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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.



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