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Using a recently installed impurity powder dropper (IPD), boron powder (μm) was injected into lower single null (LSN) L-mode discharges in WEST. IPDs possibly enable real-time wall conditioning of the plasma-facing components and may help to facilitate H-mode access in the full-tungsten environment of WEST. The discharges in this experiment featured I p = 0.5 MA, B T = 3.7 T, q 95 = 4.3, t pulse = 12–30 s, n e,0 ∼ 4 × 1019 m−2, and P LHCD ∼ 4.5 MW. Estimates of the deuterium and impurity particle fluxes, derived from a combination of visible spectroscopy measurements and their corresponding S/XB coefficients, showed decreases of ∼50% in O+, N+, and C+ populations during powder injection and a moderate reduction of these low-Z impurities (∼50%) and W (∼10%) in the discharges that followed powder injection. Along with the improved wall conditions, WEST discharges with B powder injection observed improved confinement, as the stored energy W MHD, neutron rate, and electron temperature T e increased significantly (10%–25% for W MHD and 60%–200% for the neutron rate) at constant input power. These increases in confinement scale up with the powder drop rate and are likely due to the suppression of ion temperature gradient (ITG) turbulence from changes in Z eff and/or modifications to the electron density profile.

High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in a harsh environment with considerable neutron/nuclear radiation and high thermal load. Essential components of the HF sensor system, including inductive coil, electron cyclotron heating (ECH) shield, electrical cabling and termination load, have been designed to meet ITER measurement requirements. System performance (e.g. frequency response, thermal conduction) has been assessed. A prototyping campaign was initiated to demonstrate the manufacturability of the designed components. Prototypes have been produced according to the specifications. A series of lab tests have been performed to examine assembly issues and validate electrical and thermo-mechanical aspects of the design. In-situ microwave radiation test has been conducted in the MISTRAL test facility at IPP-Greifswald to experimentally examine the microwave shielding efficiency and structural integrity of the ECH shield. Low-power microwave attenuation measurement and scanning electron microscopic inspection were conducted to probe and examine the quality of the metal coating on the ECH shield.

Disruptions in large size tokamak like ITER must be mitigated to reduce detrimental effects on the device. Massive impurity injection prior to disruption is a promising mitigation technique. Many injector designs have already been tested on nowadays tokamaks. A novel concept of injector has been designed and tested on Tore Supra. It is based on the use of high pressure cartridges sealed with a bursting disc. For firing, an electric arc is generated inside the cartridge close to the rupture disc. This initiates a shock wave which breaks the disc. Tests show that the opening is achieved in less than 400 μs. A detailed modeling of the fast injector operation is presented. One result is that the gas is expelled in about 2 ms and an outflow rate as large as 4.2 × 10 25 atoms/s can be achieved at the nozzle exit (1 cm diameter). Massive neon gas injections have be performed on stable plasma to validate the concept. The gas penetration into the plasma prior to disruption is followed by a fast camera. Comparison with a ferromagnetic-valve based injector was also carried out. It is found that the velocity of the cold front penetration into the hot plasma is larger using this new injector.

The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R&D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R&D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed. (C) 2013 Elsevier B.V. All rights reserved.

Fiber optic current sensors have been recently considered for current measurements in ITER. These sensors exploit Ampere's law by measuring the change of light polarization along an optical fiber subjected to a magnetic field aligned along the fiber axis. Through collaboration between CEA and SCK·CEN, such fiber optic current sensors have been installed on Tore Supra tokamak as a first demonstration for possible future applications in ITER. This paper aims at reporting the results obtained on Tore Supra. It discusses the technique used to separate the effects of the intrinsic linear birefringence and the magneto-induced circular birefringence, the latter enabling the determination of the current flowing through the area defined by the optic fiber loop. Three optical fibers have been installed on Tore Supra for assessment of their linearity response on a wide range of plasma current from 0.3 to 1.4 MA as well as their long term stability. The extrapolation of the results and the issues related to a future application on ITER are finally discussed.

This paper presents the multiple set of requirements for the ITER magnetic diagnostic systems and the current status of the various R&D activities performed by the EU partners.

The Tore Supra tokamak is devoted to long-duration, high-performance plasma research. The real-time measurements and control (RTMC) system has been developed to address the basic tokamak controls, such as plasma equilibrium and density control. Over the years, more and more sophisticated and demanding controls have been implemented, allowing the improvement of plasma performance and machine protection. This includes current profile control to enhance the performance and to avoid magnetohydrodynamic instabilities, infrared monitoring of plasma facing components to prevent overheating, and disruption detection and mitigation techniques. Most of these improvements are relevant to the plasma operation in a full steady-state regime. This paper describes the present status of the Tore Supra RTMC system, detailing recent progress and highlighting the advantages of the various control schemes implemented so far.

This work deals with the implementation and integration of advanced feedback control schemes on the Tore Supra experiment, reports the technical details and shows the first positive results that have been achieved. Encouraging outcomes have been obtained in the field of current and pressure profiles control. A new feedback control algorithm, whose purpose is to maximize the plasma performance, has been effectively tested too. These and other promising results, whose detailed description will be given in the article, have been obtained thanks to real-time availability of several diagnostic systems. A key issue is the possibility to integrate these controls in such a way as to cope with different requests at the same time. For example, the simultaneous control of the total 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 profile's width through the parallel refractive index has been successfully accomplished. The Tore Supra real-time feedback control system showed robust, flexible and effective.

Using the Tore Supra infrared thermography diagnostics, a new real time feedback control has been successfully implemented to maximize additional RF power while preventing plasma facing components (PFCs) from overheating and damage. As a first step, a thermography feedback control has been used to detect and extinguish electric arcs on lower hybrid current drive (LHCD) launchers. Secondly, heating sources on PFCs have been identified highlighting the role of the power from each ion cyclotron resonance heating (ICRH) antenna and LHCD launcher and the interactions between them. A new feedback control algorithm was developed to control the additional power. The real time feedback control of PFC temperatures which makes part of an integrated feedback controller, is a reliable tool routinely used as a basic protection system. Furthermore, it has proven its capability to operate in parallel with other control schemes such as the current profile control.

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.

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° 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.

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.

The main objective of the WEST project is to test the actively cooled tungsten plasma facing components under high flux conditions and to provide relevant information to ITER [1]. In this framework, two new power supplies will be installed in Tore Supra in order to feed the new upper and lower divertor coils. These new coils are installed inside the vacuum vessel to modify the magnetic configuration (X point) and to allow long pulse operation in H-mode.

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.

The Tore Supra gigajoule (GJ) scenario was developed during the 2002 experimental campaign, with the goal of realizing long pulses in a stationary state, at a loop voltage therefore close to zero. This challenge has led to very successful results, with a record injected energy of 0.75 GJ and perfect control of the plasma density over 4 min 25 s. The performance is however limited by the lower hybrid system reliability and MHD limits. This paper focuses on the three kinds of MHD activity that were observed in the development of this scenario. The first one, probably a manifestation of the feedback control on the plasma position, is benign and, once triggered, stays at a low level. The second one is the triggering of a fast growing instability, which leads to a strong degradation of the confinement in the core region and to a redistribution of the plasma current. After such events, a regime with a persistent MHD activity sets in and represents the third kind of MHD activity in the GJ experiments. This paper reports on the analysis of this MHD activity. Perspectives for the future development of the GJ scenario in Tore Supra are drawn.

A magnetic diagnostic has been installed in Tore Supra during the year 2000. It has been designed to be compatible with the configuration Composants Internes Et Limiteur (CIEL) requirements. It consists of a set of 424 dedicated magnetic sensors, pickup coils, and flux loops. The entire system is placed in the tokamak vacuum vessel at low pressure, high temperature, and is exposed to plasma radiation. The magnetic field is determined by attaching an analog integrator to each sensor. These electronic cards were developed for long pulse operation up to 1000 s. Finally the data are acquired by a system using the VME standard. The diagnostic has been successfully operated since the restart of Tore Supra with the CIEL configuration in September 2001. It has demonstrated the capability to control the plasma during discharges of several hundreds seconds and it participates to the physics program of the CIEL Project. The technology employed for the diagnostic achievement is robust, safe, and it can be extrapolated to other tokamaks.

In the year 2000, a new version of the analog integrator used in Tore Supra for magnetic measurements has been designed. Although based on the same principles, this new integrator was improved in order to satisfy the new long duration of pulses, up to 1000 s, expected in near future on Tore Supra. After recalling the basic principles of the analog integrators and recent improvements, this article explains the main characteristics of the new integrators, tests and obtained results. The conclusion underlines the most important results and, in particular, the capability of the new integrator to answer to the International thermonuclear experimental reactor (ITER) requirements.

International audience; The nuclear fusion research goal is to demonstrate the feasibility of fusion power for peaceful purposes. In order to achieve the conditions similar to those expected in an electricity-generating fusion power plant, plasmas with a temperature of several hundreds of millions of degrees must be generated and sustained for long periods. For this purpose, RF antennas delivering multi-megawatts of power to magnetized confined plasma are commonly used in experimental tokamaks. In the gigahertz range of frequencies, high power phased arrays known as " Lower Hybrid " (LH) antennas are used to extend the plasma duration. This paper reviews some of the technological aspects of the LH antennas used in the Tore Supra tokamak and presents the current design of a proposed 20 MW LH system for the international experiment ITER.

Third harmonic generation (THG) efficiency is shown to be a greatly enhanced at the onset of inelastic scattering of electrons on optic phonons. Scaling experiments are performed on n-type InP at the pump wave frequency of 9.43 GHz at 80 K. Monte Carlo modeling is employed for scaling the effect to the 3(rd) harmonic frequency of 1 THz. The THG efficiency in n-type GaAs and InP as well as in the wurtzite phase of n-type InN and GaN compound crystals is compared to that in n-type Si. The efficiency maximum is found to weaken due to the quasi-elastic scattering on acoustic phonons and elastic scattering on ionized impurities. Nevertheless, the THG efficiency at 1 THz in InP crystals cooled down to liquid nitrogen temperatures is predicted to be 2 orders of magnitude higher than the reference value of 0.1% experimentally recorded up to now in n-type Si.

A parametric experimental study of fluctuations has been performed in L-mode plasmas during steady-state phases in Tore Supra. Simultaneous measurements of density and magnetic fluctuations in the gradient region of the discharge show that both components are strong functions of the local electron temperature decreasing lengths. The wavenumber spectrum of density fluctuations is obtained with the scattering diagnostic both in the bulk and at the edge of the discharge. In the bulk, in the L-mode compared to ohmic plasmas, the k spectrum increases below and is globally shifted towards low k values. At the edge (r/a > 0.9) an increase on all k values is observed. At r/a = 0.5, the level of both types of fluctuations is found to be linearly correlated with the local heat diffusivity . The heat flux ratio advected by electrostatic eddies in the bulk of the discharge is estimated by using fluctuation measurements and generic predictions for the parametric dependency of the thermal flux. The calculated level of density fluctuation necessary to account for the heat transport at midradius is found to be compatible with values measured in other tokamaks. It is also found that the heat flux ratio behaviour conducted by electrostatic fluctuations is compatible with most of the experimental heat flux ratio behaviour in the plasma core (0.4 < r/a < 0.6). This property does not hold at the edge of the plasma (r/a = 0.7). The similar parametric dependency of magnetic and density fluctuation level does not allow one to make a clear statement about their relative importance in the transport process and suggests that they could be coupled.

The WEST project (W - for tungsten - Environment in Steady-state Tokamak) consists in transforming the French tokamak Tore Supra into an X-point divertor one and aims at testing the ITER technology of actively cooled high heat flux tungsten components during ITER-relevant long pulses. In addition to the integration in the vacuum vessel of a lower divertor and an upper one, this metamorphosis has a strong impact on diagnostics initially present in Tore Supra. Indeed the introduction of a divertor in the upper part of the machine leads to block the lines of sight of diagnostics previously installed in the upper ports. Therefore new diagnostics playing a paramount role for the safe tokamak operation need to be implemented, such as an infrared thermography diagnostic for the protection of plasma-facing components (PFCs) and heating antennas, and a visible spectroscopy system for the protection of tungsten PFCs. Besides several existing diagnostics need to be modified to adapt their lines of sight to the new divertor targets and plasma shape. This paper presents the main issues linked to the integration of diagnostics deriving from the WEST upgrade project. First the changes involved by the upgrade on the set of diagnostics currently implemented in Tore Supra are described. Then a particular focus is put on the integration of the two new optical systems, namely the infrared system and the visible spectroscopy one. These two systems being located within the vacuum vessel, a special care must be paid to their design, to allow them to withstand their harsh environmental conditions. Other new or updated fundamental diagnostics are also briefly presented. Finally the resulting new implementation of diagnostics envisaged to reach the scientific and technological goals of the WEST project is described.

The feedback control of fusion experiments in tokamak devices is entering a new area driven by the increase of control requirements for obtaining burning plasmas under safe operation conditions. A project aiming at setting up a flight simulator for the development of advanced controllers has recently started at CEA. This simulator reuses most of the components of the European integrated tokamak modelling (ITM) simulation platform. Thus, it benefits from the development made by the task force and it will be able to offer a development platform for the new controllers of present day European tokamaks and future machines. The paper will report on the different software developments performed in order to set up a flexible and user friendly design and test of feedback control loops.

International audience

Present work discusses the issues of creating the instrumentation for testing the semiconductor magnetic field sensors during their irradiation with neutrons in nuclear reactors up to fluences similar to neutron fluences in steady-state sensor locations in ITER. The novelty of the work consists in Hall sensor parameters being investigated: first, directly during the irradiation (in real time), and, second, at high irradiation levels (fast neutron fluence > 1018n/cm2). Developed instrumentation has been successfully tested and applied in the research experiments on radiation stability of magnetic sensors in IBR-2 (JINR, Dubna) and VVR-M (PNPI, Saint-Petersburg) reactors.

Magnetic diagnostics of the ITER tokamak must fulfill demanding specifications, because their accuracy and reliability affects margins to the machine engineering limits and there-fore operational flexibility. This paper describes the challenging issues related to the implementation of the magnetic diagnostics in a tokamak environment. We focus on nuclear radiations as they can significantly affect the measurement through Radiation Induced Electromotive Force (RIEMF) or Thermally Induced Electromotive Force (TIEMF). Thermal modeling of magnetic sensors and associated design studies are also reported as the thermal gradient in the sensors must be reduced to avoid TIEMF. Alter-native magnetic sensors such as fiber optic current sensors (FOCS) or steady state magnetic field sensors are also discussed because they serve as a backup to the usual inductive magnetic measurements. We conclude by a brief review of the development needs for magnetic diagnostics‥

Attempts to run non-inductive plasma discharges on Tore Supra sometimes fail due to the triggering of magneto-hydro-dynamic (MHD) instabilities that saturate at a large amplitude, producing degraded confinement and loss of wave driven fast electrons (the so-called MHD regime (Maget et al 2005 Nucl. Fusion 45 69–80)). In this paper we investigate the transition to this soft (in the sense of non-disruptive) MHD limit from experimental observations, and compare it with non-linear code predictions. Such a comparison suggests that different non-linear regimes, with periodic relaxations or saturation, are correctly understood. However, successful non-inductive discharges without detectable magnetic island at q = 2 cannot be reproduced if realistic transport coefficients are used in the computation. Additional physics seems mandatory for explaining these discharges, such as diamagnetic effects, that could also justify cases of abrupt transition to the MHD regime.

The current status of the ITER magnetics diagnostic is summarised. The various risks are discussed on the basis of when they could be present during ITER operation and appropriate mitigation is reviewed. An implementation risk or common mode operation failure risk to the magnetic diagnostic can only be mitigated by diversity. All identified risks are being handled with ongoing research and development to minimise these risks. A reasonable level of design and manufacturing diversity has been therefore introduced in the procurement and installation plans, in order to optimise overall reliability avoiding dispersion of resources. (C) 2009 Elsevier B.V. All rights reserved.

The Integrated tokamak modelling taskforce was set up to provide the European scientific community with simulation tools for preparing and analysing discharges of fusion experiments. We will report on recent progress made on the taskforce project on equilibrium and linear stability. A generic data structure has been devised to describe the geometry of a machine and physical processes in the discharge. This data structure is used to interface all individual analysis program within the taskforce. One of the analysis tools, the equilibrium code EFIT__ITM, based on the EFIT code written by L. L. Lao, has been completely rewritten in order to make it suitable for the ITM. It has algorithm enhancements to increase execution speed, and the ability to treat anisotropic pressure and deviation from axisymmetry. The reconstruction code is now completely independent of the machine description. First results on veriflcation and validation of the new tool are presented.

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 [1]. 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 [2], a sawtooth‐like behaviour of the (2,1) Double‐Tearing Mode as first seen in TFTR [3], or a large amplitude (2,1) tearing mode that severely degrades the energy confinement, as reported in Tore Supra [4], JET [5] or DIII‐D [6]. Situations where qmin≈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 derived in [7] could explain linear stability, but such effect vanishes in presence of heat transport [8]. Electron diamagnetic rotation effect is investigated as a possible mechanism for n = 1 mode stabilization.

We evaluate the possibility to use a low-birefringence fibre for current sensing application in harsh environmental conditions as expected in quasi steady plasma burning reactor. Experimental results show that the Faraday rotation of the low-birefringence fibre is not significantly degraded by a gamma irradiation upto 5.4 MGy and when the fibre is kept at ambient temperature.

The recent experiments in Tore Supra has focused on the optimisation of the RF heating systems (ICRH and LHCD) in order to achieve high power levels simultaneously over long durations. This leads to njected power levels n the range of 10 MW, which is representative of ITER operation in terms of average power density and heat exhaust capability. Successful operation has been achieved up to 9–10 MW for 20–30 s, using real‐time control of the antenna front face temperatures from nfrared measurements, and security on Cu and Fe brightness. The wave coupling was optimised for the two RF systems simultaneously, while at the same time minimising the hot spot formation due to RF sheath effects and interaction by fast particles (fast electrons accelerated n the near field of the LH wave and fast ions with large orbits created by ICRH). Local modification of the scrape‐off layer density around an ICRH antenna impacts on magnetcally connected LHCD antennas. The LH coupling either improves or deteriorates locally, depending on the relative radial and poloidal positon of the LHCD waveguides and the ICRH antenna. The surface temperature related to the acceleration of electrons n the near‐field of the LH wave evolves accordingly, i.e. it either increases or decreases as a response to the increase or decrease in scrape‐off layer density.

During operation of present fusion devices, the plasma facing components (PFCs) are submitted to large heat fluxes within a range of 10-20 MW/m2. Understanding and preventing overheating of these components during long pulse discharges is a crucial issue. The surface temperature of the PFCs in the inner chamber of a tokamak is measured by infrared (IR) cameras interfaced with complex optical systems. Due to the complexity of the observed IR scenes and the large amount of data produced, a computer-aided analysis of infrared images will be a key point for ITER, the next fusion device for both real-time applications and deferred-time analysis. This paper describes the current state of IR image analysis based on region of interest (ROI) approach at Tore Supra with an extrapolation to the future ITER tokamak.

The paper deals with radiation resistant sensors and their associated measuring instrumentation developed in the course of R and D activities carried out in the framework of an international collaboration. The first trial tests of three-dimensional (3D) probes with Hall sensors have been performed in European tokamaks TORE SUPRA (2004) and JET (2005). Later in 2009 six sets of 3D probes were installed in JET and now continue to operate. The statistical analysis performed in 2014 on the basis of the JET database have demonstrated stable long term operation of all 18 sensors of 3D probes. The results of measurements conducted at the neutron fluxes of nuclear reactors have demonstrated the operability of the sensors up to high neutron fluences of F > 1018n cm-2 that exceeds the maximum one for the locations of steady state sensors in ITER over its total lifetime.

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.

During the 2003 experimental campaign, the aim of realizing reliable fully non inductive discharges has been successfully completed in Tore Supra. One the major difficulties in optimizing this scenario comes from MHD stability issues. Indeed, the current profile obtained with LH antennas in these experiments is prone to the triggering of single or double tearing modes. Under certain conditions, the discharges enters in a regime of permanent MHD activity. The present work investigates the MHD properties of these fully non-inductive discharges in the operational space defined by the toroidal magnetic field B, the total plasma current Ip, and the mean parallel index of LH waves . Linear MHD stability properties divide the operational space in regions separated by rational values of the minimum of the safety factor. We discuss in our work the localisation of the transition between benign MHD activity and the MHD regime., Comment: 12th International Congress on Plasma Physics, 25-29 October 2004, Nice (France)

Auto antibodies against factor VIII procoagulant activity are uncommon. Three cases are reported in three different situations: peri-partum, penicillin and idiopatic inhibitor.

A high power radiation source in the THz range with long pulse and narrow line width is required for diagnosing fusion type plasmas by collective Thomson scattering. Gyrotrons currently meet the requirements concerning power, pulse length and line width when operating in the 50-150 GHz range. They may have the potential also to do this in the THz range. In this paper two methods based on harmonic operation are investigated. Both look promising, if high conversion efficiency can be attained. This, however, still has to be demonstrated. (C) 1999 Published by Elsevier Science B.V. All rights reserved.

International audience

This paper describes what we can learn on the regimes of spontaneous electron temperature oscillations discovered in Tore Supra from the analysis of MHD activity. Since the first observations of this oscillating behaviour of plasma equilibrium, and its interpretation as a predator-prey system involving lower hybrid waves power deposition and electron confinement, analysis of MHD modes has confirmed the reality of safety factor profile oscillations. This points towards the importance of rational values of the safety factor in the transition to transport barriers in reversed magnetic shear plasmas.

International audience

Inductive magnetic sensors are widely used for plasma equilibrium reconstruction and control. However, their measurements involve electronic integration and can therefore experience drift leading to inaccurate plasma positioning. For this reason, we have studied an original drift-free approach to estimate the plasma equilibrium. The principle of this correction is based on modulation of the plasma position and current at three independent frequencies and analysing the modulated magnetic signals to provide additional estimation on the equilibrium. Using a plasma model based on current wires, the accuracy of such a method is assessed for Tore Supra in terms of the signal to noise ratio. The plasma position is recovered within a precision of 5 mm fora signal to noise ratio better than 80 dB. Applying our approach to dedicated experiments performed on the Tore Supra tokamak, we confirm the quality of the result and find that we can estimate the radial and vertical positions of the plasma to 1 mm with a one standard deviation confidence interval. (C) 2009 Elsevier B.V. All rights reserved.

We investigate experimental dependence of the third harmonic generation efficiency in the n-type Si crystals on the geometrical dimensions of the sample, polarization and power of the fundamental wave. The efficiency increases monotonically with the rise of the sample thickness up to a threshold value, and decreases dramatically above the threshold. At shorter propagation distances the generation efficiency could be correctly simulated using the layered medium approximation and the numerically calculated electron drift velocity response to the pumping wave electric field to describe the change of the semiconductor properties under high-power microwave irradiation.