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

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

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

Corresponding Author:

Gasparotto, Maurizio

Wendelsteinstraße 1,D- 17491 Greifswald, Germany

- A - Current and Next Step Devices


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

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

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

Corresponding Author:

Morikawa Junji

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

- A - Current and Next Step Devices


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

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

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

Corresponding Author:

Tanchuk Victor

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

- A - Current and Next Step Devices.


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

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

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

Corresponding Author:

Groot de, Bart

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

- A - Current and Next Step Devices



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

Selected also for oral presentation O3B-A-330

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

Corresponding Author:


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

- A - Current and Next Step Devices


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

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

Selected also for oral presentation O3B-A-449

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

Corresponding Author:

Todd, Thomas

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

- A - Current and Next Step Devices.


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

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

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

Corresponding Author:


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

- A - Current and Next Step Devices


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

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

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

Corresponding Author:

Iida Hiromasa

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

- A - Current and Next Step Devices


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

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

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

Corresponding Author:

Guérin Olivier

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

- A - Current and Next Step Devices


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

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

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

Corresponding Author:

Fardeau Agnes

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

- A - Current and Next Step Devices


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

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

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

Corresponding Author:

Lyraud Charles

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

- B - Plasma Heating and Current Drive.


Montgomery Grimes, David Terry Ron Parker Dexter Beals

Same as corresponding address

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

Corresponding Author:

Montgomery Grimes

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Xiaokang, Yang

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Piosczyk, Bernhard

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Callis, R.W.

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

J. Hosea (1)

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Jürgen Alex

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

- B - Plasma Heating and Current Drive.


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

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

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

Corresponding Author:

Saigusa Mikio

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

- B - Plasma Heating and Current Drive.

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