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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



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