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

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

Corresponding Author:

Fantz Ursel

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Deirdre Boilson(2)

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Brand Peter

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Bruschi Alessandro

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

J. H. Belo (1)

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

- B - Plasma Heating and Current Drive.


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

same as above

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

Corresponding Author:


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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Speth, Eckehart

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

- B - Plasma Heating and Current Drive.


SEKI Masami, MAEBARA Sunao, MORIYAMA Shinichi and FUJII Tsuneyuki

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

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

Corresponding Author:

SEKI Masami

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

K. Vulliez

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Testoni Pietro

Electrical and Electronics Engineering Dept. - University of Cagliari

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Zaccaria Pierluigi

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Fuentes Candida

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Kasparek, Walter

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

- B - Plasma Heating and Current Drive.


Mirizzi Francesco, Maria Laura Apicella(1), Philippe Bibet(2), Giuseppe Calabrò(1), Luigi Panaccione(1), Vincenzo Pericoli Ridolfini(1), Salvatore Podda(1), Angelo Antonio Tuccillo(1)

(1)Associazione EURATOM ENEA sulla Fusione, C.R. Frascati, via Enrico Fermi 45, 00044 Frascati, (Rome) Italy (2)Association CEA-EURATOM sur la Fusion, Centre d'Etude de Cadarache, France

Selected also for oral presentation O3A-B-392

A scaled prototype of the Passive Active Multijunction (PAM) launcher actually proposed for the LHCD system of ITER, has been realised and successfully tested on FTU in the frame of a collaboration between ENEA Frascati and CEA Cadarache. A power density of about 80 MW/m2 at the launcher mouth (corresponding to 50 MW/m2 in ITER at 5 GHz) has been routinely achieved with a power reflection coefficient r ? 2% at the launcher input. A very good coupling has been obtained also with plasma density, in front of the launcher, close to the cut-off value. Direct comparisons with the performances of a conventional grill in a different FTU port, thus in the same operative conditions, are available. The PAM is characterised by thick vertical walls between adjacent columns of active (transmitting) waveguides that assure a good mechanical stiffness to the structure, while ducts drilled in these walls, allow an effective water cooling. These thermo-mechanical characteristics make the PAM launcher very attractive for the use in the harsh plasma environment of ITER. The periodicity of a conventional multijunction launcher is restored by interposing columns of passive (reflecting) waveguides between the active ones at the mouth of the PAM. The directivity of the launcher is improved at low plasma density, where the cross coupling between active and passive waveguides is increased by strong RF power reflection conditions. This allows safe but efficient operations with the launcher positioned far from the plasma scrape-off layer where also the thermal loads are smaller. A full-scale test of the technological aspects of the PAM is now in preparation on Tore Supra. The proposed LHCD launcher for ITER will couple to the plasma a total power of 20 MW at 5 GHz. The launcher is composed of four independent units, each one including 12 PAM modules arranged in 4 poloidal rows and 3 toroidal columns. Each module is made of three rows with 8 active waveguides and 8 passive ones per row. The intrinsic phase shift between adjacent columns of active waveguides in the same module is set to 270 to have an N|| (peak) = 2 and an N|| = 1.9÷2.1 by changing the feeding phase between modules on the same row in the range ±90 . The maximum RF power density at the mouth is limited to 33 MW/m2, a value that has been largely demonstrated as safe during the test on FTU. The paper reports the main results of this test and their fall-out on the main features of the proposed launcher for ITER.

Corresponding Author:

Mirizzi Francesco

Associazione EURATOM-ENEA sulla Fusione, CR Frascati Via Enrico Fermi 45, 00044 Frascati (Rome), Italy

- B - Plasma Heating and Current Drive.


C Rotti, P K Jaykumar, K Balasubramanian, A K Chakraborty, S K Mattoo and NBI team

Institute for Plasma Research, Bhat, Gandhinagar-382 428, Gujarat, India Non-Ferrous Materials Technology Development Centre, Kanchan bagh, Hyderabad-500 058, Andhra Pradesh, India

Heat Transfer Elements (HTEs) based upon Cu-Cr-Zr alloy are used for thermal management of the beamline of the neutral beam injector for SST-1. It requires a thermo mechanical treatment during its fabrication which consists of EB welding and milling. Following the recognized material production process of Cu-Cr-Zr alloy with solution heat treatment, quenching and subsequent aging for ~ 4 hrs at 470 C has yielded mechanical properties of the material which are in the low end of the published database. In this paper we show that introduction of a significant percentage of cold work on the alloy yields remarkable enhancement of mechanical properties. A cold work of 60 – 90% on solution heat treated alloy (980 C for 20 minutes) of Cu-Cr (0.8%)-Zr(0.08%) leads to optimum properties (UTS >400 MPa, YS> 300 MPa and % elongation~22%) of the alloy suitable for fabrication of HTE’s. Suitability for fabrication was benchmarked by detailed characterization of EB weld joints for joint strength, micro hardness across the joint and metallographic properties and subsequently a full scale prototype of HTE. These results shall be discussed in this paper, with a recommendation a large database may be built for this route of material production.

Corresponding Author:

C Rotti

NBI Group, Institute for Plasma Research, Bhat, Gandhinagar-382 428, Gujarat, India

- B - Plasma Heating and Current Drive.


Toigo Vanni, A. De Lorenzi, E. Gaio, F. Milani, L. Zanotto

Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova, Italy

This paper describes an alternative scheme developed for the ITER Neutral Beam Injector (NBI) Power Supply System. The main modification proposed regards the Ion Source Power Supply (ISPS), which presents quite high current levels and very low voltages. In the Reference Scheme, this power supply is divided in two sections: in the first, the required power is obtained from ground referenced power regulators and then raised to the high voltage level (-1 MV) through eight individual insulating transformers. Then the power is transmitted via an SF6 insulated HV Transmission Line to a large tank, named High Voltage Deck (HVD), insulated in high pressure SF6 as well. The HVD contains the final step-down transformers and diode rectifiers, which are connected to the Ion Source by means of a second Transmission Line. The inspection of the devices placed inside the HVD is not an easy task, due to the SF6 environment and the closeness to the neutronic area. For this reason, and to allow easier tuning and setting up of the whole system, maintenance, trouble shooting, fault inspections and implementation of further improvements, the possibility to eliminate the HVD and thus guarantee full accessibility to all the ISPS devices has been analyzed. As a result, in this alternative scheme all the ISPS components are installed inside an air insulated Faraday Cage (–1MV to ground), the HVD is removed and the power is transmitted to the ion source via a unique SF6 insulated HV Transmission Line. Only one main insulating transformer is required, placed upstream the whole system. An interesting aspect of this solution is the possibility to operate for tuning the Ion Source at lower voltage and also without the acceleration power supply, with a very easy connection between the Faraday Cage and the Ion Source; also voltage tests and system conditioning will be greatly simplified, using a test generator directly connected to the Faraday Cage. The main drawback of this solution is the fact that the high current conductors are present not only in the last part of the Transmission Line, but in the whole HV line. The paper will describe in detail the alternative power supply scheme, will discuss advantages and disadvantages with respect to the Reference Scheme, and will present the motivations behind the design choices and their implications. In particular, the impact on the HV Transmission Line structure and the general power supply system layout will be presented and discussed.


Corresponding Author:

Toigo Vanni

Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova, Italy

- B - Plasma Heating and Current Drive.


A. Serikov, U. Fischer(1), Y. Chen(1), K. Lang(1), R. Heidinger(1), Y. Luo(2), E. Stratmanns(1), H. Tsige-Tamirat(1)

(1)Association FZK-Euratom, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, P.O. Box 3640, D-76021 Karlsruhe, Germany. (2)Computer and Information College, Hefei University of Technology, Hefei, Anhui 230009, PR China

The design of an Electron Cyclotron Wave (ECW) launching system for the ITER port is currently under development by a working group from various Euratom associations. The development work includes the ECW launcher with the waveguides, the main structural components such as the port plug, shielding and frame, and the torus window serving as vacuum closure in the waveguides. The major neutronics tasks are (i) to assess the neutron streaming in the waveguide channels for proofing the design limit for the radiation load to the CVD diamond window can be met and (ii) to assess and optimize the shielding of the launching system to ensure the radiation loads to adjacent components such as the vacuum vessel and the super-conducting magnet coils are tolerable. In addition, it must be assured that the radiation dose levels during shut down periods are tolerable to allow maintenance personnel access to the area inside the cryostat surrounding the ECW launcher port. This paper presents results of neutronics analyses conducted for the ECW launching system in the ITER upper port based on the launcher design of FOM Rijnhuizen with a twisted arrangement of 8 straight waveguides. A dedicated two-step approach is used for the neutron streaming calculations with the Monte Carlo code MCNP in the ITER 3D geometry. In the first step, a surface source is calculated at the region of the ECW launcher front using the standard ITER plasma volume source. In the second step, the surface source is used for calculating the neutron flux profiles along the waveguide channels by using point detector estimators. Shielding calculations are performed with MCNP using the importance sampling technique to assess the required dimensions of the shield material around the waveguides. The radiation dose levels during shut-down periods are calculated on the basis of the rigorous 2-step (R2S) computational scheme for MCNP based shut-down dose rate calculations. With all the calculations, use is made of a standard MCNP model of ITER (20 torus sector) to which the ECW launching system was integrated. The MCNP model of the launcher was generated from a suitably modified CATIA model by conversion into the geometry representation of the MCNP code using a newly developed interface programme.

Corresponding Author:

A. Serikov

Association FZK-Euratom, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, P.O. Box 3640, D-76021 Karlsruhe, Germany

- B - Plasma Heating and Current Drive.

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