV neutrons. The neutron flux depends on the distance (r) from the collision point and decreases as a function of 1/r2. The cryogenic target will be located at about 10 mm far from the D-T reaction point and be kept at 4.5 K by a small refrigeration system whose capacity is 0.5 W at 4.2 K. The compressor and data acquisition system are installed in the other room to reduce the neutron irradiation and operated automatically. Samples of Nb3Sn, NbTi, Nb3Al and pure copper wires will be attached on the target plate and be irradiated up to the fluence of 1016 n/cm2 at cryogenic temperature. According to the reference data, it is expected that Nb3Sn and NbTi will show no change in such neutron fluence. However, the electric resistance of pure copper will be increased and it is important to clarify the change in resistance at cryogenic temperature, for the pure copper is commonly used as a stabilizer for the superconducting strand. At the same time, organic materials will be irradiated in a room temperature space together with glass fiber reinforced plastics (GFRP) and the mechanism of the decomposition and the interlaminar shear strength will be discussed.


Corresponding Author:

Nishimura Arata

Fusion Engineering Research Center, National Institute for Fusion Science, Oroshi 322-6, Toki, Gifu 509-5292 Japan

- E - Magnets and Power Supplies.

P2T-E-106 THE EUROPEAN NB3SN ADVANCED STRAND DEVELOPMENT PROGRAMME

Vostner Alexander, E. Salpietro (1)

(1) EFDA Close Support Unit - Garching, Boltzmannstr. 2, 85748 Garching, Germany

Significant progress in the field of Nb3Sn strand manufacture has been made over the last few years. Strands relevant for fusion with high critical current densities and moderate hysteresis losses have been developed and already produced on industrial scale for the KSTAR project. Based on these achievements EFDA CSU – Garching has launched a Nb3Sn strand development and procurement action inside Europe in order to assess the current status of the Nb3Sn strand production capability. All six addressed companies replied positively to our strand R&D programme which includes the three major Nb3Sn production techniques namely the bronze, internal-tin and powder-in-tube (PIT) route. According to the strand requirements for the ITER TF conductor a critical current density of 800 A/mm2 (at 12 T, 4.2 K and 10 µV/m) and overall strand hysteresis losses below 500 kJ/m3 have been specified as the minimum guaranteed strand performance. The second major objective of this programme is to motivate the strand manufacturers in utilising the technical advances to develop and design advanced state-of-the-art Nb3Sn strands optimised for the ITER conductor. For this purpose, a target critical current density of 1100 A/mm2 has been added to the specification. This paper describes the strategy behind the strand development programme, the actual status of the strand production as well as first preliminary results obtained from the strand suppliers.


Corresponding Author:

Vostner Alexander

EFDA Close Support Unit - Garching, Boltzmannstr. 2, 85748 Garching, Germany

- E - Magnets and Power Supplies.

P2T-E-111 DESIGN AND DEVELOPMENT OF THE POWER SUPPLY SYSTEM FOR HL-2A TOKAMAK

Yao Lieying, Xuan Weimin Li Huajun Chen Yuhong Bu Mingnan Shao Kuei Hu Haotian Mao Xiaohui Wang Shujin Ren Juqian

Southwestern Institute of Physics, P.O. Box 432,Chengdu,Sichuan,610041, P.R.China

The HL-2A is the first divertor tokamak in China. Its construction is based on the main components of ASDEX from IPP and an entirely new power supply system is required to power its magnetic field coils and the plasma heating systems. The most important parameters of the HL-2A are toroidal field of 2.8T, plasma current of 480 kA with a flat top of 5s. Thus, the peak power required is 300MVA and the energy content is about 1200MJ per shot. Three flywheel motor-generators (MG) are used to transfer the power and energy from the HV grid. To get sufficient released energy, two identical existing MG have been modified by replacing original flywheel to a big one. Raising the maximum speed and increasing the speed drop of the total shaft are the other ways adopted to increase the released energy. After modification, the maximum apparent power for each generator can increase to 90MVA from 80MVA and released energy can rise to 500MJ from 100MJ.Two modified MG are used to power the toroidal field coils via a 12 pulse diode rectifier. Another MG with output power of 125MVA is used to power the poloidal field system with transformers and thyristor rectifiers. In order to check the initial design and optimize the feedback control system parameters, all the important parts of the power supply system have been simulated with EMTP code. A digital trigger circuit with the precision of 0.04 degree and a reliable protection system are developed to ensure the performances of the power supply. The feedback control of the plasma current and position were worked successfully both in limiter and divertor operations in 2003. The primary tests show that the design and development of the HL-2A power supply system basically meet the requirements of the operation of the HL-2A.


Corresponding Author:

Yao Lieying

Southwestern Institute of Physics, P.O. Box 432,Chengdu,Sichuan,610041, P.R.China

- E - Magnets and Power Supplies.

P2T-E-112 THE ITER THERMAL SHIELDS FOR THE MAGNET SYSTEM: DESIGN EVOLUTION AND ANALYSIS

BYKOV VICTOR, Yu.Krasikov(2), S.Grigoriev(2), V.Komarov(2), V.Krylov(2), A.Labusov(2), V.Pyrjaev(2), S. Chiocchio(3), V.Smirnov(2), V.Sorin(2), V.Tanchuk(2)

(2)D.V. Efremov Scientific Research Institute, St.Petersburg 196641, Russia (3)ITER IT, Boltzman Str 2, 85748 Garching, Germany

The ITER thermal shield (TS) system is designed as a continuous barrier, that reduces by over two orders of magnitude the heat loads transferred by thermal radiation and conduction from warm components to the components and structures that operate at 4.5K. Active cooling of the TS by 80K gaseous He, and the provision of silver coated surfaces with low emissivity facilitate the removal of the residual 6 kW heat load on the magnet system during normal conditions by the cryoplant of reasonable capacity. By making the TS components independent toroidally continuous structures the number of TS supports and hence conductive heat transfer are minimised. After step-by-step modifications the ITER TS consists of three main subcomponents: (1) the central TS, which comprises the vacuum vessel TS (VVTS) around the hot vacuum vessel, central cryostat and transition TS; (2) the upper cryostat TS suspended from the cryostat lid; and (3) the lower cryostat TS supported on the cryostat floor. The TS system also includes the support TS side panels, that block heat loads to and from the magnet gravity supports (MGS) and thermal anchors in the MGS. The efficiency of the TS system depends strongly on the interface between its components, therefore minimisation of the number of the TS components and reduction of heat loads through interfaces is the main approach of the design evolution. The requirements for access to magnet components for repair cannot be excluded for the ITER machine. Locating the cryostat TS just outside the PF coils provides the required space for in-cryostat repair activity outside delicate TS surfaces, while the incorporation of removable TS panels and modification of the outboard Central TS support make access to the TS/Magnet interspace for assembly and disassembly relatively easy. The modern design of the VVTS with extruded, profiled cooling pipes hidden in between double panels improves surface smoothness and thermal efficiency of the structure. Complex tube tracing avoids twisting of the pipes. This paper presents the rationale for the TS design evolution since 2002. Details of the recent modifications that affect the TS cooling panels, the Central TS ports and support system, interface labyrinths and TS structural joints as well as the modern results of thermal-hydraulic, thermal, seismic, static and dynamic structural analyses, that involve submodeling and substructuring finite element analysis techniques, are presented.


Corresponding Author:

BYKOV VICTOR

ITER IT, 801 Mukouyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan

- E - Magnets and Power Supplies.

P2T-E-121 QUALITY ASSURANCE PROCEDURES IN THE EAST MAGNETS MANUFACTURING PROCESS

Chen Siyue, Gao Daming, Yu Jie, Wu Jiefeng, Zhang Pin, Tao Yuming

Institute of Plasma Physics, Chines Academy of Sciences. Hefei, Anhui, China

EAST is a full super conducting Tokamak being constructed in Hefei, China. At the beginning of next year it will have been assembled. Three toroidal field (TF) magnets have been made so far and the first one has passed all the examination. The operating current of TF magnet system is 14.3 KA and its toroidal field is 3.5 T. All the coils are winded by Cable-in-conduit conductors (CICC). The mechanical property, position and dimension precision, the electric, cooling and vacuum performance of the magnets are guaranyeed by quality assurance procedures in fabricating process. ISO9001 quality assurance model is applied in the design and fabricating process. The priority of quality control is design and manufacturing personnel training. Suitable machining and testing tool and clean environment are necessary. The magnets are manufactured on four production lines, namely CICC jacketing line, coil winding line, vacuum pressure impregnation line and mechanical machining line. Every production line has its detailed quality plan. It describes the ways and the standards of acceptance inspection and testing, the detailed techniques, parameters and the testing standard of every working procedure, the track recording forms, the testing result forms, the identifying ways. Because these magnets are not standard products, the technology and testing standard are based on a great deal of experiments. To ensure high reliability, sometimes many testing ways are applied in one working procedure.


Corresponding Author:

Chen Siyue

Institute of Plasma Physics, Chinese Academy of Sciences. Hefei, Anhui, China

- E - Magnets and Power Supplies.

P2T-E-126 THYRISTOR CROWBAR SYSTEM FOR THE HIGH CURRENT POWER SUPPLIES OF ASDEX UPGRADE

Claus-Peter Käsemann (1), Lou van Lieshout (2) Michel Huart (1) Christof Sihler (1)

(1) Max-Planck-Institut für Plasmaphysik (IPP), EURATOM Association, Boltzmannstrasse 2, D-85748 Garching, Germany (2) Imtech Vonk BV, Modem 30, NL 7741 MJ Coevorden, The Netherlands

The ohmic heating system and the poloidal field coils of ASDEX Upgrade (AUG) are supplied by 15 thyristor converter units with an installed apparent power of 600 MVA. A Thyristor Crowbar System (TCS) consisting of 15 units (TCU) was designed, installed and commissioned. These will be used for protecting the thyristor converters against DC overvoltage arising from abnormal operations and resulting damages caused by the large energy stored in the AUG magnet coils. The TCS has to fulfil three main objectives: Reliability - intervention in case of overvoltage but no tripping due to false alarms; Modularity - independent operation of all units; Flexibility - selection of triggering voltage taking account of the different DC system voltages. Each TCU is connected to the DC output terminals of one of the thyristor converters. There are three types of TCU, characterised by their DC rated voltage, namely 2400 V, 1500 V and 500 V. The DC rated current is 45 kA. The TCU is triggered by a DC overvoltage of either polarity and suitable to carry DC current of both polarities. In case of overvoltage the trigger circuit fires a thyristor that transfers the current from the converter resp. load coil to a resistor where the energy is dumped. Each dump resistor is composed of series connected resistor banks that are each characterised by a rated pulsed energy of 5 MJ and a nominal resistance of 25 mOhm. To increase reliability each TCU comprises two modules parallel connected that include their own overvoltage detection, trigger circuit, dump resistor and a ´cross-firing´ between the two modules. The trigger-level is chosen half way between the DC rated voltage and the thyristor blocking voltage. If the converter configuration is changed the trigger voltage can easily be adapted by changing the overvoltage detection board. Each TCU includes its own instrumentation and interlock to ensure all necessary interfaces with the AUG control system and the thyristor converters interlock system. This paper describes the design and testing of the Thyristor Crowbar System representing the DC converter overvoltage protection system. It will present the layout, analyse the results of measurements obtained during commissioning, compare them to the calculated (design) values and report on the first experience on the AUG coils improving the safety of the equipment.


Corresponding Author:

Claus-Peter Käsemann (1)

Max-Planck-Institut für Plasmaphysik (IPP), EURATOM Association, Boltzmannstrasse 2, D-85748 Garching, Germany

- E - Magnets and Power Supplies.

P2T-E-186 OPTIMIZATION OF THE POWER SUPPLY FOR A HELIAS REACTOR SUPERCONDUCTING COIL SYSTEM

Harmeyer, Ewald (1), Wieczorek, Andreas (2) Wobig, Horst (1)

(1) Max-Planck-Institut fur Plasmaphysik, EURATOM Association, D-17491 Greifswald, Germany. (2) FH-University of Applied Science, D-93049 Regensburg, Germany.

For magnetic confinement of a hot fusion plasma Stellarator magnetic field configurations of the Helias type have been developed. A Helias Reactor coil system with 4 field periods and a major radius of 18m is applied. This coil system comprises 40 modular coils in total, 10 coils per field period. The electrical circuit consists of 5 coil groups, each of them with 8 equally-shaped coils connected electrically in series. These 5 coil systems will be powered individually by 5 power supplies of the thyristor type. The power supply units must generate currents up to 40kA to achieve a magnetic field of 5T on axis resulting in total stored magnetic energy of about 100GJ in the coil system. All systems will be powered direct from the medium voltage utility interface for auxiliary systems. The grid is loaded by the operation of the line commutated converters with reactive power and harmonics. Because of high power levels associated with this application, it is important to reduce the harmonic currents generated on the ac side of the converter. This is accomplished by means of a 12-pulse converter operation. If several converters are connected to the same supply mains, they will affect one another through the commutation notches. Operation directly in parallel is not possible, they must therefore be decoupled by transformer inductances. A power supply system for feeding the superconducting coils of the Helias reactor has been investigated. This multiconverter supply system has been optimized, in view of low losses in the components and only little impact to the power grid. The design of the optimized multiconverter supply system was studied by means of computer simulations, using the SIMPLORER code. The influence of the passive structures on operation of the power supply system was taken into account. The influence of induced eddy currents in the coil structure during transient processes are transformed into electric network analyses by means of the Finite Element Network (NET) method. This approximation allows the investigation of the entire coil system including power supplies and passive structures.


Corresponding Author:

Harmeyer, Ewald (1)

Max-Planck-Institut fur Plasmaphysik, EURATOM-Association, D-17491 Greifswald, Germany

- E - Magnets and Power Supplies.

P2T-E-198 QUENCH CURRENT MEASUREMENT AND PERFORMANCE EVALUATION OF THE EAST TOROIDAL FIELD COILS

Weng Peide, Z.M.Chen, Y.Wu, Y.N.Pan, W.G.Chen, Z.R.Ouyang, H.Y.Bai, X.N.Liu, P.Fu, L.W.Xue, Y.F.Tan

Institute of Plasma Physics Chinese Academy of Sciences, p.o.box 1126, Hefei,Anhui 230031,China

Selected also for oral presentation O2A-E-198

EAST device (original name is HT-7U) is a superconducting tokamak constructing in Institute of Plasma Physics Chinese Academy of Sciences. The TF magnet system of the device is consisting of 16 D shape TF coils made of NbTi Cable In Conduit Conductor. The nominal operating current of the coil is 14.3 kA. It is planed to test all of the 16 TF coils this year. The test program included a number of items such as cryogenic-hydraulic property, electro-magnetic property and quench current measurement at different temperature. Up to now, 11 TF coils have been tested in our test facility. Each coil was cooled down up to 4.5 K and charged to 16 kA, the magnet field on the coil, internal joint resistance of the coil, mass flow rate and pressure drop of each cooling channel, were measured at same time. After that, the quench current of TF coil was tested, due to limitation of power supply, we have to use higher temperature, the Helium temperature were increased to more than 7.5 K and the coil excitation were performed again till the coil quench. The measurement results and coil performance evaluation are presented in this paper.


Corresponding Author:

Weng Peide

Institute of Plasma Physics Chinese Academy of Sciences, p.o.box 1126, Hefei,Anhui,230031,China

- E - Magnets and Power Supplies.

P2T-E-209 THERMAL AND STRUCTURAL ANALYSIS OF THE W7-X MAGNET HEAT RADIATION SHIELD

Nagel, Michael, Seong Yeub Shim Felix Schauer

The magnet system of the fusion experiment WENDELSTEIN 7-X comprises 70 superconducting coils. In order to reduce the heat load on the coils, in addition to high vacuum an efficient thermal insulation is required which basically covers the outside of the plasma vessel, the inside of the outer vessel, and the outside of the port walls. The insulation consists of multi-layer insulation (MLI) and a thermal shield which is cooled by gaseous helium. Detail engineering of the plasma vessel insulation has been finished, and its production has started. The paper presents the mechanical design as well as the cooling concept of the shields, and shows the resulting temperature distributions for different design options. Calculations are based on finite element models of the outer and plasma vessel as well as the port shields. The shields are all subdivided into panels which are described with thermal SHELL elements. FLUID PIPE elements are used to model helium in the cooling tubes. Heat load on the panels in normal operation is assumed to be 6 W/m2, with a uniform distribution. The helium is warmed up from about 40 K at the inlet to around 70 K at the outlet. Structural analysis of the thermal shield is carried out in order to define its mechanical strength as well as the required number and positions of the supports. Main loads are electromagnetic forces resulting from a rapid shut down of the magnet system, the weight of the shield, and treading on during cryostat assembly works. The eddy current forces induced in the shields during a rapid shut down are calculated in a two step procedure. First the magnetic field and the corresponding vector potential data are calculated using the well known EFFI code based on the Biot-Savart law. In the second step, the vector potential is used as input parameter for calculating the induced currents on the shields with the finite element code ANSYS. For the transient calculation, an exponential decay of the magnetic field is assumed. That way only a finite element model of the shields is required, reducing the model size and required calculation time. Results for different options are discussed. It is shown that the chosen shield design fulfils the requirements.


Corresponding Author:

Nagel, Michael

Max- Planck- Institut fuer Plasmaphysik, Wendelsteinstrasse 1, D- 17491 Greifswald, Germany

- E - Magnets and Power Supplies.

P2T-E-212 FILAMENT POWER SUPPLY (AC TO AC CONVERTER) FOR LONG PULSE NEUTRAL BEAM INJECTOR OF SST-1

P.J. Patel, O.Raja, N.P.Singh, V.Sharma, U.K.Baruah ,S.K.Mattoo and NBI Team Institute for Plasma Research, Gandhinagar, India – 382428

Filament Ion Sources used for Neutral Beam Injectors use AC heating. For long pulse operation, AC filament heating power is advantageous. To minimize the ripple on plasma density, number of phases of the heater supply is usually made large. This paper presents the design and performance of AC to AC converters for filament power supply of the ion source for the long pulse Neutral Beam Injector of SST-1 (Steady-state Superconducting Tokamak-1). The input is from the utility mains, the input stage unity power factor controller circuit maintains the total harmonic distortion of line within 5% and power factor at unity. A PWM inverter along with output filters generates the output. Three phase AC output is controllable from 40-400V(rms), 400Hz, sinusoidal, at 7.0 kVA (max.). The secondary from a 3-phase step down transformer (ratio 22:1) placed at the output of the converter is connected to one filament. The stability at the output is 0.1 % with variations in input line or fluctuations in the load. Overall efficiency is approximately 90%. Eight power supplies, all capable of being synchronized with an external trigger pulse are used to generate a 24-phase filament heating system for the ion source. The PWM generation carrier waveform is generated by a digital scheme, with a start trigger for each cycle. For synchronized operation of all converters, the output can be locked to a phase reference input TTL pulse train, whereby a 24-phase (or any user defined) system is realized within 1.0 degree (electrical) accuracy. The control is matched to meet the filament temperature stability with compensation for actual current in the filament. The design uses novel power topology, a combination of a high frequency inverter and a front-end power factor controller for each phase. IGBTs are used for both the input unity power factor controller and inverter sections.


Corresponding Author:

P.J. Patel

Institute for Plasma Research,Bhat, Dist. Gandhinagar-382428

- E - Magnets and Power Supplies.

P2T-E-219 TRANSIENT ELECTRICAL BEHAVIOUR OF THE ITER TF COILS DURING FAST DISCHARGE AND TWO FAULT CASES

Stefan Fink (1), Tullio Bonicelli (3) Walter H. Fietz (1) Amir M. Miri (2) Xiangming Quan (2) Albert Ulbricht (1)

(1) Forschungszentrum Karlsruhe, ITP, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany (2) Universität Karlsruhe, IEH, Kaiserstr. 12, 76128 Karlsruhe, Germany (3) EFDA-CSU Garching, Boltzmannstr. 2, 85748 Garching, Germany

Insulation faults are regarded from the ITER International Team as the most probable cause of magnet failure. Considering the difficulties involved in the replacement of a TF coil in the ITER magnet system and the different problems occurring during high voltage tests of the ITER model coils further improvements in several aspects of high voltage technology for the realisation of the ITER magnets are indispensable. One of these aspects is the consideration of the transient electrical behaviour because it is well known that fast changes of voltages (e. g. lightning and switching impulses) may cause a non linear voltage distribution on the coil turns and possibly excite resonances within a large coil. Such high voltage stress can cause local overloading and irreversible destruction of the insulation system. This paper will present the calculation of the terminal voltages within the ITER TF coil system and the voltage stress of the three insulation types (ground, radial plate and conductor insulation) within an individual ITER TF coil for the fast discharge and two fault cases. An electrical network model for the ITER TF coil was developed and simulated with the code PSpice. The internal inductances and capacitances as well as the capacitances to ground for the establishment of this network model were determined. Skin and proximity effect as well as the damping caused by eddy currents in the stainless steel radial plates, in which the conductor is embedded, were calculated by the FEM code Maxwell. For the complete TF circuit, composed of 18 TF coils and 9 fast discharge units, an additional network model was set up and implemented with the code PSpice. Due to the large size of the individual ITER TF magnets the resonance frequency is lower than for the TF model coil. It was also determined that the three types of insulation within a single TF coil are stressed with a nonlinear voltage distribution under a fast discharge condition. The non linear voltage distribution is enhanced in case of fast excitations applied in consequence of ground faults. Therefore insulation coordination and test voltages have to be defined in consideration of the stresses caused by fast discharges and applicable and realistic fault cases to ensure a reliable operation during the foreseen ITER lifetime. Hence some proposals for the high voltage test procedures will be discussed based on the calculated voltage stress and the experiences gained during the ITER TF Model Coil test.


Corresponding Author:

Stefan Fink (1)

Forschungszentrum Karlsruhe - ITP, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen

- E - Magnets and Power Supplies.

P2T-E-223 STUDIES ON THE BEHAVIOR OF MULTISECONDARY TRANSFORMERS USED FOR REGULATED HV POWER SUPPLIES

N P Singh, U K Baruah, P J Patel, S K Mattoo & NBI Team

Multisecondary winding transformers have been in use for the input stage of modular High voltage power supplies for a long time now. These power supplies have been widely used for Neutral Beam Injectors, RF& Microwave devices. This design is also used in pulsed electric field applications. For tokamak experiments of steady state operation, suitability of these transformers needs to be analysed. Secondary windings of multisecondary transformer are displaced axially on the core in the form of stacks and each stack comprises of number of coils permitted by winding insulation. Combination of non-uniform and uniform insulation provides required sequential voltage gradient among multisecondaries and to the ground. The secondary windings are designed to achieve identical transformation ratio and impedances. Against each stack of the secondary, the primary winding is designed for ampere-turns to balance short-circuit forces during a fault. In a developed and tested design of a transformer, conventional bushing for large number of terminals is replaced by a compact resin cast bushing plate with embedded terminations. Accommodating large numbers of secondaries (~20 or more) in a compact transformer design leads to formations of various capacitances among secondaries (few tens of nanofarads) and to the ground (few hundred picofarads). As the output voltage is built up by switching of a rectifier and IGBT at the secondary, the potential of the winding is lifted to the DC potential. As switching of different stages progresses, the winding potential fluctuates at submultiples of the switching frequency of the IGBT. The winding stray capacitances charges and discharges, consequently, a high frequency leakage current is induced. Effect of these stray capacitances has been observed on the output performance of the power supply, it is likely to determine suitability of the transformers for long term operation. This paper presents a systematic study of the interactions of the stray capacitances on the behavior of the power supply. Results from simulation and experiments on a pair of 300kVA, 415/330V (20 secondary), 30kVDC isolation transformers used for generating 14kVDC, 35Amp output are presented. Possible effects on the performance of the transformer and the design considerations are discussed. Additional dielectric losses, voltage swings at the switching frequency of the semiconductor devices and the observed effects of these factors are presented.


Corresponding Author:

N P Singh

NBI Group, Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar, Gujarat (INDIA) 382428

- E - Magnets and Power Supplies.

P2T-E-236 HIGH POWER IGBT BRIGE APPLICATION FOR THE HARMONIC SUPPRESSION IN THE POWER SUPPLY SYSTEM OF THE SPANISH STELLARATOR TJ-II.

Kirpitchev Igor, P. Mendez1, M. Blaumoser1, M.Visiers2, A. Agudo2, J. Iglesias3

1. Asociación EURATOM-CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid, Spain, 2.ENERTRON, S.A. C/Amsterdam Pol. Ind. Torres de Alameda, 28813 Madrid, Spain, 3.CEDEX Alfonso XII, 3 y 5, 28014 Madrid, Spain

The magnetic system of TJ-II is equipped with one toroidal and six poloidal mutually coupled coils. The central poloidal coils system consists of a solenoid and two helical coils which spiral around this solenoid. These three coils are placed inside the toroidal system very close to the vacuum vessel and consequently to the plasma (about 10cm), and that's why the power supply system have to guarantee very low current perturbation in these coils in order to avoid negative influence on the plasma confinement. The coils systems are supplied separately by 12 pulse controlled thyristor converters with the maximum DC current 12 kA for the poloidal system and 32 kA for the toroidal one. The thyristor rectifiers are fed by a fly-wheel generator. Its output voltage is 15kV and its output frequency varies from 100Hz at the beginning of the experimental pulse to 80Hz at his end. The thyristor rectifiers can vary the current in the coils from zero to their maximum values and theirs phase control regulators produce a high frequency harmonics in the DC current. The oscillations of the various regulation systems operating simultaneously and small asymmetries of generator and transformers produce low frequency sub-harmonics also. The current ripple requirements have been specified to be kept to a very low level, 1% of the actual coil current in all coil systems but not more then +25 A in the poloidal coils and not more than +50 A in the toroidal coil. Six years of TJ-II operation demonstrate that current ripple requirements are being met, but nevertheless a further reduction of about one order of magnitude namely to +2A is necessary. The investigations of different methods show that the active filter is the most appropriate way for the current ripple reduction. Different computer simulations have been carried out in order to confirm the feasibility of the technical solution and to define the main parameters of the filter. The results of the calculations demonstrate that the active filter connected in parallel to the load and based on IGBT H-bridge can reduce the current ripple of the coil to the specified +2A limit. Spanish company ENERTRON has manufactured the active filter and the tests at factory have confirmed the correct operation of the equipment. The paper describes in detail the design of the filter, the computer simulations, the results of a test circuit at factory and the results of the commissioning tests on site.


Corresponding Author:

Kirpitchev Igor

Asociación EURATOM-CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid, Spain

- E - Magnets and Power Supplies.

P2T-E-259 MANUFACTURE AND TEST OF THE NON-PLANAR COILS FOR WENDELSTEIN 7-X

Rummel, Thomas, Konrad Risse Hartmut Ehmler

Max-Planck-Institut für Plasmaphysik, Euratom Association, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald, Germany

The standard magnetic configuration of WENDELSTEIN 7-X (W7-X) is formed by 50 non-planar superconducting coils. 20 additional planar superconducting coils allow to modify the magnetic configuration. Due to the symmetric arrangement of 5 equal modules, each being composed of two mirror-symmetric half-modules, 5 differently shaped types of non-planar coils are sufficient. The nominal current is 17.6 kA for all non planar coils, which can be varied between 14.5 kA for 2.5 Tesla operation up to 18.3 kA for the 3 Tesla operation in the low shear scenario. The winding of the coils is made of 108 turns of a forced flow cable-in-conduit conductor using a NbTi superconductor. It consists of a rope with 243 strands enclosed in an aluminium jacket with a void fraction of 37 %. The outer dimensions of the jacket are 16 mm x 16 mm. One strand has a diameter of 0.57 mm and is made of 144 NbTi filaments stabilized by copper with a copper to non-copper ratio of 2.7. The specified critical current of the superconductor is 32 kA at 4.2 K and 6 Tesla. To withstand the electromagnetic forces of up to 400 t each winding pack is stiffened by a massive steel casing, which leads to a total weight of a non-planar coil of about 5.5 tons. Typical dimensions of a non-planar coil are about 3.5 m x 2.5 m x 1.5 m. The contribution gives a report about the status of the production comprising the production of the winding packs and the casings as well as the assembly of the coil, including instrumentation. Special attention had to be given to the quench detection wiring. The design of the quench detection cables, which consist of 6 single wires each, was changed in order to ensure a better electrical strength also under vacuum conditions. Several coils are already finished and tested at room temperature at the manufacturer’s site. The test procedure and typical test results will be presented. After production all coils will be tested under cryogenic conditions, too. Main tests are a nominal current test, a quench test by increasing the temperature, a high voltage test, a helium leak test and measurement of the stresses in the casing, the shrinkage during cool down and the deformations due to electromagnetic forces. The test procedure and the results of the first tests of the coils will be presented and compared with the expectations.


Corresponding Author:

Rummel, Thomas

Max-Planck-Institut für Plasmaphysik, Euratom Association, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald, Germany

- E - Magnets and Power Supplies.

P2T-E-266 V-I CHARACTERISTICS WITH BUMPS IN THE MEDIUM SIZE NBTI CICC CABLES.

Bruzzone Pierluigi, Boris Stepanov (1) Elena Zapretilina (2)

(1) EPFL-CRPP, CH-5232 Villigen-PSI, Switzerland (2) NIIEFA (D.V.Efremov Institute), 186641, St. Petersburg, Russian Federation

In the scope of conductor R&D program for the poloidal field coils of the ITER fusion project, a number of short samples of sub-size NbTi cable-in-conduit conductors with 336 strands (CICC) and full size (1440 strands) have been tested in the SULTAN facility (CRPP Villigen, Switzerland). The cabling and jacketing work is done at VNIIKP (Moscow). The dc test (critical current and current sharing temperature) was carried out with supercritical helium at 10 bar, over a broad range of operating temperature (4.5 – 7 K) and background field (4 – 7 T). The paper reports about two different cases of “bumpy behaviour observed in the voltage-current characteristics. During the tests at increased operating temperature, an abnormal behaviour of the voltage – current characteristic has been observed. In some cases, instead of smooth, monotonous transition (power law V-I characteristic) a wave-like voltage development and recovery (“bumps”) have been seen. The most likely reason for this abnormality seems to be minor temperature variations, in the range of few hundredths of degree, caused by minor pressure waves in the supercritical helium circuit. The paper discusses conditions at which the effect could be observed, and presents some analysis which reproduces the observed behaviour and support the hypotheses about ‘thermal’ origin of the voltage ‘bumps’. Another, different “bumpy” behaviour in the current-voltage characteristic is reported, where the reason is discussed to be linked with a current re-distribution phenomenon, also supported by the response of Hall sensors monitoring the self-field of the sample.


Corresponding Author:

Bruzzone Pierluigi

EPFL-CRPP, CH-5232 Villigen-PSI, Switzerland

- E - Magnets and Power Supplies.

P2T-E-284 HIGH TEMPERATURE SUPERCONDUCTORS FOR THE ITER MAGNET SYSTEM AND BEYOND

Fietz, Walter H.(1), Stefan Fink(1) Reinhard Heller(1) Peter Komarek(1) Vipulkumar L. Tanna(1) Gernot Zahn(1) Gabriel Pasztor(2) Rainer Wesche(2) Ettore Salpietro(3) Alexander Vostner(3)

(1) Forschungszentrum Karlsruhe, Institut für Technische Physik, Karlsruhe, Germany. (2)Centre de Recherches en Physique des Plasmas, Villigen, Switzerland. (3) European Fusion Development Agreement, Close Support Unit, GARCHING, Germany.

Selected also for Oral Presentation O2A-E-284

Operation currents up to 68 kA have to be transferred from room temperature (RT) to 4.5 K for the superconducting magnet system of ITER. This current transfer is made using specially designed current leads (CL). With the conventional design of such CL, ohmic losses cause high heat loads to the refrigeration system which is critical in the low temperature range where the efficiency of the refrigerator is reduced according to the Carnot rule. For ITER a cooling power of 64 kW at 4.4 K is foreseen taking more than 20 MW of electric power. This large power consumption can be reduced drastically by the use of High Temperature Superconductor (HTS) current leads for ITER and future fusion machines because these HTS CL have no ohmic losses in the range of 4.5 K to 70 K. In the frame of the European Fusion Technology Programme, the Forschungszentrum Karlsruhe and the CRPP Villigen have designed and built a 70 kA current lead using HTS material. This HTS current lead was installed and tested in the TOSKA facility of the Forschungszentrum Karlsruhe. The experiment covered the electrical and thermal behaviour under steady state conditions and in case of a quench. To characterize the performance of the current lead, the temperature profile, the contact resistances, the required cooling power, and the critical current were evaluated. To check extreme conditions a complete loss of He-flow was studied, too. Results of the experiments carried out in TOSKA facility are presented. In addition an outlook of future prospects of HTS material applications in a fusion machine will be given. An obvious possibility is to introduce HTS in the RT bus bar system to reduce losses allowing a much lower effort for thermal shielding and possibly alternative thermal insulation concepts. A preliminary layout of a HTS bus bar system is shown. However, the real challenge is to use HTS materials for the whole magnet system. Even when such a magnet system is operated at 20 K, those fusion machines would be much more efficient due to the reduction of electric power consumption for cryogenics by a factor of 5-10. An improved version would be a machine with a magnet system operating at 65 K to 77 K, because in this case liquid nitrogen could be used as coolant. An overview about status, promises and challenges of HTS conductors on the way to an HTS fusion magnet system beyond ITER is given.


Corresponding Author:

Fietz, Walter H.(1)

Forschungszentrum Karlsruhe, Institut für Technische Physik, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen

- E - Magnets and Power Supplies.

P2T-E-299 ANALYSIS OF THE RESISTIVE TRANSITION IN NB-TI CABLE-IN-CONDUIT CONDUCTORS VIA AN EXTENDED 1-D MODEL

Zamboni Walter, Pierluigi Bruzzone (1), Luca Bottura (2), Claudio Marinucci (1)

(1) CRPP-Technologie de la Fusion, CH-5232 Villigen-PSI, Switzerland (2) CERN, AT-MTM, CH-1211 Geneva 23, Switzerland

In the frame of the research on applied superconductivity, the resistive transition of superconducting cables is one of the most discussed issues in the recent years. Experimental tests performed on NbTi middle-size cable-in-conduit conductors (CICCs) at CRPP show that the resistive transition parameters, i.e. the critical current and the n-index, are strongly affected by the self (magnetic) field effect and current redistribution. The strong self-field gradients, due to large operating current, induce a current redistribution among strands. The whole phenomenon is strongly dependent on the values of transverse resistivity of the cable, which is able to either avoid or promote current redistribution. In order to investigate the influence of the self field, we simulate the current sharing process in a double stage model of the cable, which takes into account longitudinal and transverse resistive effects. To this purpose, we retain a power-law model for the basic superconducting element. In the evaluation of the electromagnetic behavior of the cable, we adopt an extended 1-D approach. It consists of a Multiconductor Transmission Line (MTL) Model with constant inductive and resistive transverse coupling between elements of the MTL. The magnetic model is coupled to an assessed thermohydraulic one and numerically solved by CryoSoft Thea®. We tested our model against the critical current experiments performed on different conductors, which mainly differ in the interstrand resistance. The results show that the model, although not complete and “self-consistent” as a 3-D one is, can be a reliable and relatively fast tool in the investigations on the self-field and current redistribution effects on CICCs, once the strands and the basic components have been characterized.


Corresponding Author:

Zamboni Walter

Association EURATOM-ENEA-CREATE, DIEL, Università degli Studi di Napoli "Federico II", via Claudio 21, 80125 Napoli Italy

- E - Magnets and Power Supplies.

P2T-E-304 EFFECTIVE BENDING STRAIN ESTIMATED FROM IC TEST RESULT OF D SHAPED NB3AL CICC COIL FABRICATED WITH A REACT AND WIND PROCESS FOR THE NATIONAL CENTRALIZED TOKAMAK

ANDO Toshinari, KIZU Kaname(1) MIURA Yuushi(1) TSUCHIYA Katsuhiko(1) MATSUKAWA Makoto(1) TAMAI Hiroshi(1) ISHIDA Shinichi(1) KOIZUMI Norikiyo(1) OKUNO Kiyoshi(1)

(1)Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukouyama, Naka-machi, Naka-gun, Ibaraki-ken 311-0193, Japan

In Japan the National Centralized Tokamak is being planned, based on a modification of JT-60 with superconducting coil. The design of the TF coil is characterized by the maximum magnetic field of 7.4 T at a nominal operating current of 19.4 kA and by the use of a react and wind process with the maximum bending strain of 0.4 % on the Nb3Al cable using a Nb3Al CIC conductor. The bending strain is defined from theNb3Al cable diameter divided by the winding diameter. Nb3Al is insensitive to strain on Ic in comparison with Nb3Sn. The conductor consists of 216 Nb3Al strands and 108 copper wires inserted into the circular hole of a rectangular stainless steel conduit. In order to demonstrate the applicability of Nb3Al conductor with the react and wind process to the TF coil, a two turns-D shaped Nb3Al coil whose height is about 2 m, with the full size CIC conductor has been fabricated and tested by installing its corner part wound with a bending strain of 0.4 %, into a split coil as background field. In this test the strain corresponding to the degradation of Ic on the Nb3Al conductor due to its bending, so called the effective bending strain to be converted into the axial strain, was investigated and estimated. From Ic results in this test the total strain on the conductor was estimated as – 0.57 %. The estimation was carried out taking account of the magnetic field and strain distribution within the Nb3Al conductor. On the other hand, the axial strain due to the thermal stress from the stainless steel conduit on the Nb3Al filament was found to be -0.57 % from the other experiment. This means that the degradation of Ic due to the bending is neglect. Therefore, the effective bending strain on the Nb3Al conductor was zero. It is considered that the strands in the cable slipped each other toward the reduction of bending strain during the bending. This result is very useful for the fusion coil fabrication with the react and winding process. In this paper, the effect of bending strain on Ic in Nb3Al cable-in-conduit conductor is discussed.


Corresponding Author:

ANDO Toshinari

Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukouyama, Naka-machi, Naka-gun, Ibaraki-ken 311-0193, Japan

- E - Magnets and Power Supplies.

P2T-E-306 ELIMINATION OF VARIABLE HARMONICS ON MOTOR GENERATOR CIRCUIT FOR EXPERIMENTAL FUSION FACILITY

Yamada Shuichi, Nakanishi Yosuke (1) Kojima Hiroshi (1) Hiue Hisaaki (2) Uede Toshio (2) Mito Toshiyuki

(1) Fuji Electric Advanced Technology Co. Ltd., Fuji-machi, Hino, Tokyo 191-8502, Japan. (2) Fuji Electric Systems Co. Ltd., 1-1, Tanabeshinden, Kawasaki, 210-9530, Japan

In an experimental fusion device, a large electric power is needed for producing high temperature plasma in the high density regime. Since the motor generator with a flywheel (FW-MG) can generate large electric power without giving the turbulence to the commercial power grid, it is used for the back power source of the heating devices such as the NBI, ECH and/or ICRH. The frequency of FW-MG changes almost factor of two between starting phase and running down phase during a pulse. When the power supply of the heating device is composed of full wave rectifiers using the thyristors, the harmonic currents of the 5th, 7th and other higher components appear on the output circuit of the FW-MG. The frequencies of these harmonic currents also change the same in proportion to the fundamental frequency of the FW-MG. These variable harmonic currents may threaten to damage the windings of the generator and/or the transformers of heating devices caused by the abnormal temperature increase. To avoid these deteriorations, a special active filter, which can eliminate the variable harmonic currents in continuity, was investigated. It has the following major functions; 1) the detection of the variable frequency of the power line, 2) the extraction of current component of fundamental frequency, 3) the operation of current component of higher harmonics, and 4) the compensation of harmonic current by generating the counter-flow current. A special algorithm of the band-pass filter was developed for the extraction of current component of fundamental frequency. Dynamic simulations for the active filter, FW-MG and power supplies of heating devices for the experimental fusion device of LHD has been conducted by using the analysis tool of the PSCAD/EMTDC. We confirmed that the harmonic currents with the amplitudes of 20% were suppressed to less than 2 % through the operational frequency range from 95 Hz to 55 Hz by using this active filter.


Corresponding Author:

Yamada Shuichi

National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, Japan

- E - Magnets and Power Supplies.

P2T-E-311 FATIGUE ASSESSMENT OF THE ITER TF COIL CASE BASED ON JJ1 FATIGUE TESTS

Hamada Kazuya, Nakajima Hideo, Katsutoshi Takano and Okuno Kiyoshi

801-1, Muko-yama, Naka-machi, Naka-gun, Ibaraki, Japan

In the International Thermonuclear Experimental Reactor (ITER), a structure material for Toroidal Field (TF) coil case at the inboard leg requires a high strength (0.2% yield strength>1000 MPa) and toughness (fracture toughness KIC(J) >200 MPam0.5) at 4.5K. Japan Atomic Energy Research Institute (JAERI) has developed JJ1 (0.05C-12Cr-12Ni-10Mn-5Mo-0.2N) for this application. Since 60,000 cycles of electromagnetic load will be loaded on the ITER TF Coil case during coil life, a fatigue characteristic of TF coil is important subject for structural design. There are two approaches for evaluation of fatigue life assessment. One is a fatigue crack growth rate (FCGR) evaluation, which has been performed based on the measurement of the FCGR. The other is a fatigue life assessment based on Stress-fatigue life(S-N curve), which we think more appropriate approach and had not been performed well because measurement data were not enough for establishment of S-N curve at 4.5K. JAERI has measured the fatigue life of the base metal and welded joint of JJ1 at 4.5K, based on JIS Z2283 ‘Method of low cycle fatigue testing for metallic material in liquid helium.’ The fatigue test has been performed in a fully reversed axial - strain controlled method. The strain range and number of cycle are -0.6% to +0.6% and 10,000 to 2,000,000 cycles, respectively. Total 17 samples have been tested. As a result, failure cycle at welded joint is evaluated to be 60,000 at stress amplitude of around 740 MPa from the S-N curve established. The S-N curve of JJ1, together with the safety factor of 20 for failure cycle and 2 for stress amplitude indicates that the stress amplitude of TF coil case should be kept less than 370 MPa to achieve required operation cycle of 60,000 in case of JJ1. In TF coil case inboard leg, the severest cyclic stress occurs at poloidal shear key region located at the top and bottom corners. The recent stress analysis indicates that maximum principal stress is 600 MPa and cyclic stress is 368 MPa. This stress condition corresponds to the equivalent alternating stress of 250 MPa, using well-known Goodman’s diagram method, which is below the allowable value. Therefore, it is concluded that the JJ1 satisfied the specified operation cycle of 60,000.


Corresponding Author:

Hamada Kazuya

801-1, Muko-yama, Naka-machi, Naka-gun, Ibaraki, Japan

- E - Magnets and Power Supplies.

P2T-E-326 EFFECT OF ELECTRICAL CHARACTERISTICS OF SIC POWER DEVICE ON OPERATIONAL EFFICIENCY OF AC/DC CONVERTER

Tatsuya Matsukawa, Hirotaka Chikaraishi (1) Yoshihisa Sato (2) Ryuichi Shimada (3)

(1)National Institute for Fusion Science, Oroshi-cho, Toki, Gifu, JAPAN (2)Daido Institute of Technology, Takiharu-cho, Minami-ku, Nagoya, JAPAN (3)Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Ookayama, Meguro, Tokyo, JAPAN

The large capacity AC/DC converter, which is to output high DC current for energizing the magnetic field coil of nuclear fusion experimental facility or SMES system, is designed to minimize its operational loss for high efficiency operation. SiC power device is a new power electronics device based on crystallized SiC material, and is one of the most promising switching elements to be applied to large capacity AC/DC converter. SiC power device is expected to have more excellent electrical characteristics than those of conventional Si power device. Typical advantages of SiC power device in electrical characteristics are high withstanding voltage, low on-state resistance, high operational temperature, fast switching speed, etc. Concerning the efficiency improvement of AC/DC converter, the temperature dependence of electrical characteristics of switching element is the important issue on operational loss reduction. On-state resistance and allowable operational temperature of switching element are directly related to the operational loss of AC/DC converter. The former one is the main parameter of conductive loss, which is dominant one of the operational loss of high current AC/DC converter, and it is effective to reduce the conductive loss for efficiency improvement. The latter one is related to the cooling capability of cooling equipment, which is a main auxiliary component of large capacity AC/DC converter. The temperature dependence of on-state resistance of SiC power device will be expected to reduce the conductive loss and to simplify the cooling equipment of AC/DC converter. The allowable operational temperature of SiC power device is higher than that of Si power device, therefore it allows also to minimize the cooling capability. Both electrical characteristics of SiC power device will contribute to reduce the operational loss and to improve the operational efficiency of AC/DC converter. Previously, the operational loss of large capacity AC/DC converters of ITER class power supply was studied based on the predicted on-state resistance of future SiC power device. In this paper, with the results of measurement in some experiments for present unipolar SiC device, the effect of electrical characteristics of SiC power device is mainly discussed on efficiency improvement. And, the temperature dependence of the electrical characteristics will be mentioned in comparison with conventional Si power device.


Corresponding Author:

Tatsuya Matsukawa

Department of Electrical and Electronic Engineering, Mie University, 1515, Kamihama-cho, Tsu, Mie, 514-8507, JAPAN

- E - Magnets and Power Supplies.

P2T-E-338 DESIGN REQUIREMENT, QUALIFICATION TESTS AND INTEGRATION OF A THIN SOLID LUBRICANT FILM OF MOS2 FOR COLD MASS SUPPORT STRUCTURE OF THE STEADY STATE SUPERCONDUCTING TOKAMAK SST-1.

Doshi Bharat1, B. Sarkar1, Pratima Rewatkar1, Dasharath Sonara1, C.Ramdas1, K.J.Thomas1, Saxena Y.C.1, V.R.Bhaskar2, Senthil Kumar2 and S.Nagbhushanam2

(1) Institute for Plasma Research, Bhat, Gandhinagar, Gujarat, India 382 428 (2) ISRO Satellite Center, Bangalore, India The SST-1 is a super conducting tokamak, which is in the final phase of assembly and commissioning. The super conducting magnet system of SST1 comprises of Toroidal field (TF) and Poloidal field (PF) coils. The 16 TF coils are nosed and clamped towards the in-board side and are supported toroidally with inter-coil structure at the out-board side, forming a rigid body system. The 9 PF coils are clamped on the TF coils structure. The integrated system of TF coils & PF coils forms the cold mass of @ 50 Ton weight. This cold mass is accommodated inside the cryostat and freely supported on the rigid support ring at 16 locations and support ring in-turn supported on 8 columns of machine support structure. During the operation this cold mass attains a cryogenic temperature of 4.2K in the hostile environment of high vacuum. The thermal excursion of cold mass and its supporting structure during this cool down results into severe frictional forces at the supporting surfaces. There is a design requirement of introducing a thin layer of solid lubricant film of MOS2 having coefficient of friction 0.05 between the sliding surfaces to control the stress contribution due to the friction. To ascertain the compatibility of molybdenum disulphide (MOS2) as a solid lubricant in high vacuum and very low temperature environment, we have carried out qualification tests on various samples and measured the coefficient of friction in both the room temperature conditions and at high vacuum & after thermal shocking to 4.2K temperatures. After successful qualification tests actual components are fabricated and integrated in the cold mass support structure assembly. This paper presents the design requirement, qualification tests performed and details about the integration of thin solid lubricant film of MOS2.


Corresponding Author:

Doshi Bharat1

Institute for Plasma Research, Near Indira Bridge, Bhat, Gnadhinagar 382 428 INDIA

- E - Magnets and Power Supplies.

P2T-E-344 HOW SHOULD WE TEST THE ITER TF COILS ?

LIBEYRE Paul, CIAZYNSKI Daniel DUCHATEAU Jean-Luc SCHILD Thierry (1) FIETZ Walter H. (2) ZAHN Gernot (2)

(1) CEA/DSM/DAPNIA CEA Saclay F-91191 Gif-sur-Yvette cedex (France) (2) Association Euratom-Forschungszentrum Karlsruhe, Institut für Teschnische Physik, Helmholtzplatz 1 D-76344 Eggenstein-Leopoldshafen (Germany)

The ITER TF coils are a major piece of the ITER magnetic system. It is of prime importance that they operate reliably during the whole life of the machine, since a failure in these coils during operation would cause a major breakdown in the programme and would lead to a difficult repair procedure. The manufacture of these coils will thus include a very strict quality assurance system at each step from the strand production to the final closure welding of the case around the winding-pack in order to avoid any defect. Nevertheless, these coils will be the largest Nb3Sn coils ever built and will operate at high magnetic field with high current and will be submitted as well to very large mechanical loads as high voltages. It is therefore necessary that their behaviour be well established. The paper addresses the advantages and drawbacks of the main options which can be considered. Particular attention is put on the information provided by testing at helium temperature each completed coil with respect to extensive testing of component samples. The critical points to be checked are the dielectric strength of the insulation, the internal joints resistance, the temperature margin of the conductor, the hydraulic resistance of the cooling circuit. Several testing configuration are considered and their impact in terms of cost and time schedule are estimated. An other important aspect of the cold tests is to verify the correct operation of the coil safety system in relevant conditions. The cold tests are the final reception tests of the coils and should therefore bring the guaranty that the coil will be able to operate safely at its nominal field and current. When assembled together to form a toroid, the operating conditions of the TF coils are highly depending on the other coils of the magnet. The maximum field applied to the conductor is reaching 11.8 T at a nominal current of 68 kA when the coil is inside the TF magnet, whereas when tested alone, the maximum field is hardly exceeding 6 T. On the other hand, in a single coil test the coil experiences only in-plane loading, whereas during plasma operation both in-plane and out-of-plane loads are applied to the coil. In the purpose of achieving more relevant operating conditions, an investigation of the possibility of testing simultaneously several coils is performed and discussed.


Corresponding Author:

LIBEYRE Paul

Association Euratom-CEA CEA/DSM/DRFC CEA Cadarache F-13108 St Paul lez Durance cedex (France)

- E - Magnets and Power Supplies.

P2T-E-379 CYCLIC TESTING OF SHEAR KEYS FOR THE ITER MAGNET SYSTEM

Rossi Paolo, L.F. Moreschi (ENEA CR Brasimone) A. Pizzuto (ENEA CR Frascati) S. Storai (ENEA CR Brasimone) C. Sborchia (EFDA CSU)

ENEA CR Frascati, PB 65, 00044 Frascati, (Roma), Italy ENEA CR Brasimone, PB 1, 40032 Camugnano BO), Italy EFDA Close Support Unit, Boltzmannstrasse 2, D-85748 Garching, Germany

Shear keys are to be used to support the out-of-plane loading of the Toroidal Field (TF) coils during a plasma pulse in ITER. At the Inner Intercoil Structures (IIS) a set of poloidal shear keys is used to take the shear load at each connection between adjacent TF coils. Solid circular keys have been selected as reference. At the Outer Intercoil Structures (OIS) adjustable conical shear keys and friction joint based shear panels are used to take the shear load. Low voltage electrical insulation is required at the flanges of the IIS and OIS, plus for all the bolts, poloidal keys and adjustable keys. This electrical insulation has to withstand large compression associated with some shear or slippage. A ceramic coating was selected for this purpose. The main scope of the experimental campaign was the mechanical testing of the shear keys and the electrical insulation in operational conditions relevant to ITER. Both keys were made of Inconel 718, provided with a ceramic alumina coating and inserted into flanges made of cast AISI 316 LN. The adjustable conical shear key was pre-loaded at room temperature and subject to cyclic shear loads of 2.5 MN for a large number of cycles (about 30,000) at cryogenic temperature (77 K). The conical key and the alumina coating resulted undamaged after test. Another test campaign was then performed with higher shear loads (up to 3 MN) to reach a sufficient safety margin even with the friction effect due to the pre-load. A set of 15,000 cycles were completed followed by some cycles at higher loads to reach the ultimate limit, which is the shear load to be experienced by the key in case of a Poloidal Field (PF) coil short.


Corresponding Author:

Rossi Paolo

ENEA CR Frascati, PB 65, 00044 Frascati, (Roma), Italy

- E - Magnets and Power Supplies.

P2T-E-390 MODULAR COIL DESIGN DEVELOPMENTS FOR THE NATIONAL COMPACT STELLARATOR EXPERIMENT (NCSX)

WILLIAMSON, David E., A. Brooks (1), T. Brown (1), J. Chrzanowski (1), M. Cole (2), H-M. Fan (1), K. Freudenberg (3), P. Fogarty (2), T. Hargrove (4), P. Heitzenroeder (1), G. Lovett (5), P. Miller (5), R.L. Myatt (6), B. Nelson (2), W. Reiersen (1), D. Strickler (2)

(1) PPPL, PO Box 451, Princeton, NJ 08540 (2) ORNL, PO Box 2008, Oak Ridge, TN 37831 (3) BWXT, PO Box 2009, Oak Ridge, TN 37831 (4) Hargrove Engr, Scottsboro, AL 35768 (5) MK Tech, Knoxville, TN 37930 (6) Myatt Consulting, Norfolk MA 02056

The National Compact Stellarator Experiment (NCSX) is a quasi-axisymmetric facility that combines the high beta and good confinement features of an advanced tokamak with the low current, disruption-free characteristics of a stellarator. The experiment is based on a three field period plasma configuration with an average major radius of 1.4-m, a minor radius of 0.32-m, and a toroidal magnetic field on axis of up to 2-T. The modular coils are one set in a complex assembly of four coil systems that surround the highly shaped plasma. There are six each of three coil types in the assembly for a total of 18 modular coils. The coils are constructed by winding flexible, copper conductor onto a stainless steel winding form that has been cast and machined to high accuracy, so that the current center of the winding pack is within +/-1.5-mm of theoretical. The modular coils operate at 80-K and produce the primary magnetic shaping field for a flat-top pulse length of 0.5-s at a current of 820-kAT. Due to geometry constraints, the coil windings must operate at high current density and are subject to rapid heating and thermal stress during a pulse. In addition, the coils experience electromagnetic forces of up to 1.2-MN/m. This paper will discuss the progression of the coil design from physics targets to filamentary models to prototype components, which are currently being fabricated. Advances in inspection technology, field error compensation analysis, and assembly simulation will be highlighted.


Corresponding Author:

WILLIAMSON, David E.

Oak Ridge National Laboratory, Post Office Box 2008, Oak Ridge, TN 37831-6169

- E - Magnets and Power Supplies.

P2T-E-394 CONCEPTUAL DESIGN OF SPHERICAL TORUS WITH TF-CS HYBRID COILS BASED ON VIRIAL THEOREM

TSUTSUI Hiroaki, NOMURA Shinichi, TSUJI-IIO Shunji, SHIMADA Ryuichi

Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo 152-8550, Japan

A conceptual design of a spherical torus (ST) device with a new type of toroidal field (TF) coils and a central solenoid (CS) whose stress is reduced to the theoretical limit determined by the virial theorem is proposed. In the last decade, we had developed a tokamak with force-balanced coils (FBCs) which are multi-pole helical hybrid coils combining TF coils and a CS coil. The combination reduces the net electromagnetic force in the direction of major radius by canceling the centering force due to the TF coil current and the hoop force due to the CS coil current. This excellent feature of FBC and its capability of tokamak operation were investigated and demonstrated by the first FBC tokamak “Todoroki-I'', while working stress in coils has not yet been investigated whereas the net electromagnetic force is reduced. Next, we had extended the FBC concept using the virial theorem which shows that strength of magnetic field is restricted by working stress in the coils and their supporting structure. High-field coils should accordingly have same averaged principal stresses in all directions, whereas conventional FBC reduces stress in the toroidal direction only. In that work, we had obtained the poloidal rotation number of helical coils which satisfied the uniform stress condition, and named the coil as virial-limit coil (VLC). VLC with a circular cross section of aspect ratio A=2 reduces maximum stress to 60% compared with that of TF coils. A tokamak discharge in VLC was also demonstrated by the first VLC tokamak “Todoroki-II”. Recently, we have developed a VLC concept with a non-circular cross section, and reduce the maximum stress to 30% compared with that of TF coils in a two dimensional analysis. Moreover, the VLC configuration has a low aspect ratio and a strongly elongated cross section with a triangularity, and is similar to that of a ST, while a VLC is a helical coil. In this work, we analyze three dimensional stress distributions, and evaluate operation scenarios in VLC ST.


Corresponding Author:

TSUTSUI Hiroaki

Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo 152-8550, Japan

- E - Magnets and Power Supplies.

P2T-E-405 EMI ON DIAGNOSTICS AND CONTROL CIRCUITS DUE TO SWITCHING POWER SUPPLIES

Gaio Elena, R. Piovan, V. Toigo

Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, C.so Stati Uniti 4, 35127 Padova, Italy

In fusion experiments, the use of switching power supplies is becoming more and more frequent. These equipments are characterized by instantaneous output voltages quickly varying with the switching frequency; moreover, the output terminals, usually connected to the machine windings, present common mode voltages varying very fast as well. The typical voltage variation speed is in the order of some kV / microseconds. The common mode voltages can excite the circulation of noise currents through parasitic capacitances, which couple different parts of the circuit designed to be isolated one to each other. The harmonic spectra of these currents range between some hundreds of kHz to some MHz. The windings are usually connected to ground through high impedances; nevertheless, the parasitic capacitances between the coils, the mechanical supporting structure and ground represent low impedance electric connections in high frequency ranges. As a consequence, currents can flow in the reference potential grids, which disturb the plasma diagnostic equipments; moreover, direct capacitive coupling between the switching power supply loads and the diagnostics themselves can cause the circulation of noise currents in the measurements circuits. These general problems have been analyzed in detail for the RFX case, where switching power supply have been introduced both in the poloidal and toroidal circuits for performing various control actions on the plasma. The analyses showed that without any provisions the common mode currents in the machine reference potential conductors can reach values in the order of amperes. In RFX, EM interferences, again referable to the same phenomenon, were observed in the control sections of the power supplies too. Faults on the drivers of dc static current breakers happened, due to the presence of very high noise currents in the isolated power supplies of the switching devices’s firing system; the currents were produced by the fast varying common mode voltage present at the semiconductors power terminals. Equivalent electric schematizations of these phenomena have been derived and numerical simulations have been worked out, which explain the induced noises and have been utilized to identify suitable correction measures; in the paper these analyses are reported and the experience developed in coping with the reduction of these types of EMI interferences is described.


Corresponding Author:

Gaio Elena

Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, C.so Stati Uniti 4, 35127 Padova, Italy

- E - Magnets and Power Supplies.

P2T-E-406 THE CONTROL SYSTEM OF THE TOROIDAL POWER SUPPLY OF RFX

Piovan Roberto, V. Toigo (1), L. Zanotto (1), M. Perna (2), A. Coffetti (2), M. Freghieri(2), M. Povolero (2)

(1)Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 (2)ASIRobicon, Viale Sarca 336 - 20126 Milano, Italy

This paper deals with the control system of the rebuilt toroidal field power supply of RFX. In a Reversed Field Pinch, such as RFX, the waveform of the toroidal field is much more complicated than in a tokamak, as the current in the toroidal field winding needs to be reversed in polarity and then regulated to produce either rotating m=0 harmonics or Poloidal Current Drives. The toroidal field power supply of RFX, which represents the first example of static power supply for fusion experiment based on components named IGCTs, combines a very complex circuit topology and a fast, reliable and flexible control system. The peculiarity of this system is the integration of many control functions in a compact solution; the control design has required particular efforts due to the different kind and nature of devices working together, which includes large capacitor banks and fast and unconventional power electronics apparatus, such as dc/ ac inverters, choppers and static circuit breakers. Both slow and fast control algorithms regarding the supervision of the plant, the protection system and the current and voltage regulation in the winding sectors are implemented in a unique hardware arrangement. The design criteria of the toroidal power supply control system will be discussed in the paper. The most important guideline in defining the specifications has been the flexibility, which is strictly related to the possibility of easily changing the experimental set-up according to requests coming from different scenarios. Moreover, to simplify the commissioning of the power supply, efforts have been put in designing a special part of the control section to perform local tests. In such a way an easy setting-up of the system parameters is possible, thus allowing to test both a single device and the whole system during an experimental sequence. Another important issue discussed in the paper is the definition of the active plant protection strategies: the complexity of the system, which presents many different fault cases and operative scenarios, has led to very fast and sophisticated protection algorithms, integrated in the control structure. The control hardware architecture derived from the above-mentioned considerations is based on two VME crates, each including a PowerPC board and some FPGA general-purpose boards. The paper will present a detailed description of the control functions and of the hardware structure; the software architecture will be also described.


Corresponding Author:

Piovan Roberto

Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127

- E - Magnets and Power Supplies.

P2T-E-417 COMPONENTS AND SYSTEM TESTS ON THE RFX TOROIDAL POWER SUPPLY

Perna Mauro, V. Toigo (1), L. Zanotto (1), E. Gaio (1), P. Bordignon (2), A. Coffetti (2), R. Novaro (2), P. Bertolotto (3), E. Rinaldi (3), G. Villa (3)

(1) Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova, Italy (2) ASIRobicon, Viale Sarca 336 - 20126 Milano, Italy (3) Passoni&Villa, viale Suzzani 229 - 20162 Milano, Italy

The final Site tests of the new RFX toroidal power supply system have been concluded by the end of 2003 and the system is now ready for the integration tests with RFX machine. This system represents the first example of static power supply for fusion experiments, based on recent power semiconductors named IGCTs (Integrated Gate Commutated Thyristors). Besides producing the time-dependent toroidal magnetic field required to set up the RFP configuration, it is also devoted to generate and control the toroidal magnetic field waveforms required to produce rotating m=0 field harmonics for applying a torque to the plasma or special Poloidal Current Drives. The system design was peculiar not only for the single components, which are developed “ad hoc” and characterised by a high degree of technological innovation, but also for the system coordinated operation, which is unique and very articulated, due to the high flexibility level required. Special factory tests on the main device prototypes have been performed to verify the critical aspects of the design. In the static breaker (4 kV, 16 kA, 128 MA2s), which represents a remarkable example of static dc current interruption technology at high power, it was necessary to realize five parallel IGCTs branches. Reaching good performances in terms of current sharing and limited reapplied overvoltages was not so straightforward. Also for dc/ac inverters (3 kV, 6 kA), composed of three single-phase IGCTs H-bridges in parallel, current sharing optimization has been a very ambitious goal: they represent the first industrial realization of parallel connection among IGCTs bridges; for both, the test results were very satisfactory. In the capacitor bank design (4 kV, 16 mF, 128 kJ), the most peculiar aspect is the protection against internal faults: the fuse design had to satisfy many different requirements: very fast intervention, less than 100 ms, fault discrimination at different bank voltage levels, no explosion in the worst fault conditions; also in this case, special tests have been performed. To achieve the required coordinated operation of all these devices was also a big task; all the necessary tests were performed on Site; the control system was designed to assure a high flexibility level and allowed the integration of one device at a time. The most peculiar aspects of the integration tests and the optimisation of the whole system operation will be described and discussed in the paper.


Corresponding Author:

Perna Mauro

ASIRobicon, Viale Sarca 336 - 20126 Milano, Italy

- E - Magnets and Power Supplies.

P2T-E-420 COMMISSIONING AND OPERATION OF 130KV/130A SWITCHED-MODE HV POWER SUPPLIES WITH THE UPGRADED JET NEUTRAL BEAM INJECTORS

David C Edwards, Marco Bigi (1), Denis Brown (1), Daniel Ganuza (2), Francisco Garcia (2), Zachary Hudson (1), Timothy Jones (1) and Alberto Perez (2)

(1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB UK (2) Grupo JEMA, E-20160 Lasarte-Oria, Guipúzcoa, Spain

The JET neutral beam (NB) upgrade increased the current of eight Positive Ion Neutral Injector (PINI) beam sources from 30A to 60A at 130kV acceleration voltage. This involved: (1) procurement of two new HV Power Supply (HVPS) units each rated at 130kV/130A (20s pulse length, 1/30 duty cycle) to feed two 130kV/60A PINIs; (2) reconfiguration of four existing HVPS modules (rated at 160kV/60A) to feed four individual PINIs. The new HVPS units are of all solid-state switched-mode type, where 120 high-frequency IGBT invertor modules feed 120 isolation transformers whose rectified outputs are connected in series. Regulation and fast switching for control and load protection are done on the LV side, and the isolation transformers provide a passive barrier to the transfer of energy in case of IGBT failure. Up to 255 re-applications allow for repetitive HV breakdowns in the load, detected when the current exceeds a pre-defined first threshold. Two independent optically triggered thyristor crowbars across the HV output act if the load current reaches a second threshold. The design of the HVPS will be discussed, including load protection characteristics in comparison to conventional switch-tube designs. The paper describes on-site testing, commissioning and operation of the new HVPS units. Both units were factory tested at full power and pulse length into dummy resistive load. Following on-site installation, the factory tests were repeated. The transition from dummy-load testing to PINI load operation required full integration of the HVPS within the overall JET control system, and rigorous testing of the co-ordinated actions and protections of all PINI power supplies (filament and arc for plasma source and negative suppression grid). The implementation of the ‘fast logic’ electronics for these functions will be described. Extensive use was made of parasitic integrated test pulses, where the other PINIs could be operated normally, and the HVPS was energised under full remote control together with the corresponding PINI plasma sources, but with the HVPS connected to dummy load (or open circuit). The amount of NB operation time dedicated to commissioning was thereby minimised, yet gave a high degree of confidence of readiness for HV energisation of the PINI, and first beam extraction occurred within less than 24 hours of HV connection to the PINI. Finally, the routine operating experience, such as performance and reliability, of the new HVPS units will be described.


Corresponding Author:

David C Edwards

Building J20/1/36, Culham Science Centre, ABINGDON OX14 3DB UK

- E - Magnets and Power Supplies.

P2T-E-427 MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: TESTS ON LOW VOLTAGE CIRCUIT BREAKERS AND CONTACTORS

Grando Luca, Antonio De Lorenzi (1) Giulio Bettanini (1) Daniele Desideri (2)

(1) Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova Italy (2) Dept. of Electrical Engineering, University of Padova, Via Gradenigo 6/a, 35131 Padova Italy

The ITER tokamak building will be permeated by an almost constant magnetic flux density up to 70 mT, generated by the superconducting winding system. This operation condition asked for a research activity aimed at assessing the magnetic field immunity of components forming the Low Voltage power supply system. In this work, the behaviour of some electro-mechanical components and of a Low Voltage cubicle under static or slowly variable magnetic field has been investigated. A selection of the most recent industrial low voltage current breakers and contactors of the major European manufactures in the 50 ÷ 250 A range, equipped with different kind of drivers and protection relays, has been tested by applying a nearly uniform and static magnetic flux density B up to 80 mT. Off load, on load, protection trip and life tests have been performed for each component following a specific test procedure and for life tests the directions given by the EN Standards have been applied. In general, the results showed that the components are sensitive to amplitude, direction and versus of the magnetic flux density starting from 10 mT. The test campaign has been completed investigating the behaviour of a fully equipped power center module, consisting in a iron cubicle equipped with different circuit breakers and contactors (of the same type of those previously tested), remotely controlled by a PLC via Profibus optical interface –a structure similar to what is expected in the ITER Low Voltage installations – and immersed in a 40 mT magnetic applied with a repetitive waveform. In order to assess the cubicle shielding efficiency, a comparison between the magnetic flux density values, computed in vacuum and the magnetic field measured inside the cubicle has also been performed. Besides, nearly 500 operations have been carried out switching on and off some resistive loads during the B field pulse. Under these conditions, the overall behaviour was consistent with the results obtained with the individual components. This work is performed under the R&D contract 02/1010 between the European Fusion Development Agreement (EFDA) and the Consorzio RFX.


Corresponding Author:

Grando Luca

Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova Italy

- E - Magnets and Power Supplies.

P2T-E-442 FIRST INTEGRATED TEST OF THE SUPERCONDUCTING MAGNET SYSTEMS FOR THE LEVITATED DIPOLE EXPERIMENT (LDX)

A. Zhukovsky, Joseph V. Minervini (1) P. C. Michael (1) J. H. Schultz (1) B. A. Smith (1) J. Kesner (1) A. Radovinsky (1) D. Garnier (2) M. Mauel (2)

(1) MIT Plasma Science and Fusion Center 77 Massachusetts Avenue Cambridge, MA 02139 USA (2) Columbia University Department of Applied Physics and Applied Mathematics Room 210 S. W. Mudd Building New York, NY 10027 USA

The Levitated Dipole Experiment (LDX) is an innovative approach to explore the magnetic confinement of a fusion plasma offering the possibility of a fusion power source with near-classical energy confinement. In this concept a magnetic dipole field is created by a superconducting solenoid which is magnetically levitated for up to 8 hours in the center of a 5-meter diameter vacuum vessel. The Floating Coil (F-coil) is designed for a maximum field of 5.3 T. A react-and-wind Nb3Sn conductor was selected to enable continuous field generation as the coil warms from an initial temperature of about 5 K at the start of the experimental day up to a final temperature of about 10 K at the end of the operating day. The F-coil is maintained in the center of the plasma volume by the Levitation Coil (L-coil). This coil is made from high temperature superconductor (Bi-2223) to minimize the electrical and cooling power needed for levitation. It is a 2800 turn, double pancake winding that supports the weight of the F-coil and controls its vertical position within the vacuum chamber. Since the F-coil must operate in a levitated position it is not desirable to have electric or cryogenic feeders serving the coil through the plasma. For this reason, the coil is inductively charged/discharged and cooled cryogenically in a lower charging station. The operating current in the F-coil is induced by the Charging Coil (C-coil) when it is resting in the charging port at the bottom of the LDX vacuum vessel. The C-Coil is a superconducting solenoid using NbTi operating in a liquid helium cryostat that surrounds the 1157 mm diameter charging station. It stores 8 MJ of energy at an operating peak field in the winding pack of 4 T. The L-coil and C-coil have each been independently tested. The F-coil cryogenic test is under preparation. This paper describes the results of the final assembly of the LDX experiment and the first integrated test of the F-coil and the C-Coil in the normal LDX operating configuration.


Corresponding Author:

ALEX ZHUKOVSKY

Massachusetts Institute of Technology, Plasma Science and Fusion Center, 77 Massachusetts Avenue, NW22, Cambridge, MA 02139 U.S.A.

- E - Magnets and Power Supplies.

P2T-E-462 MODELING AC LOSSES IN THE ITER NBTI FULL SIZE JOINT SAMPLES USING THE THELMA CODE

ZANINO Roberto, M.Bagnasco 1, F.Bellina 2, P.Gislon 3, P.L.Ribani 4, L.Savoldi Richard 1

1 Dipartimento di Energetica, Politecnico, Torino, Italy 2 Dipartimento d’Ingegneria Elettrica, gestionale e Meccanica, Università di Udine, Italy 3 ENEA, Frascati, Italy 4 Dipartimento di Ingegneria Elettrica, Università di Bologna, Italy

THELMA code is a tool for the numerical simulation of the behaviour of cable-in-conduit multistrand superconductors (CICCs), like those to be used for the magnets of the International Thermonuclear Experimental Reactor (ITER). Compared to other similar codes, the peculiarity of THELMA is the possibility to analyse long CICC lengths, electrically connected each other by means of resistive joints, as foreseen for the ITER magnets. The model solution is obtained solving simultaneously the electromagnetic and the thermal-hydraulic coupled problems. THELMA is presently under debugging phase, and its first results have been obtained and compared with experimental results. The paper presents the results of the analysis of two NbTi Full-Size Joint Samples, tested at the Sultan facility of PSI in Villigen (CH). The first is the “PF-FSJS”, tested in 2002, and the second is the “PFCI-FSJS”, tested in 2004, in the framework of the ITER R&D activities. Both the samples are made of two “legs” (two straight parallel vertical pieces of CICC), each approximately 3.5 m long, electrically connected in series at their bottom with a resistive joint. The samples are cooled with super-critical helium in forced convection at typically 5 K and 1 MPa. In the PF-FSJS the flow, in the whole CICC cross section, is directed downwards, whereas in the PFCI-FSJS, the flow is restricted to the annular region and directed upwards. For each sample, the two legs are of a different type: in the PF-FSJS the main difference consists of the type of strand, while in the PFCI-FSJS the major difference is the presence or the absence of wrappings around the CICC subcables. Both the samples are fully instrumented with magnetic field, voltage, and temperature sensors. Therefore an exhaustive comparison between the experimental and the computed quantities is possible. The tests are mainly aimed at the CICC and the joint characterization. We concentrate here on the measurement of the AC losses, induced in the cable by the magnetic field generated by a pulsed coil. In the THELMA model, one leg is considered, and the CICC is represented as 6 cable elements corresponding to the CICC petals. The magnetic coupling between the two legs and with the AC field coils is taken into account. In the paper, the distribution temperature and losses are reported as a function of time and space, and the comparison with the experimental results is presented.


Corresponding Author:

ZANINO Roberto

Dip. di Energetica, Politecnico, Troino, I-10129 Italy

- E - Magnets and Power Supplies.

P2T-E-471 POWER DISSIPATION AND ENERGY TRANSFER DURING TESTING OF THE ITER TOROIDAL FIELD MODEL COIL

Marchese Vito, W.H. Fietz, R. Heller, M. Süsser, F. Wüchner, G. Zahn

The test of the ITER Toroidal Field Model Coil (TFMC) in the background field of the EURATOM-LCT coil took place in autumn 2002 at the TOSKA facility of the Forschungszentrum Karlsruhe in the framework of the ITER R&D programme. The maximum currents in the two coils, in combined operation, were 16 kA in the LCT coil and 80 kA in the TFMC respectively. Eddy currents are generated in the passive structures (e.g., stainless steel radial plates and coil cases) which are magnetically coupled to the coil windings, leading to eddy current losses during current ramp up and down, as well as during a fast discharge. Due to the ripple of the power supplies, based on twelve pulse ac-dc thyristor converters, losses were also generated during flat top. Two He refrigerators of 2 kW and 0.5 kW respectively were used for the primary loop of the test configuration. The heat load of both coils, including the eddy current losses in the passive structures and the Joule losses due to the joint resistances, was removed by a secondary loop of forced flow supercritical He. Both the TFMC and the LCT coil windings were cooled in series with their respective coil cases and Inter Coil Structure (ICS) in order to reduce the overall mass flow rate. About 2% of the stored energy was transferred to the cryogenic system after all the safety discharges and quenches of both coils together. Most of the energy (about 98%) was extracted and transferred to the dump resistors of both coils, located outside the vacuum vessel. The evaluation of the eddy currents and the power losses in the two coil cases and in the TFMC radial plates, the latter one cooled indirectly through the TFMC windings, for different operating conditions, has been performed with a SIMULINK computer code based on the full inductance and resistance matrices. In addition to the two coil windings, a linear model of the power supplies and their current controllers and the dump resistors was applied. The program computes also the heat load of the secondary loops based on a simplified thermo hydraulic model, incorporating energy conservation and the heat transfer equations (0D), whose parameters have been validated experimentally with calorimetric measurements. The program has been used to evaluate the energy losses transferred to the cryogenic plant and to the external power circuit for the simultaneous ramping down of the currents in both coils and for the safety discharge.


Corresponding Author:

Marchese Vito

Institute für Technische Physik, Forschungszentrum Karlsruhe GmbH, P.O. Box 3640, D - 76021 Karlsruhe, Germany

- E - Magnets and Power Supplies.

P2T-E-490 DC AND TRANSIENT CURRENT DISTRIBUTION ANALYSIS FROM SELF-FIELD MEASUREMENTS ON ITER PFIS CONDUCTOR

Formisano Alessandro, T. Bonicelli (2), Yu. Ilyin (3), A. Nijhuis (3), C. Marinucci (4), R. Martone (1), L. Muzzi (5), M. Polak (6), C. Sborchia (2), B. Stepanov (4), S. Turtù (5)

(2) EFDA Close Supp. Unit, Garching, Germany. (3) Univ. of Twente, Enschede, the Netherlands. (4) EPFL-CRPP, Villigen PSI, Switzerland. (5) ENEA Frascati, Frascati, Italy. (6) Inst. Electr. Eng., Slovak Academy of Sciences, Bratislava, Slovakia.

Nowadays a number of studies are being performed to investigate actual Cable-in-Conduit Conductors (CICC) behaviour under condition of practical interest for the ITER (International Thermonuclear Experimental Reactor) magnets. Many aspects of such conductors are under examination, among which the impact of working conditions on the current distribution among the conductor sub-structures. Unfortunately, direct measurements of the current profile inside the conductor are not possible. An indirect approach for the current profile estimate could be based on the measurement of the magnetic self-field around the CICC. The current distribution should then be reconstructed using inverse problems methodology. In the present work, two approaches to such problem are presented and compared. The first model adopted in the inverse problem formulation is based on magnetostatic equations, and is able to take into account the internal structure of the cable (3D), as well as the effect of external stray fields and other disturbance sources. In contrary, the second model uses somewhat simplified geometry (2D) of the conductor, thus saving time for modelling and computation. A comparison between two models should reveal an impact of the accuracy in the geometrical representation of a cable on the result of the current reconstruction. The performance of the measurement system is crucial, due to the extremely ill conditioning of the inverse problem; consequently, Hall Probes assemblies (called “heads”) must be optimally designed and placed around the cable, in order to reach a satisfactory trade-off between sensitivity and robustness in the measurement process. Some considerations about the Hall sensors quality and the impact of geometrical uncertainties on the reconstruction process are also discussed in the paper. The reconstruction procedures have been applied on the PFIS (Poloidal Field InsertConductor Sample) tested in the SULTAN test facility under various working conditions. Three Hall Probe heads have been placed on the conductor sections being in a high field region and in a low field one. Details about the experiment and the results of the current reconstruction by two analytical approaches are presented and discussed.


Corresponding Author:

Formisano Alessandro

Dip. di Ingegneria dell’Informazione, Seconda Universita' di Napoli, Via Roma 29, I-81031, Aversa (CE), Italy

- E - Magnets and Power Supplies.

P2T-E-491 THE MAGNET SYSTEM OF THE KTM TOKAMAK

Bondarchuk Eduard, E.N. Bondarchuk (1) A.A. Malkov (1) V.A. Korotkov (1) S.A. Krasnov (1) Y.M. Krivchenkov (1) V.A. Krylov (1) A.B. Mineev (1) A.K. Cherdakov (1) E.A. Azizov (2) V.N. Dokouka (2)

(1) D.V. Efremov Institute, STC ‘Sintez’, Metallostroy, Doroga na Metallostroy 1, 196641 St. Petersburg, Russia (2) Troitsk Institute for Innovation’s and Thermonuclear Researches, 142190 Troitsk, Russia

Magnet system of the KTM tokamak has been designed to provide certain shape and evolution of single null plasma with aspect ratio A = 2. It makes possible, from one hand side, to create a compact and, relatively, cheap machine that enables to solve requested tasks, and, from another hand side, to get required physical parameters of plasma-magnet configurations that are between values typical for classical (A ? 2.5) and, so called, spherical (A ? 1.6) tokamaks. Magnet system is a resistive pulsed system and consists of 20 toroidal field coils (TF), 6 poloidal field coils (PF) and central solenoid (CS). The PF coils are located inside the TF system. To make possible the installation of the one-piece welded vacuum vessel the TF coils are designed to have joints. The TF coils produce a field of 1 Tesla at the plasma center of 0.86 m. Maximum plasma current is 0.75 MA. Pulse duration is around 4 seconds. The cooling of the windings is ensured by pumping water through the cooling channels. Magnet system has been designed to operate for the various of loading conditions including temperature distributions in windings due to Ohmic heating, weight loads, electromagnetic forces acting at the normal operation as well as for the scenario of central or vertical plasma disruption. Results of structural analysis of the magnet system show that stresses in the conductors and support structures are within the allowable limits. The CS coil is 4 layer wound coil and has 8 cooling passes. To satisfy the requirements on the mechanical strength the silver bronze is selected as the material for both the CS and central part of the TF conductors. Magnet system has been designed to withstand 20000 full size pulses.


Corresponding Author:

Bondarchuk Eduard

Efremov Institute, STC 'Sintez', Metallostroy, Doroga na Metallostroy 1, P.O. Box 42, 196641 St. Petrsburg, Russia

- E - Magnets and Power Supplies.

P2T-E-511 OPTIMISATION OF THE CURRENT DISTRIBUTION IN THE IGNITOR POLOIDAL FIELD COILS AND EVALUATION OF THE COILS TEMPERATURES AND RESISTANCE DURING THE REFERENCE OPERATING SCENARIO

Rita Camillo, Cocilovo Walter (1), Cucchiaro Antonio (1), Galasso Giuseppe (2), Pizzuto Aldo (1), Ramogida Giuseppe (1), Roccella Massimo (1), Prof. Coppi Bruno (3)

(1) Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy (2) Ansaldo Ricerche, Corso Perrone 25, 16152 Genova (GE), Italy (3) MIT, 02139 Cambridge (Ma), USA

The Ignitor Poloidal Field Coil (PFC) system consists of 13 pairs of coils up down symmetric, closely distributed around the plasma column. The main relevant components are a Central Solenoid (CS) that includes 7 coil pairs located between the Toroidal Field Coils (TFC) and the Central Post (CP), and a set (6 pairs) of external coils. The high number of independent coil pairs guarantees robust elongated plasma equilibria while optimising the flux coupling with the plasma. The balance between the PFC flux capability and the plasma flux requirement is one of the most crucial issues in designing compact high field toroidal machines. The high current values in the coils needed to assure the plasma flux requirements could produce high forces and temperature increases on the coils resulting in significant stresses in the mechanical structure. The design of the IGNITOR tokamak requires a careful analysis because of the structural performance of the machine relies on an optimised combination of bucking, between TF coils and the central solenoid, and wedging, among the toroidal magnet legs. In this context the evaluation of the poloidal fields temperature and resistances is of relevant importance for the IGNITOR design, because the low initial temperature of the coils (30 K) and the high magnetic fields involved increase the role played by the magneto-resistance effect, that produces a significant temperature gradient in the central solenoid coils. To calculate the poloidal coils temperatures and resistances, during the whole IGNITOR reference scenario has been used a 2D axisymmetric integral code developed in ENEA and its results have been compared with those one achieved with the MAXWELL FEM code. Our code calculates both electromagnetic forces and resistivity for every turn of the coils as function of the magnetic field and temperature through all the expected current scenario, for the reference IGNITOR discharge at 11 MA. It is then calculated the increase of temperature due to the Joule effect to get the maximum temperature value for each coil. Due to the short duration of the IGNITOR discharge the heat transfer mechanism can be approximated as completely adiabatic, obtaining so a conservative approach. The code results has evidenced critical temperature values on the inner central coils, suggesting the opportunity of a current density redistribution. This goal has been obtained using a grading technique for such coils.


Corresponding Author:

Rita Camillo

Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy

- E - Magnets and Power Supplies.

P2T-E-528 SAFETY ASSESSMENT OF THE ITER COILS SYSTEM

Raboin Serge, J.-L. Duchateau (1) and EISS Team

(1) Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

In the ITER experimental reactor, the thermonuclear plasma is magnetically confined by means of a complex system of superconducting magnets. The use of the superconducting technology plays a pivotal role in the viability of the fusion technology, such magnets allowing substantial gains in electrical power consumed in the operational phase of thermonuclear reactors. Superconducting coils however lead to severe operational constraints, imposing mainly to maintain a homogenous and very low temperature inside the whole volume of each of the 4 ITER coil sets. Taking into account on the one hand, that the coils are the seat of strong electromagnetic energy transfers which may endanger their structure, and on the other hand, that they are necessarily very close to the plasma chamber, their implementation in the nuclear environment of ITER has imposed a drastic level of reliability and safety. The design and the safety assessment of the ITER magnets result of a systematic and thorough approach. Potential initiating faults of every main component (conductor, coil, structure, electrical and cryogenic services, instrumentation and control) have been identified and a phenomenological analysis has allowed to define a set of conservative sequences. The analysis of these sequences shows a limited environmental potential impact, far below the safety guidelines.


Corresponding Author:

Raboin Serge

Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France

- E - Magnets and Power Supplies.

P2T-E-534 WINDING MACHINES FOR THE MANUFACTURING OF SUPERCONDUCTIVE COILS OF THE MAIN EUROPEAN FUSION RESEARCH MACHINES

Cazzaniga Rodolfo, R. Cazzaniga (1) R. Penco (2) C.E. D’urso (2)

(1) TPA s.r.l. via Ettore Paladini 16, 23891 Barzanò (LC), Italy (2) ANSALDO Superconduttori, Corso Perrone 73r, 16152 Genova, Italy

The successful completion of large magnet projects passes through the development and application of non-conventional manufacturing processes. The conductor winding process is a difficult and delicate step for the manufacturing of superconducting coils. It is one of the most challenging and advanced and has to be especially tailored for each project. The first task required for the winding process is the carrying out of unusual, large, mostly three-dimensional coils. The second is to maintain as far as possible the section of the cable unaltered in order not to damage the strands and to maintain the design features of the cable. The third is to assure the suitable repetitiveness and speed for an industrial process. The manufacturing solution is a system of different machines linked and tuned together and specially designed for each coil. Each machine must be previously tested. A tailored software assures the overall process control. TPA provided ANSALDO Superconduttori with the winding systems for all the major European projects: TFMC (of NET), CMS (of CERN), WENDELSTEIN (of Max Planck IPP). The considerable experience gained in this field by both companies, TPA and ANSALDO Superconduttori, has been just acknowledged by the CERN with “The CMS gold Award of the Year 2004”. The paper reports the progress of the winding technology. It describes the main features of the winding machines, the main problems, the layouts of the systems used in the above-mentioned projects and the new ideas for the forthcoming ones


Corresponding Author:

Cazzaniga Rodolfo

TPA BRIANZA Via Paladini, 16 23891 Barzanò (Lecco) Italy

- E - Magnets and Power Supplies.

P2T-E-539 A SUCCESS STORY: LHC CABLE PRODUCTION AT ALSTOM MSA

Grunblatt Gerard, P. Mocaer (1) C. Kohler (1)

(1) ALSTOM 3 Av des trois chenes 90000 Belfort France

ITER ,when constructed ,will be the equipment using the largest amount of superconductor strand ever built (Nb3Sn and NbTi). ALSTOM Magnets and Superconductors SA, “MSA” received in 1998 the largest orders to date for delivery of superconductor strands and cables (3100 km of cables and various strands) for the Large Hadron Collider being built at CERN. These orders to MSA correspond to more than 600 metric tons of superconducting strand ,amount to be compared to around 600 metric tons of Nb3Sn strands and 250 metric tons of NbTi strands necessary for ITER. Starting from small and short R&D programs in the early nineties, MSA has reached its industrial targets and has as of March 2004 delivered more than 60% of the whole orders with products meeting high quality standards. Production is going on at contractual delivery rate and with satisfactory financial results to finish deliveries around mid 2005 We will explain how we succeed to transform a “cottage industry” (25 people in 1997) to a very high “world class” production activity (170 people in 2004). Main industrial problems now solved such as investments and industrial set up and ramp up to reach plateau production as well as more technical problems closely linked to industrial ones such as multifilament wire breaks , strand magnetization, coating process (0.15µm controlled ) and others will be addressed and the various methods used to solve such problems will be reviewed.


Corresponding Author:

Grunblatt Gerard

ALSTOM 3 Av des trois chenes 90000 Belfort France

- F - Plasma Facing Components.

P4C-F-8 TILES CHAMFERING AND POWER HANDLING OF THE MK II HD DIVERTOR

SALAVY Jean-François, P. Chappuis (1) P.J Lomas (2) V. Riccardo (2)

(1) CEA Cadarache, Direction des Science de la Matière, F-13108 St Paul Lez Durance, France (2) UKAEA, Culham Science Centre, Abingdon, OXON, OX14 3DB, United Kingdom

The JET HD (High Delta) Divertor is an upgrade of the actual JET divertor consisting of two modified toroidal segments which are: a new Load Bearing Septum Replacement Plate (LB-SRP) tile located at the septum position, and a High Field Gap Closure (HFGC) tile protecting inboard diagnostic cabling. The aim of the upgrade is to allow high power operation and a wider range of plasma lower triangularities. This paper describes the optimisation of the tile chamfering for LB-SRP and HFGC (including edge shadowing) and of the power handling evaluation for a set of planned plasma configurations. The LB-SRP and HFGC tiles have a slope in the toroidal direction to hide any edge of the next tile from the impinging plasma. They are machined from blanks of Carbon Fibre Composites materials and are attached to the carrier through the JET usual system of dumbbell, tie rod and disc springs. The precise design of the tile faces is based on 12 plasma configurations given by the JET team, and on two sets of mechanical tile tolerances, issued by the JET drawing office. The PROTEUS code (magnetic equilibrium by finite element) is used to calculate the various field lines angles, which are inputs for the chamfering angle calculation process. The design of the LB-SRP tile has been optimised to increase the power handling (roughly from 1 to 80MJ) for the different plasma configurations and for the various tile tolerances (chamfering angle varies linearly along the tile profile). After calculating the chamfering angle values of each face, a checking exercise has been realised on the 3D CATIA models of the tiles by putting them at their extreme tolerance positions and validation if the shadowing is insured for a angle calculated to take into account the worst possibilities. With the final chamfering angle value for each face, the power handling of the tiles has been estimated with final elements calculations. Power handling are given either with the critical time to reach 1800 C at the tile surface for a total injected power of 40 MW, or with the maximum total injected power allowable for a 10 seconds run without reaching 1800 C. Chamfering angles for the new LB-SRP and HFGC tiles of the MK II HD divertor have been calculated, optimised and checked in order to insure a good shadowing of the edges for each of the 12 reference plasma configurations. The consequent power handling has been estimated and gives promising results in regards to the JET EP project objectives.


Corresponding Author:

SALAVY Jean-François

CEA Saclay, DEN/DM2S/SEMT/BCCR, F-91191 Gif sur Yvette, France

- F - Plasma Facing Components.

P4C-F-12 THERMAL AND MECHANICAL ANALYSIS OF THE EAST PLASMA FACING COMPONENTS

SONG,Yuntao, D.M Yao, S.T Wu, P.D Weng

Institute of Plasma Physics, Chinese Academy of Sciences P.O.Box 1126, Anhui, Hefei, P.R.China, 230031

The EAST (Experimental Advanced Superconducting Tokamak) is an advanced steady-state plasma physics experimental device to be built in P.R.China. It has a long pulse (~1000s) capability and will be able to accommodate divertor heat loads that make it an attractive test for the development of advanced tokamak operating modes. Now, the engineering designs for the EAST plasma-facing components (PFC) are in progress. The EAST PFCs consist of a plasma-facing surface affixed to a cooled support plate. All the PFCs are made of copper alloys (CuCrZr) on which Carbon-carbon (C-C) composites tiles are mechanically attached. A thin piece 0.3mm of flexible graphite sheet is used between the tile and the heat sink to improve the contact conductance. This paper will introduce the thermal and mechanical analysis, which included the structure intensity and optimization hydraulic parameters by the finite element method for the different type of cooling and mounting structure between the heat sink and C-C tiles using for EAST plasma facing components. Since EAST is a long pulse machine, all of the analysis was done for steady-state conditions. To reduce the radiation enhanced sublimation, the maximum temperature are controlled to be less than 1000Ž. In order to reduce the impurity sources to plasma, the minimum wall temperature has to be more than 100Ž.The maximum heat flux 7MW/m2 on these components is chosen based on the simulation of the plasma operation. In this study three types of PFC mounting structures are considered: 1) the C-C tiles bolted to the water-cooled copper plates through the thin piece of flexible graphite sheet. The copper alloys are actively cooled by water flowing the cooling channel drilled by a special punch. 2) C-C tiles bolted to the copper plates through the thin piece of flexible graphite sheet and welding a water-cooling tube on the other side of copper plates. 3) C-C tiles bolted to the copper plates, which brazing on a water-cooled and supporting stainless steel structure. Based on the optimization analysis results the type 1 is chosen as the mounting structure between the tiles and heat sinks for the PFCs of EAST device, the maximum water-cooling velocity chose as 7m/s. Under these conditions the maximum thermal stresses in copper plate is 220MPa,which is less than the allowable stress based on the design criteria ASME code. The maximum temperature of C-C tiles is 829Ž, which also have been proved by a prototype test.


Corresponding Author:

SONG,Yuntao

P.O.Box 1126, Anhui, Hefei, P.R.China, 230031,Institute of Plasma Physics, Chinese Academy of Sciences

- F - Plasma Facing Components.

P4C-F-13 THE DYNAMIC ERGODIC DIVERTOR IN TEXTOR – A NOVEL TOOL FOR STUDYING MAGNETIC PERTURBATION FIELD EFFECTS

O. Neubauer, B. Giesen, P.W. Hüttemann, H.T. Lambertz and the TEXTOR Team

Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, Association EURATOM/FZJ, Trilateral Euregio Cluster, D-52425 Jülich, Germany

Recently TEXTOR has been upgraded by installation of the Dynamic Ergodic Divertor (DED). The purpose of the DED is to influence transport parameters in plasma edge and core and to study the resulting effects on heat exhaust, edge cooling, impurity screening, plasma confinement and stability. Alternatively, the DED creates static or rotating multipolar helical magnetic perturbation fields of different mode patterns. A set of 16 helical coils has been installed on the inboard high-field side of the vacuum vessel covering about one third of the plasma surface. Thus, in contrast to similar experiments, the DED produces a clear mode spectrum with mode numbers of m/n = 12/4, 6/2, 3/1 or a superposition of those. For the first time rotating fields of up to 10 kHz can be generated. The peak coil current is 15 kA for pulse duration of up to 10 s. A novel coil design has been developed which fulfils the various mechanical, electrical, high frequency, thermal, and vacuum requirements. The coils consist of twisted copper wires in a corrugated stainless steel tube. A combined Helium / water cooling system removes the energy of adiabatic heating of the coils during a pulse. Coaxial vacuum feedthroughs have been designed which allow for compensated current feeding as well as supply of the cooling media. During a major shut down the DED components have been integrated. For this purpose, after removing diagnostics, TEXTOR has been split into two parts; the liner has been removed, modified and reintegrated, followed by the mounting of DED components. In parallel the power supply system has been fabricated, installed and tested on a full size dummy load. Particularly the AC operation has been realised by modular IGBT series resonant inverters. A sophisticated control system guarantees sufficient stability of amplitude, frequency and phase of the coil currents. In addition to the various technical aspects of the DED design, implementation and commissioning, highlights of recent experiments will be presented. In particular the impact of the perturbation field on MHD stability and plasma rotation will be addressed.


Corresponding Author:

O. Neubauer

Forschungszentrum Jülich GmbH, IPP, 52425 Jülich, Germany

- F - Plasma Facing Components.

P4C-F-24 EAST(HT-7U) IN-VESSEL COMPONENTS DESIGN

Yao Damao, J.G.Li; Y.T.Song; S.J.Du; J.L.Chen; P.D.Weng

P.O.Box1126 Hefei Anhui 230031, P.R.China

In vessel components are very important parts of EAST (HT-7U) superconducting tokamak. The primary purpose of these components is to protect the vacuum vessel, RF system and diagnostic components from the plasma particles and heat loads. Other function of in vessel components is additional to particles and heat loads management. The divertor is designed to provide particles exhaust into the divertor cryopump, provide recycling control, and impurity control. The passive plates stabilize the plasma vertical stability. Heat loads and electric-magnetic forces are quite complicate for in-vessel components. Considering the possibility of plasma operation condition for EAST different operation phases, the initial operation of EAST is focus on physics scenarios, plasma elongation experiment explore and plasma control experiment explore etc. The long pulses and steady state operation is planed in the later time of first phase. During the first stage plasma heating power will be reached 10MW, and maximum heat flux will not permit more than 1MW/m2 for steady state operation. Brazed tile are not employed for divertor plate. All boron doped graphite tiles with SiC coating are bolted to active water-cooling CuCrZr heat sink. The bolted structure is also used for inner toroidal limiter and passive stabilizer to handling heating power up to 10MW when the plasma is operated in circle cross-section. EAST divertor layout is designed as up-down symmetry to accommodate both double null and single null plasma configuration. It is provide a large experimental flexibility and capable of running in a stand scenario, with power conducted along the field lines to the target plates, or in a radiative divertor mode. The geometry of the divertor is based on simulations obtained using the EFIT code and references the experiences of physic design, engineering design and experiment of other tokamak. The vertical target is inclined so as to intercept the magnetic field lines of the separatrix at an acute angle. A “V” shape was formed and expect particles remain in this region to aids heat load uniform distribute on divertor plate so as to reduce the peak heat flux on divertor targets. Passive stabilizer is a single turn saddle coil with active water cooling and a single vertical current bridge. Consider the flexibility of plasma control several copper saddle coils will be used for plasma equilibrium control, error field corraction and RWM control.


Corresponding Author:

Yao Damao

P.O.Box1126 Hefei Anhui 230031, P.R.China

- F - Plasma Facing Components.

P4C-F-27 HOT RADIAL PRESSING: AN ALTERNATIVE TECHNIQUE FOR THE MANUFACTURING OF PLASMA-FACING COMPONENTS

Visca Eliseo, S. Libera (1), G. Mazzone(1), A. Pizzuto(1) C. Testani (2)

(1)Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, IT-00044 Frascati (2)CSM S.p.A., IT-00128 Castel Romano, Roma

The Hot Radial Pressing (HRP) manufacturing technique is based on the radial diffusion bonding principle performed between the cooling tube and the armour tile. The bonding is achieved by pressurizing the cooling tube while the joining interface is kept at the vacuum and temperature conditions. This technique has been used for the manufacturing of relevant mock-ups of the ITER divertor vertical target. Tungsten monoblock mock-ups were successfully tested to high heat flux thermal fatigue (18 MW/m2 for 1000 cycles). After these good results the activity is now focused on the developing of a canister suitable for the CFC monoblock mock-ups. The stainless steel canister reutilization is also one of the main objectives. A FE calculation was performed to investigate the stress involved in the CFC tiles during the process and to avoid the CFC fracture. The influence of the process parameters was also investigated in order to keep the process bonding temperature as low as possible to preserve the copper alloy thermo-mechanical properties. The design improvement of the canister and the equipment will be reported in the paper as well as the results obtained by FE calculation and by the HRP manufacturing of the monoblock mock-ups.


Corresponding Author:

Visca Eliseo

Associazione EURATOM-ENEA sulla Fusione - Via E. Fermi, 45 - 00044 Frascati RM

- F - Plasma Facing Components.

P4C-F-28 HETS PERFORMANCES IN HE COOLED POWER PLANT DIVERTOR

Aldo Pizzuto, Panos Karditsas (+), Claudio Nardi (*), Stamos Papastergiou (x)

(*) ENEA – CR Frascati – Via E. Fermi 45 – I-00044 Frascati (Roma) Italy (+) UKAEA – Culham Science Centre, Abingdon, Oxfordshire – OX14 3DB, UK (x) EURATOM at ENEA Frascati

In the frame of the activities aimed to evacuate the performances of a He cooled divertor in the future fusion power plant, the High Efficiency Thermal Shield (HETS) concept has been proposed. This concept relays on an abrupt change of momentum of the fluid in order to increase the turbulence in the gas, and therefore the heat transfer. This concept, initially developed for water, has been extended in the past years to He, and studies have been performed in order to evaluate his suitability in such environment. The requirements for the power plant divertor are to sustain a thermal flux of at least 10 MW/m2, without exceeding limits in stress and temperature. A further limitation is given by the required pumping power, because the attractiveness of the power plant is strictly related to the power output of the plant itself. As the PPS divertor must operate with an energy production plant, the He must have such a characteristics to be used directly in the energy production cycle (gas turbine), therefore the reference values are of 10 MPa He pressure and an inlet temperature of 600 C and outlet temperature 800 C (temperature rise of 200 C, this value could be re-evaluated in order to optimize the characteristics. Analytical studies developed in ENEA and UKAEA showed that the HETS concept can sustain a thermal flux of 10 MW/m2, keeping low pressure drops (and therefore pumping power). A thermal flux as high as 15 MW/m2 can be easily sustained, increasing the pumping power up to the proposed limit (10% of the thermal power to the divertor). Because of the relevance of pressure drop in the structure for the performances of the divertor, experimental validations are required for the HETS, as in literature reliable assessments for this characteristic can not be found. The preliminary experiments have been performed using air at room temperature and high pressure, showing values of pressure drop in line with the parameters used in the calculations. Studies aimed to optimize the shape of the channel, in order to further reduce the pressure drop are in progress. At the present stage the HETS appear to be a promising solution for the He cooled power plant divertor.


Corresponding Author:

Aldo Pizzuto

ENEA - Via E. Fermi 47 - 00044 Frascati (Roma)

- F - Plasma Facing Components.

P4C-F-32 THE INFLUENCE OF IRRADIATION REGIMES ON RETENTION HYDROGEN ISOTOPES IN STRUCTURAL MATERIALS

Zaluzhnyi Alexander Georgievich, Kopytin Vladimir Platonovich Suvorov Alexander Leonidovich

B. Cheryomushkinskaya 25, Moscow 117259, Russia

Researching of the hydrogen isotopes retention in candidate materials, which connects to possible loosing of valuable fuel (tritium) and service safety of fusion reactor, becomes very important because of elaboration of the first-wall materials for fusion reactor. In the present work was investigated the influence of irradiation regimes on retention hydrogen isotopes in samples of austenitic steel during heating. The samples of studied materials were irradiated both in the reactor and by hydrogen isotopes ions of different energies and fluencies bombardment in an accelerator. Kinetic of hydrogen release from the samples worked with deuterium plasma was investigated. The following results were obtained. Heating the irradiated samples of steel (irradiated in the reactor or by hydrogen isotopes ions bombardment), which have been kept in normal temperature during quite a long period after the irradiation, a shift of the diffusion peak of hydrogen release to higher temperatures, comparing to no irradiated samples, was observed. It means that atoms of hydrogen in the irradiated sample were caught by radiation defects, which are very effective as traps for hydrogen atoms till quite high temperatures (700 K). The worked out analysis of the received results supposes that vacancy complexes. On thermodesorption curves of hydrogen release from irradiated samples of austenitic steels a high temperature peak (900-1000 K) was observed because of dissociation of hydrogen containing compounds in micro pores. During investigations of hydrogen release from irradiated samples of austenitic steel, after it had been saturated with hydrogen plasma, abnormally big blisters were registered with cover thickness of about 1mkm. Three peaks were observed on the thermodesorption curves of hydrogen release from irradiated samples, contained blisters. The low temperature spike (~500 K) was showed to correspond to hydrogen release because of its resolution from blisters, where it was in molecular form. The high temperature peak (?900 ?) corresponds to hydrogen release from dissociating blisters, which contain hydrocarbons. The mechanism of abnormal blisters generation is offered. Inasmuch methane is not soluble in metals in temperatures lower then temperature of its dissociation, it behaves as a noble gas, when heated to temperatures lower then temperature of its dissociation.


Corresponding Author:

Zaluzhnyi Alexander Georgievich

Kashirskoe Shosse 31, Moscow 115409, Russia

- F - Plasma Facing Components.

P4C-F-33 MANUFACTURING TECHNOLOGY DEVELOPMENT FOR THE VACUUM VESSEL AND PLASMAFACING COMPONENTS

Laitinen Arttu, Jari Liimatainen * Mika Korhonen * Pentti Hallila * Seppo Tähtinen **

* Same address as for the corresponding author ** VTT Industrial Systems, P.O.Box 1704, FIN 02044 VTT Finland

Vacuum vessel and plasma facing components of the Iter construction including shield modules and primary first wall panels have great impact on the production costs and reliability of the installation. From the manufacturing technology point of view, accuracy of shape, properties of the various austenitic stainless steel/austenitic stainless steel interfaces or CuCrZr/ austenitic stainless steel interfaces as well as stress corrosion and fatigue properties of the base materials are crucial for technical reliability of the construction. The current approach in plasma facing components has been utilization of solid-HIP technology and solid-powder-HIP technology. Due to the large size of especially shield modules shape control of the internal cavities and cooling channels is extremely demanding. This requires strict control of the raw materials, especially powder size distribution, encapsulation technology and hot isostatic pressing pressing cycle forms. For the vacuum vessel components there are several optional manufacturing routes including narrow gap and beam welding (laser or electron beam) of the bended thick walled plates and forgings, or optionally use of forgings and thich walled plates together with hot isostatically pressed components in order to reduce total amount of welding and to be able to position welds in a way that minimizes distortion risks during assembly welding. In this presentation, different manufacturing methods area compared and their relative attractiveness is evaluated and discussed.


Corresponding Author:

Laitinen Arttu

Rieväkatu 2, P.O.Box 1100, FIN-33541 Tampere Finland

- F - Plasma Facing Components.

P4C-F-41 ENGINEERING AND THERMAL-HYDRAULIC DESIGN OF PFC COOLING FOR SST-1 TOKAMAK

Chaudhuri Paritosh, P. Santra, N. Rabi Prakash, S. Khirwadkar, G. Ramash, S. Dubey, A. Arun Prakash, D. Chenna Reddy, and Y. C. Saxena

Institute for Plasma Research, Bhat, Gandhinagar, Gujarat, INDIA

Steady state Superconducting Tokamak (SST-1) is a medium size tokamak with superconducting magnetic field coils. It is a large aspect ratio tokamak with a major radius of 1.1 m and minor radius of 0.20 m. SST-1 is designed for plasma discharge duration of ~1000 seconds to obtain fully steady state plasma with total input power upto 1.0 MW. PFC is one or the major sub-systems of SST-1 tokamak consisting of divertors, passive stabilizers, baffles, and poloidal limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC is the steady state heat removal of upto 1MW/m2. In addition to remove high heat fluxes the PFC are also designed to be compatible for high temperature baking at 350 C. Extensive studies, involving different flow parameters and various cooling layouts have been examined to select the final cooling parameters and layout, compatible for cooling and baking. Design considerations included 2-D steady state and transient thermal analysis of PFC during plasma operation. Thermal analysis is carried out with the purpose of evaluating the thermo-mechanical behavior of the PFC. Both 1-D analytical and 2-D Finite Element (FE) analysis were carried out to determine the temperature distribution and the thermally induced stresses in PFC. The results of the calculation led to a good understanding of the temperature and thermal stress distribution in various parts of the PFC. Since the tiles are mechanically attached to the back plate, the fitting technique must provide the highest mechanical stress so that thermal transfer efficiency is maximized. Proper brazing of cooling tube on the copper back plate is necessary for the efficient heat transfer from the tube to the back plate. The contact at the brazed joint of the tube to the backplate is critical for the above application. Using an infra-red-camera, spatial and temporal evaluation of the temperature profile has been studied under various flow parameters to evaluate the quality of the brazed joint of the manufactured modules. The optimized thermal-hydraulic design and the effect of stress and strain on different material used in PFC were examined and discussed in this paper.


Corresponding Author:

Chaudhuri Paritosh

Institute for Plasma Research, Bhat, Gandhinagar, Gujarat, INDIA

- F - Plasma Facing Components.

P4C-F-49 THE USE OF COPPER ALLOY CUCRZR AS A STRUCTURAL MATERIAL FOR ACTIVELY COOLED PLASMA FACING AND IN VESSEL COMPONENTS

Lipa Manfred (1), A. Durocher (1), R. Tivey (2), Th. Huber (3), B. Schedler (3), J. Weigert (4)

(1) CEA/Cadarache, DRFC, F-13108 SAINT PAUL LEZ DURANCE, France (2) ITER Garching Joint Work Site, D-85748 Garching, Germany (3) Metallwerk Plansee GmbH, A-6600 REUTTE, Austria (4) PTR Präzisionstechnik GmbH, D-63461 Maintal-Doernigheim, Germany

Within the European fusion community there is a wide experience of the use of copper chromium zirconium (CuCrZr). Precipitation hardened CuCrZr has been used as structural material for actively cooled plasma facing components for 15 years in Tore Supra (TS), and more than 20 years in JET. In the future CuCrZr will be used for the actively cooled divertor plates of the W7X stellarator and has been selected as the heat sink material for the ITER divertor. In all cases TS, JET, W7X and ITER the components are cooled using pressurised hot water and as they operate in ultra high vacuum have to remain leak tight in operation, during which they are exposed to cyclic thermal loads, and in the case of TS and ITER large electro dynamic forces, and specifically for ITER neutron irradiation. The CuCrZr pre-material for TS and JET component fabrication has been procured from several suppliers and delivered in product shapes such as: drawn rods, bars, hollow-profiles and in forged plates or rolled sheets. In the case of TS the material has always been delivered in the solution annealed, quenched and age hardened state, whereas JET prefers to use solution annealed material only performing hardening operations towards the end of the fabrication process. This paper presents our experience with this material in the procurement and fabrication of actively cooled components for TS (pump limiter fingers, guard limiter heat sink hollow-profiles, ripple protection tubes, endoscope heads), JET (accelerations grids and beam scrapers for the beam-lines) and divertor prototypes tested for ITER. By highlighting failed manufacturing processes it is hoped that future users can avoid repetition of costly and time-consuming failures that might occur both during manufacture and during subsequent operation. To this end the technical specification for procurement of CuCrZr is discussed including the supposed influence of chemical impurities, alloying elements, material production process and heat treatments on material properties. The general properties of different material grades, procured from different suppliers for various component applications, are also given. The associated mechanical characterisation of component joining, focused on fusion welding using electron beams both of CuCrZr to SS via a nickel adaptor and CuCrZr to CuCrZr, is presented. The behaviour of actively cooled CuCrZr components during normal and especially during accidental operation conditions is described.


Corresponding Author:

Lipa Manfred (1)

CEA Cadarache, DRFC, 13108 Saint Paul lez Durance, France

- F - Plasma Facing Components.

P4C-F-54 MANUFACTURING OF THE W7-X DIVERTOR AND WALL PROTECTION

Streibl Bernhard, J. Boscary, P. Grigull 2), H. Greuner, J. Kißlinger, C. Li, B. Mendelevitch, T. Pirsch, N. Rust, S. Schweizer, M. Weißgerber

Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, D-85748 Garching 2) Max-Planck-Institut für Plasmaphysik, Euratom Association, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald

The W7-X stellarator is designed for steady state operation with an input power of 10 MW and transient discharges up to 20 MW for 10 seconds. In the first step an 'open divertor' will uncouple the plasma core from the wall. According to the structure of the plasma boundary 10 water-cooled divertor units, one lower and one upper per field period, are arranged such that leading edges are avoided. Each divertor unit, composed of horizontal and vertical modules adjacent to the pumping slot, is 5 m long and spans over 70% of the field period. Baffle modules increase the neutral density in front of the target plates. To improve the particle control a closed divertor chamber is formed by toroidal and poloidal end plates and two cryo-pump units behind the horizontal target increase the pumping speed. Behind the vertical baffle a control coil is arranged for compensating symmetry breaking error fields and sweeping the strike points. The remaining poloidal and toroidal area of the plasma vessel is covered by a water-cooled wall protection. On the outboard, where the lowest heat flux is expected, steel panels are applied. They protect also the access ports over the length of their typical diameter. On the inboard side of the plasma vessel heat sinks armoured with clamped graphite tiles are arranged. The same design is applied for the baffle modules and two of the 9 horizontal target modules with a considerably reduced heat load of 1 MW/m2. Only the width ratio of heat sinks and tiles is adjusted appropriately to the heat flow. The remaining target modules are composed of 6 to 12 elements, equipped with CFC tiles of Sepcarb® NB31 to take up 10 MW/m2. In total 940 elements of this kind, including 50 spares, are required and will be manufactured by the company Plansee. Profit will be taken from the CIEL experience via the collaboration with CEA. The company MAN-DWE will manufacture the steel panels. Also an external company produces the 10 control coils. By IPP itself, the modules with clamped tiles and the cryo-pumps will be manufactured and qualified with pre-series tests. In addition IPP will perform all acceptance tests: the high heat flux tests for the target elements with an ion beam facility, the vacuum and out gassing tests in a large vacuum oven, the water flow and pressure tests and the final tests of the control coils at nominal electrical current. Special technological aspects will be qualified in close collaboration between the manufacturers and IPP.


Corresponding Author:

Streibl Bernhard

Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, D-85748 Garching

- F - Plasma Facing Components.

P4C-F-59 STUDIES ON GRAPHITE SURFACES DETRITIATION BY PULSED REPETITION RATE NANOSECOND LASERS

SEMEROK Alexandre, WEULERSSE J.-M., BRYGO F., LASCOUTOUNA CH., HUBERT CH. , LE GUERN F., *TABARANT M.

CEA Saclay, DPC/SCP/LILM, 91191 Gif sur Yvette CEDEX, France *CEA Saclay, DPC/SCP/LRSI, 91191 Gif sur Yvette CEDEX, France

The detritiation of plasma-facing components is regarded as one of the crucial problems in a future thermo-fusion reactor design and construction. A fast heating of an exposed surface by a focused laser beam allows to obtain the temperature higher than 1000K on a thin (1 – 100 µm) near-surface layer, thus, resulting in detritiation either by hydrogen desorption from the surface or by ablation of this near-surface layer. Thus, visible and near-IR lasers with 100-500W average power that can be transmitted by optical fibers may provide a completely automatic unattended system for reactor vacuum chamber surfaces in situ cleaning. Low and high repetition rate nanosecond laser benches provided with a sufficiently complete control and measurement equipment were developed and applied for studies on graphite and co-deposited layer heating and ablation. Heating and ablation regimes of detritiation (dehydrogenization) with pulsed lasers were distinguished by ablation threshold fluency that was determined experimentally for graphite samples with D/H isotopes from TexTor and TORE SUPRA (CEA Cadarache). Ablation threshold fluencies were Fth = 0.4 ± 0.1 J cm-2 for a co-deposited layer and Fth = 1 J cm-2 for graphite surfaces.For graphite samples from TORE SUPRA and TexTor, the ablation efficiencies were determined to be different: (0.025 µm/J cm-2) for graphite and (0.2 µm/J cm-2) for co-deposited layer. The particular features of the graphite and co-deposited layer ablation (different ablation thresholds and laser ablation efficiencies) are discussed with relation to the procedures that could ensure self-controlled laser surface cleaning. As the result of our investigations, the conclusion was made that detritiation rate of 1 m2 per hour can be obtained for 20 µm co-deposited layer with high repetition rate Nd-YAG laser beam of 250 W mean power. In this case, the laser fluency should be 1 J/cm2 to provide the maximum ablation efficiency of 0.2 µm/Jcm-2. At the same time, we concluded that detritiation of a thick co-deposited layer by laser heating is much more efficient with a continuous wave Nd-YAG laser radiation. Experimental and theoretical studies on laser heating and ablation with different types of co-deposited layers and Nd-YAG lasers (cw and pulsed) in controllable environmental conditions (gas composition, pressure) are in progress and will be presented in the paper.


Corresponding Author:

SEMEROK Alexandre

CEA Saclay, DPC/SCP/LILM, B.467, 91191 Gif sur Yvette CEDEX France

- F - Plasma Facing Components.

P4C-F-66 STEADY STATE AND TRANSIENT THERMAL-HYDRAULIC ANALYSES ON ITER DIVERTOR MODULE

Dell'Orco Giovanni *, Ancona Antonella**, Di Maio Pietro Alessandro**, Merola Mario***, Vella Giuseppe**,

* ENEA, P.O. Box 1, 40032 Camugnano (Bo) Italy, ** Università di Palermo - DIN, V.le delle Scienze, 90128 Palermo Italy, *** EFDA CSU Garching Boltzmannstr. 2, D-85748 Garching-Germany

The Divertor is one of the most challenging components of the next step ITER nuclear fusion reactor. It is aimed at reducing the impurities in the plasma and at sustaining the heat and particle fluxes during normal and transient operations as well as during disruption events. The ITER Divertor consists of 54 cassettes and three plasma-facing components (PFCs), namely the inner vertical target, the outer vertical target and the dome liner. The water maximum total flow rate should be 1000 kg/s, with 100-150 C inlet/outlet temperatures, 4.2 MPa inlet pressure and a maximum pressure drop of 1.4 MPa. The PFCs are cooled in series, with a maximum water velocity in the channel of 11 m/s, whilst the water coolant is routed via the cassette body. Each PFC consists in a number of plasma facing units, cooled in parallel and assembled onto a supporting structure. Due to the extremely high heat loads expected onto the PFCs (up to 20 MW/m2 over 20 s), the hydraulic design of the DIVERTOR is particularly demanding. It shall ensure that the foreseen flow rate actually reaches each plasma-facing unit to ensure an adequate cooling and to prevent any risk of Critical Heat Flux (CHF). Sufficient margin ( > 40 %) to avoid the reaching of a CHR limit on the PFCs could be obtained by using hypervapotron design inside the flat channels and swirl flow turbulence tape promoters inside the cooling tubes. Furthermore the overall pressure drop and flow rate shall be within the specified design limit to avoid an unduly high pumping power. Another important issue is the definition of a proper procedure to drain the coolant and dry the Divertor components prior to the maintenance operations as well as to refill them with water after maintenance ensuring a complete elimination of gas bubbles. Due to the complex flow scheme of the hydraulic circuit, a pure theoretical study does not appears sufficient to address all the above-mentioned items and an experimental validation of the models is mandatory. In addition to that, the assembly of the PFCs onto the cassette body as well as their integration by welding the coolant connections of the PFCs, also represent a critical step to be investigated. The paper presents both the steady sate and transient theoretical thermal hydraulic analyses, carried out by RELAP code, on the Divertor module for the: i) flow distribution, pressure drop and Critical Heat Flux margin; ii) draining and drying of the Divertor components.


Corresponding Author:

Dell'Orco Giovanni *

ENEA, P.O. Box 1, 40032 Camugnano (Bo) Italy

- F - Plasma Facing Components.

P4C-F-69 APPLIED TECHNOLOGIES AND INSPECTIONS FOR THE W7-X PRE-SERIES TARGET ELEMENTS

BOSCARY Jean, H. Greuner K. Scheiber(1) B. Streibl B. Schedler (1) B. Mendelevitch J. Schlosser (2)

(1) Plansee Aktiengesellschaft, Technology Center, A-6600 Reutte, Austria (2) CEA Cadarache, Euratom Association, F-13108 St Paul-lez-Durance, France

The WENDELSTEIN 7-X (W7-X) divertor is designed to remove 10 MW during steady state operation. 23 m2 of the target area are approximated by 890 water-cooled target elements of 14 various types. The water-cooling characteristics are optimized to sustain 10 MW/m2 and to remove 100 kW maximum per element. Results of finite element calculations show the maximum allowable transient up to steady state heat load for standard and diagnostic elements. The manufacturing process uses well-known technologies but applied to the particular W7-X geometry. All plasma facing target elements are covered with CFC Sepcarb® NB31 flat tiles (standard tile). They are bonded to Cu by Active Metal Casting (AMC®). The heat sink is made of CuCrZr. All these processes will be qualified during the pre-series phase. The characterization of the first delivered CFC batch of 150 kg shows that the thermal and mechanical properties of blocks are compatible with the AMC® process. More than 2/3 of the elements are shielded by a L-shaped front tile against 1 MW/m2. This solution allows the 3D fitting along the divertor pumping space. 30 diagnostic elements are equipped with lateral tiles specified for 3 MW/m2. These particular tiles require to try two technologies for the bonding of the AMC®-NB31 tile to the heat sink, namely electron beam (EB) welding and hot isostatic pressing (HIP). On one hand, HIP guarantees a better thermal contact of the tile to the heat sink and avoids the access difficulty of EB-welding for lateral tiles. On the other hand, HIP has never been applied to such a large production. The selection between the two technologies is one of the issues of the pre-series phase. The four pre-series elements and relevant specimens manufactured with this aim will be described. Non-destructive examinations of the bond between the tile and the heat sink are integrated throughout manufacturing and applied at a very early stage of the fabrication. AMC®-NB31 standard tiles are examined by X-ray and lock-in thermography. The bond of these tiles to the heat sink block is checked by ultrasonic and lock-in. Lock-in is finally applied for the completed target element, equipped with cooling channels. The crucial issue of the pre-series phase is the definition of the acceptance criteria applied to the series production.


Corresponding Author:

BOSCARY Jean

Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, D-85748 Garching bei München, Germany

- F - Plasma Facing Components.

P4C-F-76 OVERVIEW OF THE ENGINEERING DESIGN OF THE ITER DIVERTOR

Tivey Richard, V. Chuyanov, E. D’Agata, G Federici, H. Heidl, A. Makhankov, M. Merola, J. Palmer

ITER Joint Work Site, Boltzmannstr.2, 85748 Garching, Germany D.V. Efremov Research Institute, St. Petersburg, Russia; EFDA Close Support Unit, Boltzmannstr.2, 85748 Garching, Germany

There have been significant developments in the design and R&D of the ITER divertor since it was last reported in the ITER Final Design Report 2001 (FDR 2001). The main developments will be presented and these are outlined below. The construction and testing of prototypical components has demonstrated the capability of handling, with sufficient margin, the predicted steady state heat flux on carbon-fibre composite (CFC) and tungsten-armoured surfaces. There has been feedback from European industry on the manufacturing difficulties associated with building of these large-scale pieces. As a direct consequence, a number of design options are under consideration aimed at simplifying manufacture and/or reducing costs. These include the option to keep as separate pieces the CFC- and tungsten-armoured sub-components until late in the manufacturing cycle. Furthermore, simplifications of the divertor geometry are presented that avoid the need to manufacture many special cassettes, for example that might be required to accommodate the wide range of diagnostics incorporated into the divertor. Intermediate ducts have been introduced into the design to bridge the gap between the four pumping ports and the divertor cassettes. These ducts, in combination with gas seals between the sidewalls of the cassettes, localize the high pressure (1-10Pa), hydrocarbon-rich exhaust gas preventing it from entering other divertor level ports and the area behind the divertor. Apart from helping to localize potential C-H deposits, the design reduces the overall burden of gas to be removed during the dwell time, inhibits the exhaust gas from re-entering the main chamber during a pulse, and avoids the need to develop elaborate gas seals around the diagnostics and viewing systems that are integrated into other ports at the divertor level. In response to the feedback from remote maintenance specialists on the difficulties foreseen in in-vessel handling of the FDR 2001 design of divertor cassettes, a cassette to vessel attachment scheme aimed at simplifying the maintenance operations is proposed. Instead of clamping the cassettes to rails attached to the vessel walls, the cassettes are maintained in position by a spring on the cassette pressing features on the cassette into recesses in the vessel wall. The paper introduces the design, discusses the implications for remote installation and build tolerances, and outlines the planned programme of work aimed at validating it.


Corresponding Author:

Tivey Richard

ITER Garching JWS, Boltzmannstr. 2, D-85748 Garching, germany

- F - Plasma Facing Components.

P4C-F-92 TOWARDS THE DEVELOPMENT OF WORKABLE ACCEPTANCE CRITERIA FOR THE DIVERTOR CFC MONOBLOCK ARMOUR.

D'Agata Elio, Tivey Richard

Boltzmannstrasse, 2 D-85748 Garching Germany

The plasma-facing components (PFCs) of the divertor are subjected to high heat flux (HHF). Carbon-fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets. Failure of the heat sink to armour heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFC to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints. This paper discusses the implications on the operation of ITER of both the failure and the sub-standard performance of components. Based on this understanding, acceptance criteria are proposed. Firstly, to ensure that the erosion rate and hence the carbon released into the divertor channel is within tolerable limits, and secondly, to make it extremely unlikely that, because of defects in the structure and/or poor thermal conductivity, the critical heat flux (CHF) will be exceeded leading to an ingress of coolant event (ICE) into the main chamber. Most promising are the thermographic techniques such as those developed by CEA and Plansee. These have shown that defects can be detected in relatively thin-walled (» 5 mm) armour. However, with thick-walled armour with anisotropic properties like that proposed for the ITER divertor, system errors, largely due to variations in thermal conductivity, mean only relatively large defects can be detected with any certainty. This means that although defects that might lead to an ICE can be detected, smaller defects that will contribute to a reduced armour lifetime cannot. One solution to this problem is to use acceptance criteria that take account of the standard deviation in the thermal conductivity of the CFC. This paper will propose such limits that should ensure a satisfactory performance of the ITER divertor. It is important that these limits be defined in advance of any manufacturing contract with industry, and that they are workable without reducing the demands on the manufacturer to provide a high quality product.


Corresponding Author:

D'Agata Elio

Boltzmannstrasse, 2 D-85748 Garching Germany

- F - Plasma Facing Components.

P4C-F-100 RESULTS AND ANALYSIS OF HIGH HEAT FLUX TESTS ON A FULL SCALE VERTICAL TARGET PROTOTYPE OF ITER DIVERTOR

MISSIRLIAN Marc, F. Escourbiac (1) M. Merola (2) I. Bobin-Vastra(3) J. Schlosser (1) A. Durocher (1)

(1) CEN Cadarache, 13108 Saint-Paul-Lez-Durance (FRANCE) (2) EFDA Close Support Unit, Garching (GERMAN) (3) FRAMATOME, Le Creusot (FRANCE)

An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a Full Scale Vertical Target (VTFS) prototype of ITER divertor has been tested by means of the FE200 electron beam European facility at Framatome in Le Creusot (France). This component was 1000 mm long and contained all the main features of the corresponding ITER design. Four units using entirely the monoblock technology were assembled in parallel and actively water cooled. The armour upper part of the prototype was made of an alloy of tungsten (W-1%La2O3) lamellae whereas the lower part was made of Carbone Fibre reinforced Carbon (CFC-NB31). The heat sink was in precipitation hardened copper (CuCrZr) equipped with a swirl insert into the straight part of the cooling channel. The manufacturing technology was Active Metal Casting (AMCâ) followed by an Hot Isostatic Pressing (HIP) step. The rear side of each monoblock tiles is machined in order to allow slidings in its stainless steel support structure. Several steps of fatigue cycling on CFC and W armoured regions were planned on this prototype taking into account ITER safety margin requirements in terms of thermal fatigue. CFC monoblocks were tested up to 23 MW/m2 x 2000 cycles (10 s heating phase/10 s dwell phase) on the straight part without any indication of failure. W monoblocks endured 10 MW/m2 before a first water leak after ~600 cycles and 15 MW/m2 before a second water leak after ~100 cycles. After these high heat flux experiments, metallographic examination were undertaken on the damaged units of the prototype. This paper summarises the main test results and describes the numerical simulation of the thermomechanical behaviour of the VTFS mock-up during the production process as well as during the thermal fatigue loading. The purpose of the thermomechanical analyses coupled to damage valuation is to allow a reasonable interpretation of the occurred phenomena during this fatigue cycling campaign. Thermal and mechanical stress analyses have been performed using the CAST-3M finite element code including transient and steady state thermal analyses as well as fatigue life time evaluation under ITER operating conditions.


Corresponding Author:

MISSIRLIAN Marc

CEN Cadarache, 13108 Saint-Paul-Lez-Durance (FRANCE)

- F - Plasma Facing Components.

P4C-F-115 STRUCTURAL AND FRACTURE MECHANICS ANALYSIS OF ITER TOROIDAL FIELD COIL

Shuji Iki, K.Watanabe(1) K.Takiue(2) M.Saito(2)

(1)JFE Steel Corporation(East Japan Works),Chiba,260-0835,Japan (2)Univ.of Tsukuba,Tsukuba,Ibaraki,305-8573,Japan

One of structural issues of toroidal field (TF) coil of a tokamak fusion device is a considerably large transverse displacement due to out-of-plane electromagnetic loads induced by the interaction of TF coil current with the magnetic fields. The magnetic fields are varying with central solenoid (CS) coil current, poloidal field (PF) coil current and plasma current. These currents are controlled to realize the assumed operation scenario. The first purpose of the study is evaluations of maximum displacement and maximum stress of TF coil to confirm the structural integrity of TF coil. The TF coils without shear panels are examined using standard operation scenario. The lower end of a gravity support is fixed as a boundary condition and in the bonded region with adjacent TF coil, a periodical boundary condition is employed. The numerical results are: TF coil leans toward one direction in beginning and leans toward the reversed direction in nearly end of the operation scenario. The maximum transverse displacement is beyond 110mm and appears in the upper curved part of TF coil. From the viewpoint of plasma physics, the magnetic field disturbed by the TF coil displacement may be significant. Thus in ITER, the shear panels are installed to reduce the transverse displacement. The maximum Mises stress of about 590MPa appears in the lower bonded region in nearly end of the scenario. The lower bonded region is loaded by a tension-compression cycle in one shot of plasma. Since the maximum stress area is restricted locally in the structural material of TF coil case, there appears no highly stressed area inside the coil case, that is, the superconducting material is not so strained. However, because of the stress amplitude in every shot of plasma, the lower bonded region should be examined from the viewpoint of the possibility of fatigue crack propagation. The second purpose of this study is the detailed evaluation of stress field around the initial crack assumed in the lower bonded region. The stress intensity factor K is employed to investigate the crack growth. From fracture mechanics analysis, there is no significant crack growth in the anticipated cycle of pulses in ITER. In conclusions, globally, in a sense of structural mechanics, the electromagnetic load is large and gives a considerably large transverse displacement of TF coil, but locally, in a sense of fracture mechanics, the electromagnetic load is not so large and gives only a small value of K.


Corresponding Author:

Shuji Iki

Saito-lab,Institute of Engineering Mechanics and Systems,University of Tsukuba,Tsukuba,Ibaraki 305-8573,Japan

- F - Plasma Facing Components.

P4C-F-116 CRACK PROPAGATION BEHAVIOR AROUND DSCU/SS316 HIP BONDED INTERFACE BY THERMAL FATIGUE

Takahiro Oyama, Akira Yamamoto(1) Kensuke Mhori(2) Masakatsu Saito(1)

(1)Saito-lab, Institute of Engineering Mechanics and Systems, Univ. of Tsukuba, Tsukuba, Ibaraki, 305-8573, Japan (2)Kawasaki Heavy Industries, Ltd., Tokyo, 136-8588, Japan

The pulse operation is assumed in ITER. If the first wall has a defect, a crack may be propagated. The first wall is composed of DSCu and SS316. HIP bonded process is employed to fabricate the first wall. This study deals with the crack propagation behavior around HIP bonded interface by thermal fatigue in order to confirm integrity of HIP bonded joints. Thermal fatigue experiments were carried out by use of the EB gun as a heat source. Specimens are DSCu/SS316 HIP bonded plates. Their thicknesses are 3.0mm and 5.0mm respectively. A surface crack of nearly 0.5mm depth is introduced in DSCu. DSCu surface including an initial crack was cyclically irradiated by heat flux. SS316 surface was cooled by the cooling plate. The maximum temperature difference was about 500degree. As a result, the crack propagation direction changed perpendicularly near HIP bonded interface. Cracks along HIP bonded interface were propagated on the surface which left the thin copper layer (thickness about 20micro meter) to SS316. The cracks did not penetrate into SS316. The crack propagation rate on HIP bonded interface (estimate over 1micro meter/cycle) was much larger than the crack propagation rate in DSCu (over 0.1micro meter/cycle). 2-dimensional elasto-plastic thermal stress analysis of DSCu/SS316HIP bonded plates was carried out by use of finite element code MARC. First, this analysis explains the change of crack propagation direction around HIP bonded interface from the viewpoint of fracture mechanics. Delta J hat, the amplitude of efficient J hat integral, was calculated by use of deLorenzi's virtual crack extension method. Two delta J hat, delta J_x hat along the crack surface and delta J_y hat perpendicular to the crack surface, were compared. When the crack tip in DSCu is far from HIP bonded interface, delta J_x hat is larger than delta J_y hat. When the crack tip comes near to HIP bonded interface, delta J_y hat increases rapidly. At HIP bonded interface delta J_y hat is much larger than delta J_x hat. Second, stress intensity factor K_i defined along the interface was computed from the stress distribution. K_i decreases as the crack tip’s highest temperature decreases.


Corresponding Author:

Takahiro Oyama

Saito-lab, Institute of Engineering Mechanics and Systems, Univ. of Tsukuba, Tsukuba, Ibaraki, 305-8573, Japan

- F - Plasma Facing Components.

P4C-F-133 SIMULATION OF MANY-ATOMIC INTERACTIONS IN W-O-H SYSTEM WITH THE MD CODE CADAC

I.S. Landman,

In the vessel of future tokamak reactors such as ITER, tungsten is the first candidate material for the divertor armour and perhaps for the first wall. Chemical erosion of tungsten surfaces under impact of scrape-off-layer (SOL) deuterium-tritium plasma containing impurities, for instance oxygen, is an important issue for its use in the reactor components. The impurities coming to the irradiated surface can form volatile molecular complexes WxOy. The H-atoms retained near the implantation layer can create the volatile complexes HxOy, which reduces the formation of WxOy and mitigates the corrosion effect (H states for tritium or deuterium). The impact of the hot H-ions of the SOL plasma can destroy the complexes and influence the surface chemistry drastically. To investigate this W-O-H system the molecular dynamics (MD) simulation code CADAC was recently developed. In this work, the previous version of CADAC that contains a pair-atomic interaction algorithm is extended to include many-atomic interactions. Now CADAC can simulate such important effects as the molecular organization of atoms. To achieve this generalisation of the code, some practical however rather general new approach that enabled to tackle complexity of the system is introduced. MD is naturally combined with the concept of valence. The model approximates the chemical reactions using atomic valences and available data on the atom-atom pair-wise interactions. In this work the new model itself is described for the first time, some elementary atomic configurations built of H-, O- and W-atoms are analyzed, such important atomic systems as O2 and H3 gases and the W-lattice with many-atomic potentials are simulated at room temperature, and the first results on the chemical erosion of tungsten surface with the many-atomic interactions taken into account are presented.


Corresponding Author:

I.S. Landman

Forschungszentrum Karlsruhe, Institute for Pulsed Power and Microwave Technology, Post Box 3640, 76021 Karlsruhe, Germany

- F - Plasma Facing Components.

P4C-F-135 DESIGN OF A LIMITER FOR THE JET EP ICRH ANTENNA

Chappuis Philippe, Christophe Portafaix(1) Eric Thomas(1) Bernard Bertrand(1) Robert Walton(2) Valeria Riccardo(2) Richard Baker(2) Ian Barlow(2) Alan Kaye(2) Axel Lorenz(3),

(1)Euratom-CEA Cadarache, CEA/DSM/DRFC, 13108 St Paul lez Durance, France (2)Euratom-UKAEA/Fusion, Culham, OX14 3DB Abingdon, United Kingdom (3)EFDA-JET CSU-Culham, OX14 3DB Abingdon, United Kingdom

A new set of poloidal limiters have been designed and manufactured to allow the installation of the upgraded ICRH system on the JET machine within the available wall space. The aim of the ICRH upgrade is to demonstrate adequate power handling capability in conditions relevant to the ones to be found on ITER-FEAT (large distance between launcher and plasma last closed flux surface and ELMs). The Antenna is surrounded by a frame of Carbon tiles allowing for a full protection against any impinging particle flux on the straps. The tiles are attached to an Inconel frame designed to control efficiently the flow of current between the plasma and the vessel. The frame is connected to the vessel by mechanical supports allowing for the vessel distortions and adjustment during installation, baking and to a lesser extent operation. The limiter beams and all the tiles are installed in the vessel using Remote Handling (RH). The plasma facing carbon tiles were shaped to comply with the existing limiters and with various plasma scenarios. Based on magnetic equilibrias calculated using the finite element code Proteus, the optimisation was achieved through the CFPflux field line tracing code to ensure the correct shadowing of all edges. Consequently the designed surface temperature should remain lower than 900 C in the worst case. The Eddy and Halo current distribution in the limiter frame was determined with the ANSYS using design current sinks & sources and field transients based on previous operational experience. The total current flow is controlled by the use of resistive straps between the different limiter elements. The mechanical calculations linked to the interaction of these currents with the magnetic field led to the optimisation of the limiter weight (RH constrains) while maintaining allowable stresses in all elements. As part of the ICRF Antenna project, the design of the Poloidal limiter was carried out by CEA, supervised by EFDA CSU JET in close co-operation with the JET operator (UKAEA). After agreement by the operator Quality Assurance (QA) system and selection of a proper qualified company by way of an international tender, the manufacturing of all elements is ongoing and should be achieved in August 2004.


Corresponding Author:

Chappuis Philippe

CEA DRFC, CEA Cadarache, 13108, St Paul lez durance, France

- F - Plasma Facing Components.

P4C-F-145 EROSION OF TUNGSTEN MACROBRUSH ARMOR AFTER MULTIPLE INTENSE TRANSIENT EVENTS IN ITER

Bazylev Boris, G.Janeschitz (1) I.S. Landman (1) A. Loarte (2) S.E. Pestchanyi(1)

(1) Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe, Germany (2) EFDA-CSU, Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching, Germany

In the future tokamak ITER, tungsten is foreseen as one of perspective materials for the divertor and the dome. The main disadvantage of bulk tungsten armour is surface cracking under high heat loads typical for the intense transient events such as disruptions and ELMs. In one ITER discharge about 1000 ELMs are expected. During ITER operation several hundred disruptions may occur. One possibility to mitigate the surface cracking is the tungsten macrobrush armour (W-brushes). However during the transient events, when a significant part of confined plasma is dumped onto the macrobrush elements, it may result in surface melting of them. The melt motion may produce significant surface roughness and droplet splashing thus causing erosion of the elements. The separatrix strike position (SSP) at the surface can substantially vary in sequential ELMs, which correspondingly changes the distributions of the heat flux and the pressure of impacting plasma at the target. The results of fluid dynamics simulations for the melt motion erosion of W-brushes after multiple stochastically varied plasma heat load pulses typical of the ITER regime are presented for the following ranges of the surface energy deposition Q and the pulse duration t: Q = 5–100 MJ/m2 and t = 1–10 ms (disruption), Q = 1–5 MJ/m2 and t = 0.1–0.5 ms (ELM). The heat loads are calculated applying the two-dimensional MHD code FOREV-2D taking into account the vapour shield in front of the target and radiation transport in the ITER magnetic field configuration for both the face- and lateral structures of W-brushes. The target melt motion erosion is calculated by the fluid dynamics code MEMOS-1.5D. The surface tension and the viscosity of molten metal as well as the Lorentz forces due to the currents crossing the melt layer are taken into account. The geometric peculiarities of W-brushes and the melt motion along the gap edges as well as the features of energy deposition in such a complicated geometry are implemented in MEMOS-1.5D remaining in frame of the “shallow water” approximation on which the code is based. The erosion features of the tungsten bulk armour and that of the W-brushes are compared. Latest validations of the codes against the experiments on plasma guns and the tokamak JET are described.


Corresponding Author:

Bazylev Boris

IHM, Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe, Germany

- F - Plasma Facing Components.

P4C-F-146 DEVELOPMENT OF AN ORIGINAL ACTIVE THERMOGRAPHY METHOD ADAPTED TO ITER PLASMA FACING COMPONENTS CONTROL

Alain Durocher, N.Vignal(a), F.Escourbiac(a), J.L.Farjon(a) , J.Schlosser(a), F.Cismondi(b)

(a)Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 SAINT PAUL LEZ DURANCE, France (b) Université de Toulon et du Var, BP 132 83957, LA GARDE, France

The interface inspection by active infrared thermography of actively cooled components has been a technique used at TORE SUPRA for several year in the field of the High Heat Flux (HHF) components. An infrared thermography test bed named SATIR (Station Acquisition Traitement InfraRouge) has been developed specially by CEA in order to evaluate the manufacturing process quality of actively water-cooled plasma facing components. The technical specifications for the supply of ITER Divertor Vertical Targets (DVT) stated that all Cu cast layers on W or CFC armour shall be subjected to 100% thermographic examination, such as the CEA developed SATIR test. Today, the current SATIR facility does not allow control for the large-sized HHF components such as those ITER DVT. In the past, the control of full-scale ITER DVT mock-up showed the limitations of the SATIR test bed in term of measuring accuracy, flow rate capability (2m3/h), low pressurization of water loop (3bar), and useful hot water volume (1.2m3). However the last studies realised on SATIR allowed to define the features of a new installation named "SATIRPACA" adapted to the control of ITER DVT: - Obviously, the principle of the SATIR test bed by internal thermal excitation remains a very interesting method and must be preserved because it measures exactly ability of the component to be cooled. Moreover in 2004 important improvements have been performing about the detection sensitivity. - An unique Non Destructive Examination method at this level of high technology is not sufficient. The original coupling of SATIR with a lock-in thermography system will be realized in this study, which will allow to have two infrared thermographic inspection methods on the same test bed. An improvement of the global reliability of the coupled facility is expected by merging the data produced by the two techniques. Afterwards the using of the cooling channel of HHF component during the test will improve the sensitivity of detection by the lock-in thermography method. - The heat transfer convective coefficient will be also improved by installing an upgraded flow rate device. The new proposed enhanced test bed is designed for the full scale Non Destructive Examination of the HHF ITER components. The control of Wendelstein-7X HHF components expected from the beginning 2005 will allow to validate SATIRPACA test bed. This paper proposes a detail overview of improvements which will equip this new infrared test bed.


Corresponding Author:

Alain Durocher

Association Euratom-CEA, CEA/DSM/DRFC,CEA/Cadarache, F-13108 SAINT PAUL LEZ DURANCE, France

- F - Plasma Facing Components.

P4C-F-162 PLASMA SPRAYED TUNGSTEN-BASED COATINGS AND THEIR PERFORMANCE UNDER FUSION RELEVANT CONDITIONS

Matejicek, Jiri, Vladimir Weinzettl (1) Yoshie Koza (2)

(1) Institute of Plasma Physics, Za Slovankou 3, 18221 Praha, Czech Republic (2) Forschungszentrum Juelich, IWV-2, D-52425 Juelich, Germany

Tungsten is one of the candidate materials for plasma facing components for ITER and other fusion devices. Plasma spraying is among prospective fabrication technologies, thanks to its ability to coat large areas and the possibility of in-situ repair. This paper reports on the development of tungsten and tungsten+copper plasma sprayed coatings and their behavior under high heat fluxes and in tokamak plasma. Several tungsten-based coatings were produced at IPP, using water-stabilized plasma spraying from different powders and under various spraying conditions. Their basic properties (structure, composition, etc.) were characterized by SEM, XRD and other techniques. The coatings contain varying amount of porosity and oxides; these factors are subject to further process optimization. The behavior of plasma sprayed coatings and a solid tungsten sample under high temperature plasma conditions was investigated at the small-size CASTOR tokamak at IPP (ohmic heating 30 kW, pulse length of 30 ms, electron temperature ~30-60 eV, ion temperature ~10-20 eV, plasma density 0.5-1x10-19 m-3). The samples were inserted into plasma at various radii and exposed to standard plasma discharge conditions. Main plasma parameters were observed, together with impurity radiation in the XUV, VUV and visible ranges. Only a local influence of the tungsten presence was found. Heating and cooling rates were measured between the shots. No melting and very limited surface modification of the materials was observed. Selected coatings were tested under high heat fluxes at the electron beam facility JUDITH at FZJ, to simulate disruption conditions. The samples were subjected to different incident beam current (30-125 kV) and loading time (5-10 ms) over the area of ~7 mm2, while the absorbed current, surface temperature and particle emission were recorded. The induced changes were observed by surface profilometry, SEM and optical microscopy. These were namely the removal of oxide scale at lower incident energies, surface melting at intermediate and deep melting at high energies. The coatings were able to absorb about 0.5 GW/m2 (2.5 MJ/m2) in thermal shock loading without significant damage.


Corresponding Author:

Matejicek, Jiri

Institute of Plasma Physics, Za Slovankou 3, 18221 Praha, Czech Republic

- F - Plasma Facing Components.

P4C-F-167 HIGH TEMPERATURE STRESSES IN ITER RELEVANT BRAZED GLIDCOP/W MODEL STRUCTURES

Coppola* Roberto, C. Nardi 1, M. Valli 2

(1) ENEA-Frascati, FUS, CP 2400, 00100 Roma, Italy (2) ENEA-“Clementel“, V. Don Fiammelli 2, 40129 Bologna, Italy

It is well known that plasma-facing components of near term machines, such as ITER, must be designed to withstand huge and long (1000 s) heat charges with consequent thermo-mechanical stresses. Therefore, the knowledge of the stress field in components such as the divertor is essential to define the engineering parameters required to design real-scale components. This contribution will present the results of an experimental study on high temperature stress evolution in brazed Glidcop/W model structures. The samples, approximately 23 x 23 x 8 cm3 in volume, were obtained by brazing a W and a Glidcop platelet at 650 C using TiCu Ag alloy as a filler. Neutron diffraction was utilized to determine the strains, then the stresses in the bulk of the samples between room-temperature and 500 C; unstrained Glidcop and W reference samples were measured as well. The measurements were carried out at the D1A diffractometer available at the High Flux Reactor of the ILL-Grenoble. The experimental results provide relevant engineering information such as the zero strain temperature. The adopted experimental procedure will also be discussed in view of its possible use for real-scale components.


Corresponding Author:

Coppola* Roberto

ENEA-Casaccia, FIS, CP 2400, 00100 Roma - I

- F - Plasma Facing Components.

P4C-F-176 A MATURE INDUSTRIAL SOLUTION FOR ITER DIVERTOR PLASMA FACING COMPONENTS: HYPERVAPOTRON COOLING CONCEPT ADAPTED TO TORE SUPRA FLAT TILE TECHNOLOGY

ESCOURBIAC FREDERIC, V.Kuznetsov(2) M.Missirlian(1) B.Schedler(3) J.Schlosser(1)

(1) Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 SAINT PAUL LEZ DURANCE, France (2) Efremov institute, Doroga na Metallostroy, St. Petersburg, 196641, Russia (3) Plansee AG, 6600 Reutte, Austria

The design heat flux for specific plasma facing components in ITER is in the same range (10-20 MW/m²) than observed in electron tubes. Historically, concepts with enhanced cooling capabilities implying boiling/condensation effects due to a fin design named hypervapotron were developed by Thomson CSF tube Company for such purposes and later designed for neutral beam heating systems. This cooling concept adapted to a CuCrZr heat sink armoured with CFC or W was envisaged for the vertical targets of the ITER divertor since the beginning of ITER EDA, but finally abandoned for two main reasons : it was suspected that the joint temperature between CFC or W and CuCrZr may be too high as well as a possible occurrences of a “cascade tile failure” effect. Last experimental results accompanied with progress in modelling have shown excellent behaviour of hypervapotron based armours with regard to the two mentioned supposed disadvantageous arguments : temperature of the armour/heat sink joint - strongly dependent on the flow velocity – can be driven below a tolerated limit of 500 C and cascade tile failure occurrence was not experimentally observed. In order to validate the hypervapotron concept as a design solution for the ITER divertor, thermal fatigue testing has been performed on two medium scale mock-ups. They were manufactured by Plansee AG with respect to the main technological features of a TORE SUPRA toroidal limiter finger element. One of these mock-ups was tested in the European facility FE200 (Elecron Beam 200 kW) and the other one in the Russian facility TSEFEY-M (Elecron Beam 60 kW). Both testing campaigns have shown that the mock-ups were able to sustain with margins corresponding to the divertor requirements in terms of thermal fatigue : 3000 cycles at 15 MW/m², 800 cycles at 25 MW/m² and a critical heat flux limit higher than 30 MW/m². Analyses of tests results will be reported in this paper. (*)High heat flux testing partially supported by EFDA


Corresponding Author:

ESCOURBIAC FREDERIC

(1) Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 SAINT PAUL LEZ DURANCE, France

- F - Plasma Facing Components.

P4C-F-192 CONCEPTUAL DESIGN OD A HIGH-TEMPERATURE WATER-COOLED DIVERTOR FOR A FUSION POWER REACTOR

GIANCARLI Luciano, J.P. Bonal (1), A. Li Puma (1), B. Michel (2), J.F. Salavy (1), P. Sardain (3)

(1) CEA/Saclay, DEN/DM2S, 91191 Gif-sur-Yvette, France (2) CEA/Cadarache, DEN/DER, 13108 St.Paul-lez-Durance, France (3) EFDA, CSU-Garching, Boltzmannstr. 2, D-85740 Garching, Germany

The large effort devoted in recent years to the divertor development for ITER reactor has shown that the divertor is a critical reactor component because of the severe operating conditions which have to be withstand, such as very high surface heat fluxes, interaction with energetic plasma particles, and complex geometry. Additional requirements need to be fulfilled for a power reactor divertor such as the resistance to high neutron fluxes and fluences, use of high temperature coolant for achieving an acceptable overall reactor thermal efficiency, and, possibly, the use of low activation materials. This paper presents the studies performed in the framework of the EU Power Plant Conceptual Study (PPCS) concerning the development of the conceptual design of a water-cooled divertor using low-activation martensitic steel (EUROFER) as structural material, water coolant at PWR conditions (15.5 MPa pressure and 325 C outlet temperature), and W-alloy monoblock as armour. The concept consists of a series of EUROFER pipes for coolant flow, each of them surrounded by a W-alloy monoblock, attached to a common EUROFER back plate. The concept is able to withstand a continuous surface heat flux of 15 MW/m2, reaching an acceptable maximum structure temperature of about 516 C and showing acceptable stresses, provided an appropriate interface between pipes and monoblock is used. EUROFER has been selected due to its expected capability of withstanding neutron damages higher than 70 dpa (eq. Fe). However, because of its relatively low thermal conductivity and its differential thermal expansion with the W-alloy, direct joints EUROFER/W-alloy cannot be used in order to avoid too high thermal stresses. Therefore it is proposed to add at the interface a thermal barrier on the front half of the pipes, made of pyrolitic graphite to enhance the thermal flux repartition, and of a compliance layer made of soft graphite “papyex”. In this case, thermal stresses become acceptable and the maximum W-alloy temperature is about 2000 C. The main issues of this divertor concept are the manufacturing process of the steel/W interface and the behaviour under irradiation of graphite materials. Experimental data up to 30 dpa (eq. C) have been collected in the literature and their assessment, presented in this paper, shows that the behaviour of such materials, when used as thin layers without mechanical functions, could be acceptable.


Corresponding Author:

GIANCARLI Luciano

CEA/Saclay, DEN/CPT, 91191 Gif-sur-Yvette, France

- F - Plasma Facing Components.

P4C-F-195 DEVELOPMENT OF A COPPER ALLOY TO BERYLLIUM HIP BONDING TECHNOLOGY FOR THE ITER FIRST WALL

P. Sherlock (1), A. T. Peacock (2) A. D. Mc Callum (1)

(1) NNC Limited, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England (2) EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching, Germany

The Primary First Wall (PFW) modules of the ITER blanket concept are covered with separable PFW panels. The PFW panels comprise a bi-metallic copper alloy / stainless steel 316L water-cooled heatsink faced with a plasma facing material. Dispersion Strengthened copper (DS-Cu) and precipitation strengthened CuCrZr are options for the copper alloy. One option for the plasma facing material is beryllium, in the form of tiles. Over recent years, the technology needed to bond beryllium tiles to the copper alloy of the heatsink has been developed. During this development, solid HIP bonding has been employed as one method to produce the heatsink base and bond the beryllium tiles in place. The development of the manufacture is typically done in three stages. Small samples are first produced in the laboratory to show that beryllium can be bonded to CuCrZr under ideal conditions using the selected parameters. Larger mock-ups are then produced which have some geometrical aspects that are similar to those of the full size panels. This shows the selected bonding parameters can be used under manufacturing conditions. The final stage is to produce full size prototype PFW panels to prove the technology at this scale [1]. During the first two stages structural analysis, mainly in the form of finite element analysis, has been used to assess the mechanical behaviour during the manufacturing process. For the small samples, the analysis models the stresses that result from the differential thermal expansion between the beryllium and copper alloy. This assists in the selection of the compliant layer which strains to accommodate the expansion and reduce residual stress. On the larger mock-ups, the interaction between the HIP can and the mock-up itself during the HIP processes is modelled to progress the design of the HIP can / mock-up assembly. This paper describes the small samples and larger mock-ups produced by NNC during the development of the copper alloy / beryllium HIP bonding technology. It demonstrates how structural analyses were used to gain an understanding of the bonding process and develop the HIP can / mock-up assembly design. [1] Manufacture of blanket shield modules for ITER, P. Lorenzetto et. al., this conference.


Corresponding Author:

P. Sherlock (1)

NNC Limited, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England

- F - Plasma Facing Components.

P4C-F-228 AN ADVANCED HE-COOLED DIVERTOR CONCEPT: DESIGN, COOLING TECHNOLOGY, AND THERMOHYDRAULIC ANALYSES WITH CFD

Ihli, Thomas (1), R. Kruessmann (1), I. Ovchinnikov (2), P. Norajitra (1), V. Kuznetsov (2), R. Giniyatulin (2)

(1) Institute for Materials Research III, Forschungszentrum Karlsruhe GmbH, P.O. Box 3640, 76021 Karlsruhe, Germany (2) D.V. Efremov Institute, Scientific Technical Center "Sintez", 196641 St. Petersburg, Russia

An advanced modular helium-cooled divertor concept for near-term reactor models like DEMO is being investigated at Forschungszentrum Karlsruhe (FZK). It is based on the multiple jet impingement cooling technology which is efficiently applied in the gas turbine sector. The major challenge for the divertor concept is to handle target heat loads of up to 10-15 MW/m². Hot helium is chosen as coolant due to its advantageous safety characteristics. It allows for a high exit temperature of at least 700 to 750 C, which is suitable for the power conversion system that uses a gas turbine cycle. A highly effective and reliable cooling system is necessary to fulfil the requirements. Nevertheless, the pumping power for the coolant should be kept as low as possible and the thermal stresses caused by extreme temperature differences in the heat-loaded and cooled parts have to be kept below acceptable limits. This leads to a segmented bodywork for the targets which consist of small multiple finger units (e.g. 9 fingers in their own housing). The fingers consist of small W-amour parts, that are brazed onto separate pressure-carrying components (caps). The interior of the caps is cooled by arrays of helium jets supplied from impingement hole arrays in cartridges which are inserted into the caps. The finger units can be tested separately before fixing them to stripe units which form the targets. The divertor system presented is extremely flexible and can be adapted to all kinds of gas-cooled reactors by adjusting the jet hole configuration and the numbers of parallel and series connections of the small multiple finger units. It ensures a high heat transfer coefficient at a well-balanced mass flow rate. The performance of the concept is investigated by means of finite-element (FE) and computational fluid dynamics (CFD) analyses. The results of a CFD parameter study, focusing on the minimum helium mass flow rate required for cooling and the respective temperature distributions are incorporated in the thermohydraulic design. They provide for an iterative approach comprising CFD and FE calculations for design improvement and the prediction of pressure loss and heat transfer coefficients prior to experimental investigations of the concepts. In this study, the design and cooling method shall be described briefly. Design performance and layout examples shall be highlighted using results of CFD calculations performed at FZK and the Efremov Institute.


Corresponding Author:

Ihli, Thomas (1)

Institute for Materials Research III, Forschungszentrum Karlsruhe GmbH, P.O. Box 3640, 76021 Karlsruhe, Germany

- F - Plasma Facing Components.

P4C-F-239 THE NEW ELECTRON BEAM TEST FACILITY JUDITH II FOR HIGH HEAT FLUX EXPERIMENTS ON PLASMA FACING COMPONENTS.

Majerus Patrick, Reiner Duwe (2) Takeshi Hirai (1) Winfried Kühnlein (2) Jochen Linke (1) Manfred Rödig (2)

(1) IWV 2, Forschungszentrum Jülich GmbH, EURATOM-Association, 52425 Jülich, Germany (2) B-Z, Forschungszentrum Jülich GmbH, EURATOM-Association, 52425 Jülich, Germany

The Juelich Divertor Test Facility in Hot Cells JUDITH I has been operating successfully in the Research Centre Jülich since the early nineties. It represents a unique high heat flux experiment for simulating thermal loads on neutron activated plasma facing components. The potential for testing toxic materials, like beryllium and for simulating all ITER relevant thermal loads (including disruptions, VDEs and ELMs) made JUDITH I break its capacity during the recent years. To extend the parameter range and because of the urgent need of additional testing capacity a new electron beam facility JUDITH II is being build up. Beyond the ability to perform the same type of experiments as in JUDITH I, a whole range of optimised features is included into the new facility. An approximately three times higher nominal power of 200 kW, combined with a beam scanning angle of ±14 enables to test larger components up to 0,5 x 1 m². The relatively small acceleration voltage, adjustable between 30 and 60 kV, reduces volumetric heating for the benefit of a more plasma like surface heating. Due to a very flexible and individual programmable system for electron beam pattern generation, a highly homogeneous load distribution can be achieved. Two different testing modes, pulsed and beam sweeper mode, allow rather realistic simulations of the ITER relevant transient loads. Power densities up to 10 GW/m² and minimum event times as short as 2 µs are the only limiting parameters. This offers a full range of new possibilities in simulating ELMs with deposited energy densities in the order of 1 MJ/m² and pulse durations of several hundred microseconds. ELMs have only recently been identified as possibly life-time limiting events in future confinement experiments, such as ITER. Furthermore it will become possible to combine static and transient loads in one single experiment. Besides IR-diagnostics, high resolution and fast image grabbing will allow an enhanced study of the effects caused by transient loads on armour candidate materials. Especially brittle destruction will be addressed, using a spectrometer in the visible range, a photodiode array and acoustic emission. The first two methods serve to analyse the emitted particle while acoustic emission shall give information on the onset of brittle destruction. With a combination of the applied methods it is expected to additionally measure the energy release rate per emitted particle.


Corresponding Author:

Majerus Patrick

IWV-2, Forschungszentrum Jülich GmbH, EURATOM-Association, 52425 Jülich, Germany

- F - Plasma Facing Components.

P4C-F-253 FORMATION OF CRYSTALLINE NANOSTRUCTURES DURING DEUTERIUM PLASMA INTERACTION WITH TUNGSTEN-BASED MATERIALS IN SIMULATED GAS DIVERTOR CONDITIONS.

Guseva Mariya, V.M. Gureev, L.S. Danelyan, B.N. Kolbasov, S.N. Korshunov, V.B. Petrov, B.I. Khripunov

Nuclear Fusion Institute, RRC Kurchatov Institute, Kurchatov sq. 1, 123182 Moscow, Russia

Sputtering of W-based materials considered as candidates for ITER divertor armour manufacturing [W–10Re, W–1La2O3, W–13I and W(111)] by deuterons with subthreshold energy (5 eV) in a dense steady-state plasma of the LENTA facility was studied by weighing technique at armour temperatures from 1250 to 1520 K. It has been found that sputtering of these materials occurs above 1250 K. At 1520 K and irradiation dose of 1.5x10*26 m*-2, the sputtering yields for W(111), W–13I, W–10Re and W–1La2O3 are 1.1x10*-4, 2.6x10*-4, 2.9x10*-4 and 5.3x10*-4 respectively. Specimen weight loss decreases with increase in irradiation dose, e.g. the weight loss of the W–10Re specimen reduced fivefold when the irradiation dose doubled (to 3x10*26 m*-2). Microstructure studies and X-ray diffraction analysis suggest that such an effect is due to the formation of some special structures on the W surfaces. At an irradiation dose of 3x10*26 m*-2, surfaces of the specimens acquire a block nanostructure consisting of various polygons, including pentagons, hexagons and heptagons of different areas. The smallest blocks (~100-nm) were observed on the W–13I surface and the largest (0.1-3 mm) – on the W–1La2O3 surface. The presence of pentagons, hexagons and heptagons may be attributed to the condensation of sputtered atoms and the formation of a new type of substance under irradiation by deuteron flux of 10*18 cm*-2s*-1. X-ray diffraction analysis suggests that the W structures on the specimen surfaces are strongly textured in the <110> direction. The W lattice spacing (3.165 Å) is the same for all the specimens within the limits of experimental error. The structures on the surfaces of W–10Re, W–13I and W(111) are practically continuous and have a good adhesion. Those covering W–1La2O3 (PLANSEE) have 1–2-mm holes of irregular shapes. Deuterium (D) and protium (H) were detected in a narrow, ~20 nm thick, near-surface film layer using elastic recoil detection analysis. D content in this layer is insignificant (<0.05 at.%). H concentration is 2.5 at.%. Thus, our findings suggest that surface structures with negligible D content, preventing further erosion of W-based ITER divertor elements would arise on the surface of these elements even after 7-8 normal (400-s) pulses, provided the elements’ temperature is kept in the range of 1250-1520 K. It would be of interest to look into the possibility of using the nanostructures forming on the W surfaces for technological purposes.


Corresponding Author:

Guseva Mariya

Nuclear Fusion Institute, RRC Kurchatov Institute, Kurchatov sq. 1, 123182 Moscow, Russia

- F - Plasma Facing Components.

P4C-F-265 ACTIVITY OF THE EUROPEAN HIGH HEAT FLUX TEST FACILITY: FE200

I. Bobin-Vastra, F. Escourbiac (2), M. Merola (3), P. Lorenzetto (3)

2) CEA-DRFC-SIPP, Cadarache, 13108 St Paul lez Durance (F) 3) EFDA Close Support Unit, Boltzmannstr. 2, D-85748 Garching, Germany

FE200 is an Electron beam (EB) 200KW test facility stemming from partnership between Framatome Technical Center in Le Creusot (F), and Tore Supra team in CEA Cadarache (Euratom/CEA association), dedicated to high heat flux testing of plasma facing components for fusion devices. Since 1992, an extensive development program has been carried out in the FE200 high heat flux facility especially for the ITER and TORE SUPRA tokamaks. In this framework, more than 100 000 cycles for thermal fatigue tests, 400 critical heat fluxes for different hydraulic conditions, 200 disruptions and 2 tests with glancing incidence (cascade failure) were performed on materials such as Cu-Al25 and CuCrZr alloys, Carbon Fibre Composite (CFC) or Tungsten (W) monoblocks and tiles, and plasma spray-W coatings for various actively cooled plasma facing component designs. The tests concerned small mock-ups as well as large components with a length ranging from a few tens of millimeters up to 1m in length. The facility includes a 200KW EB gun (200KV, 1A) which is able to deliver continuously during more than one hour from 0.1 to >100 MW/m² heat flux for thermal fatigue testing and up to 10GJ/m² during a few milliseconds for disruptions. A programmable sweeping allows several kinds of energy repartition (uniform and peaked), on a 13 shooting angle. In the 8m3 vacuum chamber, the maximum allowable length for components to be tested, is 1m if the surface is perpendicular to the beam, 2m when the component is tilted. The component is connected to a pressurised loop working from 0.2 to 3.3 MPa, at temperatures between 50 to 230 C, up to a maximum flow rate of 6kg/s. This large range of parameters gives a flexibility to the pressurised loop, which allows LOFA tests (Loss of Flow Accident) with successive cooling rate modifications during the test. Instrumentation gives information for diagnostics on calorimetry balance (absorbed heat flux), surface temperature till 2300 C (Infrared camera and pyrometers with remote positioning during test), visual aspect or behaviour (CCD camera with remote focusing). The paper illustrates the FE200 capabilities through several testing scenarios on different mock-ups and components tested in this facility, namely cascade failure configuration results on CFC and W, thermal fatigue tests performed on Primary First Wall (PFW) type components and divertor component, highest critical heat fluxes and LOFA tests on hypervapotron designed mock-ups.


Corresponding Author:

I. Bobin-Vastra

Framatome-anp Centre Technique (groupe AREVA), FE200, Porte Magenta, BP181, 71205 Le Creusot Cedex (F)

- F - Plasma Facing Components.

P4C-F-274 PROPOSAL OF LITIZATION OF FTU VACUUM VESSEL BY USING A LITHIUM LIMITER

Apicella Maria Laura, G. Mazzitelli (1) V.B. Lazarev (2) E.A. Azizov (2) S.V. Mirnov (2) V.G. Petrov (2) V.A. Evtikhin (3) I.E. Lyublinski (3) A.V. Vertkov (3) F. Lucca (4)

(1) Ass. ENEA-EURATOM sulla Fusione CR Frascati (2) Troitsk Inst. for Innovation and Fusion Research, Troitsk, Moscow Reg., RF (3) State Enterprise «Red Star» - Prana-Center Co, Moscow, RF (4) L.T. Calcoli SaS, Via C. Baslini, 13 - 23807 Merate (LC)

A new promising idea for the application of liquid lithium as plasma facing material in fusion reactors has been recently proposed and began to be tested. It is based on the surface tension forces in capillary channels that may be used to compensate forces induced in liquid lithium by the JxB effect under the plasma MHD events in tokamaks. The new structure, called CPS (Capillary Porous System) has been realized as a matt from wire meshes of Stainless Steel 304 with pore average radius 15 micron and wire diameter 30 micron. The liquid lithium flows inside these capillaries from on side of the system, which is in contact with a liquid lithium reservoir, to the other side that is faced to the plasma. The main features of CPS are the high stability and resistance to surface damage and the self-regeneration of the lithium surface through capillary forces. This last property becomes very important for divertor plates and the wall protection of ITER-like or post-ITER tokamak that will operate in the presence of ELMs which are the main reason of enhanced erosion. FTU, a medium size tokamak, represents a very good opportunity to test for the first time CPS configuration for a litization experiment by using a liquid lithium limiter. This experiment consists in the wall coating with a thin lithium film produced during a plasma discharge by a displacement of the LCMS (Last Closed Magnetic Surface) towards the lithium limiter. Before its installation on FTU, foreseen for the beginning of 2005, a detailed study of plasma scenario and lithium limiter operations have been done. In addition, a full electromagnetic analysis of the JxB forces and their influence on liquid lithium confinement in the capillary-porous limiter under disruption in FTU (Ip=1.6 MA, BT = 8 T) has been carried out. These results indicate that, assuming a realistic shape of CPS (three modules nearly semi cylindrical) and a thickness of CPS layer equal to 1mm, the peak amplitude of electromagnetic pressure can reach the value of 10 kPa which is a factor 5 lower than the capillary pressure able to retain liquid lithium in porous structure. The study of the lithium limiter experiment on FTU has been completed with a thermal and thermal-mechanical analysis performed by ANSYS 5.2 code.


Corresponding Author:

Apicella Maria Laura

ENEA C.R. Frascati - Via E. Fermi, 45-00044 Frascati-Roma-Italia

- F - Plasma Facing Components.

P4C-F-278 DESIGN, PERFORMANCE AND CONSTRUCTION OF A 2 MW ION BEAM TEST FACILITY FOR PLASMA FACING COMPONENTS

Greuner, Henri, B. Boeswirth, T. Franke, P. McNeely, N. Rust

A new ion beam test facility for the testing of plasma facing components (PFCs) under high heat fluxes is presently under construction at IPP Garching. The aim of this facility is to provide thermal testing capabilities for high heat loaded PFCs with both active water cooling and large outer dimensions. Start of operation is planned to be late summer 2004. Long pulse and cycling heat load tests of WENDELSTEIN 7-X divertor target elements and of complete target modules are the main activities planned for the next years to ensure the successful development, manufacturing and operation of these components. The experience gained from these extensive tests can be used to later adapt the facility to the requirements of effective HHF testing for ITER divertor components. The facility consists of a water-cooled vacuum vessel with a diameter of 1.5 m, a length of 3.7 m and is equipped with 2 ion sources. Initially, only one of the two individually controlled RF ion sources with 1.1 MW maximum beam power will be used for heat load tests in an operating regime between 5 and approximately 65 MW/m² at the target position. The water-cooled ion source allows for high power, long pulse operation facilitating cycling tests of large components. The water cooling system of the facility is designed for testing of components with cooling water consumption of up to 8 l/s and a pressure drop of 15 bar. This paper describes the technical characteristics and operating conditions of the facility. The vacuum- and cooling system, the power supply and control system, the target diagnostic and data acquisition system are described in detail.


Corresponding Author:

Greuner, Henri

Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, D- 85748 Garching, Germany

- F - Plasma Facing Components.

P4C-F-279 SPECTROSCOPIC STUDIES OF HOMOGENEOUS CARBON FLAKES WITH A HIGH DEUTERIUM CONTENT FORMED IN TOKAMAK T-10

Stankevich Vladimir, N.Yu. Svechnikov (1), A.M. Lebedev (1), K.A. Menshikov (1), B.N. Kolbasov (1), L.N. Khimchenko (1), N.M. Kocherginsky (2), D. Rajarathnam (2), Yu. Kostetski (2)

(1) Russian Research Center “Kurchatov Institute”, Kurchatov sq. 1, 123182 Moscow, Russia (2) Faculty of Engineering, National University of Singapore, Singapore

Carbon films redepositing on plasma facing elements in tokamaks attract the attention of investigators mainly as accumulators of hydrogen isotopes, especially tritium. Homogeneous deuterated carbon a-C:D films with a high deuterium content [1], redeposited under deuterium plasma discharges inside the T-10 tokamak vacuum chamber, have been studied using thermogravimetric analysis (TGA), electron spin resonance (ESR) , Fourier-transform infrared (IR) reflection, as well as luminescence spectroscopy in vacuum ultraviolet and visible light ranges, including luminescence excitation by synchrotron radiation in the range 4-18 eV at 300 K. TGA measurements have revealed that a mass loss of up to 30% occurs at 450 C mainly at the expense of carbon and water. ESR spectroscopy results point to a high density of free radicals (~10*19 spins) and a low anisotropy g-factor. As for IR spectroscopy results, we found a significant decrease of deformational vibrations of CHx aromatic groups (out of plane and in plane) in the wavelength range below 1000 cm-1 after baking at 450 C. It indicates that high amount of aromatic groups, including protium was desorbed during baking. On another hand, C-H and C-D stretching modes have shown different behaviour: the peak corresponding to sp3 C-H stretching mode (2925 cm-1) increased after baking at 450 C, possibly due to the decay of OH stretching modes with a subsequent H hopping to C-radicals, while the peak of C-D stretching mode slightly decreased. The photoluminescence effect, observed at 390-530 nm for the first time, could be related to C2p pi-pi transitions within aromatic rings with a subsequent electron-hole recombination. The luminescence was quenched at 450 C – apparently due to the formation of disordered aromatic network within the gap states. The latter possibility is supported by a high value of spins formed on the in-gap defect states. The luminescence excitation spectra of the tokamak films are similar to those of fullerene C60 films for peaks at 3.4, 6.5 and 8.5 eV of a C2p pi- and sigma- character, which appeared to be common for tokamak a-C:D films and C60 systems. [1]. P.V. Romanov, B.N. Kolbasov, V.Kh. Alimov, et al. J. Nucl. Mater. 307-311 (2002) 1294.


Corresponding Author:

Stankevich Vladimir

Russian Research Center “Kurchatov Institute”, Kurchatov sq. 1, 123182 Moscow, Russia

- F - Plasma Facing Components.

P4C-F-280 VACUUM PLASMA-SPRAYED TUNGSTEN ON EUROFER AND 316L - RESULTS OF CHARACTERISATION AND THERMAL LOADING TESTS -

Bolt, Harald, H. Greuner, B. Boeswirth, S. Lindig, W. Kühnlein (1), T. Huber (2), K. Sato (3), S. Suzuki (3)

(1) FZ Jülich, Euratom Association, Forschungszentrum Jülich, B-NZ Heisse Zellen, 52428 Jülich, Germany (2) PLANSEE AG, A-6600 Reutte/ Tirol, Austria (3) Blanket Engineering Laboratory, JAERI, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 JAPAN

Tungsten is being considered as a potential plasma facing material for future fusion devices, primarily due to its low erosion rate and heat resistance. Vacuum plasma spraying (VPS) of tungsten is an effective industrial technique for coating actively cooled plasma facing components made of low activation steels or stainless steel 316L. The coated material would be a potential candidate for first wall components receiving moderate heat load up to 1MW/m². A development programme examined the manufacturing and suitability of W-VPS coatings as plasma facing material on up to 1 MW/m² heat loaded first wall components. Mock-ups made of martensitic steels EUROFER and F82H as well as austenitic steel 316L were coated with 2 mm thick W-VPS layers. Mixed tungsten/steel interlayers were applied to both reduce the residual and thermal stresses at the substrate-coating interface and to improve the adhesion of the coating,. The characterisation of the W-VPS layers included the evaluation of the coating micro structure, the measurement of physical and mechanical properties (thermal conductivity, density, hardness, bending strength and Young’s modulus etc.) and the metallographical examination before and after heat load tests. Thermal loading tests were carried out at the JUDITH facility at the Research Centre Jülich and in parallel in the JEBIS facility at JAERI for the F82H mock-up. Successfully completed cycling tests with heat loads of 2 MW/m² and screening tests up to 2.5 MW/m² confirm the thermomechanical suitability of W-VPS coatings for plasma facing first wall components made of steel.


Corresponding Author:

Bolt, Harald

Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, D- 85748 Garching, Germany

- F - Plasma Facing Components.

P4C-F-285 CAN TOKAMAK DEVICES SURVIVE ELMS DURING NORMAL OPERATION? A SIMULATION STUDY

Konkashbaev, Isak, HASSANEIN, Ahmed

Argonne National Laboratory, 9700 S. Cass Ave., Bldg. 308, Argonne, IL 60439, USA

During normal operation of H-mode, edge-localized modes (ELMs) are serious concern for divertor and nearby plasma-facing components (PFCs) of the next generation tokamaks. During ELMs part of the total plasma energy is released and deposited on divertor surface in duration 0.1-1 ms with a frequency of 1-20 Hz depending on ELM type. The power from scrape-off-layer (SOL) to PFC in ITER-like devices can then increase from 5 MW/m2 to ¡Ö 300-3000 MW/m2. Erosion lifetime strongly depend on ELM power deposited. However, the resulting evaporated material can reach the core and disrupt the plasma. In addition, with higher ELMs frequency, thermal cycling takes place and can result in thermal stresses and fatigue. At high ELM power, the resulting high surface temperature causes vapor cloud formation with similar consequences to disruptions. Vapor shielding decreases energy deposition at the surface but increases radiation flux to nearby components. Metallic PFC can melt and liquid metal flow instabilities occur with mass losses due to both MHD splashing effects and vaporization. In this study a comprehensive two-fluid model is developed to integrate Core and SOL parameters during ELMs with PFC surface evolution (melting, vaporization, vapor cloud hydrodynamics and mixing with plasma particles, and macroscopic spallation) for low and high ELM power using HEIGHTS numerical simulation package. Initial results indicate that high-power, i.e., Giant ELMs in ITER-like machines can cause serious damage to PFCs, may terminate plasma in disruptions, and because of large contamination may affect subsequent plasma operations. A comparison of modeling results with available data from current machines is also addressed.


Corresponding Author:

Konkashbaev, Isak

Argonne National Laboratory, 9700 S. Cass Ave., Bldg. 308, Argonne, IL 60439, USA

- F - Plasma Facing Components.

P4C-F-294 EU R&D ON DIVERTOR COMPONENTS

Merola Mario, W. Daenner (1) M. Pick (1)

(1) EFDA, Boltzmannstr. 2, D-85748 Garching, Germany

Selected also for oral presentation O4A-F-294

Since the last SOFT conference held in Helsinki in 2002, substantial progress has been made in the EU R&D on the divertor components. A number of activities have been completed and new ones have been launched. The present paper gives an update of the works carried out by the EU Participating Team in support of the development of the divertor, which is one of the most challenging components of the next step ITER machine. One of the most impressive achievement was the further development and consolidation of suitable technologies for the production of high heat flux components with both CFC and tungsten armour joined onto a copper alloy heat sink, namely CuCrZr. This long lasting effort culminated with the manufacturing of a near full-scale vertical target prototype, which was high heat flux tested well above the ITER design loads. To ensure competition among the EU industries, different technologies were developed like HIP’ing, brazing and Hot Radial Pressing. Work is now focused on the preparation of EU industry to the unprecedented ITER series production. Another recent achievement was the completion of the post-irradiation testing of divertor mock-ups and material samples. This activity demonstrated that the proposed technologies are able to perform above the requirements, even after being neutron irradiated at 0.2 and 1.0 dpa at 200 C. A substantial design effort was also carried out in collaboration with the ITER IT, with the EU Associations and EU industries. The final outcome is a comprehensive ITER divertor design capable to withstand all the expected loads with minimum manufacturing costs, minimum waste and maximum performances. The manufacturing of a complete set of full-scale divertor prototypes with dummy armour was launched and is progressing according to schedule. After their delivery, the PFCs will be used to validate a software tool, which was recently specifically developed by an EU Association to simulate the hydraulics of the ITER divertor including the draining and drying. In preparation for writing the procurement specification for the ITER vertical target PFCs, an activity is in progress in the EU with the objective of defining workable acceptance criteria for the PFC armour joints. It will be the experimental basis for the final definition of the maximum acceptable defects as well as to assess if and how these defects can be detected by means of non-destructive testing techniques.


Corresponding Author:

Merola Mario

EFDA, Boltzmannstr. 2, D-85748 Garching, Germany

- F - Plasma Facing Components.

P4C-F-305 THERMAL MODELING OF W ROD ARMOR SUBJECTED TO ELMS

NYGREN, Richard E.,

Sandia National Laboratories* has been developing and testing mockups armored with tungsten (W) rods for most of the last decade and Sandia pioneered the initial development of W rod armor for ITER in the 1990's. Plasma facing components (PFCs) with W rod armor have been designed for the ITER-FEAT divertor and are the reference design for the FIRE divertor. Water-cooled heat sinks armored with tungsten rod armor can endure heat fluxes near 25 MW/m2 without cracking, melting or debonding of the armor. Results from earlier have been reported at SOFT and elsewhere and we continue the testing program. We have also developed 3-D thermal models of the W rod-armored PFCs and applied the model to both short pulse testing to simulate ITER ELMs (edge localized modes) and thermal performance in steady state. The basic model is 1/6 of an individual rod with mirror boundaries along the cut sides. Variations include (1) a model with cells graduated to 10 microns in thickness at the top of the rod to follow the shallow and rapid thermal penetration of very high heat laods (1-5 GW/m2) necessary to model the high energy density from ELMS, (2) a rod based upon a W rod cluster to be tested in the DiMES probe in DIII-D, and (3) aggregates of several rods to study the effects of uneven heating on rod groups. The mesh was created in PATRAN and the model is run using ABAQUS 6.3.1. Heat loss due to thermal radiation is included as are temperature dependent properties of materials. The heat transfer coefficient at the water cooling boundary follows a specified boiling curve that depends on the coolant temperature, pressure and the presence of any heat transfer enhancement such as a twisted tape insert. This paper briefly describes the model and focuses on the thermal modeling of ELMs. For example, the threshold energy density for melting was studied for various values of steady state heat flux (i.e., starting surface temperature). Also, in the case of repeated ELMs that caused some melting, the model accounts for the enthalpy stored in melting and its effect on the thermal response during ELMS repeated with frequencies of 1-10 Hz. Applications of the model to a W rod DiMES probe to various heat loads and the thermal performance of W-rod-armored mockups are also mentioned. *Sandia is a multi-program laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy under Contract DE-AC04-94AL85000


Corresponding Author:

NYGREN, Richard E.

Sandia National Laboratories MS1129, PO Box 5800, Albuquerque NM 87185 USA

- F - Plasma Facing Components.

P4C-F-315 CRITICAL HEAT FLUX TESTING ON SCREW COOLING TUBE MADE OF RAFM-STEEL F82H FOR DIVERTOR APPLICATION

Koichiro EZATO, Satoshi Suzuki, Masayuki Dairaku, Kazuyoshi Sato, and Masato Akiba

As part of development of Plasma-Facing Components (PFCs) for fusion machines, JAERI has been developing high performance cooling tubes with pressurized water flow. Along this line, a cooling tube with a helical triangular fin on its inner surface has been proposed recently for application to a DEMO reactor. Since the fin can be machined by a simple mechanical threading, this tube is called as a screw tube. In our previous experiments, it was reported that heat removal performance of the screw tube made of pure Cu is twice as high as that of a smooth tube. In DEMO designs, Reduced Activation Ferritic Martensitic (RAFM) steel such as F82H is one of the candidate materials for a cooling structure of PFCs instead of Cu-alloy. As thermal conductivity of F82H is about ten times smaller than that of Cu-alloy, this study is intended to examine heat removal capability of the screw tube made of F82H. For this purpose, we have carried out Critical Heat Flux (CHF) testing under one-sided heating conditions by using a hydrogen ion beam. The test samples are the screw tubes with M10 of 1.5-mm-pitch. The M10 threads are directly shaped in F82H and OFHC-Cu tubes with the outer diameter of 12 mm. The minimum wall thickness of each tube is 1 mm. Inlet temperature and local pressure of cooling water are room temperature and 1MPa. Flow velocity ranges from 2 to 12 m/s. Incident heat flux at the sample position has a Gaussian profile and its maximum value ranges from 8 to 48MW/m2. Incident CHF (ICHF) of the F82H screw tube is reduced to about half of the OFHC-Cu tube. For instance, ICHF of the F82H tube is 13 MW/m2 at the flow velocity of 4 m/s and that of the Cu tube is 25MW/m2. Numerical analyses show that the critical heat flux at the inner surface of the cooling tube is almost the same for both tube materials. This means that the incident heat flux is highly concentrated for the F82H cooling tube because of its low thermal conductivity. It is also found that the ratios of the heat flux at the inner surface of the cooling tube to the incident heat flux are around 1.6 for the F82H tube and 1.1 for the Cu tube. Based on these results it turns out that application of F82H to PFC cooling structures needs to enhance dispersion of the incident heat flow, for example, to be covered with armor material with higher heat conductivity such as tungsten.


Corresponding Author:

Koichiro EZATO

Japan Atomic Energy Research Institue, 801-1 Mukoyama, Naka-machi, Naka-gun, Ibaraki-ken 311-0193, Japan

- F - Plasma Facing Components.

P4C-F-328 STATUS OF HE-COOLED DIVERTOR DEVELOPMENT FOR DEMO

NORAJITRA , PRACHAI, GINIYATULIN, RADMIR (b) IHLI, THOMAS (a) KRAUSS, WOLFGANG (a) KRUESSMANN, REGINA (a) KUZNETSOV, VLADIMIR (b) MAZUL, IGOR (b) OVCHINNIKOV, IVAN (b)

(a) FORSCHUNGSZENTRUM KARLSRUHE, P.O. BOX 3640, D-76021 KARLSRUHE, GERMANY (b) D.V. Efremov Institute, Scientific Technical Centre “Sintez”, 196641 St. Petersburg, Russia

The helium-cooled divertor is considered a suitable option for fusion power plants, as it is compatible with likewise He-cooled blanket systems. It is also recommended for those blankets, where water-cooling of in-vessel components would lead to considerable concerns in terms of safety (e.g. steam-beryllium reaction with H production). Furthermore, it allows for a relatively high gas outlet temperature, i.e. a high thermal efficiency of the power conversion systems. He-cooled modular divertor concepts with integrated flow promoters in the form of a pin (HEMP) or slot (HEMS) array are being developed at the Forschungszentrum Karlsruhe within the European Power Plant Conceptual Study. In parallel, an alternative design HEMJ is under investigation, which is based on multiple jet impingement cooling without flow promoter. The modular design helps to reduce thermal stresses. Tungsten is considered the most promising material to withstand the high heat load, due to its high melting point, high thermal conductivity, and low thermal expansion. The proposed HEMP/S divertor concept employs small W tiles of quadratic or hexagonal shape, which are brazed to a thimble structure of W alloy below. A flow promoter of W is brazed underneath each thimble to increase the cooling surface. The structure is made of high-temperature ODS RAFM. The development and optimisation of the divertor concepts require a close link of and iterative approach comprising the main issues of design, analyses, materials, fabrication technology, and experiments. Predicting the temperatures and stresses by means of computational fluid dynamics and finite element computer codes is indispensable to ensure that the engineering design limits are not exceeded. The divertor working temperature window is restricted by ductile-brittle transition temperature at the lower and recrystallisation temperature of the W structure at the upper limit. Enlarging this temperature window is a challenging task of materials development. For manufacturing divertor components of W and W alloy, EDM, ECM, laser, and PIM are considered promising methods. Experiments on W/W and W/steel joining were performed successfully at Efremov. A helium loop will be built at Efremov this year for high-heat-flux (HHF) integral testing of the divertor design variants. An electronic beam facility there allows for the HHF simulation of 10 MW/m² at least. The status of development in the above areas of work shall be outlined in this report.


Corresponding Author:

NORAJITRA , PRACHAI

FORSCHUNGSZENTRUM KARLSRUHE, P.O. BOX 3640, D-76021 KARLSRUHE, GERMANY

- F - Plasma Facing Components.

P4C-F-333 NUMERICAL AND EXPERIMENTAL STUDY OF DEMO HE-COOLED DIVERTOR TARGET MOCK-UPS

Rumyantsev Mikhail, Kuznetsov Vladimir (1) Ovchinnikov Ivan (1) Filatov Vladimir (1) Janeschitz Guenter (2)

(1) The D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, 3 Doroga na Metallostroy, Promzona "Metallostroy", St. Petersburg 196641, Russian Federation (2) Forschungszentrum Karlsruhe, P.O Box 3640, D 76021 Karlsruhe, Germany

The helium cooled divertor with modular conception is proposed for the fusion reactor DEMO. It is planned to use helium (600oC, 10MPa) for divertor cooling, where a high surface heat flux (up to 15MW/m2) must be removed from vertical targets. For the best cooling of modules the smooth surfaces which interacts with helium flow are increased with help of additional pins or slots. The more area of the surface the better cooling, but at the same time pumping power may be increased, i.e. useful power will be decreased. Selection of an optimal geometry of the cooling surface is current problem of project. Optimization of cooling surface at expected reactor conditions was performed with help numerical code. The following options of modules are presented: - a first option is modules with straight radial slots on cooling surface. An optimization of height, width and numbers of slots was performed for this option. Influence of increasing in target dimensions and helium mass flow rate on thermal state were analyzed here, too. By results of these analyses a best geometry of slots was selected. Thermal stresses were examined for this geometry. - a second geometry option is module with pins. Also paper describes experimental gas puffing facility GPF2, where the analyzed mock-ups were tested. Reversed heat flux was proposed for the test facility which consists from 2 loops, helium (600oC, 10 MPa), and water (RT, 5 MPa). In this case, the mock-up with the hot helium flow is intensively cooled by water from the plasma-facing side. This approach give possibility to study: different mock-up designs can be compared with respect to pressure drop and cooling efficiency. The GPF2 works with helium pulses that are longer by 2-3 orders of magnitude to reach stationary flows in the mock-up. The simulation approach, methods and data processing are described. Test results obtained for different mock-ups at fluxes of 5-15 MW/m2 are presented and discussed. In addition, behavior of the mock-ups in GPF2 was simulated by CFD and the calculation results are compared with the experiments.


Corresponding Author:

Rumyantsev Mikhail

Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA) 3 Doroga na Metallostroy, Promzona "Metallostroy", Metallostroy, St. Petersburg 196641

- F - Plasma Facing Components.

P4C-F-343 MEASUREMENTS OF H/D DIFFUSIVITY IN AND SOLUBILITY THROUGH TUNGSTEN IN THE TEMPERATURE RANGE OF 600 C TO 800 C

Aiello Antonio, Gianluca Benamati (1) Andrea Ciampichetti (2)

(1) ENEA C.R. Brasimone - Bacino del Brasimone, 40032 Camugnano (BO)- Italy (2) Politecnico di Torino – DENER – Corso duca degli Abruzzi 24, 10129 Torino - Italy

Estimation of tritium permeation through the plasma facing materials is an important issue for fusion reactors. Because of their refractory nature and good thermal properties, tungsten and tungsten-alloys are considered to be alternatives to graphite as plasma-facing materials for ITER. Tungsten has a very high threshold for sputtering as well as a high melting point. Tungsten is expected to be used in areas where the energy of plasma particles can be kept well below the sputtering threshold, removing the plasma impurities problem associated with the use of this material. Experimental campaigns on the characterisation of hydrogen isotopes transport and solubility parameters of tungsten have been conducted by several laboratories, but results are often in disagreement. An extensive experiment have been conducted in ENEA using a permeation device named PERI 2. Permeation tests were carried out using membranes of tungsten separating two volumes in the PERI II apparatus, an high pressure volume and an high vacuum volume. Hydrogen gas was charged in the high pressure side and measuring the pressure evolution in the low pressure side it was possible to determine the hydrogen transport parameters in the sample. A permeated gas analysis by means of a mass quadrupole was also performed. Experiments conducted in the temperature range between 350 C and 800 C indicated the low permeability of tungsten. The obtained results together with the experimental procedure adopted are herein presented and discussed.


Corresponding Author:

Aiello Antonio

ENEA C.R. Brasimone - Bacino del Brasimone - 40032 Camugnano (BO) Italy

- F - Plasma Facing Components.

P4C-F-367 TESTING OF ACTIVELY COOLED MOCK-UPS IN SEVERAL HIGH HEAT FLUX FACILITIES – AN INTERNATIONAL ROUND ROBIN TEST

Roedig, Manfred, I. Bobin-Vastra (2), S. Cox (3), F. Escourbiac (4), A. Gervash (5), A. Kapoustina (1), W. Kuehnlein (1), V. Kuznetsov (5), M. Merola (6), R. Nygren (7), D.L. Youchison (7)

(1) FZJ, Jülich, Germany (2) Framatome, Le Creusot, France (3) JET, Abingdon, UK (4) CEA-DRFC, Cadarache, France (5) Efremov Inst., St. Petersburg, Russia (6) EFDA Close Support Unit, Garching, Germany (7) Sandia Nat. Lab. Albuquerque, USA

In next step fusion devices like ITER, the first wall and divertor component will be exposed to high heat fluxes up to more than 10MW/m2. Hence a large R&D effort is being carried out to develop suitable high heat flux components. In order to test components under operational relevant conditions, several electron or ion beam facilities have been used worldwide. Up to a certain degree these machines are comparable. They consist of a beam generator, a beam sweeping system, a vacuum test chamber, and a number of diagnostic devices. But some machine parameters like the beam generation and system, calibration techniques and diagnostic systems are quite different. In order to assess the influence of these differences on testing results, a round robin test has been performed on five electron beam facilities a few years ago. The aim was, to study the influence of specific layouts of these machines on the results of high heat flux tests. The comparison was carried out by high heat flux testing of actively cooled CFC samples at identical target loading conditions, and the surface temperatures were used as a criterion for the assessment of the results. This former test campaign was not planned as a round robin test from the very beginning, and some questions stayed open. Hence a new test campaign has been initiated by the EFDA team. For this new test campaign special actively-cooled mock-ups have been produced, and testing parameters have been planned with respect to the testing facilities involved. In the beginning, only electron beam facilities were intended to take part. But later the program was extended, and tests at the JET neutral beam injector testbed have been included. In this testing campaign, a set of actively cooled CFC monoblock mock-ups has been loaded in the different facilities at comparable power densities. The temperature response during these loadings on the surface (IR cameras, pyrometers) and inside the mock-ups (thermo couples) has been registered and used as a criteria for comparison. Furthermore finite element calculations have been carried out for the temperature fields at different power densities. Most of the surface temperatures were found in a relatively narrow scatter band. Only one of the electron beam facilities shows somewhat higher temperatures compared to the other machines. At higher power densities, the JET-NBI data are on the upper side of the scatter band. This may be explained by the peaked beam profile in this machine.


Corresponding Author:

Roedig, Manfred

Forschungszentrum Juelich, 52425 Juelich, Germany

- F - Plasma Facing Components.

P4C-F-376 STUDY OF TECHNOLOGICAL AND MATERIAL ASPECTS OF HE-COOLED DIVERTOR FOR DEMO REACTOR

GERVASH Alexander, R.Giniyatulin 1, W.Krauss 2, A.Makhankov1, I. Mazul 1, P.Norajitra 2

1 Efremov Research Institute, 196641 St. Petersburg, Russia 2 Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe, Germany

Study of technological and materials aspects of He-cooled divertor for DEMO reactor Further development of a helium-cooled divertor concept for fusion reactor like DEMO depends strongly on the progress in selection of suitable materials and technologies of their manufacturing and joining as well. Design proposes small tungsten tiles joined to a thimble structure made of tungsten alloy. High-temperature ferritic steel is proposed for supported structure. Presented paper gives recent results of technological trials to manufacture helium-cooled divertor module. To enhance inner thimble surface with pin and slot array several fabrication techniques were checked. In particularly, the numbers of prototypes were produced by electric discharge machining (EDM), electrochemical milling (ECM), laser ablation and chemical vapour deposition (CVD) methods. The accuracy of required dimensions, cost reasons, possibilities of serial production of such prototypes were compared and discussed. To select most suitable W-alloy for the thimble production paper gives the main results of comparative testing of candidates (W-single crystal, W-1%La2O3, CVD-tungsten, W-Cu composite, forged/rolled sintered tungsten). Investigating the problem of joining W-thimble to ferritic steel structure the number of W/ferritic steel specimens produced by e-beam welding, diffusion bonding, high temperature brazing and locking with cast copper were manufactured and tested. The main results are presented and discussed. Summarizing presented data authors inform about nearest further steps of their investigation.


Corresponding Author:

GERVASH Alexander

Efremov Research Instutute, 196641 Saint Petersburg, Russia

- F - Plasma Facing Components.

P4C-F-386 DESIGN AND THERMAL PERFORMANCE OF SURFACE-BOLTLESS MECHANICALLY ATTACHED MODULE FOR DIVERTOR PLATE OF LHD

Kubota Yusuke, Masuzaki Suguru(1), Morisaki Tomohiro(1), Tokunaga Kazutoshi(2), and Noda Nobuaki(1)

(1)National Institute for Fusion Science, Oroshicho 322-6, Toki 509-5292, Japan (2)Research Institute for Applied Mechanics, Kyushu Univ.,Kasugai 816-8580, Japan

Abstract: To prevent plasma collapse during plasma confinement experiment, especially in steady state operations, suppressions of outgassing and high Z impurity emission from the first walls are required strongly. According to the requirements, high performance mechanically attached module has been developed as the next type of divertor plate for large helical device(LHD). The most advantage of the module is not to have any bolts for fix on the armor tile surface to avoid high Z impurity emission different from the previous one1) used in the LHD since the third campaign(FY 1999). The new one consists of two armor tiles made of iso-graphite, a thin super graphite sheet, and a SS cooling pipe. A couple of armor tiles sandwiches the cooling pipe through a super graphite sheet to improve thermal contact between two materials, and fixed tightly with only two TZM(an alloy of molybdenum) bolts. This simple structure without a copper heat sink allows smooth heat flow from the tile surface to the cooling pipe different from the previous one. Using a test facility ACT2) with a 100kW electron gun, steady high heat flux tests up to 1.2 MW/m2 were carried out for the new one without any trouble although the previous one was limited about 0.3 MW/m2. Moreover, outgassing from the new one during high heat flux tests up to 0.5 MW/m2 decreased to about one-third of that of the previous one. Low outgassing of the new one may originate in the simple structure without a copper heat sink and use of super graphite sheet with an excellent thermal conductivity. Thermal fatigue test up to 500 cycles under steady heat flux of 1 MW/m2 for the new one is scheduled to carry out using ACT before full-scale application to divertor plates of LHD. The design, thermal performance, thermal analysis by 3D CAD, and outgassing of the newly developed mechanically attached module are presented. References 1) N.Noda, S.Sakamoto, Y.Kubota et al., J.Plasma Fusion Res. SERIES, Vol.3(2000)180. 2) Y.Kubota, N.Noda, A.Sagara et al., Fusion Engineering and Design 56-57(2001)205.


Corresponding Author:

Kubota Yusuke

National Institute for Fusion Science, Oroshicho 322-6, Toki 509-5292, Japan

- F - Plasma Facing Components.

P4C-F-413 MANUFACTURE OF BLANKET SHIELD MODULES FOR ITER

Lorenzetto Patrick, Boireau B. (2), Boudot C. (2), Bucci Ph. (3) Furmanek A. (1), Ioki K. (4), Liimatainen J. (5), Peacock A. (1), Sherlock P. (6), Tähtinen S. (7)

(1) EFDA CSU Garching, Germany. (2) FRAMATOME ANP, Le Creusot, France. (3 ) CEA, Grenoble, France. (4) ITER IT, Garching, Germany. (5) Metso Powdermet, Tampere, Finland. (6) NNC Ltd, Knutsford, England. (7) VTT Industrial Systems, Espoo, Finland.

Selected also for oral presentation O4A-F-413

The ITER Blanket-shield concept is a modular configuration mechanically attached onto the vacuum vessel and consists of Limiter modules and Primary First Wall (PFW) / Shield modules. The latter consist of a water-cooled 316L(N)-IG Stainless Steel (SS) Shield Block and separable PFW panels mechanically attached onto the Shield Block. The PFW panels consist of a bi-metallic structure with a 316L(N)-IG SS backing plate and a Copper (Cu) alloy heat sink layer. There are two Cu alloy candidates: Dispersion Strengthened Cu-Al25 and Precipitation Hardened CuCrZr alloys. Beryllium (Be) tiles are joined to the Cu alloy heat sink as plasma facing material. A Research and Development programme for the ITER Blanket-shield has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups and scale-one prototypes of Shields and PFW panels. These prototypes aim at demonstrating the fabricability of the components. Two methods are being considered in Europe for the manufacture of the Shield blocks. The first method is based on conventional fabrication techniques using drilling, machining and welding. The second method uses a more advanced technique based on Hot Isostatic Pressing (HIPping) of 316L(N)-IG SS powder and 316L(N)-IG SS solid parts. One Shield prototype made from powder HIPping is already complete and a second prototype made from mixed powder and solid HIPping is under fabrication. Two methods are also being considered in Europe for the manufacture of the bi-metallic structure of the PFW panels: solid and powder HIPping. With solid HIPping, the 316L(N)-IG SS backing plate, the Cu alloy plates and the 316L(N)-IG SS tubes are joined together with one single HIP cycle. With powder HIPping, a first HIP cycle is used to consolidate the 316L(N)-IG SS powder with embedded 316L(N)-IG SS tubes. A second HIP cycle is then performed for consolidating and joining the CuCrZr powder. Beryllium tiles are then joined by HIPping or brazing; high temperature HIPping or furnace brazing for PFW panels with CuAl25 alloy heat sink material, and low temperature HIPping or inductive brazing for PFW panels with CuCrZr alloy heat sink material. Three panel prototypes have already been completed. Two more are under fabrication. This paper describes the main fabrication steps for the above Shield and PFW panel prototypes.


Corresponding Author:

Lorenzetto Patrick

EFDA CSU Garching, Boltzmannst. 2, D-85748 Garching, Germany

- F - Plasma Facing Components.

P4C-F-426 THE MAST IMPROVED DIVERTOR

Darke Andrew, R J Hayward G F Counsell K Hawkins

The Mega Amp Spherical Tokamak (MAST) at Culham is one of the leading world machines studying the spherical tokamak (ST) concept. At the time of the initial construction in 1998 little was known about the sort of divertor structures that would be required in an ST. The machine was therefore provided with relatively rudimentary structures that were designed mostly to protect important components from the hot plasma. While these have served the machine well it was accepted that they might not be suitable when operating MAST to its full potential. The years of experience of operating MAST have led to the design, manufacture and now installation of a new divertor, the MAST Improved Divertor or MID, that should be able to cope with the full performance of the machine. The design is based on imbricated (fan-shaped) rings of tiles at the top and bottom of the machine for the outer strike points, giving an excellent compromise between power handling and diagnostic access, with substantial new centre column strike point armour and a shaped plate in between. High purity graphite is chosen as the plasma facing material in preference to CFC since in this case it has a better balance of performance and cost. The lower imbricated ring is insulated in alternate sectors for studies of divertor biasing and extensive diagnostics and additional inboard gas injection are included. This work was funded jointly by the United Kingdom Engineering and Physical Sciences Research Council and by EURATOM.


Corresponding Author:

Darke Andrew

EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB, UK

- F - Plasma Facing Components.

P4C-F-429 OXYGEN IMPURITY EFFECTS ON HYDROGEN ISOTOPE RELEASE FROM PLASMA CHEMICAL VAPOR DEPOSITION BORON COATING

Makoto Oyaidzu (1), Makoto Oyaidzu(1), Akira Yoshikawa(1), Yoshihiro Onishi(1), Hhiromi Kimura(1), Yasuhisa Oya(2), Masao Matsuyama(3), Akio Sagara(4), Nobuaki Noda(4), and Kenji Okuno(1)

(2)RI Center, The University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-0032, Japan (3)HIRC, Toyama University, Gofuku 3190, Toyama 930-8555, Japan (4)NIFS, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan

The reduction of impurities, particularly oxygen, in plasma has been one of the key issues for the large tokamaks, since plasma impurities cause dilution of fuel particles. To reduce oxygen impurities in plasma, boronization has been developed. As a result of boronization, oxygen impurities have been dramatically trapped in boron coating and boron coatings contained oxygen are formed on the PFM. During boronization and/or PCVD processes, hydrogen originated from borane gases is mixed with the boron coating. Moreover, in D-T fusion reactors, hydrogen isotopes, deuterium and tritium, are implanted into boron coatings. Therefore it is important to elucidate hydrogen isotopes release behavior from boron coating contained oxygen from the viewpoint of tritium retention, hydrogen recycling, and characterization. In the present study, oxygen impurity effects on hydrogen isotopes release behavior from boron coating contained oxygen prepared using PCVD technique was studied by XPS and TDS. Boron coatings contained oxygen deposited on silicon substrates using a decaborane gas (vol. 20%) diluted with helium (Vol. 50%) and oxygen (vol. 30%) gases by PCVD were used as samples. The atomic composition ratio in samples was estimated by XPS. TDS technique was also applied to evaluate release behavior of hydrogen isotopes, namely hydrogen mixed in the samples during preparation and deuterium implanted into the sample. XPS measurements showed that the atomic ratio of boron to oxygen is almost unity. After TDS measurement after preparation, hydrogen release spectrum was found to consist of a shoulder at around 400 K and a peak at around 530 K. In the hydrogen release from pure boron coating, the temperature of low and high temperature side is around 450 and 650 K, respectively [1]. As the result of numerical analysis, it was found that the amount of hydrogen retention in boron coating contained oxygen is almost five times as much as that in the pure boron coating and the hydrogen release spectrum was divided into two peaks, namely peaks A and B. Therefore, it is suggested that oxygen existing in boron coating lead to increase hydrogen retention and hydrogen release temperature region wholly shifts to low temperature side. In the presentation, oxygen impurities effects on hydrogen isotopes release behavior from boron coating will be discussed in detail, taking into account for results of deuterium release behavior. [1] H. Kodama, et. al., J. Nucl. Mater., in press.


Corresponding Author:

Makoto Oyaidzu (1)

(1)RRL, Faculty of Science, Shizuoka University, 836 Oya, Shizuoka 422-8529, Japan

- F - Plasma Facing Components.

P4C-F-431 IMPLANTATION TEMPERATURE DEPENDENCE ON DEUTERIUM BEHAVIOR IN HIGHLY ORIENTED PYROLITIC GRAPHITE

Hiromi Kimura, Yasutomi Morimoto Hiroshi Kodama Makoto Oyaidzu Akira Yoshikawa Tsuyoshi Takeda Kenji Okuno

Radiochemistry Research Laboratory, Faculty of Science, Shizuoka University, Ohya, Shizuoka 422-8529, Japan

For evaluation of tritium safety in fusion reactors, it is very important to know how tritium implanted into plasma facing materials (PFMs) behaves chemically. In PFMs, tritium chemically interacts with damages induced by energetic atom and/or irradiation. Chemical interactions are governed by two processes, thermal processes, such as thermal release of deuterium and the thermal annealing of graphite structure, and high energy process. In our previous study using TDS and XPS techniques, recovery of graphite structure damaged by ion implantation was observed above 573 K. In addition, it was found that deuterium implantation at lower temperature than RT is very useful to investigate those processes. In present study, dynamics of deuterium implanted into graphite in various temperatures is studied from the kinetic point of view, using TDS and XPS techniques. The sample used in the present study was a HOPG crystal purchased from Pechiney Co. Ltd. Deuterium ions were implanted into HOPG with an energy of 1.0 keV D2+, a flux of 1.0×1018 D+ m-2 s-1, and a fluence of 6.4×1021 D+ m-2 at various temperatures region from 173 to 773 K. To investigate implantation temperature dependence on deuterium retention, the sample after implantation was heated up to ~1400 K with heating rate of 0.5 K s-1. To estimate activation energy of deuterium desorption, the sample was heated with heating rate in the range from 0.083 to 1.0 K s-1. From the results of TDS experiments after deuterium ion implantation from 173 to 773 K, deuterium implanted into HOPG was released from ~800 K. However, the deuterium retention was decreased as implantation temperature increased. The activation energies in the higher temperature than 237 K were determined to be ~2.3 eV, which were approximately coincident with the literature values [1, 2], while the activation energy at 173 K was estimated to be 4.4 eV. It was consider that this value also would include movement of carbon. In present study, taking into account for carbon mobilization in HOPG, dynamics of deuterium implanted into graphite will be discussed in detail. [1] T. Tanabe, et. al., J. Nucl. Mater., 179-181 (1991) 231-234. [2] K. Ashida, et. al., J. Nucl. Mater., 128-129 (1984) 792-797.


Corresponding Author:

Hiromi Kimura

Radiochemistry Research Laboratory, Faculty of Science, Shizuoka University, Ohya, Shizuoka 422-8529, Japan

- F - Plasma Facing Components.

P4C-F-445 MANUFACTURING AND TESTING IN REACTOR RELEVANT CONDITIONS OF BRAZED PLASMA FACING COMPONENTS OF THE ITER DIVERTOR

Grattarola Marco, M. Bisio (1) V. Branca (1) M. Di Marco (2) A. Federici (1) G. Gualco (1) P. Guarnone (1) U. Luconi (2) M. Merola (3) C. Ozzano (1) G. Pasquale (2) S. Rizzo (1) F. Varone (1)

(1) Ansaldo Ricerche s.r.l., C.so Perrone 25, I-16161 Genova, Italy (2) FN s.p.a. SS 35 bis dei Giovi km 15, I-15062 Bosco Marengo (AL), Italy (3) EFDA CSU Garching,Boltzmannstr. 2, D-85748 Garching, Germany

Selected also for oral presentation O4A-F-445

A fabrication route based on brazing technology has been developed for the realization of the High Heat Flux Components for the ITER Vertical Target and Dome-Liner. The divertor vertical target is based on a monoblock design with CfC and tungsten armour in the lower straight part and in the upper curved part, respectively. The cooling tubes are made of precipitation hardened copper alloy CuCrZr. A pure copper interlayer between the heat sink and the armour mitigates the joint interface stress due to the thermal expansion mismatch between the CuCrZr and the armour material. The plasma facing units of the Dome component are based on a tungsten flat tile design with hypervapotron cooling. The heat sink is taken from a bimetallic plate composed of precipitation hardened copper alloy CuCrZr and stainless steel joined together by means of the explosion bonding process. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production (patent pending), has been used for the CFC/Cu joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the CuCrZr alloy after brazing. The yield stress, the ultimate tensile strength and the average grain size measured after the brazing thermal cycle are 263 MPa, 404 MPa and 28 microns respectively. The CuCrZr/steel explosion bonding has been qualified by means of an extensive metallurgical and mechanical test program. Non destructive examination methods based on ultrasonic techniques have been developed and qualified to inspect the brazed joints. The fabrication route of plasma facing components for the ITER Vertical Target and Dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions. Flat tile and monoblock mock-ups with both CFC and tungsten armour material have been successfully tested by thermal fatigue tests with incident heat flux higher than 15 MW/m2.


Corresponding Author:

Grattarola Marco

Ansaldo Ricerche s.r.l., C.so Perrone 25, 16161 Genova, Italy

- F - Plasma Facing Components.

P4C-F-447 DEVELOPMENT OF THE PLASMA FACING COMPONENTS FOR THE DOME-LINER COMPONENT OF THE ITER DIVERTOR

Luconi Umberto, V. Branca (1) E. Cordano (1) M. Di Marco (2) A. Federici (1) M. Grattarola (1) M. Merola (3) C. Ozzano (1)

(1) Ansaldo Ricerche s.r.l., C.so Perrone 25, I-16161 Genova, Italy (2) FN s.p.a. SS 35 bis dei Giovi km 15, I-15062 Bosco Marengo (AL), Italy (3) EFDA CSU Garching,Boltzmannstr. 2, D-85748 Garching, Germany

On the basis of the design and the specification of the Dome-Liner elaborated by EFDA, a manufacturing route has been developed and proved by means of the fabrication and testing of several samples and mock-ups. The dome is supported on four posts and is protected with a 10-mm thick tungsten tile armour. These tungsten tiles are joined onto a plate (heat sink) made of precipitation hardened copper-chromium-zirconium alloy (CuCrZr). Due to constraints posed by the electromagnetic loads, the thickness of the CuCrZr plate can not exceed 10 mm; as a consequence the remaining part of the heat sink shall be made of stainless steel. Therefore the heat sink is obtained from a CuCrZr/AISI 316 L bimetallic plate realized by explosion bonding. The hypervapotron cooling channels are obtained by machining the plate from the steel side crossing the CuCrZr/AISI 316 L explosion bonded joint, then the channels are closed by welding a steel rear closure plate. Both the explosion bonded joint and the rear plate welding must comply the strict ITER vacuum tightness requirements. The brazed joint between the tungsten tiles and the CuCrZr heat sink has been qualified by means of thermal fatigue tests on small-scale mock-ups performed by the Efremov Institute (St. Petersburg, Russia) in reactor relevant conditions (1000 cycles at 15 MW/m2). The CuCrZr/AISI 316L explosion bonding process utilized to join the front CuCrZr plate to the rear steel backing has been qualified by means of an extensive metallurgical and mechanical test program according to the specification provided by EFDA. The test program included the execution of hot pressure helium leak tests, cyclic pressure and burst tests on several relevant mock-ups. The mock-ups sustained He leak tests at 250 C and 6 MPa (pressurized He inside the components), before and after the repeated pressure test at 8 MPa for 100 cycles. After the previous tests, the mock-ups sustained an internal pressure higher than 500 bar during the bust tests. The dimensional stability during the fabrication route has been investigated by means of the realization of a relevant curved component that has been dimensionally tested after the completion of each step of the manufacturing route. The results of the experimental activity are presented and discussed in this paper.


Corresponding Author:

Luconi Umberto

Ansaldo Ricerche s.r.l., C.so Perrone 25, 16161 Genova, Italy

- F - Plasma Facing Components.

P4C-F-474 THERMAL PROPERTY CHANGES OF ERODED AND REPETITIVELY LOADED CFC

Schmalz, F., D.S. D’Hulst, J.G. van der Laan

NRG, P.O.Box 25, NL 1755 ZG PETTEN

The power handling characteristics of CFC tiles may change after long term exposure to plasmas and repetitive power fluxes. In ITER such effects may appear in an early operation phase. Samples taken from graphite and CFC tiles exposed to large plasma fluences in various EU tokamaks are subjected to short pulse high heat fluxes, chosen below the material ablation limit. Their transient heat load responses are evaluated by applying Nd:YAG laser pulses of 0.2 up to 10 ms duration, and measuring the surface temperature response with a fast IR pyrometer. Specimen temperatures are varied between 300 and 1000 K and testing is performed in vacuum. Power densities are chosen to have peak temperatures in the range of 1200 to 1700 K. In order to quantify the effect of morphology changes, un-exposed specimens are tested at similar heat loads. The materials shorter pulse response is more sensitive to near surface properties (< 0.1 mm). In selected cases thermal diffusivity is measured in addition. ITER CFC grades are subjected to power fluxes that approach marginal erosion conditions in ELMs. The possible deterioration of the material under repetitive pulses (up to 10^3 pulse at energy densities up to 1.2 MJ/m²) is investigated. The focus is placed on particular microstructural effects, like matrix-fibre detachment and enhanced erosion for energy fluxes lower than those causing ‘brittle damage’. The paper will report on the detailed experimental procedure, thermal property determination and the results of the specimen damage evaluations.


Corresponding Author:

Schmalz, F.

NRG, P.O.Box 25, NL 1755 ZG PETTEN

- F - Plasma Facing Components.

P4C-F-475 STUDIES OF HEAT CONDUCTION IN LIQUID LITHIUM CAPILLARY POROUS SYSTEM

Azizov Englen, A.G. Alekseyev, V.B. Lazarev, S.V. Mirnov V.A. Evtikhin, I.E. Lyublinski, A.V. Vertkov

1 TRINITI, Troitsk, Moscow reg., 142190 Russia 2 State Enterprise «Red Star» - «Prana-Center» Co, Moscow, Russia

An application of liquid metals to the design of divertor plates is considered now as a promising approach for the future fusion devices [1-3]. Liquid Lithium (LL) capillary-porous system (CPS) proved to be one of the most reliable technical solutions, which is able to withstand up to 20 MW/m2 power loading. A number of rail limiters with LL CPS were developed and tested in the T-11M tokamak starting from 1997 [4-6]. Recent progress in this R&D program includes a quasi-stationary limiter providing the saturation of temperature gradient in CPS for the plasma discharges longer than 0.1 sec [6]. One of the most essential problems of CPS design for the steady state heat loading is providing a reliable thermal contact between liquid and solid parts of the heat sink, which might be subjected to an occasional destruction under the thermal cycling and heavy local power loading. The last version of LL CPS limiter installed into the T-11M tokamak has demonstrated an ability to improve the thermal contact after the long-term uniform thermal heating, and even to recover an initial situation after the local damage. Some other aspects of heat conduction in the LL CPS under the powerful loading are discussed also. References 1. L.G. Golubchikov, et al. J. Nucl. Mater., v.233-237 (1996), 667-672. 2. V.A. Evtikhin, I.E. Lublinski, et al., Proc. 16th Int. Conf. on Fusion Energy, Montreal, 7-11 Oct. 1996, Fusion Energy 1997, IAEA, Vienna, 1997, vol. 3, 659-665. 3. V.A. Evtikhin, I.E. Lyublinski, A.V. Vertkov, et al., Fusion Energy 1998, IAEA, Vienna, 1999, vol. 4, p. 1039-1313. 4. V.B. Lazarev, E.A. Azizov, A.G. Alekseyev, et al. 26th EPS Conf. on Contr. Fusion Plasma Physics, ECA, 1999, vol. 23I, pp. 845-848. 5. A.M. Belov, V.B. Lazarev, A.G. Alekseyev, S.V. Mirnov, I.N. Makashin, 28th EPS Conf. on Controlled Fusion and Plasma Physics, Madeira, Portugal, 2001. 6. V.B. Lazarev, et al, 30th EPS Conference on Contr. Fusion and Plasma Phys., St.Petersburg, 7-11 July 2003 ECA, Vol. 27A, P-3.162.


Corresponding Author:

Azizov Englen

TRINITI, Troitsk, Moscow reg., 142190 Russia

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-4 DESIGN AND DEVELOPMENT TOWARDS A PARALLEL WATER HYDRAULIC WELD/CUT ROBOT FOR MACHINING PROCESSES IN ITER VACUUM VESSEL

wu huapeng, Pekka Pessi, Heikki Handroos, Janne Kovanen, (2)Lawrence Jones

Department of Mechanical Engineering Lappeenranta University of Technology 53851 Lappeenranta, Finland (2)EFDA Close Support Unit, IPP-Garching, Boltzmannstrasse 2, D-85748, Germany

Selected also for oral presentation O4B-G-4

For the assembly of the ITER vacuum vessel sector, precise positioning of welding end-effectors, at some distance in a confined space from the available supports, will be required, which is not possible using conventional machines or robots. This paper presents a special robot, able to carry these welding and also machining processes from inside the ITER vacuum vessel, consisting of a five-degree-of-freedom parallel robot mounted on a carriage driven by electric motor/gearbox on a rack. The robot carries the machining tool or welding gun such as a TIG, hybrid laser or e-beam welding gun to weld the inner and outer walls of the ITER vacuum vessel sectors. The kinematic design of the robot has been optimised for ITER access and a hydraulically actuated pre-prototype built. The machining force analysis and the optimisation of the machining processes is discussed in the paper. The welding process with a 3-D seam tracker is introduced. The motion control of the robot is challenging problem due to the nonlinear behaviour of the mechanical structure and hydraulic system. A hybrid controller is designed for hydraulics and control system of the robot, including position, speed and pressure feedback loops to achieve high control accuracy and high dynamic performances. Finally the experimental results are presented and discussed. Keywords: Parallel robot, ITER vacuum vessel, Machining/welding, water hydraulic,and Control


Corresponding Author:

wu huapeng

P.O.Box 20, FIN-53851 Lappeenranta, Finland

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-14 BAKING SYSTEM FOR EAST VACUUM VESSEL

Y.L Cheng, Y.T. Song D.M Yao

Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126,Hefei,Anhui 230031,China

EAST is a medium-size tokamak with super-conducting magnetic field coils. The primary function of the vacuum vessel are to provide a high quality vacuum for the plasma discharge. Baking technique is applied to EAST with a view to removing impurities which are mostly due to the release of containants from limiters and walls during the startup of the plasma discharge. Two baking systems are considered at one time resulting from the quite different structural characteristics. One is that the vacuum vessel of double-wall structure is heated by a flow of hot nitrogen gas between the inner and outer shell. The other is an electric heater roiling around the ports of the vessel. While the systems provide the maximal baking temperature of the vacuum vessel to be equal to 250ŽC to ensure to achieve acceptable thermal stresses and deformation due to temperature gradients, we must attach much importance to the non-uniform heating. The baking systems are presented. Some simulated analyses have been done to make sure to achieve their design specification and to expect some demolishing factors in the operating process.


Corresponding Author:

Y.L Cheng

P.O.Box 1126, Anhui, Hefei, P.R.China, 230031,Institute of Plasma Physics, Chinese Academy of Sciences

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-18 LASER MEGAJOULES CRYOGENIC TARGET DEVICES

BRISSET Didier, Valerie LAMAISON (1) Gael PAQUIGNON (1) Eric BOULEAU (2) Denis CHATAIN (2) Jean MANZAGOL (2) Jean Paul PERIN (2)

(1) Centre Etudes Scientifiques et Techniques Aquitaine /Departement des Lasers de Puissance B.P. 2 33114 Le Barp FRANCE (2)Commissariat Energie Atomique de Grenoble DRFMC/Service des Basses Températures 17 Rue des Martyrs 38054 GRENOBKLE Cedex 9

LMJ program claims to obtain Deuterium-Tritium (DT) mixture combustion leading to a fusion gain of ten. The cryogenic targets for inertial confinement, driven by 240 laser beams, are hollow spheres. Their internal shell are covered with a solid cryogenic fuel layer. The success of DT combustion depends on quality of the fuel layer geometry. Cryogenic targets must be cooled and kept at temperatures near the triple point (19 K) with a very good stability (1 mK) for many hours. In the French concept, the targets will be transferred at 22K +/-2K to the cryotarget positioner using another cryostat wich is carrying target from TargetLab to LMJ building . Besides that sharp thermal characteristics, the Cryogenic Target Positioner (CTP) displays others technical challenges like very strong mechanical specifications. The CTP deals with the target handling and positioning in the center of the 5 m radius experimental target chamber with a precision of few microns. The target transfer and positioning must be entierely remote controlled and done under vacuum. In order to validate our current CTP conception, we have manufactured a One Scale Cryostat Demonstrator to confirm all CTP thermal challenges, such as sharp thermal regulation, cooling autonomy and cryogenic target transfer. First results obtained with this prototype will be presented.


Corresponding Author:

BRISSET Didier

CEA/CESTA BP2 33114 LE BARP FRANCE

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-29 IRRADIATION TESTS ON WATER HYDRAULIC COMPONENTS

T. Hernandez, and E.R. Hodgson

Euratom / CIEMAT Fusion Association, 28040 Madrid, Spain

Remote handling operations in ITER will require the use of hydraulic systems for lifting and moving activities, such as those envisaged for the CMM (Cassette Multifunctional Mover) and the CTM (Cassette Toroidal Mover). These hydraulic systems will make use of high pressure water up to 210 bar, rather than oil. Hence the polymer materials employed as seals, glide rings, and wipers on the pistons and cylinders will be subjected to gamma irradation in the presence of water or high humidity. It is therefore necessary to study the combined effect of radiation and water on the degradation of the polymers, as well as possible enhanced water corrosion due to gases such as chlorine or fluorine released from the irradiated polymers. The use and trade names of the materials which have been tested are: Piston seal: Dehoplast PE-1000 (UHMW-PE). Glide ring: Merkel Hard fabric HGW HG517 (Freudenberg Simrit). Wiper: Merkel Novathan 95 AU V149 (polyurethane). Gamma irradiations have been performed in the CIEMAT 60Co pool facility (Nayade), which allows irradiation in a closed sample chamber at a controlled temperature and in a controlled atmosphere. Tests were carried out on the materials to 1.0, 3.3, and 10.0 MGy at 9 Gy/s, 300 K to evaluate the modification / degradation of the physical properties of the polymers. The irradiations were performed in two different environments: dry nitrogen and in deionised water. In addition control samples were put in deionised water without irradiation during the time needed to reach 1, 3.3, and 10 MGy in order to compare the effects of water alone, and water plus radiation. Following the tests microstructure, microhardness, chemical analysis, and water absorption were examined. Results for UHMW-PE (Ultra high molecular weight- Polyethylene) seals, NBR (Acrinolnitrile-butadiene rubber) O-rings and Polyurethane wiper rings demonstrated no notable degradation up to 10 MGy. However, by 10MGy the hard fabric material for the glide ring (piston bearing) exhibited undesirable levels hardening, production of fluorides in the water and the fibre matrix appears to have been dissolved.


Corresponding Author:

T. Hernandez

Euratom / CIEMAT Fusion Association, 28040 Madrid, Spain

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-53 EXPERIMENTAL RESULT OF THE LASER IN VESSEL VIEWING AND RANGING SYSTEM (IVVS) FOR ITER

Neri Carlo, L. Bartolini, A. Coletti, M. Ferri de Collibus, G. Fornetti, F. Pollastrone, M. Riva, L. Semeraro

Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome,Italy

A prototype of Laser in Vessel Viewing and Ranging System (IVVS) was developed at ENEA laboratories. It is based on an amplitude modulated (AM) laser radar specifically designed to withstand with the severe ITER conditions. The system is able to perform, at the same time, viewing&ranging of in-vessel surface. The target is scanned using a radiation resistant laser beam deflection system coping the ITER thermonuclear and mechanical constraints. The most critical parts of the system, which is the fiber optic optical encoder, has been successfully tested by SCK-CEN laboratories in Mol under radiation at 15 KGy/h up to a dose of 2.47 MGy. A series of viewing and ranging tests have been performed on the system to better evaluate its main characteristics that can be resumed in auto-illumination, large field of view, zoom capability, range measurement capability, high dynamic range (thousand of grey levels can be resolved), relative immunity on speed fluctuation of the scanning mechanism. It was verified the large field of view of the system that is able to take images in very wide angles (quasi-spherical images) as well as in restricted areas. Resolution charts have been scanned to evaluate the maximum resolution of the system, which has been able to distinguish rows of 0.55 mm at a medium distance of 3.5 m; the results was in accordance with the theoretical one <1mm in the range 2-7 m although speed fluctuation of about 20% has been observed. Test has been performed on the ranging capability of the system using a testing sample made of three group of stairs of 10 mm, 5 mm and 1 mm of step; the obtained image allowed to see well all the stairs groups both in amplitude and in range. A Standard Deviation of the range measure of 320 µm with a 3 ms stay time of the laser beam was reached comparing the measurement with the CAD representation of the sample. The paper shortly describe the overall architecture of the system then the radiation test performed on the more critical components and in more detail presents the experimental viewing and ranging results obtained, showing the main characteristics and the advantages of the system.


Corresponding Author:

Neri Carlo

Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome,Italy

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-56 ITER VACUUM VESSEL SECTOR MANUFACTURING DEVELOPMENT IN EUROPE

Lawrence P Jones, Aldo Bianchi (1), Alain Cros (2), Enrico di Pietro (3), Benoit Giraud (2), Kimihimo Ioki (4), Lubomir Junek (5), Bruno Parodi (1), Michael Pick (3), Gian-Paolo Sanguinetti (1), Richard Tivey (4), Yuri Utin (4)

(1) Ansaldo Ricerche - Corso Perrone 25, 16161, Genova, Italy, (2) Framatome ANP - 10 rue Juliette Récamier, 69456, Lyon, France, (3) EFDA CSU - Garching, (4) ITER JCT - Garching, (5) Inst. of App. Mechanics Brno - Veveri 95, 61139, Brno, Czech Republic

The ITER Vacuum Vessel provides the first Tritium and Vacuum boundary and supports the first wall blanket and divertor modules, the attachment requirements of which complicate the construction of the vessel and place manufacturing tolerances on the VV several times smaller than usual in relation to it’s large size so that there is a risk of rejections after manufacture due to out-of tolerance for the 9 Vacuum Vessel Sectors. Utilising large bracing fixtures stiffer and heavier than the vessel itself, European Industry has proposed a manufacturing route for the Sector construction, and the ITER International Team has accepted this method as the reference. However, the achievement of the required tolerances remains challenging and has to be validated prior to their procurement. The central part of this validation is the placement of a contract for the procurement of a full-size, poloidal segment, consisting of a 40 degree, 5 metre high, 20 Ton part of the inboard section, fabricated according to the manufacturing route, including bracing fixtures, welding applications, restraint effects, and fit-up aspects. A steel beam structure with stiffness comparable to the missing part of the sector is included, as this has an important influence on the distortion, which will be measured at the end of the fabrication. The finished segment is used as the basis for a mock-up that simulates the joining of two sectors. Since the goal of this program is to be able to use the model to extrapolate the segment distortions to the actual ITER VV sector, numerical simulations by a European Association, using the SYSWELD program, is used with a new module, incorporating local models of instrumented welding coupons, the results of which are included in a series of simplified global models. This paper describes the results of the manufacturing development programme so far.


Corresponding Author:

Lawrence P Jones

EFDA CSU - Boltzmannstr. 2, 85748, Garching, Germany

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-71 STRUCTURAL UPGRADE OF IN-VESSEL CONTROL COIL ON DIII D*

Anderson, P.M., A.G. Kellman, E.E. Reis

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

For most of 2003, DIII-D operated 12 new in-vessel outer wall mounted control coils. The single turn, rectangular coils are mounted in 2 levels of 6 coils each. The coils were used for many experiments such as suppression of the resistive wall mode, for correction of magnetic field imperfections and for creation of an ergodic edge magnetic field for the suppression of edge localized modes. During operation with a maximum current of 4.5 kA at 100 Hz, one coil developed a leak through the stainless barrier that separates the nitrogen blanketed insulated conductor from vessel vacuum. A pair of coils was taken out of operation for the last month of the year. This paper describes the failure investigation, design, analysis, component testing, repair, system testing and new interlocks for the system that will see significant use in 2004. The crack in the stainless barrier was attributed to low cyclic fatigue related to operation of the coil at 100 Hz, a frequency near the vertical natural frequency of the coaxial lead. Finite element analysis (FEA) after the failure showed that electromagnetic forces on the single conductor section were sufficient to excite the coaxial lead in a vertical mode. The crack likely developed in less than a second. Repair options were limited. The coils are mounted to the walls and removal of PF tiles was discouraged in order to minimize the repair time. Welding on the coil was limited in order to protect the internal polyamide insulator from overheating. Repairs included: 1) seal the leak in the faulted coil, 2) increase the stiffness of the single conductors near the coaxial transition and 3) significantly increase the first natural frequency of the coaxial leads to allow operation to 1000 Hz. In-vessel vibration testing was done at each stage of repair to compare the natural frequency of the three types of leads with that determined by FEA models. Verification testing was done prior to vessel closure. The test included temporary installation of field tolerant strain gages to monitor strain in the stainless for comparison with model results. Permanent vacuum vessel port deflection monitors were added with the hope that excessive lead vibrations could be detected by port deflections for interlock purposes. All 12 coils were successfully repaired, upgraded and test results are encouraging. *Work was supported by the U.S. Department of Energy under DE-FC02-04ER54698.


Corresponding Author:

Anderson, P.M.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-89 MANUFACTURING OF CRYOSTAT FOR EAST SUPERCONDUCTING TOKAMAK

YU Jie,

EAST superconducting Tokamak is under fabrication and in procurement phase at Institute of Plasma Physics, Chinese Academy of Sciences. One of the main parts of ESAT is the cryostat which provides the thermal protection of the coil system. The cryostat consists of a cylindrical section bolted to dished lid wall at top and to base plate at bottom by flanges with special C clamps. The lid wall of the cryostat is a dished configuration for reasonable stress distribution. The support of the cryostat on the base has been designed for transferring the loads of the vacuum vessel and magnets to the basis of the machine test hall directly. All parts of cryostat withstand the design basic loads, which include external pressure at most operating condition (1bar), dead weight, electromagnetic forces and seismic load. The manufacturing of the cryostat is under way and will be finished in July, 2004. Totally sixteen horizontal ports and thirty-two vertical ports are designed for the requirement of diagnostics and operation. With physical diagnostics and test specifications, three types of horizontal ports with different shapes and size are needed, in which three of them are capable for tangential neutron beam injection and physical diagnostics. In addition, there are eight horizontal man ports for future maintenance. There are 16 vertical access ports and four maintenance ports (the same four ports in the down section of middle ring) in lid wall. There are sixteen horizontal ports to the machine vacuum vessel at the machine equator. There are two types sixteen vertical ports, eight cryogenic ports and one helium exhaust port on the base plate. The cryostat is made of 304L stainless steel. The cryostat is 7592mm outside diameter and 7095mm height (not including main support). The cryostat is manufactured by Shanghai Boiler Works, Ltd (SBWL). The inside wall radius is 3661mm and the height of cylindrical section in middle ring 4460mm. The lid is made of 28mm 304L stainless steel, which minimum thickness is 25mm. The non-standard dished head spherical and knuckle radii are 10000mm and 850mm, respectively. The dished lid was fabricated at the end of 2002 by using a set of head ramming machine and a set of head flanging machine (BOLDRINI) imported from Italy with working diameter from 1800mm to 8000mm.


Corresponding Author:

YU Jie

P.O.Box 1126, Hefei Anhui 230031, P.R.China

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-117 SPECIAL BLANKET DESIGN IN THE NB REGION OF ITER

ELIO Filippo, K. Ioki , Y. Utin, M. Morimoto

ITER International Team Boltzmannstrasse 2 85748 Garching Deutschland

The ITER blanket is a 45 cm thick nuclear shield mounted on the vessel as 440 modules with regular shape and a weight below 4.5 tonnes. The modules have a straight profile 85-120 cm long allowing the production of the first wall (FW) in flat panels, 25-40 cm wide, four per module. To maintain a trapezoidal front view, the boundaries between the modules have been aligned with the edges of the equatorial and upper ports of the vessel. The design of the FW panels, the main stainless steel body and the mechanical attachment to the vessel wall has been developed and tested for this configuration. Unfortunately the 3 Neutral Beam (NB) openings violate the 20 cyclic symmetry of the regular ports and would require a different blanket segmentation. These openings are smaller and cross the FW in the plane of the toroidal field coil. In the past the NB module design lacked a satisfactory attachment on the vessel and was not fully integrated with the standard blanket portions. The requirements of the NB openings have been reviewed and compared with the design constraints of the vessel, of the blanket cooling manifolds, of the module and its attachment. A new layout has been developed which appears to be a good compromise for all components and a good basis for the detailed design. The height of the NB opening has been squeezed symmetrically from 136 to 116 cm resulting in a higher, though still acceptable beam heat flux on the sides of the duct. The equatorial portion of the blanket cooling manifolds has been curved toroidally around the beam. The blanket module on the side of the NB opening exposed to the plasma has been extended 80 cm along the inside of the port. Thus it has a wider region for the attachment and completely protects the vessel corner from the nuclear radiation. The poloidal segmentation is now 2 modules as it is elsewhere between the equatorial ports, the toroidal spacing is also close to the regular angular pitch of 10 with a deviation imposed by the slanted sides parallel to the NB axis. All these benefits require a new effort to develop the corner blanket module, which is special and exposed to nuclear radiation and surface heating on two sides. The paper explains the blanket design evolution in the NB region, lists the requirements of all interacting components, presents their individual design and discuss it. The implications for the handling and the feasibilty of a separable FW with its reduced waste are also reported.


Corresponding Author:

ELIO Filippo

ITER Team, Boltzmannstrasse 2, 85748 Garching, Deutschland

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-137 USE OF ELECTRONIC AND OPTOELECTRONIC INDUSTRIAL SYSTEMS FOR MAINTENANCE TOOLS OF ITER FUSION EXPERIMENTAL REACTOR

GIRAUD Alain, Marco Van UFFALEN(1) Francis BERGHMANS(1)

(1) SCK CEN, Department Instrumentation Boeretang 200, B-2400 Mol, Belgium

The environmental constraints encountered by the maintenance tools of the future ITER Fusion Reactor such as CMM, SCEE, DTP and DRP, could reach a few MGy of total dose irradiation while temperature could rise up to 150 C. The necessary remote handling systems will generate a huge number of wires to connect the sensors and actuators to the control room. The use of embedded electronic and optoelectronic systems could be an appropriate solution, for the future designers, to reduce the size of umbilical and connectors, facilitate their movements and limit inside failures. Involved since fifteen years in the developments of industrial electronic systems for civilian nuclear activities (AREVA, EDF, …), our laboratory has gained its knowledge in the permanent understanding of the radiation behavior of components and the optimized design of electronic architecture to extend the lifetime of on board systems. In the same way, SCK laboratory has developed an similar approach for optoelectronic components and optical fibers. Taking into account the common use of sensors with analog output signals, a experimental data link was built in 2002 for communications between an embedded sensor and the control room. A mock-up including electronic and optoelectronic components was designed to digitalize the analog output signal and transfer it, after an electric to optical conversion, through optical fiber. The expected tolerance to the environment was established without significant degradation of converted signals. The continuous survey of emerging “off the shelf” technology allowed us to propose now a pre-prototype of a data multiplexer able to generate a 16 bits frame. Data inserted represent the digitalization of analog output coming from well-known sensors like LVDT or resolvers, limit switches or proximity sensors. The transfer to the control room can be done either by a bifilar or optical support at a frame rate of 32kHz which meets most of the remote processes needs. Validation to radiative environment was performed without any failure. The regularity of all analog conversions was correctly assumed. The final prototype could be proposed to demonstration and final validation by the end of this year. Machine designers and end-users will be provided timely with appreciable tools to limit wires and simplify umbilical problems by proceeding, all along ITER life, to a regular exchange of “electronic black boxes”.


Corresponding Author:

GIRAUD Alain

CEA/DRT/LIST/DTSI/SARC Bat 451 CEA/Saclay 91191 Gif-sur-Yvette FRANCE

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-199 NON-DESTRUCTIVE TESTING OF BONDED STRUCTURES FOR PLASMA FACING COMPONENTS

ONOZUKA, Masanori, K. Kikuchi, A. Kirihigashi, Y. Oda, and K. Shimizu

Mitsubishi Heavy Industries, Ltd. Kobe Shipyard & Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe, 652-8585 Japan

Because of material characteristics requirements and adequate heat removal capability, bonded structures have been employed for plasma-facing components in the International Thermonuclear Experimental Reactor. To ensure the structural integrity of the bonded structures, non-destructive testing is to be conducted. Amongst the various types of non-destructive testing, ultra-sound testing (UT) was examined, with emphasis on UT inspection of the first-wall panel of the blanket module. Three test samples simulating the first-wall panel were fabricated. Two plates made of Cu-Cr-Zr alloy were bonded with stainless steel (SS) cooling pipes that were inserted between the two plates to form a heat sink structure. The heat sink structure was then bonded to a SS structural block. Thus, there are three bonded interfaces: the first between the Cu alloy plates, the second between the Cu alloy plate and the SS block, and the third between the Cu alloy plates and the SS pipe. In the test samples, several artificial defects were applied along the bonding interfaces. Three types of UT probes have been tested. A vertical UT probe and a phased array UT probe were used to detect defects between the Cu alloy plates, and between the Cu alloy plate and the SS block. Both the probes were applied on the Cu alloy surface or on the SS block surface. To detect defects along the SS pipes, a beam-focused type UT probe has been applied. The focused-type probe was inserted into the pipe for detection. In addition, to attain a better signal to noise ratio (S/N), a noise reduction technique has been applied using digital data processing. Using the above UT probes, artificial defects as small as 2 mm in size have been successfully detected at a S/N ratio of more than 2. Details of the study will be presented at the symposium.


Corresponding Author:

ONOZUKA, Masanori

Mitsubishi Heavy Industries, Ltd. Nuclear Systems Engineering Department, Minatomirai 3-3-1, Nishi-ku, Yokohama, 220-8401 Japan

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-240 VERTICAL DIPLACEMENT EVENTS SIMULATIONS FOR TOKAMAK PLASMAS

Paccagnella Roberto, T.Bolzonella, M.Cavinato, S.Ortolani, G.Pautasso(1), W. Schneider (1), V.Lukash(2), H. Strauss(3)

(1) Max Planck Institute, IPP (Garching, Germany) (2)RRC Kurchatov Institute (Moscow, Russia) (3)Courant Institute for Mathematical Sciences (NY, USA)

In this paper we study the so called Vertical Displacement Events (VDEs) in an elongated and diverted tokamak plasma. The study is carried out using two numerical codes: DINA [1] which is a nonlinear magneto-hydro-dynamic (MHD) 2D code (assuming plasma axi-symmetry) evolving the plasma through equilibrium states which satisfies the toroidal Grad-Shafranov equation and which takes into account of the electromagnetic interaction with metal walls and external coils; M3D [2] a 3D multi-level toroidal code which evolves in time the full MHD equations and takes into account a resistive wall surrounding the plasma by matching the internal plasma magnetic field with the external vacuum solution. In this paper we compare the time evolution of a VDE for the axi-symmetric (DINA) and the non axi-symmetric (M3D) cases starting from the same initial equilibrium. We analyze ITER-like equilibria and also some equilibria relevant for the ASDEX-U experiment. For the simulations in the M3D code a “virtual casing” method is used with the coil’s currents held fixed at their initial values, while in the DINA code the coils currents can or cannot be constant in time during the evolution. In the case of ASDEX-U a comparison of the experimental data with the 2D simulation results is performed while for the 3D case the study is focusing on the identification of suitable conditions able to represent the effect of the passive structures. This study is particularly important in order to estimate the symmetric (DINA) and non axi-symmetric (M3D) structure of halo currents during VDEs. This can have an import impact on the mechanical design of future experiments like ITER. The requirements for the VDE control system in such devices can be also affected by the computations presented here. [1 ] R.R. Khayrutdinov, V.E. Lukash, Studies of Plasma Equilibrium and Transport in a Tokamak Fusion Device with the Inverse-Variable Technique, Journal of Comp. Physics 2, 106, (1993) [2 ]Park W., et al., Phys.Plasmas 6, 1796 (1999)


Corresponding Author:

Paccagnella Roberto

Consorzio RFX Corso Stati Uniti 4 35127 Padova

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-249 DESIGN OF THE ITER HOT CELL BUILDING

J Hayward (1), D Maisonnier (1) O Asuar (2) T Fisher (3) T Eurajoki (4) and Jöelle Elbez-Uzan (5)

(1) EFDA Close Support Unit, Garching; (2) EFET, IBERTEF-EA; (3) EFET, NNC; (4) EFET-Fortum; (5) Euratom-CEA Association

The Hot Cell building is a reinforced concrete building, which provides the facilities required for a variety of operations on activated in-vessel components and systems. Due to the erosion of the plasma-facing components by the high thermal loads, the divertor will need to be replaced and upgraded several times during the life of ITER. To minimise the amount of activated waste, the divertor is of modular design and based on the use of reusable cassettes, which can be removed from the Tokamak to be repaired and refurbished with new plasma-facing components. The in-line repair and refurbishment of divertor cassettes is the main operational requirement of the Hot Cell, but the repair and refurbishment of other Tokamak components, including diagnostic plugs, blanket modules and RF heating port plugs, are also part of the specified Hot Cell operations. Additionally, the Hot Cell has to process and store waste accumulated from the Tokamak during its operational lifetime. At present, the size and layout of the building is determined principally by the maintenance requirements and to provide all the facilities necessary for equipment storage, repair and testing, exchange and maintenance of remote handling tools, and waste processing and storage. A number of studies and reviews of remote maintenance activities have resulted in a simplification of the remote handling methodologies and of the concepts for the refurbishment tasks. The design of the Hot Cell building must also be reviewed with respect to the applicable regulatory codes, the possible requirement for a later extension of the facility to cater for the dismantling of the in-vessel component systems during the de-activation phase of the project, and other specific functions, such as in-vessel component docking, dust cleaning, transfer cask storage, atmosphere confinement control, and atmosphere de-tritiation. The paper will describe the results of these reviews and proposed changes to the design to meet both the functional and regulatory requirements.

WITHDRAWN


Corresponding Author:

J Hayward (1)

EFDA Close Support Unit, Max-Planck-Institut für Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching-bei-München, Germany

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-264 RECENT DEVELOPMENTS TOWARDS ITER 2001 DIVERTOR MAINTENANCE

PALMER James, Mikko Siuko(1) Pietro Agostini (2) Roland Gottfried (3) Michael Irving (2) Alessandro Tesini (4) Marco Van Uffelen (5)

((1) Tampere University of Technology, Tampere, Finland (2) ENEA CR Brasimone, Bologna, Italy (3) Framatome ANP GmbH, Erlangen, Germany (4) ITER International Team, Naka, Japan (5) SCK-CEN, Mol, Belgium

The divertor assembly for ITER consists of 54 rail-mounted cassettes located in the bottom region of the vacuum vessel. Due to the erosion of the plasma-facing components and the possible need for improving the divertor design, its periodic replacement is foreseen a number of times during ITER’s 20 year operational lifetime. In moving from the ITER’98 to the more compact ITER 2001 design, although the general principles of divertor RH remained intact, the detailed design of almost all the divertor handling equipment had to be significantly changed, mainly due to the reduction in space between the cassettes and the inner wall of the vacuum vessel. This feature prevents the use of the simple “trolley-like” cassette carriers developed for ITER’98 (and modelled in the Divertor Test Platform (DTP) at Brasimone), but necessitates the use of cantilevered cassette handling using a more complex device known as the “Cassette Multifunctional Mover” (CMM). In this new approach the vertical position of the cassette during its passage along the vacuum vessel duct is no longer simply related to a fixed set of radial rails but has to be continually adjusted in free space by a serial chain of robotic joints. Added to this, the nominal clearances between the cassette and vacuum vessel duct are only 30 mm which sets extreme demands for the mover control system. Since early 2003 the EU Participant Team has been engaged in the detailed design of the CMM together with its set of specialised end-effectors. The efficiency of this process and integration of the CMM design into that of the ITER machine, have been greatly enhanced by the use of state-of-the-art virtual reality and virtual prototyping techniques using Igrip and ADAMS software. During the same period a new RH mock-up facility, designated DTP2, has been designed and specified. Its main purpose will be to allow demonstration and refinement of the CMM design and related operational procedures in preparation for procurement of the actual RH equipment to be used in ITER. This paper will briefly describe the current ITER divertor replacement rationale, report on the latest cassette mover designs and outline the nature and objectives of the new DTP2 test facility.


Corresponding Author:

PALMER James

EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-293 MANAGEMENT OF A WATER LEAK ON ACTIVELY COOLED FUSION DEVICES

SAMAILLE Frank, M. Chantant, D. van Houtte, J.J. Cordier, L. Gargiulo

Association Euratom-CEA, CEA/DSM /DRFC, CEA- Cadarache, 13108 Saint Paul Lez Durance, France,

ITER will be the most important machine equipped with actively cooled Plasma Facing Components (PFCs). In case of abnormal events during a discharge, the PFC will be submitted to localized transient phenomena (high power densities, run away electrons, etc ), leading, in the worst case, to the degradation of the PFC wall and possibly to a water leak. In any case, a leak will have important consequences for the PFCs and equipments located in the vacuum vessel or connected to the ports such as seals, pumping systems or diagnostics. A great experience of these events has been gained at Tore supra since more than 10 years and it will be useful for the next step machines. During 2002 and 2003 experimental campaigns, several leaks occurred at Tore Supra. In the most important one, 2000 liters of water were spilled in the vacuum. During this major event, specifics actions have been done to limit the damages and especially to preserve the aluminum seals. In the first seconds after a leak occurs, if possible, a fast and reliable research of the leaking circuit has to be carried out. The isolation valves of the circuit are then closed in order to reduce the inside pressure in the lines and to limit the water vapor flow into the vacuum vessel. In the case of a large leak, all the circuits of the cooling loop connected to the Tokamak which are located in the Torus Hall are drained off by using a wired safety device. A drain off procedure has been defined and it is continuously improved. At the present time, the drain of the PFCs fed by the upper part circuit is not fully satisfactory because of several lines connected in parallel. Preliminary experiments have been performed to improve it and led to encouraging results. Once the leaky PFC has been emptied, some effort is required to evacuate the water from the vacuum vessel. After removal of water by gravity, and then, by baking and pumping in the vessel, the identification and repair of the leaky PFC start. The quick restart of the systems after the most severe leak of September 2002 with a pump-out without any air leak at the level of one hundred Aluminum seals confirms the convenience of the performed actions. The paper will present the description of the procedures applied to put the system in safety depending on the gravity of the leak. It will also present the methods used at Tore Supra to drain-off the primary loop circuits and to determine the leaky PFC.


Corresponding Author:

SAMAILLE Frank

Association Euratom-CEA, DSM / DRFC, CEA- Cadarache, 13108 Saint Paul Lez Durance

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-296 DESIGN PROGRESS OF THE ITER VACUUM VESSEL AND PORTS

Utin Yuri, V. Chuyanov(1), F. Elio(1), K. Ioki(1), L. Jones(2), V. Komarov(3), E. Kuzmin(3), M. Morimoto(1), M. Nakahira(4), G. Sannazzaro(1)

(1) ITER IT, Boltzmannstr. 2, 85748 Garching, Germany (2) EFDA, Boltzmannstr. 2, 85748 Garching, Germany (3) NTC "Sintez", Efremov Inst., 189631 St. Petersburg, Russia (4) JAERI, Naka Fusion Research Establishment, Naka, Ibaraki, 311-0193, Japan

The ITER vacuum vessel (VV) is a torus-shaped double-wall structure with stiffening poloidal/toroidal ribs between the shells. The VV main function is to provide the high-vacuum and primary safety confinement boundary. The vessel also supports the blanket and the divertor components. Along with the in-vessel components, the VV provides radiation shielding – the neutron heat is removed by the water circulating between the shells. To provide access inside the vessel for auxiliary plasma heating, diagnostics, vacuum pumping and other needs, the VV is equipped with upper, equatorial, and lower ports. Approaching the ITER construction phase, the VV design has been improved and developed in more detail with the focus on simplified manufacture and reduced cost. Options of the general fabrication scheme have been considered in cooperation with the industrial companies and the design has been updated in conformity with the main manufacture requirements and recommendations. To simplify the design, the inboard triangular supports of the blanket modules have been eliminated and the design of the outboard supports refined. Another important improvement is that the VV supporting system has been modified to provide better access to the main supporting components after assembly of the machine. For some ports, a single-wall construction will be used at a certain distance from the main vessel, where the neutron load is less intense. This approach simplifies the port manufacture and maintenance. For the upper and equatorial ports, the in-port space is occupied by an integrated subassembly - the port plug, which, apart from its functional purposes, provides the vacuum/pressure boundary for the in-vessel volume. Special attention was paid to the design of the supporting and sealing components between the plug and the port with the focus on improved structural performance and maintenance. For further cost reduction, the number of large lower ports is halved, with the ports between every other toroidal field coil. Where there are no ports, there are only pipe feedthroughs and local small penetrations. The VV must withstand those loads directly induced in the vessel and those transmitted from the in-vessel components. Based on the performed structural analyses, additional reinforcements have been incorporated into the main vessel/ports where required. Details of the current VV design and results of the related analyses are reported in this paper.


Corresponding Author:

Utin Yuri

ITER IT, Garching JWS, Boltzmannstr. 2, 85748 Garching, Germany

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-348 1200 MM BORE VOLTAGE BREAK OF THE NB DUCT FOR KSTAR

B.J. Yoon, T. Nagayama(1) S.R. In(2) B.H. Oh(2)

(1)Hitachi Haramachi Electronics, Hitachi-shi, 317-0072, Japan (2)Korea Atomic Energy Research Institute, Daejeon, 305-353, Korea

The beam duct connecting the NBI system to the KSTAR (Korea Superconducting Tokamak Advanced Research) vacuum vessel consists of a large gate valve, a voltage break, transition tubes, and beam stoppers. The voltage break keeps peripheral devices safe from the potential difference of higher than 10 kV generated between the NBI and the torus during plasma disruptions and power faults in superconducting magnets and NBI high voltage system. The voltage break is composed of a ceramic ring, a bellows and fitting flanges. The voltage break is mechanically very delicate component because it must accommodate thermal and mechanical relative displacements of the vacuum vessel side and NBI side structures, occurred due to the system baking and misalignments in the assembly. Therefore, the bellows should be flexible enough to absorb a shift of a few cm in the axial direction and ~5 mm in transverse and be rigid to withstand the atmospheric pressure exerted on sidewalls of convolutions. The bellows is designed to be a welded type and have a size of 1200 I.D, 75 mm width, 14 mm pitch and 15 convolutions to fulfil above requirements. The ceramic ring made of alumina (Al2O3) is the key part of the voltage break which has a design breakdown voltage of 30 kV. The ceramic ring has dimensions of 1200 mm I.D., 50 mm thickness and 45 mm width. The ceramic ring is brazed with KOVAR sleeves on both flat sides, and the sleeves are welded to the bellows and the flange assembly. The ceramic ring is the most fragile part in the voltage break at the bonding boundary between the sleeve and the alumina ring. There has been no successful experience of fabricating a 1200 mm bore alumina ring so far in the world. At the first step a ceramic ring is formed with the alumina powder by pressing and sintering at the Kyocera Kagoshima factory, and then bonded to KOVAR sleeves using the active metal brazing method in the vacuum brazing furnace at the Mitzubishi Hiroshima factory. The welding of bellows and flanges to the ceramic ring is carried out at the Hitachi factory. The entire procedure is managed by Hitachi. The surface of the ceramic ring is checked with the ink penetration method to find cracks before the brazing procedure. After brazing the ceramic ring is checked with the eye and leak-tested. The voltage break assembly is tested by pressurizing inner volume of the break up to 1.5 atm. The strength between the ceramic ring and the KOVAR sleeve is expected to be more than 80 MPa.


Corresponding Author:

B.J. Yoon

Korea Atomic Energy Research Institute, Daejeon, 305-353, Korea

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-355 MANUFACTURE OF THE PLASMA VESSEL AND THE PORTS FOR WENDELSTEIN 7-X

Reich, Jens, Willi Gardebrecht (1) Bernd Hein (1) Bernd Missal (1) Joerg Tretter (1) Franz Leher (2) Stefano Langone (3)

(1) Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (2) MAN DWE GmbH Deggendorf, Werftstraße 17, D-94469 Deggendorf (3) Romabau-Gerinox AG, Fohlenweide, CH-8570 Weinfelden

Selectesd also for Oral Presentation O4B-G-355

WENDELSTEIN 7-X (W7-X) is a superconducting helical advanced stellarator which is presently under construction at the Greifswald branch of IPP. Thermal insulation of the 70 coils requires a cryostat. It is being composed of a plasma vessel, an outer vessel, ports to observe and heat the plasma, cooled shields and multilayer insulation. The German company, MAN-DWE, is responsible for manufacture of the plasma vessel, the outer vessel and the thermal protection. The Romabau-Gerinox AG delivers the ports. Following the symmetry of the magnetic configuration the cryostat is composed of five almost equal modules. The shape of the plasma vessel has to closely follow the twisted shape of the plasma and has a cross section which continuously varies between triangular and bean shape. The plasma vessel is composed of 10 half-modules. Each half-module is again divided into two sectors to allow stringing of the coils during assembly. For each half-module 20 steel rings are precisely bent to the required shape and carefully welded to represent the changing cross-section of the plasma vessel. For local areas, which cannot be approximated by bending steel, sheets are fitted by hot forming. By end of March the four sectors required for the first module of the plasma vessel were delivered. The contours of the sectors were measured by laser tracker system and met well the given narrow tolerances. Vacuum tightness of the welds was checked by an integral helium leak test of each whole sector. Precise cutting of the holes for the ports was performed by water jet technique. Water pipes around the outside of the vessel allow its temperature to be controlled during plasma operation and for bake out. A total of 299 ports are used for evacuating the plasma vessel for plasma diagnostics and plasma heating and for feeding supply lines and sensor cables. Cross sections of the ports range from 100 mm circular to 1000 x 400 mm rectangular. Bellows balancing movements of the plasma vessel during bake-out and final adjustment. All ports are surrounded by water pipes to control their temperature. By spring of this year 60 ports have been delivered. The paper will summarise the design activities and give a short description of the fabrication status of the main cryostat components.


Corresponding Author:

Reich, Jens

Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald, Germany

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-360 DYNAMIC IDENTIFICATION OF THE HYDRAULIC ITER MAESTRO MANIPULATOR - RELEVANCE FOR MONITORING

BIDARD Catherine, C. Libersa (1) D. Arhur (1) Y. Measson (1) J.-P. Friconneau (1) J.-D. Palmer (2)

(1) Robotics and Interactive Systems Unit - CEA, BP 6, F92265 Fontenay-aux-Roses Cedex, France (2) EFDA CSU Graching, Boltzmannstrasse 2, 85748, Graching, Germany

Maintenance tasks at ITER divertor level requires use of powerful Force feed back Remote Handling device such as hydraulic Manipulator. CEA in collaboration with CYBERNETIX and IFREMER has developed the hydraulic manipulator MAESTRO (Modular Arm and Efficient System for TeleRObotics). Force control of the robot allows for force-reflective telemanipulation, which is required for better manipulation and sensing of the environment during remote operation tasks. When considering ITER vacuum Vessel conditions, radiation level will reduce significantly the possibility to operate remote tools with good vision feed back. Therefore, model based monitoring of the manipulator will be required during operational period to enhance feedback to the operator. Therefore, it is required to control that the manipulator inside the Vacuum Vessel is well operating and to warn the operator of possible failures. Without adding any sensors, it is possible to monitor the relation between the robot torques and trajectory. In this paper we present the experimental identification of the parameters of the dynamical model of the MAESTRO arm. The control of joints was done on the MAESTRO arm with flow control servo-valves. The axes torques were derived from pressure measurements in joints chamber, and the axes positions measured by resolver sensors. The first part presents the experimental set-up and trajectories. The second part deals with the joints friction models: nonlinear viscous behaviour was observed and explained by possible non laminar effects of flow inside the hydraulic joints. One axis with multiple seals showed Stribeck effect. The third part presents the identification of the articulated dynamics model. We used the regressor form of the dynamic equation to get a least-square solution. Then the prediction capability of the identified model is tested on a test trajectory, similar to the combined trajectories used for identification, and a robot movement obtained using manual command via a master arm. Finally the capability to detect perturbations is tested on movements where the robots contacted and pushed some objects. As conclusion, we examine the capability to monitor for unpredicted behaviour such as collisions or friction with an unknown environment. This capability is however limited to the case when the robot is not moving. Furthermore, when using pressure-controlled servo-valves, the dynamical model will also detect internal failure in the servo-actuators.


Corresponding Author:

BIDARD Catherine

Robotics & Interactive Systems Unit - CEA , BP6, F92265 Fontenay-aux-Roses Cedex, France

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-361 GENERIC CONTROL SYSTEM DESIGN FOR THE CASSETTE MULTIFUNCTION MOVER AND OTHER ITER REMOTE HANDLING EQUIPMENT

Michael IRVING, J.Palmer (1) M.Siuko (2)

(1) EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany (2)Tampere University of Technology, PO Box 589, 33101 Tampere, Finland

A fundamental difference between ‘robotic’ type control systems for normal industrial environments and those intended for nuclear environments, is how they react to failures in the equipment they control. Normally, troubleshooting and recovery from failure are carried out locally to the failed equipment, which for the remote handling (RH) equipment used to carry out in-vessel ITER maintenance, is not possible due to the radiation levels involved. Control systems for these environments need to be designed with fault detection, diagnosis and recovery at the outset, since it is quite possible for a failure in the RH equipment to render it unrecoverable and therefore compromise the whole project. Past experience in this field indicates that mechanical handling equipment controllers are generally made by the company supplying the mechanics, and if as is likely in ITER, different items of RH equipment are supplied by different manufacturers, each will possess a different type of controller with different philosophies running different HMI’s, as has already happened. Across the entire suite of RH equipment, this requires an enormous amount of technical knowledge to give effective support. The alternative approach is to recognise that from the RH control perspective, equipment whose size, function and purpose may be totally disparate, are likely to be drivable using virtually identical controllers. With RH personnel needing in-depth knowledge of a now reduced range of equipment, incorporating modern reliability and recovery techniques, it should be possible to detect failures before the task being carried out is compromised, and certainly while recovery is still possible. Furthermore, controller subsystems designed and built in modules will ease later upgrades, as well as making repairs simple and quick. This paper will explore the design and operation of RH control systems specifically for ITER-type RH applications, with a view to identifying guidelines to the equipment suppliers which will constrain the proliferation of controllers that would otherwise naturally occur. This approach is particularly timely, since a new suite of RH equipment for ITER is currently being specified, starting with the Cassette Multifunction Mover (CMM) to be used to transport the divertor cassettes into and out of the vessel. Using this as an initial example, a generic type of control system will be presented, which could be used as a basis for control of other ITER RH equipment.


Corresponding Author:

Michael IRVING

Remote Maintenance Group (UTS Tecnologie Fisiche Avanzate), ENEA CR Brasimone, 40032 Camugnano (BO), Italy

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-374 ANALYSES OF THE ITER VACUUM VESSEL WITH THE USE OF A NEW MODELLING TECHNIQUE

Rozov Vladimir, E. D'Agata, K. Ioki, M. Morimoto, G. Sannazzaro, R. Tivey, Yu. Utin

Same as Corresponding Address

The on-going design development of the ITER Vacuum Vessel (VV) has been supported and accompanied recently by extensive analytical studies of its different structural aspects. The use of a newly developed parametric model of the ITER VV sector with port structures has enabled various assessments of this structure in detail. Implementation of new features in the model and new approach for its generation eases updates and modifications of the model, reducing a time necessary to introduce them into the model and allowing it to keep pace with discussions on planned design change. It helps to keep the model up-to-date with the latest status of the design and to minimise the time lag between the latest reference design documentation and the available set of analyses validating it. A number of assessments have been accomplished recently for the ITER VV. An assessment of its confinement function shows that it is able to fulfil its function and to keep the scale of any mechanical failures limited to local areas even in the cases of a deflagration inside the chamber. An assessment of the VV under hydrostatic test coolant pressure load has enabled an estimate to be made of its performance as a double-wall structure pressurised in the interspace between the shells. Definition of the forces developed in different regimes in the housings of all blanket supports which act as local links between the inner and the outer shells has provided grounds for their further design improvement, correlated with the problem of the VV manufacturing. Some analyses of the local reinforcements have been done since the new concept of the VV support has been adopted. The implementation of complementary specialised tools for the automatic allocation of the distributed masses has enabled a realistic representation to be made of the mass and inertia properties of the real object and its parts in the program-generated model of the VV. This makes it possible to perform assessments of the frequency and dynamic characteristics of the VV and other inertia-related issues. An assessment of the triangular support region following recent design changes has been conducted using a complementary model developed as a result of the implementation of the new modelling system for the purpose of generation of local detailed models consisting of solid elements. The paper describes the main results of these latest assessments of the ITER VV as part of the analytical support of the on-going design work.


Corresponding Author:

Rozov Vladimir

ITER Joint Work Site, Boltzmannstr. 2, D-85748 Garching, Germany

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-389 ITER ARTICULATED INSPECTION ARM (AIA): GEOMETRIC CALIBRATION ISSUES OF A LONG-REACH FLEXIBLE ROBOT.

D. ARHUR (1), Y. Perrot(1) C. Bidard(1) J.P. Friconneau(1) J.D. Palmer(2) C. Talarico(3)

(1)Robotics and Interactive Systems Unit – CEA. BP6 F-92265 Fontenay aux Roses Cedex France (2)EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany (3)ENEA –CR Frascati Fusion Technology Division – Via E. Fermi, 45 00044 Frascati Italy

This project takes place in the Remote Handling (RH) activities for the next step of the fusion reactor ITER. Close inspection task of the Vacuum Vessel first wall of ITER with a long reach robot motivates improvements on accuracy of the end effector’s position. The aim of the R&D program performed under EFDA work program is to develop a flexible model of an IVP-like system. The first phase of the project concerned the development of a IVP-like physical model. As the characteristics of the structure are the big dimensions, the high number of joints with reduction of mass, significant compliance of the structure occurs. However, geometric parameters but also non-geometric parameters such as stiffness must be taken into account in the flexible model. The output of the model is the geometric configurations of the arm including the end effector’s position. As our goal is to simulate the behaviour of a IVP-like system, a precise calibration of the parameters is essential. The set of parameters is obtained using a non linear and multivariable optimisation method. Its aim is to reduce the average distance between the end effector’s position stemming from the model and the measured position by optimising a set of parameters. The identification method is first tested on simulated positions to validate the feasibility of the approach and lastly applied to practical experimentation performed on the IVP first module. On the practical experimentations, we show that taking into account the flexibilities improves accuracy on the end effector’s position of the first module. The clearance on the rotation axis and the repeatability of the mechanics affects the results of the optimisation because those phenomenons are not predicted by the model. The mechanics must be improved to tend towards better results. Nevertheless, the calibration method is more general, faster and improves the precision on the end effector’s position of the first module. The extrapolation of the model to the whole arm will enable to compensate errors from more than 1 meter range in some configurations to a final accuracy of few centimetres. Once the whole arm will be available, a global identification will have to be performed with the same method validated on the first segment.This paper presents the physical model of the robot, the results stemming from the simulation and measurements test campaign and at last the benefits of using this kind of model on the accuracy of the end effector’s position.


Corresponding Author:

D. ARHUR (1)

Robotics and Interactive Systems Unit – CEA. BP6 F-92265 Fontenay aux Roses Cedex France

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-393 ITER ARTICULATED INSPECTION ARM (AIA) : R&D PROGRESS ON VACUUM AND TEMPERATURE TECHNOLOGY FOR REMOTE HANDLING.

Yann PERROT (1), J.J. Cordier(2) J.P. Friconneau(1) L. Gargiulo(2) E. Martin(3) J.D. Palmer(4)

(1) DTSI - CEA BP6 92265 Fontenay aux Roses France (2) DRFC – CEA Cadarache, 13108 St Paul Lez Durance France (3) EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany (4) ITER International Team, Boltzmannstrasse 2, 85748, Garching, Germany

This project takes place in the Remote Handling (RH) activities for the next step of the fusion reactor ITER. The aim of the R&D program performed under EFDA workprogramme is to demonstrate the feasibility of close inspection (e.g. for viewing and leak testing) of the Divertor cassettes and the Vacuum Vessel first wall of ITER. We assumed that a long reach and limited payload carrier penetrates the first wall using the 6 penetrations evenly distributed around the machine designed for the In-Vessel Viewing System (IVVS). A first phase of the project concerned the analysis of the requirements to perform a realistic operation inside the Vacuum Vessel, a conceptual and detailed design of a manipulator called IVP (In Vessel Penetrator).A scale one mock up was manufactured, focusing what is the most demanding and what requires to be demonstrated (electro mechanics in air and at room temperature). This demonstrator of an IVP module (2 degrees of freedom) was finally successfully tested and gave confidence to meet ITER requirements. In parallel, a feasibility study of limited maintenance operation under vacuum and temperature with the IVP system was performed. This study was completed and the possible applicable technologies were selected. Some of these are directly suitable for the design of IVP under ITER’s vacuum and temperature requirements but some others needed developments which were validated on proof of principles mock ups.The next step of the study is the design, manufacture and testing of a vacuum and temperature hardened prototype called Articulated inspection Arm (AIA) which is foreseen to be tested in real Tokamak conditions at Tore Supra. As well as demonstrating the potential for the application of an AIA type device in ITER, this program will also serve to explore the necessary robotic technologies applicable to ITER’s IVVS deployment system. This paper presents the whole AIA robot concept, the results of the test campaign on the prototype vacuum and temperature demonstrator module.


Corresponding Author:

Yann PERROT (1)

Robotics and Interactive Systems Unit – CEA. BP6 F-92265 Fontenay aux Roses Cedex France

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-404 ASSESSMENT OF A COOPERATIVE MAINTENANCE SCHEME FOR ITER DIVERTOR COOLING PIPE

FRICONNEAU JEAN-PIERRE, O. David CEA - LIST J.P. Martins CEA LIST J.D. Palmer EFDA CSU A. Tesini ITER Naka

EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany ITER International Team, ITER Naka Joint Work Site, 801-1, Muouyama, Naka-machi, Naka-gun, Iberaki-ken 311-0193, Japan

Divertor cassettes are components that require scheduled maintenance in the lifetime of the experimental fusion reactor ITER. During maintenance phases, the cassettes are disconnected from the reactor frame and moved outside the vessel for refurbishment in hot cells. Connection and disconnection of the cassette means removal of the electrical connector and cutting or welding of the cassette cooling pipes. The ability for the maintenance tool to cut or weld these cooling pipes therefore becomes an essential part of the maintenance process. Previous studies of the maintenance task clearly identified operations that will need to be performed remotely. They can be sorted into assembly operations: clamping of the tool, alignment or release of the pipe… and into process operations: cutting, welding, non- destructive testing of the weld quality. The latest evolutions of the ITER design focused on a reduction of the divertor cooling pipes from the initial 6” straight pipes to 2.5” bent pipes. As a consequence of this new definition, the initial maintenance scheme where maintenance is completely performed from the inside of the pipe by one carrier becomes obsolete. Reflections are now converging on a cooperative work between a carrier sent into the pipe to perform part of the maintenance operations and an orbital tool positioned by a slave manipulator mounted on the Cassette Toroidal Mover (CTM) and used to perform general maintenance tasks in the divertor. While access to the cutting location from the inside of the pipe was chosen during previous studies because the way was always clear, access to the cooling pipes with a slave manipulator mounted on the CTM is a more challenging operation. Definition of the reference maintenance scheme involves assessment of operational conditions such as: feasibility of the operation, complexity of the environment, space to operate with the manipulator in relation to its size and poor viewing conditions. Identification of the critical areas and zones where access to the pipe is critical is made by means of digital mock-up analysis of the Vacuum Vessel divertor region. Extraction of a reference working area is made after examination of all pipe ducts and cassettes arrangement. Simulations of maintenance scenario take place in this reference area with a manipulator model based on a Maestro like architecture. Accessibility towards the working area with a virtual cutting or welding orbital tool is then checked and discussed.


Corresponding Author:

FRICONNEAU JEAN-PIERRE

Robotics and Interactive Systems Unit – CEA-LIST DTSI-SCRI. BP6 F-92265 Fontenay aux Roses Cedex France

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-422 RF TESTS OF THE ELECTRICAL INSULATIONS FOR THE TOROIDAL STRUCTURES OF RFX

Sonato Piergiorgio, A.Masiello, G.Mella, C.Taccon

The new load assembly of RFX is characterized by a 3mm thick copper shell composed of two halves joined on the outer equatorial plane by copper plates, which short-circuit the gap on the outer side, whereas on the inner it remains electrically insulated. In the region of the single poloidal gap, two copper layers overlap each other for about 23 toroidal degrees and the electrical insulation between them is secured by a 2mm PTFE layer. The maximum toroidal loop voltage attainable with the magnetizing winding of RFX is 700V and in the case of fast plasma current termination the observed peak reaches roughly the same value. The maximum poloidal loop voltage is instead about one order of magnitude lower than the toroidal one. All the insulating gaps of the shell need to be tested during the assembly phases of the torus, to verify the good conditions of the insulation. In particular the poloidal gap on the shell shall be tested at least at 1kV. Since the toroidally shaped shell is a single conducting structure with only one insulated gap in both direction (poloidal and toroidal), it is not possible to apply an electric potential difference across the poloidal gap by means of a conventional DC or AC generator. Different methods have been studied, based on a capacitor bank discharge, but the results showed that the pulse duration and voltage level required for the test would impose the realization of an expensive custom-made high voltage circuit. On the other hand it was found that, using a relatively small RF amplifier (100 W) tuned at the natural frequency of resonance of the shell, sufficiently high voltage could be reached. This innovative system was initially set up on the old RFX shell, which has approximately the same geometry of the new one, to select the best matching equipment for the RF amplifier and to verify both the maximum attainable potential difference at the poloidal gap and the ground effect. In fact it was found that the shell behaves like a magnetic antenna and that the ground proximity reduces strongly the electromagnetic field, thus also the voltage that can be applied to the poloidal gap. The RF system was then successfully used for the acceptance test of the insulating poloidal gap of the shell at more than 1 kV, while 40W was the maximum supplied RF power. These tests demonstrated that the RF method for applied-voltage tests is of general use and it can be applied to closed conducting structures with a single insulating gap.


Corresponding Author:

Sonato Piergiorgio

Consorzio RFX, Corso Stati Uniti, 4 – I35127 Padova - Italy

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-435 OPERATIONAL EXPERIENCE FEEDBACK IN JET REMOTE HANDLING

DAVID Olivier (1), A.B. Loving (2) J.D. Palmer (3) S. Ciattaglia (3) J.P. Friconneau (1)

(1)Robotics and Interactive Systems Unit CEA BP6 F-92265 Fontenay aux Roses Cedex France (2)UKAEA Fusion Association, Culham Science Centre, Abingdon Oxfordshire, OX 14 3DB (3) Close Support Unit Garching –Boltzmannstr. 2, D-85748 Garching Germany

The radiation levels expected in ITER during the latter stages of machine operation are such that maintenance work cannot be carried out by human intervention. Remote Handling was therefore defined by the ITER collaborators at the beginning of the project, as the nominal solution for the maintenance of the reactor. The feedback provided by RH platforms developed for ITER within the L6 and L7 projects such as the Divertor Test Platform (DTP) at Brasimone (Italy) and the In Vessel Transporter (IVT) at Naka (Japan) is clearly a step further for the definition of ITER’s RH in several fields. But one has to admit that there is still a significant amount of work to be done, especially considering the evolution in the machine design from ITER’98 to ITER 2001. JET is the only operating platform within fusion where RH techniques have been developed to a stage that allows in-vessel maintenance work to be carried out fully remotely. JET’s RH team developed a methodology and a rational approach that helped them to succeed in the task. JET’s experience clearly shows that the gap between the first prototype and its upgrades to make it ready for operational use and perform maintenance work in a safe and repeatable way has a manpower cost which is often under-estimated just like the aspect of having a local team that you can rely on to solve any occurring problems and develop new applications. As a simple rule of thumb, one can assume that following its original design and manufacture, the effort required to properly prepare the RH equipment for real operations involved as much work as creating the equipment in the first place. This paper presents some general rules that can be used to make the distinction between the needs for hands-on activities and Remote Handling according to JET’s experience and ITER needs and makes a summary of the experience gained by JET people during the development and operation of their RH equipment which could be directly applied to ITER. Finally starting from the JET example, this document tries to give ITER some references for the evaluation of the amount of work and of the manpower cost that is really needed for a complete Remote Handling system to become fully operational and reliable.


Corresponding Author:

DAVID Olivier (1)

Robotics and Interactive Systems Unit – CEA BP6 F-92265 Fontenay aux Roses Cedex France

- G - Vessel-in vessel Engineering and Remote Handling.

P4T-G-509 IGNITOR PLASMA CHAMBER STRUCTURAL DESIGN WITH DYNAMIC LOADS DUE TO PLASMA DISRUPTION EVENT

Cucchiaro Antonio, Bianchi Aldo (2), Crescenzi Claudio (1), Linari Mauro (3), Lucca Flavio (3), Marin Anna (3), Mazzone Giuseppe (1), Parodi Bruno (2), Pizzuto Aldo (1), Ramogida Giuseppe (1), Roccella Massimo (1), Prof. Coppi Bruno (4)

(1) Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy (2) Ansaldo Ricerche, Corso Perrone 25, 16152 Genova (GE), Italy (3) L. T. Calcoli, Piazza Prinetti 26/B, 23805 Merate (LC), Italy (4) MIT, 02139 Cambridge (Ma), USA

The new reference plasma disruption for IGNITOR produces a significant increase of electromagnetic (EM) loads requiring a dynamic elastic-plastic structural analysis of the IGNITOR plasma chamber (PC). The EM loads due to the worst disruption event (VDE) was calculated using the MAXFEA 2D code. The uniform 26 mm PC thickness (as envisaged in the previous design) did not comply with the new E.M. requirements in term of stresses and deformation. New Design Plasma Chamber wall thickness distribution has been defined approaching a step by step optimization with the aim to minimize the vertical displacement complying with the allowable plastic strains. The analysis is non-linear, due to boundary conductions and material properties, and it’s necessary to modelling the entire (360 ) PC structure because the various load components are distributed with different poloidal periodicity. As result the IGNITOR PC is capable to withstand, according to the ASME III code rules, several hundred of cycles under plasma disruption conditions. The main results of the analysis demonstrate that the plastic deformation is below the ASME limits and the maximum permanent displacement is limited to few millimeters.


Corresponding Author:

Cucchiaro Antonio

Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy

- H - Fuel Cycle.

P1C-H-17 ADVANCED PROCEDURES FOR TWO-STAGE REPETITIVE PELLET INJECTOR.

Pavarin Daniele, Francesconi Alessandro Niero Federico Rondini Davide Angrilli Francesco

Centero of Studies and Activities for Space (CISAS G.Colmbo) University of Padua Via Venezia 15 35131 Padova Italy

Two stage pellet injector for fusion experiments are powerful machine to accelerate refuelling pellet at very high velocity ensuring penetration in the plasma core also through very dense and energetic plasmas. Besides the possibility of accelerating pellet at very high velocity they are effected by several problems as: gas following the pellet, which may contaminate the plasma, projectile strength and system reliability. CISAS has developed a two-stage unit based on the concept of fusion injector which is able to accelerate plastic projectile weighting 100 mg at 5.5 km/s. The achievement of this goal required the set-up of advanced control and diagnostic procedures and new technology solutions which may become useful for the development of the future fusion pellet injectors. Particularly CISAS Gun implements high release pressure check valve which could be useful to reduce the gas following the projectile, diagnostic procedure are applied to check the gun status with no needs of disassembling the unit, and finally active piston techniques are applied to control the shape of the pressure profile behind the projectile. In the paper a detailed description of the procedure set-up in the CISAS facility is presented and a possible application to future fusion experiments discussed.


Corresponding Author:

Pavarin Daniele

CISAS University of Padua Via Venezia 15 35131 Padova Italy

- H - Fuel Cycle.

P1C-H-38 STUDIES OF PELLET DELIVERY AND SURVIVABILITY THROUGH CURVED GUIDE TUBES FOR FUSION FUELING AND IMPLICATIONS FOR ITER

Combs, Stephen, L. R. Baylor, C. R. Foust, T. C. Jernigan, and D. A. Rasmussen

Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6169

Injection of solid hydrogen pellets will be the primary technique for depositing fuel particles into the core of ITER burning plasmas. Gas fueling is calculated to have much lower fueling efficiency in ITER than present day experiments, thus pellet fueling will be particularly important. For many years, pellets have been injected from the outside mid-plane of experimental fusion devices (magnetic low-field side for tokamaks), and researchers strived for high pellet speeds (up to ~4 km/s) to achieve deep pellet penetration and better fueling efficiency [1]. In the past several years, pellet technology development and experiments have concentrated on pellet fueling from the magnetic high-field side for tokamaks [2], with significantly deeper mass deposition and improved fueling efficiency observed at relatively low pellet speeds (<300 m/s). These injection schemes require the use of curved guide tubes to route the pellets from the acceleration devices to the inside wall launch or vertical launch locations; and thus, the speed must be limited to ensure pellet survivability. To more thoroughly understand and document the capability of curved guide tubes for the alternative injection schemes, experimental studies have been carried out at the Oak Ridge National Laboratory (ORNL), including mock-up tests of guide tube installations for DIII-D, JET, LHD, and FIRE. For the actual installations on the machines and fueling experiments, the speed limits have proved to be in good agreement with the results from the ORNL mock-up tests. For these installations, the speeds of deuterium pellets must be limited to a few hundreds of meters per second for reliable delivery of intact pellets. Presently, an experimental simulation of the ITER guide tube installation for inside pellet launch is being set up in the lab and will be tested with 3-mm deuterium pellets. The test results from the previous mock-ups will be summarized in the paper, and the new data from the ITER mock-up will be presented and discussed. The implications of these results for ITER pellet fueling will also be discussed. [1] A. Géraud et al., Proc. 20th Symposium on Fusion Technology (1998) 941–944. [2] P.T. Lang et al., Phys. Rev. Lett. 79 (1997) 1487–1490.


Corresponding Author:

Combs, Stephen

Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6169

- H - Fuel Cycle.

P1C-H-93 PELLET INJECTORS FOR STEADY STATE FUELLING

Vinyar Igor, A.Géraud(2), H.Yamada(3), A.Lukin(1), R.Sakamoto(3), S.Skoblikov(1), A.Umov(1) Y.Oda(4), G.Gros(2), I.Krasilnikov(1), P.Reznichenko(1), V.Panchenko(1)

(1)PELIN Laboratory,20a,Berezhkovskaya Emb.,Moscow,123995,Russia (2)Euratom-CEA,CEA/Cadarache,F-13108,St-Paul Lez Durance,France (3)National Institute for Fusion Science,322-6,Toki,509-5292,Japan (4)Mitsubishi Heavy Industries,Kobe,652-8585, Japan

Selected also for oral presentation O1B-H-93

Current successful operation of TORE SUPRA and LHD, as well as ITER in the future, should be supported by pellet injection in order to produce high performance plasmas with discharge durations up to 1000s. This paper presents pneumatic pellet injectors and its implementation on the LHD and TORE SUPRA, and a new centrifuge injector for steady state fuelling. All pellet injectors are based on the screw extruder technology developed by the PELIN Laboratory. This feed system, well suited for continuous operation, is coupled to a pellet cutting system forming a steady state pellet generator. Two conventional launchers are used for pellet acceleration: pneumatic gun and centrifuge. A distinguishing feature of the LHD pellet injector is a cooling system based on two cryorefrigerators (3 W at 4.2 K in total). The pellet injection is available 4 hours after the start of cooling down from the room temperature. Hydrogen is continuously pushed out from the extruder in the shape of cylindrical ice rod of 2 or 2.5 mm diameter with the maximum rate of 15 mg/s. Pellets can be cut off from the rod and ejected at frequency up to 11 Hz and velocities from 150 to 650 m/s with a reliability better than 99%. The amount of gas is suppressed to less than 1.5 Pam3 per one pellet launch. The injector has been employed for LHD plasma fuelling and demonstrated quasi-steady state high density operation (8 x 1019m-3) for 2 s. This duration is limited by heating capability not by pellet injection. The TORE SUPRA pellet injector was equipped with a pellet size regulator for real-time fuelling adjustment during the plasma discharge using feedback control data. Pellet diameters are 1.7 mm, but pellet lengths can be changed from 1.5 to 3.5 mm. Operational injector characteristics in steady state mode are: (2-6)1020 at/pellet, 120-700 m/s velocity, 0-10 Hz frequency, 98% reliability. High field side, vertical and low field side injections were performed in plasma experiments in 2003-2004. A distinguishing feature of the centrifuge injector is a curved barrel for pellet acceleration whose entrance section is aligned with the rotational axis of the rotor. So, the barrel entrance section, into which a pellet is fed, does not replace relatively to the fuel rod extruded from a screw extruder. This design allows pellets to be fed in the barrel entrance section in any moment. This injector is under testing now.


Corresponding Author:

Vinyar Igor

PELIN Laboratory, Ltd., 20a, Berzhkovskaya Emb., Moscow, 123995, Russia

- H - Fuel Cycle.

P1C-H-107 JET CONTRIBUTIONS TO THE ITER FUEL CYCLE ISSUES.

C. Grisolia,

Since 2000, the JET fusion machine is operated in the frame of European Fusion Development Agreement. It is the unique world fusion device licensed for operating with tritium fuelling and beryllium. A dedicated task force on fusion technology (TF-FT) aims at answering to ITER relevant technological issues making use of JET facilities and of the related operating experience. In the frame of tritium in the tokamak, surface analyses allow an estimation of Be and C deposition on plasma facing components. According to the obtained results, Be is transported towards the upper tiles of the inner divertor where it is stacked. Carbon, after deposition, is re-eroded through chemical sputtering and transported towards the inner flat tiles. Improved calibrations indicate that the total Be deposited in the inner divertor is of the order to 35-40 g leading to total carbon influx of 570 to 870 g. This value fits by less than a factor two to the spectroscopic evaluation. Surface analyses dedicated to 13C characterisation just after 13CH5 puffing plasma experiments show that 40% of the 13C is trapped in the inner divertor tiles confirming the flow in the scrape off layer. Estimation of ITER in-vessel tritium retention have shown that overnight in situ detritiation is needed during operation. Detritiation processes based on laser or flash lamp are being investigated for JET configuration. According to laser studies, ablation energy threshold is of 0.4 J/cm2 for a co-deposited layer whereas is of 1J/cm2 for graphite surfaces. A removal rate of 1 m2/hour is obtained for 20 µm co-deposited layer from TEXTROR and Tore Supra. Flash lamp in situ tests are scheduled in April 2004 to confirm ex-situ results (cleaning rate > 3m2/hour for 50microns). On gas exhaust control, the JET Active Gas Handling System has been used to test a prototype cryosorption panels during Trace Tritium Experiment. Results will be addressed in the paper. In the frame of the design project of a water detritiation facility for JET, key components for such a system are being studied. The most promising catalyst/packing mixtures for the Liquid Phase Catalytic Exchange (LPCE) column are currently tested in endurance tests. After recalling briefly the TF-FT activities and especially those devoted to ITER licensing involving Plasma Facing Components, Tritium processes and safety, this paper will be focused on the presentation of a comprehensive picture of the Tritium fuel cycle in a fusion facility.


Corresponding Author:

C. Grisolia

Assoc. Euratom-CEA sur la Fusion, CEA Cadarache, DSM/DRFC/STEP, 13108 Saint Paul Lez Durance, France

- H - Fuel Cycle.

P1C-H-153 COMPARISON OF MODELLING OF TRITIUM RELEASE FROM CERAMIC BREEDER MATERIALS

Munakata Kenzo, Yokoyama Yoshihiro (1) Penzhorn R. -D. (2) Oyaidzu Makoto (3) Okuno Kenji (3)

(1) Kyushu University, Kasuga 816-8580, Japan (2) Research Center Karlsruhe, Tritium Laboratory Karlsruhe, 76021 Karlsruhe, Germany (3) Shizuoka University, Radiochemistry Research Laboratory, Faculty of Science, Szuoka, 422-8529, Japan

In most current designs of D-T fusion reactor blankets employing ceramic breeder materials, the use of a helium sweep gas containing 0.1 % of hydrogen is contemplated to extract tritium efficiently via isotopic exchange reactions. However, at lower temperatures, the release process of tritium from the breeders is dominated by the desorption of tritiated water and is therefore rather slow since the rate of isotope exchange reactions is considerably low. With this background, the authors investigated the effect of water vapor on the releases of tritium from a lithium silicate ceramic breeder material. Out of pile tritium release experiments were conducted using the ceramic breeder irradiated in a research reactor. Tritium was released from the lithium silicate breeder material using a nitrogen sweep gas with 0.1 % of water vapor. As a result, it was found that the addition of water vapor to the sweep gas greatly enhances the release rate of tritium from the ceramic breeder. These are probably caused by the acceleration of the exchange reaction that takes place on the surface of the breeder material. The result of tritium release experiment with the wet sweep gas was analyzed using several numerical models. The surface effect was eliminated in the model, since the surface reactions were thought being significantly enhanced by the wet gas purge. The results of the analysis indicate that simple linear driving force model or diffusion model cannot reproduce the experimental tritium release curve. Thus, the trapping of tritium caused by trapping sites was also considered in the model. Then, it was found that both of the models with the trapping site effect well reproduce the experimental tritium release curve. Moreover, the authors tested another model in which a resistance between the bulk crystal phase and the surface of the crystal grain was considered. It is known that the disordered layer close to the surface becomes a resistance to the migration of hydrogen in metals. As a result, this model was also found to reproduce the experimental tritium release curve. In the presentation, the details of the model were explained, and the reproducibility of the experimental result was compared.


Corresponding Author:

Munakata Kenzo

Kyushu University, Interdisciplinary Graduate School of Engineering Science, Kasuga 816-8580, Japan

- H - Fuel Cycle.

P1C-H-217 ASSESSMENT OF THE ITER DWELL EVACUATION

Wykes Michael, Federici Gianfranco

ITER International Team, Boltzmannstrasse 2, D-85748 Garching, Germany

During the “dwell” period between ITER plasma current pulses the pressure in the plasma chamber has to be pumped down to a level of ~ 0.1 mPa to allow for an orderly pre-fill and breakdown to initiate the subsequent plasma current pulse. The shortest reference evacuation time corresponds to the dwell between successive ITER 400 s burn pulses, each dwell being 1400 s long. The dwell evacuation is dominated by the outgassing from the plasma-facing components, the majority of which are armoured with beryllium (~700 m2), with a minority in CFC on the lower divertor vertical targets (~50 m2). During plasma discharges, impinging deuterons and tritons load the implantation layer of the armour to near-saturation conditions. From previous experimental measurements and theoretical studies, it is known that the outgassing rate decays from an initial value that depends on the armour material and temperature, and the energy and implantation time of the incident particles, according to a power law in elapsed time (t-n), the exponent n being ~0.7. During the subsequent dwell evacuation, the implanted atoms are desorbed and constitute the main load on the primary torus cryo-sorption pumps, particularly during the latter stage of evacuation when the pressure is low. The results of parametric studies are reported to assess the effect of various factors that may affect the outgassing rate (e.g., mixing of material, effect of temperature). These latter results delineate the domain in which the terminal pressure is acceptable and indicate the amount of additional pumping that will be provided by the neutral beam cryo-sorption pumps in order to attain the required dwell terminal pressure under the most adverse, but realistic, outgassing characteristics.


Corresponding Author:

Wykes Michael

ITER International Team, Boltzmannstrasse 2, D-85748 Garching, Germany

- H - Fuel Cycle.

P1C-H-247 STRATEGY FOR DETERMINATION OF ITER IN-VESSEL TRITIUM INVENTORY

Murdoch, David, Cristescu, Ioana-Ruxandra (1); Lässer Rainer (2)

(1) TLK, Forschungszentrum Karlsruhe, Postfach 3640, D76021 Karlsruhe, Germany. (2) EFDA Garching CSU, Max Planck IPP, Boltzmannstr. 2, D-85748 Garching, Germany

Tracking of tritium inventory transfers on ITER will be essential to ensure that the safety limits established for the mobilizable tritium inventory in the vacuum vessel (VV) are not violated. The large and highly variable fuelling and exhaust flow rates, and the complex chemical composition of the tokamak exhaust stream, will compromise the precision of direct flow rate and composition measurements. Thus a complementary procedure to derive the in-vessel inventory at regular intervals is required, and this is described in the paper. In ITER, the fuel cycle, the VV and the hot cell building are within the same tritium Material Balance Area (MBA). The global tritium inventory of this MBA at any selected time can be derived from the previous determination by measuring all tritium quantities entering and leaving the MBA and calculating the quantities of tritium created (by breeding) and destroyed (by fusion reactions and decay) within it. This depends on the fact that reliable measurements and/or calculations of each of these source terms can be made, and the methods proposed for this will be outlined in the paper. At predetermined intervals the tritiated gases in all systems of the fuel cycle will be transferred to the Storage and Delivery System (SDS), and tritium quantities measured using in-bed calorimetry, to give the total inventory of mobile tritium in the fuel cycle system by addition. The calculated global inventory of the MBA less the measurable fuel cycle inventory represents the trapped tritium inventory, both inside and outside the VV. The ex-vessel portion of tritium bound in materials such as catalysts, molecular sieves, pump oils, and component walls will be determined by building up a comprehensive experimental data base on the evolution of tritium content in these materials from the start of operations, enabling the in-VV portion to be derived by difference. The procedure for transfer of the mobilizable gases to the SDS, assay of the tritium and redistribution of the gases before restart will be outlined and estimates of the time required and the achievable accuracy of the determinations presented. Although the VV and the hot cell building are within the same MBA, tritium inventories in activated components, PFC flakes and dust transferred to the Hot Cell, and in refurbished components returned to the VV, will be assessed in order to adjust the in-vessel inventory. Proposed methods for achieving this will be discussed.


Corresponding Author:

Murdoch, David

EFDA Garching CSU, Max Planck IPP, Boltzmannstr. 2, D-85748 Garching, Germany

- H - Fuel Cycle.

P1C-H-275 REQUIREMENTS AND SELECTION CRITERIA FOR THE MECHANICAL PUMPS FOR THE ITER TRITIUM PLANT

C J Caldwell-Nichols, M Glugla (1), S Welte (1), D Murdoch (2)

(1) Tritium Laboratory, Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe, Germany (2) EFDA CSU, MPI, Boltzmannstraße 2, D-85748 Garching, Germany

The tritium plant for ITER requires many mechanical gas pumps for its operation. The correct selection of pumps is important not only to meet the requirements of the various processes in terms of pressures, flow rates etc. but also must be satisfy more nebulous criteria of reliability and maintainability. In common with many ITER components, they will not be purchased for several years but will then be required to operate over the 20 years of the operational lifetime of ITER and possibly beyond this time. It must be expected that there will be failures over the lifetime of the tritium plant, so attention must be paid to the methods for maintenance, noting that all components will be installed in glove boxes or other enclosures. A good prediction of the availability of replacement pumps or components over this period will have to be made. An analysis of the requirements of all pumping stages within the tritium plant will be made and these will be compared with performance of the available pump types. Factors to be considered are whether to use single compound pumps in certain applications or several individual smaller and simpler pumps. This analysis may lead to the reduction of the types of pumps used. The types of pump motors will also be examined , noting that brushed motors are not acceptable inside glove boxes due to the creation of carbon dust. The results of the analyses and tests of candidate pumps in representative conditions (inclusion of filters, high Reynold’s number flow, long pipe lengths) will be presented.


Corresponding Author:

C J Caldwell-Nichols

Tritium Laboratory, Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe, Germany

- H - Fuel Cycle.

P1C-H-303 HIGH-POWER PULSED FLASHLAMP CLEANING OF CO-DEPOSITED HYDROCARBON FILMS FROM PLASMA FACING COMPONENTS

K.J. Gibson, G.F. Counsell (2) M.J. Forrest (2) M.J. Kay

(2) Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon,, Oxon.OX14 3DB, UK

The use of carbon-based materials for first wall components in the divertor region of tokamaks has been observed to result in the formation of significant amorphous hydrocarbon deposits on both plasma facing components and sub-divertor pumping assemblies. Such deposits, which take the form of thin films of varying morphology, could lead to a high rate of in-vessel tritium retention in future fusion devices. Laboratory based experiments at UMIST have demonstrated that photonic cleaning can provide a clean, efficient and rapid method for removing such hydrocarbon films and represents a good candidate technology for the periodic removal of co-deposits in ITER. The use of high power Xenon flashlamps as a source for this cleaning has been demonstrated in air, inert atmosphere and vacuum, with effective film removal occurring at a fluence threshold of between 1.9 and 2.5 J/cm2. The by-products of the cleaning process, both particulates and gases, have been characterised using particle sizing spectrometry and quadrupole mass spectrometry respectively. It is found that hydrogen, methane, acetylene, ethylene, ethane and carbon dioxide are the principal gaseous products produced during the cleaning process, which also produces a significant fraction of particulates in the size range 2-20mm. In an extension to this work, a series of experiments are described in which co-deposits on divertor tiles from the MAST and Asdex Upgrade tokamaks have been removed using a newly assembled flashlamp source. This source, capable of delivering instantaneous powers of up to 1GW/m2 and with pulse duration of approximately 100ms, allows a more complete study of the scaling of the removal process at higher fluences (up to 10J/cm2, well above the threshold) as well as a comparison of the relative importance to the cleaning efficiency of the UV and visible components of the flashlamp spectral output. Finally we outline plans for in-situ, full scale demonstrations of flashlamp removal of co-deposited films from the JET and Tore-Supra tokamaks. This work was jointly funded by the UK Engineering and Physical Sciences Research Council and EURATOM. The authors would also like to thank Dr Rudolf Neu of the Max-Plank Institut für PlasmaPhysik, Garching for the loan of Asdex Upgrade tiles.


Corresponding Author:

K.J. Gibson

Department of Physics, UMIST, Manchester M60 1QD, UK

- H - Fuel Cycle.

P1C-H-358 GAS PUFFING BY MOLECULAR BEAM INJECTION IN ADITYA TOKAMAK

S. B. Bhatt, Ajai Kumar, K. P. Subramanian*, P. K. Atrey, C. V. S. Rao and Aditya Team

Institute for Plasma Research, Bhat, Gandhinagar-382428, Gujarat, India * Physical Research Laboratory, Ahmedabad – 380 009, INDIA

Plasma density control is a prime requirement of tokamak plasma. Various methods are used for fuelling the gas for tokamak plasma. In normal gas puffing system, there is large angular distribution/ velocity profile in injected gas molecules. Due to this, considerable number of rundown gas molecules are adsorbed on the surface of wall and limiter. These adsorbed gas molecules are released from the surface during the plasma discharge and causing more recycling of fuel gas and some time control of density is difficult. A new method of gas fuelling in tokamak with molecular beam injection is developed for fast gas puff during the plasma discharge. The molecular beam of fuel gas is formed by the expansion of the high-pressure gas through the nozzle and skimmer. The penetration depth injected beam is more than normal gas puffing due to its energy. Due more penetration and very less divergence of beam, 1.wall loading due to fuel gas reduces causing reduction in recycling, and 2, there is an increase in the fuelling efficiency. A molecular beam injection system for Aditya tokamak is developed in house by modifying a Piezo-electric gas inlet valve and using100 µm diameter nozzle at 2 to 5 kg/cm2 gas pressure. This valve is mounted directly on tokamak. Series of experiments are performed with this molecular beam system by puffing hydrogen gas of different pulse duration, different number of pulses at various time of tokamak discharge. It is observed that there is increase, in plasma density more than 1.5 times, bolometer signal, soft X-ray and reduction in H? signal. This paper presents the results of molecular beam injection during discharge.


Corresponding Author:

S. B. Bhatt

Institute for Plasma Research, Bhat, Gandhinagar-382428, Gujarat, India

- H - Fuel Cycle.

P1C-H-438 INFLUENCE OF DEUTERIUM ON THE DESIGN OF THE JET WATER DETRITIATION SYSTEM

Cristescu Ion, I-R Cristescu 1, L. Dörr 1, M. Glugla 1, A .Bell 2, D. Brennan 2, D. Murdoch 3

1 Forschungszentrum Karlsruhe, Tritium Laboratory, Germany 2 JET, Abigdon, UK 3 EFDA CSU Garching

The development of a Water Detritiation System (WDS), i.e. configuration, performance testing of critical components and system design is essential for both JET and ITER. The WDS at JET is foreseen to process the tritiated water accumulated during operations and generated during decommissioning. The WDS for JET will be based on the Combined Electrolysis Catalytic Exchange (CECE) process employing a Liquid Phase Catalytic Exchange (LPCE) column. A direct combination of the WDS with a Cryogenic Distillation (CD) unit is also under planning. The final goal is to convert tritiated water to tritium and deuterium enriched molecular hydrogen and to further enrich tritium by cryogenic distillation of the hydrogen-deuterium –tritium mixture, followed by recovery of pure tritium in the GC based isotope separation system of the AGHS of JET. A decontamination factor of 104 for tritium is required along the striping section of the LPCE column in order to discharge and essentially tritium free molecular hydrogen isotope product into the environment. The process for the JET WDS was preliminary evaluated considering the deuterium concentration in the tritiated water to be at natural level. A detailed analysis of the direct combination of the CECE-CD processes revealed the necessity to consider also the deuterium content in the water to be processed. Therefore eleven samples from different drums with tritiated water at JET have been measured for their deuterium concentration, which was found to be in the range of 0.2-0.5% atomic ratio D/ (H+D+T). The presence of deuterium in tritiated water to be processed will change the tritium distribution on molecular species. Therefore instead of two molecular species, such as H3 and HT when deuterium is negligible, in the CECE and CD processes five molecular species such as H3, HT, D2, HD, DT has to be considered. The constant equilibrium and separation factor for CECE and CD processes are very different from one molecular specie to another which means that the length of the LPCE and CD columns depends upon the molecular species which contain tritium in order to achieve a required decontamination factor of tritium. Based on these measurements the interface between the CECE and CD was evaluated in detailed and the optimum values for deuterium and tritium composition at this interface were established.


Corresponding Author:

Cristescu Ion

Forschungszentrum Karlsruhe GmbH, Postfach 36 40, 76021 Karlsruhe

- H - Fuel Cycle.

P1C-H-441 EXPERIMENTAL VALIDATION OF A METHOD FOR PERFORMANCE MONITORING OF THE FRONT-END PERMEATORS IN THE TEP SYSTEM OF ITER

B. Bornschein, M. Glugla (1) K. Guenther (1) T.L. Le (1) K.H. Simon (1)

(1) Forschungszentrum Karlsruhe, TLK, P.O. Box 3640, D76021 Karlsruhe, Germany

The Tokamak Exhaust Processing (TEP) system within the Tritium Plant of ITER need to be designed such that tritium is recovered from all exhaust gases produced during different modes and operational conditions of the vacuum vessel. The reference process for the TEP system of ITER is called CAPER and comprises three different, consecutive steps to recover hydrogen isotopes at highest purity for direct transfer to the cryogenic Isotope Separation system. A tritium removal efficiency of about 1E+8 is required for TEP and is regularly achieved in experiments with the semi-technical facility CAPER at the Tritium Laboratory Karlsruhe (TLK). Expressed in terms of tritium concentrations the decontamination required by TEP corresponds to an outlet concentration of about 1E-4 g/m^3 (1 Ci/m^3). The CAPER process developed at the TLK employs a Pd/Ag permeator battery as the 1st step to separate more than 95% of the un-burnt DT fuel from impurities like helium, hydrocarbons and water. These so called front-end permeators have to cope with a feed flow rate of about 80 mol/h per 1 m^2 effective surface area. They have to be operated under conditions that avoid coking of the permeation membranes by hydrocarbon cracking, since this process would lead to a reduction of the effective surface area and therefore to a reduction of the performance of the permeator. In a series of tritium experiments with the CAPER facility at TLK a method has been developed to determine the actual hydrogen isotope permeability of the front-end permeator. During this experimental campaign the permeator has been operated with DT (typically 50%T) mixed with tritiated methane under conditions that promote coking by hydrocarbon cracking. The reduction of the permeator performance has then been determined by measurements of so-called break-through curves. The front end permeator could all the times be successfully regenerated after coking the membranes. The experimental results will be presented and the feasibility of performance monitoring of the ITER front-end permeators will be described. Details of the regeneration process will be reported and possible consequences for the design of the TEP system will be discussed.


Corresponding Author:

B. Bornschein

Forschungszentrum Karlsruhe, TLK, P.O. Box 3640, D76021 Karlsruhe, Germany

- H - Fuel Cycle.

P1C-H-461 PROTECTION OF THE PRIMARY CIRCUITS AND EFFECT ON THE DESIGN OF THE INNER DEUTERIUM / TRITIUM FUEL CYCLE OF ITER

M. Glugla, C. Caldwell-Nichols (1), I.R. Cristescu (1), L. Doerr (1), G. Hellriegel (1), D. Murdoch (2), P. Schaefer (1)

(1)Forschungszentrum Karlsruhe, Tritium Laboratory, PO Box 3640, D 76021 Karlsruhe, Germany (2)EFDA CSU, MPI fuer Plasmaphysik, Boltzmannstr. 2, D 85748 Garching, Germany

The inner deuterium / tritium fuel cycle of ITER comprises different, but strongly interlinked subsystems, namely the Tokamak Exhaust Processing (TEP) system, the Isotope Separation System (ISS) and the Water Detritiation System (WDS), the Storage and Delivery System (SDS) together with the Long Term Storage (LTS) and the Analytical System (ANS). Detailed Process Flow Diagrams (PFD’s) and even Pipe and Instrumentation Diagrams (P&ID’s) have been prepared and were included in the Final Design Report of ITER (2001). For each of the subsystems an initial Failure Mode and Effect Analysis (FMEA) was carried out individually. An Outline Flow Diagram for the inner fuel cycle was prepared for design integration and eventually an FMEA considering the inner fuel cycle as a whole became available. Confinement of tritium is certainly the ultimate safety goal. However, the protection of components such as sensors, pumps or vessels against over-pressure or over-temperature is of great concern, even at levels significantly below values at which the sensors or components would loose their mechanical integrity. The design shall take into account the necessity to validate and test the protection measures, noting the contamination of the primary system with tritium and the restricted access due to secondary containments. In view of primary safety, the subsystems of the ITER fuel cycle have been designed on the basis of the safety philosophy established and practiced at the Tritium Laboratory Karlsruhe (TLK). However, modern international standards for functional safety management such as the IEC 61508 are now available and should be evaluated for applicability in the ITER Tritium Plant. The general philosophy and the principles for over-pressure and over-temperature protection within the design of the inner fuel cycle of ITER will be explained in detail. Examples will be presented for selected subsystems.


Corresponding Author:

M. Glugla

Forschungszentrum Karlsruhe, Tritium Laboratory, PO Box 3640, D 76021 Karlsruhe, Germany

- H - Fuel Cycle.

P1C-H-467 EVALUATION OF SUPER CRITICAL HELIUM AS A COOLANT FOR DIII-D TYPE CRYOCONDENSATION

Baxi, C.B.,

DIII–D tokamak uses three cryocondensation pumps for plasma density control. Each DIII–D pump consists of a series of concentric stainless steel tubes. The pumping surface is a 10 m long 25 mm diameter stainless steel tube. The pumping surface of each of the three cryocondensation pumps is about 1 sq m in area and is maintained below 5 K by cooling with a two phase helium (1.3 atm, 4.35 K). The two-phase helium (TP) was chosen for DIII-D because it is available on DIII-D site and is used for NB and other applications. The pumping speed is about 30000 l/s per pump. The three pumps inside DIII-D have performed as expected for last several years. Super conducting machines under construction such as KSTAR and SST-1 have supercritical (SC) helium available on site and would prefer to use it for cooling the cryocondensation pumps. The typical condition of the available helium is 0.4 MPa (3.94 atm) pressure and 4.2 K temperature. The design of DIII-D cryocondensation pump is simple, robust, inexpensive and reliable. This study was undertaken to evaluate if the supercritical helium can be used as a coolant for GA design of the cryocondensation pump. Thermodynamic, thermal hydaulic and stability evaluation was done. It is concluded that with super critical helium a flow rate of 50 to 60 g/s (compared to 5 to 10 gm/s with two phase helium) will be required to achieve a similar performance. The co-axial insert used in DIII–D helium panel will not be required with SC helium.


Corresponding Author:

Baxi, C.B.

General Atomics, P.O. Box 85608, San Diego, California 92186-5608

- I - Materials Technology and Breeding Blankets.

P1C-I-1 USE OF THE SPIRAL 2 FACILITY FOR MATERIAL IRRADIATIONS WITH 14 MEV ENERGY NEUTRONS

MOSNIER Alban, R. Anne (1) Y. Huguet (1) X. Ledoux (4) M. Lipa (2) Ph. Magaud (2) G. Marbach (2) F. Pellemoine (1) D. Ridikas (3) M.G. Saint-Laurent (1) A.C.C. Villari (1)

(1) GANIL, BP 55027, 14076 Caen, France (2) CEA/DSM/DRFC, CEA/Cadarache, 13108 Saint Paul Lez Durance, France (3) CEA/DSM/DAPNIA, CEA-Saclay, 91191 Gif-sur-Yvette, France (4) CEA/DAM/DPTA, BP 12, 91680, Bruyeres-le-Chatel, France

The primary goal of an irradiation facility for fusion applications will be to generate a material irradiation database for the design, construction, licensing, and safe operation of a fusion demonstration reactor (e.g., DEMO). This will be achieved through testing and qualifying material performance under neutron irradiation that simulates service up to the full lifetime anticipated in the demonstration fusion reactors. Preliminary investigations of 14 MeV neutron effects on different kinds of fusion material could be assessed by the SPIRAL 2 project at GANIL (Caen-France) with first beams expected by 2009. This would allow to prepare and validate experiences which are scheduled in IFMIF during 2010-2015. It concerns e.g. small specimen size optimisation based on FEM calculations, micro toughness modelling and qualification of small specimen test technology towards accepted standards. In SPIRAL2, a deuteron beam of 5 mA and 40 MeV interacts with a rotating carbon disk producing high energy neutrons (in the range between 1-40 MeV) via C(d,xn) reactions. This facility, which will produce neutron-rich fission fragments for RNB physics studies, could be used for 3-4 months a year for material irradiation purposes. Estimations, taking into account this exposure time for a fusion dedicated irradiation plug, lead to damage rates in the order of 1-2 dpa/y (in Fe) in a volume of ~10 cm3. Therefore the use of miniaturised specimens is essential in order to effectively utilize the available irradiation volume in SPIRAL2. Sample package irradiation temperature would be in the range of 250 C to 1000 C. The irradiation level of 1 dpa/y with 14 MeV neutrons (average energy) may be interesting for micro-structural and metallurgical investigations (e.g., mini-traction, small punch tests, etc.) and possibly for the understanding of specimen size/geometric effects of critical material properties. Due to the small test cell volume, sample in situ experiments are not foreseen. However sample packages would be, if required, available each month after transfer in a special hot cell on site. The SPIRAL 2 project as well as the possible implementation of the dedicated area for material irradiations is briefly presented including expected irradiation characteristics.


Corresponding Author:

MOSNIER Alban

CEA/DSM/DAPNIA, CEA-Saclay, 91191 Gif-sur-Yvette, France

- I - Materials Technology and Breeding Blankets.

P1C-I-10 SCIENTIFIC AND TECHNICAL FOUNDATIONS AND TECHNOLOGIES OF REDUCTION OF MHD-RESISTANCE OF DUCTS WITH HEAVY LIQUID METAL COOLANTS IN MAGNETIC FIELD OF BLANKET AND DIVERTER OF TOKAMAK

Pinaev Sergey, Beznosov Alexandr (1) Muraviev Evgeni (2) Orlov Yury (3)

(1) Nizhny Novgorod State Technical University, Minin st. 24, 603600, Nizhny Novgorod, Russia (2) Research and Design Institute of Power Engineering, P.O. Box 788, 101000, Moscow, Russia (3) IPPE, Bondarenko sq. 1, 249020, Obninsk, Kaluga Region, Russia

Developing of liquid metals as coolants of blanket and diverter of tokamaks can lead us to choose coolant with higher safety standards than lithium. Heavy liquid metal coolants such as lead, gallium, eutectic lead-bismuth and lead-lithium ensure higher safety because they don't burn in the air and don't react with water and steam like alkaline metals. For cooling diverter channels possible choice is gallium, lead based coolants are candidates for blankets. Electroinsulating coating formation on the inner surface of ducts is an efficacious solution of high MHD-resistance problem. Heavy liquid metal coolants facilitate formation of oxygen based electroinsulating coating and help maintain their stability. Structure of coating is oxygen-containing compound of coolant, structural material and coolants impurities. Research of MHD flow of heavy liquid metal coolants in a transverse magnetic field and methods of MHD-resistance reduction by electroinsulating coating due to oxide layer formation on the inner surface of ducts are carried out in the department of “Nuclear and Thermal Power Stations” of the Nizhny Novgorod State Technical University. Formation of oxygen electroinsulating coating were executed by two main methods of oxygen delivery to pipes surface: injection of oxygen-containing gas mixture into the coolant flow; leading oxygen-containing gas mixture into expansion vessel over the free surface of coolant, with further inflow of oxygen into the coolant and delivery with flow of coolant to surfaces of structural material. Content of “free” oxygen into the coolant was controlled by galvanic concentration cell. Conclusions of the most effective methods of electroinsulating coatings formation are based on results of direct measure of MHD-resistance of different heavy liquid metal coolants. It has been proven experimentally that value of MHD-resistance of heavy liquid metal flow in round steel ducts with formed electroinsulating coatings in transverse magnetic field is between theoretical values for conductive walls and fully nonconducting walls. Electroinsulating coatings created in heavy liquid metal coolants are able to decrease the value of MHD-resistance by more than 5 – 10 times (depending on the coolant and the technology of coatings formation).


Corresponding Author:

Pinaev Sergey

Nizhny Novgorod State Technical University, Minin st. 24, 603600, Nizhny Novgorod, Russia

- I - Materials Technology and Breeding Blankets.

P1C-I-26 EFFECT OF UNDERSIZED SOLUTE ATOMS ON MICROSTRUCTURE CHANGE

Ryazanov Alexander, V.A.Egorov-a, H.Matsui-b

-a-Russian Research Center” Kurchatov Institute”,123182, Moscow, Russia, -b- Institute for Materials Research, Tohoku University, Katahira 2-1-1, Aoba-ku, Sendai 980-8577, Japan

Vanadium-based alloys are considered as one of candidate structure materials for fusion reactors and so understanding of physical mechanisms of an effect of solute atoms in these alloys on microstructure change is very important for development of fusion material technology. Experimental investigations show that in irradiated binary vanadium alloys V-A (A=Fe, Cr and Si) the number densities of self interstitial (SIA) loops are found to be much higher that in pure vanadium. This indicates that solute atoms trap SIAs and enhance dislocation loop nucleation. In the present paper, the di-atomic cluster nucleation model is extended to describe the formation of SIA loops in irradiated binary vanadium alloys, including the effect of undersized solute atoms on SIA loop nucleation and growth. In this model undersize solutes are considered to have strong binding with SIAs and can act as the loop nucleation sizes. The suggested model takes into account also the effect of solute segregation to loops and dislocation lines. The segregation of impurity atoms (undersized solute atoms) at dislocation lines and SIA loops modifies the dislocation bias, the sign of the bias correction being opposite to that of impurity misfit. Such bias modification affects the nucleation and growth SIA loops too. The influence of these two factors on nucleation and growth dislocation loops in binary vanadium alloys are presented here. It is shown that under irradiation the density of dislocation loops increases with increasing concentration of undersized solute atoms and growth kinetics of SIA loops in these alloys is different too. The predictions of the model and performed numerical calculations are compared with observed experimental data on dislocation loop formation and growth under electron irradiation. It is found that the model is able to describe the main features of the experimentally observed nucleation and growth of SIA loops in binary vanadium alloys.


Corresponding Author:

Ryazanov Alexander

Russian Research Centre"Kurchatov Institute",123182,Moscow,Kurchatov Sq.1,Russia

- I - Materials Technology and Breeding Blankets.

P1C-I-39 RADIATION INDUCED CONDUCTIVITY AND SURFACE ELECTRICAL DEGRADATION OF PLASMA SPRAYED SPINEL FOR NBI SYSTEMS

A. Moroño, and E.R. Hodgson

Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain

Plasma sprayed spinel is being considered as a possible candidate material for the insulator rings of the ITER neutral beam injector bushing. During ITER operation the insulating material will be subjected to a radiation field due to the plasma and the NBI accelerator itself. The radiation will cause an increase in the electrical conductivity (Radiation Induced Conductivity RIC) and possibly permanent radiation induced electrical degradation both within the volume and/or on the surface. The electrical insulating rings in the bushing will have one side facing high vacuum and the other exposed to a pressurized gas (SF6, N2, or dry air). The electrical behaviour of the surface of the insulator under irradiation will depend strongly on the environment (pressurized gas or high vacuum) surrounding the material. The paper describes experiments performed to evaluate the RIC and surface electrical degradation of plasma sprayed spinel from LWK-PlasmaCeramic. The experiments have been performed in the beam line of a 2 MeV Van de Graaff accelerator, where 10x10x0.7 mm2 LWK spinel samples were irradiated in high vacuum (10-6 bar) with 1.8 MeV electrons, at dose rates from 7 to 70 Gy/s and temperatures between 50 and 350 C. The electron beam was perpendicular to the 10x10 mm2 faces. Platinum central and guard electrodes were sputtered on one of the 10x10 mm2 surfaces and a single earth electrode on the other. The experimental set-up permitted an electric field of up to 1 MV/m to be applied to the sample and to measure both volume and surface conductivity during and after irradiation . The volume RIC at 70 Gy/s and 350 C for LWK spinel is 0.30x10-9 S/m. This value was not observed to change for doses up to 10 MGy. However the material exhibits severe surface electrical degradation in vacuum when heated up to temperatures between 150 and 400 C The threshold temperature for surface degradation depends on the surface studied. For the surface perpendicular to the material growth direction the critical temperature for significant electrical surface degradation in vacuum was found to be 400 C. In the case of the surface parallel to the growth direction, the temperature for surface degradation was 150 C. As this represents the surface which will be exposed to the vacuum, this material should not be used as an electrical insulator in vacuum at temperatures higher than 150 C.


Corresponding Author:

A. Moroño

Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain

- I - Materials Technology and Breeding Blankets.

P1C-I-43 BLANKET MANUFACTORING TECHNOLOGIES : THERMOMECHANICAL TESTS ON HCLL BLANKET MOCKS UP

CACHON Lionel, DIEPPOIS Jean Paul* TALAND Rémi* LAFFONT Guy* POITEVIN Yves**

*CEA Centre de Cadarache, Bât. 204, DTN/STPA/LTCG, 13108 St PAUL lez Durance, FRANCE **CEA Centre de Saclay, DM2S/SERMA/LCA , 91400 SACLAY, France

In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the “HCLL blanket concept” have motivated the present study. The aim of this study, carried out in the frame of EFDA Workprogramme, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-ups under ITER TBM representative conditions.The first step of this experimental program is the design & manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling, taking the opportunity of synergies with the gas-cooled fission reactor R&D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the relevancy of the DIADEMO engineering design for HCLL thermomechanical tests . In particular, the He loop is an isobaric device with an operating pressure of 80 bar. A blower allows an He flow rate of 30 g/s, and the test section is fed by helium at a temperature between 300 and 400 C. After the test section the He temperature can be higher than 500 C. To optimise the power balance of the loop, an heat exchanger of 50 KW will be used. The inlet He temperature of the blower has to be lower than 60 C. So a cooler was designed to evacuate 40 KW maximum, by mean of the glycol/water loop of DIADEMO. The “DIADEMO HCLL” loop will be in operation at the end of 2004.


Corresponding Author:

CACHON Lionel

CEA Centre de Cadarache, Bât. 204, DTN/STPA/LTCG, 13108 St PAUL lez Durance, FRANCE

- I - Materials Technology and Breeding Blankets.

P1C-I-58 HIGH ENERGY PROTON DEGRADATION IN KU1 QUARTZ GLASS

CONSTANTINESCU BOGDAN, -

-

We studied the 3 and 12.6 MeV proton - room temperature - irradiation-induced modifications on ultraviolet transmission properties on KU1 quartz glass samples. We started with 3 MeV proton irradiation at room temperature. Two 0,8 mm thick samples provided by CIEMAT, Madrid have been implanted using the Bucharest HVEC Tandem accelerator at 8 „e 1013 and 1,5 „e1014 protons, respectively. The optical transmission properties (absorption, transmission and reflectivity) in the UV region have been measured with a Cary 4 VARIAN spectrophotometer. For the lower dose, absorption peaks at 215 nm and 240 nm, similar in shape, but smaller in intensity to gamma irradiation case, can be observed. For higher dose, a supplementary 202 peak appeared, splitting the 215 nm peak. The 3 MeV protons produce considerable ionization, which is the main cause of the energy loss at such low energies. The result of the overall ionization is the 215 nm band. As concerning the number of induced defects, our doses, equivalent to 4 „e 1015 p/cm2 and 7,5 „e 1015 p/cm2 produced 4 „e 10-6 and 7,5 „e 10-6 dpa, repectively, in the 3 MeV protons range in quartz (80 ƒÝm), equivalent to 5 „e 10-4 and 9 „e10-4 dpa for 1 cm. We continued our KU1 quartz glass studies using 12.6 MeV proton irradiation. The superiority of 12.6 MeV irradiation as compared to 3 MeV is evident, due to the relative homogeneity of the induced defects (vacancies produced) across the target depth. We irradiated a 0.8 mm thick KU1 sample in the same conditions as for 3 MeV irradiation, at a total dose of 2 x 1014 protons (12.6 MeV energy). We observed the presence of 215 nm peak (due to both electron and nuclear collisions stopping) and a big reduction of 240 nm peak (due only to nuclear collisions). We evaluated the dose rate of our 12.6 MeV proton irradiations at 200 Gy/s and the total irradiation doses at 10 and 20 MGy. Comparing our spectra (mainly the intensity of 215 nm peak) with the results for gamma and high energy electron irradiations, we can conclude for the 12.6 MeV proton irradiation that the saturation effect in absorption is obtained after a 10 MGy dose, as compared with 4-5 MGy for gamma and with 11-12 MGy for electrons, suggesting the ionization process is essential for defect absorption centers in all the cases.


Corresponding Author:

CONSTANTINESCU BOGDAN

INSTITUTE OF ATOMIC PHYSICS, POB MG-6, Bucharest, Romania

- I - Materials Technology and Breeding Blankets.

P1C-I-85 EXPERIMENTAL STUDY OF LITHIUM MHD FLOW IN SLOTTED CHANNEL FROM V-4TI-4CR ALLOY

Lyublinski Igor, V.A. Evtikhin, A.V. Vertkov, N.I. Ezhov, V.M. Shcherbakov

FSUE “Red Star”, 1a, Elektrolytniy Proezd, Moscow, 115230, Russia

MHD pressure drop in flowing liquid metal for a tokamak with high magnetic fields is a key concern regarding the development of lithium self-cooled test module for ITER and lithium breeding blanket for DEMO-type projects. MHD losses on liquid metal pumping can be most efficiently reduced by applying of an electrically insulating coating to the inner surface of the channels. The choice of materials and coating applying technology requires an experimental procedure for the assessment of coating characteristics in conditions close to reactor conditions. The developed method of estimation of electrically insulating coating properties on the V-4Ti-4Cr channel internal surface is based on the pressure drop measurement in liquid lithium forced circulation system with MHD test section. The tests were conducted on channels from V-4Ti-4Cr alloy with insulating coating based on AlN and without coating. The dimensions of the test section were 6×20×320 mm. Data were taken at lithium flow velocity up to 5 m/s, temperature up to 500oC and a uniform transverse magnetic field up to 1.6 T. Measured hydraulic resistance ? depending on the MHD interaction parameter Ha2/Re have shown five times reduction for coated wall in comparison with conducting wall channel. Methods of improving the electrically insulating coating characteristics on the vanadium alloys are considered.


Corresponding Author:

Lyublinski Igor

FSUE “Red Star”, 1a, Elektrolytniy Proezd, Moscow, 115230, Russia

- I - Materials Technology and Breeding Blankets.

P1C-I-88 A NEUTRONIC INVESTIGATION OF HE-COOLED LI-BREEDER BLANKETS FOR FUSION POWER REACTOR

Kim, Yonghee, Hong, Bong Guen

150 Deokjin-dong, Yuseong-gu, Daejeon 305-353, Republic of Korea

In Korea, a liquid metal blanket is being studied as an option of the ITER test blanket. The R&D efforts are tuned to He-cooled and Li-breeder blankets after assessment of various liquid metal blanket concepts. Major technical rationale for the selection is in that the concept is virtually free from the tritium (T) permeation problem and the MHD-related issues. This paper is concerned with neutronic investigation of the liquid metal blanket concepts. The He-cooled blanket is adopting a multi-layered design concept in the radial direction. Both the first wall and the breeding zone are cooled by the He coolant. In the design, a thin Li layer is placed between the first wall and the coolant in order to prevent tritium permeation into the coolant channel from the plasma zone. For an efficient T breeding, a static neutron multiplier is also introduced into the blanket. In order to minimize the neutron leakage to the vacuum vessel (VV), a neutron reflector and a neutron absorber are placed after the breeding zone. Finally, a gamma shield is put between the absorber and VV. For T recovery, the Li breeder is circulated very slowly such that the MHD pressure drop might not be an issue. Various optimization studies have been performed with a neutron transport code in a one-dimensional cylindrical geometry mainly from the neutronic point of view. As the performance measure of the blanket, three design parameters were considered: tritium breeding ratio, Li-6 enrichment (or Li volume), and the energy multiplication in blanket. Three neutron multiplier options (Be, PbO, and W) were evaluated in terms of the three performance measures. In a fusion reactor, it is crucial to maintain a self-sustaining tritium cycle. Meanwhile, energy production of the (n,T) reaction of Li-6 is quite significant in a T-self-sufficient cycle ( about 4.8 MeV per Li-6 (n,T) reaction). To investigate the impact of neutron spectrum on the blanket performance, a graphite moderator was assessed in this paper. Also, the blanket performance was also evaluated in terms of the Li-6 enrichment. In addition, conversion of Li-7 to Li-6 was investigated.


Corresponding Author:

Kim, Yonghee

150 Deokjin-dong, Yuseong-gu, Daejeon 305-353, Republic of Korea

- I - Materials Technology and Breeding Blankets.

P1C-I-96 MICROSTRUCTURAL CHARACTERISATION OF EUROFER-ODS RAFM STEEL IN THE NORMALIZED AND TEMPERED CONDITION AND AFTER THERMAL AGING IN SIMULATED FUSION CONDITIONS

Paúl, Antonio, O. M. Montes (1) E. Alves (2) L. C. Alves (2) R. Lindau (3) J. A. Odriozola (1)

(1)Instituto de Ciencia de Materiales de Sevilla. Avda. Américo Vespuccio s/n, 41092 Sevilla, Spain (2) ITN, Estrada Nacional 10, Sacavém Portugal (3) EFDA, Garching, Germany

ODS RAFM steels are promising candidates to be used as structural materials in fusion reactors, mainly due to their creep and swelling resistance. In this work we present the results of our research on the microstructure of EUROFER based ODS using different characterization techniques. Preliminary results with optical microscopy on the as-received material indicate that the microstructure is ferritic. In addition, ion microprobe studies reveal a homogeneous distribution of yttrium particles in the ferritic microstructure. The austenitisation temperature will be determined by in-situ high temperature XRD in order to find the most adequate normalization and tempering treatments to obtain a fully ODS martensitic microstructure. Samples in the normalized and tempered condition will be characterized by means of XRD, optical microscopy, SEM and TEM so that a complete characterisation of the microstructure will be obtained. Special attention will be paid to the yttrium oxide dispersion and the presence of precipitates. During operation in nuclear fusion plants the structural materials will be exposed to high temperatures. Experiments of thermal aging at 700 C in He/H gas mixture up to about 5000 hours are being performed. The microstructure after thermal aging will be compared to that of the original material. Relevant results of this research will be presented in the conference.


Corresponding Author:

Paúl, Antonio

Instituto de Ciencia de Materiales de Sevilla. Avda. Américo Vespuccio s/n, 41092 Sevilla, Spain

- I - Materials Technology and Breeding Blankets.

P1C-I-102 NON-DESTRUCTIVE ANALYSIS OF MINIATURIZED FUSION MATERIALS SAMPLES AND IRRADIATION CAPSULES BY X RAY MICRO-TOMOGRAPHY

Tiseanu Ion, Teddy Craciunescu Bogdan N. Mandache

National Institute for Laser, Plasma and Radiation Physics Plasma Physics and Nuclear Fusion Laboratory

Recently, at the Association EURATOM-MECT (Romania) a laboratory for X-ray microtomography was established with European Community support. Its research is focused on NDT inspection of miniaturized samples of fusion materials and irradiation capsules for IFMIF environment conditions. Computer-aided tomography (CAT) systems are configured to take many views (radiographies) of the object in order to build a 3-D model of its internal structure. X-ray tomography as an NDT tool for fusion material samples can provide information on: density variations, micro-cracks development by mechanical/thermal cycling, permeability of porous materials, components microstructure integrity, 3-D accurate geometrical measurements. Our tomographic facility consists of an open type microfocus X-ray source, a five axis micrometric translation/rotation manipulator and optionally large area, high resolution image intensifier or amorphous silicon flat panel as X-ray detection system. This setup permits high resolution cone-beam tomography of miniaturized samples as well as an innovative oblique view inspection of the flat samples or components as irradiation capsules, IFMIF Li-target backplate etc. 3-D tomographic reconstructions are obtained by a proprietary computer code based on a modified Feldkamp algorithm. The reconstruction software also incorporates efficient techniques for beam hardening reduction and ring artifacts elimination. By numerous experiments it was established that our system can be used for a large range of samples with regards to size, material and complexity. For the individual miniaturized samples the microtomography analysis is guaranteed for feature recognition down to few microns. A space resolution of tens of microns for irradiation capsules of around 100 mm characteristics dimension is currently obtained. The microtomography facility is available for EFDA Technology Programme. The future activities will be focused on CAT structural integrity assessment of instrumented capsules and rigs and the development of real-time micro-radiography of miniaturized samples under mechanical/thermall stress. In addition to the transmission tomography studies one presents a conceptual design of an emission tomographic system for already irradiated miniaturized samples and capsules. Numerical simulations validated by experimental tests show that the design parameters for space resolution and isotope selectivity are well within reach.


Corresponding Author:

Tiseanu Ion

National Institute for Laser, Plasma and Radiation Physics, Plasma Physics and Nuclear Fusion Laboratory, Atomistilor Str. No 111, P.O. Box MG-36, R-76900 Bucharest, Magurele, Romania

- I - Materials Technology and Breeding Blankets.

P1C-I-108 INNER STRUCTURES OF COMPRESSED PEBBLE BEDS DETERMINED BY X-RAY TOMOGRAPHY

Reimann Joerg, 1)R. A. Pieritz, 2)M. di Michiel, C. Ferrero

1) Applied Research Solutions, 15 place du Charmeyran, F-38700 La Tronche, France 2) European Synchrotron Radiation Facility,B.P. 220,F-38043 Grenoble CEDEX, France

In the Helium-cooled Pebble Bed (HCPB) blanket, beryllium in form of pebbles is planned to be used as a neutron multiplier. During operation, thermal stresses will result in a compression of these beds which influences significantly the pebble bed thermal conductivity. For the blanket design, the accurate knowledge of the dependence of this thermal conductivity on the compression state is of large importance. Currently, this dependence is measured in uniaxial compression tests considering the pebble bed as a “black box”. For the extrapolation of data and the improvement of currently available heat transfer correlations the knowledge of the number of pebble contacts and corresponding contact zones within the bed is of great interest. Experiments were performed first in the Forschungszentrum Karlsruhe where cylindrical pebble beds were pre-compressed to different strain levels in uniaxial compression tests. For higher measurement accuracy, spherical 3.5 aluminium pebbles were used instead of the (for fusion applications) usual1 mm beryllium pebbles. In the European Synchrotron Radiation Facility (ESRF) Grenoble, a special microtomography experimental setup was then used allowing the computer aided reconstruction of 3-D images of the attenuation coefficient of the X-ray synchrotron radiation beam within the pebble beds. By post-processing the data, very useful information was obtained on both radial and axial void fraction distributions in the samples as well as the detailed information on pebble contact numbers and contact zones. In the paper, the microtomographic technique as well as the first results of the analyses are presented and critically discussed in special regard to future investigations.


Corresponding Author:

Reimann Joerg

Forschungszentrum Karlsruhe, Institut für Kern- und Energietechnik, P.O. Box 3640, D-76021 Karlsruhe, Germany

- I - Materials Technology and Breeding Blankets.

P1C-I-109 THERMAL CREEP OF BERYLLIUM PEBBLE BEDS

Harsch Heinrich, Joerg Reimann

Forschungszentrum Karlsruhe, Institut für Kern- und Energietechnik, P.O. Box 3640, D-76021 Karlsruhe, Germany

Present ceramic breeder blanket designs are based on ceramic breeder and beryllium pebble beds. During operation, thermal stresses arise from different thermal expansions of the pebble beds and structural materials, and pebble bed swelling due to irradiation. Thermal creep of pebble beds will partly release the build-up of stresses, might improve heat transfer due to increased contact areas between the pebbles and might compensate a further stress build-up due to irradiation induced swelling. Therefore, the knowledge of thermal creep is of prime importance for both types of pebble beds. In the past, thermal creep investigations were restricted to different granular ceramic breeder materials In this paper, first results for beryllium pebble beds consisting of 1mm NGK pebbles are presented. These experiments were performed in the uniaxial test facility HECOP II in a temperature range between 450 and 650 C and uniaxial stresses up to 3.6MPa. Thermal creep strain was described by a correlation of the type ecr = A exp(B/T) sp tn. Compared to ceramic breeder materials, both the coefficient A and exponents p and n differ. However, it is interesting to note that for relatively short creep periods and blanket relevant temperatures, the creep strain is very similar for Li4SiO4 and beryllium pebble beds. With these new results, for the first time a complete set of data exists required for the description of the thermomechanical interaction of solid breeder and beryllium pebble beds with the structural material of blanket elements.


Corresponding Author:

Harsch Heinrich

Goraieb Versuchstechnik, In der Tasch 4a, D-76227 Karlsruhe, Germany

- I - Materials Technology and Breeding Blankets.

P1C-I-110 THERMAL CREEP BEHAVIOR OF THE EUROFER97 RAFM STEEL AND TWO EUROPEAN ODS-EUROFER97 STEELS

Nadine Baluc, Nobuyasu Nita (1) Gang Yu (2)

(1) Matsui Lab., Institute for Material Research Tohoku University Katahira 2-1-1, Aoba-ku Sendai, Japan 980-8577 (2) Fusion Technology-Materials, CRPP - EPFL , AssociationEURATOM-Confederation Suisse, 5232 Villigen PSI, Switzerland

The reduced activation ferritic/martensitic (RAFM) steel EUROFER97 and the oxide dispersion strengthened (ODS) steel ODS-EUROFER97, with the EUROFER97 as matrix material and 0.3 wt.% Y2O3 particles as reinforcement material, are foreseen to be used as structural materials in fusion power reactor at temperatures up to about 550 C and 650 C, respectively. Their creep behavior is one of the key issues for their future application. Thermal creep tests have been conducted on the EUROFER97 and two kinds of ODS-EUROFER97, which were manufactured using slightly different powder metallurgy procedures, at the Centre of Research in Plasma Physics (Switzerland) and at the CEA-Grenoble (France), respectively. Thermal creep experiments were conducted under constant stress at temperatures between 450ºC and 750ºC, in an argon flow, up to rupture. They were complemented with post-testing microscopic observations of the specimen and rupture surfaces. It was found that the ODS-EUROFER97 exhibits significantly higher creep strength than the EUROFER97, and could clearly be used 100 degrees above the EUROFER97. Creep exponents have been determined. A creep exponent of about 4 was found for the ODS-EUROFER, which is characteristic of a climb dislocation mechanism at obstacles, i.e. the Y2O3 particles. A creep exponent of about 14 was found the EUROFER97, which indicates that the stress sensitivity of the strain rate is much less for the ODS-EUROFER97 than for the EUROFER97, which is beneficial for its future use.


Corresponding Author:

Nadine Baluc

Fusion Technology-Materials, CRPP - EPFL , AssociationEURATOM-Confederation Suisse, 5232 Villigen PSI, Switzerland

- I - Materials Technology and Breeding Blankets.

P1C-I-122 SEGREGATED VOID SWELLING IN A HETEROGENEOUS MATERIAL: IMPLICATIONS FOR FUSION MATERIALS

Sergei Dudarev (1), Alexei Semenov(2) and Chung Woo(2)

(2)Department of Mechanical Engineering, Polytechnic University of Hong-Kong, Hung Hom, Kowloon, Hong-Kong

Nucleation and growth of voids and gas bubbles in materials under irradiation is known to represent one of the key factors limiting