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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.



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