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Sample records for tokamak tore supra

  1. Edge plasma diagnostics on Tore Supra tokamak

    International Nuclear Information System (INIS)

    Fujita, Junji

    1991-01-01

    From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)

  2. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

    Aymar, R.; Bareyt, B.; Bon Mardion, G.

    1980-10-01

    This document describes the basic design for the main components of the Tokamak system of Tora Supra. As such, it focuses on the engineering problems, and refers to last year report on Tora Supra (EUR-CEA-1021) for objectives and experimental programme of the apparatus on one hand, and for qualifying tests of the main technical solutions on the other hand. Superconducting toroidal field coil system, vacuum vessels and radiation shields, poloidal field system and cryogenic system are described

  3. Scrape-off layer flows in the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Gunn, J.P.; Loarer, T.; Saint-Laurent, F.; Bucalossi, J.; Devynck, P.; Hertout, P.; Moreau, P.; Nanobashvili, I.; Rimini, F.; Duran, I.; Fuchs, V.; Panek, R.; Stockel, J.; Adamek, J.; Dejarnac, R.; Hron, M.; Sarkissian, A.

    2005-01-01

    Near-sonic parallel flows are systematically observed in the scrape-off layer (SOL) of the limiter tokamak Tore Supra, as in many X-point divertor tokamaks. The poloidal variation of the Mach number of the parallel flow has been measured by moving the contact point of a small circular plasma onto limiters at different poloidal angles. The resulting variations of flow are consistent with the existence of a poloidally nonuniform core-to-SOL out-flux concentrated near the outboard midplane. Strong variations of the SOL width up to a factor of 10 suggest that this localized out-flux is due to enhanced radial transport. The plasma that gets ejected into the SOL can expand radially to the wall if magnetic field lines have long connection lengths and pass unobstructed across the outboard midplane. (authors)

  4. A step towards controlled fusion reactors: Tore Supra tokamak with superconducting magnets

    International Nuclear Information System (INIS)

    Turck, B.

    1988-01-01

    Tore Supra technology has to solve all the problems related to the development and the installaion of superconducting coils and associated cryogenic devices. Tore Supra will allow to get a significative experience to prepare next machines. Specifications and needs of tokamaks concerning the superconducting coils of future machines are recalled [fr

  5. Pneumatic injector of deuterium macroparticles for TORE-SUPRA tokamak

    International Nuclear Information System (INIS)

    Vinyar, I.V.; Umov, A.P.; Lukin, A.Ya.; Skoblikov, S.V.; Reznichenko, P.V.; Krasil'nikov, I.A.

    2006-01-01

    The pneumatic injector for periodic injection of fuel-solid-deuterium pellets into the plasma of the TORE-SUPRA tokamak in a steady-state mode is described. The deuterium pellet injection with an unlimited duration is ensured by a screw extruder in which gaseous deuterium is frozen and squeezed outwards in the form of a rod with a rectangular cross section. A cutter installed on the injector's barrel cuts a cylinder with a diameter of 2 mm and a length of 1.0-3.5 mm out from this rod. The movement of the cutter is controlled by a pulsed electromagnetic drive at a pulse repetition rate of 10 Hz. In the injector's barrel, a compressed gas accelerates a deuterium pellet to a velocity of 100-650 m/s [ru

  6. Study of heat flux deposition in the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Carpentier, S.

    2009-02-01

    Accurate measurements of heat loads on internal tokamak components is essential for protection of the device during steady state operation. The optimisation of experimental scenarios also requires an in depth understanding of the physical mechanisms governing the heat flux deposition on the walls. The objective of this study is a detailed characterisation of the heat flux to plasma facing components (PFC) of the Tore Supra tokamak. The power deposited onto Tore Supra PFCs is calculated using an inverse method, which is applied to both the temperature maps measured by infrared thermography and to the enthalpy signals from calorimetry. The derived experimental heat flux maps calculated on the toroidal pumped limiter (TPL) are then compared with theoretical heat flux density distributions from a standard SOL-model. They are two experimental observations that are not consistent with the model: significant heat flux outside the theoretical wetted area, and heat load peaking close to the tangency point between the TPL and the last closed field surface (LCFS). An experimental analysis for several discharges with variable security factors q is made. In the area consistent with the theoretical predictions, this parametric study shows a clear dependence between the heat flux length λ q (estimated in the SOL (scrape-off layer) from the IR measurements) and the magnetic configuration. We observe that the spreading of heat fluxes on the component is compensated by a reduction of the power decay length λ q in the SOL when q decreases. On the other hand, in the area where the derived experimental heat loads are not consistent with the theoretical predictions, we observe that the spreading of heat fluxes outside the theoretical boundary increases when q decreases, and is thus not counterbalanced. (author)

  7. Study of heat flux deposition in the Tore Supra Tokamak; Etude des depots de chaleur dans le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Carpentier, S.

    2009-02-15

    Accurate measurements of heat loads on internal tokamak components is essential for protection of the device during steady state operation. The optimisation of experimental scenarios also requires an in depth understanding of the physical mechanisms governing the heat flux deposition on the walls. The objective of this study is a detailed characterisation of the heat flux to plasma facing components (PFC) of the Tore Supra tokamak. The power deposited onto Tore Supra PFCs is calculated using an inverse method, which is applied to both the temperature maps measured by infrared thermography and to the enthalpy signals from calorimetry. The derived experimental heat flux maps calculated on the toroidal pumped limiter (TPL) are then compared with theoretical heat flux density distributions from a standard SOL-model. They are two experimental observations that are not consistent with the model: significant heat flux outside the theoretical wetted area, and heat load peaking close to the tangency point between the TPL and the last closed field surface (LCFS). An experimental analysis for several discharges with variable security factors q is made. In the area consistent with the theoretical predictions, this parametric study shows a clear dependence between the heat flux length lambda{sub q} (estimated in the SOL (scrape-off layer) from the IR measurements) and the magnetic configuration. We observe that the spreading of heat fluxes on the component is compensated by a reduction of the power decay length lambda{sub q} in the SOL when q decreases. On the other hand, in the area where the derived experimental heat loads are not consistent with the theoretical predictions, we observe that the spreading of heat fluxes outside the theoretical boundary increases when q decreases, and is thus not counterbalanced. (author)

  8. Parametric dependences of impurity transport in the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Parisot, Th.

    2007-09-01

    During this Ph.D. work, a full setup of tools for an experimental investigation of impurity transport has been developed on the Tore Supra tokamak. It includes a laser blow-off system for metallic impurity injections and developments for ITC (Impurity Transport Code), a transport code which allows the extraction of the experimental impurity transport coefficients (diffusion and convection velocity). This tool has been used to perform and analyse several experiments, to evidence parametric dependences of impurity transport. In a first experiment, a confinement time law for nickel in Tore Supra has been obtained as a function of collisionality ν * and normalized Larmor radius ρ * . Then the impurity charge Z role has been investigated in various conditions: ohmic regime with or without sawteeth, and sawtooth less L-mode with LH power. No Z effect is observed, consistently with theoretical predictions, whether neoclassical (NCLASS) or for turbulent transport with both non linear gyro-fluid (TRB) and quasilinear gyrokinetic (QuaLiKiz) simulations. An exception is found for LH heated plasmas where the confinement time seems to decrease for the heaviest impurities. This is not explained by any model available. The observed transport is close to neoclassical between sawtooth relaxations, in the centre (r q-1 ) of ohmic plasmas, turbulent outside. Without sawteeth, it is turbulent in the whole plasma, for ohmic or L mode discharges. The profile shape of the diffusion coefficient is here qualitatively different, with a stronger and deeper transition between the low diffusion central region and a more turbulent peripheral region for LH heated plasmas. (author)

  9. Upgraded ECE radiometer on the Tore Supra Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Segui, J.L.; Molina, D.; Goniche, M.; Maget, P.; Udintsev, V.S. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Antar, G.Y. [Center for Energy Research, UCSD, La Jolla CA (United States); Kraemer-Flecken, A. [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Plasmaphysik

    2004-07-01

    An upgraded 32-channel heterodyne radiometer, 1 GHz spaced, is used on the Tore-Supra tokamak to measure the electron cyclotron emission (ECE) in the frequency range 78-110 GHz for the ordinary mode (O1) and 94-126.5 GHz for the extraordinary mode (X2). From now radial resolution is essentially limited by ECE relativistic effects related to electron temperature and density, not by the channels frequency spacing. For example, this leads to precise electron temperature mapping during magneto hydrodynamic activities (MHD). In the equatorial plane, we use a dual polarisation Gaussian optics lens antenna. It has low spreading and a perpendicular line-of-sight that gives ECE measurements very low refraction and Doppler effects. Assuming that the plasma is a black body and there is no overlap between ECE harmonics, one can deduce the electron temperature profile by using the first harmonic ordinary mode (O1) or the second harmonic extraordinary mode (X2). The principle radio frequency emitter (RF) has its frequencies down shifted into intermediary frequencies (IF) that span from 2 to 18 GHz in the single side band mode (SSB). It is amplified by low noise IF amplifiers before forming channels. A separate O/X mode RF front-end allows the use of an IF electronic mode selector. This gives the potentiality of simultaneous O/X mode measurements in the 94-110 GHz. RF and IF filters reject the gyrotron frequency (118 GHz) in order to perform electron temperature measurements during electron cyclotron resonance heated plasmas. A precise absolute spectral calibration is performed outside the tokamak vacuum vessel by using a 600 deg C black body hot source, a double coherent digital signal averaging (trigger, turn and clock) on the waveform generated by a mechanical chopper, and a simulated tokamak window. The use of differential electronics and strong electromagnetic shielding improves also the calibration precision. The fast and slow data acquisition systems are free of aliasing

  10. Upgraded ECE radiometer on the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Segui, J.L.; Molina, D.; Goniche, M.; Maget, P.; Udintsev, V.S.; Kraemer-Flecken, A.

    2004-01-01

    An upgraded 32-channel heterodyne radiometer, 1 GHz spaced, is used on the Tore-Supra tokamak to measure the electron cyclotron emission (ECE) in the frequency range 78-110 GHz for the ordinary mode (O1) and 94-126.5 GHz for the extraordinary mode (X2). From now radial resolution is essentially limited by ECE relativistic effects related to electron temperature and density, not by the channels frequency spacing. For example, this leads to precise electron temperature mapping during magneto hydrodynamic activities (MHD). In the equatorial plane, we use a dual polarisation Gaussian optics lens antenna. It has low spreading and a perpendicular line-of-sight that gives ECE measurements very low refraction and Doppler effects. Assuming that the plasma is a black body and there is no overlap between ECE harmonics, one can deduce the electron temperature profile by using the first harmonic ordinary mode (O1) or the second harmonic extraordinary mode (X2). The principle radio frequency emitter (RF) has its frequencies down shifted into intermediary frequencies (IF) that span from 2 to 18 GHz in the single side band mode (SSB). It is amplified by low noise IF amplifiers before forming channels. A separate O/X mode RF front-end allows the use of an IF electronic mode selector. This gives the potentiality of simultaneous O/X mode measurements in the 94-110 GHz. RF and IF filters reject the gyrotron frequency (118 GHz) in order to perform electron temperature measurements during electron cyclotron resonance heated plasmas. A precise absolute spectral calibration is performed outside the tokamak vacuum vessel by using a 600 deg C black body hot source, a double coherent digital signal averaging (trigger, turn and clock) on the waveform generated by a mechanical chopper, and a simulated tokamak window. The use of differential electronics and strong electromagnetic shielding improves also the calibration precision. The fast and slow data acquisition systems are free of aliasing

  11. Poloidal asymmetries of flows in the Tore Supra tokamak

    Science.gov (United States)

    Vermare, L.; Hennequin, P.; Gürcan, Ö. D.; Garbet, X.; Honoré, C.; Clairet, F.; Giacalone, J. C.; Morel, P.; Storelli, A.; Tore Supra Team

    2018-02-01

    Simultaneous measurements of binormal velocity of density fluctuations using two separate Doppler backscattering systems at the low field side and at the top of the plasma show significant poloidal asymmetry. The measurements are performed in the core region between the radii 0.7 Supra tokamak. A possible generation mechanism by the ballooned structure of the underlying turbulence, in the form of convective cells, is proposed for explaining the observation of these poloidally asymmetric mean flows.

  12. Study of the electron heat transport in Tore-Supra tokamak; Etude du transport de la chaleur electronique dans le Tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Harauchamps, E

    2004-07-01

    This work presents analytical solutions to the electron heat transport equation involving a damping term and a convection term in a cylindrical geometry. These solutions, processed by Matlab, allow the determination of the evolution of the radial profile of electron temperature in tokamaks during heating. The modulated injection of waves around the electron cyclotron frequency is an efficient tool to study heat transport experimentally in tokamaks. The comparison of these analytical solutions with experimental results from Tore-Supra during 2 discharges (30550 and 31165) shows the presence of a sudden change for the diffusion and damping coefficients. The hypothesis of the presence of a pinch spread all along the plasma might explain the shape of the experimental temperature profiles. These analytical solutions could be used to determine the time evolution of plasma density as well or of any parameter whose evolution is governed by a diffusion-convection equation. (A.C.)

  13. Development and calibration of the fast neutral particle analyzer of the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Siri, B.

    1989-10-01

    The design and construction of an analyzer for the Tore Supra tokamak fast neutral particles are presented. The energy analysis of the hydrogen and deuterium fast neutrals from the plasma allows the obtention of the plasma ionic temperature. The principle of the analysis is described. The analysis maximal energy is 300 keV for the protons and 150 keV for the deuterons. The measurement of the flow of neutrals in a given energy gap requires the knowledge of the energy of analysis, energy resolution and efficiency of the analyzer. The determination of these parameters needed the utilization of a neutral particle beam of 0 to 50 KeV energy. The energy spectra of the neutrals and the plasma ionic temperature at Tore Supra were obtained [fr

  14. Opportunist combination of electronic technologies for real time calculations in the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Barbuti, A.; Gil, C.; Pastor, P.; Spuig, P.; Vincent, B.; Volpe, D.

    2013-06-01

    The Tore Supra tokamak real-time plasma control is based on measurements coming from various diagnostics. The complexity of all the events that occur during plasma is at the origin of measurements disturbances which have to be corrected in real time in order to ensure an optimal control. The signal correction does not just mean processing but requires complex algorithms. Electronics does not only need to process and adapt electrical signals, but it has to include corrections by mathematical calculation. The FPGA (field-programmable gate array) technology, with the help of basic adapted electronics, allows integrating the entire real time calculation and digital data transmission on the network. FMC (FPGA Mezzanine Card) coupled with in-house motherboard, which is used both as the interface with Tore Supra specific systems and as the support for other signals processing options, is the perfect answer to this request. The FMC includes a FPGA, memory, Ethernet port and multiple I/O for interfacing with the motherboard and Tore Supra signals. The algorithms are developed in VHDL (Very high speed integrated circuit Hardware Description Language), parallel process management that promotes faster calculation than a common μC (Micro-controller) in one clock pulse. The flexibility, the low cost and the implementation speed allow fitting a large number of various applications in fields where no 'off-theshelf' component can be found. And more specifically, in research and experimentation, algorithms can be continuously improved or modified for new requirements. (authors)

  15. Articulated inspection arm for ITER, a demonstration in the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Cordier, J.J.; Gargiulo, L.; Grisolia, C.; Samaille, F.; Palmer, J.D.

    2003-01-01

    The aim of this program is to demonstrate for ITER the feasibility of an in-vessel remote handling inspection using a long reach, limited payload carrier (1 to 10 kg) for penetration of the ITER chamber through the openings. This device is dedicated to close inspection of the Plasma Facing Components (PFC). An articulated demonstrator called articulated inspection arm (AIA) has been manufactured. A feasibility study of a full AIA operation in Tore Supra was performed, taking into account ITER reference requirements. A scale one demonstration of the AIA under ITER relevant condition is feasible on Tore Supra and would give significant improvement in research results for ITER remote Handling equipment. The test of the AIA demonstrator behaviour is foreseen in 2005 in real Tokamak conditions. The paper presents the full robot concept, the results of the first test campaign, the AIA new design and its integration on Tore Supra. Several potential uses of the AIA for the in vessel components inspection are being studied such as PFC visual inspection, water loop leak testing, laser ablation for wall detritiation and carbon dust and flakes removal are foreseen as utilities to be placed at the AIA head. These various systems are described in the paper

  16. Articulated inspection arm for ITER, a demonstration in the Tore Supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, J.J.; Gargiulo, L.; Grisolia, C.; Samaille, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Friconneau, J.P.; Perrot, Y. [CEA Fontenay-aux-Roses, LIST Robotics and Interactive Systems Unit, 92 (France); Palmer, J.D. [Max-Planck-Institut fuer Plasmaphysik Boltzmannstr.2, Garching (Germany)

    2003-07-01

    The aim of this program is to demonstrate for ITER the feasibility of an in-vessel remote handling inspection using a long reach, limited payload carrier (1 to 10 kg) for penetration of the ITER chamber through the openings. This device is dedicated to close inspection of the Plasma Facing Components (PFC). An articulated demonstrator called articulated inspection arm (AIA) has been manufactured. A feasibility study of a full AIA operation in Tore Supra was performed, taking into account ITER reference requirements. A scale one demonstration of the AIA under ITER relevant condition is feasible on Tore Supra and would give significant improvement in research results for ITER remote Handling equipment. The test of the AIA demonstrator behaviour is foreseen in 2005 in real Tokamak conditions. The paper presents the full robot concept, the results of the first test campaign, the AIA new design and its integration on Tore Supra. Several potential uses of the AIA for the in vessel components inspection are being studied such as PFC visual inspection, water loop leak testing, laser ablation for wall detritiation and carbon dust and flakes removal are foreseen as utilities to be placed at the AIA head. These various systems are described in the paper.

  17. MHD limits in non-inductive tokamak plasmas: simulations and comparison to experiments on Tore Supra

    International Nuclear Information System (INIS)

    Maget, P.; Huysmans, G.; Ottaviani, M.; Garbet, X.; Moreau, Ph.; Segui, J.-L.; Luetjens, H.

    2008-01-01

    Non-inductive tokamak discharges with a flat or hollow current profile are prone to the triggering of large tearing modes when the minimum of the safety factor is just below a low order rational. This issue is of particular importance for discussing the optimal safety factor for MHD modes avoidance in Steady-State reactor plasmas. Different non-linear regimes of such magnetic configurations in Tore Supra are studied using the full MHD code XTOR. Numerical simulations show that the non-linear stage of the Double-Tearing Mode (DTM) is governed by the full reconnection model, but a single tearing mode in a low magnetic shear configuration can have a similar impact on the confinement. The different regimes observed experimentally are recovered in the simulations: a small amplitude (2,1) DTM for close resonant surfaces as seen in Tore Supra, a sawtooth-like behaviour of the (2,1) Double-Tearing Mode as first seen in TFTR, or a large amplitude (2,1) tearing mode that severely degrades the energy confinement, as reported in Tore Supra, JET or DIII-D. Situations where q min ≅1.5 with a stable n = 1 mode, as seen in Tore Supra longest discharges, seem to put specific constraints on the MHD model that is used. Indeed, curvature stabilisation without transport terms as could explain linear stability, but such effect vanishes in presence of heat transport. Electron diamagnetic rotation effect is investigated as a possible mechanism for n = 1 mode stabilization.

  18. Pioneering superconducting magnets in large tokamaks: evaluation after 16 years of operating experience in tore supra

    International Nuclear Information System (INIS)

    Duchateau, J.L.; Gravil, B.; Tena, E.; Henry, D.; Journeaux, J.Y.; Libeyre, P.

    2004-01-01

    The toroidal field (TF) system of Tore Supra (TS) is superconducting. After 16 years of operation it is possible to give an overview of the experience gained on a large superconducting system integrated in a large Tokamak. Quantitative data will be given, about the TF system for the cryogenic system and for the magnet system as well, concerning the number of plasmas shots and the availability of the machine. The origin and the number of breakdowns or incidents will be described, with emphasis on cryogenics, to document repairs and changes on the system components. Concerning the behaviour during operation, the Fast Safety Discharges (FSD) in operation are of particular interest for the Tokamak operation, as they interrupt it on a significant time of the order of one hour. This aspect is particularly documented. The approach followed to decrease the number of these FSD will be reported and explained. The Tore Supra Tokamak was the first important meeting between Superconductivity and Plasma Physics on a large scale. Overall, despite the differences in design and size, the accumulated experience over 16 years of operation is a useful tool to prepare the manufacturing and the operation of the ITER magnets. (authors)

  19. Collection and Characterization of Particulate from the Tore Supra Tokamak (Dec. 1999 Vent)

    International Nuclear Information System (INIS)

    Sharpe, John Phillip

    2002-01-01

    Particulate generated during the operation of a fusion device contributes to the radiological source term associated with accident scenarios in the device. Understanding the mechanisms generating the particulate and correctly describing its physical and chemical behavior is essential for safety analyses of future fusion devices. Knowledge of these mechanisms is gained by collecting and characterizing particulate matter from operating fusion facilities. Tokamak dust, the particulate matter generated during the operation of a tokamak fusion device, was collected from Tore Supra in December 1999, during the initial phase of the scheduled shutdown for installation of advanced plasma facing components. Tore Supra, located at CEA Cadarache, France, is presently the third largest operating tokamak with the capability of long-pulse operation. Eighteen super-conducting coils produce the 4.5 T magnetic field inside a vessel 2.4 m in major radius and 1.2 m in minor radius. Limiter and divertor regimes of operation are possible. In the divertor regime, the circular magnetic configuration is ergodized by six outboard resonant divertor modules that are covered with 12 m2 of carbon fiber composite (CFC) tiles. In addition, some field lines are diverted to actively cooled neutralizing plates made of CuCrZr bars covered with B4C. In the limiter regime, the plasma leans on the actively cooled inboard first wall or on a set of inertially cooled pumped limiters. The first wall area of 12 m2 is covered with both polycrystalline graphite tiles (83%) and CFC tiles (17%). The single outboard limiter is constructed of pyrolitic graphite, and the four toroidally symmetric bottom limiters are constructed of CFC. Figure 1.1 displays the relative location of plasma facing components within the plasma chamber of Tore Supra. In this report, we present in Section 2 the methods used to collect and analyze this dust and describe the selection of sampling locations. Section 3 includes a discussion

  20. Ion temperature measurements in the scrape-off layer of the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Kocan, M.

    2009-10-01

    The thesis describes measurements of the scrape-off layer (SOL) ion temperature T i with a retarding field analyzer (RFA) in the limiter tokamak Tore Supra. Considerable emphasis is placed on study of the instrumental effects of RFAs and their influence on T i measurements. In general, the influence of instrumental effects on T i measurements is found to be relatively small. The instrumental study is followed by systematic measurements of T i (as well as other parameters) in the Tore Supra SOL. This includes the scaling of SOL temperatures and electron density with the main plasma parameters (such as the plasma density, toroidal magnetic field, working gas, and the radiated power fraction). Except at very high densities or in detached plasmas, SOL T i is found to be higher than T e by up to a factor of 7. While SOL T i is found to vary by almost two orders of magnitude, following the variation of the core temperatures, SOL T e changes only little and seems to be decoupled from the core plasma. The first continuous T i /T e profile from the edge of the confined plasma into the SOL is constructed using data from different tokamaks. It is shown that T i /T e > 1 in the SOL but also in the confined plasma, and increases with radius. The first evidence of poloidal asymmetry of the radial ion and electron energy transport in the SOL is reported. Implications for ITER start-up phase are discussed. Correlation of the asymmetries of SOL T i and T e measured from both directions along the magnetic field lines with changes of the parallel Mach number is studied. SOL T i was measured for the first time in Tore Supra by charge exchange recombination spectroscopy (CXRS) and compared to RFA data. A factor of 4 higher T i measured by CXRS is a subject of further analysis. (A.C.)

  1. Tore Supra: technical description

    International Nuclear Information System (INIS)

    1985-08-01

    This report is aimed, after a brief recall of physics and technologic perspectives of Tore Supra, at giving a detailed description of the basic machine; details of each component are defined. Volume 1 is specifically concerned with the general aspects of Tore Supra and the toroidal field system [fr

  2. Near infrared thermography by CCD cameras and application to first wall components of Tore Supra tokamak

    International Nuclear Information System (INIS)

    Moreau, F.

    1996-01-01

    In the Tokamak TORE-SUPRA, the plasma facing components absorbs and evacuate (active cooling) high power fluxes (up to 10 MW/m 2 ). Their thermal behavior study is essential for the success of controlled thermonuclear fusion line. The first part is devoted to the study of power deposition on the TORE-SUPRA actively cooled limiters. A model of power deposition on one of the limiters is developed. It Takes into account the magnetic topology and a description of the plasma edge. The model is validated with experimental calorimetric data obtained during a series of shots. This will allow to compare the surface temperature measurements with the predicted ones. The main purpose of this thesis was to evaluate and develop a new surface temperature measurement system. It works in the near infrared range (890 nm) and is designed to complete the existing thermographic diagnostic of TORE-SUPRA. By using the radiation laws (for a blackbody and the plasma) ant the laboratory calibration one can estimate the surface temperature of the observed object. We evaluate the performances and limits of such a device in the harsh conditions encountered in a Tokamak environment. On the one hand, in a quasi ideal situation, this analysis shows that the range of measurement is 600 deg. C to 2500 deg. C. On the other hand, when one takes into account of the plasma radiation (with an averaged central plasma density of 6.10 19 m -3 ), we find that the minimum surface temperature rise to 900 deg. C. In the near future, according to the development of IR-CCD cameras working in the near infrared range up to 2 micrometers, we will be able to keep the good spatial resolution with an improved lower limit for the temperature down to 150 deg. C. The last section deals with a number of computer tools to process the images obtained from experiments on TORE-SUPRA. A pattern recognition application was especially developed to detect a complex plasma iso-intensity structure. (author)

  3. Scrape-off layer flows in the Tore Supra tokamak

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Boucher, C.; Dionne, M.; Ďuran, Ivan; Fuchs, Vladimír; Loarer, T.; Pánek, Radomír; Saint Laurent, F.; Stöckel, Jan; Adámek, Jiří; Bucalossi, J.; Dejarnac, Renaud; Devynck, P.; Hertout, P.; Hron, Martin; Nanobashvili, I.; Rimini, F.G.; Sarkissian, A.

    2006-01-01

    Roč. 812, - (2006), s. 27-34 ISSN 0094-243X. [AIP Conference Proceedings. Opole-Turawa, 06.09.2006-09.09.2006] R&D Projects: GA ČR GP202/03/P062 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * scrape-off layer * plasma flow * radial transport * Mach probe Subject RIV: BL - Plasma and Gas Discharge Physics http://proceedings.aip.org/dbt/dbt.jsp?KEY=APCPCS&Volume=812&Issue=1

  4. Study of plasma turbulence by ultrafast sweeping reflectometry on the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Hornung, Gregoire

    2013-01-01

    The performance of a fusion reactor is closely related to the turbulence present in the plasma. The latter is responsible for anomalous transport of heat and particles that degrades the confinement. The measure and characterization of turbulence in tokamak plasma is therefore essential to the understanding and control of this phenomenon. Among the available diagnostics, the sweeping reflectometer installed on Tore Supra allows to access the plasma density fluctuations from the edge to the centre of the plasma discharge with a fine spatial (mm) and temporal resolution (μs), that is of the order of the characteristic turbulence scales.This thesis consisted in the characterization of plasma turbulence in Tore Supra by ultrafast sweeping reflectometry measurements. Correlation analyses are used to quantify the spatial and temporal scales of turbulence as well as their radial velocity. In the first part, the characterization of turbulence properties from the reconstructed plasma density profiles is discussed, in particular through a comparative study with Langmuir probe data. Then, a parametric study is presented, highlighting the effect of collisionality on turbulence, an interpretation of which is proposed in terms of the stabilization of trapped electron turbulence in the confined plasma. Finally, it is shown how additional heating at ion cyclotron frequency produces a significant though local modification of the turbulence in the plasma near the walls, resulting in a strong increase of the structure velocity and a decrease of the correlation time. The supposed effect of rectified potentials generated by the antenna is investigated via numerical simulations. (author) [fr

  5. Multi scale study of carbon deposits collected in Tore-Supra and TEXTOR tokamaks; Etude multi echelle des depots carbones collectes dans les tokamaks Tore Supra et TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Richou, M

    2007-06-15

    Tokamaks are devices aimed at studying magnetic fusion. They operate with high temperature plasmas containing hydrogen, deuterium or tritium. One of the major issue is to control the plasma-wall interaction. The plasma facing components are most often in carbon. The major drawback of carbon is the existence of carbon deposits and dust, due to erosion. Dust is potentially reactive in case of an accidental opening of the device. These deposits also contain H, D or T and induce major safety problems when tritium is used, which will be the case in ITER. Therefore, the understanding of the deposit formation and structure has become a main issue for fusion researches. To clarify the role of the deposits in the retention phenomenon, we have done different complementary characterizations for deposits collected on similar places (neutralizers) in tokamaks Tore Supra (France) and TEXTOR (Germany). Accessible microporous volume and pore size distribution of deposits has been determined with the analysis of nitrogen and methane adsorption isotherms using the BET, Dubinin-Radushkevich and {alpha}{sub s} methods and the Density Functional Theory (DFT). To understand growth mechanisms, we have studied the deposit structure and morphology. We have shown using Transmission Electron Microscopy (TEM) and Raman micro-spectrometry that these deposits are non amorphous and disordered. We have also shown the presence of nano-particles (diameter between 4 and 70 nm) which are similar to carbon blacks: nano-particle growth occurs in homogeneous phase in the edge plasma. We have emphasised a dual growth process: a homogenous and a heterogeneous one. (author)

  6. Measurements of RF-induced sol modifications in Tore Supra tokamak

    International Nuclear Information System (INIS)

    Kubic, Martin; Gunn, James P.; Colas, Laurent; Heuraux, Stephane; Faudot, Eric

    2012-01-01

    Since spring 2011, one of the three ion cyclotron resonance heating (ICRH) antennas in the Tore Supra (TS) tokamak is equipped with a new type of Faraday screen (FS). Results from Radio Frequency (RF) simulations of the new Faraday screen suggest the innovative structure with cantilevered bars and 'shark tooth' openings significantly changes the current flow pattern on the front of the antenna which in turn reduces the RF potential and RF electrical field in particular parallel to the magnetic field lines which contributes to generating RF sheaths. Effects of the new FS operation on RF-induced scrape-off layer (SOL) modifications are studied for different plasma and antenna configurations - scans of strap power ratio imbalance, phasing, injected power and SOL density. (authors)

  7. Mechanical design and manufacture of magnetic ergodic divertor for the TORE SUPRA tokamak

    International Nuclear Information System (INIS)

    Lipa, M.; Aymar, R.; Deschamps, P.; Hertout, P.; Portafaix, C.; Samain, A.

    1989-01-01

    A configuration of six equally spaced ergodic divertors has been chosen to control the plasma impurities in the TORE SUPRA tokamak since the control of these impurities is essential to the long pulse duration envisioned for the machine. Each of the six indentical modules is composed of (8) conductor bars arranged in a poloidal direction forming a resonant helical winding. The proximity of the conductors to the plasma requires that each copper assembly be water cooled, enclosed in a stainless steel casing and protected by pure graphite tiles attaches to the inner surface of the casing. Particles which drift between the coil bars are neutralized on actively water cooled neutralizer plates and then pumped out by titanium getter pumps which are located on each toroidal end of a divertor modul. (author). 5 refs.; 7 figs.; 1 tab

  8. Surface modification and hydrogen isotope retention in CFC during plasma irradiation in the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Begrambekov, L.; Brosset, C.; Bucalossi, J.; Delchambre, E.; Gunn, J.P.; Grisolia, C.; Lipa, M.; Loarer, T.; Mitteau, R.; Moner-Garbet, P.; Pascal, J.-Y.; Shigin, P.; Titov, N.; Tsitrone, E.; Vergazov, S.; Zakharov, A.

    2007-01-01

    The uniform layer with thickness at least 50-100 μm was found on the CFC tiles from the inboard midplane after more than four years of tokamak operation. The upper part of the uniform layer was amorphous, but at the depth of ∼5 μm a structure consisting of micro-size regions with aromatic chains located parallel to the surface was found. Gradual transition from uniform layer to underlying CFC structure was observed. The reciprocating material probe was used for installation of CFC samples in the Tore Supra deuterium plasma. The thermal desorptional spectra of these samples are compared with the spectra of the samples irradiated in the laboratory stand and with the spectra of hydrogenated carbon film. The peculiarities of hydrogen isotope trapping under plasma irradiation and at the atmosphere are presented and discussed

  9. Particle control studies on Tore Supra

    International Nuclear Information System (INIS)

    Mioduszewski, P.

    1987-01-01

    The report consists of viewgraphs. The goal of the particle control program at Tore Supra is to study plasma performance with strong pellet fueling and corresponding particle exhaust in a limiter tokamak

  10. Study of heat transfer and particle transport in Tore Supra and HL-2A tokamaks

    International Nuclear Information System (INIS)

    Song, S.

    2011-12-01

    This thesis reports on experimental studies of heat and particles transport performed on 2 large tokamaks: Tore Supra (based at CEA/Cadarache, France) and HL-2A (based at the Southwestern Institute of Physics, Chengdu, China). The modulated source is the Electron Cyclotron Resonance Heating (ECRH) for the heat pinch and density pump-out studies, while the non-local transport experiments use the Supersonic Molecular Beam Injection (SMBI) as source of modulation. The emphasis is put on the inward heat pinch. In the off-axis ECRH modulation experiments on Tore Supra with low frequency (1 Hz), strong heat inward transport has been observed, in particular for low density. Two transport models have been applied in order to analyze the experimental behavior. The first one is the linear pinch model (LPM) and the second one is an empirical model based on micro-instabilities theory, named Critical Gradient Model (CGM). Good agreement has been found for all harmonics between the experimental data and the simulation using LPM. On the other hand, good agreement has not been achieved using CGM. The density pump-out with large particles and energy losses during ECRH is commonly observed in tokamaks. A new dynamic approach using the modulation technique has been used in HL-2A for analyzing the transient phase of the density pump-out. A correlation between the turbulence increase and the density pump-out has been found. The non-local transport phenomenon, characterized by a fast transient process compared to the normal diffusive response to the perturbation is observed. Both phenomena, i.e., pump-out and non-locality, show as simultaneous variation of density and temperature. This can be an inspiration for the usage of a transport matrix which considers the density and temperature evolution together. Simulations with a simple transport matrix, with non-diagonal terms coupling temperature and density qualitatively reproduce the non-local and pump-out effects qualitatively

  11. Calibration of Fabry-Perot interferometers for electron cyclotron emission measurements on the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Javon, C.; Talvard, M.

    1990-01-01

    The electron temperature is routinely measured on TORE SUPRA using Fabry-Perot cavities. These have been calibrated using a technique involving coherent addition and Fourier analysis of a chopped black-body source. Comparison with conventional techniques is reported

  12. Implementation of FCI heating system to the control system of Tore-Supra; Integration du systeme de chauffage FCI au sein du reseau de controle commande du Tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Wisniewski, S

    2001-11-01

    This report presents the implementation of the ion cyclotron resonance heating system (FCI) to the instrumentation and control system of the Tore-Supra tokamak. The new plasma heating system involves 3 antennas delivering 12 MW that are required to maintain fusion reactions. This paper is divided into 8 chapters: 1) thermonuclear fusion and Tore-Supra tokamak; 2) hardware system around Tore-Supra, in this chapter the control system and the data acquisition and processing systems are presented; 3) functional analysis, this analysis defines the different needs concerning timing and pilot-controlling, a preliminary proposition of hardware equipment is made; 4) operating modes of FCI; 5) communication within the control system network; 6) communication with the supervisory system of the power stations; 7) management of data exchange with SMX generators; and 8) control of the rate of stationary waves during the injection of power into the plasma.

  13. Study of fusion plasma microturbulence by reflectometry in Tore Supra tokamak

    International Nuclear Information System (INIS)

    Gerbaud, Th.

    2008-11-01

    Fast-sweeping reflectometry in extraordinary mode allows direct measurement of radial wave-number local spectra S(δ n /n](k r ,r), and radial profiles of density fluctuations, on Tore Supra tokamak. Wavelet-based approach - a mathematical tool for position-frequency analysis - made possible to consider the strong radial variation of the measured turbulence. Special consideration was given to the validation of spectra and turbulent profiles measurements, by comparing with experimental measurements (reflectometry, probes) and numerical non-linear gyrokinetic simulations. This density fluctuations measurement method has been used to analyse the local transport, by performing a dimensionless scaling on collisionality, ν * . The scaling experiments allow direct comparisons of plasmas from different tokamaks. A clear decrease of the normalized confinement time of the plasma energy with the normalized collisionality was observed: Bτ E ∼ ν *-0.5±0.15 . These new measurements of density fluctuations profiles have shown an intense rise of the edge turbulence (r/a > 0.8) when increasing - also observed by Doppler reflectometry diagnostic - providing a physical explanation of the loss of confinement with the normalized collisionality. More central regions did not present apparent variations (δ n /n, χ(eff)). Core plasma simulations (linear stability code KineZero and non-linear gyrokinetic GYRO) were performed, in order to analyse the experimental behaviour of the plasma. (author)

  14. Cherenkov-type diamond detectors for measurements of fast electrons in the TORE-SUPRA tokamak

    International Nuclear Information System (INIS)

    Jakubowski, L.; Sadowski, M. J.; Zebrowski, J.; Rabinski, M.; Malinowski, K.; Mirowski, R.; Lotte, Ph.; Gunn, J.; Pascal, J-Y.; Colledani, G.; Basiuk, V.; Goniche, M.; Lipa, M.

    2010-01-01

    The paper presents a schematic design and tests of a system applicable for measurements of fast electron pulses emitted from high-temperature plasma generated inside magnetic confinement fusion machines, and particularly in the TORE-SUPRA facility. The diagnostic system based on the registration of the Cherenkov radiation induced by fast electrons within selected solid radiators is considered, and electron low-energy thresholds for different radiators are given. There are some estimates of high thermal loads, which might be deposited by intense electron beams upon parts of the diagnostic equipment within the TORE-SUPRA device. There are some proposed measures to overcome this difficulty by the selection of appropriate absorption filters and Cherenkov radiators, and particularly by the application of a fast-moving reciprocating probe. The paper describes the measuring system, its tests, as well as some results of the preliminary measurements of fast electrons within TORE-SUPRA facility.

  15. Plasma density control with ergodic divertor on Tore Supra; Controle de la densite du plasma en presence du divertor ergodique dans le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Meslin, B

    1998-04-30

    Plasma density control on the tokamak Tore Supra is important for the optimization of every experimental scenario dealing with the improvement of plasma performances. Specific conditions are required both in the plasma bulk and at the edge. Within the framework of the present study, a magnetic configuration is used in the e plasma edge of Tore Supra: the ergodic divertor configuration. A magnetic perturbation which is resonant with the permanent field destroys the plasma confinement locally, opening the field lines onto the material components. They aim of the study is the characterization of the edge density in every relevant scenario for Tore Supra. The first part of this work is dedicated to density and temperature measurements by a series of fixed Langmuir probes located at the very edge of the plasma. Thanks to them, density regimes have been put in evidence during experiments where the volume averaged density , an usual control parameter of the plasma, was varied. The analysis of heat and particle transport through the plasma edge region explains the mechanisms leading to those regimes. The essential factor in our analysis is the dependence of the electron conductivity and ionization depth on temperature. While heat conduction governs the heat transport, the edge density varies linearly according to . Below a critical temperature, reached when the ion flux amplification at constant power density is large enough, a parallel temperature gradient appears leading to a density gradient in the opposite direction in order to maintain the pressure constant along the field lines. A high recycling regime is obtained and the edge density varies like {sup 3}. The pressure conservation is no more satisfied during the detachment of the plasma, which is characterized by a high neutral density at low temperatures leading to a ion momentum loss by friction against the neutrals. The edge density drops in those conditions. These regimes are similar

  16. Major progress on tore supra toward steady state operation of tokamaks

    International Nuclear Information System (INIS)

    Saoutic, Y.

    2003-01-01

    During winter 2000-2001, a major upgrade of the internal components of Tore Supra has been completed that increased the heat extraction capability to 25 MW in steady state. Operating Tore Supra in this new configuration has produced a wealth of new results. The highlights of the 2002 long duration discharges campaign are: 4 minutes 25 seconds long discharges with an integrated energy of 0.75 GJ, which is three time higher than the old Tore Supra world record; recharge of the primary transformer by Lower Hybrid Current Drive (LHCD) for about 1 minute; 4 minutes long LHCD pulses; 1 minute long Ion Cyclotron Resonant Heating (ICRH) pulse (0.11 GJ of ICRH injected energy). Beyond the quantitative step, significant qualitative progress in the steady state nature of the discharge has been accomplished: contrary to the situation in the old Tore Supra configuration, the plasma density is perfectly controlled by active pumping over the overall shot duration. The duration of Tore Supra discharges is sufficient to allow the complete diffusion of the resistive current. Surprising new physics is revealed in such discharges when approaching zero loop voltage. Slow central electron temperature oscillations have been observed in a variety of situations. Such oscillations are not likely to be linked to any MHD instabilities and probably results from an interplay between current profile shape, LHCD power deposition and transport. Analysis of the temperature gradient in the core region shows a very interesting behaviour and the normalised temperature gradient length is compared to the critical thresholds. Finally, the performance of heating and current drive systems and the observations made of the interior of Tore Supra after the long duration discharges campaign are reported. (author)

  17. Steady-state operation of tokamaks: Key physics and technology developments on Tore Supra

    International Nuclear Information System (INIS)

    Jacquinot, J.

    2005-01-01

    Important technological and physics issues related to long pulse operation required for a reactor are now being addressed in Tore Supra. experimental results in conditions where all the plasma facing components are actively cooled during pulses exceeding six minutes. Important physics issues related to continuous operation are observed in non inductively driven plasmas. (author)

  18. Tore Supra: technical description

    International Nuclear Information System (INIS)

    1985-08-01

    Cryogenic system of Tore Supra is described with its principal functions and operation modes. Data control and acquisition with on line data processing is presented. Radiation dose and induced radioactivity evaluation is studied. Cooling system is detailed together with characteristics of facilities and circuits to cool. Then machine assembly and buildings are presented [fr

  19. Multi scale study of carbon deposits collected in Tore-Supra and TEXTOR tokamaks

    International Nuclear Information System (INIS)

    Richou, M.

    2007-06-01

    Tokamaks are devices aimed at studying magnetic fusion. They operate with high temperature plasmas containing hydrogen, deuterium or tritium. One of the major issue is to control the plasma-wall interaction. The plasma facing components are most often in carbon. The major drawback of carbon is the existence of carbon deposits and dust, due to erosion. Dust is potentially reactive in case of an accidental opening of the device. These deposits also contain H, D or T and induce major safety problems when tritium is used, which will be the case in ITER. Therefore, the understanding of the deposit formation and structure has become a main issue for fusion researches. To clarify the role of the deposits in the retention phenomenon, we have done different complementary characterizations for deposits collected on similar places (neutralizers) in tokamaks Tore Supra (France) and TEXTOR (Germany). Accessible microporous volume and pore size distribution of deposits has been determined with the analysis of nitrogen and methane adsorption isotherms using the BET, Dubinin-Radushkevich and α s methods and the Density Functional Theory (DFT). To understand growth mechanisms, we have studied the deposit structure and morphology. We have shown using Transmission Electron Microscopy (TEM) and Raman micro-spectrometry that these deposits are non amorphous and disordered. We have also shown the presence of nano-particles (diameter between 4 and 70 nm) which are similar to carbon blacks: nano-particle growth occurs in homogeneous phase in the edge plasma. We have emphasised a dual growth process: a homogenous and a heterogeneous one. (author)

  20. Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak

    International Nuclear Information System (INIS)

    Costanzo, L.

    2001-10-01

    The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor γ was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that γ=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a major advantage

  1. Experimental study of the MHD activity associated to the mode m=2, n=1 in the Tore Supra tokamak; Etude experimentale de l`activite MHD associee au mode m=2, n=1 dans le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Turlur, S.

    1996-09-20

    In tokamaks such as Tore Supra, the plasma confinement magnetic structure can be severely affected when Magnetohydrodynamic (M.H.D.) instabilities are destabilized. Experimentally, these instabilities are detected as magnetic fluctuations with captors located against the inner wall of the vacuum vessel. Fourier analysis provides amplitude, frequency and wave numbers of magnetic modes. In case of fast or transient phenomena, the analysis of magnetic fluctuations is completed using the singular value decomposition. In this dissertation, these analysis techniques are used to study two specific examples of M.H.D. activity related to the m = 2, n = 1 mode. On Tore Supra, the onset of this mode have strong consequences on the stability of partially or fully non inductive discharges. A regular and persistent sawtooth-like regime is observed on the electronic temperature leading to a significant degradation of the central confinement. Heat exhaust and particle balance are also essential parameters to achieve stationary discharges. On Tore Supra, these are studied with the ergodic divertor which produces stochastic magnetic field lines at the plasma edge. For optimal operating conditions of the ergodic divertor, the growth of the m = 2, N = 1 mode can lead to sudden destruction of magnetic equilibrium. For both cases, understanding and characterization of mechanisms leading to the observed m = 2, n = 1 M.H.D. activity are fundamental to obtain stationary discharges. (author). 115 refs.

  2. Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak; Etude experimentale des aspects topologiques du divertor ergodique de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Costanzo, L

    2001-10-01

    The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor {gamma} was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that {gamma}=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a

  3. High power lower hybrid current drive experiment in TORE SUPRA tokamak

    International Nuclear Information System (INIS)

    Peysson, Y.

    2001-01-01

    A review of the Lower Hybrid (LH) current drive experiments carried out on the TORE SUPRA tokamak is presented. This work highlights the issues for an effective application of the LH wave at high power in reactor relevant conditions. Very promising performances have been obtained with the new launcher that is designed to couple up to 4 MW during 1000 s at a power density of 25 MWm -2 . The heat load on the guard limiter of the antenna and the fast electron acceleration in the near electric field of the grill mouth remain at a low level, while the mean reflection coefficient never exceeds 10%. The powerful diagnosis capabilities of the hard x-ray (HXR) fast electron bremsstrahlung tomography has led to significant progresses in the understanding of the LH wave dynamics. The role of the fastest electrons driven by the LH wave is clearly identified. From HXR measurements, an increase of the LH current drive efficiency with the plasma current is predicted and confirmed by a direct determination at zero loop voltage. LH power absorption is observed to be off-axis in almost all plasma conditions, and its radial width clearly depends of antenna phasing conditions. A correlation between the HXR profiles and the onset of an improved core confinement is identified in fully non-inductive discharges. This regime ascribed to some vanishing of the magnetic shear is found to be transient and usually ends when the minimum of the safety factor becomes very close to 2, leading to a large MHD activity. Experimental observations and numerical simulations suggest that LH power is absorbed in a few number of passes. However, besides toroidal mode coupling, additional mechanisms may likely contribute to a spectral broadening to the LH wave. (author)

  4. Shielding study of a fusion machine. Elaboration of a global shielding calculation scheme for the Tokamak tore Supra

    International Nuclear Information System (INIS)

    Diop, C.M'B.

    1984-01-01

    This thesis presents a global shielding calculation scheme for neutron and gamma rays arising from the Tokamak TORE SUPRA fusion device, in which a deuterium plasma is used. To study the shield parameters we have elabored a important chaining of neutron and gamma transport codes, TRIPOLI, ANISN, MERCURE 4, allowing to evaluate the radial and skyshine components of the dose rate behind the concrete shield. The study of thermonuclear neutron activation is fundamental to define a tokamak exploitation strategy. For this, two formalisme have been developed. They are based on a modelization of the activation reaction rates according to TRIPOLI, ANISN, and MERCURE 4 codes capabilities. The first one calculates, in one dimensional geometry, the desactivation gamma dose rate inside the vacuum chamber. The second one is a tridimensional model which determines the spatial variation of the gamma dose rate in the machine room. The problem of the existence of runaway electrons and associated secondaries radiations, bremsstrahlung gamma rays particularly, is approched. The results which are presented have contributed to define the parameters of the concrete shield and a strategy for TORE SUPRA Tokamak exploitation [fr

  5. Tore supra first wall conditioning

    International Nuclear Information System (INIS)

    Gauthier, E.; Achard, M.H.; Grosman, A.; Monier, P.

    1989-01-01

    The procedures and the results obtained concerning impurity and isotopic control in Tore Supra tokamak are summarized. The conditioning of the vessel, mainly achieved by glow discharges, is described. The impurity control of the discharge was monitored with a VUV-X spectrometer. The in situ blasting degassing procedure applied is explained. In the sequence of the conditioning process, the hydrogen and the helium glow discharges and the carbonization method are discussed. The He glow discharges allowed to limit the H content of the He plasma shot below 20%

  6. Near infrared thermography by CCD cameras and application to first wall components of Tore Supra tokamak; Thermographie proche infrarouge par cameras CCD et application aux composants de premiere paroi du tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, F.

    1996-06-07

    In the Tokamak TORE-SUPRA, the plasma facing components absorbs and evacuate (active cooling) high power fluxes (up to 10 MW/m{sup 2}). Their thermal behavior study is essential for the success of controlled thermonuclear fusion line. The first part is devoted to the study of power deposition on the TORE-SUPRA actively cooled limiters. A model of power deposition on one of the limiters is developed. It takes into account the magnetic topology and a description of the plasma edge. The model is validated with experimental calorimetric data obtained during a series of shots. This will allow to compare the surface temperature measurements with the predicted ones. The main purpose of this thesis was to evaluate and develop a new temperature measurement system. It works in the near infrared range (890 nm) and is designed to complete the existing thermographic diagnostic of TORE-SUPRA. By using the radiation laws (for a blackbody and the plasma) and the laboratory calibration one can estimate the surface temperature of the observed object. We evaluate the performances and limits of such a device in the harsh conditions encountered in a Tokamak environment. On the one hand, in a quasi ideal situation, this analysis shows that the range of measurements is 600 deg. C to 2500 deg. C. On the other hand, when one takes into account of the plasma radiation (with an averaged central plasma density of 6.10{sup 19} m{sup -3}), we find that the minimum surface temperature rise to 900 deg. C instead of 700 deg. C. In the near future, according to the development of IR-CCD cameras working in the near infrared range up to 2 micrometers, we will be able to keep the good spatial resolution with an improved lower limit for the temperature down to 150 deg. C. The last section deals with a number of computer tools to process the images obtained from experiments on TORE-SUPRA. A pattern recognition application was developed to detect a complex plasma iso-intensity structure. 87 refs.

  7. Study of the chemical sputtering in Tore-Supra; Etude de l'erosion chimique dans le tokamak Tore-Supra

    Energy Technology Data Exchange (ETDEWEB)

    Cambe, A

    2002-06-28

    The work presented in this thesis focuses on the interactions between energetic particles coming from thermonuclear plasma and the inner components of a fusion machine. This interaction induces two major problems: erosion of the wall, and tritium retention. This report treats the erosion of carbon based materials. The first part is devoted to chemical sputtering, that appears to be the principal erosion mechanism, compared to physical sputtering and radiation enhanced sublimation that both can be limited. Chemical sputtering has been studied in situ in the tokamak Tore-Supra for ohmic and lower hybrid (LH) heated discharges, by means of mass spectrometry and optical spectroscopy. We have shown that it is necessary to take into account both methane and heavier hydrocarbons (C{sub 2}D{sub x} and C{sub 3}D{sub y}) in the determination of the chemical sputtering yield. It is found that for the ohmic discharges, the sputtering yield of CD{sub 4} (Y{sub CD4}) is highly flux ({phi}) dependent, showing a variation of the form: Y{sub CD4} {proportional_to} {phi}{sup -0.23}. The experimental study also reveals that an increase of the surface temperature induces an augmentation of Y{sub CD4}. The interpretation and the modelling of the experimental results have been performed with a Monte Carlo code (BBQ. In the second part of this work, we have developed and installed an infrared spectroscopy diagnostic in the 0.8-1.6, {mu}m wavelength range dedicated to the measurement of surface temperature, and the identification of atomic and molecular lines emitted during plasma/wall interactions. In the third part, we present the feasibility study of an in situ tungsten deposition process at low temperature(<80 deg C) in order to suppress the chemical sputtering. This study shows that, with this method call Plasma Assisted Chemical Vapor Deposition (PACVD), we are able to coat the whole inner vessel of a tokamak with 1 {mu}m of tungsten. (author)

  8. Ion temperature profiles along a hydrogen diagnostic beam in a TORE SUPRA tokamak plasma

    International Nuclear Information System (INIS)

    Romannikov, A.; Petrov, Yu.; Platts, P.; Khess, V.; Khutter, T.; Farzhon, Zh.; Moro, F.

    2002-01-01

    By means of corpuscular diagnostics one studies temperature of ions along a diagnostic hydrogen beam. Paper presents comparison of temperature of plasma (deuterium) basic ions measures by means of the active corpuscular diagnostics with temperature of C + carbon ions along a beam. One studies behavior peculiarities of T i ion temperature profiles for TORE-SUPRA different modes, such as: formation of plane and even hollow T i profiles for ohmic modes, variation of T i profiles under operation of an ergodic diverter, difference of temperature of basic ions measured by means of the active corpuscular diagnostics from C +5 temperature. Paper offers clear explanation of these peculiarities [ru

  9. Nuclear fusion TORE SUPRA, a new stage

    International Nuclear Information System (INIS)

    Gregoire, M.; Laurent, L.

    1995-01-01

    Since almost forty years, the scientists try and neutralize in a pacific aim thermonuclear fusion energy and therefore they use the magnetic confinement of hot plasmas.In France, since 1960 the achieved studies permitted in 1988 to bring into service the TORE SUPRA TOKAMAK, which used, for the first time a superconducting magnet to generate the confinement magnetic field. TORE SUPRA, which didn't still explore its maximal potentialities, will be one of the apparatuses which will be used as basis of the international project ITER development. 5 figs

  10. TORE SUPRA: programme of development, qualifying tests

    International Nuclear Information System (INIS)

    Aymar, R.; Bareyt, B.; Bon Mardion, G.

    1982-02-01

    TORE SUPRA is a Tokamak under construction in France. It is designed as a useful complement to JET. In this paper, we review the development work and test programme, and summarize its main objectives and conclusions. This concerns first the more conventional parts of the Tokamak (vacuum vessel technology and high power circuit breaker for the poloidal field system) followed by a large amount of work about the superconducting coils

  11. Infrared surface temperature measurements for long pulse operation, and real time feedback control in Tore-Supra, an actively cooled Tokamak

    International Nuclear Information System (INIS)

    Guilhem, D.; Adjeroud, B.; Balorin, C.; Buravand, Y.; Bertrand, B.; Bondil, J.L.; Desgranges, C.; Gauthier, E.; Lipa, M.; Messina, P.; Missirlian, M.; Mitteau, R.; Moulin, D.; Pocheau, C.; Portafaix, C.; Reichle, R.; Roche, H.; Saille, A.; Vallet, S.

    2004-01-01

    Tore-Supra has a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T proportional to √(t)) during the discharge until radiation cooling is equal to the incoming heat flux (T > 1800 K). Such an environment is found in most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra infrared thermography system is composed of 7 infrared endoscopes, this system is described in details in the paper, the new JET infrared thermography system is presented and some insights of the ITER set of visible/infrared endoscope is given. (authors)

  12. Infrared surface temperature measurements for long pulse operation, and real time feedback control in Tore-Supra, an actively cooled Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Guilhem, D.; Adjeroud, B.; Balorin, C.; Buravand, Y.; Bertrand, B.; Bondil, J.L.; Desgranges, C.; Gauthier, E.; Lipa, M.; Messina, P.; Missirlian, M.; Mitteau, R.; Moulin, D.; Pocheau, C.; Portafaix, C.; Reichle, R.; Roche, H.; Saille, A.; Vallet, S

    2004-07-01

    Tore-Supra has a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T proportional to {radical}(t)) during the discharge until radiation cooling is equal to the incoming heat flux (T > 1800 K). Such an environment is found in most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra infrared thermography system is composed of 7 infrared endoscopes, this system is described in details in the paper, the new JET infrared thermography system is presented and some insights of the ITER set of visible/infrared endoscope is given. (authors)

  13. Disruption mitigation on Tore Supra

    International Nuclear Information System (INIS)

    Martin, G.; Sourd, F.; Saint-Laurent, F.; Bucalossi, J.; Eriksson, L.G.

    2004-01-01

    During disruptions, the plasma energy is lost on the first wall within 1 ms, forces up to hundred tons are applied to the structures and kA of electrons are accelerated up to 50 MeV (runaway electrons). Already sources of concern in present day tokamaks, extrapolation to ITER shows the necessity of mitigation procedures, to avoid serious damages to in-vessel components. Massive gas injection was proposed, and encouraging tests have been done on Textor and DIII-D. Similar experiments where performed on Tore Supra, with the goal to validate their effect on runaway electrons, observed during the majority of disruptions. 0.1 mole of helium was injected within 5 ms in ohmic plasmas, up to 1.2 MA, either stable, or in a pre-disruptive phase (argon puffing). Beneficial effects where obtained: reduction of the current fall rate and eddy currents, total disappearance of runaway electrons and easy recovery for the next pulse, without noticeable helium pollution of following plasmas. Analysis of the 4 ms period between injection and disruption indicates that to reach these goals, one need to inject enough helium to keep it only partially ionised. It corresponds to 0.1 g for Tore Supra, and extrapolate to hundreds of grams for ITER. (authors)

  14. Disruption mitigation on Tore Supra

    International Nuclear Information System (INIS)

    Martin, G.; Sourd, F.; Saint-Laurent, F.; Bucalossi, J.; Eriksson, L.G.

    2005-01-01

    During disruptions, the plasma energy is lost on the first wall within 1 ms, forces up to hundred tons are applied to the structures and kA of electrons are accelerated up to 50 MeV (runaway electrons). Already sources of concern in present day tokamaks, extrapolation to ITER shows the necessity of mitigation procedures, to avoid serious damages to in-vessel components. Massive gas injection was proposed, and encouraging tests have been done on Textor and DIII-D. Similar experiments where performed on Tore Supra, with the goal to validate their effect on runaway electrons, observed during the majority of disruptions. 0.1 mole of helium was injected within 5 ms in ohmic plasmas, up to 1.2 MA, either stable, or in a pre-disruptive phase (argon puffing). Beneficial effects where obtained: reduction of the current fall rate and eddy currents, total disappearance of runaway electrons and easy recovery for the next pulse, without noticeable helium pollution of following plasmas. Analysis of the 4 ms period between injection and disruption indicates that to reach these goals, one need to inject enough helium to keep it only partially ionised. It correspond to 0.1 g for Tore Supra, and extrapolate to hundred's of grams for ITER. (author)

  15. Ciel, a new breath for Tore Supra

    International Nuclear Information System (INIS)

    Garin, P.

    1998-01-01

    In view of preparing next generation of tokamaks, a new goal has been given to Tore Supra, initially designed for 30 second operation. In order to fulfill this target, with the ultimate aim of injecting in the plasma 25 MW for 1000 s shots: all existing components (including the 6 ergodic divertor modules) present inside the inner vessel of Tore Supra will be removed by the end of 1999, they will be replaced in 2000 by a set of up to date technology components, designed to withstand an overall power of 25 MW for shots up to 1000 s constituting the so-called 'CIEL' project (French acronym for 'Composants Internes Et Limiteur'). It is planned to stop present Tore Supra's operation in autumn 1999, and start again with CIEL environment at the end of 2000. Full capacity (pumping and cooling) will be available in 2002. (author)

  16. Ultra fast frequency sweep heterodyne reflectometer on the Tore Supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, Ph.; Clairet, F.; Chareau, J.M.; Paume, M.; Laviron, C

    1998-09-01

    A new O-Mode dual frequency heterodyne reflectometer has been installed on Tore Supra, introducing outstanding improvement of the phase determination due high dynamic heterodyne detection, and ultra fast sweeps capabilities (10 {mu}s) provided by the solid state source HTOs. This system operates with O-mode electric field polarization in the range of 26 - 36 GHz and has been designed for density profile measurements. The reflectometer launches simultaneously into the plasma two frequencies separated by 320 MHz. The heterodyne detection improves the dynamic, up to 60 db, and is associated with an I.Q. detection to allow separate analysis of amplitude and phase of each reflected signals of both probing waves. Therefore, to calculate the density profile, the group delay can be defined in two different ways: from the derivative of the absolute phase of one of the two probing waves or by calculating the phase difference between the two probing waves. It is explained how ultra fast sweep operations (down to 10 {mu}s) significantly reduce the influence of the plasma turbulence upon the phase measurements. It is also pointed out the importance of filtering carefully the detected signal in order to keep only the information coming from the reflection at the cut-off and to get rid of parasitic reflections. It is shown that the phase difference technique does not remove completely fluctuations such as long radial correlation MHD perturbation. Density profiles determined by the heterodyne reflectometer are in good agreement with the measurements from the other diagnostics of Tore Supra. (author) 16 refs.

  17. Study of production, transport and radiation of carbon impurities near the ergodic divertor in Tore-Supra tokamak

    International Nuclear Information System (INIS)

    Corre, Y.

    2001-11-01

    In a Tokamak thermonuclear reactor, the impurity control is essential to keep the plasma in fusion conditions. Indeed, the impurities which enter the bulk are responsible for the fuel dilution (and hence a significant reduction of the number of fusion reactions). The pollution of fusion plasma by impurities has thus to be as low as possible. In the Tokamak Tore Supra, as in most present day Tokamaks, the main impurity is carbon. First, we have studied the carbon production mechanisms on the Neutralizer Plates of the Tore Supra Ergodic Divertor (where the plasma surface interaction is the most important). For this purpose we have used an endoscope during an experimental campaign in order to measure spectral line brightnesses emitted in the visible wavelength range by low charge carbon ions. The quantitative analysis of the pictures provided by this endoscope, together with measurements by other plasma edge diagnostics, has allowed us to estimate the atom flux extracted from the neutralizer plates during the various density regimes accessed in ED configuration. We have deduced from these calculations an experimental sputtering yield. A comparison with the theoretical sputtering yield allows us to determine the dominant erosion mechanism as a function of the edge plasma density and temperature. This comparative analysis shows that when the edge electron temperature is above 30 eV, the self-sputtering process is the dominant phenomenon for impurity production and bulk contamination. When T e bord is high, the effective erosion is bigger than the erosion due to deuterium ion impacts. This information has then been used to study transport and radiation of this impurity near the neutralizer plates with the 3-D Monte Carlo code BBQ (grill) and other carbon radiation measurements. It has allowed us to characterise the circulation of carbon in the plasma edge and to determine the carbon fraction which enters the confined plasma and the fraction which is rapidly driven back

  18. 2-D mapping of ICRF-induced SOL perturbations in Tore Supra tokamak

    International Nuclear Information System (INIS)

    Colas, L.; Gunn, J.P.; Nanobashvili, I.; Petrzilka, V.; Goniche, M.; Ekedahl, A.; Heuraux, S.; Joffrin, E.; Saint-Laurent, F.; Balorin, C.; Lowry, C.; Basiuk, V.

    2007-01-01

    ICRF-induced SOL modifications are mapped for the first time in 2-D around Tore Supra ICRF antennas using reciprocating Langmuir probes. When probe heads are magnetically connected to powered antennas, radical modifications of floating potentials V float , effective temperatures T eff and ion saturation currents are observed. V float perturbations are located radially near antenna limiters, with a typical extension 2 cm. Poloidally they are locally minimal near the equatorial plane, and maximal near antenna box corners. Two possible interpretations for increased T eff are proposed: localised electron heating and RF loop voltage induced along probe circuit. Both interpretations rely on the generation of parallel RF fields by parallel RF currents on the antenna structure. The topology of such currents could explain the 2-D structure of T eff maps. Both interpretations also imply a positive DC biasing of the antenna environment. Differential biasing of nearby flux tubes drives DC E x B 0 convection that could explain 2-D density patterns

  19. Evidence for a poloidally localized enhancement of radial transport in the scrape-off layer of the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Gunn, J.P.; Boucher, C.; Dionne, M.; Duran, I.; Fuchs, V.; Loarer, T.; Nanobashvili, I.; Panek, R.; Pascal, J.-Y.; Saint-Laurent, F.; Stoeckel, J.; Rompuy, T. van; Zagorski, R.; Adamek, J.; Bucalossi, J.; Dejarnac, R.; Devynck, P.; Hertout, P.; Hron, M.; Lebrun, G.; Moreau, P.; Rimini, F.; Sarkissian, A.; Oost, G. van

    2007-01-01

    Near-sonic parallel flows are systematically observed in the far scrape-off layer (SOL) of the limiter tokamak Tore Supra, as in many L-mode X-point divertor tokamak plasmas. The poloidal variation of the parallel flow has been measured by moving the contact point of a small circular plasma onto limiters at different poloidal angles. The resulting variations of flow are consistent with the existence of a poloidally localized enhancement of radial transport concentrated in a 30 deg. sector near the outboard midplane. If the plasma contact point is placed on the inboard limiters, then the SOL expands to fill all the space between the plasma and the outboard limiters, with density decay lengths between 10 and 20 cm. On the other hand, if the contact point lies on the outboard limiters, the localized plasma outflux is scraped off and the SOL is very thin with decay lengths around 2-3 cm. The outboard radial transport would have to be about two orders of magnitude stronger than inboard to explain these results

  20. Runaway electrons dynamics and confinement in Tore-Supra

    International Nuclear Information System (INIS)

    Chatelier, M.; Geraud, A.; Joyer, P.; Martin, G.; Rax, J.M.

    1989-01-01

    The lack of energy of runaway electrons, confined in Tore Supra tokamak, is studied. Ohmic discharges, obtained with helium gas, exhibit a small amount of runaway electrons on both hard X-ray monitors and neutron sensors. The observations show an important lack of energy for runaway electrons confined in Tore Supra. It is assumed to be dued to a small pitch-angle scattering (a few degrees), and many candidates for this are compared: the strongest known one collisions seems not to be enough by an order of magnitude. Density and magnetic scans on Tore Supra are needed to discriminate between enhanced collisional scattering processes and purely magnetic phenomena

  1. Snake studies on Tore Supra

    International Nuclear Information System (INIS)

    Cristofani, P.; Desgranges, C.; Garbet, X.; Geraud, A.; Gil, C.; Hoang, G.T.; Joffrin, E.; Pecquet, A.L.

    1995-01-01

    Snakes have been achieved after pellet injection in Tore Supra during ohmic as well as ICRH discharges as it has already been observed in other machines. On Tore Supra, high speed H 2 pellets were injected into D 2 plasmas under the specified experimental conditions, the matter is deposited in the centre and snakes are produced in 50% of the cases, but they are created on a second much more internal q=1 surface leading probably to a non monotonic current profile. The properties of the snake, induced current modification and the important role of the bootstrap current in the snake formation are described. (K.A.) 5 refs.; 7 figs

  2. Plasma Edge Control in Tore Supra

    International Nuclear Information System (INIS)

    Evans, T.E.; Mioduszewski, P.K.; Foster, C.; Haste, G.; Horton, L.; Grosman, A.; Ghendrih, P.; Chatelier, M.; Capes, H.; Michelis, C. De; Fall, T.; Geraud, A.; Grisolia, C.; Guilhem, D.; Hutter, T.

    1990-01-01

    TORE SUPRA is a large superconducting tokamak designed for sustaining long inductive pulses (t∼ 30 s). In particular, all the first wall components have been designed for steady-state heat and particle exhaust, particle injection, and additional heating. In addition to these technological assets, a strict control of the plasma-wall interactions is required. This has been done at low power: experiments with ohmic heating have been mainly devoted to the pump limiter, ergodic divertor and pellet injection experiments. Some specific problems arising in large tokamaks are encountered; the pump limiter and the ergodic divertor yield the expected effects on the plasma edge. The effects on the bulk are discussed

  3. M.H.D. activity associated with the q=1 surface in the Tore-Supra tokamak; Activite M.H.D. associee a la surface q=1 dans le tokamak Tore-Supra

    Energy Technology Data Exchange (ETDEWEB)

    Cristofani, P.

    1996-02-12

    In order to increase the temperature, density and confinement time of the plasma energy inside tokamak devices, several heating and fuel injection techniques have been used. However, the increase of the energy content of the central part of the plasma leads to instabilities in the confinement magnetic structure which can degrade the confinement properties and the temperature performances. Inside the plasma, the ``q=1`` surface plays an important role in the confinement process. The aim of this thesis is to study the experimental physics related to this surface with the analysis of the ``saw-tooth`` periodical internal relaxations and of the ``snake`` structure. The first chapter gives a general introduction about thermonuclear fusion and a description of the plasma and of its equilibrium. Chapter 2 is devoted to the description of the soft X-ray tomography, the diagnostic technique used in this work. In chapter 3, a theoretical presentation of plasma stability and a comparison with experimental results obtained in the Tore-Supra tokamak are given. The observations of saw-tooth instabilities are presented with the principal theoretical models which are used to explain this phenomenon. The snake density instability localized in the central part of the plasma is described in chapter 4 with an attempt of interpretation. The equation of the size evolution of a magnetic island was modified to test different models which can explain the snake stability. One model is based on the modification of the bootstrap current induced by the presence of the snake, and on the local modification of the current induced by the accumulation of impurities inside the snake. (J.S.). 107 refs.

  4. M.H.D. activity associated with the q=1 surface in the Tore-Supra tokamak

    International Nuclear Information System (INIS)

    Cristofani, P.

    1996-01-01

    In order to increase the temperature, density and confinement time of the plasma energy inside tokamak devices, several heating and fuel injection techniques have been used. However, the increase of the energy content of the central part of the plasma leads to instabilities in the confinement magnetic structure which can degrade the confinement properties and the temperature performances. Inside the plasma, the ''q=1'' surface plays an important role in the confinement process. The aim of this thesis is to study the experimental physics related to this surface with the analysis of the ''saw-tooth'' periodical internal relaxations and of the ''snake'' structure. The first chapter gives a general introduction about thermonuclear fusion and a description of the plasma and of its equilibrium. Chapter 2 is devoted to the description of the soft X-ray tomography, the diagnostic technique used in this work. In chapter 3, a theoretical presentation of plasma stability and a comparison with experimental results obtained in the Tore-Supra tokamak are given. The observations of saw-tooth instabilities are presented with the principal theoretical models which are used to explain this phenomenon. The snake density instability localized in the central part of the plasma is described in chapter 4 with an attempt of interpretation. The equation of the size evolution of a magnetic island was modified to test different models which can explain the snake stability. One model is based on the modification of the bootstrap current induced by the presence of the snake, and on the local modification of the current induced by the accumulation of impurities inside the snake. (J.S.)

  5. Resistive evolution of current profile in tokamaks, application to the optimization of Tore-supra plasma discharges; Evolution resistive du profil de courant dans les Tokamaks, application a l'optimisation des decharges de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Bregeon, R

    1999-03-01

    In Tokamak plasma physics, current profile shaping has now become a key issue to improve the confinement properties of the plasma discharge. The objective of this work is to study the processes governing the current diffusion when non-inductive current are playing a major role in the discharge. Ultimately, this study aims to identify the key parameters to control the plasma current density profile with external current drive heating systems such as Lower Hybrid Current drive (LHCD) or self generated current drive such as the bootstrap current. Principles of non inductive current drive and heating systems are introduced as well as bootstrap current mechanisms. Then we present the experimental study of plasma parallel electric conductivity to validate existing models. Using these results, the poloidal magnetic field flux diffusion is modelled, using toroidal co-ordinates in order to give an accurate description of the current density profiles evolution. The initial and boundary conditions required for numerical resolution of the diffusion equation are also presented. Finally, we conclude this work with the simulations of two discharges: one with Fast Wave Electron Heating and the second using Lower Hybrid Current Drive. These simulations have multiples aims: validity test of our numerical tool and to show some limits of cylindrical models. Test of electric conductivity and bootstrap current models. To identify the key parameters involved in the current diffusion processes of a high performance plasma discharge on Tore Supra. Such simulations are crucial to determine the amount of non-inductive current required to control and sustain long plasma discharges in steady state. (author)

  6. Evidence for a poloidally localized enhancement of radial transport in the scrape-off layer of the Tore Supra tokamak

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Boucher, C.; Dionne, M.; Ďuran, Ivan; Fuchs, Vladimír; Loarer, T.; Nanobashvili, I.; Pánek, Radomír; Pascal, J.-Y.; Saint-Laurent, F.; Stöckel, Jan; Van Rompuy, T.; Zagórski, R.; Adámek, Jiří; Bucalossi, J.; Dejarnac, Renaud; Devynck, P.; Hertout, P.; Hron, Martin; Lebrun, G.; Moreau, P.; Rimini, F.; Sarkissian, A.; Van Oost, G.

    363-365, - (2007), s. 484-490 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/17th./. Hefei, 22.05.2006-26.05. 2006] R&D Projects: GA ČR GP202/03/P062 Institutional research plan: CEZ:AV0Z20430508 Keywords : Cross-field transport * Edge plasma * Plasma flow * Tore Supra Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.643, year: 2007

  7. Control and monitoring of the Tore Supra toroidal superconducting coils

    International Nuclear Information System (INIS)

    Prou, M.

    1989-07-01

    Light nuclei controlled fusion reactions are seen as a possible way to produce nuclear energy. For this reason, the interest in hot plasma researches in tokamaks has increased. The Tore Supra main characteristic is related to the superconducting magnet coils. They allow a suitable energy balance, however, they require an accurate and preventive fault detection. The Tore Supra machine and the different methods to detect a transition (from superconducting to normal mode) in the toroidal coils are described. The voltage of the coils, the pressure of the helium superfluid at 1.8 K and the electric current in the circuit parallel resistances, are measured. A computer aided control system allows the toroidal field monitoring (current in the coils, fault detection). The superconducting magnet configuration chosen for Tore Supra seems to be suitable for future large Tokamak devices [fr

  8. Scrape-off layer power flux measurements in the Tore Supra tokamak

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Dejarnac, Renaud; Devynck, P.; Fedorczak, N.; Fuchs, Vladimír; Gil, C.; Kočan, M.; Komm, Michael; Kubič, M.; Lunt, T.; Monier-Garbet, P.; Pascal, J.-Y.; Saint-Laurent, F.

    2013-01-01

    Roč. 438, suppl (2013), S184-S188 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/20./. Aachen, 21.05.2012-25.05.2012] Institutional support: RVO:61389021 Keywords : plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.016, year: 2013 http://www.sciencedirect.com/science/article/pii/S0022311513000639#

  9. Self consistent description of plasma equilibrium evolution in Tore Supra

    International Nuclear Information System (INIS)

    Blum, J.; Le Foll, J.; Leloup, C.

    1984-01-01

    A model is presented which describes in a self-consistent way the evolution of the plasma equilibrium in a Tokamak. Numerical simulations are presented for ohmic heating discharges, neutral beam injection, lower hybrid electron heating and current drive in Tore Supra. The various control systems (plasma current, shape and position, coil current sharing) are tested with the code. (author)

  10. Heat Loads On Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Chantant, M.; Beaumont, B.; Ekedahl, A.; Goniche, M.; Moreau, P.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency launchers plasma facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. Lessons are drawned both with regards to Tore Supra possible operational limits and to ITER ICRF launcher design

  11. Determination of electromagnetic modes in oversized corrugated waveguides on the electron cyclotron resonance heating installation at the tokamak Tore Supra; Determination de modes electromagnetiques de guides d'ondes corrugues surdimensionnes sur l'installation de chauffage des electrons de tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Courtois, L

    2001-03-09

    Electron cyclotron resonance heating (ECRH) in the Tore Supra tokamak constitutes an important step in the research aimed at obtaining thermonuclear fusion reactions. Electron heating is achieved by transmitting an electromagnetic wave from the oscillators (gyrotrons) to the plasma via the fundamental mode, propagating in oversized corrugated waveguides. Maximizing the proportion of the gyrotron power coupled to the fundamental waveguide mode is essential for the good functioning of the transmission line and for maximizing the effect on the plasma. This thesis gives all necessary tools for finding the proportion of the fundamental mode and all other modes present in passive components and at the output of the gyrotron as installed in the Tore Supra ECRH plant. This characterisation is based on obtaining amplitude and phase diagrams of the electric field on a plane transverse to the propagation axis. The most difficult part of obtaining these diagrams is measuring the phase which, despite the very short wavelength, is measured directly at low power levels. At high power levels the phase is numerically reconstructed from amplitude measurements for gyrotron characterisation. A complete theoretical study of the phase reconstruction code is given including its validation with theoretical diagrams. This study allows the realisation of a modal characterisation unit electromagnetic for measurement of radiated beams and usable in each part of the ECRH installation. At the end, the complete modal characterisation is given at low level for a mode converter and also at high level for the first series gyrotron installed at TORE SUPRA. (author)

  12. Internal magnetic fluctuations and electron heat transport in the TORE SUPRA Tokamak. Observation by cross polarisation scattering

    International Nuclear Information System (INIS)

    Colas, L.; Paume, M.; Zou, X.L.; Chareau, J.M.; Guiziou, L.; Hoang, G.T.; Michelot, Y.; Gresillon, D.

    1997-03-01

    Magnetic fluctuations (radial size ∼ 5 mm) are measured by a cross polarisation scattering (CPS) diagnostic in TORE SUPRA. These fluctuations are investigated quantitatively in the ohmic and low confinement regimes over a wide range of plasma currents, densities and additional heating powers. Simultaneously, electron heat diffusivities expected from these fluctuations are compared to those obtained by profile analysis. A radial profile of the magnetic fluctuations in the gradient region ( 0.3 e mag = πqRv th (δ B r / B) 2 . Both the order of magnitude and the parametric dependence of χ e mag show similarities with electron diffusivities determined by transport analysis. In particular, a threshold is observed for the dependence of fluctuation-induced heat fluxes on the local temperature gradient, which is close to the critical gradient observed for the measured heat fluxes. (author)

  13. First ECRH modulation experiments in Tore Supra

    International Nuclear Information System (INIS)

    Zou, X.L.; Artaud, J. F.; Bouquey, F.; Clary, J.; Darbos, C.; Giruzzi, G.; Lennholm, M.; Magne, R.; Segui, J.L.

    2003-01-01

    The modulation of Electron Cyclotron Resonance Heating (ECRH) is a powerful tool to investigate the crucial problem of heat transport in tokamaks. This method consists in producing perturbations in the electron temperature by means of highly localised electron heating. By studying the propagation of the heat pulse associated with such perturbations, it is possible to determine both the heat transport coefficient and, as a by-product, the ECRH power deposition. In this work, three methods have been used for electron transport analysis: i) the well-known FFT method; ii) an analytical method, consisting in the simulation of the temperature modulation by an analytical solution of the heat transport equation in cylindrical geometry; iii) the power balance method with the profile analysis. The three methods are applied to the analysis of recent experiments performed on Tore Supra, and their advantages and drawbacks are discussed. (authors)

  14. Particles pumping in Tore Supra

    International Nuclear Information System (INIS)

    Bonnel, P.; Chappuis, P.; Lipa, M.

    1989-01-01

    TORE SUPRA and its peripheral equipments are provided with routine clean high vacuum by turbomolecular pumping. During plasma discharges large quantity of very hot gases activating at plasma edge and plasma density in scrape off layer has to be controlled before they strike violently solid wall provoking increase in impurities content and make density up to disruptive level. A Magnetic Ergodic Divertor made of six winding structures - MED - six Vertical Pumped Limiters - VPL - and one Horizontal Pumped Limiter - HPL - are set in the vacuum chamber in order to cope with plasma-wall interactions and neutral gas recycling. Each apparatus is equipped at front side with thermal shield respectively made of polycristallin and pyrolitic graphite bolted on stainless steel support for MED and HPL whereas for VPL it is made of CFC Aerolor 05 brazed on hardened copper. The total heat removal capacity of these plasma facing components is 12 MW. Design of particles collection openings and ducts conductance allow 10% of capture efficiency, that means for TORE SUPRA a flux of 3 x 10 21 particles/second has to be sorbed by water cooled titanium getter pumps, settled at rear side. All those facilities were put into plasma operation at the beginning of 1989 for a short time. Preliminary observations go along with theoretical predictions, that actions in scrappe-off layer may provoke effects in bulk plasma. Very first results drawn out, show that particle collection and heat removal were effective by MED, VPL and HPL and that plasma behaviour was not disturbed by their presence and actions but instead tendency to improvement was observed

  15. Pellet injection requirements for TORE SUPRA

    International Nuclear Information System (INIS)

    Lafferranderie, J.

    1986-01-01

    The main parameters of TORE SUPRA are outlined and pellet injection requirements to meet plasma density goals are discussed. Topics considered include plasma buildup, plasma refueling and penetration depth

  16. Study and impact of fast electrons diagnosed by electron cyclotron radiation on Tore-Supra tokamak; Etude et influence des electrons rapides diagnostiques par le rayonnement cyclotronique electronique sur le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, P

    1999-12-01

    This thesis aims at characterizing the dynamics of fast electrons generated by the Landau absorption of the hybrid wave and studying their effects on electron cyclotron radiation. The different processes involved in the propagation and resonant absorption of the hybrid wave in plasmas are described. A method such as ray-tracing allows the characterization of the dynamics of heating but this method relies on the hypothesis of geometrical optics. Whenever absorption rate is low as it is in Tore-Supra, the hybrid wave undergoes a series of successive reflections on the edge of the plasma before being completely absorbed. These reflections generate an electromagnetic chaos in which geometrical optics hypothesis are no longer valid. A statistical treatment of the Fokker-Planck equation allows the calculation of the mean distribution function of electrons in the plasma submitted to hybrid wave. The electron cyclotron radiation is then deduced and by assuming that plasma behaves like a black body, a theoretical radiative temperature is calculated. The confrontation of this theoretical temperature profile with experimental values allows the validation of this modeling and the estimation of the effects of fast electrons on temperature measurements. (A.C.)

  17. Heat loads on Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Beaumont, B.; Chantant, M.; Goniche, M.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency (ICRF) launchers plasma-facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. It is found that the most critical items for Tore-Supra operation are localized heat loads on the Faraday screen top left corner and vertical edges. Warming up close to maximum temperature limit originally set for protection of the plasma-facing components is found of high power pulses, but no erosion was observed after detailed inspection of the launcher in Tore-Supra vessel. Yet, the associated heat loads could be limiting for Tore-Supra operation in the future, and some dedicated work is under progress to improve the understanding of these power fluxes, pointing out the importance of getting a better knowledge of particle flows in the scrape of layer

  18. Remote experiment participation on Tore-Supra

    International Nuclear Information System (INIS)

    Theis, J.-M.; Larsen, J.-M.

    2004-01-01

    The DRFC has traditionally had a very large external collaboration involvement. In particular, 15% of the DRFC work is directed towards the JET programme. As a consequence substantial telecommunications facilities have been installed [F.E.D. 60/3 (2002) 449; F.E.D. 60/3 (2002) 459]. A specific station for remote communication has been set up in the Tore-Supra control room, closely coupled to a collaborating team at INRS Que., Canada. This paper describes our pilot experience with the Canadian participation, which gives details of the communication and data sharing tools used to fully work on Tore-Supra

  19. Experience from Tore Supra acquisition system and evolutions

    International Nuclear Information System (INIS)

    Guillerminet, B.; Buravand, Y.; Chatelier, E.; Leroux, F.

    2004-01-01

    The Tore Supra tokamak has been upgraded to explore long duration plasma discharges up to 1000s. Since summer 2001, the acquisition system operates in continuous mode apart of the data processing which is still done after the pulse. In the first part, we explore a few solutions to process continuously the data during the pulse, based on parallel processing on a Linux farm and then on a CONDOR system. The second part is devoted to the Web service exposing the Tore Supra operation. In the last part, the VME acquisition system has been redesigned to keep up with the high data rates required by a few diagnostics. The workflow is now distributed among a few computers. At the end, we give the current status of the realisation and the future planning

  20. Transport and turbulence in TORE SUPRA ohmic discharges

    International Nuclear Information System (INIS)

    Garbet, X.; Payan, J.; Laviron, C.; Devynck, P.; Saha, S.K.; Capes, H.; Chen, X.P.; Coulon, J.P.; Gil, C.; Harris, G.; Hutter, T.; Pecquet, A.L.

    1992-01-01

    The mechanisms underlying the energy confinement behaviour in ohmic tokamak discharges are not yet understood. It is well known that the confinement time increases with the average density and saturates above a critical value of the density, but several explanations exist for this saturation. The present study is an analysis of a set of ohmic discharges in Tore Supra with I p =1.6 MA, B=4 T, R=2.35 m and a=0.78 m, where the average density was increased from 0.9 to 4.2 10 19 m -3 . For these plasma parameters, the energy confinement time given by magnetic measurements saturates for e > ≥ 2.5 10 19 m -3 . It is emphasized here that the onset of ionic turbulence is unlikely in Tore Supra. This conclusion relies on a transport analysis and turbulence measurements by CO 2 laser scattering, whose results are presented in this paper

  1. Steady state operation and control experiments on Tore Supra

    International Nuclear Information System (INIS)

    Saint-Laurent, F.

    2000-01-01

    The main programme of the Tore Supra tokamak is to investigate the route towards long pulse plasma discharges. Tore Supra is thus equipped with a superconducting toroidal magnet, a full set of actively cooled plasma facing components, and a heating and current drive capability based on high power RF systems connected to actively cooled antennas. After pioneering investigations using the LHCD system alone (2 min and zero loop voltage discharges), recent efforts have concentrated on finding scenarios to couple the two RF heating systems in order to perform high power, long duration discharges. To this aim, 6.5 MW, 25 s as well as 4 MW, 60 s discharges have been successfully achieved. At these high power levels, the plasma-wall interaction becomes a critical issue, and recycling fluxes must be controlled to maintain density and to avoid plasma contamination. All these results contributed to the validation of the upgrade of the Tore Supra first wall components (CIEL project) scheduled for 2000. (author)

  2. ORNL compact loop antenna design for TFTR and Tore Supra

    International Nuclear Information System (INIS)

    Taylor, D.J.; Baity, F.W.; Bryan, W.E.; Hoffman, D.J.; McIlwain, R.L.; Ray, J.M.

    1987-01-01

    The goal supplemental ion cyclotron resonance heating (ICRH) of fusion plasma is to deliver power at high efficiencies deep within the plasma. The technology for fast-wave ICRH has reached the point of requiring ''proof-of-performance'' demonstration of specific antenna configurations of specific antenna configurations and their mechanical adequacy for operating in a fusion environment. Oak Ridge National Laboratory (ORNL) has developed the compact loop antenna concept based on a resonant double loop (RDL) configuration for use in both Tokamak Fusion Test Reactor (TFTR) and the Tore Supra ICRH programs. A description and a comparison of the technologies developed in the two designs are presented. The electrical circuit and the mechanical philosophy employed are the same for both antennas, but different operating environments result in substantial differences in the design of specific components. The ORNL TFTR antenna is designed to deliver 4 MW over a 2-s pulse, and the ORNL Tore Supra antenna is designed for 4 MW and essentially steady-state conditions. The TFTR design embodies the first operations compact RDL antenna, and the Tore Supra antenna extends the technology to an operational duty cycle consistent with reactor-relevant applications. 7 refs., 5 figs

  3. Ion Temperature Measurements in the Tore Supra Scrape-Off Layer Using a Retarding Field Analyzer

    International Nuclear Information System (INIS)

    Kocan, M.; Gunn, J.P.; Pascal, J.Y.; Gauthier, E.

    2010-01-01

    The retarding field analyzer (RFA) is one of the only widely accepted diagnostics for measuring the ion temperature T i )in the tokamak scrape-off layer. An overview of the outstanding RFA performance over ten years of operation in Tore Supra tokamak is given and the validation of T i measurements is addressed. The RFA measurements in Tore Supra are found to be well reproducible. The ion-to-electron temperature ratio is higher than one at low-to-moderate ion-electron collisionality regime and converges to unity at high collisionality regime. (authors)

  4. Electron cyclotron emission measurement in Tore Supra

    International Nuclear Information System (INIS)

    Javon, C.

    1991-06-01

    Electron cyclotron radiation from Tore-Supra is measured with Michelson and Fabry-Perot interferometers. Calibration methods, essential for this diagnostic, are developed allowing the determination of electron temperature in the plasma. In particular the feasibility of Fabry-Perot interferometer calibration by an original method is demonstrated. A simulation code is developed for modelling non-thermal electron population in these discharges using measurements in non-inductive current generation regime [fr

  5. Study of plasma-wall interactions in Tore-supra; Etude des phenomenes d'interaction plasma/paroi dans Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Ruggieri, R

    2000-01-01

    In tokamaks the interaction between wall and plasma generates impurities that affect the thermonuclear fusion. This thesis is divided into 2 parts. The first part describes the physico-chemical processes that are involved in chemical erosion, the second part deals with the study of the wear of Tore-supra's walls due to chemical erosion. Chapter 1 presents the wall-plasma interaction and reviews the different processes between plasma and carbon that occur in Tore-supra. Chapter 2 considers the various crystallographic and electronic structures of the carbon that interferes with Tore-supra plasma, the evolution of these structures during irradiation and their temperature dependence are studied. Chapter 3 presents a crystallo-chemical study of graphite samples that have undergone different surface treatments: ionic bombardment, annealing and air exposure. This experimental study has been performed by using energy-loss spectroscopy. It is shown that air exposure modifies the crystallo-chemical structure of surfaces, so it is necessary to prevent air from contaminating wall samples from Tore-supra. Chapter 4 presents a parametric study of chemical erosion rate of plasma facing components (LPM) of Tore-supra. A relation such as Y{sub cd4}{alpha}{gamma}{sup -0.1} gives a good agreement for chemical erosion rate between measurements and the numerical values of the simulation. (A.C.)

  6. Implementation of FCI heating system to the control system of Tore-Supra

    International Nuclear Information System (INIS)

    Wisniewski, S.

    2001-11-01

    This report presents the implementation of the ion cyclotron resonance heating system (FCI) to the instrumentation and control system of the Tore-Supra tokamak. The new plasma heating system involves 3 antennas delivering 12 MW that are required to maintain fusion reactions. This paper is divided into 8 chapters: 1) thermonuclear fusion and Tore-Supra tokamak; 2) hardware system around Tore-Supra, in this chapter the control system and the data acquisition and processing systems are presented; 3) functional analysis, this analysis defines the different needs concerning timing and pilot-controlling, a preliminary proposition of hardware equipment is made; 4) operating modes of FCI; 5) communication within the control system network; 6) communication with the supervisory system of the power stations; 7) management of data exchange with SMX generators; and 8) control of the rate of stationary waves during the injection of power into the plasma

  7. Study of the non inductive current generation in Tore Supra and application to the operational scenario of a continuous tokamak; Etude de la generation de courant non inductive dans Tore Supra et application aux scenarios operationnels d`un tokamak continu

    Energy Technology Data Exchange (ETDEWEB)

    Kazarian-Vibert, F.

    1996-07-05

    Lower Hybrid Current Drive in tokamak plasmas allows to obtain continuous operations, which constitute a necessary step towards a definition of a thermonuclear fusion reactor. The objectives of this work is to define and study fully non inductive steady-state scenarios on Tore Supra. The current diffusion equation is solved to determined precisely the inductive and non inductive current density profiles and their influence on thee time evolution of a discharge. Then, a new operation mode is studied theoretically and experimentally. In this scenario, the transformer primary circuit voltage is controlled in such a way that the flux consumption vanishes. It allows to achieve full steady-state discharges in a fast and reproducible manner. A theoretical flux consumption scaling law during plasma current ramp-up assisted by Lower-Hybrid waves is presented and validated by experimental data, in view to minimized this consumption. The influence of a non monotonic current profile on the confinement and the transport of energy in the plasma is also clearly illustrated by experiments. (author). 138 refs., 16 figs., 1 tab.

  8. Global analysis of ICRF wave coupling on Tore Supra

    International Nuclear Information System (INIS)

    Goniche, M.; Bremond, S.; Colas, L.

    2003-01-01

    The Tore Supra tokamak is equipped with a multi-megawatt ion cyclotron range of frequency (ICRF) system for heating and current drive. The coupling of the fast wave to the plasma, characterized by the distributed coupling resistance along the radiating straps, is a crucial issue in order to launch large RF powers. Many factors can have an effect on ICRF wave coupling. Quantitative prediction from theoretical modelling requires the knowledge of the local inhomogeneous plasma density profile in front of the antenna for running sophisticated antenna codes. In this work, we have rather followed a 'global' approach, based on Tore Supra experimental results, for the parametric study of the coupling resistance. From a large data base covering seven experimental campaigns (∼2250 shots), a scaling law of the coupling resistance including the main parameters of the plasma and of the antenna configuration is established. This approach is found to be reliable for the analysis of coupling in the different scenarios: He/D 2 gas filling, gas/pellets for plasma fuelling, plasma leaning on inner wall/low field side limiter, limiter/ergodic divertor configuration, minority heating/direct electron heating. From one scenario to another, a significant variation of the coefficients of the scaling law is found. The study of these variations allows to get some insight on the main physical mechanisms which influence the ICRF wave coupling in a tokamak operation, such as the wall conditioning and recycling conditions, RF sheaths or frequency. (author)

  9. The acquisition system for Tore Supra 1000 s discharges

    International Nuclear Information System (INIS)

    Guillerminet, B.; How, J.

    2000-01-01

    Long duration discharges are planned for Tore Supra in the near future. A study has been made to detect and correct all the possible limitations of the data acquisition system. Results and analysis of a few 1000 s 'dry-run' test pulses are presented in this paper as well as the solutions foreseen for Tore Supra

  10. Current diffusion and flux consumption in Tore Supra

    International Nuclear Information System (INIS)

    Van Houtte, D.; Talvard, M.; Agostini, E.; Gil, C.; Hoang, G.T.; Lecoustey, P.; Parlange, F.; Rodriguez, L.; Vallet, J.C.

    1991-01-01

    TORE SUPRA has been designed to study long pulse plasmas (t > 30 s) at high plasma current (Ip < 2 MA) associated with high additional power (20 MW). Current diffusion studies are essentially based on the analysis of the plasma discharge paths. The current diffusion rate during the current rise phase is analysed with a numerical code using plasma resistivity profiles from Te profiles measured by the ECE diagnostic. Owing to the fact that the quantity of magnetic flux available in a tokamak is limited, perfect knowledge is required of the various components of the flux consumed in order to minimize consumption and to be able to define a suitable transformer size for future high current tokamak projects

  11. The TORE SUPRA fast reciprocating RF probe

    International Nuclear Information System (INIS)

    Thomas, C.E. Jr.; Harris, J.H.; Haste, G.R.

    1994-01-01

    A fast reciprocating ICRF (Ion Cyclotron Range of Frequencies) probe was installed and operated on TORE SUPRA during 1992/1993. The body of the probe was originally used on the ATF experiment at ORNL. The probe was adapted for use on TORE SUPRA, and mounted on one of the two fast reciprocating probe mounts. The probe consists of two orthogonal single-turn wire loops, mounted so that one loop senses toroidal RF magnetic fields and the other senses poloidal RF magnetic fields. The probe began operation in June, 1993. The probe active area is approximately 5 cm long by 2 cm, and the reciprocating mount has a slow stroke (5 cm/sec) of 30 cm by 2 cm, and the reciprocating mount has a slow stroke (5 cm/sec) of 30 cm and a fast stroke (1.5 m/sec) of about 10 cm. The probe was operated at distances from the plasma edge ranging from 30 cm to -5 cm (i.e., inside the last closed flux surface). The probe design, electronics, calibration, data acquisition and data processing are discussed. First data from the probe are presented as a function of ICRF power, distance from the plasma, loop orientation, and other plasma parameters. Initial data shows parametric instabilities do not play an important role for ICRF in the TORE SUPRA edge and scrape-off-layer (SOL) plasmas. Additionally it is observed that the probe signal has little or no dependence on position in the SOL/plasma edge

  12. Microwave transmission measurements in Tore Supra

    International Nuclear Information System (INIS)

    Segui, J.L.; Giruzzi, G.

    1991-01-01

    A microwave transmission diagnostic system below the electron gyrofrequency is now operating on Tore-Supra. In low density Lower-Hybrid current drive discharges the refraction effects are weak and the measurement of the wave transmission coefficient provides an estimate of the non inductive current. For higher densities, the signal is modulated by refraction effects related to sawteeth and MHD modes. In this case, more elaborated computational techniques are required in order to isolate the absorption effect from the apparent wave attenuation due to refraction. Conversely, at frequencies low enough to avoid cyclotron absorption, microwave transmission can be used as a diagnostic of sawteeth and MHD activity

  13. Ripple losses during ICRF heating in Tore Supra

    International Nuclear Information System (INIS)

    Basiuk, V.; Eriksson, L.-G.; Bergeaud, V.; Chantant, M.; Martin, G.; Nguyen, F.; Reichle, R.; Vallet, J.C.; Delpeche, L.; Surle, F.

    2004-01-01

    The toroidal field coils in Tore Supra are supra-conducting, and their number is restricted to 18. As a result, the ripple is fairly large, about 7% at the plasma boundary. Tore Supra has consequently been equipped with dedicated ripple loss diagnostics, which has allowed ripple loss studies. This paper reports on the measurements made with these diagnostics and provides an analysis of the experimental results, comparing them with theoretical expectations whenever possible. Furthermore, the main heating source accelerating ions in Tore Supra is ion cyclotron resonance range of frequency (ICRF) heating, and the paper provides new information on the ripple losses of ICRF accelerated ions. (author)

  14. Controlled thermonuclear fusion: Tore Supra back bone of the EURATOM-CEA programme for the next ten years

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The decision to grant priority operation status to the French Tokamak Tore Supra will make it possible to start on the construction of this large machine and to bring together at the Cadarache Nuclear Study Centre all the facilities of the CEA for their research on fusion by magnetic confinement. The work is scheduled to begin in 1982 and to last until 1985. The financing is indicated and Tore Supra is briefly described [fr

  15. Evolution of the bonding defect reported on the tiles of the toroidal pumped limiter of the Tore Supra tokamak with infrared analysis

    International Nuclear Information System (INIS)

    Cai Laizhong; Gauthier, Eric; Corre, Yann; Loarer, Thierry; Missirlian, Marc; Martin, Vincent; Moncada, Victor

    2012-01-01

    The bonding of plasma-facing component (PFC) tiles and their possible defects need to be monitored to evaluate the safety during long pulse plasma operations and prevent critical failure, which is very important for ITER and next-step fusion devices. The defect evolutions of two toroidal pumped limiter (TPL) tiles are investigated by analysing the infrared images obtained during Tore Supra experiments from 2006 to 2010 (about 10 000 plasma discharges). The evolution of the defect is characterized by the surface temperature reached in stationary discharge conditions. The comparison of the defect tiles with neighbouring tiles (with no defect) and the evolutions are carried out through the thermal time constants of the tiles. The results show that the two tiles are slowly deteriorating during plasma operation and no plateau of the deterioration is observed in the considered period. By comparing the defect evolutions with the modelling results calculated by CAST3M, the current state of the defect tiles is identified. The defects are significant (about 30% of the tile length or area) and need to be followed and monitored further in the next experimental campaigns of Tore Supra.

  16. Effect of Magnetic Field on Pellet Penetration in Tore Supra

    International Nuclear Information System (INIS)

    Klaywittaphat, P.; Pegourie, B.; Onjun, T.; Imbeaux, F.; Artaud, J. F.; Ge raud, A.; Picha, R.; Poolyarat, N.

    2011-06-01

    Full text: In this work, we aim to study the effect of magnetic field on pellet penetration in Tore Supra Tokamak using CRONOS integrated predictive modeling code. The Tore Supra plasma discharges with LFS pellet injection (discharges 45072-45085 and 45716-45734) are simulated. In these discharges, the toroidal magnetic field varies from 1.9 to 3.8 tesla. Consecutively, 4 to 6 pellets per discharge are applied. The experimental pellet deposition and penetration can be measured by using CCD camera and interferometer. The CRONOS code with pellet module is used to simulate these experimental results. In these simulations, the core transport is calculated using a combination of the GLF23 anomalous core transport model and the NLCASS neoclassical transport model. The pellet ablation in the hot plasma is described using NGPS pellet ablation model. It is found that the simulation results of pellet penetration agree well with experimental data obtained from the CCD camera. In addition, the result shows that the toroidal magnetic field does not have strong influence on pellet penetration

  17. Towards high-power long-pulse operation on Tore Supra

    International Nuclear Information System (INIS)

    2000-01-01

    The Tore Supra tokamak was given the main mission to investigate the route towards long pulse plasma discharges. This includes the problem of heat exhaust and particle control (via the development of high-performance plasma facing components), and in parallel the physics of fully non inductive discharges and its optimization with respect to the confinement. Tore Supra is thus equipped with a superconducting toroidal magnet (maximum magnetic field on axis 4.5T), a full set of actively cooled plasma facing components (PFC), and a heating and current drive capability based on high power RIF systems connected to actively cooled antennas. The encouraging results already obtained, as well as recent progress in PFC, allowed us to envisage a significant improvement in the heat exhaust capability of Tore Supra. The so-called CIEL-project consists in a complete upgrade of the inner chamber of Tore Supra, planned to be installed during the year 2000. The present paper deals with the experimental and modeling activity linked to the preparation of the long-pulse high-power discharges using the present Tore Supra equipment: heating and current drive scenarios, power coupling, confinement and transport studies, discharge control... An overview of the results obtained in that field is presented, as well as the progress required in the coming years, and the expected performance, for the CIEL phase, in terms of current drive and confinement. (author)

  18. Towards high-power long-pulse operation on Tore Supra

    International Nuclear Information System (INIS)

    2001-01-01

    The Tore Supra tokamak was given the main mission to investigate the route towards long pulse plasma discharges. This includes the problem of heat exhaust and particle control (via the development of performant plasma facing components), and in parallel the physics of fully non inductive discharges and its optimisation with respect to the confinement. Tore Supra is thus equipped with a superconducting toroidal magnet (maximum magnetic field on axis 4.5T), a full set of actively cooled plasma facing components (PFC), and a heating and current drive capability based on high power RF systems connected to actively cooled antennas. The encouraging results already obtained, as well as recent progress in PFC, allowed us to envisaged a significant improvement in the heat exhaust capability of Tore Supra. The so-called CIEL-project consists in a complete upgrade of the inner chamber of Tore Supra, planned to be installed during the year 2000. The present paper deals with the experimental and modelling activity linked to the preparation of the long-pulse high-power discharges using the present Tore Supra equipment: heating and current drive scenarios, power coupling, confinement and transport studies, discharge control,... An overview of the results obtained in that field is presented, as well as the progress required in the coming years, and the expected performance, for the CIEL phase, in terms of current drive and confinement. (author)

  19. Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

    International Nuclear Information System (INIS)

    Gargiulo, L.; Cordier, J.J.; Friconneau, J.P.; Grisolia, C.; Palmer, J.D.; Perrot, Y.; Samaille, F.

    2007-01-01

    The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools

  20. Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

    Energy Technology Data Exchange (ETDEWEB)

    Gargiulo, L. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)], E-mail: laurent.gargiulo@cea.fr; Cordier, J.J. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Friconneau, J.P. [CEA-LIST Robotics and Interactive Systems Unit, BP6 F-92265 Fontenay aux Roses Cedex (France); Grisolia, C. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Palmer, J.D. [EFDA CSU, Max-Planck-Institut fuer Plasma Physik Boltzmannstr. 2, D-85748 Garching (Germany); Perrot, Y. [CEA-LIST Robotics and Interactive Systems Unit, BP6 F-92265 Fontenay aux Roses Cedex (France); Samaille, F. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)

    2007-10-15

    The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools.

  1. Edge plasma density convection during ICRH on Tore Supra

    International Nuclear Information System (INIS)

    Becoulet, M.; Colas, L.; Gunn, J.; Ghendrih, Ph.; Becoulet, A.; Pecoul, S.; Heuraux, S.

    2001-11-01

    The 2D edge plasma density distribution around ion cyclotron resonance heating (ICRH) antennae is studied experimentally and numerically in the tokamak Tore Supra (TS). A local density decrease in front of the loaded ICRH antenna ('pump-out' effect) is demonstrated by Langmuir probe measurements in a low recycling regime. An up-down asymmetry in the heat-flux and in the antenna erosion is also observed, and is associated with poloidal variations of the local density. These density redistributions are ascribed to an ExB convection process linked with RF-sheaths. To assess this interpretation, the 2D transport code CELLS was developed for modeling the density distribution near an antenna. The code takes into account perpendicular diffusion, parallel transport and convection in RF-sheath-driven potentials given by the 3D-antenna code ICANT. The strong density differences obtained in simulations reproduce up-down asymmetries of the heat fluxes. (authors)

  2. The Tore-Supra ion source

    International Nuclear Information System (INIS)

    Fumelli, M.; Jequier, F.; Bottiglioni, F.

    1986-01-01

    The Tore-Supra neutral beam injectors required an ion source with an extraction area of 113 x 6.6 cm 2 . 40 A in deuterium must be available during up to 30 s. Fitted to these requirements, a plasma source with a multipolar magnetic configuration has been constructed and operated at FAR, firstly in H 2 , on a test stand and then, in D 2 , on the prototype injector. The nominal ion current density of 150 mA/cm 2 has been achieved with a uniformity better than 10%, in pulses of 1 s. Two discharge regimes, depending on different magnetic topologies, have been studied with regard to the discharge efficiency, deuterons yield and plasma uniformity. Proton contents up to 85% at 100 mA/cm 2 have been measured. (author)

  3. The Tore-Supra ion source

    International Nuclear Information System (INIS)

    Fumelli, M.; Jequier, F.; Bottiglioni, F.

    1987-01-01

    The Tore-Supra neutral beam injectors require an ion source with an extraction area of 113 x 6.6 cm 2 . 40 A in deuterium must be available during up to 30 s. Fitted to these requirements, a plasma source with a multipolar magnetic configuration has been constructed and operated at FAR, firstly in H 2 , on a test strand and then, in D 2 , on the prototype injector. The nominal ion current density of 150 mAcm 2 has been achieved with a uniformity better than 10%, in pulses of 1 s. Two discharge regimes, depending on different magnetic topologies, have been studied with regard to the discharge efficiency, deuterons yield and plasma uniformity. Proton contents up to 85% at 100 mAcm 2 have been measured

  4. Pellet fuelling in Tore Supra long discharges

    Science.gov (United States)

    Géraud, A.; Bucalossi, J.; Loarer, T.; Pégourié, B.; Grisolia, C.; Gros, G.; Gunn, J.

    2005-03-01

    A new pellet injector, able to inject continuously hydrogen or deuterium pellets, was installed on Tore Supra in 2003 and preliminary experiments aiming to fuel long discharges were performed. In combination with Lower Hybrid (LH) Current Drive, pure pellet fuelled discharges lasting up to 2 min were achieved. The LH power was switched off just before each pellet injection (LH notching) to maintain a relatively deep pellet penetration by reducing the energy of the super-thermal electrons driven by the LH wave. A comparison, based on a particle balance study, between two comparable pellet fuelled and gas fuelled discharges has been done. In the two cases, the volume average density is the same and the analysis shows that the particle source, the pumped flux and the wall retention are similar and appear to be independent of the fuelling method for the low plasma current and density conditions considered ( Ip = 0.6 MA, = 1.5 × 10 19 m -3).

  5. Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

    International Nuclear Information System (INIS)

    Laurent Gargiulo, L.; Cordier, J.-J.; Samaille, F.; Grisolia, Ch.; Perrot, Y.; Olivier, D.; Friconneau, J.-P.; Palmer, J.

    2006-01-01

    The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier. This project called AIA (Articulated Inspection Arm) is currently being developed at CEA under a European EFDA work program. The paper describes the detailed design, the manufacturing processes and the results of the first module test campaign in the CEA Tore Supra ME60 facility, at representative vacuum, temperature and nominal loading conditions. The second part of this work that is reported in the paper, concerns the description of the whole integration of the device on the Tore Supra tokamak that is foreseen to be operated on Tore Supra early 2007. The deployer system and the 10 m long storage vacuum vessel are presented. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development is presented in the paper. It will allow close inspection of the Tore Supra Plasma Facing Components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. Such viewing process could be used on ITER during the early stage of operation under a limited radiation level. The AIA technology is also showing promising potential for generic application in alternative systems for ITER. The feasibility study for viewing inspection of the beam line components in the neutral beam test facility is presented. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be

  6. Integration of advanced feedback control techniques on Tore Supra

    International Nuclear Information System (INIS)

    Barana, O.; Basiuk, V.; Bucalossi, J.

    2006-01-01

    Tore Supra tokamak plays an important role in development and optimisation of steady-state scenarios. Its real-time feedback control system is a key instrument to improve plasma performances. For this reason, new feedback control schemes have been recently put into operation and others are being developed. This work deals with the implementation in Tore Supra of these advanced algorithms, reports the technical details and shows the first positive results that have been achieved. For instance, encouraging results have been obtained in the field of profiles control. Controls of the full width at half maximum of the suprathermal electrons local emission profile at very low loop voltage and of the maximum of the thermal Larmor radius, normalised to the characteristic length of the electron temperature gradient, have been attained. While the first quantity can be directly associated to the current profile, the second one characterises the pressure profile. A new feedback control algorithm, employed to maximise a given quantity by means of a '' Search Optimisation '' technique, has been effectively tested too: the hard X-ray width has been maximised with simultaneous use of lower hybrid heating power and wave parallel index as actuators. These and other promising results, whose detailed description will be given in the article, have been obtained thanks to the real-time availability of several diagnostic systems. Using a shared memory network as communication layer, they send their measurements to a central computing unit that, in its turn, dispatches the necessary requirements to the actuators. A key issue is the possibility to integrate these controls in such a way as to cope with different requests at the same time. As an example, simultaneous control of the plasma current by means of the lower hybrid heating power, of the loop voltage by means of the poloidal field system and of the hard X-ray width through the lower hybrid heating phase shift has been successfully

  7. Water leak localization and recovery in Tore Supra

    International Nuclear Information System (INIS)

    Martinez, A.; Samaille, F.; Chantant, M.; Hatchressian, J.C.

    2007-01-01

    For almost 20 years at Tore Supra, plasma facing components (PFCs) are actively cooled by a pressurized water primary loop. Tore Supra can be considered as ITER relevant on this particular aspect. During plasma operation, it happened, that unexpected localized power deposits damaged a PFC leading to more or less large water leaks in the vacuum vessel. The improvement of the procedure to localize the leaky circuits, the investigation of technical solutions for minimizing the amount of water from steam condensation in the vacuum vessel and the optimisation of the quality of the draining and the drying of Tore Supra cooling loop circuits are the result of the experience gained during several years by the analysis of the water leak from plasma facing components (PFCs) and their consequences. Different new specific systems designed and installed during this year to fulfil these objectives are described in this paper

  8. New developments on Tore Supra data acquisition units

    International Nuclear Information System (INIS)

    Leroux, F.; Caulier, G.; Ducobu, L.; Goniche, M.; Antar, G.

    2012-01-01

    The Tore Supra data acquisition system (DAS) was designed in the early eighties and has considerably evolved since then. Three generations of data acquisition units still coexist. As cost and maintenance of different operating systems is expensive, it was decided to explore an alternative solution based on an open source operating system (OS) with a disk-less system for the fourth generation. In 2010, Linux distributions for VME bus and PCI bus systems have been evaluated and compared to Lynx OS TM real time OS. The results obtained allowed to choose a version of Linux for VME and PC platform for DAS on Tore Supra. In 2011, the Tore Supra DAS dedicated software was ported on a Linux disk-less PCI platform. The new generation was successfully tested during real plasma experiments on one diagnostic, called DCEDRE. (authors)

  9. Ohmic discharges in Tore Supra - Marfes and detached plasmas

    International Nuclear Information System (INIS)

    Vallet, J.C.

    1990-01-01

    The Tore Supra plasma characteristics are given. The observed discharges are either leaning on the graphite inner first wall or limited by movable pump limiters located outboard and at the bottom of the vacuum chamber. The particular plasma conditions which lead to marfes and detached plasmas in ohmically heated He and D2 discharges limited by the inner wall are investigated. The results show that the ratio of radiated power to ohmic power increase linearly with M.g. As M.g rises, attached plasma, marfe and detached plasma are sequentially observed. Detached plasma with an effective radius as small as. 7 times the limiter radius was observed on Tore Supra

  10. Numerical modelling of pump limiter biasing on TEXTOR-94 and Tore Supra

    International Nuclear Information System (INIS)

    Gerhauser, H.; Claassen, H.A.; Mank, G.; Zagorski, R.; Loarer, T.; Gunn, J.; Boucher, C.

    2002-01-01

    The two-dimensional multifluid code TECXY has been used to model the biasing (with respect to the first wall) of the toroidal belt limiter ALT-II on the tokamak TEXTOR-94 and of the new toroidal pump limiter being installed on Tore Supra tokamak in the framework of the CIEL project. It is well known that the edge flow pattern can be influenced by the poloidal electric drifts from imposing radial electric fields. The modelling with TECXY introduces imprinted bias currents in the scrape-off layer (SOL) for the case of negative (limiter) biasing, and imprinted bias potentials for the case of positive biasing. This allowed us to simulate sufficiently well the experimental I-V characteristics for either biasing of ALT-II and also reproduced the essential features and trends of the observed plasma profiles in the SOL of TEXTOR-94. For negative biasing a moderate improvement of the pumping exhaust efficiency can be achieved in the case of TEXTOR. For Tore Supra, however, only a negligible improvement of the limiter performance with biasing can be predicted, which is explained by the relatively weak drift flows in Tore Supra. (author)

  11. Management of water leaks on Tore Supra actively cooled fusion device

    International Nuclear Information System (INIS)

    Hatchressian, J.C.; Gargiulo, L.; Samaille, F.; Soler, B.

    2005-01-01

    Up to now, Tore Supra is the only fusion device fully equipped with actively cooled Plasma Facing Components (PFCs). In case of abnormal events during a plasma discharge, the PFCs could be submitted to a transient high power density (run away electrons) or to a continuous phenomena as local thermal flux induced by trapped suprathermal electrons or ions). It could lead to a degradation of the PFC integrity and in the worst case to a water leak occurrence. Such water leak has important consequence on the tokamak operation that concerns PFCs themselves, monitoring equipment located in the vacuum vessel or connected to the ports as RF antennas, diagnostics or pumping systems. Following successive water leak events (the most important water leak, that occurred in September 2002, is described in the paper), a large feedback experience has been gained on Tore supra since more than 15 years that could be useful to actively cooled next devices as W7X and ITER. (authors)

  12. Experience gained from high heat flux actively cooled PFCs in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.; Bayetti, P.; Brosset, C.; Bucalossi, J.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Ghendrih, Ph.; Guilhem, D.; Gunn, J.; Loarer, T.; Lipa, M.; Mitteau, R.; Pegourie, B.; Reichle, R.; Schlosser, J.; Tsitrone, E.; Vallet, J.C.

    2005-01-01

    The implementation of actively cooled high heat flux plasma facing components (PFCs) is one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW/m 2 of nominal convected heat flux. Technical information is drawn from the whole development up to the industrialisation and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non-destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final 'burial' of about half of the injected gas during long discharges

  13. Technical and human feedbacks about Tore Supra control system

    Energy Technology Data Exchange (ETDEWEB)

    Fejoz, P.; Baudet, J.; Lebourg, P. [CEA Cadarache, IRFM, 13 - Saint-Paul-lez-Durance (France)

    2009-07-01

    Full text of publication follows: During the 23 years of the Tore Supra tokamak operation, we went through several technological boom in many domains such as electronics, software development, robotics and automatics. Even if the initial control-command system fitted all the requirements and satisfied the users, it was not possible to maintain such a system over the years without deep changes. Important issues have appeared: some software or hardware were no more maintained by the suppliers, the performances were not up to date, the human interface were outdated, requirements evolved. The goals were to upgrade and make durable the application programming interface park and the related human-machine interfaces. To reach these objectives we had to face several challenges: -) The training of the staff (programmers and end-users); -) The integration of new equipment in the current architecture; -) The cohabitation between new and old systems; -) The continuity of the operation; -) Make the good choice to go along without any new overhaul. (authors)

  14. Improvement of the Gyrotron TH 1506B for Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Magne, R.; Bouquey, F.; Clary, J.; Darbos, C.; Jung, M.; Lambert, R.; Lennholm, M.; Roux, D. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Alberti, S.; Hogge, J.P. [Association Euratom-Confederation Suisse, EPFL - CRPP, Lausanne (Switzerland); Bariou, D.; Legrand, F.; Lievin, C. [Thales Electron Devices, Microwave Tubes and Devices, 78 - Velizy (France); Arnold, A.; Thumm, M. [Association Euratom-FZK, IHM, Karlsruhe (Germany)

    2004-07-01

    For heating and current drive experiments by electron cyclotron resonance interaction on Tore Supra tokamak, a new generator is presently under construction. The generator is made of 6 gyrotrons TH-1506B. Tests made on the first series gyrotron have revealed some unexpected limitations: the most visible was an insufficient cooling of the mirror box, another one was the observation of a spurious oscillation at the output of the gyrotron at a frequency close to 119,7 GHz. The design of the mirror box has been modified, it was decided to use a double-wall structure with water flowing between. As for preventing the spurious oscillation, calculations have shown that it would be sufficient to add a very small conical angle to the launcher profile. A new generator has been built according to this new design. Tests have shown that previous limitations have disappeared but the output beam seems to be not perfectly Gaussian probably due to the launcher. Studies are being made to understand this problem in order to optimize the design for the following 5 gyrotrons. (A.C.)

  15. Simulation of core turbulence measurement in Tore Supra ohmic regimes

    NARCIS (Netherlands)

    Hacquin, S.; Citrin, J.; Arnichand, H.; Sabot, R.; Bourdelle, C.; Garbet, X.; Kramer-Flecken, A.; Tore Supra team,

    2016-01-01

    This paper reports on a simulation of reflectometry measurement in Tore Supra ohmic discharges, for which the experimental observations as well as gyrokinetic non-linear computations predict a modification of turbulence spectrum between the linear (LOC) and the saturated ohmic confinement (SOC)

  16. Runaway electrons dynamics and confinement in TORE-SUPRA

    International Nuclear Information System (INIS)

    Chatelier, M.; Geraud, A.; Joyer, P.; Martin, G.; Rax, J.M.

    1989-01-01

    Ohmic discharges in TORE-SUPRA are sufficiently long (∼ 6 s) for runaway electrons (R.E.) to reach a steady energy state: their energy limit is determined by the balance between parallel electric field acceleration (20 MeV/V.s in TORE-SUPRA) and radiation losses due to the curvature of the trajectories. When R.E. energy is supposed to be only parallel, this provides estimate of order of 70 MeV (value usually called 'synchrotron limit') reached in less than 2 seconds. Experimental observations on TORE-SUPRA of photoneutron emission together with residual induced radioactivity in the first wall components tend to prove that the actual value is much lower than 70 MeV (i.e. 15-35 MeV). Earlier observations in ORMAK, PLT and TFR already showed R.E. energy a slightly less than expected from standard loop voltage acceleration calculations. Explanations given for this lack of energy (as skin-effect lowering the electric field during the ramp-up phase or balance between continuous creation and losses) seems not to hold on TORE-SUPRA and therefore another mechanism must be considered to explain the R.E. energy limitation. 4 refs., 2 figs

  17. Lower hybrid current drive in Tore Supra and JET

    International Nuclear Information System (INIS)

    Moreau, D.; Gormezano, C.

    1991-01-01

    Recent Lower Hybrid Current Drive (LHCD) experiments in TORE SUPRA and JET are reported. Large multijunction launchers have allowed the coupling of 5MW to the plasma for several seconds with a maximum of 3.8 kW/cm 2 . Measurements of the scattering matrices of the antennae show good agreement with theory. The current drive efficiency in TORE SUPRA is about 0.2 x 10 20 Am -2 /W with LH power alone and reaches 0.4 x 10 20 Am -2 /W in JET thanks to a high volume-averaged electron temperature (1.9 keV) and also to a synergy between Lower Hybrid and Fast Magnetosonic Waves. At n e = 1.5 x 10 19 m -3 in TORE SUPRA, sawteeth are suppressed and m = 1MHD oscillations the frequency of which clearly depends on the amount of LH power are observed on soft X-rays, and also on non-thermal ECE. In Jet ICRH produced sawtooth free periods are extended by the application of LHCD and current profile broadening has been clearly observed consistent with off-axis fast electron populations. LH power modulation experiments performed in TORE SUPRA at n e = 4 x 10 19 m -3 show a delayed central electron heating despite the off-axis creation of suprathermal electrons, thus ruling out the possibility of a direct heating through central wave absorption. A possible explanation in terms of anomalous fast electron transport and classical slowing down would yield a diffusion coefficient of the order of 10 m 2 /s for the fast electrons. Finally, successful pellet fuelling of a partially LH driven plasma was obtained in TORE SUPRA, 28 successive pellets allowing the density to rise to n e = 4 x 10 19 m -3 . This could be achieved by switching the LH power off for 90 ms before each pellet injection, i.e. without modifying significantly the current density profile

  18. Engineering studies for the installation of an axi-symmetric metallic divertor in Tore Supra

    International Nuclear Information System (INIS)

    Doceul, L.; Portafaix, C.; Bucalossi, J.; Saille, A.; Bertrand, B.; Lipa, M.; Missirlian, M.; Jiolat, G.; Samaille, F.; Soler, B.

    2011-01-01

    Tore Supra (TS) has been designed to operate using technologies that allow long plasma operation (a few minutes), by means of superconducting magnets and actively-cooled high heat flux plasma facing components (PFCs). Actively cooled tungsten PFC will be used in the baffle area of the first ITER divertor. In order to validate such a technology fully (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axi-symmetric divertor in the tokamak Tore Supra has been studied . With this second major upgrade, Tore Supra should be able to address the problematic of long plasma discharges with a metallic divertor. The proposed divertor is made up of two stainless steel casings containing a copper coil winding located at the top and bottom area of the vacuum vessel. These casings are firmly maintained by connection beams and protected by PFC. This paper describes the mechanical design of this major component and its integration in TS, the associated electromagnetic and thermomechanical analysis, the manufacturing issues and finally the integration of ITER representative PFCs.

  19. Analysis of density fluctuations in the Tore Supra tokamak. Up-down asymmetries and limiter effect on plasma turbulence; Etude des fluctuations de density dans les plasmas du tokamak Tore Supra. Asymetries haut-bas et effet du limiteur sur la turbulence

    Energy Technology Data Exchange (ETDEWEB)

    Fenzi, Ch

    1999-10-29

    In magnetic fusion devices, the optimisation of the power deposition profile on plasma facing components crucially depends on the heat diffusivity across the magnetic field fines, which is determined by the plasma edge turbulence. In this regard, spatial asymmetries of plasma edge turbulence are of great interest. In this work, we interest in up-down asymmetries of density fluctuations which are usually observed in Tore Supra, using a coherent light scattering experiment. It is shown that these asymmetries are correlated to the plasma edge geometrical configuration (plasma facing components, limiters). In fact, the plasma-limiter interaction induces locally in the plasma edge and the SOL (r/a > 0.9) an additional turbulence with short correlation length along the magnetic field fines, which spreads in the plasma core (0.9 {>=} r/a {>=} 0.5). The resultant up-down asymmetry weakly depends on density, increases with the edge safety factor, and inverts when the plasma current direction is reversed. Such up-down asymmetry observations bring strong impact on edge turbulence and transport models, which usually predict a ballooning of the turbulence in the high-field side but not an up-down asymmetry. A possible model is proposed here, based on the Kelvin Helmholtz instability. (author)

  20. Synergy between electron cyclotron and lower hybrid current drive on Tore Supra

    International Nuclear Information System (INIS)

    Giruzzi, G.; Artaud, J.F.; Dumont, R.J.; Imbeaux, F.; Bibet, P.; Berger-By, G.; Bouquey, F.; Clary, J.; Darbos, C.; Ekedahl, A.; Hoang, G.T.; Lennholm, M.; Maget, P.; Magne, R.; Segui, J.L.; Bruschi, A.; Granucci, G.

    2005-01-01

    Improvement (up to a factor ∼ 4) of the electron cyclotron (EC) current drive efficiency in plasmas sustained by lower hybrid (LH) current drive has been demonstrated in stationary conditions on the Tore Supra tokamak. This was made possible by feedback controlled discharges at zero loop voltage, constant plasma current and density. This effect, predicted by kinetic theory, results from a favorable interplay of the velocity space diffusions induced by the two waves: the EC wave pulling low-energy electrons out of the Maxwellian bulk, and the LH wave driving them to high parallel velocities. (author)

  1. Rapid global response of the electron temperature during pellet injection on Tore Supra

    International Nuclear Information System (INIS)

    Talvard, M.

    1993-06-01

    During pellet injection in the Tore Supra tokamak, a very quick electron temperature drop in the whole plasma column has been observed by means of a fast acquisition ECE Fabry-Perot interferometers system. The time delay of the temperature drop between plasma edge and center is less than 20 microseconds, corresponding to a propagation velocity of the order of 25 km/s, much larger than both the pellet velocity and the ordinary diffusion velocity. A model of neutral atom diffusion, in which charge-exchange process plays a key role, is proposed to explain such phenomena

  2. Methodology for the design of diagnostic windows for Tore Supra

    International Nuclear Information System (INIS)

    Missirlian, Marc; Lipa, M.; Portafaix, C.; Gil, C.; Rey, G.

    2003-01-01

    The next generation of fusion experiments will operate for much longer pulse lengths within a context of high power density, introducing several new requirements for diagnostics compared with existing experiments. In this context, the upgrade of Tore Supra (CIEL Project) foresees high power and high radiating plasma scenario during very long pulse operation. This long plasma operation imposes the improvement of diagnostic systems and the design of thermally resistant in-vessel components with reliable window assemblies. Hence, the specific requirements and the methodology adopted to design Tore Supra/CIEL (TS) high thermal loaded diagnostic windows are summarised in this paper. Thermal and mechanical analyses are reported for different window materials and assembling methods

  3. The plasma facing components of the Tore Supra ICRF antenna

    International Nuclear Information System (INIS)

    Beaumont, B.; Agarici, G.; Gauthier, E.; Kuus, H.; Schlosser, J.

    1994-01-01

    Two generations of Faraday shields for the Tore Supra ICRH antennas interacting with the edge plasma are presented. The last one, using a film of boron carbide as protective material performs well, proving the relevance of this technique for in vessel equipment submitted to low power fluxes. The different lateral protections used on Tore Supra are submitted to high power fluxes. Finite element calculations allow to assess their performances. One type, using Boron Carbide, can be used to measure the local heat flux. The estimation of this flux confirm the specificity of the edge/RF interaction, which is more than one order of magnitude above the exponential decay observed in ohmic plasmas. (author) 11 refs.; 1 fig

  4. MHD activity triggered by monster sawtooth crashes on Tore Supra

    International Nuclear Information System (INIS)

    Maget, P; Artaud, J-F; Eriksson, L-G; Huysmans, G; Lazaros, A; Moreau, P; Ottaviani, M; Segui, J-L; Zwingmann, W

    2005-01-01

    The crash of monster sawteeth in Tore Supra ion cyclotron resonance heated plasmas is observed to trigger long-lived magneto hydrodynamic (MHD) activity, dominated by a (m = 3, n = 2) magnetic perturbation at the edge. This phenomenon is reminiscent of the triggering of neoclassical tearing modes, although in Tore Supra the MHD activity decays and eventually vanishes. It can be explained by the linear destabilization of the (3, 2) mode as the current sheet developed in the non-linear stage of the internal kink relaxation gets closer to q = 3/2. However, the lifetime of the (3, 2) island is longer than the period of linear instability. We find that the neoclassical drive is essential for explaining the observed lifetime and width of the island, although the overall dynamics is controlled by the relaxation of the current profile on a resistive time scale

  5. The TORE SUPRA 300 W - 1.75 K refrigerator

    International Nuclear Information System (INIS)

    Gistav, G.M.

    1986-01-01

    The TORE SUPRA refrigerator design and manufacture have both been governed by several strict design criteria. These include reliable operation over long periods (8,000 hours per year), pulsed thermal loads at 1.75 K and 4.0 K (every 4 minutes), fully automatic control in the various operating modes, low operating costs and acting as a technical demonstration so that larger future designs could be extrapolated from this base. The paper reviews these criteria, and presents the current status

  6. 3400 m/s deuterium pellet injector for Tore Supra

    International Nuclear Information System (INIS)

    Perin, J.P.; Geraud, A.

    1995-01-01

    This paper reports on the Tore Supra high velocity pellet injector which was built in Grenoble and after qualification tests installed on Tore Supra Tokomak where it is used for plasma and ablation studies. By using a two stage light gas gun (TSG) and two cells (φ = 3 mm or 4 mm), unsupported pellets pellets (1 to 3.5 10 21 atoms) made directly in the gun by > [1] have been launched into Tore Supra plasma at speeds between 2400m/s and 3400m/s with a reliability of 80%. These higher pellets velocities (> 2500 m/s) [2] are obtained by the optimization of a TSG and the search for the cryogenic conditions of freezing deuterium with good mechanical properties. In particular, the impurities concentration in deuterium during the condensation process has been studied. Several tens pellets have been injected into ohmically and ICR heated plasma and during LH current drive experiments with a good reliability in the range of 3000m/s. These experiments allowed us to extend significantly the ablation data base. Central penetrations can be reached even for high temperatures plasma (3-5 keV) and very peaked density profiles have been obtained in ohmically and ICR heated plasmas A transient improved confinement regime is then observed, which presents some features similar to the PEP regime obtained on JET. (orig.)

  7. Excitation of beta Alfven eigenmodes in Tore-Supra

    International Nuclear Information System (INIS)

    Nguyen, C; Garbet, X; Sabot, R; Goniche, M; Maget, P; Basiuk, V; Decker, J; Elbeze, D; Huysmans, G T A; Macor, A; Segui, J-L; Schneider, M; Eriksson, L-G

    2009-01-01

    Modes oscillating at the acoustic frequency and identified as beta Alfven eigenmodes (BAEs) have been observed in Tore-Supra under ion cyclotron resonant heating. In this paper, the linear excitation threshold of these modes, thought to be driven by suprathermal ions, is calculated and compared with Tore-Supra observations. Similar studies of the linear excitation threshold of energetic particles driven modes were carried out previously for toroidal Alfven eigenmodes or fishbones. In the case of BAEs, the main point is to understand whether the energetic particle drive is able to exceed ion Landau damping, which is expected to be important in the acoustic frequency range. For this, the BAE dispersion relation is computed and simplified in order to derive a tractable excitation criterion suitable for comparison with experiments. The observation of BAEs in Tore-Supra is found to be in agreement with the calculated criterion and confirms the possibility to trigger these modes in the presence of ion Landau damping. Moreover, the conducted analysis clearly puts forward the role of the global tunable parameters which play a role in the BAE excitation (the magnetic field, the density etc), as well as the role of some plasma profiles. In particular, the outcome of a modification of the shear or of the heating localization is found to be non-negligible and it is discussed in the paper.

  8. Lower hybrid current drive in Tore Supra and jet

    International Nuclear Information System (INIS)

    Moreau, D.; Gormezano, C.

    1991-07-01

    Recent Lower Hybrid Current Drive (LHCD) experiments in TORE SUPRA and JET are reported. Large multijunction launchers have allowed the coupling of 5 MW to the plasma for several seconds with a maximum of 3.8 kW/cm 2 . Measurements of the scattering matrices of the antennae show good agreement with theory. The current drive efficiency in TORE SUPRA is about 0.2 x 10 20 Am -2 /W with LH power alone and reaches 0.4 x 10 20 Am -2 /W in JET thanks to a high volume-averaged electron temperature (1.9 keV) and also to a synergy between Lower Hybrid and Fast Magnetosonic Waves. At average n e = 1.5 x 10 19 m -3 in TORE SUPRA, sawteeth are suppressed and m = 1 MHD oscillations the frequency of which clearly depends on the amount of LH power are observed on soft X-rays, and also on non-thermal ECE. In JET ICRH produced sawtooth-free periods are extended by the application of LHCD (2.9 s. with 4 MW ICRH) and current profile broadening has been clearly observed consistent with off-axis fast electron populations. LH power modulation experiments performed in TORE SUPRA at average n e = 4 x 10 19 m -3 show a delayed central electron heating despite the off-axis creation of suprathermal electrons, thus ruling out the possibility of a direct heating through central wave absorption. A possible explanation in terms of anomalous fast electron transport and classical slowing down would yield a diffusion coefficient of the order of 10 m 2 /s for the fast electrons. Other interpretations such as an anomalous heat pinch or a central confinement enhancement cannot be excluded. Finally, successful pellet fuelling of a partially LH driven plasma was obtained in TORE SUPRA, 28 successive pellets allowing the density to rise to average n e = 4 x 10 19 m -3 . This could be achieved by switching the LH power off for 90 ms before each pellet injection, i.e. without modifying significantly the current density profile

  9. Tore-Supra infrared thermography system, a real steady-state diagnostic

    International Nuclear Information System (INIS)

    Guilhem, D.; Bondil, J.L.; Bertrand, B.; Desgranges, C.; Lipa, M.; Messina, P.; Missirlian, M.; Portafaix, C.; Reichle, R.; Roche, H.; Saille, A.

    2005-01-01

    Tore-Supra Tokamak (I p = 1.5 MA, B t = 4 T) has been constructed with a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components (PFCs) for high-performance long pulse plasma discharges. When not actively cooled, plasma facing components can only accumulate a limited amount of energy since the temperature increases continuously during the discharge until radiation cooling equals the incoming heat flux. Such an environment is found in the JET Tokamak [JET Team, IAEA-CN-60/A1-3, Seville, 1994] and on TRIAM [M. Sakamoto, H. Nakashima, S. Kawasaki, A. Iyomasa, S.V. Kulkarni, M. Hasegawa, E. Jotaki, H. Zushi, K. Nakamura, K. Hanada, S. Itoh, Static and dynamic properties of wall recycling in TRIAM-1M, J. Nucl. Mater. 313-316 (2003) 519-523] [Y. Kamada, et al., Nucl. Fusion 3 (1999) 1845]. In Tore-Supra, the surface temperature of the actively cooled plasma facing components reach steady state within a second. We present here the Tore-Supra thermographic system, made of seven endoscope bodies equipped so far with eight infrared (IR) cameras. It has to be noted that this diagnostic is the first diagnostic to be actively cooled, as required for steady state. The main purpose of such a diagnostic is to prevent the plasma to damage the actively cooled plasma facing components (ACPFCs), which consist of the toroidal pumped limiter (TPL), 7 m 2 , and of five radio-frequency antennae, 1.5 m 2 each

  10. Experimental demonstration of synergy between electron cyclotron and lower hybrid current drive on Tore Supra

    International Nuclear Information System (INIS)

    Artaud, J.F.; Giruzzi, G.; Dumont, R.J.; Imbeaux, F.; Bibet, P.; Bouquey, F.; Clary, J.; Ekedahl, A.; Hoang, G.T.; Lennholm, M.; Magne, R.; Segui, J.L.

    2004-01-01

    Non-inductive current drive (CD) has two main applications in tokamaks: sustainment of a substantial fraction of the toroidal plasma current necessary for the plasma confinement and control of the plasma stability and transport properties by appropriate shaping of the current density profile. For the first kind of applications, lower hybrid (LH) waves are known to provide the highest efficiency (defined as the ratio of the driven current to the injected wave power), although with limited control capability. Conversely, electron cyclotron (EC) waves drive highly localized currents, and are therefore particularly suited for control purposes, but their CD efficiency is much lower than that of LH waves (typically, an order of magnitude in present day experiments). Various calculations have demonstrated an interesting property: the current driven by the simultaneous use of the two waves, I(LH+EC), can be significantly larger than the sum I(LH)+I(EC) of the currents separately driven by the two waves in the same plasma conditions. This property, called synergy effect. The peculiar experimental conditions attainable on the Tore Supra tokamak have allowed the first experimental demonstration of the synergy between EC and LH current drive. The significant improvement of the electron cyclotron current drive (ECCD) efficiency in the presence of low hybrid current drive (LHCD), predicted by kinetic theory and confirmed by stationary experiments on Tore Supra, opens up the possibility of using ECCD as an efficient current profile control tool in LHCD plasmas

  11. The Tore Supra Lower Hybrid Test Bed : improvements and applications

    International Nuclear Information System (INIS)

    Delpech, L.; Achard, J.; Beaumont, B.

    2006-01-01

    Within the CIMES project framework in Tore Supra, a klystron TH2103C (3.7 GHz) is under development at THALES ELECTRON DEVICES. It differs from the previous klystrons used in Tore Supra generator mainly in that it has no modulating anode, the RF output power will reach 700 kW CW, by raising the High Voltage value to 76 kV and a beam current up to 23 A. The Tore Supra test bed is a dedicated facility used for high power tests on RF components or on RF transmitters. It has been improved to integrate the TH2103C klystron and a specific 100 kV solide state switch which control the beam current. Since April 2005, the integration of the first tube (without modulating anode) and the 100 kV switch has been completed in the Test Bed and has allowed the modifications and tests of the interfaces and security system for the devices. Improvements were also made on the cooling loop flow to dissipate a power of 1750 kW CW. With these devices, the RF power routinely available in the Lower Hybrid Test Bed is 400 kW CW. With the development of the TH2103C, detailed studies and tests on RF components which will be used up to 750 kW CW on match load or 700 kW on VSWR = 1.4, are necessary to evaluate their performances and thermal behaviour. The test a crucial component, the recombiner, which adds the RF powers coming from the two RF outputs of the TH2103C and inject the resulted power into one WR284 waveguide to a test load or to the plasma, was completed. Two tests have been performed : a thermal study with 400 kW during 1000 s, and RF pulsed tests on short cuts to increase the value of the electric field inside the component. The experiments and calculations (ANSYS and HFSS codes) validate the use of this device with the TH2103C. A module made with two different Beryllium Oxide RF windows, has been under test. The losses on each window are measured by calorimetric measurements and evaluated by computation with HFSS and ANSYS code. The results are compared. In this paper, the

  12. First results from the Tore-Supra prototype injector

    International Nuclear Information System (INIS)

    Fumelli, M.; Bottiglioni, F.; Jequier, F.; Pamela, J.

    1987-01-01

    The first results of the Tore-Supra neutral injector prototype line are presented. The beam was extracted from a 6.6 x 113 cm 2 surface, the plasma source being grounded. The residual fast ion component was electrostatically deflected out of the beam at the exit of the neutralizer. In the first phase of the experiment we operated the injector up to 70 kV and 17 A of extracted deuterium beam. The measured neutral beam power fraction transmitted at 6.4 meter on a 24 x 40 cm 2 target was about 45% of the total extracted power, for a beam divergence of 0.7 0 .te

  13. First results from the Tore-Supra prototype injector

    International Nuclear Information System (INIS)

    Fumelli, M.; Bottiglioni, F.; Jequier, F.; Pamela, J.

    1986-01-01

    The first results of the Tore-Supra neutral injector prototype line are presented. The beam extracted from a 6.6 x 113 cm 2 surface, the plasma source being grounded. The residual fast ion component was electrostatically deflected out of the beam at the exit of the neutralizer. In the first phase of experiment we operated the injector up to 70 kV and 17 A of extracted deuterium beam. The measured neutral beam power fraction transmitted at 6.4 meter on a 24 x 40 cm 2 target was about 45% of the total extracted power, for a beam divergence of 0.7 0 . (author)

  14. Cybele: a large size ion source of module construction for Tore-Supra injector

    International Nuclear Information System (INIS)

    Simonin, A.; Garibaldi, P.

    2005-01-01

    A 70 keV 40 A hydrogen beam injector has been developed at Cadarache for plasma diagnostic purpose (MSE diagnostic and Charge exchange) on the Tore-Supra Tokamak. This injector daily operates with a large size ions source (called Pagoda) which does not completely fulfill all the requirements necessary for the present experiment. As a consequence, the development of a new ion source (called Cybele) has been underway whose objective is to meet high proton rate (>80%), current density of 160 mA/cm 2 within 5% of uniformity on the whole extraction surface for long shot operation (from 1 to 100 s). Moreover, the main particularity of Cybele is the module construction concept: it is composed of five source modules vertically juxtaposed, with a special orientation which fits the curved extraction surface of the injector; this curvature ensures a geometrical focalization of the neutral beam 7 m downstream in the Tore-Supra chamber. Cybele will be tested first in positive ion production for the Tore-Supra injector, and afterward in negative ion production mode; its modular concept could be advantageous to ensure plasma uniformity on the large extraction surface (about 1 m 2 ) of the ITER neutral beam injector. A module prototype (called the Drift Source) has already been developed in the past and optimized in the laboratory both for positive and negative ion production, where it has met the ITER ion source requirements in terms of D-current density (200 A/m 2 ), source pressure (0.3 Pa), uniformity and arc efficiency (0.015 A D-/kW). (authors)

  15. Calibration procedures of the Tore-Supra infrared endoscopes

    Science.gov (United States)

    Desgranges, C.; Jouve, M.; Balorin, C.; Reichle, R.; Firdaouss, M.; Lipa, M.; Chantant, M.; Gardarein, J. L.; Saille, A.; Loarer, T.

    2018-01-01

    Five endoscopes equipped with infrared cameras working in the medium infrared range (3-5 μm) are installed on the controlled thermonuclear fusion research device Tore-Supra. These endoscopes aim at monitoring the plasma facing components surface temperature to prevent their overheating. Signals delivered by infrared cameras through endoscopes are analysed and used on the one hand through a real time feedback control loop acting on the heating systems of the plasma to decrease plasma facing components surface temperatures when necessary, on the other hand for physics studies such as determination of the incoming heat flux . To ensure these two roles a very accurate knowledge of the absolute surface temperatures is mandatory. Consequently the infrared endoscopes must be calibrated through a very careful procedure. This means determining their transmission coefficients which is a delicate operation. Methods to calibrate infrared endoscopes during the shutdown period of the Tore-Supra machine will be presented. As they do not allow determining the possible transmittances evolution during operation an in-situ method is presented. It permits the validation of the calibration performed in laboratory as well as the monitoring of their evolution during machine operation. This is possible by the use of the endoscope shutter and a dedicated plasma scenario developed to heat it. Possible improvements of this method are briefly evoked.

  16. Particle retention during long discharges in Tore Supra and JET

    International Nuclear Information System (INIS)

    Loarer, T.; Tsitrone, E.; Brosset, C.; Bucalossi, J.; Gunn, J.; Joffrin, E.; Monier-Garbet, P.; Pegourie, B.; Thomas, P.; Lomas, P.; Ongena, J.

    2003-01-01

    The particle balances and the associated particle retentions for the long discharge experiments performed in Tore-Supra and for the L and H mode discharges carried out in JET are reported in this paper. From the reported experiments, the same particle retention behaviors are observed in Tore-Supra and JET in spite of the differences between the plasma geometry and the confinement mode (respectively limiter L-mode and divertor H-mode). A particle retention up to 70-80% of Γ(puff) for the larger gas injection has been obtained in JET. The particle retention behavior observed with the gas puff appears to be strongly dominant in the particle retention process. Indeed, no influence has been noticed from the active pumping, the saturation of the recycling area (0.4 D/C), the precedent discharges history (in terms of total 'particles retained' in the vessel) and even from the disruptions (conditioning). Also, the outgassing between discharges becomes negligible in terms of particle recovering when Γ(puff) and/or the discharge duration are increased. Finally, neither the edge localized modes (ELMs type-I or III) nor the disruptions modify the reported behaviour. For ITER, the particle retention is strictly limited and from the presented results it seems that strong gas injection should be avoided. (A.C.)

  17. Study on the materials for mirrors and back mirror reflectors of thermonuclear reactors and their testing in Tore-Supra

    International Nuclear Information System (INIS)

    Schunke, B.; Voytsenya, V.; Gil, C.; Lipa, M.

    2003-01-01

    Plasma diagnostics using visible or ultra-violet or infra-red radiations require mirrors to probe the plasma. These mirrors have to sustain very hostile environment and despite that must maintain good optical properties. Mirror samples made of 3 different metals: copper, stainless steel and molybdenum have been designed and installed in Tore Supra tokamak and will be exposed to plasmas till mid 2004. This project will allow fusion engineers to assess the impact of plasma ion bombardment on mirror reflectivity. Optical properties and parameters concerning the surface state of the samples have been measured before the installation in Tore Supra and are presented in the paper. Simulations with a Monte-Carlo code predict the particle flux and spectra near the samples. A specific back mirror reflector has been designed to probe mirror reflectivity changes. (A.C.)

  18. High heat flux actively cooled plasma facing components development, realization and first results in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.

    2004-01-01

    The development, design, manufacture and testing of actively cooled high heat flux plasma facing components (PFC) has been an essential stage towards long powerful tokamak operations for Tore-Supra, it lasted about 10 years. This paper deals with the toroidal pumped limiter (TPL) that is able to sustain up to 10 MW/m 2 of nominal heat flux. This device is based on hardened copper alloy heat sink structures covered by a carbon fiber composite armour, it resulted in the manufacturing of 600 elementary components, called finger elements, to achieve the 7.6 m 2 TPL. This assembly has been operating in Tore-Supra since spring 2002. Some difficulties occurred during the manufacturing phase, the valuable industrial experience is summarized in the section 2. The permanent monitoring of PFC surface temperature all along the discharge is performed by a set of 6 actively cooled infrared endoscopes. The heat flux monitoring and control issue but also the progress made in our understanding of the deuterium retention in long discharges are described in the section 3. (A.C.)

  19. High heat flux actively cooled plasma facing components development, realization and first results in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    2004-07-01

    The development, design, manufacture and testing of actively cooled high heat flux plasma facing components (PFC) has been an essential stage towards long powerful tokamak operations for Tore-Supra, it lasted about 10 years. This paper deals with the toroidal pumped limiter (TPL) that is able to sustain up to 10 MW/m{sup 2} of nominal heat flux. This device is based on hardened copper alloy heat sink structures covered by a carbon fiber composite armour, it resulted in the manufacturing of 600 elementary components, called finger elements, to achieve the 7.6 m{sup 2} TPL. This assembly has been operating in Tore-Supra since spring 2002. Some difficulties occurred during the manufacturing phase, the valuable industrial experience is summarized in the section 2. The permanent monitoring of PFC surface temperature all along the discharge is performed by a set of 6 actively cooled infrared endoscopes. The heat flux monitoring and control issue but also the progress made in our understanding of the deuterium retention in long discharges are described in the section 3. (A.C.)

  20. Test model of the fast thyristor circuit breaker, for TORE SUPRA

    International Nuclear Information System (INIS)

    Bareyt, B.; Leloup, C.; Rijnoudt, E.

    1984-01-01

    The tokamak TORE SUPRA, permits, owing to the toroidal superconducting coils and to the poloidal field system performances, long discharges (30 s and more), for a plasma current of typically 2 MA. The poloidal field system uses the magnetic energy initially stored, for the ignition and the fast rise of the plasma current, by forcing the primary current to flow through a resistor after breaking the main rectifier current by a fast thyristor circuit breaker. In order to test the technical capabilities of such a breaker system made of fast thyristors, in series and in parallel, after a single thyristor test model T1 the series arrangement was studied on a 24 thyristor test model T2 and the parallel arrangement problems, led the manufacturer CGEE Alsthom, to build a new test model T3. (author)

  1. Runaway electron damage to the Tore Supra Phase III outboard pump limiter

    International Nuclear Information System (INIS)

    Nygren, R.; Lutz, T.; Walsh, D.; Martin, G.; Chatelier, M.; Loarer, T.; Guilhem, D.

    1996-01-01

    Operation of the Phase III outboard pump limiter (OPL) in Tore Supra in 1994 was terminated prematurely when runaway electrons during the current decay following a disruption pierced leading edge tube on the electron side and caused a water leak. The location, about 20 mm outside the last closed flux surface during normal operation, and the infrared (IR) images of the limiter indicate that the runaways moved in large outward steps, i.e. tens of millimeters, in one toroidal revolution. For plasma (runaway) currents in the range of 155 to 250 kA, the drift orbits open to the outside. Basic trajectory computations suggest that such motion is possible under the conditions present for this experiment. Activation measurements made on sections of the tube to indicate the area of local damage are presented here. An understanding of this event may provide important guidance regarding the potential damage from runaways in future tokamaks

  2. Thermal and non-thermal particle interaction with the LHCD launchers in Tore Supra

    International Nuclear Information System (INIS)

    Ekedahl, A.; Goniche, M.; Balorin, C.; Basiuk, V.; Bibet, Ph.; Chantant, M.; Colas, L.; Delpech, L.; Desgranges, C.; Eriksson, L.-G.; Joffrin, E.; Kazarian, F.; Lowry, C.; Moreau, Ph.; Petrzilka, V.; Portafaix, C.; Prou, M.; Roche, H.

    2007-01-01

    The interaction between the lower hybrid current drive (LHCD) launchers and the plasma has been studied during long pulse, high power operation in the Tore Supra tokamak. The main diagnostics used for characterising the plasma-launcher interaction are calorimetry of the energy extracted by the launchers and infrared (IR) imaging of the launchers and their side limiters. The calorimetry has allowed to identify three different heat sources on the LHCD launchers, namely the RF losses in the waveguides, a fraction (∼0.8%) of the total injected energy and, finally, fast ion losses during ion cyclotron resonance heating (ICRH), accounting for ∼1% of the injected ICRH energy. The interaction by fast ions is identified by infrared imaging of the LHCD launchers as a localised hotspot on the ion drift side, below or at the mid-plane

  3. Replacement of the instrumentation and control system of Tore Supra

    International Nuclear Information System (INIS)

    Leveque, P.

    1995-02-01

    The control system of the Tore-Supra is a wide and complex system that cannot be interrupted while running without significant consequences on the operating of the machine. Replacing the current system cannot be achieved in a global way without immobilisation and high costs. Therefore partial changes have been decided on. This work presents the detailed analysis of the arrangements and the operating of the system that will be replaced: the pro's and con's that have appeared through experience are related. The possibilities that the new apparatus offers are also examined. A method of step by step replacements had to be set up in order to assess the means, funds, term of achievement, performance and quality of the overall project. (TEC). 15 refs., 29 figs

  4. ITER-like PAM launcher for tore supra's LHCD system

    International Nuclear Information System (INIS)

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

    2004-01-01

    A new launcher based upon the PAM concept (Passive-Active Multijunction) already proposed for ITER has been developed and is currently under realisation at Tore Supra. It was designed for an injection power capability of 2.7 MW CW at 3.7 GHz, a power density of 25 MW/m 2 , to radiate 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 on a wide range of electron densities. In this paper an overall description of the antenna as well as the foreseen manufacturing and assembling processes are given, followed by the results from studies and optimizations of its RF components, and by a stress analysis of its thermomechanical behaviour. (authors)

  5. Ergodic divertor impact on Tore Supra plasma edge

    International Nuclear Information System (INIS)

    Grosman, A.; Ghendrih, P.; Agostini, E.; Bruneau, J.L.; Michelis, C. De; Fall, T.; Gil, C.; Guilhem, D.; Hess, W.; Hutter, T.

    1990-01-01

    Present ergodic divertor experiments in TORE SUPRA have been devoted to benchmarking the operational regimes of the apparatus. Two major effects are reported; on one hand, strong changes occur in the ergodized boundary layer (up to 20% of the minor radius), and on the other hand, the central plasma and especially the confinement is not directly affected, i.e. the observed modifications are induced by edge effects. The basic trends, which are recorded are a decrease of both the edge electronic temperature and the edge density gradient while the radiated power is increased at the very edge of the ergodic region. The latter feature is in agreement with the impurity line emission characterized by an increase of the peripheral lines with a strong decrease of the central lines

  6. An advanced plasma control system for Tore Supra

    International Nuclear Information System (INIS)

    Wijnands, T.; Martin, G.

    1996-01-01

    First results on plasma control with the new plasma control system of Tore Supra are presented. The system has been especially designed for long pulse operation: plasmas are controlled on reference signals, which can be varied in real time by using diagnostic measurements. On line determination of the global plasma equilibrium has enabled new operation scenarios in which both the power from the poloidal field generators and the total Lower Hybrid (LH) power are used to control the plasma. Experiments with feedback control of the safety factor on the plasma boundary, control of the LH driven current, control of the flux on the plasma boundary and control of the internal inductance are discussed. (author)

  7. Current density profile inside q=1 on Tore Supra

    International Nuclear Information System (INIS)

    Joffrin, E.; Desgranges, C.; Sabot, R.; Dubois, M.A.

    1995-01-01

    The Tore Supra polarimeter used to measure the poloidal field distribution is described. The current density profiles are computed in two different ways using the interferometric and polarimetric data in conjunction with the magnetic data and the location of the inversion radius determined by the soft X-ray camera. The current density inside the q=1 surface is investigated for normal and monster sawteeth. Its variation are also measured by the polarimeter and compared with that predicted by the current diffusion equation assuming complete reconnection. Finally, the safety factor profile is compared with that obtained with the striation data of the pellet ablation. The results of the evolution of the q profile during sawteeth are in good agreement with those obtained in other devices. (author) 9 refs.; 4 figs

  8. Deuterium inventory in Tore Supra: reconciling particle balance and post-mortem analysis

    Science.gov (United States)

    Tsitrone, E.; Brosset, C.; Pégourié, B.; Gauthier, E.; Bouvet, J.; Bucalossi, J.; Carpentier, S.; Corre, Y.; Delchambre, E.; Desgranges, L.; Dittmar, T.; Douai, D.; Ekedahl, A.; Escarguel, A.; Ghendrih, Ph.; Grisolia, C.; Grosman, A.; Gunn, J.; Hong, S. H.; Jacob, W.; Kazarian, F.; Kocan, M.; Khodja, H.; Linez, F.; Loarer, T.; Marandet, Y.; Martinez, A.; Mayer, M.; Meyer, O.; Monier Garbet, P.; Moreau, P.; Pascal, J. Y.; Pasquet, B.; Rimini, F.; Roche, H.; Roure, I.; Rosanvallon, S.; Roubin, P.; Roth, J.; Saint-Laurent, F.; Samaille, F.; Vartanian, S.

    2009-07-01

    Fuel retention, a crucial issue for next step devices, is assessed in present-day tokamaks using two methods: particle balance performed during shots and post-mortem analysis carried out during shutdowns between experimental campaigns. Post-mortem analysis generally gives lower estimates of fuel retention than integrated particle balance. In order to understand the discrepancy between these two methods, a dedicated experimental campaign has been performed in Tore Supra to load the vessel walls with deuterium (D) and monitor the trapped D inventory through particle balance. The campaign was followed by an extensive post-mortem analysis phase of the Tore Supra limiter. This paper presents the status of the analysis phase, including the assessment of the D content in the castellated tile structure of the limiter. Indeed, using combined surface analysis techniques, it was possible to derive the relative contributions of different zones of interest on the limiter (erosion, thick deposits, thin deposits), showing that the post-mortem inventory is mainly due to codeposition (90% of the total), in particular due to gap deposits. However, deuterium was also evidenced deep into the material in erosion zones (10% of the total). At the present stage of the analysis, 50% of the inventory deduced from particle balance has been found through post-mortem analysis, a significant progress with respect to previous studies (factor 8-10 discrepancy). This shows that post-mortem analysis can be consistent with particle balance provided specific procedures are implemented (dedicated campaign followed by extensive post-mortem analysis). Both techniques are needed for a reliable assessment of fuel retention in tokamaks, giving complementary information on how much and where fuel is retained in the vessel walls.

  9. Deuterium inventory in Tore Supra: reconciling particle balance and post-mortem analysis

    International Nuclear Information System (INIS)

    Tsitrone, E.; Brosset, C.; Pegourie, B.; Gauthier, E.; Bouvet, J.; Bucalossi, J.; Carpentier, S.; Corre, Y.; Delchambre, E.; Dittmar, T.; Douai, D.; Ekedahl, A.; Ghendrih, Ph.; Grisolia, C.; Grosman, A.; Gunn, J.; Hong, S.H.; Desgranges, L.; Escarguel, A.; Jacob, W.

    2009-01-01

    Fuel retention, a crucial issue for next step devices, is assessed in present-day tokamaks using two methods: particle balance performed during shots and post-mortem analysis carried out during shutdowns between experimental campaigns. Post-mortem analysis generally gives lower estimates of fuel retention than integrated particle balance. In order to understand the discrepancy between these two methods, a dedicated experimental campaign has been performed in Tore Supra to load the vessel walls with deuterium (D) and monitor the trapped D inventory through particle balance. The campaign was followed by an extensive post-mortem analysis phase of the Tore Supra limiter. This paper presents the status of the analysis phase, including the assessment of the D content in the castellated tile structure of the limiter. Indeed, using combined surface analysis techniques, it was possible to derive the relative contributions of different zones of interest on the limiter (erosion, thick deposits, thin deposits), showing that the post-mortem inventory is mainly due to codeposition (90% of the total), in particular due to gap deposits. However, deuterium was also evidenced deep into the material in erosion zones (10% of the total). At the present stage of the analysis, 50% of the inventory deduced from particle balance has been found through post-mortem analysis, a significant progress with respect to previous studies (factor 8-10 discrepancy). This shows that post-mortem analysis can be consistent with particle balance provided specific procedures are implemented (dedicated campaign followed by extensive post-mortem analysis). Both techniques are needed for a reliable assessment of fuel retention in tokamaks, giving complementary information on how much and where fuel is retained in the vessel walls.

  10. Edge plasma control: Particle channeling in Tore Supra pump limiter and ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, P.; Samain, A.; Grosman, A.; Capes, H.; Morera, J.P.

    1989-01-01

    Improved pumping efficiency can be achieved on Tore Supra by channeling process for particles, i.e. channeling of neutrals in the throat of pump limiters and channeling of plasma towards neutralizer plates in the ergodic divertor. The plugging length for the pump limiter throat is computed and numerical evidence of plasma flux channeling between the conductor bars of the ergodic divertor is presented. The effect of the Tore Supra ergodic divertor on edge plasma state and edge plasma transport is discussed. (orig.)

  11. Replacement of the instrumentation and control system of Tore Supra; Remplacement du systeme de controle commande de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Leveque, P.

    1995-02-01

    The control system of the Tore-Supra is a wide and complex system that cannot be interrupted while running without significant consequences on the operating of the machine. Replacing the current system cannot be achieved in a global way without immobilisation and high costs. Therefore partial changes have been decided on. This work presents the detailed analysis of the arrangements and the operating of the system that will be replaced: the pro`s and con`s that have appeared through experience are related. The possibilities that the new apparatus offers are also examined. A method of step by step replacements had to be set up in order to assess the means, funds, term of achievement, performance and quality of the overall project. (TEC). 15 refs., 29 figs.

  12. Comparison of two regularization methods for soft x-ray tomography at Tore Supra

    International Nuclear Information System (INIS)

    Jardin, A; Mazon, D; Bielecki, J

    2016-01-01

    Soft x-ray (SXR) emission in the range 0.1–20 keV is widely used to obtain valuable information on tokamak plasma physics, such as particle transport, magnetic configuration or magnetohydrodynamic activity. In particular, 2D tomography is the usual plasma diagnostic to access the local SXR emissivity. The tomographic inversion is traditionally performed from line-integrated measurements of two or more cameras viewing the plasma in a poloidal cross-section, like at Tore Supra (TS). Unfortunately, due to the limited number of measured projections and presence of noise, the tomographic reconstruction of SXR emissivity is a mathematical ill-posed problem. Thus, obtaining reliable results of the tomographic inversion is a very challenging task. In order to perform the reconstruction, inversion algorithms implemented in present tokamaks use a priori information as additional constraints imposed on the plasma SXR emissivity. Among several potential inversion methods, some of them have been identified as well suited to tokamak plasmas. The purpose of this work is to compare two promising inversion methods, i.e. the minimum fisher information method already used at TS and planned for WEST configuration, and the alternative 2nd order Phillips–Tikhonov regularization with smoothness constraints imposed on the second derivative norm. Respective accuracy of both reconstruction methods as well as overall robustness and computational time are studied, using several synthetic SXR emissivity profiles. Finally, a real case is studied through tomographic reconstruction from TS SXR database. (paper)

  13. Tore supra: towards the 'long time' fusion. Press journey; Tore Supra: vers la fusion 'longue duree'. Voyage de presse

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    After a recall of the interest in the fusion for the development of energy sources, the document presents the fusion from the solar reaction to the nuclear fusion in laboratory. Then it discusses the great challenges of this technology and the Tore Supra installation. The last part is devoted to ITER and DEMO projects. (A.L.B.)

  14. Snake-like phenomena in Tore Supra following pellet injection

    International Nuclear Information System (INIS)

    Pecquet, A.L.; Cristofani, P.; Mattioli, M.; Garbet, X.; Laurent, L.; Geraud, A.; Gil, C.; Joffrin, E.; Sabot, R.

    1996-01-01

    Snakes are observed in Tore-Supra, after injection of high velocity solid hydrogen or deuterium pellets ablated inside the q=1 surface. They are detected, immediately after the ablation, as oscillations on the line integrated densities of the central interferometer channels. The corresponding oscillations on the soft X-ray signals detach from the noise about 70 ms later. Snakes survive sawtooth crashes, but are nevertheless affected by them. Variations, during the about 500 ms long lifetime, of the snake radius τ s , of the rotation frequency and of the rotation direction are discussed, stressing the effects of the sawtooth crashes. In many snakes τ s /τ q =1 is of the order of 0.5. Since the snake has a m=1, n=1 helicity, this points out the existence of a flat or inverted safety factor profile, confirmed by calculation of the current profile using Spitzer's resistivity. Combined simulations of the snake oscillations on both interferometer and soft X-ray signals have indicated that, starting about 80 ms after the snake formation, the impurity (carbon) density inside the snake is much larger than outside it. Since a change of regime seems to appear about 80 ms after the snake formation on the soft X-ray, it seems plausible that impurity (carbon) accumulation takes place at this time. A stability criterion taking into account both impurity and bootstrap effects is presented, the result agrees with the model proposed by Wesson. (authors)

  15. Turbulence and energy confinement in TORE SUPRA ohmic discharges

    International Nuclear Information System (INIS)

    Garbet, X.; Payan, J.; Laviron, C.; Devynck, P.; Saha, S.K.; Capes, H.; Chen, X.P.; Coulon, J.P.; Gil, C.; Harris, G.; Hutter, T.; Pecquet, A.L.

    1992-06-01

    Results on confinement and turbulence from a set of ohmic discharges in Tore Supra are discussed. The attention is focused on the saturation of the energy confinement time and it is emphasized that this saturation could be explained by a saturation of the electron heat diffusivity. Ion behaviour is indeed governed by dilution and equipartition effects. Although the ion heat transport is never neoclassical, there is no enhanced degradation at the saturation. This behaviour is confirmed by turbulence measurements given by CO 2 laser coherent scattering. The density fluctuations level follows the electron heat diffusivity variations with the average density. Waves propagating in the ion diamagnetic direction are always present in turbulence frequency spectra. Thus, the saturation cannot be explained by the onset of an ion turbulence. The existence of an ion turbulence at the edge at all densities cannot be excluded. However, this ion feature in scattering spectra could be explained by a Doppler shift associated to an inversion point of the radial electric field at the edge

  16. Structural design of the Tore Supra Phase III limiter head

    International Nuclear Information System (INIS)

    Dempsey, J.F.; Koski, J.A.; Watson, R.D.

    1992-01-01

    The Sandia National Laboratories Fusion Technology Division has designed and fabricated an actively cooled pumped limiter for use in Tore Supra. The limiter head is composed of an integrated high tolerance contoured surface of pyrolytic graphite (PG) tiles brazed to 14 independent OFHC (oxygen free high conductivity) copper tube segments. The limiter is designed to absorb up to 2 MW of incident power and peak heat fluxes up to 30 MW/m 2 under nominal plasma operating conditions. The limiter will also receive vibrational loads induced by plasma disruptions. This paper discusses innovative support techniques that are used to prevent misalignment of the tubes in the limiter head during plasma operations. Thermal misalignment of the limiter with respect to the toroidal field will damage the limiter, therefore, all tubes in the tube bank must be maintained in good alignment with each other. In order to minimize limiter damage due to misalignment and dynamic chattering loads without increasing eddy current loads, a design which weaves the limiter head together with flexible strands of inconel wire evolved

  17. Multichannel and multicolor infrared thermography in Tore Supra

    International Nuclear Information System (INIS)

    Reichle, R.; Pocheau, C.; Balorin, C.; Delchambre, E.; Desgrange, C.; Guilhem, D.; Messina, P.; Roche, H.

    2004-01-01

    An imaging spectrometer using a sapphire prism as dispersing element has been conceived at Tore Supra for the spectral range of 1 - 4 μm. It measures simultaneously at various wavelengths the temperature on distributed high heat-flux elements under plasma impact with 36 optical fibres, 4 of which are ZrF 4 fibres. It employs an InSb focal plane array detector (256*320 pixels) behind a silicon filter and a ZnS window yielding a dynamic range of 200 to 1500 deg C with 20 ms temporal resolution. The fibre transmission and the spatial variation of gain and background of the camera are calibrated using a light source with integrating sphere. With a black body source one determines the non linearity of the average gain and controls its stability during operation. The spectral dispersion of about 30 nm/pixel is determined with interference filters and controlled with a spectral lamp. The measurements at various wavelengths allow to determine the temperature distribution in the held of view. (author)

  18. Power deposition to the facing components in Tore-Supra

    International Nuclear Information System (INIS)

    Guilhem, D.; Chatelier, M.; Chappuis, P.; Koski, J.; Watkins, J.

    1990-01-01

    The modifications of the power scrape-off length, λ q and power deposition are studied for various configurations in ohmic Tore-Supra plasmas. The plasma is either touching the horizontal limiter alone, the full set of six pump limiters or the inner bumper limiter. All configurations are with and without the ergodic divertor system energized. From a comparison of the infrared images of the limiter we derived that the λ q for power deposition was slightly less than 9±1 mm in ohmic plasmas which is in agreement with the predicted design value of 10 mm. Using the six limiters, instead of one, does not modify λ q significantly, but leads to small asymmetries. The power is shared by all the limiters and the maximum surface temperature on the horizontal limiter decreases. These λ q values have been independently determined by calorimetric measurements on the integrated energy deposition on the horizontal limiter and other internal structures 5 cm into the scrape-off layer. These values agree with the infrared measurements in the two cases. In the presence of the ergodic divertor we observe a broadening of the scrape-off layer, the e-folding length for power deposition reaching 2.5 cm. Large asymmetries in the power deposition can be seen on the front face of the limiter, leading to the formation of hot spots at the leading edges. (orig.)

  19. Long Discharge Particle Balance and Fuel Retention in Tore Supra

    Science.gov (United States)

    Pégourié, B.; Brosset, C.; Delchambre, E.; Loarer, T.; Roubin, P.; Tsitrone, E.; Bucalossi, J.; Gunn, J.; Khodja, H.; Lafon, C.; Martin, C.; Parent, P.; Reichle, R.

    In the new CIEL configuration of Tore Supra, all the plasma facing components are actively cooled. The surface area of the wall that is covered with carbon measures about 15 m2 (2 to 4 m2 of which in close interaction with the plasma). Steady-state plasma conditions up to 4 min 25 s have been maintained in this configuration. In these experiments, the required gas injection to maintain the prescribed density remains constant during the whole discharge. The exhausted flux is also constant and equal to 40 A~· 50% of the injected flux. Therefore, 50 A~· 60% of the injected particles remain trapped in the vessel, the total retention being proportional to the plasma duration. Since the amount of gas recovered between shots or by He-glow discharges does not always balance the injected gas, it follows that a quantity of deuterium remains indefinitely trapped in the vessel, which appears as an infinite reservoir. This reservoir is believed to be dominated by co-deposited layers, as observed in several places of the vessel. The thickest deposits (up to 800 Î 1/4 m) are observed on the leading edge of the neutralizers of the pump limiter. They display a column-like shape (typical growth rate âe 1/4 20 nm/s) and have a graphite-like structure. Their deuterium concentration is D/C âe 1/4 1%. Conversely, in regions that are shadowed from the direct plasma flux, the deposits show a smoother shape and their deuterium content is typically âe 1/4 10 A~· 15%.

  20. Validation of the LH antenna code ALOHA against Tore Supra experiments

    International Nuclear Information System (INIS)

    Hillairet, J.; Ekedahl, A.; Kocan, M.; Gunn, J. P.; Goniche, M.

    2009-01-01

    Comparisons between ALOHA code predictions and experimental measurements of reflection coefficients for the two different Lower Hybrid Current Drive (LHCD) antennas (named C2 and C3) in Tore Supra are presented. A large variation of density in front of the antennas was obtained by varying the distance between the plasma and the antennas. Low power ( 2 ) was used in order to avoid non-linear effects on the wave coupling. Results obtained with ALOHA are in good agreement with the experimental measurements for both Tore Supra antennas and show that ALOHA is an efficient LH predictive tool.

  1. Spatially resolved charge exchange flux calculations on the Toroidal Pumped Limiter of Tore Supra

    International Nuclear Information System (INIS)

    Marandet, Y.; Tsitrone, E.; Boerner, P.; Reiter, D.; Beaute, A.; Delchambre, E.; Escarguel, A.; Brezinsek, S.; Genesio, P.; Gunn, J.; Monier-Garbet, P.; Mitteau, R.; Pegourie, B.

    2009-01-01

    A spatially resolved calculation of the charge exchange particle and energy fluxes on the Toroidal Pumped Limiter (TPL) of Tore Supra is presented, as a first step towards a better understanding and modelling of carbon erosion, migration, as well as deuterium codeposition and bulk diffusion of deuterium in Tore Supra. The results are obtained with the EIRENE code run in a 3D geometry. Physical and chemical erosion maps on the TPL are calculated, and the contribution of neutrals to erosion, especially in the self-shadowed area, is calculated.

  2. Identification and real time control of current profile in Tore-supra: algorithms and simulation; Identification et controle en temps reel du profil de courant dans Tore Supra: algorithmes et simulations

    Energy Technology Data Exchange (ETDEWEB)

    Houy, P

    1999-10-15

    The aim of this work is to propose a real-time control of the current profile in order to achieve reproducible operating modes with improved energetic confinement in tokamaks. The determination of the profile is based on measurements given by interferometry and polarimetry diagnostics. Different ways to evaluate and improve the accuracy of these measurements are exposed. The position and the shape of a plasma are controlled by the poloidal system that forces them to cope with standard values. Gas or neutral ions or ice pellet or extra power injection are technical means used to control other plasma parameters. These controls are performed by servo-controlled loops. The poloidal system of Tore-supra is presented. The main obstacle to a reliable determination of the current profile is the fact that slightly different Faraday angles lead to very different profiles. The direct identification method that is exposed in this work, gives the profile that minimizes the square of the margin between measured and computed values. The different algorithms proposed to control current profiles on Tore-supra have been validated by using a plasma simulation. The code Cronos that solves the resistive diffusion equation of current has been used. (A.C.)

  3. Spectral broadening measurement of the lower hybrid waves during long pulse operation in Tore Supra

    Science.gov (United States)

    Berger-By, G.; Decampy, J.; Antar, G. Y.; Goniche, M.; Ekedahl, A.; Delpech, L.; Leroux, F.; Tore Supra Team

    2014-02-01

    On many tokamaks (C-Mod, EAST, FTU, JET, HT-7, TS), a decrease in current drive efficiency of the Lower Hybrid (LH) waves is observed in high electron density plasmas. The cause of this behaviour is believed to be: Parametric Instabilities (PI) and Scattering from Density Fluctuations (SDF). For the ITER LH system, our knowledge must be improved to avoid such effects and to maintain the LH current drive efficiency at high density. The ITPA IOS group coordinates this effort [1] and all experimental data are essential to validate the numerical codes in progress. Usually the broadening of the LH wave frequency spectrum is measured by a probe located in the plasma edge. For this study, the frequency spectrum of a reflected power signal from the LH antenna was used. In addition, the spectrum measurements are compared with the density fluctuations observed on RF probes located at the antenna mouth. Several plasma currents (0.6 to 1.4 MA) and densities up to 5.2 × 1019 m-3 have been realised on Tore Supra (TS) long pulses and with high injected RF power, up to 5.4 MW-30s. This allowed using a spectrum analyser to make several measurements during the plasma pulse. The side lobe amplitude, shifted by 20-30MHz with respect to the main peak, grows with increasing density. Furthermore, for an increase of plasma current at the same density, the spectra broaden and become asymmetric. Some parametric dependencies are shown in this paper.

  4. Actively cooled pump limiters and power scrape-off length measurements in Tore-Supra

    International Nuclear Information System (INIS)

    Guilhem, D.; Seigneur, A.; Chappuis, P.; Chatelier, M.; DeMichelis, C.; Deschamps, P.; Grosman, A.; Hess, W.; Lecoustey, P.; Loarer, T.; Poutchy, L.; Schlosser, J.

    1992-01-01

    TORE-SUPRA is a superconducting Tokamak aimed at studying long plasma pulses (>30 s). It is equipped with two types of pump limiters (PL). A provisional type, semi-inertially cooled between shots, has been used for plasma scrape off characterization. The e-folding length λq for power deposition on these components has been unfolded (1.0cm 19 m -3 19 m -3 ), of power level up to 4 MW and of toroidal magnetic field (1.5 T -1/2 ). The second type used for long pulse operation, is actively cooled during shots, its thermal time constant being less than 2 seconds. Experiments using this ITER relevant technology are presented. Three of the actively cooled limiters have been successfully tested in a steady state regime with a surface temperature less than 1000 deg C (I p =1.6 MA). The design value for power removal on this type of limiters has been obtained. Peak power fluxes of 10 MW/m 2 have been estimated. This represents a breakthrough for high heat flux components since critical heat flux and burnout with subcooled flow boiling are major aspects for this kind of design

  5. First lower hybrid current drive experiments at 3.7 GHz in Tore Supra

    International Nuclear Information System (INIS)

    Tonon, G.; Goniche, M.; Moreau, D.

    1989-01-01

    The results of electromagnetic waves injection in the Tore Supra plasma, at a frequency of 3.7 GHz, are reported. The process is applied for current generation and plasma heating, through Landau damping on the electron population. The experimental set-up is described. The lower hybrid current drive experiments in Tore Supra are carried out under the following conditions: major and minor radii of the plasma are respectively 2.37 m and 0.77 m and the toroidal magnetic field is 1.8 Teslas. A multijunction-grill composed of 128 waveguides is applied. Up to 1.25 MW of rf power is injected in Tore Supra, after less than 30 plasma shots. The results lead to the conclusion that the coupling, not yet optimized, is good enough for safe klystron operation with no circulator. The measured value RIp P RF -1 (δV L /V L ) obtained on Tore Supra (Bt = 1.8 T) is closed to one observed on PETULA-B (Bt = 2.75 T) at the same frequency and density

  6. Radio frequency additional heating systems issues for the TORE-SUPRA WEST project

    NARCIS (Netherlands)

    Guilhem, D.; Argouarch, A.; Bernard, J.M.; Bouquey, F.; Colas, L.; Delpech, L.; Durodié, F.; Ekedahl, A.; Helvoirt, J.; Hillairet, J.; Joffrin, E.; Litaudon, X.; Magne, R.; Milanesio, D.; Moerel, J.; Mollard, P.; Wittebol, E.; Achard, J.; Armitano, A.; Berger-By, G.; Charabot, N.; Goniche, M.; Jacquot, J.; Lombard, G.; Prou, M.; Traisnel-Corbel, E.; Volpe, R.; Vulliez, K.

    2013-01-01

    This year TORE-SUPRA celebrated its 25 years of operation. During this long time a number of technologies have been developed [1]. First of all it was mandatory to develop reliable superconducting magnets at ∼ - 4 K, with superfluid helium as efficient coolant. For the production of steady state

  7. Evolution of the Tore Supra Lower Hybrid Current Drive System for WEST

    Energy Technology Data Exchange (ETDEWEB)

    Delpech, Léna, E-mail: lena.delpech@cea.fr [CEA, IRFM, F-13108 St Paul-Lez-Durance (France); Achard, Joelle; Armitano, Arthur; Berger-By, Gilles; Ekedahl, Annika; Gargiulo, Laurent; Goniche, Marc; Guilhem, Dominique; Hertout, Patrick; Hillairet, Julien; Magne, Roland; Mollard, Patrick [CEA, IRFM, F-13108 St Paul-Lez-Durance (France); Piluso, P. [CNIM Industrial Systems, 83507 La Seyne-sur-Mer (France); Poli, Serge; Prou, Marc; Saille, Alain; Samaille, Franck [CEA, IRFM, F-13108 St Paul-Lez-Durance (France)

    2015-10-15

    Highlights: • Describe the state of the Lower Hybrid heating system for the WEST project. • Detailed the experiments to assess the coupling in WEST configuration. • Give the modifications required on the launchers to be adapted to WEST configuration. • Detailed the technical modifications with the CNIM company on the launchers. - Abstract: The WEST-project (W-tungsten Environment in Steady-state Tokamak) involves equipping Tore Supra with a full tungsten divertor, capable of withstanding heat load of 10 MW/m{sup 2} in steady-state conditions, in discharges sustained by Lower Hybrid Current Drive (LHCD). The LHCD generator, recently upgraded to deliver 9.2 MW/1000 s, is equipped with sixteen TH2103C klystrons powering two launchers. The WEST transformation involves reducing the plasma volume, thus moving the launchers ∼10 cm closer to the tokamak centre. The toroidal curvature of the launchers no longer fits the plasma curvature due to the strong magnetic field ripple effect, leading to a degradation of the LH wave coupling, especially with the Full Active Multijunction Launcher (FAM). The toroidal curvature radius of the FAM launcher mouth will therefore be reshaped from 1700 mm to 2300 mm. The machining process is described in this article. In order to improve the coupling of the LH wave, the local gas injection has been modified to help to meet the requirement of 7 MW/1000 s of LH power coupled to the plasma in the WEST scenarios. Finally, the curvature radius of the waveguide septa are rounded to minimize the excitation of suprathermal electrons near the plasma edge, which can induce high power loads on the plasma facing components.

  8. The 118-GHz electron cyclotron heating system on Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Darbos, C.; Magne, R.; Bouquey, F.; Lambert, R.; Lennholm, M.; Traisnel, E. [CEA, IRFM, F-13108 St Paul Les Durance, (France); Arnold, A.; Thumm, M. [Univ Karlsruhe, Inst Hochfrequenztech and Elect, Karlsruhe, (Germany); Prinz, H.O.; Thumm, M. [EURATOM, Forschungszentrum Karlsruhe, Inst Hochleistungsimpuls und Mikrowellentech, Karlsruhe, (Germany); Hogge, J.P. [Assoc Euratom Confederat Suisse, Lausanne, (Switzerland); Lievin, C. [Thales Electron Devices, Velizy Villacoublay, (France)

    2009-07-01

    An electron cyclotron resonance heating (ECRH) system capable of delivering 2.4 MW cw has been designed to be built at Commissariat a l'Energie Atomique. Cadarache, for the Tore Supra (TS) experiment, to provide plasma heating and current drive by electron cyclotron resonance interaction. The planned system was composed of a generator using six gyrotrons 500 kW for 5 s or 400 kW cw working at 118 GHz. Six transmission lines made of corrugated waveguide, 63.5-mm diameter, carry. the HE11 mode to one antenna composed of six fixed mirrors and three independently movable mirrors for the adjustment of the injection angles of the rf beams. The antenna was built and installed in TS., and all transmission line components ordered and installed between the gyrotron locations and the antenna. In the same way, the required six oil tanks, the six cryo-magnets, and the six modulating anode devices were designed and manufactured. In parallel, after demonstration in the factory of proper operation of the prototype gyrotron, the manufacture of a first so-called series gyrotron was made. Bill this gyrotron experienced hard limitations (overheating inducing prohibited outgassing, parasitic oscillations) during the long-pulse tests in Cadarache, and the achieved performance was 300 kW for 110 s. A new study was then carried out in collaboration with Thales Electron Devices, the EURATOM-CEA Association. and the EURATOM-Confederation Suisse Association to understand and overcome the limitations, which led to the construction of a nest, modified gyrotron. During the tests in factory of this new gyrotron, the output beam showed two peaks, a pattern never predicted by simulations. The gyrotron was nevertheless transferred to Cadarache for long-pulse testing, but all arc oil the windows definitely stopped the tests. To understand the cause of the observed two peaks, various low-level tests were then performed oil a model of the mode converter with different shapes for the launcher; but

  9. Experimental study of turbulence on Tore Supra by plasma micro-waves interaction; Etude experimentale de la turbulence sur Tore Supra par interaction plasma micro-ondes

    Energy Technology Data Exchange (ETDEWEB)

    Colas, L

    1996-09-23

    Internal small-scale magnetic turbulence is a serious candidate to explain the anomalous heat transport in tokamaks. This turbulence is badly known in the gradient region of large machines. In this work internal magnetic fluctuations are measured on Tore Supra with an original diagnostic : Cross Polarization Scattering (CPS). This experimental tool relies on the Eigenmode change of a probing polarised microwave beam scattered by magnetic fluctuations, close to a cut-off layer for the incident wave. In this work, the diagnostic is first qualified to assess its sensitivity to magnetic fluctuations, and the spatial localisation for its measurements. The magnetic fluctuation behaviour is then analysed over a wide range of plasma current, density and additional power, and interpreted with a simple 1-D scattering model. A scan of the plasma density or magnetic field is used to move the CPS measurement location from r/a = 0.3 to r/a = 0.75. A fluctuation radial profile is obtained by two means. In L-mode discharges, the relation between magnetic fluctuations, temperature profiles and local heat diffusivities is investigated. With all measurements, it is also possible to look for a local parameter correlated to the turbulence in a large domain of plasma conditions. The fluctuation-induced local heat diffusivity expected from the measured fluctuations is estimated using the non-collisional quasi-linear formula: X{sup mag}{sub e} = {pi}qRV{sub te}({delta}B / B){sup 2}. Both the absolute values and the parametric dependence of calculated X{sup mag}{sub e} are close to the electron thermal diffusivities Xe determined by transport analysis. In particular, a threshold is evidenced in the dependence of fluctuation-induced heat fluxes on local {nabla}T{sub e}, which is analogous to the critical gradient for measured heat fluxes. The experimental setup is also sensitive to the Thomson scattering of the probing wave by density fluctuations. Its measurements are analysed as the

  10. Conceptual design of a high heat flux toroidal pumped limiter for Tore Supra

    International Nuclear Information System (INIS)

    Doceul, L.; Schlosser, J.; Chappuis, Ph.; Chatelier, M.; Cocat, J.P.; Deck, C.; Faisse, F.; Grosman, A.; Mitteau, R.; Tonon, G.

    1994-01-01

    In the frame of the Tore-Supra upgrade, where it is planned to inject up to 25 MW during a time up to 1000 s, a complete toroidal pumped limiter covered of CFC (Carbon Fiber Composite) tiles is being designed. The design is based on the important experience gained from the operation on Tore Supra of actively cooled plasma facing components and pumped limiters. This toroidal limiter covers 7.5 m 2 of the bottom part of the inner vessel and is composed of 576 elementary components. Each element is built from dispersion strengthened copper (DSCu) protected by brazed CFC flat tiles and cooled by pressurised water at 150 deg C. This limiter is designed to sustain 15 MW of convective power. (author) 7 refs.; 5 figs., 3 tabs

  11. First results of the Tore Supra ITER like ICRF antenna prototype

    International Nuclear Information System (INIS)

    Vulliez, K.; Bosia, G.; Bremond, S.; Agarici, G.; Beaumont, B.; Lombard, G.; Millon, L.; Mollard, P.; Volpe, D.

    2005-01-01

    The project of an ITER-like (IL) ion cyclotron (IC) prototype antenna was initiated in mid 2002 in Cadarache, to assess the operational characteristics of the IL scheme in a Tore Supra (TS)-sized array. The prototype was developed by modifying an existing Tore Supra resonant double loop (RDL) antenna. This strategy, chosen to reduce cost and manufacturing time, allowed the completion of the array construction at the end of 2003. The array was installed on TS in January 2004, at the beginning of the new experimental campaign. After a few comments on the design, the paper reports results on low-power characterisation, high-power commissioning and preliminary tests on TS plasmas, prior to the array dis-assembly for inspection

  12. Carbon source from the toroidal pumped limiter during long discharge operation in Tore Supra

    International Nuclear Information System (INIS)

    Dufour, E.; Brosset, C.; Lowry, C.; Bucalossi, J.; Chappuis, P.; Corre, Y.; Desgranges, C.; Guirlet, R.; Gunn, J.; Loarer, T.; Mitteau, R.; Monier-Garbet, P.; Pegourie, B.; Reichle, R.; Thomas, P.; Tsitrone, E.; Hogan, J.; Roubin, P.; Martin, C.; Arnas, C.

    2005-01-01

    A better understanding of deuterium retention mechanisms requires the knowledge of carbon sources in Tore-Supra. The main source of carbon in the vacuum vessel during long discharges is the toroidal pumped limiter (TPL). This work is devoted to the experimental characterisation of the carbon source from the TPL surface during long discharges using a visible spectroscopy diagnostic. Moreover, we present an attempt to perform a carbon balance over a typical campaign and we discuss it with regards to the deuterium in-vessel inventory deduced from particle balance and the deuterium content of the deposited layers. The study shows that only a third of the estimated deuterium trapped in the vessel is trapped in the carbon deposits. Thus, in the present state of our knowledge and characterisation of the permanent retention, one has to search for mechanisms other than co-deposition to explain the deuterium retention in Tore Supra. (A.C.)

  13. The response of the Tore Supra edge plasma to supersonic pulsed gas injection

    Czech Academy of Sciences Publication Activity Database

    Pánek, Radomír; Gunn, J. P.; Bucalossi, J.; Ďuran, Ivan; Geraud, A.; Hron, Martin; Loarer, T.; Pégourié, B.; Stöckel, Jan; Tsitrone, E.

    337-339, č. 16 (2005), s. 530-534 ISSN 0022-3115. [Plasma Surface Interactions /16./. Portland, 24.5.2005-28.5.2005] R&D Projects: GA ČR(CZ) GP202/03/P062 Institutional research plan: CEZ:AV0Z20430508 Keywords : Edge plasma * Gas injection and fuelling * probes * Plasma flow * Tore Supra Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.414, year: 2005

  14. A long-period analog integrator for magnetic measurements in Tore Supra

    International Nuclear Information System (INIS)

    Prou, M.; Barth, E.; Couturier, P.; Ouvrier-Buffet, P.

    1998-01-01

    A new analog integrator, called 'Integrateur 2000', has been developed for precise integration over long periods for the magnetic signals of Tore Supra with the aim of 1000 s discharges for the CIEL project. This new integrator is already in routine use for the present pulse lengths of up to 200 s in T-S, and tests have been carried out up to 1000 s with less than 2 mV of drift. (author)

  15. MHD stability and mode locking in pre-disruptive plasmas on TORE SUPRA

    International Nuclear Information System (INIS)

    Vallet, J.C.; Edery, D.; Joffrin, E.; Lecoustey, P.; Mohamed-Benkadda, M.S.; Pecquet, A.L.; Samain, A.; Talvard, M.

    1991-01-01

    Experiments devoted to the study of MHD activity have been carried out on TORE SUPRA. The observed disruptions are preceded by the growth of an m=2 N=1 rotating mode which locks when the magnetic field perturbation exceeds a critical value. The mode locking is interpreted as a bifurcation of the mode frequency. In addition, stabilization of the m=2 N=1 tearing mode has been obtained with the Ergodic Divertor (ED)

  16. Towards high-power long-pulse operation on Tore Supra

    International Nuclear Information System (INIS)

    Becoulet, A.

    1999-01-01

    The present paper deals with the experimental and modelling activity linked to the preparation of the long-pulse high-power discharges using the present Tore Supra equipment: heating and current drive scenarios, power coupling, confinement and transport studies, discharge control,... An overview of the results obtained in that field is presented, as well as the progress required in the coming years, and the expected performance, for the CIEL phase, in terms of current drive and confinement

  17. Experimental results of Tore Supra neutral beam injector in the line testing system

    International Nuclear Information System (INIS)

    Fumelli, M.; Jequier, F.

    1991-04-01

    Results of the tests carried out on one of the six Tore Supra neutral beam injectors are reported. Several minor modifications of the injector design allowed us to operate up to 92 keV - 30 A beams limited by the high voltage power supplies. Results of studies on different topics like new titanium pumping system, neutron yield from neutraliser and target, beam conditioning and breakdown statistical analysis are also reported [fr

  18. ICRF plasma production in Tore Supra: analysis of antenna coupling and plasma properties

    International Nuclear Information System (INIS)

    Beaumont, B.; Becoulet, A.; Lyssoivan, A.

    1999-01-01

    A study of RF plasma production frequency range ω. 2ω ci has been undertaken on Tore Supra taking into account antenna coupling predictions of theory and the TEXTOR-94 database. Two scenarios for RF discharges have been tested (fixed frequency of the RF generator): operation with pure toroidal magnetic field, at standard and lower B T and operation in the magnetic configuration with a small vertical (B V ) field superimposed on B T (B V T ). (authors)

  19. The Tore Supra Front Steering Long Pulse ECRH Antenna. Past Experience - Present Evolution - Future Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Lennholm, M; Bouquey, F; Chantant, M; Chappuis, P; Clary, J; Darbos, C; Doceul, L; Faisse, F; Jung, M; Lambert, R; Magne, R; Montecot, A; Poli, S; Roux, D; Samaille, F; Traisnel, E; Villedieu, E [Association Euratom-CEA, DSM/DRFC, CEA/Cadarache, 13108 St. Paul-lez-Durance (France)

    2005-01-01

    The long pulse capability of Tore Supra and its ECRH system makes it an ideal machine to prove steady state feedback control as required in ITER. Although Neoclassical Tearing Modes (NTMs) have not yet been observed on Tore Supra, the control of other MHD modes represents a very similar task from a control point of view and the stabilisation of such modes for long periods using ECRH will provide essential experience for the implementation of such control schemes on ITER. For this work to progress on Tore Supra, it must be possible to vary the injection angles in real time under feedback control from measured plasma parameters. At Tore Supra the front mirror position - and hence the injection angles - is adjusted using stepper motors controlled through a serial link. The use of a serial link limiting the sampling time for the control system to 50-100 ms and the dynamic response of the stepper motors results in a system frequency response <5 Hz. For following the evolution of the current profile this seems fully acceptable though it could prove somewhat slow for reacting to fast beta changes. In any case the time constants typically associated with growth rate of NTMs on ITER would not require a significantly faster system. Initial commissioning of such real time control was performed in 2004. Unfortunately, following a limited number of cycles and prior to using the system for plasma experiments, one water bellows in the antenna ruptured indicating a major weakness in the design of the water-cooling system for the antenna. As a consequence the water-cooling of the mobile mirror has been redesigned. Careful calculations and subsequent tests have been used to optimise the trajectories of the flexible water connections and a more robust layout is being implemented, ready for operation in the summer 2005.

  20. The Tore Supra control, computer system : six years of operation and improvements

    International Nuclear Information System (INIS)

    Journeaux, J.Y.; Badie, O.; Chatelier, E.; Hennion, F.; Lebourg, P.; Leveque, P.; Hernandez, M.; Moulin, D.

    1995-01-01

    The Tore Supra control computer system has been providing a good operation of the Tore Supra machine for six years. It controls all of the sub-systems, the continuous ones as well as the sequential ones, and the automatic operation is very efficient. The control system has been programmed by the users themselves thanks to its user-friendly qualities, in order to keep the full control and knowledge of the automatisms. Nevertheless, some improvements are now necessary. Their main principles are : to choose ergonomic and powerful tools, industrial standards, and to keep the users's participation. The whole control system will be upgraded : the automatism level as well as the display level and the communication networks. The operator's driving job is analysed as supervisory and diagnostic tasks which have an effect on the Tore Supra machine efficiency. So a very powerful driving software has been chosen and linked with an expert system, which is to be designed and implemented with the aim to give an immediate accurate and global understanding of the process and situations, in particular in case of trouble. The method is based on an artificial intelligence approach, and it exploits both the process' informations and automatisms' steps, to determine the process state, next possible states and diagnosis of the process troubles. (orig.)

  1. The data acquisition and interlock system for Tore Supra infrared imaging

    International Nuclear Information System (INIS)

    Moulin, D.; Balorin, C.; Buravand, Y.; Caulier, G.; Ducobu, L.; Guilhem, D.; Jouve, M.; Roche, H.

    2003-01-01

    The data acquisition for the infrared measurement system on Tore-Supra is a key element in ensuring the supervision of the new actively-cooled plasma facing components of the CIEL project. It will allow us to follow the thermal evolution of components of Tore-Supra, in particular the toroidal pumped limiter (LPT) (360 deg.-15 m long) and the five additional heating launchers. When fully installed, the infrared measurement system will be composed of 12 digital 16-bit infrared cameras. They cover a 100-1200 deg.C temperature range and each picture has a definition of 320x240 pixels with a 20 ms time resolution. The objectives of the data acquisition system is real-time recording and analysis of each view element for further post-pulse analysis in order to understand the physics phenomenon and ensure the supervision of the plasma facing components and also to be part of the global feedback control system of Tore Supra

  2. Infrared reflection properties and modelling of in situ reflection measurements on plasma-facing materials in Tore Supra

    International Nuclear Information System (INIS)

    Reichle, R; Desgranges, C; Faisse, F; Pocheau, C; Lasserre, J-P; Oelhoffen, F; Eupherte, L; Todeschini, M

    2009-01-01

    Tore Supra has-like ITER-reflecting internal surfaces, which can perturb the machine protection systems based on infrared (IR) thermography. To ameliorate this situation, we have measured and modelled in the 3-5 μm wavelength range the bi-directional reflection distribution function (BRDF) of wall material samples from Tore Supra and conducted in situ reflection measurements and simulated them with the CEA COSMOS code. BRDF results are presented for B 4 C and carbon fibre composite (CFC) tiles. The hemispherical integrated reflection ranges from 0.12 for the B 4 C sample to 0.39 for a CFC tile from the limiter erosion zone. In situ measurements of the IR reflection of a blackbody source off an ICRH and an LHCD antenna of Tore Supra are well reproduced by the simulation.

  3. First plasma experiments in Tore Supra with a new generation of high heat flux limiters for RF antennas

    International Nuclear Information System (INIS)

    Agarici, G.; Beaumont, B.; Bibet, Ph.; Bremond, S.; Bucalossi, J.; Colas, L.; Durocher, A.; Gargiulo, L.; Ladurelle, L.; Lombard, G.; Martin, G.; Mollard, P.

    2000-01-01

    During the 1997 and 1998 Tore Supra shutdown, a first set of new antenna guard limiters was installed on one of the three ion cyclotron resonance heating (ICRH) antennas of Tore Supra. This limiter, which was one of the main technological studies of the 1998 campaign, was widely experimented in real plasma conditions, thus allowing the validation in situ, for the first time, of the technology of active metal casting (AMC) for plasma facing components. The huge improvement in the thermal response of the new limiter generation, compared to the old one, is shown on plasma pulses made identical in terms of antenna position and injected RF power profile. By using the infrared cameras installed inside Tore Supra and viewing the antennas front, the power density fluxes received by the carbon fibre composite (CFC) surface of the limiter were evaluated by correlation with the heat load tests made on the electrons beam facility of CEA/Framatome

  4. Progress in ergodic divertor operation on Tore Supra

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Colas, L.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Azeroual, A.; Basiuk, V.; Beaumont, B.; Becoulet, A.; Bremond, S.; Bucalossi, J.; Capes, H.; Corre, Y.; Costanzo, L.; Michelis, C. de; Devynck, P.; Feron, S.; Friant, C.; Garbet, X.; Giannella, R.; Grisolia, C.; Hess, W.; Hogan, J.; Ladurelle, L.; Laugier, F.; Martin, G.; Mattioli, M.; Meslin, B.; Monier-Garbet, P.; Moulin, D.; Nguyen, F.; Pascal, J.Y.; Pecquet, A.L.; Pegourie, B.; Reichle, R.; Saint-Laurent, F.; Vallet, J.C.; Zabiego, M.

    1999-09-01

    Upgrade of the Tore ergodic divertor has led to significant progress in ergodic divertor physics. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching 2 on the magnetic surface q = 2 + 3 / 12. This experimentally observed robustness allows one to operate at very low safety factor on the separatrix (q ∼ 2). Numerical analysis of ballooning turbulence in a stochastic layer indicates that the decay of the density fluctuations is in associated with an increase of the fluctuating electric drift velocity. The bottom line is then an enhanced cross-field transport in the vicinity of the target plates. This lowering of confinement appears to be compensated by an intrinsic transport barrier on the electron temperature. The 3-D response of the temperature field is computed with a fluid code. The intrinsic transport barrier at the separatrix, reported experimentally, can be recovered together with small amplitude temperature modulations in the divertor volume. Experimental evidence of the 3 density regimes (linear, high recycling and detachment) is reported. The low critical density values for these transitions indicate that similar parallel physics govern the axisymmetric and ergodic divertor, despite the open configuration of the latter. Measurement and understanding of these density regimes provide a means for feedback control of plasma density and an improvement in ICRH coupling scenarios. Experimental data also indicated that particle control with the vented target plates is effective. Increase of impurity control and radiation efficiency are recalled. Global power balance has been analysed. These results confirm the enhanced radiation capacity of the ergodic divertor. (author)

  5. Experimental study of turbulence on Tore Supra by plasma micro-waves interaction

    International Nuclear Information System (INIS)

    Colas, L.

    1996-01-01

    Internal small-scale magnetic turbulence is a serious candidate to explain the anomalous heat transport in tokamaks. This turbulence is badly known in the gradient region of large machines. In this work internal magnetic fluctuations are measured on Tore Supra with an original diagnostic : Cross Polarization Scattering (CPS). This experimental tool relies on the Eigenmode change of a probing polarised microwave beam scattered by magnetic fluctuations, close to a cut-off layer for the incident wave. In this work, the diagnostic is first qualified to assess its sensitivity to magnetic fluctuations, and the spatial localisation for its measurements. The magnetic fluctuation behaviour is then analysed over a wide range of plasma current, density and additional power, and interpreted with a simple 1-D scattering model. A scan of the plasma density or magnetic field is used to move the CPS measurement location from r/a = 0.3 to r/a = 0.75. A fluctuation radial profile is obtained by two means. In L-mode discharges, the relation between magnetic fluctuations, temperature profiles and local heat diffusivities is investigated. With all measurements, it is also possible to look for a local parameter correlated to the turbulence in a large domain of plasma conditions. The fluctuation-induced local heat diffusivity expected from the measured fluctuations is estimated using the non-collisional quasi-linear formula: X mag e = πqRV te (δB / B) 2 . Both the absolute values and the parametric dependence of calculated X mag e are close to the electron thermal diffusivities Xe determined by transport analysis. In particular, a threshold is evidenced in the dependence of fluctuation-induced heat fluxes on local ∇T e , which is analogous to the critical gradient for measured heat fluxes. The experimental setup is also sensitive to the Thomson scattering of the probing wave by density fluctuations. Its measurements are analysed as the fluctuations of the amplitude and the phase of

  6. Copper alloy-stainless steel bonds and recent developments for Tore Supra

    International Nuclear Information System (INIS)

    Lipa, M.; Chappuis, Ph.; Mitteau, R.; Reindl, G.

    1998-01-01

    High strength high conductivity copper alloys such as CuCrZr are used in Tore Supra as structural heat sink material for high heat flux plasma facing components. Although friction welded pipes on these components have shown a satisfactory in-service reliability, it came out that recently produced CuCrZr-stainless steel bonds showed very poor results in ductility. This led to a more detailed investigation of friction weldments. Finally, a more ductile joint has been developed with 'Plansee AG' on the basis of electron beam welded nickel adapters inserted between the two materials. Characterisation of such bonds is reported. (author)

  7. Controlled irradiation of CFC samples in the scrape-off layer of Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Gunn, J.P. [DRFC, Bat. 508, Association Euratom - CEA, CEA/DSM/DRFC, Centre de Cadarache, 13108 Saint Paul Lez Durance (France)]. E-mail: jamie.gunn@cea.fr; Begrambekov, L. [Moscow Engineering and Physics Institute, 115409 Moscow (Russian Federation); Brosset, C. [DRFC, Bat. 508, Association Euratom - CEA, CEA/DSM/DRFC, Centre de Cadarache, 13108 Saint Paul Lez Durance (France); Gordeev, A. [Moscow Engineering and Physics Institute, 115409 Moscow (Russian Federation); Loarer, T. [DRFC, Bat. 508, Association Euratom - CEA, CEA/DSM/DRFC, Centre de Cadarache, 13108 Saint Paul Lez Durance (France); Miljavina, E. [Moscow Engineering and Physics Institute, 115409 Moscow (Russian Federation); Shigin, P. [Moscow Engineering and Physics Institute, 115409 Moscow (Russian Federation); Khodja, H. [Laboratoire Pierre Suee, CEA/CNRS, CEA Saclay, 91191 Gif sur Yvette (France); Oddon, P. [DRFC, Bat. 508, Association Euratom - CEA, CEA/DSM/DRFC, Centre de Cadarache, 13108 Saint Paul Lez Durance (France); Pascal, J.-Y. [DRFC, Bat. 508, Association Euratom - CEA, CEA/DSM/DRFC, Centre de Cadarache, 13108 Saint Paul Lez Durance (France); Vartanian, S. [DRFC, Bat. 508, Association Euratom - CEA, CEA/DSM/DRFC, Centre de Cadarache, 13108 Saint Paul Lez Durance (France)

    2005-03-01

    The first experiments with a mobile sample holder in Tore Supra are described. It exposes 10 CFC samples to direct irradiation by the scrape-off layer plasma. The plasma parameters are measured simultaneously by two Langmuir probes, and the temperature of the samples by embedded thermocouples. The cumulated irradiation dose during the first brief campaign was enough to exceed the classical saturation of the ion stopping zone, as verified by thermodesorption spectroscopy and nuclear reaction analysis. Scanning electron microscopy of some of the samples was performed before and after irradiation in order to investigate the evolution of the surface structure due to ion bombardment.

  8. Analytical and numerical studies of sawtooth stabilization with on-axis ICRH on Tore Supra

    International Nuclear Information System (INIS)

    Zabiego, M.; Basiuk, V.; Becoulet, A.; Garbet, X.; Saoutic, B.; White, R.B.; Wu, Y.

    1995-01-01

    Sawtooth stabilization has recently been achieved with on-axis ICRH on Tore Supra. Analytic and numerical tools have been developed that allow analysis of these data, with the emphasis on the energetic-particle contribution. The latter is calculated on the basis of an original expression for the hot-ion distribution function, mostly-passing, deduced from theoretical and experimental ICRH studies, and compared to the usual mostly-trapped model. Both models account for stabilization, but for different reasons: with the mostly-passing distribution, stabilization is essentially driven by barely-passing hot ions. (author) 9 refs.; 2 figs

  9. Infrared thermography of the Tore Supra inner chamber: data acquisition and interpretation

    International Nuclear Information System (INIS)

    Fleury, I.

    1990-03-01

    Plasma-wall interactions occur in Tore Supra and undesirable effects take place in the plasma. Moreover, the damage of components may be induced by high energy concentrations. By means of the infrared Tomography diagnostic the components under high thermal flow can be controlled. The design and application of this diagnostic technique is presented. The optical and mechanical structure of the already installed endoscope are described. Three endoscopes are to be used in Tore Supra. The surveillance of the plasma limiters and ergodic divertors can be performed. The monitoring of the temperature variation on the head of the horizontal pumped limiter, during shocks, is carried out. The favourable effect of the ergodic divertor on the power stored on the head is confirmed. The heat decreasing radial length (λ Q ) and the limiter maximal temperatures (Tmax) are evaluated during shocks without a divertor (λ Q = 0.9 cm, Tmax = 440 deg C) or with a divertor current of 40 RA (λ Q = 2.5 cm, Tmax = 230 deg C). Different structures of power concentration, obtained under the effect of the magnetic perturbations induced by the divertor, are shown by means of infrared thermography [fr

  10. Thermal electron transport in regimes with low and negative magnetic shear in Tore Supra

    International Nuclear Information System (INIS)

    Voitsekhovitch, I.; Litaudon, X.; Moreau, D.; Aniel, T.; Becoulet, A.; Erba, M.; Joffrin, E.; Kazarian-Vibert, F.; Peysson, Y.

    1997-01-01

    The magnetic shear effect on thermal electron transport is studied in a large variety of non-inductive plasmas in Tore Supra. An improved confinement in the region of low and negative shear was observed and quantified with an exponential dependence on the magnetic shear (Litaudon, et al., Fusion Energy 1996 (Proc. 16th Int. Conf. Montreal, 1996), Vol. 1, IAEA, Vienna (1997) 669). This is interpreted as a consequence of a decoupling of the global modes (Romanelli and Zonca, Phys. Fluids B 5 (1993) 4081) that are thought to be responsible for anomalous transport. This dependence is proposed in order to complete the Bohm-like L mode local electron thermal diffusivity so as to describe the transition from Bohm-like to gyroBohm transport in the plasma core. The good agreement between the predictive simulations of the different Tore Supra regimes (hot core lower hybrid enhanced performance, reversed shear plasmas and combined lower hybrid current drive and fast wave electron heating) and experimental data provides a basis for extrapolation of this magnetic shear dependence in the local transport coefficients to future machines. As an example, a scenario for non-inductive current profile optimization and control in ITER is presented. (author)

  11. Near infrared spectra of carbon deposited layers from Tore Supra under plasma particle bombardment

    International Nuclear Information System (INIS)

    Delchambre, E.; Reichle, R.; Loarer, T.

    2003-01-01

    The authors present the results of laboratory investigations that show spectral luminance distributions similar to those emitted by plasma facing components in Tore-Supra experiments. The device used to produce plasma impact on the target is an Helicon source, where the gas is ionised with a 13.56 MHz RF generator. Different targets were tested: highly oriented pyrolytic graphite (HOPG) and pyrolytic graphite sampled from the MPL (modular pump limiter) neutralizer covered with flakes (a loosely attached deposited carbon layer). The target, placed in the centre of the vacuum vessel, was positively biased to ensure an electron bombardment only. NIR (near infra-red) spectral luminance deformation phenomena as observed in Tore-Supra, have been reproduced in the laboratory. Additional NIR luminance, with a maximum around 1.3 μm considering T(1.55 μm) as temperature reference, has shown up on the same carbon deposited layer that actually gave rise to the first reports on the phenomenon but not on HOPG. The phenomenon occurs in Ar and in H 2 when the current collected on the target exceeds 20 or 30 mA/cm 2 respectively. The intensity of the effect increases with growing target temperature and seemingly with growing collected current density and it disappears after electron bombardment with a time constant of 0.34 s. It shows some linear behaviour in Arrhenius plot

  12. Thermal electron transport in the regimes with low and negative magnetic shear on tore supra

    International Nuclear Information System (INIS)

    Voitsekhovitch, I.; Litaudon, X.; Moreau, D.; Aniel, T.; Becoulet, A.; Erba, M.; Joffrin, E.; Kazarian-Vibert, F.; Peysson, Y.

    1997-04-01

    The magnetic shear effect on the thermal electron transport is studied in a large variety of non-inductive plasmas of Tore Supra. An improved confinement in the region of low and negative shear was observed and quantified with an exponential dependence on the magnetic shear [Litaudon et al. in Plasma Physics and Controlled Nuclear Fusion Research, 1996, Montreal (International Atomic Energy Agency, Vienna, 1997) to be published]. This is interpreted as the consequence of a decoupling of the global modes [Romanelli and Zonka, Phys. Fluids B5 (1993), 4081] which are thought to be responsible for anomalous transport. This dependence is proposed to complete the Bohm-like L-mode local electron thermal diffusivity to describe the transition from the Bohm-like to the gyro-Bohm transport in the plasma core. The good agreement between the predictive simulations of the different Tore Supra regimes (hot core lower hybrid enhanced performance, reversed shear plasmas and combined lower hybrid current drive and fast wave electron heating) and experimental data gives a basis for the extrapolation of this magnetic shear dependence in the local transport coefficients for future machines. As an example a scenario for non-inductive current profile optimisation and control in ITER is presented. (author)

  13. Current profile control and magnetohydrodynamic stability in Tore Supra discharges with edge-plasma control by the ergodic divertor

    International Nuclear Information System (INIS)

    Zabiego, M.; Friant, C.; Ghendrih, P.; Becoulet, M.; Bucalossi, J.; Saint-Laurent, F.

    1999-01-01

    Although ergodic divertors are primarily designed to control particle and heat fluxes at the plasma edge, they also happen to affect the MHD stability of tokamak discharges. On Tore Supra, the ergodic divertor has long been known to stabilize the m/n=2/1 tearing mode induced, for instance, by edge radiation and detachment processes, thus allowing safe high-current and high-density operations. More recently, though, in discharges where ergodic divertor operations were optimised relative to the control of the edge-plasma (i.e., with large divertor perturbation), a detrimental increase in the disruptiveness has been observed. The action that the ergodic divertor has on the MHD activity is interpreted in terms of a redistribution of the current profile. The latter results from a large increase in the edge resistivity, primarily induced by the degradation of the electron energy confinement in the ergodic layer. The possibility that a transport barrier develops in the vicinity of the separatrix strongly affects the considered modelling. (authors)

  14. Absolute spectral characterization of silicon barrier diode: Application to soft X-ray fusion diagnostics at Tore Supra

    International Nuclear Information System (INIS)

    Vezinet, D.; Mazon, D.; Malard, P.

    2013-01-01

    This paper presents an experimental protocol for absolute calibration of photo-detectors. Spectral characterization is achieved by a methodology that unlike the usual line emissions-based method, hinges on the Bremsstrahlung radiation of a Soft X-Ray (SXR) tube only. Although the proposed methodology can be applied virtually to any detector, the application presented in this paper is based on Tore Supra's SXR diagnostics, which uses Silicon Surface Barrier Diodes. The spectral response of these n-p junctions had previously been estimated on a purely empirical basis. This time, a series of second-order effects, like the spatial distribution of the source radiated power or multi-channel analyser non linearity, are taken into account to achieve accurate measurements. Consequently, a parameterised physical model is fitted to experimental results and the existence of an unexpected dead layer (at least 5 μm thick) is evidenced. This contribution also echoes a more general on-going effort in favour of long-term quality of passive radiation measurements on Tokamaks

  15. The control of plasma density profile in Tore Supra. Comparison of different fueling techniques

    International Nuclear Information System (INIS)

    Commaux, N.

    2007-09-01

    The behaviour of a reactor-class plasma when fuelled using the existing techniques (gas puffing, supersonic molecular beam injection and pellet injection) is still very difficult to foresee. The present work has been initiated on Tore Supra in order to extrapolate the consequences of the different fuelling systems on ITER. Two main topics have been studied: the comparison of the plasma behaviour when fuelled using the different techniques at high Greenwald density fractions and the study of the homogenization following a pellet injection (main fuelling technique for ITER burning plasmas). The experiments at high Greenwald density fractions performed on Tore Supra showed that the plasma behaviour is very dependent on the fuelling method. The plasma energy confinement is following the scaling laws determined at low density when fuelled using pellet injection. which is better than for gas puffing and SMBI. both inducing a significant confinement loss. This behaviour is nor related to a transport modification: the ratio between effective diffusion and convection is similar to the pellet case. The difference between these shots is related only to the position of the matter source (at the edge for gas and close to the center for pellets). The study concerning the homogenization phenomena following a pellet injection aims mainly to study the ∇B-drift effect that expels the mater deposited by a pellet toward the low field side. A new phenomenon. which appears to be particularly important for the ∇B-drift during low field side injections. was then discovered: the influence of magnetic surfaces with an integer-valued safety factor (q). When the mater drifting toward low field side crosses an integer q surface. it experiences an important braking effect which stops the drift motion. It implies that the pellet material is mainly deposited on the last integer q surface crossed by the pellet during its injection. This study allows also to determine that the

  16. MHD stability of (2,1) tearing mode: an issue for the preforming phase of Tore Supra non-inductive discharges

    International Nuclear Information System (INIS)

    Maget, P.; Luetjens, H.; Huysmans, G.; Moreau, Ph.; Schunke, B.; Segui, J.-L.; Garbet, X.; Joffrin, E.; Luciani, J.F.

    2007-01-01

    The early phase of a tokamak plasma discharge can have a dramatic impact on the main heating phase. This has been a persistent problem for the development of the steady state, fully non-inductive scenario using lower hybrid current drive (LHCD) on Tore Supra. The present paper reports on recent experimental and numerical investigations showing that a tearing mode coupled to the internal kink grows on q = 2 in the ohmic phase when the total current is too low, due to the weakening of field line curvature stabilization. Then, the application of LHCD drives the island to a larger size and undermines the development of the non-inductive phase. Decreasing the edge safety factor or increasing the Lundquist number S is found to be beneficial in both the linear and non-linear MHD analyses. The experimental database, which allows covering the edge safety factor dependence, supports this interpretation

  17. The Tore Supra control, computer system: six years of operation and improvements

    International Nuclear Information System (INIS)

    Journeaux, J.Y.; Badie, O.; Chatelier, E.; Hennion, F.; Lebourg, P.; Leveque, P.; Hernandez, M.; Moulin, D.

    1994-01-01

    Some necessary improvements of the Tore Supra control computer system are discussed. Their main principles are: to choose ergonomic and powerful tools, industrial standards, and to keep the users's participation. The whole control system will be upgraded: the automatism level as well as the display level and the communication networks. A very powerful driving software has been chosen and linked with an expert system, which is to be designed and implemented with the aim to give an immediate accurate and global understanding of the process and situations, in particular in case of trouble. The method is based on an artificial intelligence approach, and it exploits both the process' informations and automatism' steps, to determine the process state, next possible states and diagnosis of the process troubles. (author) 3 refs., 2 figs

  18. Real time plasma feedback control: An overview of Tore-Supra achievements

    International Nuclear Information System (INIS)

    Martin, G.; Bucalossi, J.; Ekedahl, A.; Gil, C.; Grisolia, C.; Guilhem, D.; Gunn, J.; Kazarian, F.; Moulin, D.; Pascal, J.Y.; Saint-Laurent, F.

    2001-01-01

    Stable and reliable fusion plasma operation requires increasingly advanced control systems. This is especially true for steady-state operation in advanced modes, when several parameters are to be simultaneously optimised: e.g. the current profile, which has been related to the formation of internal transport barrier, and the density, which plays a crucial role both in the fusion power and in the plasma wall interactions. At a more technological level, good management of the power entering and leaving the plasma is required, by efficient additional heating coupling, and with a full control of radiation and convection losses and distribution to the first wall elements. For these goals, several feed-back mechanisms have been developed with success on Tore-Supra, in the past four years. Most of them are based on software, implemented in a set of micro-computers connected through a VME network. (author)

  19. Determination of q during sawtooth from inverse evolution of BAEs in Tore Supra

    Science.gov (United States)

    Amador, C. H. S.; Sabot, R.; Garbet, X.; Guimarães-Filho, Z. O.; Ahn, J.-H.

    2018-01-01

    Measuring the value of the safety factor (q) in the core during sawtooth cycles is still an open issue. A new method to measure q in Tore Supra plasma core is presented here. It relies on the analysis of the time evolution of a set of MHD modes detected after the sawtooth crashes. These modes are in the frequency range of previously observed Beta-induced Alfvén Eigenmodes, but with a frequency declining in time. The mode frequency analysis shows that the q profile is reversed when we have ICRH, after the sawtooth crash. In high current discharges (I_p>1.15 MA), the q-profile remains reversed for a longer time compared with lower plasma current discharges. Non-linear 3D MHD simulations of sawteeth performed with the XTOR-2F code (Lütjens and Luciani 2010 J. Comput. Phys. 229 8130-43) exhibit features that are similar to these observations.

  20. Raman study of CFC tiles extracted from the toroidal pump limiter of Tore Supra

    Science.gov (United States)

    Pardanaud, C.; Giacometti, G.; Martin, C.; Ruffe, R.; Angot, T.; Aréou, E.; Pégourié, B.; Tsitrone, E.; Dittmar, T.; Hopf, C.; Jacob, W.; Schwarz-Selinger, T.; Roubin, P.

    2011-08-01

    The structure of six tiles extracted from the erosion and deposition zones (thin and thick deposition) of the Tore Supra toroidal pump limiter (TPL) have been analyzed in the framework of the DITS campaign using micro-Raman spectroscopy. This post-mortem analysis gives information on both carbon structure and D content. We have found that the carbon structure is most often similar to that of plasma-deposited hard amorphous carbon layers. The role of the surface temperature during the discharge in the D content is investigated: in all locations where the temperature does not reach more than 500 °C the D content seems to be roughly uniform with D/D + C ≈ 20%.

  1. Modelling the erosion/deposition pattern of the Tore Supra Toroidal Pumped Limiter

    International Nuclear Information System (INIS)

    Panayotis, S.; Pégourié, B.; Borodin, D.; Kirschner, A.; Gunn, J.; Marandet, Y.; Mellet, N.

    2015-01-01

    This paper aims at understanding the main processes responsible for the erosion/deposition pattern observed on the surface of the Toroidal Pumped Limiter of Tore Supra, using the 3D local impurity transport code ERO. The influence of the plasma impurity content, CX-flux and surface temperature on the global carbon balance and erosion/deposition pattern is discussed. Main results are (1) that considering medium-range transport of C ions is mandatory for reproducing the main characteristics of the global C balance and erosion/deposition pattern, (2) that impurities and CX-atoms increase the erosion by a factor ⩽2 (without changing the net/gross erosion ratio), and (3) that chemical erosion is governed by the re-erosion of deposits, which depends strongly on the surface temperature

  2. Determination of ion temperatures from Zeeman broadened spectral lines in the edge of Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, C.C.; Isler, R.C.; Tobin, S.J.; Hogan, J.T. [Oak Ridge National Lab., TN (United States). Fusion Energy Div.; Hess, W.R. [Association EURATOM-CEA sur la Fusion Controlee, St-Paul-lez-Durance (France). Centre d`Etudes de Cadarache

    1994-09-01

    The authors have examined a {sup 3}P {yields} {sup 3}S multiplet of C III in Tore Supra in order to assess the possibility of determining the ion temperatures from transitions where the Zeeman effect cannot be neglected compared to the Doppler broadening. The preliminary studies lead them to believe that with good quality data the temperatures can be determined within about 20% in the 20--30 eV range and within about 50% in the neighborhood of 5 eV by fitting the entire multiplet rather than a semi-isolated feature, even though certain parameters important for the analysis, such as polarization effects of the optics, are not well characterized. In order to quantify these conclusions more precisely, future work will concentrate on developing numerical fitting routines and on examining the validity of the assumption that the distribution function for low ionization stages is Maxwellian.

  3. Determination of ion temperatures from Zeeman broadened spectral lines in the edge of Tore Supra

    International Nuclear Information System (INIS)

    Klepper, C.C.; Isler, R.C.; Tobin, S.J.; Hogan, J.T.; Hess, W.R.

    1994-01-01

    The authors have examined a 3 P → 3 S multiplet of C III in Tore Supra in order to assess the possibility of determining the ion temperatures from transitions where the Zeeman effect cannot be neglected compared to the Doppler broadening. The preliminary studies lead them to believe that with good quality data the temperatures can be determined within about 20% in the 20--30 eV range and within about 50% in the neighborhood of 5 eV by fitting the entire multiplet rather than a semi-isolated feature, even though certain parameters important for the analysis, such as polarization effects of the optics, are not well characterized. In order to quantify these conclusions more precisely, future work will concentrate on developing numerical fitting routines and on examining the validity of the assumption that the distribution function for low ionization stages is Maxwellian

  4. Measurement of the mean ionic charge on Tore supra by visible bremsstrahlung radiation

    International Nuclear Information System (INIS)

    Thieyacine Fall, M.

    1991-01-01

    The effective plasma charge Zeff (proportional to plasma impurity ratio) is of prime importance for controlled fusion by magnetic confinement because it is involved for plasma ignition. From bremsstrahlung radiation theory in the visible part of the spectrum it is shown how effective charge in the plasma is deduced. A validity criterion is established from experiments (radiation at λ = 5235 A) to obtain Zeff profile and error estimation. This profile allows the calculation of resistivity profiles from different theories which are compared to a magnetohydrodynamic code. Calculation time is reduced by a fast analysis method from global parameters given time evolution of Zeff. This last measurement is essentially used for interpretation of experimental results of the Tore Supra physical program [fr

  5. Energy measurement of fast ions trapped in the toroidal field ripple of Tore Supra

    International Nuclear Information System (INIS)

    Basiuk, V.; Becoulet, A.; Hutter, T.; Martin, G.; Pecquet, A.L.; Saoutic, B.

    1993-09-01

    During additional heating in Tore Supra (ICRF or NBI) fast ion losses due to the toroidal field ripple were clearly measured by a set of graphite probes. This diagnostic collects the flow of fast ions entering a vertical port and usually shows a maximum flux for ions originating from the vicinity of surface δ * = 0. During the monster sawteeth regime, achieved with ICRF, a remarkable phenomenon was observed: the ejection of fast ions, not correlated with any measured MHD activity. The radial distribution of these ions is quite different from that usually observed exhibiting a peak located in the central section of the plasma. In order to measure the energy distribution of these ions, from 80 keV (energy of the neutral beam injected in Tore Supra) up to 1 MeV (expected during ICRF), a new diagnostic is under construction. The principle of the diagnostic is to discriminate the ions in energy using their Larmor radius (p = 1.3 cm for 100 keV → p = 3.6 cm for 700 keV, B = 4T). The detector is made of a hollow graphite cylinder with a small entrance slot, located in a vertical port on the ion drift side. An array of six metallic collectors placed inside the graphite cylinder intercepts the ions. The current on each collector was estimated at 10 → 100 nA, during ICRF heating. The energy resolution of this diagnostic is expected to be about 20 keV for the lowest energy range and 100 keV for the highest. This type of ruggedized detector might be extrapolated for the measurements of alpha particle losses in future DT experiments. It should also be suitable for the studies of stochastic ripple diffusion. (authors). 3 refs., 9 figs

  6. Transport in the plasma edge specific connection to the wall in the Tore Supra ergodic divertor experiments

    International Nuclear Information System (INIS)

    Grosman, A.; Ghendrih, P.; DeMichelis, C.; Monier-Garbet, P.; Vallet, J.C.; Capes, H.; Chatelier, M.; Geraud, A.; Goniche, M.; Grisolia, C.; Guilhem, D.; Harris, G.; Hess, W.; Nguyen, F.; Poutchy, L.; Samain, A.

    1992-01-01

    The ergodic divertor experiments in TORE SUPRA can be analysed along two main lines. The first one refers to the change of the heat and particle transport in the ergodized zone. This is especially true for the electron heat transport which is enhanced in the edge layer. But other distinctive features give evidence of the importance of the parallel connexion length between the plasma edge and the wall. The field lines, which are stochastic in the major part of the perturbed layer (10-15 cm) are such that, in the outermost layer (3 cm), the connexion topology is regular. This has obvious effects on the particle and power deposition, but also on the plasma parameters, and consequently influences the particle recycling and impurity shielding processes. The TORE SUPRA ergodic divertor experiments are reviewed in this framework

  7. Real time control of fully non-inductive 6 minute, 1 Gigajoule plasma discharges in Tore Supra

    International Nuclear Information System (INIS)

    Houtte, D. van; Martin, G.; Bucalossi, J.; Saint-Laurent, F.

    2005-01-01

    The experimental programme of Tore Supra has been devoted in 2003 to study simultaneously heat removal capability and particle exhaust in steady-state fully non-inductive current drive discharges. This required both advanced technology integration and steady-state real time plasma control. In particular, an improvement of the plasma position within a few millimetres range, and new real time cross controls between RF power and various actuators built around a shared memory network, have allowed Tore Supra to access a powerful steady-state regime with an improved safety level for the actively cooled plasma facing components. Feedback controlled fully non-inductive plasma discharges have been sustained in a steady-state regime up to 6 minutes with a new world record of injected-extracted energy exceeding 1 GJ. Advanced tools, experimental results and brief physics analysis of these discharges are presented. (author)

  8. Measurements of scrape-off layer ion-to-electron temperature ratio in Tore Supra ohmic plasmas

    Czech Academy of Sciences Publication Activity Database

    Kočan, M.; Gunn, J. P.; Pascal, J.-Y.; Bonhomme, G.; Devynck, P.; Ďuran, Ivan; Gauthier, E.; Ghendrih, P.; Marandet, Y.; Pegourie, B.; Vallet, J.-C.

    390-391, - (2009), s. 1074-1077 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/18th./. Toledo, 26.05.2008-30.05. 2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Ion temperature * Electron temperature * Edge plasma * Tore Supra Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.933, year: 2009

  9. Deuterium Inventory in Tore Supra (DITS): 2nd post-mortem analysis campaign and fuel retention in the gaps

    International Nuclear Information System (INIS)

    Dittmar, T.; Tsitrone, E.; Pegourie, B.; Cadez, I.; Pelicon, P.; Gauthier, E.; Languille, P.; Likonen, J.; Litnovsky, A.; Markelj, S.; Martin, C.; Mayer, M.; Pascal, J.-Y.; Pardanaud, C.; Philipps, V.; Roth, J.; Roubin, P.; Vavpetic, P.

    2011-01-01

    A dedicated study on fuel retention has been launched in Tore Supra, which includes a D wall-loading campaign and the dismantling of the main limiter (Deuterium Inventory in Tore Supra, DITS project). This paper presents new results from a second post-mortem analysis campaign on 40 tiles with special emphasis on the D retention in the gaps. SIMS analysis reveals that only 1/3 of the thickness of deposits in the plasma shadowed zones are due to the DITS wall-loading campaign. As pre-DITS deposits contain less D than DITS deposits, the contribution of DITS to the D inventory is about 30-50%. The new estimate for the total amount of D retained in the Tore Supra limiter is 1.7 x 10 24 atoms, close to the previous estimate, with the gap surfaces contributing about 33%. NRA measurements show a stepped decrease of D along the gap with strong asymmetries between different gap orientations.

  10. Experience feedback from high heat flux component manufacturing for Tore Supra

    International Nuclear Information System (INIS)

    Schlosser, J.; Durocher, A.; Huber, T.; Garin, P.; Schedler, B.; Agarici, G.

    2001-01-01

    Tore Supra is involved in flat tile carbon armoured plasma facing components (PFCs) since 1985. In 1997, a third generation of components, based on the original concept developed with Plansee Company, called active metal casting (AMC[reg]), has been launched. Since 1998, 660 elementary components for the toroidal pump limiter (TPL) are in production. The route of the manufacture is rather complex and many controls were requested all along the fabrication to insure a high reliability of the elements. One of the main controls is the final infrared (IR) test allowing to determine the quality of the bonding between the carbon fibre composite (CFC) tiles and the heat sink made of copper-chromium-zirconium alloy (CuCrZr). Although results for the first batch of elements were as expected (less than 5% rejected at the final test), unexpected defects appeared with the followings batches. Investigations on the fabrication processes underlined the importance of having a better heat treatment of the pieces in copper alloy (CuCrZr), however this was not sufficient to completely explain the observed defects

  11. Power exhaust and edge control in steady state Tore Supra plasma

    International Nuclear Information System (INIS)

    Mitteau, R.

    2002-01-01

    Tore Supra is operated since 2001 with a flat limiter which is designed for 10 MW/m 2 . The limiter is located in the bottom of the vacuum vessel. It was only partial in 2001, but it is now fully toroidal without poloidal leading edges. Part of the experimental campaign of 2001 was devoted to the physical as well as technological qualification of the limiter. For 4 MW injected, the limiter extracted 2.5 MW and heat flux densities reached 2.5 MW/m 2 . It is still modest compared to the design value, but nonetheless enables a comparison to the modelling as surface temperature increased locally to 400 deg C. Thermal steady state is reached in 5-8 seconds. The values of heat flux and the deposition pattern are in very good accordance with design simulations. The heat flux pattern is a combination of parallel and perpendicular flow components which are roughly of equal magnitude. Insights on the heat flux deposition pattern as well as on the tiles behaviour are given. Operation with such a large size high heat flux component sets renewed emphasis on issues such as feed back systems, active security, cooling parameter and in situ assessment of the elements. They are dealt with in the paper. (author)

  12. Power exhaust and edge control in steady state tore supra plasma

    International Nuclear Information System (INIS)

    Mitteau, R.; Guilhem, D.; Vallet, J.C.

    2003-01-01

    Tore Supra is operated since 2001 with a toroidal limiter designed to remove continuously a heat power of 15 MW at a maximum (ITER relevant) heat flux density of 10 MW/m 2 . The limiter is located in the bottom of the vacuum vessel and is actively cooled by a pressurised water loop. For an injection of 6 MW of additional power, the limiter extracts 3.6 MW and the surface temperature stabilises typically at 400 deg C in 5 seconds during discharges lasting up to 260 seconds. The maximum heat flux density reach 3 MW/m 2 which is still modest compared to the design value, but nevertheless enables a comparison to the modelling. Heat flux and deposition pattern are in very good accordance with design simulations. Additional heat load concentrations specific to radio-frequency heating superimpose to the convection heat load at levels that are commensurable with the elements thermal capability. Insights on the tiles behaviour are given. Operation with such a large size high heat flux component sets an renewed emphasis on issues such as feed back systems, active security, cooling parameters and in situ assessment of the elements. These issues are dealt with in this paper. (author)

  13. Assessing braze quality in the actively cooled Tore Supra Phase III outboard pump limiter

    International Nuclear Information System (INIS)

    Nygren, R.E.; Lutz, T.L.; Miller, J.D.; McGrath, R.; Dale, G.

    1994-01-01

    The quality of brazing of pyrolytic graphite armor brazed to copper tubes in Tore Supra's Phase III Outboard Pump Limiter was assessed through pre-service qualification testing of individual copper/tile assemblies. The evaluation used non-destructive, hot water transient heating tests performed in the high-temperature, high-pressure flow loop at Sandia's Plasma Materials Test Facility. Surface temperatures of tiles were monitored with an infrared camera as water at 120 degrees C at about 2.07 MPa (300 psi) passed through a tube assembly initially at 30 degrees C. For tiles with braze voids or cracks, the surface temperatures tagged behind those of adjacent well-bonded tiles. Temperature tags were correlated with flaw sizes observed during repairs based upon a detailed 2-D heat transfer analyses. open-quotes Badclose quotes tiles, i.e., temperature tags of 10-20 degrees C depending upon tile's size, were easy to detect and, when removed, revealed braze voids of roughly 50% of the joint area. Eleven of the 14 tubes were rebrazed after bad tiles were detected and removed. Three tubes were rebrazed twice

  14. Modeling of neutral transport and impurity generation in the Tore Supra pump limiter

    International Nuclear Information System (INIS)

    Klepper, C.C.; Owen, L.W.; Mioduszewski, P.K.; Grosman, A.

    1989-01-01

    A system of modular pump limiters is employed in Tore Supra for particle control. It includes one large outboard module which is instrumented with an optical multi-channel analyzer viewing the neutralizer plate, in addition to pressure, temperature and particle diagnostics. The transport of neutrals in both the particle scoop mode and in the pump limiter mode with active pumping is modeled with the DEGAS code. Equilibrium neutral pressure and density, H α -intensity at the neutralizer plate, and particle back-flow into the scrape-off layer are calculated for typical projected plasma edge conditions. Effects on neutral particle transport of varying the length of the ''shelf'' separating the throat region from the pumping chamber are investigated. The generation and transport of carbon impurities arising from ion impact sputtering at the neutralizer plate are estimated. The short mean-free-path for ionization of sputtered carbon atoms suggests the possibility of spectroscopically measuring the effects of poloidal flux variations arising from shadowing of the modular limiter system. (orig.)

  15. Assessing braze quality in the actively cooled Tore Supra phase III outboard pump limiter

    International Nuclear Information System (INIS)

    Hygren, R.; Lutz, T.; Miller, J.

    1994-01-01

    This paper discusses the assessment of quality of brazing of pyrolytic graphite (PG) armor brazed to copper tubes in Tore Supra's Phase III Outboard Pump Limiter (OPL). The limiter head is a bank of 14 water-cooled copper tubes with several hundred brazed PG tiles. Braze quality was first assessed through pre-service qualification testing of individual copper/tiles assemblies. The quality of brazes was evaluated using (non-destructive) transient heating (open-quotes hot waterclose quotes) tests performed in the high temperature, high pressure flow loop at Sandia's Plasma Materials Test Facility. The surface temperatures of tiles were monitored with an infra-red (IR) camera as water at 120 degrees C water at about 2.07 MPa (300 psi) passed through a tube assembly initially at 30 degrees C. For tiles with braze voids or cracks, the surface temperatures lagged behind those of adjacent well bonded tiles. Temperature lags were correlated with flaw sizes observed during repairs using a detailed 2-D heat transfer analyses. open-quotes Badclose quotes tiles, i.e., temperature lags of 10-20 degrees C depending upon tile's size, were easy to detect and, when removed, revealed braze voids of roughly 50% of the joint area. 11 of the 14 tubes were rebrazed after bad tiles were detected and removed. Three tubes were re-brazed twice

  16. Deuterium in-vessel retention characterisation through the use of particle balance on Tore Supra

    International Nuclear Information System (INIS)

    Bucalossi, J.; Brosset, C.; Pegourie, B.; Tsitrone, E.; Dufour, E.; Eckedahl, A.; Geraud, A.; Goniche, M.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; Vallet, J.C.; Vartanian, S.

    2007-01-01

    Fuel retention inside plasma facing components will be a crucial issue not only in fusion reactors of the future, but also in ITER. The estimation of the fraction of the fuel which remains trapped inside the vessel is quite a difficult task. Particle balance analysis provides information for the whole vacuum chamber as a function of time and can be use to monitor the tritium in-vessel retention in real-time. On Tore Supra with a careful choice and position of pressure sensors, proper calibration procedures, the accuracy of the balance is around 10%. Particle balance analysis have been performed on many long pulse discharges and deuterium in-vessel retention has been found to be a constant around 5 x 10 20 D/s after several minutes of plasma. The evolution of the retention rate with plasma parameters indicates that deuterium bulk implantation and diffusion could dominate codeposition with carbon atoms. Particle balance is a powerful tool that should be implemented in ITER

  17. Kinetic Monte-Carlo modeling of hydrogen retention and re-emission from Tore Supra deposits

    International Nuclear Information System (INIS)

    Rai, A.; Schneider, R.; Warrier, M.; Roubin, P.; Martin, C.; Richou, M.

    2009-01-01

    A multi-scale model has been developed to study the reactive-diffusive transport of hydrogen in porous graphite [A. Rai, R. Schneider, M. Warrier, J. Nucl. Mater. (submitted for publication). http://dx.doi.org/10.1016/j.jnucmat.2007.08.013.]. The deposits found on the leading edge of the neutralizer of Tore Supra are multi-scale in nature, consisting of micropores with typical size lower than 2 nm (∼11%), mesopores (∼5%) and macropores with a typical size more than 50 nm [C. Martin, M. Richou, W. Sakaily, B. Pegourie, C. Brosset, P. Roubin, J. Nucl. Mater. 363-365 (2007) 1251]. Kinetic Monte-Carlo (KMC) has been used to study the hydrogen transport at meso-scales. Recombination rate and the diffusion coefficient calculated at the meso-scale was used as an input to scale up and analyze the hydrogen transport at macro-scale. A combination of KMC and MCD (Monte-Carlo diffusion) method was used at macro-scales. Flux dependence of hydrogen recycling has been studied. The retention and re-emission analysis of the model has been extended to study the chemical erosion process based on the Kueppers-Hopf cycle [M. Wittmann, J. Kueppers, J. Nucl. Mater. 227 (1996) 186].

  18. Improvement of density control by feedback on Langmuir probe signals in Tore Supra

    International Nuclear Information System (INIS)

    Gunn, J.; Bucalossi, J.; Costanzo, L.; Grisolia, C.; Ghendrih, Ph.; Grosman, A.; Loarer, T.; Martin, G.; Monier-Garbet, P.; Moulin, D.; Pascal, J.Y.; Saint-Laurent, F.

    1999-12-01

    Real time control of deuterium or helium gas injection by feedback on Langmuir probe signals is implemented in Tore Supra ergodic divertor discharges. The feedback schemes are based on the robust experimental observation that the density limit coincides with edge temperature T e ∼ 10 eV. Three control algorithms are used: (1) proportional feedback on the central line-averaged density with real-time attenuation of the system gain and security cut-off of the gas injection if the edge temperature becomes too low; (2) proportional feedback on the central line-averaged density with security cut-off controlled by the degree of detachment (DoD); (3) proportional feedback on edge temperature with security cut-off on the DoD. The DoD is defined for deuterium discharges, but not for helium since those do not detach. All three feedback modes permit operation close to the density limit and have been successfully applied for plasma currents 0.4 p p =1.4 MA with up to 4 MW of ICRH power. (author)

  19. Evaluation of the growth of carbonaceous deposit in steady state Tore Supra using infrared thermography

    International Nuclear Information System (INIS)

    Mitteau, R.; Guilhem, D.; Reichle, R.; Vallet, J.C.; Roche, H.; Buravand, Y.; Chantant, M.; Tsitrone, E.; Brosset, C.; Grosman, A.; Chappuis, P.

    2006-01-01

    Fusion devices with carbon as the main armour material are experiencing a growth in carbonaceous deposits at the surface of the plasma facing components. Tore Supra presents such deposits, and has specific features which influence their growth: long pulse operation and cooled walls. Deposits have a low thermal transfer to the cooled structure so that they appear as hot areas with the infrared imaging system looking at the elements surface temperature during plasma discharges. A 'degree of (carbon) deposit' on the toroidal pumped limiter is estimated by establishing the ratio between the apparent power on the limiter derived from the infrared measure and the actual one, deduced from a power balance analysis between the injected and the radiated power. This criterion is used to monitor the evolution of the deposit average thermal resistance. Successive shots have a similar 'degree of deposit', showing that the evaluation makes sense. Two years of data have been compiled (2003 and 2004), representing 3000 discharges (13 h of plasma, including 30 discharges longer than one minute). A three-fold increase in the 'degree of deposit' over six months is evidenced, following a limiter clean-up early in 2003. A comparison with calorimetric data produces a similar result, albeit less pronounced. Large steps in the degree of deposit are sometimes observed, usually correlated with identified events such as disruption, vessel opening, conditioning or plasma parameters change. It indicates that the deposit thermal resistance can change rapidly, although a systematic correlation with the above mentioned events could not be established

  20. Evaluation of the growth of carbonaceous deposit in steady state Tore Supra using infrared thermography

    Science.gov (United States)

    Mitteau, R.; Guilhem, D.; Reichle, R.; Vallet, J. C.; Roche, H.; Buravand, Y.; Chantant, M.; Tsitrone, E.; Brosset, C.; Grosman, A.; Chappuis, P.

    2006-03-01

    Fusion devices with carbon as the main armour material are experiencing a growth in carbonaceous deposits at the surface of the plasma facing components. Tore Supra presents such deposits, and has specific features which influence their growth: long pulse operation and cooled walls. Deposits have a low thermal transfer to the cooled structure so that they appear as hot areas with the infrared imaging system looking at the elements surface temperature during plasma discharges. A 'degree of (carbon) deposit' on the toroidal pumped limiter is estimated by establishing the ratio between the apparent power on the limiter derived from the infrared measure and the actual one, deduced from a power balance analysis between the injected and the radiated power. This criterion is used to monitor the evolution of the deposit average thermal resistance. Successive shots have a similar 'degree of deposit', showing that the evaluation makes sense. Two years of data have been compiled (2003 and 2004), representing 3000 discharges (13 h of plasma, including 30 discharges longer than one minute). A three-fold increase in the 'degree of deposit' over six months is evidenced, following a limiter clean-up early in 2003. A comparison with calorimetric data produces a similar result, albeit less pronounced. Large steps in the degree of deposit are sometimes observed, usually correlated with identified events such as disruption, vessel opening, conditioning or plasma parameters change. It indicates that the deposit thermal resistance can change rapidly, although a systematic correlation with the above mentioned events could not be established.

  1. X mode reflectometry for edge density profile measurements on Tore Supra

    International Nuclear Information System (INIS)

    Clairet, F.; Bottereau, C.; Chareau, J.M.; Paume, M.; Sabot, R.

    1999-01-01

    X mode heterodyne reflectometry associated with fast sweep capabilities demonstrates very precise measurement on Tore Supra and a high sensitivity (∼10 17 m -3 ) to density variations. Very good agreement with Thomson scattering measurement is observed. Fluctuations of the radial positions of the profile are no more than ± 0.5 cm. However, edge magnetic field ripple can be a concern since it is not easy to stand precisely for the wave trajectory into the plasma and for the toroidal position of the cutoff layer; nevertheless if the error can be estimated to be less than than 3 cm in the position of the whole profile, addition work is needed combining 3-D ray tracing and different antenna systems. Additional LH heating generates an ECE noise in the same frequency range of the reflectometer and is detected. This emission throughout the plasma is fortunately stopped by the upper X mode cutoff and is also reabsorbed by the electron cyclotron resonance. But at the very edge, due to a misalignment of the antenna to the plasma magnetic field and the low optical thickness of the plasma, the first cutoff frequency, i.e. the profile initialization, may be determined less precisely. (authors)

  2. Origin of the spectral deformation in the near infrared radiation from Tore-Supra carbon components; Origine de la deformation spectrale de la luminance proche infrarouge des composants en carbone de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Delchambre, E

    2004-03-01

    < 40 {mu}m. These results are confirmed by the laser ablation experiment's on CFC of A. Hermann et al. in which it is inferred that the size of hot spots is less than the 30 {mu}m spatial resolution of the IR camera which is used. Additional results are exhibited describing the micro-analysis of deposition samples removed from Tore Supra. The analyses have shown that the characteristics forms are due to deposition process and not to in situ surface alterations. The deposition is characterized and a ballistic model for the observed orientation and morphology is proposed and the global growth rate of these samples is evaluated. Correlations with the evolution of the infrared emission spectrum have been made, outlining the sequence of layer growth and the corresponding spectral deformation characteristics. (author)

  3. Experimental characterization and modelling of non-linear coupling of the lower hybrid current drive power on Tore Supra

    International Nuclear Information System (INIS)

    Preynas, M.; Goniche, M.; Hillairet, J.; Litaudon, X.; Ekedahl, A.; Colas, L.

    2013-01-01

    To achieve steady-state operation on future fusion devices, in particular on ITER, the coupling of the lower hybrid wave must be optimized on a wide range of edge conditions. However, under some specific conditions, deleterious effects on the lower hybrid current drive (LHCD) coupling are sometimes observed on Tore Supra. In this way, dedicated LHCD experiments have been performed using the LHCD system of Tore Supra, composed of two different conceptual designs of launcher: the fully active multi-junction (FAM) and the new passive active multi-junction (PAM) antennas. A non-linear interaction between the electron density and the electric field has been characterized in a thin plasma layer in front of the two LHCD antennas. The resulting dependence of the power reflection coefficient (RC) with the LHCD power is not predicted by the standard linear theory of the LH wave coupling. A theoretical model is suggested to describe the non-linear wave–plasma interaction induced by the ponderomotive effect and implemented in a new full wave LHCD code, PICCOLO-2D (ponderomotive effect in a coupling code of lower hybrid wave-2D). The code self-consistently treats the wave propagation in the antenna vicinity and its interaction with the local edge plasma density. The simulation reproduces very well the occurrence of a non-linear behaviour in the coupling observed in the LHCD experiments. The important differences and trends between the FAM and the PAM antennas, especially a larger increase in RC for the FAM, are also reproduced by the PICCOLO-2D simulation. The working hypothesis of the contribution of the ponderomotive effect in the non-linear observations of LHCD coupling is therefore validated through this comprehensive modelling for the first time on the FAM and PAM antennas on Tore Supra. (paper)

  4. Particle exhaust with vented structures: application to the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    Azeroual, A.

    2000-01-01

    In a thermonuclear reactor, one must continuously fuel the discharge and extract the ashes resulting from fusion reactions. To avoid the risk of discharge poisoning, α-particle concentration is limited to ∼ 10 %. To allow for steady-state conditions requires then to extract ≥2 % of the helium out flux. In Tore Supra, the ergodic divertor is the main component managing the heat and particle fluxes at the edge. Its principle consists in generating a resonant perturbation able to destroy magnetic surfaces at the plasma periphery. In this region, the field lines are open and connected at both ends to neutralizers which are wetted by the major part of the heat and particle fluxes and are the structures through which a part of the plasma out flux is pumped for maintaining the discharge in steady-state conditions. This work describes the neutral recirculation around the ergodic divertor and is based on a data base of 56 discharges. One discuss the two processes allowing for particle exhaust: the ballistic collection of ions and that of neutrals backscattered by atomic reactions. These two processes are modelled accounting for a realistic description of the divertor geometry. A comparison between simulations and experiments is presented for measurements characterising the three main actors of plasma-wall interaction: the edge plasma, the D α light emission and the neutral pressure in the divertor plenum. Last, one question how such a system can be extrapolated to next step machines, for which one must account for technical constraints linked to the presence of the shield protecting the coils from the high neutron flux. (author)

  5. The control of plasma density profile in Tore Supra. Comparison of different fueling techniques; Controle du profil de densite dans le plasma de Tore Supra. Comparaison de differentes methodes d'alimentation en particules

    Energy Technology Data Exchange (ETDEWEB)

    Commaux, N

    2007-09-15

    The behaviour of a reactor-class plasma when fuelled using the existing techniques (gas puffing, supersonic molecular beam injection and pellet injection) is still very difficult to foresee. The present work has been initiated on Tore Supra in order to extrapolate the consequences of the different fuelling systems on ITER. Two main topics have been studied: the comparison of the plasma behaviour when fuelled using the different techniques at high Greenwald density fractions and the study of the homogenization following a pellet injection (main fuelling technique for ITER burning plasmas). The experiments at high Greenwald density fractions performed on Tore Supra showed that the plasma behaviour is very dependent on the fuelling method. The plasma energy confinement is following the scaling laws determined at low density when fuelled using pellet injection. which is better than for gas puffing and SMBI. both inducing a significant confinement loss. This behaviour is nor related to a transport modification: the ratio between effective diffusion and convection is similar to the pellet case. The difference between these shots is related only to the position of the matter source (at the edge for gas and close to the center for pellets). The study concerning the homogenization phenomena following a pellet injection aims mainly to study the {nabla}B-drift effect that expels the mater deposited by a pellet toward the low field side. A new phenomenon. which appears to be particularly important for the {nabla}B-drift during low field side injections. was then discovered: the influence of magnetic surfaces with an integer-valued safety factor (q). When the mater drifting toward low field side crosses an integer q surface. it experiences an important braking effect which stops the drift motion. It implies that the pellet material is mainly deposited on the last integer q surface crossed by the pellet during its injection. This study allows also to determine that the {nabla

  6. Evident anomalous inward particle pinch in full non-inductive plasmas driven by lower hybrid waves on Tore Supra

    International Nuclear Information System (INIS)

    Hoang, G.T.; Bourdelle, C.; Pegourie, B.; Artaud, J.F.; Bucalossi, J.; Clairet, F.; Fenzi-Bonizec, C.; Garbet, X.; Gil, C.; Guirlet, R.; Imbeaux, F.; Lasalle, J.; Loarer, T.; Lowry, C.; Schunke, B.; Travere, J.M.; Tsitrone, E.

    2003-01-01

    These slides present some characteristics concerning peaked density profile observed in Tore-Supra. It appears that density profile remains peaked for more than 3 minutes in fully LHCD (lower hybrid current drive) discharges. The absence of toroidal electric field and the fact that the ware pinch has vanished across the entire plasma show that toroidal electric field and ware pinch are not the cause of the peaked profile. It is shown that peaked profile is linked to transport properties and can only be explained by a particle pinch velocity 2 orders of magnitude above the neoclassical pinch. It is also shown that the radial profile is in agreement with Isitchenko's formula. (A.C.)

  7. Flux consumption, current ramp-up and current diffusion in Tore Supra non-inductive Lower Hybrid scenarios

    International Nuclear Information System (INIS)

    Kazarian, F.; Litaudon, X.; Moreau, D.; Arslanbekov, R.; Hoang, G.T.; Joffrin, E.; Peysson, Y.; Allibert, J.P.; Ane, J.M.; Bremond, S.

    1995-01-01

    The main objective of the Lower Hybrid (LH) experiments performed on Tore Supra is to provide large flux savings for long pulse operation while controlling the plasma current density profile. This goal will be best achieved by applying LH wave directly during the current ramp-up phase. Experiments have been performed where a large fraction of the current is driven non-inductively during the ramp-up phase. A theoretical flux consumption scaling is presented and compared to experimental data. The time evolutions of the current density profiles are analysed with a new current diffusion code (CRONOS). In view to achieve fully non-inductive current drive discharges in a fast, systematic and reproducible way, experiments where the primary voltage is imposed have been carried out. In a complementary approach, an appropriate transformer flux feedback scheme has been also studied. (author) 6 refs.; 6 figs

  8. Role of wall implantation of charge exchange neutrals in the deuterium retention for Tore Supra long discharges

    International Nuclear Information System (INIS)

    Tsitrone, E.; Reiter, D.; Loarer, T.; Brosset, C.; Bucalossi, J.; Begrambekov, L.; Grisolia, C.; Grosman, A.; Gunn, J.; Hogan, J.; Mitteau, R.; Pegourie, B.; Ghendrih, P.; Reichle, R.; Roubin, P.

    2005-01-01

    In Tore Supra long pulses, particle balance gives evidence that a constant fraction of the injected gas (typically 50%) is retained in the wall for the duration of the shot, showing no sign of wall saturation after more than 6 min of discharge. During the discharge, the retention rate first decreases (phase 1), then remains constant throughout the pulse (phase 2). Phase 1 could be interpreted as implantation of particles combined with a constant codeposition rate, while phase 2 could correspond to codeposition alone, once the implanted surfaces are saturated with deuterium. This paper presents a possible contribution of charge exchange neutrals to the implantation process, based on modelling results with the Eirene neutral transport code. A complex pattern of particle implantation is evidenced, with saturation time constants ranging from less than one to several hundreds seconds, compatible with the experimental behaviour during phase 1

  9. Isotope exchange experiments on TEXTOR and TORE SUPRA using Ion Cyclotron Wall Conditioning and Glow Discharge Conditioning

    International Nuclear Information System (INIS)

    Wauters, T.; Douai, D.; Lyssoivan, A.; Philipps, V.; Bremond, S.; Freisinger, M.; Kreter, A.; Lombard, G.; Marchuk, O.; Mollard, P.; Paul, M.K.; Pegourie, B.; Reimer, H.; Sergienko, G.; Tsitrone, E.; Vervier, M.; Van Wassenhove, G.; Wuenderlich, D.; Van Schoor, M.; Van Oost, G.

    2011-01-01

    This contribution reports on isotope exchange studies with both Ion Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) in TEXTOR and TORE SUPRA. The discharges have been carried out in H 2 , D 2 (ICWC and GDC) and He/H 2 mixtures (ICWC). The higher reionization probability in ICWC compared to GDC, following from the 3 to 4 orders of magnitude higher electron density, leads to a lower pumping efficiency of wall desorbed species. GDC has in this analysis (5-10) times higher removal rates of wall desorbed species than ICWC, although the wall release rate is 10 times higher in ICWC. Also the measured high retention during ICWC can be understood as an effect of the high reionization probability. The use of short RF pulses (∼1 s) followed by a larger pumping time significantly improves the ratio of implanted over recovered particles, without severely lowering the total amount of removed particles.

  10. Characterization of the up-down asymmetry of density fluctuations induced by a lower modular limiter in Tore Supra

    International Nuclear Information System (INIS)

    Fenzi, C.; Devynck, P.; Garbet, X.; Antar, G.; Capes, H.; Laviron, C.; Truc, A.; Gervais, F.; Hennequin, P.; Quemeneur, A.

    1999-01-01

    In magnetic fusion devices, the effect of plasma facing components on plasma turbulence is a key issue for several reasons. Firstly, the edge turbulence controls the power deposition on plasma facing components. Secondly, the possible influence of the edge parameters on the core fluctuations is a central question, since the core turbulent transport is responsible for the confinement degradation. It is in practice difficult to determine whether the plasma core influences the edge, or the opposite. We show here that spatial edge asymmetries of density fluctuations, and particularly up-down asymmetries, provide a powerful tool to investigate this problem. In TORE SUPRA, previous scaling analyses with various plasma parameters have emphasized that a very clear effect on the asymmetry level appears when the plasma leans on the lower modular limiter located close to the measurement chord. We present here recent measurement results concerning that specific case. They tend to show that the limiter configuration has some effect on the core turbulence. (authors)

  11. Progress report of the Research Group. 1st part: Tore Supra. 2nd part: Fontenay-aux-Roses

    International Nuclear Information System (INIS)

    1981-01-01

    Three major events dominated the activities of the EURATOM/CEA association during 1980: the decision to launch the realization of the TORE SUPRA project, the progressive recognition of high frequency heating as a solution for the future, and the increasing support given to the development of heating methods and diagnostics in the JET project. It is estimated that project studies are sufficiently advanced and that industrial fabrication problems have been sufficiently covered for the realization of Tore Supra to begin in 1981. One of the successes of the work carried out is the complete validation for the superfluid helium cooling system. The satisfactory development of high frequency heating and the increasing credibility of this form of heating for future work are very important factors. In this context, the decision of the JET to envisage a large amount of ionic cyclotron heating is particularly important. The results obtained in 1980 are in fact very encouraging. The maximum power of the 500 kW T.F.R. generator was coupled with the plasma and it was possible to establish an energy Q-value. Even though the injection of neutral particles can now be considered as a proved heating method, studies of the accompanying physical phenomena are still important. The T.F.R. experiments carried out in this field in 1980 were very useful. The importance of the realization and development activities conducted during 1980, should not mask the enormous effort that made, both experimentally and theoretically, in order to understand key physical phenomena in plasma. The main peoccupation concerned small and large disruptions and all aspects of the associated instabilities. A detailed analysis of the experimental results using numerical models has led to improved empirical knowledge on the elementary transport phenomena taking place. Increasingly detailed studies on microinstabilities were also fruitful and have even led to a complete reversal in some of the ideas held about the

  12. Contributions to the design and to the fabrication of the magnet of the toroidal field of Tore Supra

    International Nuclear Information System (INIS)

    Turck, B.

    1992-03-01

    This report is a collection of published papers in French and in English about the design and the qualification of the magnet of the toroidal field of Tore Supra. The development test programme, the controls during conductor manufacturing and the acceptance tests have shown to be the bases for achieving a very low level of rejection for the whole production. A systematic study of the performances correlated to the fabrication conditions has provided valuable informations for the optimization of the manufacturing processes of superconductors. The tests of single coils have enabled the commissioning of a monitoring and protection system specially adapted for this magnet of 18 coils cooled in a superfluid helium bath. After the accident caused by an arcing in one coil of the Torus, and the replacement of the faulty coil, the monitoring and safety discharge system have been adapted. The current in the magnet has been increased up to 1 455 A for 9.3 T on the conductors (nominal values 1 400 A and 9 T). During the last three years (1989-1991) only one transition to normal state has been observed in one coil strongly irradiated after a severe plasma disruption. In these conditions the protection system acted very well and as expected

  13. Steady-state heat and particle removal with the actively cooled Phase III outboard pump limiter in Tore Supra

    International Nuclear Information System (INIS)

    Nygren, R.; Koski, J.; Lutz, T.; McGrath; Miller, J.; Watkins, J.; Guilhem, D.; Chappuis, P.; Cordier, J.; Loarer, T.

    1995-01-01

    Tore Supra's Phase III outboard pump limiter (OPL) is a modular actively-cooled mid-plane limiter, designed for heat and particle removal during long pulse operation. During its initial operation in 1993, the OPL successfully removed about 1 MW of power during ohmicly heated shots of up to 10 s duration and reached (steady state) thermal equilibrium. The particle pumping of the Phase III OPL was found to be about 50% greater than the Phase II OPL which had a radial distance between the last closed flux surface and the entrance of the pumping throat of 3.5 cm compared with only 2.5 cm for the Phase III OPL. This paper gives examples of power distribution over the limiter from IR measurements of surface temperature and from extensively calorimetry (34 thermocouples and 10 flow meters) and compares the distributions with values predicted by a 3D model (HF3D) with a detailed magnetic configuration (e.g., includes field ripple). ((orig.))

  14. Q-profile evolution and improved core electron confinement in the full current drive operation on Tore Supra

    International Nuclear Information System (INIS)

    Litaudon, X.; Peysson, Y.; Aniel, T.; Huysmans, G.; Imbeaux, F.; Joffrin, E.; Lasalle, J.; Lotte, Ph.; Schunke, B.; Segui, J.; Tresset, G.; Zabiego, M.

    2000-12-01

    The formation of a core region with improved electron confinement is reported in the recent full current drive operation of Tore Supra where the plasma current is sustained with the Lower Hybrid, LH, wave. Current profile evolution and thermal electron transport coefficients are directly assessed using the data of the new fast electron Bremsstrahlung tomography that provides the most accurate determination of the LH current and power deposition profiles. The spontaneous rise of the core electron temperature observed a few seconds after the application of the LH power is ascribed to a bifurcation towards a state of reduced electron transport. The role of the magnetic shear is invoked to partly stabilize the anomalous electron turbulence. The electron temperature transition occurs when the q-profile evolves towards a non-inductive state with a non-monotonic shape i.e. when the magnetic shear is reduced close to zero in the plasma core. The improved core confinement phase is often terminated by a sudden MHD activity when the minimum q approaches two. (authors)

  15. Spectroscopic measurements of the density and electronic temperature at the plasma edge in Tore Supra

    International Nuclear Information System (INIS)

    Lediankine, A.

    1996-01-01

    The profiles of temperature and electronic density at the plasma edge are important to study the wall-plasma interaction and the radiative layers in the Tokamak plasmas. The laser ablation technique of the lithium allows to measure the profile of electronic density. To measure the profile of temperature, it has been used for the first time, the injection of a fluorine neutral atoms beam. The experiments, the results are described in this work. (N.C.)

  16. Supravodivý tokamak dobyl Asii

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2006-01-01

    Roč. 54, č. 18 (2006), s. 58 ISSN 0040-1064 Institutional research plan: CEZ:AV0Z20430508 Keywords : superconducting tokamak * ITER * Tore Supra * Institute of Plasma Physics AV CR Subject RIV: BL - Plasma and Gas Discharge Physics

  17. Experiments on steady state particle control in Tore Supra and DIII-D

    Science.gov (United States)

    Mioduszewski, P. K.; Hogan, J. T.; Owen, L. W.; Maingi, R.; Lee, D. K.; Hillis, D. L.; Klepper, C. C.; Menon, M. M.; Thomas, C. E.; Uckan, T.; Wade, M. R.; Chatelier, M.; Grisolia, C.; Ghendrih, Ph.; Grosman, A.; Hutter, T.; Loarer, T.; Pégourié, B.; Mahdavi, M. A.; Schaffer, M.

    1995-04-01

    Particle control is playing an increasingly important role in tokamak plasma performance. The present paper discusses particle control of hydrogen/deuterium by wall pumping on graphite or carbonized surfaces, as well as by external exhaust with pumped limiters and pumped divertors. Wall pumping is ultimately a transient effect and by itself not suitable for steady state particle exhaust. Therefore, external exhaust techniques with pumped divertors and limiters are being developed. How wall pumping phenomena interact and correlate with these inherently steady state, external exhaust techniques, is not well known to date. In the present paper, the processes involved in wall pumping and in external pumping are investigated in an attempt to evaluate the effect of external exhaust on wall pumping. Some of the key elements of this analysis are: (1) charge-exchange fluxes to the wall play a crucial role in the core-wall particle dynamics, (2) the recycling fluxes of thermal molecules have a high probability of ionization in the scrape-off layer, (3) thermal particles originating from the wall, which are ionized within the scrape-off layer, can be directly exhausted, thus providing a direct path between wall and exhaust which can be used to control the wall inventory. This way, the wall can be kept in a continuous pumping state in the sense that it continuously absorbs energetic particles and releases thermal molecules which are then removed by the external exhaust mechanism. While most of the ingredients of this analysis have been observed individually before, the present evaluation is an attempt to correlate effects of wall recycling and external exhaust.

  18. Real time control of fully non-inductive operation in Tore Supra leading to 6 minutes, 1 giga-joule plasma discharges

    International Nuclear Information System (INIS)

    Van Houtte, D.; Martin, G.; Becoulet, A.; Saoutic, B.

    2004-01-01

    The experimental programme of Tore Supra (a = 0.72 m, R = 2.4 m, I p T < 4.5 T) has been devoted in 2003 to study simultaneously heat removal capability and particle exhaust in steady-state fully non-inductive current drive discharges. This required both advanced technology integration and steady-state real time plasma control. In particular, an improvement of the plasma position within a few millimetre range, and new real time cross controls between radio frequency (RF) power and various actuators built around a shared memory network, have allowed Tore Supra to access a powerful steady-state regime with an improved safety level for the actively cooled plasma facing components. Feedback controlled fully non-inductive plasma discharges have been sustained in a steady-state regime up to 6 minutes with a new world record of injected-extracted energy exceeding 1 GJ. Advanced tools, experimental results and brief physics analysis of these discharges are presented and discussed. (authors)

  19. Tore Supra: a tokamak with superconducting coils for the toroidal field

    International Nuclear Information System (INIS)

    Turck, B.

    1984-01-01

    It is under construction on the site of CEN/Cadarache for the EURATOM-CEA Association. The design has been lead by a group including teams of the DRFC of Fontenay-aux-Roses and Grenoble and the DPh/PE-STI of CEN/Saclay [fr

  20. Investigation of steady-state tokamak issues by long pulse experiments on Tore Supra

    Czech Academy of Sciences Publication Activity Database

    Giruzzi, G.; Abgrall, R.; Allegretti, L.; Ané, J.M.; Angelino, P.; Aniel, T.; Argouarch, A.; Artaud, J.F.; Balme, S.; Basiuk, V.; Bayetti, P.; Bécoulet, A.; Bécoulet, M.; Begrambekov, L.; Benkadda, M.S.; Benoit, F.; Berger-by, G.; Bertrand, B.; Beyer, P.; Blum, J.; Boilson, D.; Bottollier-Curtet, H.; Bouchand, C.; Bouquey, F.; Bourdelle, C.; Brémond, F.; Brémond, S.; Brosset, C.; Bucalossi, J.; Buravand, Y.; Cara, P.; Carpentier, S.; Casati, A.; Chaibi, O.; Chantant, M.; Chappuis, P.; Chatelier, M.; Chevet, G.; Ciazynski, D.; Ciraolo, G.; Clairet, F.; Clary, J.; Colas, L.; Corre, Y.; Courtois, X.; Crouseilles, N.; Darmet, G.; Davi, M.; Daviot, R.; De Esch, H.; Decker, J.; Decool, P.; Delchambre, E.; Delmas, E.; Delpech, L.; Desgranges, C.; Devynck, P.; Doceul, L.; Dolgetta, N.; Douai, D.; Dougnac, H.; Duchateau, J.L.; Dumont, R.; Dunand, A.; Durocher, A.; Ekedahl, A.; Elbeze, D.; Eriksson, L.G.; Escarguel, A.; Escourbiac, F.; Faisse, F.; Falchetto, G.; Farge, M.; Farjon, L.J.; Fedorczak, N.; Fenzi-Bonizec, C.; Garbet, X.; Garcia, J.; Gardarein, J.L.; Gargiulo, L.; Garibaldi, P.; Gauthier, E.; Géraud, A.; Gerbaud, T.; Geynet, M.; Ghendrih, P.; Gil, C.; Goniche, M.; Grandgirard, V.; Grisolia, C.; Gros, G.; Grosman, A.; Guigon, R.; Guilhem, D.; Guillerminet, B.; Guirlet, R.; Gunn, J.; Hacquin, S.; Hatchressian, J.C.; Hennequin, P.; Henry, D.; Hernandez, C.; Hertout, P.; Heuraux, S.; Hillairet, J.; Hoang, G.T.; Hong, S.H.; Honore, C.; Hourtoule, J.; Houry, M.; Hutter, T.; Huynh, P.; Huysmans, G.; Imbeaux, F.; Joffrin, E.; Johner, J.; Journeaux, J.Y.; Jullien, F.; Kazarian, F.; Kočan, M.; Lacroix, B.; Lamaison, V.; Lasalle, J.; Latu, G.; Lausenaz, Y.; Laviron, C.; Le Niliot, C.; Lennholm, M.; Leroux, F.; Linez, F.; Lipa, M.; Litaudon, X.; Loarer, T.; Lott, F.; Lotte, P.; Luciani, J.F.; Lütjens, H.; Macor, A.; Madeleine, S.; Magaud, P.; Maget, P.; Magne, R.; Manenc, L.; Marandet, Y.; Marbach, G.; Maréchal, J.L.; Martin, C.; Martin, V.; Martinez, A.; Martins, J.P.; Masset, R.; Mazon, D.; Meunier, L.; Meyer, O.; Million, L.; Missirlian, M.; Mitteau, R.; Mollard, P.; Moncada, V.; Monier-Garbet, P.; Moreau, D.; Moreau, P.; Nannini, M.; Nardon, E.; Nehme, H.; Nguyen, C.; Nicollet, S.; Ottaviani, M.; Pacella, D.; Pamela, S.; Parisot, P.; Parrat, H.; Pastor, P.; Pecquet, A.L.; Pégourié, B.; Petržílka, Václav; Peysson, Y.; Portafaix, C.; Prou, M.; Ravenel, N.; Reichle, R.; Reux, C.; Reynaud, P.; Richou, M.; Rigollet, F.; Rimini, F.; Roche, H.; Rosanvallon, S.; Roth, J.; Roubin, P.; Sabot, R.; Saint-Laurent, F.; Salasca, S.; Salmon, T.; Samaille, F.; Santagiustina, A.; Saoutic, B.; Sarazin, Y.; Schlosser, J.; Schneider, K.; Schneider, M.; Schwander, F.; Ségui, J.L.; Signoret, J.; Simonin, A.; Song, S.; Sonnendruker, E.; Spuig, P.; Svensson, L.; Tamain, P.; Tena, M.; Theis, J.M.; Thonnat, M.; Torre, A.; Travère, J.M.; Trier, E.; Tsitrone, E.; Turco, F.; Vallet, J.C.; Vatry, A.; Vermare, L.; Villecroze, F.; Villegas, D.; Voyer, D.; Vulliez, K.; Xiao, W.; Yu, D.; Zani, L.; Zou, X.L.; Zwingmann, W.

    2009-01-01

    Roč. 49, č. 10 (2009), s. 104010-104010 ISSN 0029-5515 Institutional research plan: CEZ:AV0Z20430508 Keywords : SOL * LH wave * plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://www.iop.org/EJ/ toc /0029-5515/49/10

  1. Tore Supra: technical description

    International Nuclear Information System (INIS)

    1985-08-01

    This report volume is devoted to system composed of vacuum vessel and thermal shields, and to poloidal field system. Mocks realized to study feasibility and structure properties are presented, mechanical calculations (with different cases of static and dynamic loads for vacuum vessel) are resumed, first wall components are described. Techniques designed to dissipate plasma power, to control recycling and maintain a good impuritiy rate are examined. Principal fabrication process are described together with pumping systems; and control of pumping circuits. For poloidal field, electrical power supplies, operation and use priciples, field coils and magnetic circuits are described [fr

  2. Results of water corrosion in static cell tests representing multi-metal assemblies in the hydraulic circuits of Tore Supra

    International Nuclear Information System (INIS)

    Lipa, M.; Blanchet, J.

    2007-01-01

    Full text of publication follows: Tore supra (TS) has used from the beginning of operation in 1989 actively cooled plasma facing components. Since the operation and baking temperature of all in vessel components has been defined to be up to 230 deg. C at 40 bars, a special water chemistry of the cooling water plant was suggested in order to avoid eventual water leaks due to corrosion (general corrosion, galvanic corrosion, stress corrosion, etc.) at relative high temperatures and pressures in tubes, pipes, bellows, water boxes, coils, etc. From the beginning of TS operation, in vessel components (e.g. wall protection panels, limiters, ergodic divertor coils, neutralisers and diagnostics) represented a unique combination of metals in the hydraulic circuit mainly such as stainless steel, Inconel, CuCrZr, Nickel and Copper. These different materials were joined together by welding (St to St, Inconel to Inconel, CuCrZr to CuCrZr and CuCrZr to St-St via a Ni sleeve adapter), brazing (St-St to Cu and Cu-LSTP), friction (CuCrZr and Cu to St-St), explosion (CuCrZr to St-St) and memory metal junction (Cryo-fit to Cu - only test sample). Following experiences obtained with steam generator tubes of nuclear power plants, a cooling water quality of AVT (all volatile treatment) has been defined based on demineralized water with adjustment of the pH value to about 9.0/ 7.0 (25 deg. C/ 200 deg. C) by addiction of ammoniac, and hydrazine in order to absorb oxygen dissolved in water. At that time, a simplified water corrosion test program has been performed using static (no circulation) test cell samples made of above mentioned TS metal combinations. All test cell samples, prepared and filled with AVT water, were performed at 280 deg. C and 65 bars in an autoclave during 3000 hours. The test cell water temperature has been chosen to be sufficient above the TS component working temperature, in order to accelerate an eventual corrosion process. Generally all above mentioned metal

  3. Results of water corrosion in static cell tests representing multi-metal assemblies in the hydraulic circuits of Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Lipa, M.; Blanchet, J. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Cellier, F. [Framatome, Centre Technique, 71 - Saint Marcel (France)

    2007-07-01

    Full text of publication follows: Tore supra (TS) has used from the beginning of operation in 1989 actively cooled plasma facing components. Since the operation and baking temperature of all in vessel components has been defined to be up to 230 deg. C at 40 bars, a special water chemistry of the cooling water plant was suggested in order to avoid eventual water leaks due to corrosion (general corrosion, galvanic corrosion, stress corrosion, etc.) at relative high temperatures and pressures in tubes, pipes, bellows, water boxes, coils, etc. From the beginning of TS operation, in vessel components (e.g. wall protection panels, limiters, ergodic divertor coils, neutralisers and diagnostics) represented a unique combination of metals in the hydraulic circuit mainly such as stainless steel, Inconel, CuCrZr, Nickel and Copper. These different materials were joined together by welding (St to St, Inconel to Inconel, CuCrZr to CuCrZr and CuCrZr to St-St via a Ni sleeve adapter), brazing (St-St to Cu and Cu-LSTP), friction (CuCrZr and Cu to St-St), explosion (CuCrZr to St-St) and memory metal junction (Cryo-fit to Cu - only test sample). Following experiences obtained with steam generator tubes of nuclear power plants, a cooling water quality of AVT (all volatile treatment) has been defined based on demineralized water with adjustment of the pH value to about 9.0/ 7.0 (25 deg. C/ 200 deg. C) by addiction of ammoniac, and hydrazine in order to absorb oxygen dissolved in water. At that time, a simplified water corrosion test program has been performed using static (no circulation) test cell samples made of above mentioned TS metal combinations. All test cell samples, prepared and filled with AVT water, were performed at 280 deg. C and 65 bars in an autoclave during 3000 hours. The test cell water temperature has been chosen to be sufficient above the TS component working temperature, in order to accelerate an eventual corrosion process. Generally all above mentioned metal

  4. Design, fabrication and testing of an improved high heat flux element, experience feedback on steady state plasma facing components in Tore Supra

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Chatelier, M.; Durocher, A.; Guilheim, D.; Lipa, M.; Mitteau, R.; Tonon, G.; Tsitrone, E.

    1998-01-01

    Actively cooled plasma facing components (PFC) have been developed and used in Tore Supra since 1985. One of the main technological problem is due to the expansion mismatch between graphite armour and metallic heat sink material. A first technology used graphite tiles with or without a reinforcement and a compliant layer, brazed with titanium copper-silver (TiCuAg) alloy. The next technology used carbon fiber material (CFC) tiles with a 2 mm pure copper compliant layer, since the good mechanical strength of the CFC allowed the reinforcement layer to be suppressed. No destructive inspection during the manufacturing procedure was found to be essential to insure a good reliability of the elements. (orig.)

  5. Electric field determination in the plasma-antenna boundary of a lower-hybrid wave launcher in Tore Supra through dynamic Stark-effect spectroscopy

    Science.gov (United States)

    Martin, E. H.; Goniche, M.; Klepper, C. C.; Hillairet, J.; Isler, R. C.; Bottereau, C.; Colas, L.; Ekedahl, A.; Panayotis, S.; Pegourie, B.; Lotte, Ph; Colledani, G.; Caughman, J. B.; Harris, J. H.; Hillis, D. L.; Shannon, S. C.; Clairet, F.; Litaudon, X.

    2015-06-01

    Interaction of radio-frequency (RF) waves with the plasma in the near-field of a high-power wave launcher is now seen to be an important topic, both in understanding the channeling of these waves through the plasma boundary and in avoiding power losses in the edge. In a recent Letter, a direct non-intrusive measurement of a near antenna RF electric field in the range of lower hybrid (LH) frequencies (ELH) was announced (2013 Phys. Rev. Lett. 110 215005). This measurement was achieved through the fitting of Balmer series deuterium spectral lines utilizing a time dependent (dynamic) Stark effect model. In this article, the analysis of the spectral data is discussed in detail and applied to a larger range of measurements and the accuracy and limitations of the experimental technique are investigated. It was found through an analysis of numerous Tore Supra discharges that good quantitative agreement exists between the measured and full-wave modeled ELH when the launched power exceeds 0.5 MW. For low power the measurement becomes inaccurate utilizing the implemented passive spectroscopic technique because the spectral noise overwhelms the effect of the RF electric field on the line profile. Additionally, effects of the ponderomotive force are suspected at sufficiently high power.

  6. Soft x-ray tomography for real-time applications: present status at Tore Supra and possible future developments

    Czech Academy of Sciences Publication Activity Database

    Mazon, D.; Vezinet, D.; Pacella, D.; Moreau, D.; Gabelieri, L.; Romano, A.; Malard, P.; Mlynář, Jan; Masset, R.; Lotte, P.

    2012-01-01

    Roč. 83, č. 6 (2012), 063505-063505 ISSN 0034-6748 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * tomography X-ray * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://rsi.aip.org/resource/1/rsinak/v83/i6/p063505_s1

  7. Electron cyclotron heating and supra-thermal electron dynamics in the TCV Tokamak

    International Nuclear Information System (INIS)

    Gnesin, S.

    2011-10-01

    This thesis is concerned with the physics of supra-thermal electrons in thermonuclear, magnetically confined plasmas. Under a variety of conditions, in laboratory as well as space plasmas, the electron velocity distribution function is not in thermodynamic equilibrium owing to internal or external drives. Accordingly, the distribution function departs from the equilibrium Maxwellian, and in particular generally develops a high-energy tail. In tokamak plasmas, this occurs especially as a result of injection of high-power electromagnetic waves, used for heating and current drive, as well as a result of internal magnetohydrodynamic (MHD) instabilities. The physics of these phenomena is intimately tied to the properties and dynamics of this supra-thermal electron population. This motivates the development of instrumental apparatus to measure its properties as well as of numerical codes to simulate their dynamics. Both aspects are reflected in this thesis work, which features advanced instrumental development and experimental measurements as well as numerical modeling. The instrumental development consisted of the complete design of a spectroscopic and tomographic system of four multi-detector hard X-ray (HXR) cameras for the TCV tokamak. The goal is to measure bremsstrahlung emission from supra-thermal electrons with energies in the 10-300 keV range, with the ultimate aim of providing the first full tomographic reconstruction at these energies in a noncircular plasma. In particular, supra-thermal electrons are generated in TCV by a high-power electron cyclotron heating (ECH) system and are also observed in the presence of MHD events, such as sawtooth oscillations and disruptive instabilities. This diagnostic employs state-of-the-art solid-state detectors and is optimized for the tight space requirements of the TCV ports. It features a novel collimator concept that combines compactness and flexibility as well as full digital acquisition of the photon pulses, greatly

  8. Heat flux distribution and gyro-radius smoothing effect on misaligned CFC tile in the Tore Supra tokamak

    Czech Academy of Sciences Publication Activity Database

    Corre, Y.; Dejarnac, Renaud; Gardarein, J.-L.; Gaspar, J.; Escourbiac, F.; Gauthier, E.; Gunn, J. P.; Komm, Michael; Lipa, M.; Loarer, T.; Missirlian, M.; Rigollet, F.

    2015-01-01

    Roč. 463, August (2015), s. 832-836 ISSN 0022-3115. [PLASMA-SURFACE INTERACTIONS 21: International Conference on Plasma-Surface Interactions in Controlled Fusion Devices. Kanazawa, 26.05.2014-30.05.2014] Institutional support: RVO:61389021 Keywords : Heat loads * IR thermography * misalignment * limiter Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015 http://www.sciencedirect.com/science/article/pii/S002231151400720X

  9. TORE SUPRA neutral injection system

    International Nuclear Information System (INIS)

    Bayetti, P.; Becherer, R.; Bottiglioni, F.; Jacquot, C.; Jequier, F.; Fumelli, M.; Lotte, P.; Pamela, J.; Sledziewski, Z.

    1989-01-01

    The Neutral Beam Injection on TS consists of three boxes, each housing two injectors. Each of them is designed to accelerate 40 A at 100 kV in D 2 ar 40A at 80 kV in H 2 , in shots of 30 s. The power on the plasma is expected to be 7.5 MW (full energy) and 9 MW (total) for the D-beams; 2.5 MW and 3.8 MW for H-beams. This for a species mix of 0.85 19 m -2 and a transmission efficiency of 0.7. Four injectors are in co and two through another one, and they are tilted of 20 deg from the radial direction. Injectors are designed for the energy recovery of the full energy ions. A prototype line, operated in 1987-88, has given the imput for the demensioning of the present system.First injectors are expected to be operational by the end of 1988. The present contribution describes the injection boxes, injection line, magnetic shielding, electrical circuity and power supplies, control and data acquisition systems, and the Fast Interlock Safety System (FISS). 7 refs.; 6 figs

  10. Recent results on Tore Supra

    International Nuclear Information System (INIS)

    Moreau, D.

    1995-01-01

    Recent results regarding heating, confinement, current drive and profile modifications, heat and particle exhaust are reported. Improved core confinement is obtained after pellet injection (PEP) or Lower Hybrid current drive (LHEP) and may be linked with small - or reversed - central magnetic shear. Conversely, by increasing the magnetic shear in the gradient region, both LHCD and fast wave electron heating (FWEH) have produced improved global confinement was carried by the bootstrap current. Fast wave current drive has been observed at the level of 80 kA in a 0.4 MA discharge. In the ergodic divertor configuration, stable radiative layers were obtained with neon injection. At least 80% of a total of 7 MW injected power were radiated without confinement degradation or impurity accumulation. Finally, the heat exhaust capability of the various actively cooled plasma facing components is briefly described. (author) 14 refs.; 13 figs

  11. Development of high thermal flux components for continuous operation in Tokamaks

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Coston, J.F.; Deschamps, P.; Lipa, M.

    1991-01-01

    High heat flux plasma facing components are under development and appropriate experimental evaluations have been carried out in order to operate during cycles of several hundred seconds. In Tore Supra, a large tokamak with a plasma nominal duration in excess of 30 seconds, solutions are tested that could be later applied to the NET/ITER tokamak, where peaked heat flux values of 15 MW/m 2 on the divertor plates are foreseen. The proposed concept is a swirl square tube design protected with brazed CFC flat tiles. Development programs and validation tests are presented. The tests results are compared with calculations

  12. Measurement of plasma current in Tokamaks using an optical fibre reflectometry technique

    Directory of Open Access Journals (Sweden)

    Wuilpart Marc

    2018-01-01

    Full Text Available An optical time-domain reflectometer sensitive to the polarization of light is proposed for the measurement of plasma current in the Tore Supra fusion reactor. The measurement principle relies on the Faraday effect i.e. on the generation of a circular birefringence along an optical fiber subject to an axial magnetic field. The circular birefringence induces a polarization rotation that can be mapped along the fiber thanks to an opticaltime domain reflectometer followed by an linear polarizer. A proper fitting of the measurement trace then allows determining the applied plasma current. The sensor has been experimentally validated on the Tore Supra tokamak fusion reactor for a plasma current range going from 0.6 to 1.5 MA. A maximum error of 13.50% has been observed for the lowest current.

  13. Modelling and control of a tokamak plasma

    International Nuclear Information System (INIS)

    Bremond, S.

    1995-01-01

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes

  14. Stability analysis of tokamak plasmas

    International Nuclear Information System (INIS)

    Bourdelle, C.

    2000-10-01

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  15. Runaway-ripple interaction in Tokamaks

    International Nuclear Information System (INIS)

    Laurent, L.; Rax, J.M.

    1989-08-01

    Two approaches of the interaction between runaway electrons and the ripple field, in tokamaks, are discussed. The first approach considers the resonance effect as an intense cyclotron heating of the electrons, by the ripple field, in the guiding center frame of the fast particles. In the second approach, an Hamiltonian formalism is used. A criterion for the onset of chaotic behavior and the results are given. A new universal instability of the runaway population in tokamak configuration is found. When combined with cyclotron losses one of its major consequence is to act as an effective slowing down mechanism preventing the free fall acceleration toward the synchrotron limit. This configuration allows the explanation of some experimental results of Tore Supra and Textor

  16. Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S.

    1995-10-18

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.

  17. Magnetic ripple and the modeling of lower-hybrid current drive in tokamaks

    International Nuclear Information System (INIS)

    Peysson, Y.; Arslanbekov, R.; Basiuk, V.; Carrasco, J.; Litaudon, X.; Moreau, D.; Bizarro, J.P.

    1996-01-01

    Using ray-tracing, a detailed investigation of the lower hybrid (LH) wave propagation in presence of toroidal magnetic field ripple is presented. By coupling ray tracing with a one-dimensional relativistic Fokker-Planck code, simulations of LH experiments have been performed for the Tore Supra tokamak. Taking into account magnetic ripple in LH simulations, a better agreement is found between numerical predictions and experimental observations, such as non-thermal Bremsstrahlung emission, current profile, ripple-induced power losses in local magnetic mirrors, when plasma conditions correspond to the ' 'few passes' regime. (author)

  18. Stability analysis of tokamak plasmas; Analyse de stabilite de plasmas de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bourdelle, C

    2000-10-01

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  19. New Tore Supra steady state operating scenario

    International Nuclear Information System (INIS)

    Martin, G.; Parlange, F.; van Houtte, D.; Wijnands, T.

    1995-01-01

    This document deals with plasma control in steady state conditions. A new plasma control systems enabling feedback control of global plasma equilibrium parameters has been developed. It also enables to operate plasma discharge in steady state regime. (TEC). 4 refs., 5 figs

  20. Tokamak

    International Nuclear Information System (INIS)

    Meglicki, Z.

    1995-01-01

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs

  1. Experimental investigation of turbulent transport at the edge of a tokamak plasma

    International Nuclear Information System (INIS)

    Fedorczak, N.

    2010-01-01

    This manuscript is devoted to the experimental investigation of particle transport in the edge region of the tokamak Tore Supra. The first part introduces the motivations linked to energy production, the principle of a magnetic confinement and the elements of physics essential to describe the dynamic of the plasma at the edge region. From data collected by a set of Langmuir probes and a fast visible imaging camera, we demonstrate that the particle transport is dominated by the convection of plasma filaments, structures elongated along magnetic field lines. They present a finite wave number, responsible for the high enhancement of the particle flux at the low field side of the tokamak. This leads to the generation of strong parallel flows, and the strong constraint of filament geometry by the magnetic shear. (author)

  2. Experience with high heat flux components in large tokamaks

    International Nuclear Information System (INIS)

    Chappuis, P.; Dietz, K.J.; Ulrickson, M.

    1991-01-01

    The large present day tokamaks. i.e.JET, TFTR, JT-60, DIII-D and Tore Supra are machines capable of sustaining plasma currents of several million amperes. Pulse durations range from a few seconds up to a minute. These large machines have been in operation for several years and there exists wide experience with materials for plasma facing components. Bare and coated metals, bare and coated graphites and beryllium were used for walls, limiters and divertors. High heat flux components are mainly radiation cooled, but stationary cooling for long pulse duration is also employed. This paper summarizes the experience gained in the large machines with respect to material selection, component design, problem areas, and plasma performance. 2 tabs., 26 figs., 50 refs

  3. Soft X-Ray measurements and analysis on Tokamaks in view of real-time control

    International Nuclear Information System (INIS)

    Vezinet, Didier

    2013-01-01

    This thesis focuses on measuring and interpreting the Soft X-Ray (SXR) radiation (approximately [1 keV; 15 keV]) in Tokamaks. As explained in Chapter 2, this radiation conveys information about the plasma density, temperature, magnetic equilibrium and impurity content. However, the measured data is spectrally and spatially-integrated and results from several physical phenomena affecting every ion species. Tore Supra's SXR diagnostics is based on semiconductor diodes presented in Chapter 3, along with a new gas detector successfully tested in laboratory and on Tore Supra. A new methodology for absolute spectral characterisation of photo detectors using a portable SXR tube is presented. Tomographic inversion algorithms, that grant access to reconstructions of the SXR emissivity field in a poloidal cross-section, are presented in Chapter 4. Improvements implemented on one particular algorithm are detailed with examples of application. A comparison between the position of the SXR emissivity maximum and the magnetic axis reconstructed by an equilibrium code is presented in Chapter 5. Chapter 6 presents an approach used to derive an impurity density from its SXR emissivity using the robustness of its SXR cooling factor with respect to impurity transport. The physics accounting for this robustness is studied and a first map of the domain of validity of this method is provided. Chapter 7 addresses poloidal asymmetries of the SXR emissivity field. Two types of asymmetries are presented as well as experiments conducted on ASDEX-U to verify their parametric dependences. A new type of SXR asymmetry, observed on Tore Supra is introduced. (author) [fr

  4. Design considerations of modular pump limiters for large tokamaks

    International Nuclear Information System (INIS)

    Uckan, T.; Klepper, C.C.; Mioduszewski, P.K.

    1987-11-01

    Long-pulse (>10-s) and high-power (>10-MW) operation of large tokamaks requires multiple limiter modules for particle and heat removal, and the power load must be distributed among a number of modules. Because each added module changes the performance of all the others, a set of design criteria must be defined for the overall limiter system. The relationship between individual modules must also be considered from the standpoint of flux coverage and shadowing effects. This paper addresses these issues and provides design guidelines. Parameters of the individual modules are then determined from the system requirements for particle and power removal. Long-pulse operation of large tokamaks requires that the limiter modules be equipped with active cooling. At the leading edge of a module, the cooling channel determines the thickness of the limiter blade (or head). A model has been developed for estimating the system exhaust efficiency in terms of the parameters of the leading edge (i.e., its thickness and the design heat flux) in terms of given device parameters and the power load that must be removed. The impact on module design of state-of-the-art engineering technology for high heat removal is discussed. The choice of locations for the modules is also investigated, and the effects of shadowing between modules on particle and power removal are examined. The results are applied to the Tore Supra tokamak. Conceptual design parameters of the modular pump limiter system are given. 10 refs., 5 figs

  5. Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

    Science.gov (United States)

    Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

    2017-10-01

    The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

  6. Using plasma waves to create in tokamaks the necessary quasi-stationary conditions for controlled fusion

    International Nuclear Information System (INIS)

    Moreau, D.

    1993-04-01

    It is studied, on the one hand, how using hybrid waves with frequency near from lower hybrid frequency in fusion plasma. Works about coupling waves in plasma (chap.I), their propagation and response of the plasma to the absorption of the waves (chap.II). This method is the most effective until today. Because of limits, it has been investigated, on the other hand, fast magnetosonic wave to control current density in the centre of the discharge in a reactor or a very hot plasma. Theoretical study (chap.III) and experimental results (chap.IV) are presented. Experiments are in progress or planned in following tokamaks: D3-D (USA), JET (Europe), TORE SUPRA (France), JT-60 (Japan). figs. refs. tabs

  7. Monitoring of density in tokamaks: pumping and gas injection

    International Nuclear Information System (INIS)

    Dejarnac, R.

    2002-11-01

    In thermonuclear fusion devices, controlling the Deuterium-Tritium fuel density and exhausting the Helium ashes is a crucial point. This is achieved by fuelling the discharges by different methods (gas puffing and pellet injection are the most commonly used) and by implementing pumping devices at the plasma periphery. These two issues are treated in this work, both from an experimental and a modelling point of view, using the neutral transport code EIRENE as main tool for our studies. As far as pumping is concerned, we have modelled the outboard pump limiter of the Tore Supra tokamak with the EIRENE code to which we coupled a plasma module specially developed to simulate the neutrals and the plasma in a coherent way. This allowed to validate the code against experimental data. As far as plasma fuelling is concerned, we present here an original method: the supersonic pulsed gas injection (SPGI). This intermediate method between conventional gas puff (GP) and pellet injection was designed and tested at Tore Supra. It consists of injecting very dense and short gas puffs at high speed into the plasma. Experimentally, SPGI was found to have a better fuelling efficiency than GP and to lead to a strong plasma cooling. The mechanisms responsible for this improved efficiency are analysed by modelling, using the EIRENE code to determine the ionisation source and a 1 D transport model to reproduce the plasma density response. At last, an extrapolation of the present injector is presented, discussing the possibility to obtain a radial drift of the injected matter as observed in the case of high field side pellet injection. (author)

  8. Coupled two-dimensional edge plasma and neutral gas modeling of tokamak scrape-off-layers

    International Nuclear Information System (INIS)

    Maingi, R.

    1992-08-01

    The objective of this study is to devise a detailed description of the tokamak scrape-off-layer (SOL), which includes the best available models of both the plasma and neutral species and the strong coupling between the two in many SOL regimes. A good estimate of both particle flux and heat flux profiles at the limiter/divertor target plates is desired. Peak heat flux is one of the limiting factors in determining the survival probability of plasma-facing-components at high power levels. Plate particle flux affects the neutral flux to the pump, which determines the particle exhaust rate. A technique which couples a two-dimensional (2-D) plasma and a 2-D neutral transport code has been developed (coupled code technique), but this procedure requires large amounts of computer time. Relevant physics has been added to an existing two-neutral-species model which takes the SOL plasma/neutral coupling into account in a simple manner (molecular physics model), and this model is compared with the coupled code technique mentioned above. The molecular physics model is benchmarked against experimental data from a divertor tokamak (DIII-D), and a similar model (single-species model) is benchmarked against data from a pump-limiter tokamak (Tore Supra). The models are then used to examine two key issues: free-streaming-limits (ion energy conduction and momentum flux) and the effects of the non-orthogonal geometry of magnetic flux surfaces and target plates on edge plasma parameter profiles

  9. Study of lower hybrid wave propagation and absorption in a tokamak plasma using hard X-Ray tomography

    International Nuclear Information System (INIS)

    Imbeaux, F.

    1999-01-01

    Control of the current density profile is a critical issue in view to obtain high fusion performances in tokamak plasmas? It is therefore important to be able to control the power deposition profile of the lower hybrid wave, which has the highest current drive efficiency among all other non-inductive additional methods. Propagation and absorption of this wave are investigated in the Tore Supra tokamak using a new hard x-ray tomographic system and a new ray-tracing/Fokker-Planck code. These tools are described in detail and allow to analyse the lower hybrid power deposition profile dependence as a function of various plasma parameters (density, magnetic field, current) and of the injected wave spectrum. A good agreement between the code and the measurements found when the central electron temperature is greater than about 3 keV, that is in regimes where the wave undergoes only a few reflections before being absorbed. The simulations are then used to interpret the experimental trends. The lower hybrid power deposition profile is in nearly all discharges localised at a normalised minor radius of 0.2-0.3, and is weakly sensitive to variations of plasma parameters. It is hence difficult to perform an efficient control of the current profile generated by the lower hybrid wave in Tore Supra. This goal may nevertheless be reached by using an original method, which uses an auxiliary lower hybrid wave injected by a vertical port of the torus. This method is investigated by means of the simulation code. (author)

  10. Monitoring of density in tokamaks: pumping and gas injection; Controle de la densite dans les tokamaks: pompage et injection de matiere

    Energy Technology Data Exchange (ETDEWEB)

    Dejarnac, R

    2002-11-01

    In thermonuclear fusion devices, controlling the Deuterium-Tritium fuel density and exhausting the Helium ashes is a crucial point. This is achieved by fuelling the discharges by different methods (gas puffing and pellet injection are the most commonly used) and by implementing pumping devices at the plasma periphery. These two issues are treated in this work, both from an experimental and a modelling point of view, using the neutral transport code EIRENE as main tool for our studies. As far as pumping is concerned, we have modelled the outboard pump limiter of the Tore Supra tokamak with the EIRENE code to which we coupled a plasma module specially developed to simulate the neutrals and the plasma in a coherent way. This allowed to validate the code against experimental data. As far as plasma fuelling is concerned, we present here an original method: the supersonic pulsed gas injection (SPGI). This intermediate method between conventional gas puff (GP) and pellet injection was designed and tested at Tore Supra. It consists of injecting very dense and short gas puffs at high speed into the plasma. Experimentally, SPGI was found to have a better fuelling efficiency than GP and to lead to a strong plasma cooling. The mechanisms responsible for this improved efficiency are analysed by modelling, using the EIRENE code to determine the ionisation source and a 1 D transport model to reproduce the plasma density response. At last, an extrapolation of the present injector is presented, discussing the possibility to obtain a radial drift of the injected matter as observed in the case of high field side pellet injection. (author)

  11. Coupling structure calculations for ion cyclotron heating of Tore Supra

    International Nuclear Information System (INIS)

    Bannelier, P.

    1986-12-01

    Two structures are studied: antennas and waveguides. After some recalls on transmission lines with losses, the theory is applied to antennas with inner adaptation: the problem is to calculate the impedance necessary for complete adaptation of antenna to the power line and the generator. The Faraday screen role is detailed and studied: the per-unit length loss resistance due to ohmic losses in the screen which lower the plasma-coupled maximum power. Waveguide coupling theory is also presented. Coupling between wave guide and plasma is evaluated [fr

  12. Plasma turbulence measured by fast sweep reflectometry on Tore Supra

    International Nuclear Information System (INIS)

    Clairet, F.; Vermare, L.; Leclert, G.

    2004-01-01

    Traditionally devoted to electron density profile measurement we show that fast frequency sweeping reflectometry technique can bring valuable and innovative measurements onto plasma turbulence. While fast frequency sweeping technique is traditionally devoted to electron density radial profile measurements we show in this paper how we can handle the fluctuations of the reflected signal to recover plasma density fluctuation measurements with a high spatial and temporal resolution. Large size turbulence related to magneto-hydrodynamic (MHD) activity and the associated magnetic islands can be detected. The radial profile of the micro-turbulence, which is responsible for plasma anomalous transport processes, is experimentally determined through the fluctuation of the reflected phase signal. (authors)

  13. Plasma turbulence measured by fast sweep reflectometry on Tore Supra

    International Nuclear Information System (INIS)

    Clairet, F.; Vermare, L.; Heuraux, S.; Leclert, G.

    2004-01-01

    Traditionally devoted to electron density profile measurement we show that fast frequency sweeping reflectometry technique can bring valuable and innovative measurements onto plasma turbulence. While fast frequency sweeping technique is traditionally devoted to electron density radial profile measurements we show in this paper how we can handle the fluctuations of the reflected signal to recover plasma density fluctuation measurements with a high spatial and temporal resolution. Large size turbulence related to magneto-hydrodynamic (MHD) activity and the associated magnetic islands can be detected. The radial profile of the micro-turbulence, which is responsible for plasma anomalous transport processes, is experimentally determined through the fluctuation of the reflected phase signal

  14. Recent results on current profile shaping on tore supra

    Energy Technology Data Exchange (ETDEWEB)

    Becoulet, A.

    1994-12-31

    The link between the current profile and the confinement is studied, involving various regimes: high power minority ion cyclotron resonant heating, high power lower hybrid current drive, fast wave direct electron heating and current drive and pellet enhanced performance. It is shown that the electron heat diffusivity decreases when the magnetic shear increases in the confinement zone and/or when it decreases in the plasma centre. (authors). 13 refs., 6 figs.

  15. RF heating and current drive in Tore Supra

    International Nuclear Information System (INIS)

    Litaudon, X.

    1995-01-01

    Recent RF heating and current drive experiments in the Lower Hybrid (LH) and Ion Cyclotron (IC) frequency ranges are reported. In the 4T improved confinement LHEP regime, steady-state LHCD operation has been realized with a new ''constant-flux'' scenario. A new, reversed shear, 2T improved confinement plasma regime has also been investigated when the core plasma is inaccessible to the LH waves. Stable, LH driven 0.4 MA discharges were obtained with H rlw = 2 at βp = 0.8, q o above 2 and with a reduced electron thermal diffusivity in the central reversed shear region. Efficient direct coupling of the fast magnetosonic wave to the electrons for heating and current drive is observed during 48 MHz/2T operation. Fast wave electron heating has produced improved confinement with H rlw = 2 at βp = 1.6, and a bootstrap current fraction up to 45%. Fast wave current drive has been observed at the level of 80 kA in a 0.4 MA discharge. (authors). 28 refs., 7 figs

  16. Carbon migration and deuterium retention in Tore Supra

    International Nuclear Information System (INIS)

    Panayotis, Stephanie

    2013-01-01

    Three reasons can be invoked to characterize and control plasma-surface interaction in thermonuclear fusion devices: 1/ the plasma erosion limits the lifetime of the first wall components, 2/ the penetration of eroded particles in the plasma is the cause of fuel dilution and loss of performance and 3/ part of the fuel (D/T) is trapped in the wall or layers resulting from the redeposition of eroded particles. In carbon wall devices, points 1/ and 3/ are strongly coupled due to the chemical affinity of carbon with hydrogen or its isotopes. If the erosion/redeposition balance is often obtained from post-mortem analyses of samples extracted from the vacuum chamber, two methods are currently used to build the fuel balance, and particularly to quantify the amount of which trapped in the vessel: the post-mortem analyses cited above, and discharge per discharge gas balance. Estimations by these two methods exhibit a significant discrepancy, the amount of trapped hydrogen estimated by post-mortem being typically four times lower than that obtained from gas balance. The main reason is that the former value is resolved in space (it depends of the location of the sample in the vacuum vessel) but integrated in time (it concerns the whole period during which the sample was in the device), when the latter is a global value for the whole machine but is resolved discharge per discharge. For solving the discrepancy, one must perform post-mortem analyses on a number of samples large enough for covering the whole vessel and extend gas balance measurements to the whole period during which the considered first wall elements were used, including the period in between plasmas and vents. (author) [fr

  17. Monitoring of the current profile by using cyclotronic electron waves in tokamaks; Controle du profil de courant par ondes cyclotroniques electroniques dans les tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Dumont, R

    2001-08-01

    The subject of this thesis is the study of the cyclotronic electron wave as a monitoring tool of the current profile. The first chapter is dedicated to basic notions concerning tokamak plasmas and current generation. The second chapter is centered on the use of fast electrons to generate current and on its modelling. The propagation and absorption of the cyclotronic electron wave require a specific polarization state whose characteristics must be carefully chosen according to some parameters of the discharge, the chapter 3 deals with this topic. The absorption of a wave in a plasma depends greatly on the velocity distribution of the particles that make up the plasma and this distribution is constantly modified by the energy of the wave, so this phenomenon is non-linear and its physical description is difficult. In a case of a fusion plasma, a sophisticated approximation called quasi-linear theory can be applied with some restrictions that are presented in chapter 4. Chapters 5 and 6 are dedicated to kinetics scenarios involving the low hybrid wave and the cyclotronic electron wave inside the plasma. Some experiments dedicated to the study of the cyclotronic electron wave have been performed in Tore-supra (France) and FTU (Italy) tokamaks, they are presented in the last chapter. (A.C.)

  18. Tokamak electron heat transport by direct numerical simulation of small scale turbulence

    International Nuclear Information System (INIS)

    Labit, B.

    2002-10-01

    electron normalized Larmor has been emphasized: the confinement time is inverse proportional to this parameter. Finally, the low dependence of turbulent transport with the magnetic shear and the inverse aspect ratio is also reported. Although the transport level observed in the simulations is low compared to the experiments, we have tried a direct confrontation with Tore Supra results. This tokamak is well designed to study the electron heat transport. Keeping most of the parameters from a well referenced Tore Supra shot, the nonlinear simulation gives a threshold quite close to the experimental one. The observed turbulent conductivity is a factor fifty lower than the experimental one. An important parameter can not be matched: the normalized Larmor radius, ρ * . This limitation has to be overcome in order to confirm this results. Finally, a rigorous confrontation between this result and gyrokinetic simulations has to conclude that the ETG instability cannot describe electron heat loses in tokamaks. (author)

  19. The division of plasma physics

    International Nuclear Information System (INIS)

    Evans, T.E.; Guilhem, D.; Klepper, C.C.

    1990-07-01

    The investigations presented in the 31th meeting on plasma physics were: the main results and observations during the ergodic divertor experiments in Tore Supra tokamak; the modifications of power scrape-off-length and power deposition during various configurations in Tore Supra plasmas; the results of pressure measurements and particle fluxes in the Tore Supra pump limiter

  20. Controlled thermonuclear reactions and Tora Supra program

    International Nuclear Information System (INIS)

    1988-01-01

    The research programs for the nuclear energy production by means of thermonuclear fusion are shown. TORA SUPRA, Joint European Torus, Next European Torus and those developed at the Atomic Energy Center are described. The controlled fusion necessary conditions, the energy and confinement balance, and the research of a better tokamak configuration are discussed. A description of TORA SUPRA, the ways of achieving the project and the expected delays are shown. The Controlled Fusion Research Department functions, concerning these programs, are described. The importance of international cooperation and the perspectives about the use of controlled fusion are underlined [fr

  1. Development of coupling systems at the hybrid frequency for the non-inductive current generation inside a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berio, S. [Association Euratom-CEA, Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Aix-Marseille-1 Univ., 13 - Marseille (France)

    1996-12-31

    Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead to the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a on-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (authors) 53 refs.

  2. Measurement of local, internal magnetic fluctuations via cross-polarization scattering in the DIII-D tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Barada, K., E-mail: kshitish@ucla.edu; Rhodes, T. L.; Crocker, N. A.; Peebles, W. A. [University of California-Los Angeles, P.O. Box 957099, Los Angeles, California 90095 (United States)

    2016-11-15

    We present new measurements of internal magnetic fluctuations obtained with a novel eight channel cross polarization scattering (CPS) system installed on the DIII-D tokamak. Measurements of internal, localized magnetic fluctuations provide a window on an important physics quantity that we heretofore have had little information on. Importantly, these measurements provide a new ability to challenge and test linear and nonlinear simulations and basic theory. The CPS method, based upon the scattering of an incident microwave beam into the opposite polarization by magnetic fluctuations, has been significantly extended and improved over the method as originally developed on the Tore Supra tokamak. A new scattering geometry, provided by a unique probe beam, is utilized to improve the spatial localization and wavenumber range. Remotely controllable polarizer and mirror angles allow polarization matching and wavenumber selection for a range of plasma conditions. The quasi-optical system design, its advantages and challenges, as well as important physics validation tests are presented and discussed. Effect of plasma beta (ratio of kinetic to magnetic pressure) on both density and magnetic fluctuations is studied and it is observed that internal magnetic fluctuations increase with beta. During certain quiescent high confinement operational regimes, coherent low frequency modes not detected by magnetic probes are detected locally by CPS diagnostics.

  3. ECRH and electron heat transport in tokamaks

    International Nuclear Information System (INIS)

    Zou, X.L.; Giruzzi, G.; Dumont, R.J.

    2003-01-01

    non- local. It can be concluded that the profile resilience mainly results from two effects: the first one is that the lower order Eigenmode are more favored than the higher order; the second one (volume effect) is that the central source (ohmic heating) is favored with respect to the off-axis source (ECRH) in the contribution to the temperature profile shape. We emphasize that the resilience effect on the temperature profile is a basic and natural property of the diffusion equation in cylindrical geometry. All additional effects, as the heat pinch, critical gradient, etc, can reinforce this resilience. Finally, this analytical solution has been used with success for the determination of the transport coefficient and the polarization of the EC waves during ECRH experiments in the Tore Supra tokamak. (authors)

  4. Turbulence intermittency and burst properties in tokamak scrape-off layer

    International Nuclear Information System (INIS)

    Antar, G.Y.; Devynck, P.; Garbet, X.; Luckhardt, S.C.

    2001-01-01

    Density fluctuation measured by a reciprocating Langmuir probe on the Tore Supra tokamak [Garbet et al., Nucl. Fusion 32, 2147 (1992)] is investigated. The purpose of this article is to give a rather comprehensive analysis of intermittency by using several data analyses to compare the bursts properties to that of coherent structures and avalanches. The probability distribution of the density fluctuations is found positively skewed, while a Gaussian shape for the negative values is recorded. It is shown that the fluctuation spectra possess one scaling region with a power law close to the one predicted by a Kolmogorov-Kraichnan model in the inverse cascade subrange. However, a net deviation from this law at higher moment orders is demonstrated. The deviation from the mono-fractal model is investigated by the multifractal analysis that reveals the variety of the dissipative structures similar to what is found in fully developed fluid turbulence. The spectra are found asymmetric, indicating the presence of structures not generated by a multiplicative process. Using conditional analysis, a detailed study of the intermittent bursts independently of the background is performed. The typical form of the intermittent structures is asymmetric. Furthermore, they do not conserve mass for only positive density fluctuations are recorded. Their poloidal velocity is estimated to be 70% greater than the background turbulence, suggesting that they may not result from a diffusive process

  5. LIBS for tokamak plasma facing components characterisation: Perspectives on in situ tritium cartography

    Energy Technology Data Exchange (ETDEWEB)

    Semerok, A., E-mail: alexandre.semerok@cea.fr [CEA, DEN, DPC/SEARS/LISL, F-91191 Gif-sur-Yvette (France); Grisolia, C. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)

    2013-08-21

    Feasibility of in situ LIBS remote measurements with the plasma facing components (PFCs) from the European tokamaks (TORE SUPRA, CEA Cadarache, France and TEXTOR, Julich, Germany) has been studied in laboratory using Q-switched nanosecond Nd–YAG lasers. LIBS particular properties and optimal parameters were determined for in-depth PFCs characterisation. The LIBS method was in situ tested on the Joint European Torus (JET) in the UK with the EDGE LIDAR Laser System (Ruby laser, 3 J, 690 nm wavelength, 300 ps pulse duration, intensity up to 70 GW/cm{sup 2}). Several analytical spectral lines of H, CII, CrI, and BeII in plasma were observed and identified in 400–600 nm spectral range with the optimised LIBS and detection system. The LIBS in-depth cartography is in agreement with the surface properties of the tile under analysis, thus confirming feasibility of in situ LIBS. Further LIBS technique improvements required to provide tritium concentration measurements more accurately are discussed.

  6. LIBS for tokamak plasma facing components characterisation: Perspectives on in situ tritium cartography

    International Nuclear Information System (INIS)

    Semerok, A.; Grisolia, C.

    2013-01-01

    Feasibility of in situ LIBS remote measurements with the plasma facing components (PFCs) from the European tokamaks (TORE SUPRA, CEA Cadarache, France and TEXTOR, Julich, Germany) has been studied in laboratory using Q-switched nanosecond Nd–YAG lasers. LIBS particular properties and optimal parameters were determined for in-depth PFCs characterisation. The LIBS method was in situ tested on the Joint European Torus (JET) in the UK with the EDGE LIDAR Laser System (Ruby laser, 3 J, 690 nm wavelength, 300 ps pulse duration, intensity up to 70 GW/cm 2 ). Several analytical spectral lines of H, CII, CrI, and BeII in plasma were observed and identified in 400–600 nm spectral range with the optimised LIBS and detection system. The LIBS in-depth cartography is in agreement with the surface properties of the tile under analysis, thus confirming feasibility of in situ LIBS. Further LIBS technique improvements required to provide tritium concentration measurements more accurately are discussed

  7. Approximation in generalized Hardy classes and resolution of inverse problems for tokamaks

    International Nuclear Information System (INIS)

    Fisher, Y.

    2011-11-01

    This thesis concerns both the theoretical and constructive resolution of inverse problems for isotropic diffusion equation in planar domains, simply and doubly connected. From partial Cauchy boundary data (potential, flux), we look for those quantities on the remaining part of the boundary, where no information is available, as well as inside the domain. The proposed approach proceeds by considering solutions to the diffusion equation as real parts of complex valued solutions to some conjugated Beltrami equation. These particular generalized analytic functions allow to introduce Hardy classes, where the inverse problem is stated as a best constrained approximation issue (bounded extrema problem), and thereby is regularized. Hence, existence and smoothness properties, together with density results of traces on the boundary, ensure well-posedness. An application is studied, to a free boundary problem for a magnetically confined plasma in the tokamak Tore Supra (CEA Cadarache France). The resolution of the approximation problem on a suitable basis of functions (toroidal harmonics) leads to a qualification criterion for the estimated plasma boundary. A descent algorithm makes it decrease, and refines the estimations. The method does not require any integration of the solution in the overall domain. It furnishes very accurate numerical results, and could be extended to other devices, like JET or ITER. (author)

  8. Experiments on electron temperature profile resilience in FTU tokamak with continuous and modulated ECRH

    International Nuclear Information System (INIS)

    Cirant, S.

    2002-01-01

    Experiments performed on FTU tokamak, aiming at validation of physics-based transport models of the electron temperature profile resilience, are presented. ECRH is used to probe transport features, both in steady-state and in response to perturbations, while ECCD and LHCD are used for current density profile shaping. Observed confinement behaviour shows agreement with a critical temperature gradient length modelling. Central, low gradient plasma is characterized by low stiffness and low electron thermal diffusivity. Strong stiffness and high conduction are found in the confinement region. Resilience is experimentally characterized by an index of the resistance of the profile to adapt its shape to localized ECRH, while the diffusivity and its low-high transition are measured both by power balance and heat pulse propagation analysis. A particular attention is given to the investigation of the transition layer between low-high diffusivity and low-high stiffness regions. A dependence of LTc on magnetic shear, similar to what found in Tore Supra, and consistent with ETG based anomalous transport, is found. (author)

  9. Tokamak WEST připraven ke startu!

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    Květen (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * WEST * Tora Supra * divertor Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/2014-tokamak-west-pripraven-ke- start u

  10. Intuitionistic supra fuzzy topological spaces

    International Nuclear Information System (INIS)

    Abbas, S.E.

    2004-01-01

    In this paper, We introduce an intuitionistic supra fuzzy closure space and investigate the relationship between intuitionistic supra fuzzy topological spaces and intuitionistic supra fuzzy closure spaces. Moreover, we can obtain intuitionistic supra fuzzy topological space induced by an intuitionistic fuzzy bitopological space. We study the relationship between intuitionistic supra fuzzy closure space and the intuitionistic supra fuzzy topological space induced by an intuitionistic fuzzy bitopological space

  11. Use of plasma waves to create in Tokamaks quasi-stationary conditions required for controlled fusion

    International Nuclear Information System (INIS)

    Moreau, D.

    1993-04-01

    In this thesis are studied the coupling of hybrid waves to the plasma, multijunction antennas, hybrid wave stochastic propagation, fast wave current drive and lower-hybrid current drive experiments in Tore Supra and Jet. The possibility of decoupling current density profile and temperature give one more degree of freedom for the control of plasma in a configuration which is not very flexible

  12. Control of particles flux in a tokamak with an events structure

    International Nuclear Information System (INIS)

    Tsitrone, E.

    1995-01-01

    Two key problems in the development of a controlled fusion reactor are: -the control of the ashes resulting from the fusion reaction (helium) and of the impurities coming from the wall erosion, which affect the central plasma performances by diluting the fuel and dissipating a part of the produced energy by radiation. - the removal of the heat carried to the walls by charged particles, which is highly concentrated (peak values of several tens of MW per m 2 ). Two types of systems are generally used for the plasma-wall interface: throat limiter and axisymmetric divertor. Neither is an ideal candidate to control simultaneously the heat and particle fluxes. This thesis investigates an alternative configuration, the vented limiter, tested for the first time on the Tore Supra tokamak. The vented limiter principle lies on the recycling neutrals collection by slots, in such a way that local thermal overload is avoided. It is shown experimentally that the surface temperature of the prototype installed in Tore Supra remains uniform. As far as the particle collection is concerned, even though the pressure in the vented limiter is lower than the pressure in the throat limiter by a factor 3 for deuterium and 4 helium, it is sufficient to control the plasma density. Moreover, as with a throat limiter, the pressure exhibits a quadratic evolution with the plasma density. To interpret these results, a model describing the plasma recycling on the limiter and the pumping by the slots has been developed. The model has been validated by a comparison with the experimental data. It was then used to propose an optimized version of the prototype with reshaped slots. This should improve the pumping efficiency by a factor 2, in deuterium as well as in helium, but without removing the discrepancy between both pumping efficiencies. As a consequence, even if the thermal behaviour of the vented limiter is satisfactory, its suitability for a future strongly depends on whether it is possible or

  13. A new method used to calibrate Fabry-Perot interferometers on tokamaks

    International Nuclear Information System (INIS)

    Talvard, M.; Javon, C.; Garcin, M.; Thouvenin, D.

    1993-01-01

    Fabry-Perot interferometers are routinely used on the Tore Supra tokamak in order to measure the electron cyclotron emission spectrum especially in the optically thick region (150-300 GHz) for which the intensity is proportional to the electron temperature. In order to give the electron temperature in keV, it is necessary to calibrate the spectral measurements. In practice, one has to determine the calibration curve. The power levels involved in calibration of Fabry-Perot interferometers used on large tokamaks are generally not compatible with classical data processing techniques such as coherent addition or synchronous detection. A new method is presented in this paper able to detect DC signals as low as 0.1 nV. It is based on a proper reduction of the detection frequency bandwidth. The principle of this method is the following: The source signal is chopped and coherently added over N turns of the chopper. The resulting signal is then Fourier analyzed. One finally takes the amplitude at the chopper frequency which allows a much better filtering of the noise. Incident powers less than 1 pW have thus been measured. Till now and probably due to the difficulties detailed above, absolute calibration of Fabry-Perot interferometers has not yet been achieved for applications to plasmas diagnostics in large fusion devices. The method presented here has been first used to achieve this goal. The experimental set up is briefly described and the method detailed. Results on plasma and comparison of the method with more conventional techniques are presented

  14. Status of and prospects for advanced Tokamak regimes from multi-machine comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Litaudon, X.; Becoulet, A.; Imbeaux, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Barbato, E. [Association Euratom-ENEA sulla Fusione Centro Ricerche Energia Frascati (Italy); Doyle, E.J. [California Univ., Los Angeles, CA (United States); Fujita, T. [JAERI, Naka Fusion Research Establishment, Naka (Japan); Gohil, P. [General Atomics, San Diego (United States); Sips, G. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, Garching (Germany)

    2003-07-01

    In this series of 21 slides the author presents an assessment of the present fusion performance of the advanced tokamaks (AT) regimes for non-inductive operation. These AT regimes include data from ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U and Tore-Supra. Only data from both the 'hybrid' without necessarily an ITB (internal transport barrier) or the 'steady-state' scenario have been considered because these scenarios are the 2 candidates for the ITER non inductive current drive operation. A new operational diagram is proposed: the figure of merit for fusion performance and confinement H(ITER-89P).{beta}{sub N}/q{sup 2}{sub 95} versus the bootstrap current fraction e{sup 1/2}.{beta}{sub P}. In this diagram there is a continuous progression from the 'inductive' to the 'hybrid' and 'steady-state' tokamak operating mode. The following range of performance: H(ITER-89P).{beta}{sub N}/q{sup 2}{sub 95} {approx} 0.3-0.4 at {beta}{sub P} {approx} 1, q{sub 95} {approx} 5, is expected for Q = 5 non inductive current drive operation for ITER. Fusion performances tend to decrease with the pulse duration, so extending the plasma performances achieved on a short time scale requires operating safely far from the operational limits. Other conclusions concerning the operating domain of dimensionless parameters such as Larmor radius, collisionality, Mach number and ratio of ion to electron temperature are also presented. (A.C.)

  15. Electron cyclotron waves transmission: new approach for the characterization of electron distribution functions in Tokamak hot plasmas; La transmission d`ondes cyclotroniques electroniques: une approche nouvelle pour caracteriser les fonctions de distribution electronique des plasmas chauds de Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Michelot, Y.

    1995-10-01

    Fast electrons are one of the basic ingredients of plasma operations in many existing thermonuclear fusion research devices. However, the understanding of fast electrons dynamics during creation and sustainment of the superthermal electrons tail is far for being satisfactory. For this reason, the Electron Cyclotron Transmission (ECT) diagnostic was implemented on Tore Supra tokamak. It consists on a microwave transmission system installed on a vertical chord crossing the plasma center and working in the frequency range 77-109 GHz. Variations of the wave amplitude during the propagation across the plasma may be due to refraction and resonant absorption. For the ECT, the most common manifestation of refraction is a reduction of the received power density with respect to the signal detected in vacuum, due to the spreading and deflection of the wave beam. Wave absorption is observed in the vicinity of the electron cyclotron harmonics and may be due both to thermal plasma and to superthermal electron tails. It has a characteristic frequency dependence due to the relativistic mass variation in the wave-electron resonance condition. This thesis presents the first measurements of: the extraordinary mode optical depth at the third harmonics, the electron temperature from the width of a cyclotron absorption line and the relaxation times of the electron distribution during lower hybrid current drive from the ordinary mode spectral superthermal absorption line at the first harmonic. (J.S.). 175 refs., 110 figs., 9 tabs., 3 annexes.

  16. SUPRA SOFT SEPARATION AXIOMS AND SUPRA IRRESOLUTENESS BASED ON SUPRA B-SOFT SETS

    OpenAIRE

    Abd El-latif, Alaa Mohamed; Hosny, Rodyna Ahmed

    2016-01-01

    This paper introduces supra soft b-separation axioms based on the supra b-open soft sets which are more general than supra open soft sets. We investigate the relationships between these supra soft separation axioms. Furthermore, with the help of examples it is established that the converse does not hold. We show that, a supra soft topological space (X; t;E) is supra soft b-T1-space, if xE is supra b-closed soft set in for each x 2 X. Also, we prove that xE is supra b-closed soft set for each ...

  17. Tokamak electron heat transport by direct numerical simulation of small scale turbulence; Transport de chaleur electronique dans un tokamak par simulation numerique directe d'une turbulence de petite echelle

    Energy Technology Data Exchange (ETDEWEB)

    Labit, B

    2002-10-01

    , the crucial role of the electron normalized Larmor has been emphasized: the confinement time is inverse proportional to this parameter. Finally, the low dependence of turbulent transport with the magnetic shear and the inverse aspect ratio is also reported. Although the transport level observed in the simulations is low compared to the experiments, we have tried a direct confrontation with Tore Supra results. This tokamak is well designed to study the electron heat transport. Keeping most of the parameters from a well referenced Tore Supra shot, the nonlinear simulation gives a threshold quite close to the experimental one. The observed turbulent conductivity is a factor fifty lower than the experimental one. An important parameter can not be matched: the normalized Larmor radius, {rho}{sub *}. This limitation has to be overcome in order to confirm this results. Finally, a rigorous confrontation between this result and gyrokinetic simulations has to conclude that the ETG instability cannot describe electron heat loses in tokamaks. (author)

  18. Supra-ballistic phonons

    International Nuclear Information System (INIS)

    Russell, F.M.

    1989-05-01

    Energetic particles moving with a solid, either from nuclear reactions or externally injected, deposit energy by inelastic scattering processes which eventually appears as thermal energy. If the transfer of energy occurs in a crystalline solid then it is possible to couple some of the energy directly to the nuclei forming the lattice by generating phonons. In this paper the transfer of energy from a compound excited nucleus to the lattice is examined by introducing a virtual particle Π. It is shown that by including a Π in the nuclear reaction a substantial amount of energy can be coupled directly to the lattice. In the lattice this particle behaves as a spatially localized phonon of high energy, the so-called supra-ballistic phonon. By multiple inelastic scattering the supra-ballistic phonon eventually thermalizes. Because both the virtual particle Π and the equivalent supra-ballistic phonon have no charge or spin and can only exist within a lattice it is difficult to detect other than by its decay into thermal phonons. The possibility of a Π removing excess energy from a compound nucleus formed by the cold fusion of deuterium is examined. (Author)

  19. On Supra-Additive and Supra-Multiplicative Maps

    OpenAIRE

    Jin Xi Chen; Zi Li Chen

    2013-01-01

    Let A and B be ordered algebras over ℝ, where A has a generating positive cone and B satisfies the property that b2>0 if 0≠b∈B. We give some conditions for a map T:A→B which is supra-additive and supra-multiplicative for all positive and negative elements to be linear and multiplicative; that is, T is a homomorphism of algebras. Our results generalize some known results on supra-additive and supra-multiplicative maps between spaces of real functions.

  20. Experimental study of the interaction between RF antennas and the edge plasma of a tokamak

    International Nuclear Information System (INIS)

    Kubic, Martin

    2013-01-01

    Antennas operating in the ion cyclotron range of frequency (ICRF) provide a useful tool for plasma heating in many tokamaks and are foreseen to play an important role in ITER. However, in addition to the desired heating in the core plasma, spurious interactions with the plasma edge and material boundary are known to occur. Many of these deleterious effects are caused by the formation of radio-frequency (RF) sheaths. The aim of this thesis is to study, mainly experimentally, scrape-off layer (SOL) modifications caused by RF sheaths effects by means of Langmuir probes that are magnetically connected to a powered ICRH antenna. Effects of the two types of Faraday screens' operation on RF-induced SOL modifications are studied for different plasma and antenna configurations - scans of strap power ratio imbalance, injected power and SOL density. In addition to experimental work, the influence of RF sheaths on retarding field analyzer (RFA) measurements of sheath potential is investigated with one-dimensional particle-in-cell code. One-dimensional particle-in-cell simulations show that the RFA is able to measure reliably the sheath potential only for ion plasma frequencies ω π similar to RF cyclotron frequency ω rf , while for the real SOL conditions (ω π ≥ ω rf ), when the RFA is magnetically connected to RF region, it is strongly underestimated. An alternative method to investigate RF sheaths effects is proposed by using broadening of the ion distribution function as an evidence of the RF electric fields in the sheath. RFA measurements in Tore Supra indicate that RF potentials do indeed propagate from the antenna 12 m along magnetic field lines. (author) [fr

  1. Edge turbulence effect on ultra-fast swept reflectometry core measurements in tokamak plasmas

    Science.gov (United States)

    Zadvitskiy, G. V.; Heuraux, S.; Lechte, C.; Hacquin, S.; Sabot, R.

    2018-02-01

    Ultra-fast frequency-swept reflectometry (UFSR) enables one to provide information about the turbulence radial wave-number spectrum and perturbation amplitude with good spatial and temporal resolutions. However, a data interpretation of USFR is quiet tricky. An iterative algorithm to solve this inverse problem was used in past works, Gerbaud (2006 Rev. Sci. Instrum. 77 10E928). For a direct solution, a fast 1D Helmholtz solver was used. Two-dimensional effects are strong and should be taken into account during data interpretation. As 2D full-wave codes are still too time consuming for systematic application, fast 2D approaches based on the Born approximation are of prime interest. Such methods gives good results in the case of small turbulence levels. However in tokamak plasmas, edge turbulence is usually very strong and can distort and broaden the probing beam Sysoeva et al (2015 Nucl. Fusion 55 033016). It was shown that this can change reflectometer phase response from the plasma core. Comparison between 2D full wave computation and the simplified Born approximation was done. The approximated method can provide a right spectral shape, but it is unable to describe a change of the spectral amplitude with an edge turbulence level. Computation for the O-mode wave with the linear density profile in the slab geometry and for realistic Tore-Supra density profile, based on the experimental data turbulence amplitude and spectrum, were performed to investigate the role of strong edge turbulence. It is shown that the spectral peak in the signal amplitude variation spectrum which rises with edge turbulence can be a signature of strong edge turbulence. Moreover, computations for misaligned receiving and emitting antennas were performed. It was found that the signal amplitude variation peak changes its position with a receiving antenna poloidal displacement.

  2. ANTWKB: a code for the simulation of ion cyclotron antennas in tokamaks

    International Nuclear Information System (INIS)

    Brambilla, M.

    1995-04-01

    We have developed a code which evaluates the complex input impedance, the loading, and the spectral distribution of the launched power, of metallic antennas for ion cyclotron heating of large tokamak plasmas. The current distribution along the conductors is obtained selfconsistently from a variational method. The plasma response is evaluated assuming that the WKB approximation can be used already at the plasma edge, thereby avoiding the lengthy integration of the wave equations in the plasma. This makes possible systematic scans over frequency or other parameters, while retaining a sufficiently large number of Fourier components in the radiated fields to ensure convergence of both the resistive and reactive part of the power. Optionally, the code can evaluate the antenna response in vacuum or with a dummy load, for comparison with test bank measurements. We have applied the code to a few antennas of practical interest. The code reproduces accurately the expected transmission-line-like behaviour of a simple feeder-to-short antenna, and reasonably well the measured properties of the folded antenna of the ASDEX Upgrade ICRF experiment. This antenna is found to have particularly favourable properties, since its outer conductors present to the plasma a relatively uniform current over a broad range of frequencies, which, moreover, is always larger than in the return conductors. The loading of the ''violin antenna'' recently proposed for use in ITER is found to be satisfactory in the vicinity of antenna resonance, although rather poor at other frequencies. In the case of simple strap antennas replacing the short by an adjustable capacity, as in TORE SUPRA, is confirmed to be a good way of optimizing the loading. (orig.)

  3. A new simulation framework based on the Kepler and Scicos open-source software for the design and qualification of tokamak control algorithms: first test case results

    International Nuclear Information System (INIS)

    Barana, Oliviero; Bremond, Sylvain; Ravenel, Nathalie; Moreau, Philippe; Boulbe, Cedric; Mannori, Simone

    2011-01-01

    Plasma control is recognized to be a crucial issue for the achievement of ITER objectives. One of the most challenging tasks for the preparation of the ITER operation will therefore be the design and qualification of a variety of control algorithms. This highlights the need for a simulation platform capable of supporting the design, integration and test of advanced control algorithms on complex physics models. With this aim, a generic multi-purpose 'flight' Simulator (GMFS) is being developed at IRFM (Institut de Recherche sur la Fusion par confinement Magnetique), CEA Cadarache, France. The GMFS is based on Kepler, a free interdisciplinary open-source Java software. Kepler will be used as a simulation platform to test and improve control algorithms before their actual use in the real control system. The physics and engineering codes complementary to the control algorithms will be supplied by the EFDA Integrated Tokamak Modelling Task Force (ITM-TF). The GMFS will be benchmarked, at the beginning, on the Tore Supra Tokamak. In this paper we will report on a test case suitable to demonstrate the feasibility of a part of GMFS, namely the development of workflows where to create and verify ITER plasma boundary feedback control algorithms. lt consists of: a) derivation of a linear plasma response model; b) design of a control diagram under the ScicosLab/Scicos open-source software; c) porting of the diagram under Kepler; d) substitution of the Kepler controller with a controller generated by a special Scicos extension; e) substitution of the simplified static linear model with the free-boundary equilibrium code CEDRES++. The test case demonstrated the feasibility of employing Kepler, ScicosLab/Scicos and other expressly made codes in view of the conception of valuable instruments for the active control of ITER and it can be considered as a first step in this direction. (authors)

  4. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  5. Torus II Supra

    International Nuclear Information System (INIS)

    Adam, J.; Aymar, R.; Briffod, G.

    1979-10-01

    Clearly a new device now ready for construction will be operated during roughly the same period of time as JET and it should therefore be designed as a useful complement to JET with the aim of preparing for the next phase of the programme. A number of prospective studies, in particular the Long Term Planning and the work on INTOR and NET have pointed out several fields where a large effort was needed. In some of these fields the EURATOM-C.E.A. Association is especially well prepared to bring important contributions and this is the case for: - the construction and operation of a super conducting Tokamak, - the development of Radio Frequency Heating. These two subjects have been chosen as the major items of the proposed programme. In addition the existing expertise in Tokamak physics and the characteristics of the device lead to propose two other subjects of work, namely: - the dynamics of impurities, - the study of long pulses

  6. Interaction between fast ions and ion cyclotron heating in a tokamak plasma; Interaction des ions rapides avec les ondes a la frequence cyclotronique ionique dans un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bergeaud, V

    2001-11-01

    the Tore Supra tokamak. The different elements of the thesis are applied to the optimisation of the heating scenarios in a reactor tokamak. We study a scenario in which a small fraction of helium-3 is introduced in the plasma. This so-called helium-3 minority heating scenario yields a large fraction of ion heating, that increases the fusion power produced in the plasma. Furthermore, this method reduces the parasitic absorption of ICRF power by fusion products. Lastly, it is shown that this parasitic absorption is reduced by the effect of phase correlation for high energy particles. (author)

  7. Interaction of fast ions with ion cyclotron electromagnetic waves in tokamak plasma; Interaction des ions rapides avec les ondes a la frequence cyclotronique ionique dans un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bergeaud, V

    2000-12-01

    the Tore Supra tokamak. The different elements of the thesis are applied to the optimisation of the heating scenarios in a reactor tokamak. We study a scenario in which a small fraction of helium-3 is introduced in the plasma. This so-called helium-3 minority heating scenario yields a large fraction of ion heating, that increases the fusion power produced in the plasma. Furthermore, this method reduces the parasitic absorption of ICRF power by fusion products. Lastly, it is shown that this parasitic absorption is reduced by the effect of phase correlation for high energy particles. (author)

  8. Measurement of lower hybrid hot spots using a retarding field analyzer in Tore Supra

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Petržílka, Václav; Ekedahl, A.; Fuchs, Vladimír; Gauthier, E.; Goniche, M.; Kočan, M.; Pascal, J.Y.; Saint Laurent, F.

    390-391, č. 1 (2009), s. 904-906 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/18th./. Toledo, 26.05.2008-30.5.2008] R&D Projects: GA ČR GA202/07/0044 Institutional research plan: CEZ:AV0Z20430508 Keywords : Lower Hybrid * hot spots * ELECTRONS * POWER Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.933, year: 2009 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6TXN-4VG7MW6-7&_user=6542793&_rdoc=1&_fmt=&_orig=search&_sort=d&_docanchor=&view=c&_acct=C000070123&_version=1&_urlVersion=0&_userid=6542793&md5=944ed4d95df86a149fa36439db6215db

  9. Experimental Investigation of Nonlinear Coupling of Lower Hybrid Waves on Tore Supra

    Czech Academy of Sciences Publication Activity Database

    Goniche, M.; Frincu, B.; Ekedahl, A.; Petržílka, Václav; Berger-By, G.; Hillairet, J.; Litaudon, X.; Preynas, M.; Voyer, D.

    2012-01-01

    Roč. 62, č. 2 (2012), s. 322-332 ISSN 1536-1055 R&D Projects: GA ČR GA202/07/0044 Institutional research plan: CEZ:AV0Z20430508 Keywords : LHwave * plasma * lower hybrid * wave coupling * nonlinear coupling Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.517, year: 2012

  10. Thermal and non-thermal particle interaction with the LHCD launchers in Tore Supra

    Czech Academy of Sciences Publication Activity Database

    Ekedahl, A.; Goniche, M.; Balorin, C.; Basiuk, V.; Bibet, Ph.; Chantant, M.; Colas, L.; Delpech, L.; Desgranges, C.; Eriksson, L.-G.; Joffrin, E.; Kazarian, F.; Lowry, C.; Moreau, Ph.; Petržílka, Václav; Portafaix, C.; Prou, M.; Roche, H.

    363-365, č. 8 (2007), s. 1329-1333 ISSN 0022-3115 R&D Projects: GA ČR(CZ) GA202/07/0044 Institutional research plan: CEZ:AV0Z20430508 Keywords : LH grill * plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.643, year: 2007

  11. Parallel expansion of the ablation cloud during pellet injection in Tore Supra

    International Nuclear Information System (INIS)

    Pegourie, B.; Bruneau, J.L.; Picchiottino, J.M.

    1991-01-01

    The ablated matter propagation along the field lines during pellet ablation is observed with a five chords interferometer toroidally located at -π/3 of the pellet injector. The time resolution is 16μs and the sensitivity better than 3 10 17 m -2 . The beginning of the fast acquisition is triggered by the pellet itself and its maximum duration is 20ms. About 500μs after the pellet enters the discharge, the experimental signals exhibit a steep increase. Excepted in a few cases for which a strong oscillation at a typical frequency of 0.5kHz was detected during several ms, a new quasi-steady state is reached after about lms. The importance of the measured perturbation and the details of the sequence described above depends, for each chord, on both the Q profile and pellet penetration

  12. Determination of the safety factor profile in Tore Supra from striations observed during pellet ablation

    International Nuclear Information System (INIS)

    Dubois, M.A.; Sabot, R.; Pegourie, B.; Drawin, H.W.; Geraud, A.

    1992-01-01

    Striations in H α (D α ) light observed across the ablation cloud of injected hydrogen pellets can be explained in most cases by the presence of resonant magnetic surfaces. We show that it is possible to identify them and to obtain a very detailed profile of the safety factor. This q- profile shows shear plateaus related to rational q-values, thus suggesting the presence of magnetic turbulence

  13. Theory and experiments on RF plasma heating, current drive and profile control in TORE SUPRA

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, D.

    1994-01-01

    This paper reviews the main experimental and theoretical achievements related to the study of RF heating and non-inductive current drive and particularly phenomena related to the current density profile control and the potentiality of producing stationary enhanced performance regimes: description of the Lower Hybrid (LH) and Ion Cyclotron Resonant Frequency (ICRF) systems; long pulse coupling performance of the RF systems; observation of the transition to the so-called ``stationary LHEP regime`` in which the (flat) central current density and (peaked) electron temperature profiles are fully decoupled; experiments on ICRF sawtooth stabilization with the combined effect of LHCD modifying the current density profile; diffusion of fast electrons generated by LH waves; ramp-up experiments in which the LH power provided a significant part of the resistive poloidal flux and flux consumption scaling; theory of spectral wave diffusion and multipass absorption; fast wave current drive modelling with the Alcyon full wave code; a reflector LH antenna concept. 18 figs., 48 refs.

  14. Theory and experiments on RF plasma heating, current drive and profile control in TORE SUPRA

    International Nuclear Information System (INIS)

    Moreau, D.

    1994-01-01

    This paper reviews the main experimental and theoretical achievements related to the study of RF heating and non-inductive current drive and particularly phenomena related to the current density profile control and the potentiality of producing stationary enhanced performance regimes: description of the Lower Hybrid (LH) and Ion Cyclotron Resonant Frequency (ICRF) systems; long pulse coupling performance of the RF systems; observation of the transition to the so-called ''stationary LHEP regime'' in which the (flat) central current density and (peaked) electron temperature profiles are fully decoupled; experiments on ICRF sawtooth stabilization with the combined effect of LHCD modifying the current density profile; diffusion of fast electrons generated by LH waves; ramp-up experiments in which the LH power provided a significant part of the resistive poloidal flux and flux consumption scaling; theory of spectral wave diffusion and multipass absorption; fast wave current drive modelling with the Alcyon full wave code; a reflector LH antenna concept. 18 figs., 48 refs

  15. The 8 MW lower hybrid electron mode system for the additional heating of the plasma of the FTU Tokamak

    International Nuclear Information System (INIS)

    Andreani, R.; De Marco, F.; Ferro, C.; Mirizzi, F.; Papitto, P.; Santini, F.; Segre, S.E.; Sassi, M.

    1985-01-01

    The ''Electron Mode'' regime of LH Heating, based on the same physics as the current drive, has been extensively studied and experimentally tested especially with respect to the relation between frequency and density limit. These results have largely contributed to the decision to build a CD system on TORE SUPRA. Based on the same motivations, the Lower Hybrid 'Electron Mode' Heating (frequency: 8 ''Electron Mode'' Heating (frequency: 8 GHz), has been chosen to heat the plasma of the FTU Tokamak. The RF power required (8 MW at 8 GHz) will be produced by 16 gyrotron oscillators (500 KW unit power) feeding 16 grill couplers installed on 8 equatorial ports of FTU. The dc power supplies will be ,odularly built to be compatible even with completely different sort of tubes (e.g. for IRCH). The transmission lines between the generators and the grills will be circular oversized waveguides to reduce the losses to less than 1 dB. Each grill will consist of an 8x8 matrix of rectangular waveguides pressurized and terminated by thik (one wavelength) alumina windows facing the grill mouth. Gyrotron availability has been verified through studies conducted by the two major manufacturers presently on the market. Preliminary quotations and delivery times have been obtained. The design of the grill couplers has been supplemented by a study contract with an industrial research laboratory which is producing a prototype structure and ceramic windows with very promising results. Microwave mode converters and power dividers for the transmission system have been designed and prototypes are being built and will be tested shortly. An 8 GHz, 25 KW cw test bench has been already commissioned and will be used to test all the microwave components. The power level is more than adequate also to process single channels of the coupling structures

  16. Study and optimization of magnetized ICRF discharges for tokamak wall conditioning and assessment of the applicability to ITER

    International Nuclear Information System (INIS)

    Wauters, T.

    2011-11-01

    This work is devoted to the study and optimization of the Ion Cyclotron Wall Conditioning (ICWC) technique. ICWC, operated in presence of the toroidal magnetic field, makes use of four main tokamak systems: the ICRF antennas to initiate and sustain the conditioning discharge, the gas injection valves to provide the discharge gas, the machine pumps to remove the wall desorbed particles, and the poloidal magnetic field system to optimize the discharge homogeneity. Additionally neutral gas and plasma diagnostics are required to monitor the discharge and the conditioning efficiency. In chapter 2 a general overview on ICWC is given. Chapter 3 treats the ICRF discharge homogeneity and the confinement properties of the employed magnetic field. In the first part we will discuss experimental facts on plasma homogeneity, and how experimental optimization led to its improvement. In the second part of the chapter the confinement properties of a partially ionized plasma in a toroidal magnetic field configuration with additional small vertical component are discussed. Chapter 4 gives an overview of experimental results on the efficiency of ICWC, obtained on TORE SUPRA, TEXTOR, JET and ASDEX Upgrade. In chapter 5 a 0D kinetic description of hydrogen-helium RF plasmas is outlined. The model, describing the evolution of ICRF plasmas from discharge initiation to the (quasi) steady state plasma stage, is developed to obtain insight on ICRF plasma parameters, particle fluxes to the walls and the main collisional processes. Chapter 6 presents a minimum structure for a 0D reservoir model of the wall to investigate in deeper detail the ICWC plasma wall interaction during isotopic exchange experiments. The hypothesis used to build up the wall model is that the same model structure should be able to describe the wall behavior during normal plasmas and conditioning procedures. Chapter 7 extrapolates the results to the envisaged application of ICWC on ITER

  17. On Bitopological Supra B-Open Sets

    OpenAIRE

    M.Lellis Thivagar; B.Meera Devi

    2012-01-01

    In this paper, we introduce and investigate a new class of sets and maps be- tween bitopological spaces called supra(1,2) b-open, and supra (1,2) b-continuous maps, respectively. Furthermore, we introduce the concepts of supra(1,2) locally-closed, supra(1,2) locally b-closed sets. We also introduce supra(1,2) extremely disconnected. Finally, additional properties of these sets are investigated.

  18. Tokamak formation and sustainment by tokamak injection

    International Nuclear Information System (INIS)

    Farengo, R.; Jarboe, T.R.

    1991-01-01

    The authors propose here a new helicity injection method for tokamak formation and sustainment that has high efficiency, conserves toroidal symmetry and is inductively driven. The basic idea is to inject a small tokamak (source tokamak) into a larger tokamak (steady tokamak). This current drive scheme eliminates the need for the ohmic heating transformer in the steady tokamak allowing the formation of very small aspect ratio tokamaks (Spherical Tori). Thus, steady state operation and high beta can be realized simultaneously. The method can also be applied to a larger aspect ratio tokamak and used in conjunction with the standard inductive formation technique. In order to allow for translation the ohmic heating coil used to produce the source tokamaks must be fed from one end (as in the CSS device) and the toroidal field coil must link both tokamaks. After formation the source tokamaks are accelerated towards the steady tokamak by a mirror field and the tension of the field lines that wrap around both tokamaks (producing a doublet type configuration). In a tokamak the helicity is proportional to the current. This indicates that (assuming helicity is conserved during the merging process) a steady state situation will result if the helicity supplied by the source tokamaks is equal to the helicity dissipated by the steady tokamak. Assuming that source tokamaks of helicity K s are injected with frequency f, the steady state condition can be written as: fK s = 2V t Ψ t = K t /τ K where V t , Ψ t , K t and τ K are the ohmic loop voltage, toroidal flux, helicity and helicity decay time of the steady tokamak. A simple calculation shows that the DIII-D tokamak could be sustained by injecting source tokamaks with R = 1.20 m, a = 0.23 m and I = 151 kA at a frequency of 120 Hz. 1 ref

  19. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

    Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  20. Temaet folkemord i forfatterskapet til Tore Linne Eriksen. Festskrift til Tore Linne Eriksen

    OpenAIRE

    Balsvik, Randi Rønning

    2015-01-01

    Published version also available at https://journals.hioa.no/index.php/fleks/article/view/1494 This work is licensed under a Creative Commons Attribution 4.0 International License. This article points to the responsibility historians have in the formation of what we may call the collective memory of persons, groups and states. What are the images of “the other” conveyed in textbooks and media? In Norway, the historian Tore Linné Eriksen has – more than any other scholar – used ...

  1. Radiation scattering back to the plasma by the tokamak inner wall in the energy range 50-500 keV during lower hybrid current drive

    International Nuclear Information System (INIS)

    Peysson, Y.

    1990-10-01

    We describe the wall reflectivity by the ratio between the number of photons emerging from the wall and the number entering - and determine the proportion of the reflected contribution to the detected radiations. Various emission profiles and plasma positions in the tokamak chamber have been considered. The contribution of multiple reflections has also be investigated. The wall reflectivity can lead to spurious conclusions for a peaked radial profile in the vicinity of the plasma edge. The next step is devoted to the resolution of the radiation transport equation in solid matter. As an heterogeneous medium is considered - carbon tiles brazed on an iron bulk -, the solution is determined by a numerical Monte-Carlo method. The reflectivity is greatly enhanced by a carbon layer between 50 keV and 150 keV, even for a thickness of one centimeter. The reflectivity is then nearly independent of the energy of the entering photons up to 500 KeV, and lies between 0.15 and 0.4 from a perpendicular to a nearly tangential incidence. Angular corrections have also been considered. Finally, a fully description of the X-ray reflectivity in the high energy range has been performed, taking account of the toroidal geometry and the exact solution of the radiation transport equation. Comparison between theoretical and experimental results obtained with the Tore-Supra high energy X-ray spectrometer has been done. A strong reflectivity effect is observed for the more peripheral line of sight when the plasma emission profile is peaked. There is a good agreement for the total number of detected photons with an energy greater than 100 keV The measured energy spectrum lies up to 200 keV when the photon energy spectrum of the plasma determined from the central chords extends up to 500 keV. A procedure to determine the energy threshold above which the photon energy spectrum is free of the reflected contribution is proposed

  2. SUPRA - Enhanced upset recovery simulation

    NARCIS (Netherlands)

    Groen, E.; Ledegang, W.; Field, J.; Smaili, H.; Roza, M.; Fucke, L.; Nooij, S.; Goman, M.; Mayrhofer, M.; Zaichik, L.E.; Grigoryev, M.; Biryukov, V.

    2012-01-01

    The SUPRA research project - Simulation of Upset Recovery in Aviation - has been funded by the European Union 7th Framework Program to enhance the flight simulation envelope for upset recovery simulation. Within the project an extended aerodynamic model, capturing the key aerodynamics during and

  3. EURATOM-CEA association contributions to the 18. IAEA fusion energy conference

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Peysson, Y.; Hoang, G.T.

    2000-12-01

    The 9 contributions of EURATOM-Cea association to the fusion energy conference hold at Sorrento are gathered in this document with 7 additional papers. The different titles are: 1) Ergodic divertor experiments on the route to steady state operation of Tore-Supra, 2) High power lower hybrid current drive experiments in Tore-Supra tokamak, 3) Electron transport and improved confinement on Tore-Supra, 4) ECRH experiments and developments for long pulse in Tore-Supra, 5) Impurity penetration and contamination in Tore-Supra ergodic divertor experiments, 6) Real time plasma feed-back control: an overview of Tore-Supra achievements, 7) Numerical assessment of the ion turbulent thermal transport scaling laws, 8) Design of next step tokamak: consistent analysis of plasma flux consumption and poloidal, 9) Large superconducting conductors and joints for fusion magnets: from conceptual design to test at full size scale, 10) Burst-prone transport in tokamaks with internal transport barriers, 11) Electrostatic turbulence with finite parallel correlation length and radial electric field generation, 12) Theoretical issues in tokamak confinement: internal-edge transport barriers and runaway avalanche confinement, 13) Core and edge confinement studies with different heating methods in JET, 14) Confinement and transport studies of conventional scenarios in ASDEX upgrade, 15) First test results for the ITER central solenoid model coil, and 16) Progress of the ITER central solenoid model coil program

  4. EURATOM-CEA association contributions to the 18. IAEA fusion energy conference

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Peysson, Y.; Hoang, G.T. [and others

    2000-12-01

    The 9 contributions of EURATOM-Cea association to the fusion energy conference hold at Sorrento are gathered in this document with 7 additional papers. The different titles are: 1) Ergodic divertor experiments on the route to steady state operation of Tore-Supra, 2) High power lower hybrid current drive experiments in Tore-Supra tokamak, 3) Electron transport and improved confinement on Tore-Supra, 4) ECRH experiments and developments for long pulse in Tore-Supra, 5) Impurity penetration and contamination in Tore-Supra ergodic divertor experiments, 6) Real time plasma feed-back control: an overview of Tore-Supra achievements, 7) Numerical assessment of the ion turbulent thermal transport scaling laws, 8) Design of next step tokamak: consistent analysis of plasma flux consumption and poloidal, 9) Large superconducting conductors and joints for fusion magnets: from conceptual design to test at full size scale, 10) Burst-prone transport in tokamaks with internal transport barriers, 11) Electrostatic turbulence with finite parallel correlation length and radial electric field generation, 12) Theoretical issues in tokamak confinement: internal-edge transport barriers and runaway avalanche confinement, 13) Core and edge confinement studies with different heating methods in JET, 14) Confinement and transport studies of conventional scenarios in ASDEX upgrade, 15) First test results for the ITER central solenoid model coil, and 16) Progress of the ITER central solenoid model coil program.

  5. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

    The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)

  6. On Neutrosophic Semi-Supra Open Set and Neutrosophic Semi-Supra Continuous Functions

    OpenAIRE

    R. Dhavaseelan; M. Parimala; S. Jafari; F. Smarandache

    2017-01-01

    In this paper, we introduce and investigate a new class of sets and functions between topological space called neutrosophic semi-supra open set and neutrosophic semi-supra open continuous functions respectively.

  7. Association Contributions to the 21th EPS Conference

    International Nuclear Information System (INIS)

    1995-01-01

    A wide range of topics concerning plasma physics at Tore Supra tokamak device is covered. Different ways of plasma heating, such as current drive, electron and ion cyclotron-resonance heating and lower hybrid heating, plasma confinement, different topics on transport theory and plasma diagnostics are discussed. The new ergodic divertor at Tore Supra is studied from different aspects. 35 items have been separately indexed for the INIS database. (K.A.)

  8. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  9. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  10. Status of tokamak research

    Energy Technology Data Exchange (ETDEWEB)

    Rawls, J.M. (ed.)

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design. (MOW)

  11. On Some Maps in Supra Topological Ordered Spaces

    OpenAIRE

    Al-shami, Tareq Mohammed

    2018-01-01

    In [6] the notion of supra semi open sets was presented and some of its properties were discussed. In this study, we introduce and investigate four main concepts namely supra continuous (supra open, supra closed, supra homeomorphism) maps via supra topological ordered spaces. Our findings in this work generalize some previous results in ([1], [13]). Many examples are considered to show the concepts introduced and main results obtained herein.

  12. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  13. Tokamak transport phenomenology and plasma dynamic response

    International Nuclear Information System (INIS)

    Moret, J.M.; Association Euratom CEA, Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance

    1991-07-01

    A system identification method is developed to estimate the transfer function of the system from the time evolution of its parameters to any excitation. The form of the identified transfer function is linked to a representation of the transport in terms of poles (eigenvalues) and eigenmodes. These eigenvalues and eigenvectors are thus directly deduced from the raw data with no restriction on the underlying processes and there is consequently no need to adjust any simplified transport model to the experimental data. This method is illustrated in this paper by analysing the injection of pellets on Tore Supra. The density and the temperature transfer functions were observed to share the same poles with the corresponding eigenmodes grouped in pairs with identical profiles. This implies the presence of a coupling between the particle and heat flow. A criterion is developed to select amongst the possible coupling mechanisms, based on compatibility with the observed transfer function. The selection suggests a model in which the particle diffusion coefficient depends on the density and on the temperature gradient

  14. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  15. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  16. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

    The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

  17. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    International Nuclear Information System (INIS)

    Escourbiac, F.; Missirlian, M.; Schlosser, J.; Bobin-Vastra, I.; Kuznetsov, V.; Schedler, B.

    2004-01-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m 2 with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m 2 . These results highlight the high potential of this technology for ITER divertor application

  18. Results of water corrosion in static cell tests representing multi-metal assemblies in the hydraulic circuits of Tore supra

    International Nuclear Information System (INIS)

    Lipa, M.; Blanchet, J.; Cellier, F.

    2007-01-01

    Following experiences obtained with steam generator tubes of nuclear power plants, a cooling water quality of AVT (all volatile treatment) has been defined based on demineralised water with adjustment of the pH value to about 9.0/7.0 (25 C/200 C) by addiction of ammoniac, and hydrazine in order to absorb oxygen dissolved in water. At that time, a simplified water corrosion test program has been performed using static (no circulation) test cell samples made of above mentioned TS metal combinations. All test cell samples, prepared and filled with AVT water, were performed at 280 C and 65 bars in an autoclave during 3000 hours. The test cell water temperature has been chosen to be sufficient above the TS component working temperature, in order to accelerate an eventual corrosion process. Generally all above mentioned metal combinations survived the test campaign without stress corrosion cracking, with the exception of the memory metal junction (creep in Cu) and the bellows made of St-St 316L and Inconel 625 while 316 Ti bellows survived. In contrary to the vacuum brazed Cu-LSTP to St-St samples, some of flame brazed Cu to St-St samples failed either in the braze joint or in the copper structure itself. For comparison, a spot weld of an inflated 316L panel sample, filled voluntary with a caustic solution of pH 11.5 (25 C), failed after 90 h of testing (intergranular cracking at the spot weld), while an identical sample containing AVT water of pH 9.0 (25 C) survived without damage. The results of these tests, performed during 1986 and 1997, have never been published and therefore are presented more in detail in this paper since corrosion in hydraulic circuits is also an issue of ITER. Up to day, the TS cooling water plant operates with an above mentioned water treatment and no water leaks have been detected on in-vessel components originating from water corrosion at high temperature and high pressure. (orig.)

  19. Validation of the ITER-relevant passive-active-multijunction LHCD launcher on long pulses in Tore Supra

    Czech Academy of Sciences Publication Activity Database

    Ekedahl, A.; Delpech, L.; Goniche, M.; Guilhem, D.; Hillairet, J.; Preynas, M.; Sharma, P.K.; Achard, J.; Bae, Y.S.; Bai, X.; Balorin, C.; Baranov, Y.; Basiuk, V.; Bécoulet, A.; Belo, J.; Berger-By, G.; Brémond, S.; Castaldo, C.; Ceccuzzi, S.; Cesario, R.; Corbel, E.; Courtois, X.; Decker, J.; Delmas, E.; Ding, X.; Douai, D.; Goletto, C.; Gunn, J. P.; Hertout, P.; Hoang, G.T.; Imbeaux, F.; Kirov, K.K.; Litaudon, X.; Magne, R.; Mailloux, J.; Mazon, D.; Mirizzi, F.; Mollard, P.; Moreau, P.; Oosako, T.; Petržílka, Václav; Peysson, Y.; Poli, S.; Prou, M.; Saint-Laurent, F.; Samaille, F.; Saoutic, B.

    2010-01-01

    Roč. 50, č. 11 (2010), s. 112002-112002 ISSN 0029-5515 Institutional research plan: CEZ:AV0Z20430508 Keywords : LH wave * plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.303, year: 2010 http://iopscience.iop.org/0029-5515/50/11/112002/pdf/0029-5515_50_11_112002.pdf

  20. Resistive flux saving and current profile control during lower hybrid waves assisted current rise in TORE SUPRA

    International Nuclear Information System (INIS)

    van Houtte, D.; Hoang, G.T.; Joffrin, E.; Lecoustey, P.; Moreau, D.; Parlange, F.; Tonon, G.; Vallet, J.C.

    1992-01-01

    Resistive flux saving at densities n e = (1 - 2) x 10 19 m -3 has been studied. High flux saving efficiency (0.7 x 10 13 Wb/J/m -1 ) can be achieved for a low rf power (P LH = 0.5 MW) due to the beneficial effect of the electric field on the suprathermal electrons. However for power higher than 1 MW, the efficiency is 0.25 x 10 13 Wb/J/m -1 . This flux saving efficiency is comparable to the one obtained during the flat top phase. The application of the LH power during a low density current ramp-up tends to peak the electron temperature and current density profiles. The rf power level, the parallel wavenumber and the current ramp rate allow to control the trajectories of the plasma discharges during the current rise inside the MHD stable domain

  1. Results of water corrosion in static cell tests representing multi-metal assemblies in the hydraulic circuits of Tore supra

    Energy Technology Data Exchange (ETDEWEB)

    Lipa, M. [CEA/DSM/DRFC Centre de Cadarache, 13 - Saint-Paul lez Durance (France); Blanchet, J.; Cellier, F. [Framatome, 71 - Saint Marcel (France). Centre Technique

    2007-07-01

    Following experiences obtained with steam generator tubes of nuclear power plants, a cooling water quality of AVT (all volatile treatment) has been defined based on demineralised water with adjustment of the pH value to about 9.0/7.0 (25 C/200 C) by addiction of ammoniac, and hydrazine in order to absorb oxygen dissolved in water. At that time, a simplified water corrosion test program has been performed using static (no circulation) test cell samples made of above mentioned TS metal combinations. All test cell samples, prepared and filled with AVT water, were performed at 280 C and 65 bars in an autoclave during 3000 hours. The test cell water temperature has been chosen to be sufficient above the TS component working temperature, in order to accelerate an eventual corrosion process. Generally all above mentioned metal combinations survived the test campaign without stress corrosion cracking, with the exception of the memory metal junction (creep in Cu) and the bellows made of St-St 316L and Inconel 625 while 316 Ti bellows survived. In contrary to the vacuum brazed Cu-LSTP to St-St samples, some of flame brazed Cu to St-St samples failed either in the braze joint or in the copper structure itself. For comparison, a spot weld of an inflated 316L panel sample, filled voluntary with a caustic solution of pH 11.5 (25 C), failed after 90 h of testing (intergranular cracking at the spot weld), while an identical sample containing AVT water of pH 9.0 (25 C) survived without damage. The results of these tests, performed during 1986 and 1997, have never been published and therefore are presented more in detail in this paper since corrosion in hydraulic circuits is also an issue of ITER. Up to day, the TS cooling water plant operates with an above mentioned water treatment and no water leaks have been detected on in-vessel components originating from water corrosion at high temperature and high pressure. (orig.)

  2. Novel technique and simple approach for supra-alar region and supra-alar crease correction by supra-alar cinching.

    Science.gov (United States)

    Selvaraj, Loganathan

    2016-01-01

    This technical report describes a simple and innovative surgical technique for supra-alar sidewall region constriction and supra-alar crease attenuation by cinching technique through intraoral approach.

  3. Novel technique and simple approach for supra-alar region and supra-alar crease correction by supra-alar cinching

    OpenAIRE

    Selvaraj, Loganathan

    2016-01-01

    This technical report describes a simple and innovative surgical technique for supra-alar sidewall region constriction and supra-alar crease attenuation by cinching technique through intraoral approach.

  4. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1977-01-01

    The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

  5. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  6. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  7. Research using small tokamaks

    International Nuclear Information System (INIS)

    1989-07-01

    These proceedings of the IAEA-sponsored meeting held in Nice, France 10-11 October, 1988, contain the manuscripts of the 21 reports dealing with research using small tokamaks. The purpose of this meeting was to highlight some of the achievements of small tokamaks and alternative magnetic confinement concepts and assess the suitability of starting new programs, particularly in developing countries. Papers presented were either review papers, or were detailed descriptions of particular experiments or concepts. Refs, figs and tabs

  8. Tokamak simulation code manual

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs.

  9. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  10. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  11. Cyclotron radiation as Tokamak diagnostics

    International Nuclear Information System (INIS)

    Fiedler-Ferrari, N.

    1985-01-01

    A brief introduction to the use of Electron Cyclotron Emission as diagnostics in tokamaks is made. The utilization feasibility of this dignostics in the TBR-1 and TTF2A tokamaks is discussed. (L.C.) [pt

  12. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-03-01

    This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

  13. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

  14. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  15. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

    This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

  16. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

  17. High beta tokamak instabilities

    International Nuclear Information System (INIS)

    Bateman, G.

    1977-01-01

    Theoretical predictions using the ideal MHD model indicable that large-scale ballooning modes should appear when the average beta is raised about 1 to 2% in present-day tokamak geometries or 5 to 10% in more optimized geometries. The onset of instability is predicted to be sudden and the behavior of ballooning modes to be strikingly different from the saw-tooth and Mirnov oscillations experimentally observed at low beta. Conditions close to the predicted onset were achieved in ORMAK with no noticeable change in plasma behavior. Experiments are planned for the ISX tokamak to test the beta limit. 15 references, 3 figures

  18. Long Pulse Technology Tokamak

    International Nuclear Information System (INIS)

    Jernigan, T.C.

    1978-01-01

    The LPTT tokamak is a non-circular tokamak (R = 1.5 m, a = .45 m) proposed by ORNL for extended pulse operation at high β (5%) and reactor level wall power loading (40 w/cm 2 ). The toroidal field coils are superconducting and a super-conducting bundle divertor is proposed for active impurity control. All systems are designed for continuous operation which will provide pulse lengths > 20 seconds with a 6 to 10 weber flux swing. Experimental access and flexibility in operation are primary design goals

  19. Characterisation of Supra- and Infratentorial ICP Profiles.

    Science.gov (United States)

    Moyse, Emmanuel; Ros, Maxime; Marhar, Fouad; Swider, Pascal; Schmidt, Eric Albert

    2016-01-01

    In pathophysiology and clinical practice, the intracranial pressure (ICP) profiles in the supratentorial and infratentorial compartments are unclear. We know that the pressure within the skull is unevenly distributed, with demonstrated ICP gradients. We recorded and characterised the supra- and infratentorial ICP patterns to understand what drives the transtentorial ICP gradient.A 70-year-old man was operated on for acute cerebellar infarction. One supratentorial probe and one cerebellar probe were implanted. Both signals were recorded concurrently and analysed off-line. We calculated mean ICP, ICP pulse amplitude, respiratory waves, slow waves and the RAP index of supra- and infratentorial ICP signals. Then, we measured transtentorial difference and performed correlation analysis for every index.Supratentorial ICP mean was 8.5 mmHg lower than infratentorial ICP, but the difference lessens for higher values. Both signals across the tentorium showed close correlation. Supra- and infratentorial pulse amplitude, respiratory waves and slow waves also showed a high degree of correlation. The compensatory reserve (RAP) showed good correlation. In this case report, we demonstrate that the mean value of ICP is higher in the posterior fossa, with a strong correlation across the tentorium. All other ICP-derived parameters display a symmetrical profile.

  20. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Pare, V.K.

    1983-01-01

    Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

  1. Research using small tokamaks

    International Nuclear Information System (INIS)

    1993-01-01

    This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs

  2. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  3. 50 years of tokamaks

    Czech Academy of Sciences Publication Activity Database

    Mlynář, Jan; Řípa, Milan

    2008-01-01

    Roč. 2, č. 2 (2008), s. 7-7 ISSN 1818-5355 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * history Subject RIV: BL - Plasma and Gas Discharge Physics http://www.efda.org/news_and_events/downloads/efda_newsletter/nl_2008_12.pdf

  4. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  5. Na cestě k termojaderné fúzi

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2012-01-01

    Roč. 15, č. 6 (2012), s. 18-19 ISSN 1212-1673 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Tore Supra * dark matter * dark energy * chameleons * axions * CERN Subject RIV: BL - Plasma and Gas Discharge Physics

  6. Plasma surface interactions in controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Becoulet, M.; Costanzo, L. [and others

    2000-07-01

    This report brings together all the contributions of EURATOM/CEA association to the 14. international conference on plasma surface interactions in controlled fusion devices. 24 papers are presented and they deal mainly with the ergodic divertor and the first wall of Tore-supra tokamak.

  7. Magnetic fusion; La fusion magnetique

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This document is a detailed lecture on thermonuclear fusion. The basic physics principles are recalled and the technological choices that have led to tokamaks or stellarators are exposed. Different aspects concerning thermonuclear reactors such as safety, economy and feasibility are discussed. Tore-supra is described in details as well as the ITER project.

  8. Magnetic fusion

    International Nuclear Information System (INIS)

    2002-01-01

    This document is a detailed lecture on thermonuclear fusion. The basic physics principles are recalled and the technological choices that have led to tokamaks or stellarators are exposed. Different aspects concerning thermonuclear reactors such as safety, economy and feasibility are discussed. Tore-supra is described in details as well as the ITER project

  9. EURATOM-CEA Association contributions to the 15. I.A.E.A. conference on plasma physics and controlled nuclear fusion research

    International Nuclear Information System (INIS)

    1995-01-01

    Recent results of plasma physics at TORE SUPRA TOKAMAK device are reported. The topics covered are plasma confinement, plasma heating, current drive, radiating layers, transport phenomena and steady-state plasma. 9 papers have been separately indexed for the INIS database. (K.A.)

  10. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

    1980-06-01

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  11. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

    1981-08-01

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  12. Density limits in Tokamaks

    International Nuclear Information System (INIS)

    Tendler, M.

    1984-06-01

    The energy loss from a tokamak plasma due to neutral hydrogen radiation and recycling is of great importance for the energy balance at the periphery. It is shown that the requirement for thermal equilibrium implies a constraint on the maximum attainable edge density. The relation to other density limits is discussed. The average plasma density is shown to be a strong function of the refuelling deposition profile. (author)

  13. Energy confinement in tokamaks

    International Nuclear Information System (INIS)

    Sugihara, M.; Singer, C.

    1986-08-01

    A straightforward generalization is made of the ohmic heating energy confinement scalings of Pfeiffer and Waltz and Blackwell et. al. The resulting model is systematically calibrated to published data from limiter tokamaks with ohmic, electron cyclotron, and neutral beam heating. With considerably fewer explicitly adjustable free parameters, this model appears to give a better fit to the available data for limiter discharges than the combined ohmic/auxiliary heating model of Goldston

  14. TPX tokamak construction management

    International Nuclear Information System (INIS)

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-01-01

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly

  15. Temaet folkemord i forfatterskapet til Tore Linné Eriksen

    Directory of Open Access Journals (Sweden)

    Randi Rønning Balsvik

    2015-10-01

    Full Text Available This article points to the responsibility historians have in the formation of what we may call the collective memory of persons, groups and states. What are the images of “the other” conveyed in textbooks and media? In Norway, the historian Tore Linné Eriksen has – more than any other scholar – used his research, writings and enormous capacity for work to educate students, youth and the public in general and to create a more just image of “the other”. His driving force has been an extraordinary ability to be at the forefront in spotting international research, as well as his sense of justice and respect for the non-western world. His last extensive work on global history (Globalhistorie 1750–1900 clearly demonstrates these capabilities.In the early 1980s, long before the wave of genocide studies after the Rwanda catastrophe of 1994, Eriksen’s two works on Namibia –Namibia: Kolonialisme, apartheid og frigjøringskamp i det sørlige Afrika (1982 and The Political Economy of Namibia: An Annotated Critical Bibliography (1985 – used the concept of genocide to describe German conduct in Namibia in the early 20th century. In 2007, Eriksen published his book about what he calls the first genocide of the 20th century, Det første folkemordet i det tjuende århundret. Namibia 1903–1908. German and African history is woven into the question of whether the dangerous relations that developed between German settlers and Africans can be labelled genocide. The present article attempts to present Eriksen’s arguments. An introductory section deals with the trends in the international research literature and establishes a link between colonialism and genocide. In 2008, Eriksen’s article on the extinction of the Herero people in Namibia – “Utslettelse av Hererofolket i Namibia 1903– 1908” – was published in Bernt Hagtvets collection of articles, Folkemordenes svarte bok. The following year, Tore Linné Eriksen’s article in the

  16. Tore : [luuletused] / Betti Alver ; [Impressioonid] : Viiu Härm

    Index Scriptorium Estoniae

    Alver, Betti, 1906-1989

    2004-01-01

    Sisu: Tore ; Algav päev ; Allikal ; Maru 1 ; Lähen müüjaks ; Noorus ; Tulipunane vihmavari ; Pedja ; Äkki ilm läks sulale ; Läbi lillede ; Suvi ; Sõnarine ; Ja see oli kõik ; Asjad ; Muusale ; Hambad ; "Vaim, kandes kord triumfipärgi..." ; Hääled ; Vilepuhuja ; Tüli ; Helde andja ; Pühapäevalaps ; Froufrou ; Vooruse võlu ; Sügis 1 ; Talv ; Kuuljale ; Meistrile ; Ei vaibu ; Ilus õde ; Kahepaikne ; Kannibal ; Tuhm kalender ; Raudne taevas ; Koguja ; Inimhetk ; Su nägu ; Läkitus ; Kaks saarlast ; Hing ; Kuus kildu ; Räägi tasa minuga ; Lõppude lõpuks ; Kaduv käsi ; Vana teater ; Raudsed närvid ; Linna taga ; Külm puhang lõõtsub ; Rätsep mure ; Mina ise ; Puust palitu ; Kerjus ; Öölaul ; Pilvele ; Põdural põllul ; Üle tuhande tõkke ; Päikeses ; "Ma nägin und: mind jättis jumal maha..." ; Uni ; Kiivas kuu ; Viimne soov ; Raudahjus põlesid puud ; Linnud naersid ; Arbujate aegu 1 ; Arbujate aegu 2 ; Korallid Emajões ; Udus ; Võlg ; Raugad ; Pähklikoor ; Ekstaas ; "Maailma saatust alati..." ; Tige valgus ; "Selle ilma igav kainus..." ; Sidemed ; "Jäägu teistele alandlik jaatus..." ; Mu juurde voogas ; Suurest haardest ; Laulik ; Suur Nimetu ; Ei tea ; "Mulle meenuvad kauged hommikud..." ; Peegel ; Ime 1936 ; Võrdlus ; Ime 1988 ; Suured voogajad ; Elul on väikene hingemaa ; Ebausklik ; Priiskaja ; Tähetund ; Elu on alles uus ; Kuristikulill ; Süda ; Sa hapramast hapram ; Eluhelbed ; Lootus ; Kes oli su kaitsja ; Lehekuu lumi ; Sügis 2

  17. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    NARCIS (Netherlands)

    Box, F. M. A.; Howard, J.; VandeKolk, E.; Meijer, F. G.

    1997-01-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines.

  18. Extension of an existing control and monitoring system: architecture 7

    International Nuclear Information System (INIS)

    Soulabaille, Y.

    1991-01-01

    Tore Supra Tokamak is controlled by Architecture 7. This system comprises 3 levels: Man-machine system, automatism management and exchanges with the plant. Performing it presents, nevertheless some limitations: time response is only half a second allowing to manage 95% of Tore Supra processes, the remaining 5% requires one millisecond. The first aim is the extension of functionalities by a fast automat giving one microsecond cycle. The fast automat is applied to the poloidal field. Of main concern for fusion experiments it allows the creation of a plasma current. The second aim is the possibility to use softwares found on the computer market [fr

  19. Introduction of artificial intelligence techniques for computerized management of defects in an industrial process

    International Nuclear Information System (INIS)

    Utzel, N.

    1991-06-01

    An optimized management of Tore Supra Tokamak requires a computerized defect management. The aim is the analysis of an inhibited situation not corrected by automatisms of the process and that can be handled only by human intervention. The operator should understand, make a diagnosis and act to restore the system. In this report are studied an expert system helping the operator to analyze defects of the two main cooling loops (decarbonated water and pressurized water), management of the history of malfunction and recording of diagnosises, elaboration of an adapted expert model and installation of a methodology for defect management in other processes of Tore Supra [fr

  20. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1994-01-01

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  1. The density limit in Tokamaks

    International Nuclear Information System (INIS)

    Alladio, F.

    1985-01-01

    A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs

  2. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

  3. Nonlinear gyrokinetic tokamak physics

    International Nuclear Information System (INIS)

    Brizard, A.J.

    1990-01-01

    The gyrokinetic reduced description of low-frequency and small-perpendicular-wavelength nonlinear tokamak dynamics is presented in three different versions: the reduced dynamical description of test particles moving in electromagnetic fields; the reduced gyrokinetic description of the self-consistent interaction of particles and fields through the Maxwell-Vlasov equations; and the reduced description of nonlinear fluid motion. The unperturbed tokamak plasma is described in terms of a noncanonical Hamiltonian guiding-center theory. The unperturbed guiding-center tokamak plasma is then perturbed by gyrokinetic electromagnetic fields and consequently the perturbed guiding-center dynamical system acquires new gyrophase dependence. The perturbation analysis that follows makes extensive use of Lie-transform perturbation techniques. Because the electromagnetic perturbations affect both the Hamiltonian and the Poisson-bracket structure, the Phase-space Lagrangian Lie perturbation method is used. The description of the reduced test-particle dynamics is given in terms of a non-canonical Hamiltonian gyrocenter theory. The description of the reduced kinetic dynamics is concerned with the self consistent response of the guiding-center plasma and is described in terms of the nonlinear gyrokinetic Maxwell-Vlasov equations. It is also shown that the gyrokinetic Maxwell-Vlasov system possesses a gyrokinetic energy adiabatic invariant and that, in both the linear and nonlinear (quadratic) approximations, the corresponding energy invariants are exact. The description of the reduced fluid dynamics is concerned with the derivation of a closed set of reduced fluid equations. Three generations of reduced fluid models are presented: the reduced MHD equations; the reduced FLR-MHD equations; and the gyrofluid equations

  4. Tokamak instrumentation and controls

    Energy Technology Data Exchange (ETDEWEB)

    Becraft, W. R.; Bettis, E. S.; Houlberg, W. A.; Onega, R. J.; Stone, R. S.

    1979-02-01

    The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine.

  5. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System.

  6. Tokamak instrumentation and controls

    International Nuclear Information System (INIS)

    Becraft, W.R.; Bettis, E.S.; Houlberg, W.A.; Onega, R.J.; Stone, R.S.

    1979-02-01

    The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine

  7. The TFR-600 Tokamak

    International Nuclear Information System (INIS)

    1977-11-01

    The new step of the Tokamak TFR, TFR 600, is described with its different aspects: physical objectives, modifications of the vacuum chamber and of the poloidal circuit, additionnal heatings. The nominal characteristics are: R=98 cm; a 0 or D 0 at 40 keV (power transmitted to the plasma); - ion cyclotron radiofrequency heating: 600 kW in the bandwidth 55-83 MHz; - and cluster injection: 100 KW at 600 keV (average mass of the H 0 clusters: 100-200 A.MU) [fr

  8. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    Minardi, E.

    1989-01-01

    The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

  9. Tokamak burn control

    International Nuclear Information System (INIS)

    Sager, G.T.

    1988-06-01

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  10. Supra-personal cognitive control and metacognition

    Science.gov (United States)

    Shea, Nicholas; Boldt, Annika; Bang, Dan; Yeung, Nick; Heyes, Cecilia; Frith, Chris D.

    2014-01-01

    The human mind is extraordinary in its ability not merely to respond to events as they unfold but also to adapt its own operation in pursuit of its agenda. This ‘cognitive control’ can be achieved through simple interactions among sensorimotor processes, and through interactions in which one sensorimotor process represents a property of another in an implicit, unconscious way. So why does the human mind also represent properties of cognitive processes in an explicit way, enabling us to think and say ‘I’m sure’ or ‘I’m doubtful’? We suggest that ‘system 2 metacognition’ is for supra-personal cognitive control. It allows metacognitive information to be broadcast, and thereby to coordinate the sensorimotor systems of two or more agents involved in a shared task. PMID:24582436

  11. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  12. Topology of tokamak orbits

    International Nuclear Information System (INIS)

    Rome, J.A.; Peng, Y.K.M.

    1978-09-01

    Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well

  13. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  14. Canonical profiles in tokamaks

    International Nuclear Information System (INIS)

    Dnestrovskij, Yu.N.

    2002-01-01

    We consider the problem of the canonical profiles for tokamak plasma with arbitrary cross-section, taking into account two principles: 1) the free plasma energy minimum with the constraint of total current conservation and 2) the profile consistency. We deduce the Euler differential equation for the canonical profile of μ=1/q with two types of the boundary conditions: soft and stiff. The soft conditions correspond to the Kadomtsev solution for the circular cylinder. The stiff conditions describe a fast response of the plasma over the whole cross-section on the edge impact. Using the canonical profile of the current density, we calculate the critical gradients for the temperature, and create the transport model for the electron and ion temperatures and density. We show that, when the aspect ratio is diminished, or when the elongation increases, the canonical profiles become flatten. The similar tendency for the real profiles of the electron temperature was found in analysis of JET and START experiments. The obtained critical gradients were used to analysis of the experiments in tokamaks with moderate and tight aspect ratios. (author)

  15. An Amylase-Responsive Bolaform Supra-Amphiphile.

    Science.gov (United States)

    Kang, Yuetong; Cai, Zhengguo; Tang, Xiaoyan; Liu, Kai; Wang, Guangtong; Zhang, Xi

    2016-02-01

    An amylase-responsive bolaform supra-amphiphile was constructed by the complexation between β-cyclodextrin and a bolaform covalent amphiphile on the basis of host-guest interaction. The bolaform covalent amphiphile could self-assemble in solution, forming sheet-like aggregates and displaying weak fluorescence because of aggregation-induced quenching. The addition of β-cyclodextrin led to the formation of the bolaform supra-amphiphile, prohibiting the aggregation of the bolaform covalent amphiphile and accompanying with the significant recovery of fluorescence. Upon the addition of α-amylase, with the degradation β-cyclodextrin, the fluorescence of the supra-amphiphile would quench gradually and significantly, and the quenching rate linearly correlated to the concentration of α-amylase. This study enriches the field of supra-amphiphiles on the basis of noncovalent interactions, and moreover, it may provide a facile way to estimate the activity of α-amylase.

  16. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-05-01

    The technical reports in this document were presented at the IAEA Technical Committee Meeting ''Research on Small Tokamaks'', September 1990, in three sessions, viz., (1) Plasma Modes, Control, and Internal Phenomena, (2) Edge Phenomena, and (3) Advanced Configurations and New Facilities. In Section (1) experiments at controlling low mode number modes, feedback control using external coils, lower-hybrid current drive for the stabilization of sawtooth activity and continuous (1,1) mode, and unmodulated and fast modulated ECRH mode stabilization experiments were reported, as well as the relation to disruptions and transport of low m,n modes and magnetic island growth; static magnetic perturbations by helical windings causing mode locking and sawtooth suppression; island widths and frequency of the m=2 tearing mode; ultra-fast cooling due to pellet injection; and, finally, some papers on advanced diagnostics, i.e., lithium-beam activated charge-exchange spectroscopy, and detection through laser scattering of discrete Alfven waves. In Section (2), experimental edge physics results from a number of machines were presented (positive biasing on HYBTOK II enhancing the radial electric field and improving confinement; lower hybrid current drive on CASTOR improving global particle confinement, good current drive efficiency in HT-6B showing stabilization of sawteeth and Mirnov oscillations), as well as diagnostic developments (multi-chord time resolved soft and ultra-soft X-ray plasma radiation detection on MT-1; measurements on electron capture cross sections in multi-charged ion-atom collisions; development of a diagnostic neutral beam on Phaedrus-T). Theoretical papers discussed the influence of sheared flow and/or active feedback on edge microstability, large edge electric fields, and two-fluid modelling of non-ambipolar scrape-off layers. Section (3) contained (i) a proposal to construct a spherical tokamak ''Proto-Eta'', (ii) an analysis of ultra-low-q and runaway

  17. JUST: Joint Upgraded Spherical Tokamak

    International Nuclear Information System (INIS)

    Azizov, E.A.; Dvorkin, N.Ya.; Filatov, O.G.

    1997-01-01

    The main goals, ideas and the programme of JUST, spherical tokamak (ST) for the plasma burn investigation, are presented. The place and prospects of JUST in thermonuclear investigations are discussed. (author)

  18. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.; Saramad, S.

    2012-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

  19. Alcator C-Mod Tokamak

    Data.gov (United States)

    Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...

  20. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

    2003-12-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  1. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group

    2003-01-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  2. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes

    2003-01-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  3. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

    2003-07-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  4. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    McWilliams, R.

    1988-01-01

    The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  5. The ETE spherical Tokamak project

    International Nuclear Information System (INIS)

    Ludwig, Gerson Otto; Andrade, Maria Celia Ramos de; Barbosa, Luis Filipe Wiltgen

    1999-01-01

    This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)

  6. The ETE spherical Tokamak project

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Andrade, Maria Celia Ramos de; Barbosa, Luis Filipe Wiltgen [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] [and others]. E-mail: ludwig@plasma.inpe.br

    1999-07-01

    This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)

  7. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  8. Plasma turbulence in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)

  9. The Varennes tokamak

    International Nuclear Information System (INIS)

    Gregory, B.C.; Bolton, R.A.; Pacher, H.D.

    1983-01-01

    This article is a progress report on the Varennes Tokamak (TdeV), which is the main element in the Canadian research program on magnetic confinement fusion. The project is led by a group of five institutions: the Hydro-Quebec Research Institute (IREQ), the National Research Council - Energy, the University of Montreal, CANATOM Ltd., and MPB Technologies Inc. The TdeV will cost about 40 million dollars and will be built in a large hall at the IREQ high energy laboratory in Varennes. Operation in a quasi-stationary regime has been adopted as one of the primary research areas for the TdeV. First plasma is expected at the end of 1984 [fr

  10. High Beta Tokamak research

    International Nuclear Information System (INIS)

    Navratil, G.A.; Mauel, M.E.; Ivers, T.H.; Sankar, M.K.V.; Eisner, E.; Gates, D.; Garofalo, A.; Kombargi, R.; Maurer, D.; Nadle, D.; Xiao, Q.

    1993-01-01

    During the past 6 months, experiments have been conducted with the HBT-EP tokamak in order to (1) test and evaluate diagnostic systems, (2) establish basic machine operation, (3) document MHD behavior as a function of global discharge parameters, (4) investigate conditions leading to passive stabilization of MHD instabilities, and (5) quantify the external saddle coil current required for DC mode locking. In addition, the development and installation of new hardware systems has occurred. A prototype saddle coil was installed and tested. A five-position (n,m) = (1,2) external helical saddle coil was attached for mode-locking experiments. And, fabrication of the 32-channel UV tomography and the multipass Thomson scattering diagnostics have begun in preparation for installation later this year

  11. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-09-01

    A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  12. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-01-01

    A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  13. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  14. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  15. Combined supra/infratentorial approach to tentorial meningiomas Abordagem combinada supra e infratentorial aos meningiomas tentoriais

    Directory of Open Access Journals (Sweden)

    Igor de Castro

    2005-03-01

    Full Text Available In spite of significant advancements in imaging technology, monitoring, and microsurgical techniques, complete and safe removal of tentorial meningiomas remains a challenge for most neurosurgeons. Classifications of tentorial meningiomas are revised. The combined supra/infratentorial approach to resects tentorial meningioma is discussed. This approach provides a wider exposure of the supra/infratentorial region with less brain retraction. With this approach the occipital lobe and the cerebellum are exposed along the tentorium. Two illustrative cases are presented. The patients were studied with computerized tomography, magnetic resonance and angiography. The anatomy of the transverse sinus and the confluence of the sinus could be appreciated with these studies. The operative technique is described stepwise. Emphasis is placed on pre-operative evaluation and surgical technique, leading to a total surgical removal of the lesion with margins of safety. The goal of surgical treatment of tentorial meningiomas is their complete and safe removal. With this unique approach we sought to confirm that it offers a safe means of resection not only the neoplasm but also the infiltrated dura.Apesar dos significativos avanços na tecnologia de imagens, nas técnicas de monitorização e microcirúrgicas, a ressecção completa e segura dos meningiomas tentoriais permanece um desafio para maioria dos neurocirurgiões. A abordagem supra e infra-tentorial proporciona ampla exposição das regiões supra e infratentoriais diminuindo a retração cerebral. Com esse tipo de abordagem o lobo occiptal e o cerebelo são expostos ao longo da superficie tentorial. Dois casos ilustrativos são apresentados. Os pacientes foram avaliados com tomografia computadorizada, ressonância magnética e angiografia, o que permitiu estudar a anatomia do seio transverso, a confluência e dominancia dos seios. O objetivo do tratamento cirúrgico dos meningiomas tentoriais é a remo

  16. supraHex: An R/Bioconductor package for tabular omics data analysis using a supra-hexagonal map☆

    Science.gov (United States)

    Fang, Hai; Gough, Julian

    2014-01-01

    Biologists are increasingly confronted with the challenge of quickly understanding genome-wide biological data, which usually involve a large number of genomic coordinates (e.g. genes) but a much smaller number of samples. To meet the need for data of this shape, we present an open-source package called ‘supraHex’ for training, analysing and visualising omics data. This package devises a supra-hexagonal map to self-organise the input data, offers scalable functionalities for post-analysing the map, and more importantly, allows for overlaying additional data for multilayer omics data comparisons. Via applying to DNA replication timing data of mouse embryogenesis, we demonstrate that supraHex is capable of simultaneously carrying out gene clustering and sample correlation, providing intuitive visualisation at each step of the analysis. By overlaying CpG and expression data onto the trained replication-timing map, we also show that supraHex is able to intuitively capture an inherent relationship between late replication, low CpG density promoters and low expression levels. As part of the Bioconductor project, supraHex makes accessible to a wide community in a simple way, what would otherwise be a complex framework for the ultrafast understanding of any tabular omics data, both scientifically and artistically. This package can run on Windows, Mac and Linux, and is freely available together with many tutorials on featuring real examples at http://supfam.org/supraHex. PMID:24309102

  17. Beta limits of a completely bootstrapped tokamak

    International Nuclear Information System (INIS)

    Weening, R.H.; Bondeson, A.

    1992-03-01

    A beta limit is given for a completely bootstrapped tokamak. The beta limit is sensitive to the achievable Troyon factor and depends directly upon the strength of the tokamak bootstrap effect. (author) 16 refs

  18. Bibliography of fusion product physics in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Hively, L. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sigmar, D. J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1989-09-01

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category.

  19. Asymmetric and symmetric bolaform supra-amphiphiles: formation of imine bond influenced by aggregation.

    Science.gov (United States)

    Wang, Guangtong; Wu, Guanglu; Wang, Zhiqiang; Zhang, Xi

    2014-02-18

    A series of bolaform supra-amphilphiles with different symmetries were fabricated through dynamic benzoic imine bond formation. The pH dependence of imine formations of these supra-amphiphiles were characterazied. We found that the extent of the imine formation of these supra-amphiphies were different. The supra-amphiphiles with a poorer symmetry always exhibited a lower imine formation at a given pH. Therefore, the varied extent of imine bond formation indicate the different aggregations of these supra-amphilphiles, which are controlled by the molecular symmetry of the supra-amphiphiles.

  20. Supra Arcade Downflows in the Earth's Magnetotail

    Science.gov (United States)

    Kobelski, A.; Savage, S. L.; Malaspina, D.

    2017-12-01

    Pinpointing the location of a single reconnection event in the corona is difficult due to observational constraints, although features directly resulting from this rapid reconfiguration of the field lines can be observed beyond the reconnection site. One set of such features are outflows in the form of post-reconnection loops, which have been linked to observations of supra-arcade downflows (SADs). SADs appear as sunward-traveling, density-depleted regions above flare arcades that develop during long duration eruptions. The limitations of remote sensing methods inherently results in ambiguities regarding the interpretation of SAD formation. Of particular interest is how these features are related to post-reconnection retracting magnetic field lines. In planetary magnetospheres, similar events to solar flares occur in the form of substorms, where reconnection in the anti-sunward tail of the magnetosphere causes field lines to retract toward the planet. Using data from the Time History of Events and Macroscopic Interactions during Substorms (THEMIS), we compare one particular aspect of substorms, dipolarization fronts, to SADs. Dipolarization fronts are observed as rapid but temporary changes in the magnetic field of the magnetotail plasma sheet into a more potential-like dipolar shape. These dipolarization fronts are believed to be retracting post-reconnection field lines. We combine data sets to show that the while the densities and magnetic fields involved vary greatly between the regimes, the plasma βs and Alfvén speeds are similar. These similarities allow direct comparison between the retracting field lines and their accompanying wakes of rarified plasma observed with THEMIS around the Earth to the observed morphological density depletions visible with XRT and AIA on the Sun. These results are an important source of feedback for models of coronal current sheets.

  1. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  2. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  3. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.

    2001-01-01

    A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

  4. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (n e and T e ) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  5. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1994-01-01

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  6. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  7. Tokamak SST-1: an over-view

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2002-01-01

    Steady State Tokamak SST-1 is in advanced stage of fabrication at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak with superconducting magnets. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas for 1000 s with significant elongation (K) and triangularity (δ). The choice of the parameters is dictated by the physics and technology goals viz. (a) to control and study strongly shaped single and double null divertor plasma, (b) explore advanced tokamak plasma regimes, (c) steady state particle and heat removal from the device, (d) design and operation of large volume superconducting magnets, (e) non-inductive steady state current drive, (f) methods of plasma heating and (g) material technologies

  8. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  9. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...

  10. Tokamak Plasmas: Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...

  11. Investigation of impurity confinement in lower hybrid wave heated plasma on EAST tokamak

    Science.gov (United States)

    Xu, Z.; Wu, Z. W.; Zhang, L.; Gao, W.; Ye, Y.; Chen, K. Y.; Yuan, Y.; Zhang, W.; Yang, X. D.; Chen, Y. J.; Zhang, P. F.; Huang, J.; Wu, C. R.; Morita, S.; Oishi, T.; Zhang, J. Z.; Duan, Y. M.; Zang, Q.; Ding, S. Y.; Liu, H. Q.; Chen, J. L.; Hu, L. Q.; Xu, G. S.; Guo, H. Y.; the EAST Team

    2018-01-01

    The transient perturbation method with metallic impurities such as iron (Fe, Z  =  26) and copper (Cu, Z  =  29) induced in plasma-material interaction (PMI) procedure is used to investigate the impurity confinement characters in lower hybrid wave (LHW) heated EAST sawtooth-free plasma. The dependence of metallic impurities confinement time on plasma parameters (e.g. plasma current, toroidal magnetic field, electron density and heating power) are investigated in ohmic and LHW heated plasma. It is shown that LHW heating plays an important role in the reduction of the impurity confinement time in L-mode discharges on EAST. The impurity confinement time scaling is given as 42IP0.32Bt0.2\\overline{n}e0.43Ptotal-0.4~ on EAST, which is close to the observed scaling on Tore Supra and JET. Furthermore, the LHW heated high-enhanced-recycling (HER) H-mode discharges with ~25 kHz edge coherent modes (ECM), which have lower impurity confinement time and higher energy confinement time, provide promising candidates for high performance and steady state operation on EAST.

  12. Adjoint optimization scheme for lower hybrid current rampup and profile control in Tokamak

    International Nuclear Information System (INIS)

    Litaudon, X.; Moreau, D.; Bizarro, J.P.; Hoang, G.T.; Kupfer, K.; Peysson, Y.; Shkarofsky, I.P.; Bonoli, P.

    1992-12-01

    The purpose of this work is to take into account and study the effect of the electric field profiles on the Lower Hybrid (LH) current drive efficiency during transient phases such as rampup. As a complement to the full ray-tracing / Fokker Planck studies, and for the purpose of optimization studies, we developed a simplified 1-D model based on the adjoint Karney-Fisch numerical results. This approach allows us to estimate the LH power deposition profile which would be required for ramping the current with prescribed rate, total current density profile (q-profile) and surface loop voltage. For rampup optimization studies, we can therefore scan the whole parameter space and eliminate a posteriori those scenarios which correspond to unrealistic deposition profiles. We thus obtain the time evolution of the LH power, minor radius of the plasma, volt-second consumption and total energy dissipated. Optimization can thus be performed with respect to any of those criteria. This scheme is illustrated by some numerical simulations performed with TORE-SUPRA and NET/ITER parameters. We conclude with a derivation of a simple and general scaling law for the flux consumption during the rampup phase

  13. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    1975-06-01

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  14. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...wavelength polariza- tion field produced by the curvature and field gradient drifts [15]. The growth rate is y = Vs[k/R 2 = [T(eV)/X(cm)J 2 3.3 x 105 sec

  15. TECHNOLOGIES TO OPTIMIZE ADVANCED TOKAMAK

    Energy Technology Data Exchange (ETDEWEB)

    SIMONEN, TC

    2004-01-01

    OAK-B135 Commercial fusion power systems must operate near the limits of the engineering systems and plasma parameters. Achieving these objectives will require real time feedback control of the plasma. This paper describes plasma control systems being used in the national DIII-D advanced tokamak research program.

  16. Prospects for Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.

    1995-01-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

  17. Tokamak impurity-control techniques

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1980-01-01

    A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)

  18. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254 ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  19. An enhanced tokamak startup model

    Science.gov (United States)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

    The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

  20. Computer predictions for future Tokamaks

    International Nuclear Information System (INIS)

    Duechs, D.F.

    1978-01-01

    Proceeding from a reasonable agreement with existing experimental results, this lecture presents radial particle and energy transport computations which extrapolate to large (up to reactor dimensions) future Tokamaks. Special consideration is given to the behavior of alpha-particles, the influence of high-z impurities, and the thermal stability of the plasma

  1. Supra-National Organisations and Conflict Resolution during the ...

    African Journals Online (AJOL)

    More importantly, due to the various crises that plagued the world ranging from civil wars to border clashes etc., the importance or roles of Supra-national organisations in conflict resolution through mediatory diplomacy or otherwise cannot be over-emphasised. This is especially geared towards making the World a Haven of ...

  2. Supra-auricular versus Sinusectomy Approaches for Preauricular Sinuses.

    Science.gov (United States)

    El-Anwar, Mohammad Waheed; ElAassar, Ahmed Shaker

    2016-10-01

    Introduction  Several surgical techniques and modifications have been described to reduce the high recurrence rate after excision of preauricular sinus. Objectives  The aim of this study is to review the literature regarding surgical approaches for preauricular sinus. Data Synthesis  We performed searches in the LILACS, MEDLINE, SciELO, PubMed databases and Cochrane Library in September, 2015, and the key words used in the search were "preauricular sinus," "sinusectomy," "supra-auricular approach," "methylene blue," and/or "recurrence." We revised the results of 17 studies, including 1270 preauricular sinuses that were surgically excised by sinusectomy in 937 ears and by supra-auricular approach in 333 ears. Recurrence with supra-auricular was 4 (1.3%) while sinusectomy was 76 (8.1%) with significant difference ( p  Supra-auricular approach had significantly less recurrence rate than tract sinusectomy approaches. Thus, it could be regularly chosen as the standard procedure for preauricular sinus excision. As such, it would be helpful for surgeons to be familiar with this approach.

  3. Supra-transumbilical laparotomy (STL) approach for small bowel ...

    African Journals Online (AJOL)

    Background: Supra-Transumbilical Laparotomy (STL) has been used in paediatric surgery for a broad spectrum of abdominal procedures. We report our experience with STL approach for small bowel atresia repair in newborns and review previous published series on the topic. Patients and Methods: Fourteen patients with ...

  4. Acometimento da supra-renal associado à paracoccidioidomicose

    Directory of Open Access Journals (Sweden)

    Leonardo Maurício Diniz

    1988-12-01

    Full Text Available A supra-renal foi estudada em 60pacientes com paracoccidioidomicose. Dentre eles, 10(16,7% apresentavam alterações anatômicas ou funcionais das supra-renais. As lesões glandulares associaram-se à paracoccidioidomicose disseminada, com evolução da doença de pelo menos cinco anos sem tratamento e com hipotensão arterial sistêmica. Não houve associação entre alterações anatomoclínicas supra-renais e alterações raáiolôgicaspulmonares. Esses dados revelam a importância da avaliação sistemática da função supra-renal em portadores de formas disseminadas da paracoccidioidomicose.Sixty patients with paracoccidioidomycosis (PCM were studied, with special regard to adrenal involvement. Ten patients (16.7% had adrenal ab-normalities associated with hypotension, a disseminated form of the disease and more than five years of evolution without treatment. No statistical association between adrenal anatomo-clinical alterations and radiological pulmonary abnormalities was observed. These data point out to the importance of systematic evaluation of adrenal function in patients with disseminated PCM.

  5. Myositis Ossificans Circumscripta of the Supra-orbital Region: A ...

    African Journals Online (AJOL)

    ANNALS

    Abstract. Myositis ossificans circumscripta is a pathological condition characterized by formation of bony tissue within the skeletal muscles following repeated trauma. A case of myositis ossificans circumscripta of the supra-orbital region in a 25-year-old man is presented and the pertinent literature is reviewed. To the best.

  6. Supra-Ethnic Nationalism: The Case of Eritrea | Bereketeab | African ...

    African Journals Online (AJOL)

    African Sociological Review / Revue Africaine de Sociologie. Journal Home · ABOUT THIS JOURNAL · Advanced Search · Current Issue · Archives · Journal Home > Vol 6, No 2 (2002) >. Log in or Register to get access to full text downloads. Username, Password, Remember me, or Register. Supra-Ethnic Nationalism: The ...

  7. Simulation of sub-molecular and supra-molecular fluids

    NARCIS (Netherlands)

    Frenkel, D.

    1991-01-01

    Computer simulations indicate that many forms of liquid crystalline order in lyotropic systems may be due to simple excluded volume effects. Yet, there is more to liquid crystalline ordering than simple hard-core repulsion. In order to understand liquid crystalline order in supra-molecular systems

  8. Supra-annular mitral valve replacement in children.

    Science.gov (United States)

    Kanter, Kirk R; Kogon, Brian E; Kirshbom, Paul M

    2011-12-01

    Despite improved mitral repair techniques, some children need mitral valve replacement (MVR). Due to small annulus size, supra-annular MVR is useful. From 2003 to 2010, 15 children had 23 supra-annular MVRs. At first supra-annular MVR, median age was 6.5 months (28 days to 47 months); median weight was 5.4 kg (3.3-11.8 kg). Twelve (80%) had prior operations, 8 (53%) had previous mitral repair. Eight had congenital mitral anomalies (4 with Shone's), 5 had atrioventricular septal defects, 1 had endocarditis, and 1 had a repaired anomalous left coronary artery. All primary MVRs used mechanical valves (≤ 17 mm in 9 patients). There was one early death (93% survival) in an 11-month-old with congenital pulmonary vein stenosis. One intraoperative conversion from annular to supra-annular MVR developed heart block. Three pacemakers were implanted for supraventricular rhythm disturbances. Three children had valve thrombosis early postoperatively treated medically. On follow-up of 4.3 ± 2.8 years, 8 had reoperation including redo MVR in 6 for pannus formation or thrombus (1 had three redo MVRs). At redo, a larger valve was used in 5 and a bioprosthetic valve in 4 patients. There was one late death after third redo MVR with pulmonary vein stenosis relief (overall survival 87%). Supra-annular MVR is useful for children with a small annulus. Operative survival is good with infrequent heart block. Complications are common, including redo MVR and need for left ventricular outflow tract obstruction relief. Pulmonary vein stenosis is a marker for poor outcome; all patients without pulmonary vein stenosis survive long term. Copyright © 2011 The Society of Thoracic Surgeons. Published by Elsevier Inc. All rights reserved.

  9. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-01-01

    This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

  10. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

    LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

    1988-04-01

    The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

  11. Engineering analysis of new Brazilian Tokamak

    International Nuclear Information System (INIS)

    Tuszel, A.G.

    1990-01-01

    The engineering basic headlines are described. A project for the construction of a new tokamak is being developed at the Institute of Physics, University of Sao Paulo. The tokamak named TBR-II will be a medium size tokamak using two high power generators of 15 MW each and concepted as a versatile device for plasma physics research of interest for thermonuclear fusion studies. (Author)

  12. A numerical study of tokamak edge turbulence

    International Nuclear Information System (INIS)

    Hu Shuanghui; Huang Lin; Qiu Xiaoming

    1993-01-01

    The tokamak edge turbulence which contains resistivity and impurity gradients and impurity radiation driven sources is studied numerically. The effect of ohmic dissipation on the evolution and saturation of this turbulence is investigated. The ohmic effect drops the saturation levels of fluctuations efficiently in high density tokamaks (such as Alcator), indicating that the ohmic effect plays an important role in the evolution of tokamak edge turbulence in high density devices

  13. Microwave Tokamak Experiment: Overview and status

    International Nuclear Information System (INIS)

    1990-05-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs

  14. Comprehensive numerical modelling of tokamaks

    International Nuclear Information System (INIS)

    Cohen, R.H.; Cohen, B.I.; Dubois, P.F.

    1991-01-01

    We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell

  15. Advanced commercial Tokamak optimization studies

    International Nuclear Information System (INIS)

    Whitley, R.H.; Berwald, D.H.; Gordon, J.D.

    1985-01-01

    Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units

  16. Flux driven turbulence in tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Ghendrih, P.; Ottaviani, M.; Sarazin, Y.; Beyer, P.; Benkadda, S.; Waltz, R.E.

    1999-01-01

    This work deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts, are usually observed over a broad range of time and spatial scales. The existence of these fronts provide a way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyroBohm type in spite of these large scale transport events. Some departure from the gyroBohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux than at fixed temperature gradient, with the same time averaged profile. (author)

  17. Starfire: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

    1979-01-01

    The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

  18. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Kieras, C.E.

    1982-12-01

    Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma

  19. Equilibrium Reconstruction in EAST Tokamak

    International Nuclear Information System (INIS)

    Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)

  20. Dipole Map For Divertor Tokamaks

    International Nuclear Information System (INIS)

    Ali, Halima; Punjabi, Alkesh; Boozer, Allen

    2003-01-01

    Heat flux impinging on the collector plates of divertor tokamaks can be prodigious. Therefore, the problem of spreading the heat flux on plates is a crucial issue for divertor tokamaks such as ITER. Here we use method of maps /1,2/ to investigate this problem. Magnetic field lines in non-axisymmetric divertor tokamaks are a one and a half degree of freedom Hamiltonian system /1-3/. We represent the unperturbed magnetic topology by the Symmetric Simple Map (SSM) /4/ given by yn+1 = yn + 2kxn - 2k2yn (1 - yn), xn+1 = xn - kyn (1 - yn) - 2k2yn+1 (1 - yn+1). The effects of a current carrying coil placed externally across from X-point is represented by Dipole Map (DP) /4,5/ given by x n+1 = x n + 2δs 3 x n+1 (y n - y s + s/[x n+1 2 + (y n - y s + s) 2 ] 2 ), y n+1 = y n + δs 3 x n+1 ((y n - y s + s) 2 - x n+1 2 /[x n+1 2 + (y n - y s + s) 2 ] 2 ) δ is amplitude of high MN magnetic perturbation, s is the distance of coil from last good surface across from X point, and is the y coordinate of last good surface where it crosses the axis joining X point and O point across from X point. We fix k=0.3 and s = (1/2)|y s |. We calculate the increase in width of stochastic layer and area of footprint of field lines on divertor plate as δ is increased. We also calculate how connection length, toroidal and poloidal circuits and their fractal structures, the number, location and density of hot spots change with δ. Finally, we make conclusions about how the heat flux can be possibly controlled and reduced by applying external magnetic perturbation in divertor tokamaks

  1. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

    The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)

  2. Small-angle Scattering Theory Revisited: Photocurrent and Spatial Localization

    Science.gov (United States)

    Basse, N. P.; Zoletnik, S.; Michelsen, P. K.

    2005-01-01

    In this paper theory on collective scattering measurements of electron density fluctuations in fusion plasmas is revisited. We present the first full derivation of the expression for the photocurrent beginning at the basic scattering concepts. Thereafter we derive detailed expressions for the auto- and crosspower spectra obtained from measurements. These are discussed and simple simulations made to elucidate the physical meaning of the findings. In this context, the known methods of obtaining spatial localization are discussed and appraised. Where actual numbers are applied, we utilize quantities from two collective scattering instruments The ALTAIR diagnostic on the Tore Supra tokamak [A Truc et al, "ALTAIR An infrared laser scattering diagnostic on the Tore Supra tokamak", Rev. Sci. Instrum. 63 3716 3724 (1992)] and the LOTUS diagnostic on the Wendelstein 7-AS stellarator [M Saffman et al, "CO2 laser based two-volume collective scattering instrument for spatially localized turbulence measurements", Rev. Sci. Instrum. 72 2579 2592 (2001)].

  3. WEST Physics Basis

    Science.gov (United States)

    Bourdelle, C.; Artaud, J. F.; Basiuk, V.; Bécoulet, M.; Brémond, S.; Bucalossi, J.; Bufferand, H.; Ciraolo, G.; Colas, L.; Corre, Y.; Courtois, X.; Decker, J.; Delpech, L.; Devynck, P.; Dif-Pradalier, G.; Doerner, R. P.; Douai, D.; Dumont, R.; Ekedahl, A.; Fedorczak, N.; Fenzi, C.; Firdaouss, M.; Garcia, J.; Ghendrih, P.; Gil, C.; Giruzzi, G.; Goniche, M.; Grisolia, C.; Grosman, A.; Guilhem, D.; Guirlet, R.; Gunn, J.; Hennequin, P.; Hillairet, J.; Hoang, T.; Imbeaux, F.; Ivanova-Stanik, I.; Joffrin, E.; Kallenbach, A.; Linke, J.; Loarer, T.; Lotte, P.; Maget, P.; Marandet, Y.; Mayoral, M. L.; Meyer, O.; Missirlian, M.; Mollard, P.; Monier-Garbet, P.; Moreau, P.; Nardon, E.; Pégourié, B.; Peysson, Y.; Sabot, R.; Saint-Laurent, F.; Schneider, M.; Travère, J. M.; Tsitrone, E.; Vartanian, S.; Vermare, L.; Yoshida, M.; Zagorski, R.; Contributors, JET

    2015-06-01

    With WEST (Tungsten Environment in Steady State Tokamak) (Bucalossi et al 2014 Fusion Eng. Des. 89 907-12), the Tore Supra facility and team expertise (Dumont et al 2014 Plasma Phys. Control. Fusion 56 075020) is used to pave the way towards ITER divertor procurement and operation. It consists in implementing a divertor configuration and installing ITER-like actively cooled tungsten monoblocks in the Tore Supra tokamak, taking full benefit of its unique long-pulse capability. WEST is a user facility platform, open to all ITER partners. This paper describes the physics basis of WEST: the estimated heat flux on the divertor target, the planned heating schemes, the expected behaviour of the L-H threshold and of the pedestal and the potential W sources. A series of operating scenarios has been modelled, showing that ITER-relevant heat fluxes on the divertor can be achieved in WEST long pulse H-mode plasmas.

  4. Cochlear neuropathy and the coding of supra-threshold sound.

    Science.gov (United States)

    Bharadwaj, Hari M; Verhulst, Sarah; Shaheen, Luke; Liberman, M Charles; Shinn-Cunningham, Barbara G

    2014-01-01

    Many listeners with hearing thresholds within the clinically normal range nonetheless complain of difficulty hearing in everyday settings and understanding speech in noise. Converging evidence from human and animal studies points to one potential source of such difficulties: differences in the fidelity with which supra-threshold sound is encoded in the early portions of the auditory pathway. Measures of auditory subcortical steady-state responses (SSSRs) in humans and animals support the idea that the temporal precision of the early auditory representation can be poor even when hearing thresholds are normal. In humans with normal hearing thresholds (NHTs), paradigms that require listeners to make use of the detailed spectro-temporal structure of supra-threshold sound, such as selective attention and discrimination of frequency modulation (FM), reveal individual differences that correlate with subcortical temporal coding precision. Animal studies show that noise exposure and aging can cause a loss of a large percentage of auditory nerve fibers (ANFs) without any significant change in measured audiograms. Here, we argue that cochlear neuropathy may reduce encoding precision of supra-threshold sound, and that this manifests both behaviorally and in SSSRs in humans. Furthermore, recent studies suggest that noise-induced neuropathy may be selective for higher-threshold, lower-spontaneous-rate nerve fibers. Based on our hypothesis, we suggest some approaches that may yield particularly sensitive, objective measures of supra-threshold coding deficits that arise due to neuropathy. Finally, we comment on the potential clinical significance of these ideas and identify areas for future investigation.

  5. Left Hand Dominance Affects Supra-Second Time Processing

    Science.gov (United States)

    Vicario, Carmelo Mario; Bonní, Sonia; Koch, Giacomo

    2011-01-01

    Previous studies exploring specific brain functions of left- and right-handed subjects have shown variances in spatial and motor abilities that might be explained according to consistent structural and functional differences. Given the role of both spatial and motor information in the processing of temporal intervals, we designed a study aimed at investigating timing abilities in left-handed subjects. To this purpose both left- and right-handed subjects were asked to perform a time reproduction of sub-second vs. supra-second time intervals with their left and right hand. Our results show that during processing of the supra-second intervals left-handed participants sub-estimated the duration of the intervals, independently of the hand used to perform the task, while no differences were reported for the sub-second intervals. These results are discussed on the basis of recent findings on supra-second motor timing, as well as emerging evidence that suggests a linear representation of time with a left-to-right displacement. PMID:22028685

  6. The steady state in toroidal traps

    International Nuclear Information System (INIS)

    Goldston, R.

    1997-01-01

    Experiments at the JET, TORE SUPRA, TFTR and DIII-D reactors have corroborated calculations showing that an advanced tokamak configuration with an important self-generated current, a large plasma pressure and thus a large thermonuclear power density, could allow for the construction of fusion steady state reactors with reduced size and cost. Stellarators only need external superconductive coils for reaching the steady state, but it is essential to reduce in a large proportion the plasma self-generated current

  7. Tokamak plasma position dynamics and feedback control

    International Nuclear Information System (INIS)

    Burenko, L.; Bailey, J.M.

    1983-01-01

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form

  8. Mercier criterion for high-β tokamaks

    International Nuclear Information System (INIS)

    Galvao, R.M.O.

    1984-01-01

    An expression, for the application of the Mercier criterion to numerical studies of diffuse high-β tokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the high-β tokamak approximation is derived. (L.C.) [pt

  9. Magnetic confinement by Tokamak: physical aspects

    International Nuclear Information System (INIS)

    Tachon, J.

    1980-01-01

    After describing the Tokamak configuration concept, the author provides an analysis of the principal physical aspects of this type of installation and concludes by estimating that the Tokamak concept is a 'plausible candidate' as a means of producing controlled thermonuclear fusion [fr

  10. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  11. TGV, hutě a tokamak ITER

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    Leden (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * TGV * Pulse Power Electrical Network * Steady State Electrical Network Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/1954-tgv-hute-a-tokamak-iter

  12. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    Im, K.H.; Cho, S.; Her, N.I.

    2001-01-01

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  13. Plasma detachment in divertor tokamaks

    Science.gov (United States)

    Leonard, A. W.

    2018-04-01

    Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma E× B drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.

  14. The microwave Tokamak experiment (MTX)

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Cohen, B.I.; Hooper, E.B.; Lang, D.D.; Nevins, W.M.

    1987-01-01

    A new experimental facility is being assembled at the Lawrence Livermore National Laboratory (LLNL) for studying microwave propagation and absorption in high density plasmas. A unique feature of the facility is the free electron laser (FEL) used to generate high peak power microwaves at 250 GHz, at a repetition rate so as to produce up to 2 MW of average power for up to 30 s. Called the Microwave Tokamak Experiment (MTX), the facility will be used for studies of plasma heating, current drive, and confinement

  15. Cluster storage for COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír

    2012-01-01

    Roč. 87, č. 12 (2012), s. 2238-2241 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006

  16. Design constraints on magnet systems of future tokamaks based on experiences of present s.c. magnet development

    International Nuclear Information System (INIS)

    Heinz, W.; Jeske, U.; Komarek, P.; Krauth, H.

    1983-01-01

    In view of the urgent need for superconductivity in the next generation of big fusion devices and the identified gap between aimed data and the state of the art, impressive development programs are running world-wide, e.g. the IEA-Large Coil Task (LCT) and magnets for near term experiments (T15, Tore Supra). During the development work for all these magnet systems and simultaneously running design studies, especially the INTOR-study, some critical problem areas, e.g. concerning NbTi-conductor design and manufacturing and coil fabrication could be solved, others like the limitations by fatigue stresses for coil case and support structure turned out to be more stringent than anticipated. This paper tries to show which plasma physics parameters place especially severe constraints to magnet design, like PF-pulse number and amplitude at the TF-coils, so that they should be chosen with strongest care. It further points out which technologies under these circumstances are still missing or unproven with respect to the INTOR-like generation of fusion experiments. Further effort is mainly required for fatigue load behaviour of materials and components, high field windings and poloidal field coils. (author)

  17. Nuclear fusion research at Tokamak Energy Ltd

    International Nuclear Information System (INIS)

    Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

    2017-01-01

    Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

  18. Channels in tokamak reactor shields

    International Nuclear Information System (INIS)

    Shchipakin, O.L.

    1981-01-01

    The results of calculations of neutron transport through the channels in the tokamak reactor radiation shields, obtained by the Monte Carlo method and by the method of discrete ordinates, are considered. The given data show that the structural materials of the channel and that of the blanket and shields in the regions close to it are subjected to almost the same irradiation as the first wall and therefore they should satisfy the technical requirements. The radiation energy release in the injector channel wall, caused by neutron shooting, substantially depends on the channel dimensions. At the channel large diameter (0.7-10 m) this dependence noticeably decreases. The investigation of the effect of the injector channel cross section form on the neutron flux density through the channel, testifies to weak dependence of shooting radiation intensity on the form of the channel cross section. It is concluded that measures to decrease unfavourable effect of the channels on the safety of the power tokamak reactor operation and maintenance cause substantial changes in reactor design due to which the channel protection must be developed at first stages. The Monte Carlo method is recommended to be used for variant calculations and when calculating the neutron flux functionals in specific points of the system the discrete ordinate method is preferred [ru

  19. CAT-D-T tokamaks

    International Nuclear Information System (INIS)

    Greenspan, E.; Blue, T.; Miley, G.H.

    1981-01-01

    The domains of plasma fuel cycles bounded by the D-T and Cat-D, and by the D-T and SCD modes of operation are examined. These domains, referred to as, respectively, the Cat-D-T and SCD-T modes of operation, are characterized by the number (γ) of tritons per fusion neutron available from external (to the plasma) sources. Two external tritium sources are considered - the blankets of the Cat-D-T (SCD-T) reactors and fission reactors supported by the Cat-D-T (SCD-T) driven hybrid reactors. It is found that by using 6 Li for the active material of the control elements of the fission reactors, it is possible to achieve γ values close to unity. Cat-D-T tokamaks could be designed to have smaller size, higher power density, lower magnetic field and even lower plasma temperature than Cat-D tokamaks; the difference becomes significant for γ greater than or equal to .75. The SCD-T mode of operation appears to be even more attractive. Promising applications identified for these Cat-D-T and SCD-T modes of operation include hybrid reactors, fusion synfuel factories and fusion reactors which have difficulty in providing all their tritium needs

  20. Plasma position control in TCABR Tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

    1998-01-01

    The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

  1. Estimation of Zeff in Novillo Tokamak

    International Nuclear Information System (INIS)

    Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.

    1996-01-01

    We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)

  2. Fast IR diodes thermometer for tokamak

    International Nuclear Information System (INIS)

    Chen Xiangbo

    2001-01-01

    A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices

  3. ESEM modifications to LEO SUPRA 35 VP FESEM.

    Science.gov (United States)

    Dracopoulos, Vassileios; Danilatos, Gerasimos

    2013-01-01

    A LEO SUPRA 35 VP FESEM has been modified to significantly improve image quality in the low pressure mode of operation. It is further shown that the same conversion can allow the machine to operate also as a fully fledged environmental scanning electron microscope with specimen chamber pressures in excess of 2000 Pa. This is achieved by a diamagnetic insert with a thin pressure limiting aperture at the bottom of the pole-piece. The insert allows the use of various size apertures in addition to the existing 1 mm differential diaphragm. Copyright © 2012 Elsevier Ltd. All rights reserved.

  4. Regional Themes and Global Means in Supra-National Higher Education Policy

    Science.gov (United States)

    Watson, Pam

    2009-01-01

    The supra-national level has become increasingly important in educational policy formulation. This paper describes and compares two settings in which growth in these supra-national policies is evident--in Europe and in Africa. Key themes arising in policy documents in each context are examined. A distinction is drawn in analysis between themes…

  5. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  6. Research into controlled fusion in tokamaks

    International Nuclear Information System (INIS)

    Zacek, F.

    1992-01-01

    During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

  7. Theory of incremental turbulent transport in tokamaks

    International Nuclear Information System (INIS)

    Similon, P.L.

    1991-01-01

    The goal of this research is to understand how the various aspect of turbulent transport operate in tokamaks, in the presence of low frequency fluctuations such as drift waves or trapped electron modes

  8. Definition of total bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    Ross, D.W.

    1995-01-01

    Alternative definitions of the total bootstrap current are compared. An analogous comparison is given for the ohmic and auxiliary currents. It is argued that different definitions than those usually employed lead to simpler analyses of tokamak operating scenarios

  9. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    1981-01-01

    Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

  10. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    Caldas, I.L.; Vannucci, A.

    1985-01-01

    A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.) [pt

  11. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  12. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1986-01-01

    Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

  13. Power and particle exhaust in tokamaks

    International Nuclear Information System (INIS)

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER's nominal design positions; important directions for further research are identified

  14. Atypical initial presentation of Takayasu arteritis as isolated supra-valvular aortic stenosis.

    Science.gov (United States)

    Kim, Do Yeon; Kim, Hwan Wook

    2016-01-19

    Among the vascular involvements of Takayasu arteritis, a supra-valvular aortic stenosis has been reported very rarely. We report a case of surgically corrected, supra-valvular aortic stenosis caused by Takayasu arteritis. A 32-year-old female was diagnosed with supra-valvular aortic stenosis by transthoracic echocardiography for the evaluation of cardiac murmur with constitutional symptoms. Under the impression of non-familial sporadic type of supra-valvular aortic stenosis, surgical correction was performed. However, after 1 year from the operation, we could know the cause of her disease through the findings of computed tomographic aortography that Takayasu arteritis was suspected. Takayasu arteritis should be considered in adult female patients presenting supra-valvular aortic stenosis with constitutional symptoms, even if no typical features of vascular involvement.

  15. Duration Adaptation Occurs Across the Sub- and Supra-Second Systems.

    Science.gov (United States)

    Shima, Shuhei; Murai, Yuki; Hashimoto, Yuki; Yotsumoto, Yuko

    2016-01-01

    After repetitive exposure to a stimulus of relatively short duration, a subsequent stimulus of long duration is perceived as being even longer, and after repetitive exposure to a stimulus of relatively long duration, a subsequent stimulus of short duration is perceived as being even shorter. This phenomenon is called duration adaptation, and has been reported only for sub-second durations. We examined whether duration adaptation also occurs for supra-second durations (Experiment 1) and whether duration adaptation occurs across sub- and supra-second durations (Experiment 2). Duration adaptation occurred not only for sub-second durations, but also for supra-second durations and across sub- and supra-second durations. These results suggest that duration adaptation involves an interval-independent system or two functionally related systems that are associated with both the sub- and supra-second durations.

  16. Preliminary results of the TBR small tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Fagundes, A.N.; Da Silva, R.P.; Galvao, R.M.O.; Del Bosco, E.; Vuolo, J.H.; Sanada, E.K.; Dellaqua, R.

    1982-01-01

    The paper gives a short description of the TBR - small Brazilian tokamak and the first results obtained for plasma formation and equilibrium. Measured breakdown curves for hydrogen are shown to be confined within analytically calculated limits and to depend strongly on stray vertical magnetic fields. Time profiles of plasma current in equilibrium are shown and compared with the predictions of a simple analytical model for tokamak discharges. Reasonable agreement is obtained taking Zsub(eff) as a free parameter. (author)

  17. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br

    2002-07-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  18. Suprathermal electron studies in Tokamak plasmas by means of diagnostic measurements and modeling

    International Nuclear Information System (INIS)

    Kamleitner, J.

    2015-01-01

    To achieve reactor-relevant conditions in a tokamak plasma, auxiliary heating systems are required and can be realized by waves injected in the plasma that heat ions or electrons. Electron cyclotron resonant heating (ECRH) is a very flexible and robust technique featuring localized power deposition and current drive (CD) capabilities. Its fundamental principles are well understood and the application of ECRH is a proven and established tool; electron cyclotron current drive (ECCD) is regularly used to develop advanced scenarios and control magneto-hydrodynamics (MHD) instabilities in the plasma by tailoring the current profile. There remain important open questions, such as the phase space dynamics, the observed radial broadening of the supra-thermal electron distribution function and discrepancies in predicted and experimental CD efficiency. A main goal is to improve the understanding of wave-particle interaction in plasmas and current drive mechanisms. This was accomplished by combined experimental and numerical studies, strongly based on the conjunction of hard X-ray (HXR) Bremsstrahlung measurements and Fokker-Planck modelling, characterizing the supra-thermal electron population. The hard X-ray tomographic spectrometer (HXRS) diagnostic was developed to perform these studies by investigating spatial HXR emission asymmetries in the co- and counter-current directions and within the poloidal plane. The system uses cadmium-telluride detectors and digital acquisition to store the complete time history of incoming photon pulses. An extensive study of digital pulse processing algorithms was performed and its application allows the HXRS to handle high count rates in a noisy tokamak environment. Numerical tools were developed to improve the time resolution by conditional averaging and to obtain local information with the general tomographic inversion package. The interfaces of the LUKE code and the well-established CQL3D Fokker-Planck code to the Tokamak a

  19. Suprathermal electron studies in Tokamak plasmas by means of diagnostic measurements and modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kamleitner, J.

    2015-07-01

    To achieve reactor-relevant conditions in a tokamak plasma, auxiliary heating systems are required and can be realized by waves injected in the plasma that heat ions or electrons. Electron cyclotron resonant heating (ECRH) is a very flexible and robust technique featuring localized power deposition and current drive (CD) capabilities. Its fundamental principles are well understood and the application of ECRH is a proven and established tool; electron cyclotron current drive (ECCD) is regularly used to develop advanced scenarios and control magneto-hydrodynamics (MHD) instabilities in the plasma by tailoring the current profile. There remain important open questions, such as the phase space dynamics, the observed radial broadening of the supra-thermal electron distribution function and discrepancies in predicted and experimental CD efficiency. A main goal is to improve the understanding of wave-particle interaction in plasmas and current drive mechanisms. This was accomplished by combined experimental and numerical studies, strongly based on the conjunction of hard X-ray (HXR) Bremsstrahlung measurements and Fokker-Planck modelling, characterizing the supra-thermal electron population. The hard X-ray tomographic spectrometer (HXRS) diagnostic was developed to perform these studies by investigating spatial HXR emission asymmetries in the co- and counter-current directions and within the poloidal plane. The system uses cadmium-telluride detectors and digital acquisition to store the complete time history of incoming photon pulses. An extensive study of digital pulse processing algorithms was performed and its application allows the HXRS to handle high count rates in a noisy tokamak environment. Numerical tools were developed to improve the time resolution by conditional averaging and to obtain local information with the general tomographic inversion package. The interfaces of the LUKE code and the well-established CQL3D Fokker-Planck code to the Tokamak a

  20. Tore Linné Eriksen og de store utviklingsspørsmålene

    Directory of Open Access Journals (Sweden)

    Kristen Nordhaug

    2015-10-01

    Full Text Available This article reviews some of Tore Linné Eriksen’s works within development studies/development research. In a recent introduction to development studies from 2013, he presented development research as a cross-disciplinary social science approach that addresses the grand problems of mankind. Eriksen’s own research into these grand problems has concentrated on the causes of national and international inequality and poverty. In 1974 he supported the view of the “underdevelopment school”: “Underdevelopment” in Africa and Latin America was the outcome of the inclusion of those continents in a capitalist world economy dominated by Europe. Recent works by Eriksen on the origins of the “great divergence” between Western Europe and economically advanced non-European countries (2010 and on inequality and poverty in the current world (2012 are far more complex and empirically nuanced. Still, in a recent discussion of globalization and global capitalism (2013 he reverts to some of the earlier ”underdevelopment arguments” from 1974. The article concludes that there is a tension within Eriksen’s works on the role of capitalism in development and underdevelopment. In his programmatic writings, global capitalism is seen as the main cause of inequality and poverty. In his more empirically grounded works, global capitalism is viewed rather as an important part of the bigger picture of inequality and poverty.