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Tore Supra is dedicated to long pulse operation at high power, with a record in injected energy of 1 GJ (2.8 MW × 380 s) and an achieved capability of 12 MW injected power delivered by 3 RF systems: Lower Hybrid Current Drive (LHCD), Ion Cyclotron Resonance Heating (ICRH) and Electron Cyclotron Resonance Heating (ECRH). The new WEST project (W [tungsten] Environment in Steady-state Tokamak) aims at fitting Tore Supra with an actively cooled tungsten coated wall and a bulk tungsten divertor. This new device will offer to ITER a test bed for validating the relevant technologies for actively cooled metallic components, with D-shaped H-mode plasmas. For WEST operation, different scenarii able to reproduce ITER relevant conditions in terms of steady state heat loads have been identified, ranging from a high RF power scenario (15 MW, 30 s) to a high fluence scenario (10 MW, 1000 s). This paper will focus on the evolution of the RF systems required for WEST. For the ICRH system, the main issues are its ELM resilience and its CW compatibility, three new actively cooled antennas are being designed, with the aim of reducing their sensitivity to the load variations induced by ELMs. The LH system has been recently upgraded with new klystrons and the PAM antenna, the possible reshaping of the antenna mouths is presently studied for matching with the magnetic field line in the WEST configuration. For the ECRH system, the device for the poloidal movement of the mirrors of the antenna is being changed for higher accuracy and speed.Tore Supra is dedicated to long pulse operation at high power, with a record in injected energy of 1 GJ (2.8 MW × 380 s) and an achieved capability of 12 MW injected power delivered by 3 RF systems: Lower Hybrid Current Drive (LHCD), Ion Cyclotron Resonance Heating (ICRH) and Electron Cyclotron Resonance Heating (ECRH). The new WEST project (W [tungsten] Environment in Steady-state Tokamak) aims at fitting Tore Supra with an actively cooled tungsten coated wall and a bulk tungsten divertor. This new device will offer to ITER a test bed for validating the relevant technologies for actively cooled metallic components, with D-shaped H-mode plasmas. For WEST operation, different scenarii able to reproduce ITER relevant conditions in terms of steady state heat loads have been identified, ranging from a high RF power scenario (15 MW, 30 s) to a high fluence scenario (10 MW, 1000 s). This paper will focus on the evolution of the RF systems required for WEST. For the ICRH system, the main issues are its ELM resil...

During the initial operation of the JET ITER-like wall, particular attention was given to the characterization of the Ion Cyclotron Resonance Frequency (ICRF) heating in this new metallic environment. In this contribution we compare L-modes plasmas heated by ICRF or by Neutral Beam Injection (NBI). ICRF heating as expected led to a much higher centrally peaked power deposition on the electrons and due to the central fast ion population to stronger sawtooth activity. Surprisingly, although a higher bulk radiation was observed during the ICRF phase, the thermal plasma energy was found similar for both cases, showing that a higher radiation inside the separatrix was not incompatible with an efficient central heating scheme. The higher radiation was attributed to the presence Tungsten (W). Tomographic inversion of SXR emissions allowed a precise observation of the sawtooth effect on the radiation pattern. W concentration profiles deconvolved from SXR emission showed the flattening of the profiles due to sawtooth for both heating and the peaking of the profiles in the NBI case only hinting for extra transport effect in the ICRF case.

In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall material on the JET Ion Cyclotron Resonance Frequency (ICRF) operation was assessed and also the properties of JET plasmas heated with ICRF were studied. No substantial change of the antenna coupling resistance was observed with the ILW as compared with the carbon wall. Heat-fluxes on the protecting limiters close the antennas quantified using Infra-Red (IR) thermography (maximum 4.5 MW/m2 in current drive phasing) are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can well reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. Some experimental facts indicate that main-chamber W components could be an important impurity source: the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions; the W content is also increased in ICRF-heated limiter plasmas; and Be evaporation in the main chamber results in a strong and long lasting reduction of the impurity level. The ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 20%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating efficiency; The ICRF power can be deposited at plasma centre and the radiation is mainly from the outer part of the plasma. Application of ICRF heating in H-mode plasmas started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core could be observed.

After the change over from the C-wall to the ITER-like Be/W wall (ILW) in JET, the radiation losses during ICRF heating have increased and are now substantially larger than those observed with NBI at the same power levels, in spite of the similar global plasma energies reached with the two heating systems. A comparison of the NBI and ICRF performances in the JET-ILW experiments, based on a statistical analysis of ∼3000 L-mode discharges, will be presented.

ITER's Ion Cyclotron Range of Frequencies (ICRF) system [1] comprises two antenna launchers designed by CYCLE (a consortium of European associations listed in the author affiliations above) on behalf F4E for the ITER Organisation (IO), each inserted as a Port Plug (PP) into one of ITER's Vacuum Vessel (VV) ports. Each launcher is an array of 4 toroidal by 6 poloidal RF current straps specified to couple up to 20 MW in total to the plasma in the frequency range of 40 to 55 MHz but limited to a maximum system voltage of 45 kV and limits on RF electric fields depending on their location and direction with respect to respectively the torus vacuum and the toroidal magnetic field. A crucial aspect of coupling ICRF power to plasmas is the knowledge of the plasma density profiles in the Scrape-Off Layer (SOL) and the location of the RF current straps with respect to the SOL. The launcher layout and details were optimized and its performance estimated for a worst case SOL provided by the IO. The paper summarizes the estimated performance obtained within the operational parameter space specified by IO. Aspects of the RF grounding of the whole antenna PP to the VV port and the effect of the voids between the PP and the Blanket Shielding Modules (BSM) surrounding the antenna front are discussed.

The major aspects linked to the use of the JET auxiliary heating systems: NBI, ICRF and LHCD, in the new JET ITER-like wall (JET-ILW) are presented. We show that although there were issues related to the operation of each system, efficient and safe plasma heating was obtained with room for higher power. For the NBI up to 25.7MW was safely injected; issues that had to be tackled were mainly the beam shine-through and beam re-ionisation before its entrance into the plasma. For the ICRF system, 5MW were coupled in L-mode and 4MW in H-mode; the main areas of concern were RF-sheaths related heat loads and impurities production. For the LH, 2.5 MW were delivered without problems; arcing and generation of fast electron beams in front of the launcher that can lead to high heat loads were the keys issues. For each system, an overview will be given of: the main modifications implemented for safe use, their compatibility with the new metallic wall, the differences in behavior compared with the previous carbon wall, with emphasis on heat loads and impurity content in the plasma., 21 pages, 17 figures

International audience; The upgrade of its ion cyclotron resonance (ICRH) and lower hybrid current drive (LHCD) heating systems makes the Tore Supra (TS) tokamak particularly well suited to address the physics and technology of high-power and steady-state plasma-surface interactions. High radio frequency (RF) heating powers have been successfully applied up to 12.2 MW coupled to the plasma, in which about 7.85 MW flows through the scrape-off layer. Thermal calculation based on thermography measurements gives the heat flux density distribution on the TS toroidal limiter located at the bottom of the machine. The target heat flux densities are divided by the incidence angle of the field lines with the surface and mapped to the magnetic flux surface to evaluate the power flowing in the scrape-off layer (SOL). The power profile shows a narrow component near the last closed flux surface and a wide component in the rest of the SOL. The narrow component is attributed to significant cross-field heat flux density around the plasma contact point, about 0.8% of the parallel heat flux density in the SOL, when incident angles are nearly tangential to the surface. The wide component is used to derive the experimental heat flux decay length (lambda(q)) and parallel heat flux in the SOL. The power widths are measured for a series of 1 MA/3.8 T discharges involving a scan of RF injected power 3.5

In ASDEX Upgrade H-modes with H98≈0.95, similar effect of the ICRF antenna loading improvement by local gas injection was observed as previously in L-modes. The antenna loading resistance Ra between and during ELMs can increase by more than 25% after a switch-over from a deuterium rate of 7.5⋅1021 D/s injected from a toroidally remote location to the same amount of deuterium injected close to an antenna. However, in contrast to L-mode, this effect is small in H-mode when the valve downstream w.r.t. parallel plasma flows is used. In L-mode, a non-linearity of Ra at PICRP 30% with no effect of spectrum and phase of MPs on Ra found so far. In the case ELMs are fully mitigated, the antenna loading is higher and steadier. In the case ELMs are not fully mitigated, the value of Ra between ELMs is increased. Looking at the W source modification for the impro...

Équipe 107 : Physique des plasmas chauds; International audience; In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall materials on the JET Ion Cyclotron Resonance Frequency (ICRF) operation is assessed and some important properties of JET plasmas heated with ICRF are highlighted. A similar to 20% reduction of the antenna coupling resistance is observed with the ILW as compared with the JET carbon (JET-C) wall. Heat-fluxes on the protecting limiters close the antennas, quantified using Infra-Red thermography (maximum 4.5MW/m(2) in current drive phasing), are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. The location of the tungsten ICRF specific source could not be identified but some experimental observations indicate that main-chamber W components could be an important impurity source: for example, the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions, and Be evaporation in the main chamber results in a strong reduction of the impurity level. In L-mode plasmas, the ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 15%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating performance; the power is typically deposited at the plasma centre while the radiation is mainly from the outer part of the plasma bulk. Application of ICRF heating in H-mode plasmas has started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core has been observed.

Using a reciprocating Retarding Field Analyser (RFA), Scrape-Off Layer (SOL) modifications were investigated on ASDEX-Upgrade during heating with waves in the Ion Cyclotron Range of Frequencies (ICRF), suspected for enhanced impurity production in this all-metal machine. Two quantities involved in the sputtering were measured: the current Islit on a saturated slit plate, proportional to the parallel ion flux and the mean parallel energy t of collected ions, averaged over many RF cycles. Combining multiple RFA reciprocations over a scan of q95 provided 2D poloidal/radial resolution. In the outer SOL a localized RF-perturbed zone was evidenced on the RFA side magnetically connected to an active ICRF antenna. A flat 2D Islit pattern surrounded by steep gradients was observed, correlatively with t exceeding 150eV. The centre of the zone is connected radially slightly behind the leading edge of antenna side limiters, with a radial extension up to ±2cm. The zone is broadest and t is largest near the...

The magnetic deflection of field lines to dedicated wall components in the ergodic divertor of Tore Supra generates complex patterns of power deposition. In this paper, we analyze the energy flux deposition on neutralizer plates as measured by infrared cameras and Langmuir probes. Three important features will be discussed: (1) The energy deposition during helium shots is as much as twice that for deuterium shots, for a given input power level. (2) The sheath heat transmission factor, deduced experimentally by comparison between probes and infrared measurements, increases with input power independently of the working gas from ∼7.5 (P TOT = 1 MW) to ∼10-11 for P TOT = 5 MW). In ohmic discharges, the standard value of 7 is recovered except specific cases in helium where γ can decrease to 2 or 3. (3) These anomalous values put in doubt the validity of edge temperature measurements by Langmuir probes in detached plasmas and have led to the development of a promising 'infrared' degree of detachment (Dod).

Introduction L’isotretinoine est largement utilisee comme traitement des acnes severes. Une allergie a l’arachide ou au soja constitue une contre-indication au traitement (Vidal®). Cette contre-indication vient du fait qu’il existe de l’huile raffinee de soja a hauteur de plus de 50 % dans la composition du produit fini. Nous presentons le cas d’un patient allergique vrai a l’arachide et au soja pour qui un traitement par Curacne® a ete envisage et introduit. Observations Un adolescent de 15 ans etait adresse a la consultation en mai 2015 pour une acne severe. Il presentait comme principaux antecedents un asthme, une allergie au soja, a l’arachide, aux fruits a coque et aux acariens. Il avait par ailleurs presente un choc anaphylactique en 2013 lors d’une reintroduction hospitaliere d’arachide. Il existait, pour son acne, une indication theorique a un traitement par isotretinoine. Compte tenu de ses antecedents, le laboratoire commercialisant Curacne® avait ete contacte, confirmant la presence d’huile de soja dans la composition du produit (63 %) sans fraction proteique. En l’absence de fraction proteique, des prick-tests a l’isotretinoine avaient ete realises, revenus negatifs en lecture immediate, et une introduction de Curacne® a ete faite en hospitalisation de jour sans reaction anaphylactique constatee. Le traitement a donc ete initie en janvier 2016 et poursuivi pendant 4 mois sans complication allergique immediate ou retardee. Il a cependant ete interrompu precocement pour cause de cephalees avec bilan exhaustif (ophtalmologique et scanner) negatif. Discussion L’huile de soja raffinee entrant dans la composition d’isotretinoine commercialisee ne contient pas de fraction proteique et donc aucune fraction potentiellement allergisante. Les laboratoires commercialisant le Procuta®, l’Acnetrait® et le Contracne® ont egalement confirme cette donnee. L’introduction d’isotretinoine dans notre cas clinique s’est deroulee sans complication et remet en question l’interdiction formelle d’isotretinoine chez les sujets allergiques au soja. Conclusion Ce cas permet donc de nuancer la contre-indication formelle de l’isotretinoine chez les patients connus pour avoir des allergies vraies a l’arachide et au soja.

During the 1997 and 1998 Tore Supra shutdown, a first set of new antenna guard limiters was installed on one of the three ion cycloton 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.

Magnetic fluctuations (radial size ? 5 mm) are measured by a cross polarization scattering (CPS) diagnostic in Tore Supra. In the scenario O + ? X, only the poloidal component of the magnetic fluctuations is measured, while both the radial and the poloidal component are measured in the scenario X + ? O. These fluctuations are investigated quantitatively in the ohmic and low confinement regimes over a wide range of plasma currents, densities and additional heating powers. At the same time, the electron heat diffusivities expected from these fluctuations are compared with those obtained by profile analysis. Three main results are obtained: (a) The radial profile of the poloidal magnetic fluctuations in the gradient region (0.3 < r/a < 0.7) is established from these measurements. The magnetic fluctuation levels are found to increase towards the plasma edge, and this feature is compatible with that of electron heat diffusivity. (b) A strong correlation between the measured magnetic turbulence and the local temperature gradient is observed during the additional heating. (c) The local electron heat diffusivity induced by magnetic fluctuations is estimated using the non-collisional quasi-linear formula ?mage = ?qRvth(?Br/B)2, where the radial component of the magnetic turbulence is assumed to be of the same order as the poloidal component. Both the order of magnitude and the parametric dependence of ?mage 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.

Various steady state non-inductive plasmas, with strong electron heating and significant modification of the current density profile, have been routinely obtained in Tore Supra discharges with lower hybrid current drive (LHCD) and/or fast wave electron heating (FWEH) experiments. The dependence of the electron heat diffusivity χe on the electron temperature gradient ∇Te, the magnetic shear s and the safety factor q is demonstrated. The increase of χe with ∇Te indicates the existence of a critical temperature gradient. Moreover, the current density profile affects the global confinement and the local transport. The electron heat flux qe is found to be roughly proportional to q2. The effect of magnetic shear on χe is studied in the improved confinement discharges obtained by modifying the current profile. When the magnetic shear increases in the confinement zone and/or vanishes in the plasma centre, χe decreases. The effect of the current profile is also observed in the saturated ohmic regime. The results for χe do not agree with the local transport models of Taroni et al. or Rebut-Lallia-Watkins (RLW) for all ranges of electron heating.

This study was conducted to examine whether detection of atherosclerotic aortic plaque by multiplane transoesophageal echocardiography could predict the absence or presence of significant coronary artery disease in young and elderly valvular patients.Clinical and angiographic features and transoesophageal echocardiography findings were prospectively analysed in 278 consecutive valvular patients. In 93 patients with significant coronary artery disease, 85 had thoracic aortic plaque on transoesophageal echocardiography studies. In contrast, aortic plaque existed in only 33 of the remaining 185 patients with normal or mildly abnormal coronary arteries. Therefore, the presence of aortic plaque on transoesophageal echocardiography studies had a sensitivity of 91%, a specificity of 82%, and positive and negative predictive values of 72% and 95%, respectively, for significant coronary artery disease. In the 109 patients agedor = 70 years, these sensitivity, specificity, and positive and negative predictive values were 96%, 78%, 79%, and 96%, respectively. The above high negative predictive value was the major finding of this study and indicated that the absence of thoracic plaque is a strong predictor for absence of significant coronary artery disease. There was a significant relationship between the degree of aortic intimal changes and the severity of coronary artery disease (P0.0001). Multivariate logistic regression analysis revealed that aortic plaque, angina, hypercholesterolaemia and age were significant predictors of coronary artery disease: aortic plaque was the most significant independent predictor, even in patientsor = to 70 years.This large prospective study indicates that examination of thoracic atherosclerotic plaque, by multiplane transoesophageal echocardiography, is a marker for coronary artery disease, and is a particularly powerful predictor for absence of significant coronary artery disease in valvular patients, even in the elderly.

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.

Équipe 107 : Physique des plasmas chauds; International audience; A novel method to simulate ion cyclotron wave coupling in the edge of a tokamak plasma with the finite element technique is presented. It is applied in the commercial software COMSOL Multiphysics. Its main features include the perfectly matched layer (PML) technique to emulate radiating boundary conditions beyond a critical cutoff layer for the fast wave (FW), full-wave propagation across the inhomogeneous cold peripheral plasma and a detailed description of the wave launcher geometry. The PML technique, while widely used in numerical simulations of wave propagation, has scarcely been used for magnetized plasmas, due to specificities of this gyrotropic material. A versatile PML formulation, valid for full dielectric tensors, is summarized and interpreted as wave propagation in an artificial medium. The behavior of this technique has been checked for plane waves on homogeneous plasmas. Wave reflection has been quantified and compared to analytical predictions. An incompatibility issue for adapting the PML for forward (FW) and backward (slow wave (SW)) propagating waves simultaneously has been evidenced. In a tokamak plasma, this critical issue is overcome by taking advantage of the inhomogeneous density profile to reflect the SW before it reaches the PML. The simulated coupling properties of a Tore Supra ion cyclotron resonance heating (ICRH) antenna have been compared to experimental values in a situation of good single-pass absorption. The necessary antenna elements to include in the geometry to recover the coupling properties obtained experimentally are also discussed.

After the change over from the C-wall to the ITER-like Be/W wall (ILW) in JET, the radiation losses during ICRF heating have increased and are now substantially larger than those observed with NBI at the same power levels, in spite of the similar global plasma energies reached with the two heating systems. A comparison of the NBI and ICRF performances in the JET-ILW experiments, based on a statistical analysis of ~3000 L-mode discharges, will be presented.

The most recent JET campaign has focused on characterizing operation with the "ITER-like" wall. One of the questions that needed to be answered is whether the auxiliary heating methods do not lead to unacceptably high levels of impurity influx, preventing fusion-relevant operation. In view of its high single pass absorption, hydrogen minority fundamental cyclotron heating in a deuterium plasma was chosen as the reference wave heating scheme in the ion cyclotron domain of frequencies. The present paper discusses the plasma behavior as a function of the minority concentration X[H] in L-mode with up to 4MW of RF power. It was found that the tungsten concentration decreases by a factor of 4 when the minority concentration is increased from X[H] 5% to X[H] 20% and that it remains at a similar level when X[H] is further increased to 30%; a monotonic decrease in Beryllium emission is simultaneously observed. The radiated power drops by a factor of 2 and reaches a minimum at X[H] 20%. It is discussed that poor single pass absorption at too high minority concentrations ultimately tailors the avoidance of the RF induced impurity influx. The edge density being different for different minority concentrations, it is argued that the impact ICRH has on the fate of heavy ions is not only a result of core (wave and transport) physics but also of edge dynamics and fueling.

Magnetically confined plasmas, such as those produced in the tokamak JET, contain measurable amounts of impurity ions produced during plasma-wall interactions (PWI) from the plasma-facing components and recessed wall areas. The impurities, including high- and mid-Z elements such as tungsten (W) from first wall tiles and nickel (Ni) from Inconel structure material, need to be controlled within tolerable limits, to ensure they do not significantly affect the performance of the plasma. This contribution focuses on documenting W and Ni impurity behavior during Ion Cyclotron Resonance Heating (ICRH) operation with the new ITER-Like Wall (ILW). Ni- and W-concentration were derived from VUV spectroscopy and the impact of applied power level, relative phasing of the antenna straps, plasma separatrix ­ antenna strap distance, IC resonance position, edge density and different plasma configuration, on the impurity release during ICRH are presented. For the same ICRH power the Ni and W concentration was lower with dipole phasing than in the case of -p/2 phasing. The Ni concentration was found to increase with ICRH power and for the same NBI power level, ICRH-heated plasmas were characterized by two times higher Ni impurity content. Both W and Ni concentrations increased strongly with decreasing edge density which is equivalent to higher edge electron temperatures and more energetic ions responsible for the sputtering. In either case higher levels were found in ICRH than in NBI heated discharges. When the central plasma temperature was similar, ICRH off-axis heating resulted in higher core Ni impurity concentration in comparison to on-axis ICRH in L-mode. It was also found that the main core radiation during ICRH came from W.

ITER will start its operation with non-activated Hydrogen and Helium plasmas at a reduced magnetic field of B0 = 2.65T. In Hydrogen plasmas, the two Ion Cyclotron Resonance Frequency (ICRF) heating schemes available for central plasma heating (fundamental H majority and 2nd harmonic 3He minority ICRF heating) are likely to suffer from relatively low RF wave absorption, as suggested by numerical modelling and confirmed by previous JET experiments conducted in conditions similar to those expected in ITER's initial phase. With 4He plasmas, the commonly adopted fundamental H minority heating scheme will be used and its performance is expected to be much better. However, one important question that remains to be answered is whether increased levels of Hydrogen (due to e.g. H pellet injection) jeopardize the high performance usually observed with this heating scheme, in particular in a full-metal environment. Recent JET experiments performed with the ITER-like wall (ILW) shed some light into this question and the main results concerning ICRF heating performance in L-mode discharges are summarized here.

In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall materials on the JET Ion Cyclotron Resonance Frequency (ICRF) operation is assessed and some important properties of JET plasmas heated with ICRF are highlighted. A ~20% reduction of the antenna coupling resistance is observed with the ILW as compared with the JET carbon (JET-C) wall. Heat-fluxes on the protecting limiters close the antennas, quantified using Infra-Red (IR) thermography (maximum 4.5MW/m2 in current drive phasing), are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. The location of the tungsten ICRF specific source could not be identified but some experimental observations indicate that main-chamber W components could be an important impurity source: for example, the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions and Be evaporation in the main chamber results in a strong reduction of the impurity level. In L-mode plasmas, the ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 15%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating efficiency; the power is typically deposited at the plasma centre while the radiation is mainly from the outer part of the plasma bulk. Application of ICRF heating in H-mode plasmas has started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core has been observed.

The major aspects linked to the use of the JET auxiliary heating systems: NBI, ICRF and LHCD, in the new JET ITER-like wall (JET-ILW) are presented. We show that although there were issues related to the operation of each system, efficient and safe plasma heating was obtained with room for higher power. For the NBI up to 25.7MW was safely injected; issues that had to be tackled were mainly the beam shine-through and beam re-ionisation before its entrance into the plasma. For the ICRF system, 5MW were coupled in L-mode and 4MW in H-mode; the main areas of concern were RF-sheaths related heat loads and impurities production. For the LH, 2.5MW were delivered without problems; arcing and generation of fast electron beams in front of the launcher that can lead to high heat loads were the keys issues. For each system, an overview will be given of: the main modifications implemented for safe use, their compatibility with the new metallic wall, the differences in behavior compared with the previous carbon wall, with emphasis on heat loads and impurity content in the plasma.

When using Ion Cyclotron Range of Frequency (ICRF) heating, enhanced power deposition on Plasma-Facing Components (PFCs) close to the antennas can occur. Experiments have recently been carried out on JET with the new ITER-Like-Wall (ILW) to characterize the heat fluxes on the protection of the JET ICRF antennas, using Infra-Red (IR) thermography measurement. The measured heat flux patterns along the poloidal limiters surrounding powered antennas were compared to predictions from a simple RF sheath rectification model. The RF electric field, parallel to the static magnetic field in front of the antenna, was evaluated using the TOPICA code, integrating a 3D flattened model of the JET A2 antennas. The poloidal density variation in front of the limiters was obtained from the mapping of the Li-beam or edge reflectometry measurements using the flux surface geometry provided by EFIT equilibrium reconstruction. In many cases, this simple model can well explain the position of the maximum heat flux on the different protection limiters and the heat-flux magnitude, confirming that the parallel RF electric field and the electron plasma density in front of the antenna are the main driving parameters for ICRF-induced local heat fluxes.

During the initial operation of the JET ITER-like wall, particular attention was given to the characterization of the Ion Cyclotron Resonance Frequency (ICRF) heating in this new metallic environment. In this contribution we compare L-modes plasmas heated by ICRF or by Neutral Beam Injection (NBI). ICRF heating as expected led to a much higher centrally peaked power deposition on the electrons and due to the central fast ion population to stronger sawtooth activity. Surprisingly, although a higher bulk radiation was observed during the ICRF phase, the thermal plasma energy was found similar for both cases, showing that a higher radiation inside the separatrix was not incompatible with an efficient central heating scheme. The higher radiation was attributed to the presence Tungsten (W). Tomographic inversion of SXR emissions allowed a precise observation of the sawtooth effect on the radiation pattern. W concentration profiles deconvolved from SXR emission showed the flattening of the profiles due to sawtooth for both heating and the peaking of the profiles in the NBI case only hinting for extra transport effect in the ICRF case.

In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall material on the JET Ion Cyclotron Resonance Frequency (ICRF) operation was assessed and also the properties of JET plasmas heated with ICRF were studied. No substantial change of the antenna coupling resistance was observed with the ILW as compared with the carbon wall. Heatfluxes on the protecting limiters close the antennas quantified using Infra-Red (IR) thermography (maximum 4.5MW/m2 in current drive phasing) are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can well reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. Some experimental facts indicate that main-chamber W components could be an important impurity source: the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions; the W content is also increased in ICRF-heated limiter plasmas; and Be evaporation in the main chamber results in a strong and long lasting reduction of the impurity level. The ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 20%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating efficiency; The ICRF power can be deposited at plasma centre and the radiation is mainly from the outer part of the plasma. Application of ICRF heating in H-mode plasmas started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core could be observed.

Fully dynamic Stark effect visible spectroscopy was used for the first time to directly measure the local rf electric field in the boundary plasma near a high-power antenna in high-performance, magnetically confined, fusion energy experiment. The measurement was performed in the superconducting tokamak Tore Supra, in the near field of a 1--3 MW, lower-hybrid, 3.7 GHz wave-launch antenna, and combined with modeling of neutral atom transport to estimate the local rf electric field amplitude (as low as $1--2\text{ }\text{ }\mathrm{kV}/\mathrm{cm}$) and direction in this region. The measurement was then shown to be consistent with the predicted values from a 2D full-wave propagation model. Notably the measurement confirmed that the electric field direction deviates substantially from the direction in which it is launched by the waveguides as it penetrates only a few cm radially inward into the plasma from the waveguides, consistent with the model.

20th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI), Forschungszentrum Julich, Aachen, GERMANY, MAY 21-25, 2012; International audience; The influence of RF electric fields on retarding field analyzer (RFA) measurements of sheath potential, V-sh is investigated. One-dimensional particle-in-cell simulations show that the RFA is able to measure reliably the rectified dc sheath potential only for ion plasma frequencies omega(pi) similar to the rf wave frequency omega(rf), while for real SOL conditions (omega(pi) > omega(rf)), when the RFA is magnetically connected to an RF antenna, it is strongly underestimated. An alternative method to investigate RF sheaths effects is proposed that uses broadening of the ion distribution function as evidence of the rf electric fields in the sheath. RFA measurements in Tore Supra indicate that the average effects of rf potentials do indeed propagate from the antenna 12 m along magnetic field lines.

When using Ion Cyclotron Range of Frequency (ICRF) heating, enhanced heat-fluxes are commonly observed on some plasma facing components close to the antennas. Experiments have recently been carried out on JET with the new ITER-Like-Wall (ILW) to characterize the heat flux to the JET ICRF antennas. Using Infra-Red thermography and thermal models of the tiles, heat-fluxes were evaluated from the surface temperature increase during the RF phase of L-mode plasmas. The maximum observed heat-flux intensity was ~ 4.5 MW/m2 when operating with -{\pi}/2 current drive strap phasing at power level of 2MW per antenna and with a 4 cm distance between the plasma and the outer limiters. Heat-fluxes are reduced when using dipole strap phasing. The fraction of ICRF power deposited on the antenna limiters or septa was in the range 2-10% for dipole phasing and 10-20% with +/-{\pi}/2 phasing., Comment: 22 pages, 6 figures

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 discharge, 3 types of Radio Frequency (RF) additional heating systems have been developed: Lower Hybrid Current Drive (LHCD), Ion Cyclotron Resonance Heating (ICRH) and Electron Cyclotron Resonance Heating (ECRH) [2]. To cope with long lasting discharges (up to 380 s × 2.8 MW) and large RF additional heating power (12.3 MW × 3 s), Actively Cooled (AC) Plasma Facing Components (PFC) were deployed in TORE-SUPRA for the first time in a Tokamak environment. TORE-SUPRA is now being modified into an axisymmetric tokamak with actively cooled tungsten main chamber walls and a divertor, the WEST project (W - for tungsten - Environment in Steady-state Tokamak) [3]. This new facility has the objective to offer ITER a test bed for validating the relevant actively cooled metallic technologies in D-shape H-mode plasmas. In contrast to other metallic devices such as JET and ASDEX, WEST will rely only on RF additional power systems. A set of plasma scenarios have been identified, ranging from a high total RF power scenario up to 15 MW during 30 seconds, to a high fluence scenario of 1000 seconds with up to 10 MW of injected RF power. These scenarios are able to reproduce ITER relevant conditions of steady state heat loads of 10 to 20 MW/m, to test tungsten actively cooled divertor technologies with relevant power heat fluxes and particle fluence. The paper presents the main issues regarding WEST project and especially the additional RF power injection systems (2 LHCD antennas, 3 + 4 = 7 MW continuous wave and 3 ICRH antennas, 3 × 3 = 9 MW-30 s or 3 MW-1000 s) for WEST. The front face of the LHCD antennas will be modified to account for the different plasma position and smaller toroidal field ripple, due to the more inward antenna position in the vessel. No other modifications are needed on the Passive-Active Multijunction (PAM) or the Fully-Active Multijunction (FAM) LHCD antennas, or the associated generator (2 × 8 klystrons, 600 kW each CW). Concerning the ICRH system, the main challenges are its ELM-resilience, its compatibility with continuous operation, and the interaction of the RF near fields with neighbouring plasma facing components. 3 new actively cooled antennas are being designed to be matched with an ELMs resilient electric circuit. The proposed solution is based on the JET-EP antenna and CEA prototype tested in 2007, both having identical internal conjugate-T electrical layout and a demonstrated load resilience capacity to plasma edge transients during ELMs. © 2013 IEEE.

The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40-55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R&D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper highlights progress and outstanding issues for the various system components. (C) 2013 ITER Organization. Published by Elsevier B.V. All rights reserved.

The ITER ion-cyclotron range of frequency (ICRH) heating system is required to couple 20 MW of power in the frequency range 40-55 MHz for a large range of scenarios with Edge Localized Modes. To mitigate the associated risks, it is foreseen to design and install on ITER two port-plug antennas for a total of 20 MW coupled power on long pulse operation [1]. The CEA activity to this antenna design within the CYCLE consortium was focused on the Faraday screen design (Fig. 1) and associated radio frequency (RF) sheath modelling, the reflectometers design for edge density measurement in the antenna vicinity and a contribution to the remote handling/tooling of the antenna. This paper is an overview of each item and proposes some R&D activities on key components. © 2013 Elsevier B.V. All rights reserved.

A comparison of the ASDEX Upgrade 3-strap ICRF antenna data with the linear electro-magnetic TOPICA calculations is presented. The comparison substantiates a reduction of the local electric field at the radially protruding plasma-facing elements of the antenna as a relevant approach for minimizing tungsten (W) sputtering in conditions when the slow wave is strongly evanescent. The measured reaction of the time-averaged RF current at the antenna limiters to the antenna feeding variations is less sensitive than predicted by the calculations. This is likely to have been caused by temporal and spatial fluctuations in the 3D plasma density distribution affected by local non-linear interactions. The 3-strap antenna with the W-coated limiters produces drastically less W sputtering compared to the W-coated 2-strap antennas. This is consistent with the non-linear asymptotic SSWICH-SW calculations for RF sheaths. © 2017 IOP Publishing Ltd.

For the first time, the internal magnetic turbulence is measured by a new cross polarization scattering diagnostic in the Tore Supra tokamak. The principle of this experiment is presented. It is based on the polarization change or mode conversion of the electromagnetic wave scattered by magnetic fluctuations. A strong correlation between the internal magnetic fluctuation and the additional heating is observed, contrary to the edge fluctuations. The observed fluctuations increase linearly with the poloidal beta number {beta}{sub {ital p}} in the {ital L}-mode confinement regime.

Several discharges have been analysed in Tore Supra in order to improve the understanding of anomalous transport. In ohmic discharges, it has been shown that both particle and heat diffusion coefficients improve with increasing density, then saturate. The level of edge density fluctuations behave in the same way. In the presence of additional heating, the bulk and the edge turbulences behave differently. The level of edge density fluctuations remains constant in most discharges with additional heating (up to 8 MW). In practice, it only depends on edge conditions. On the contrary, the level of core magnetic fluctuations increases systematically with any additional power. Finally, a lowering of the level of the edge density fluctuations has been observed when using the Tore Supra ergodic divertor. This improvement is accompanied by the onset of a temperature pedestal between the plasma bulk and the ergodic layer. This explains why the global confinement remains unchanged in presence of the ergodic divertor in spite of the reduced radius of the last closed magnetic surface.

Introduction Le syndrome DRESS pour drug reaction with eosinophilia and systemic symptoms est une toxidermie severe. Les PCI sont reputes donner de nombreuses reactions immediates et retardees en grande partie liees a l’osmolarite et/ou a la charge ionique. Tres peu de cas de DRESS aux PCI ont ete rapportes. Suite a deux cas nantais de DRESS imputables aux PCI avec un delai de survenu non conventionnel, nous avons consulte la base nationale de pharmacovigilance (BNPV) concernant la survenue de DRESS induite/associee a ces medicaments. Methodes Nous avons effectue une recherche dans la BNPV en utilisant le terme preferentiel « syndrome DRESS » ou « hypersensibilites medicamenteuses avec eosinophilie et symptomes systemiques » et le nom commercial un par un des PCI. Le score de REGISCAR a ete calcule pour ne conserver que les scores superieurs ou egaux a 4–5. Resultats Vingt dossiers ont ete analyses correspondant aux criteres de recherche avec une mediane d’âge de 55,5 ans (6 hommes et 14 femmes). Cinq cas (mediane d’âge = 58 [52 ; 77]) ont ete retenus car seul le PCI etait imputable au DRESS, confirmes par tests cutanes (patch-test et IDR). Un autre cas a ete confirme par TPO-IV avec recidive du DRESS sans signe de gravite. Un aspect interessant etait le delai de survenue tres rapide (mediane de 48 h [24 h ; 96 h]). Six autres cas (mediane d’âge = 58 [30 ; 77]) ont ete choisis sur l’association de plusieurs familles medicamenteuses responsables du DRESS avec un delai median de 3,5 sem (2 ; 8). De maniere interessante dans ces cas la, le delai de survenue entre l’injection du PCI et le DRESS est egalement court avec une mediane de 3,5 jours (1 ; 15). Discussion La survenue d’un DRESS syndrome est en moyenne de 15 jours a 3 mois apres l’exposition a un medicament tel que le montre la cohorte REGISCAR. Dans notre serie de cas, le delai de survenue apparait tres court. De plus, les PCI semblent avoir un effet potentialisateur sur le DRESS dans un delai tres bref. Conclusion Les PCI semblent donc pourvoyeurs de DRESS et les prises ponctuelles de ces medicaments ne doivent pas faire oublier le role de ces produits dans l’apparition de cet effet grave meme si les delais sont courts.

Introduction Le syndrome Dress « Drug reaction with eosinophilia and systemic symptoms » est une toxidermie severe avec un taux de mortalite estime entre 5 a 10 %. Les effets cutanes rapportes dans le resume des caracteristiques du produit (RCP) ou les ouvrages de reference sur la spiramycine concernent des eruptions et tres rarement des pustuloses exanthematiques aigues disseminees. Suite a un cas nantais de Dress probable survenu apres la prise de spiramycine, nous avons recherche dans la base nationale de pharmacovigilance (BNPV) la survenue de Dress. Methodes Une recherche dans la BNPV utilisant le terme preferentiel « reactions medicamenteuses avec eosinophilie et symptomes systemiques » et le principe actif « spiramycine » a ete effectuee. Le score Regiscar a ete calcule pour ne conserver que ceux superieurs ou egaux a 2–3. Une revue de la litterature ainsi que des ouvrages de reference et des RCP a ete realisee. Resultats Nous avons retenu 11 cas concernant 5 hommes et 6 femmes de 25 a 74 ans. Sept cas correspondent a un traitement par Rodogyl ® ou Birodogyl ® et 4 cas a la Rovamycine ® . Associee au metronidazole, les indications correspondent 7 fois a des pathologies dentaires. La Rovamycine ® a ete prescrite 3 fois pour pneumopathie et une fois pour une toxoplasmose. Les scores Regiscar se situent entre 3 et 7. Dans 8 cas, les tests cutanes sont non effectues ou en attente, realises, ils se revelent 1 fois negatifs et 2 fois positifs a la Rovamycine ® . Discussion Les RCP et les ouvrages de reference ne mentionnent pas de Dress, 7 cas sont associes avec le metronidazole, non pourvoyeur de Dress. Dans la litterature, 3 cas de Dress avec azithromycine et un avec roxithromycine sont rapportes, aucun avec la spiramycine. Les cas, meme rares, rapportes a la BNPV suite a l’utilisation du score de Regiscar apparaissent comme de probables Dress dus a la spiramycine. Conclusion Au regard de la litterature, l’utilisation des macrolides semble associee a un risque faible de Dress. Ce risque n’a pas ete etudie a partir de donnees de notifications spontanees ciblant la classe des macrolides, une etude de la BNPV va etre realisee suite a ce travail afin d’apporter plus de connaissances concernant cette classe de medicaments et le Dress.

International audience; Most present Ion Cyclotron Resonant Frequency (ICRF) heating codes and antenna codes assume the antenna sitting in a vacuum region and consider the fast wave only, which implicitly performs an abrupt density transition from vacuum to above lower hybrid (LH) resonance. The impact of the appearance of the LH resonance is entirely overlooked in their simulations. We studied the impact of densities that decay continuously inside the antenna box on near field patterns and power coupling. A new full wave code based on the COMSOL Finite Element Solver has been developed to investigate this topic. It is shown that: up to the memory limits of the adopted workstation, the local RF field pattern in low-density regions below the LH resonance changes with the grid size. Interestingly and importantly, however, the total coupled toroidal spectrum is almost independent on the mesh size and is weakly affected by the presence of the density profile inside the antenna box in dipole toroidal strap phasing. This suggests one can drop out this density for coupling studies to speed up the computation. Simulation also shows that varying the density gradient in the fast wave evanescence region has no significant effect on wave coupling.

International audience; Tore Supra routinely addresses the physics and technology of very long-duration plasma discharges, thus bringing precious information on critical issues of long pulse operation of ITER. A new ITER relevant lower hybrid current drive (LHCD) launcher has allowed coupling to the plasma a power level of 2.7 MW for 78 s, corresponding to a power density close to the design value foreseen for an ITER LHCD system. In accordance with the expectations, long distance (10 cm) power coupling has been obtained. Successive stationary states of the plasma current pro le have been controlled in real-time featuring (i) control of sawteeth with varying plasma parameters, (ii) obtaining and sustaining a `hot core' plasma regime, (iii) recovery from a voluntarily triggered deleterious magnetohydrodynamic regime. The scrape-off layer (SOL) parameters and power deposition have been documented during L-mode ramp-up phase, a crucial point for ITER before the X-point formation. Disruption mitigation studies have been conducted with massive gas injection, evidencing the difference between He and Ar and the possible role of the q = 2 surface in limiting the gas penetration. ICRF assisted wall conditioning in the presence of magnetic eld has been investigated, culminating in the demonstration that this conditioning scheme allows one to recover normal operation after disruptions. The effect of the magnetic eld ripple on the intrinsic plasma rotation has been studied, showing the competition between turbulent transport processes and ripple toroidal friction. During dedicated dimensionless experiments, the effect of varying the collisionality on turbulence wavenumber spectra has been documented, giving new insight into the turbulence mechanism. Turbulence measurements have also allowed quantitatively comparing experimental results with predictions by 5D gyrokinetic codes: numerical results simultaneously match the magnitude of effective heat diffusivity, rms values of density uctuations and wavenumber spectra. A clear correlation between electron temperature gradient and impurity transport in the very core of the plasma has been observed, strongly suggesting the existence of a threshold above which transport is dominated by turbulent electron modes. Dynamics of edge turbulent uctuations has been studied by correlating data from fast imaging cameras and Langmuir probes, yielding a coherent picture of transport processes involved in the SOL.

This paper treats self‐consistently the interplay between slow magnetosonic wave penetration and edge plasma DC biasing using a two‐field fluid approach. The RF and DC parts are coupled by non‐linear RF and DC sheath boundary conditions at lateral boundaries of the simulation domain. Within this approach two mechanisms of RF‐induced DC current flows are outlined, depending on the toroidal symmetry of the RF drive. The numerical implementation of the model using COMSOL, its validation against explicit simplified solutions and its planned improvements are discussed.

The interaction between a powered ICRH antenna and plasma scrape‐off layer is investigated by Langmuir probes. When a reciprocating Langmuir probe is magnetically connected to a powered ICRH antenna, large RF‐induced potentials are observed along the entire leading edge of the antennaSs poloidal protection limiters. Recent experiments show that a key parameter affecting the amplitude of these potentials is plasma density. The detailed 2D map of the potentials induced in front of the ICRH antenna is presented as well as the influence of the density on these potentials and parallel flow.

We investigate possible mechanisms for radial DC and RF current flow by running a set of simulations based on a fluid modeling with flute hypothesis. They clearly demonstrate the impact of DC transverse currents in perpendicular potential gradient regions. We propose a modeling coupling parallel currents driven by RF sheaths, both RF and DC transverse currents, and plasma density in biased flux tubes to reconstruct IV probe characteristics when they are connected to an antenna.

Heat loads from ICRF and LH wave absorption in the SOL are characterized on JET from the de‐convolution of surface temperatures measured by infrared thermography. The spatial localization, quantitative estimates, parametric dependence and physical origin of the observed heat fluxes are documented. Implications of these observations are discussed for the operation of JET with an ITER‐Like Wall, featuring Beryllium tiles with reduced power handling capability.

In the past years, one of the focal points of the JET experimental programme was on ion-cyclotron resonance heating (ICRH) studies in view of the design and exploitation of the ICRH system being developed for ITER. In this brief review, some of the main achievements obtained in JET in this field during the last 5 years will be summarized. The results reported here include important aspects of a more engineering nature, such as (i) the appropriate design of the RF feeding circuits for optimal load resilient operation and (ii) the test of a compact high-power density antenna array, as well as RF physics oriented studies aiming at refining the numerical models used for predicting the performance of the ICRH system in ITER. The latter include (i) experiments designed for improving the modelling of the antenna coupling resistance under various plasma conditions and (ii) the assessment of the heating performance of ICRH scenarios to be used in the non-active operation phase of ITER. © 2011 IOP Publishing Ltd.

The impact of anomalous radial diffusion due to turbulent magnetic fields on the electrical conductivity of a toroidal plasma is investigated. The problem is solved by evaluating an appropriate three-dimensional response function for the current density. This yields an analytic expression of the turbulent electrical conductivity including trapped particle effects. In the pressure a gradient region, the conductivity is generally enhanced with respect to the neoclassical one. This results in an anomalously low plasma resistance-a trend observed in many tokamak experiments. The analytic expression is shown to be in very good agreement with numerical calculations using a three-dimensional bounce averaged Fokker-Planck code. By a similar method, the toroidal response function for radiofrequency current drive in the presence of magnetic turbulence is also found

International audience; Measurements show that the vicinity of powered ion cyclotron range of frequency (ICRF) antennae is biased positively with respect to its environment [ J. Gunn et al., Proc. 22nd IAEA Fusion Energy Conference, Geneva 2008, EX/P6-32 ]. This is attributed to radio-frequency (rf) sheath rectification. The radial penetration of these direct current (dc) potentials from ICRF launchers into the tokamak scrape-off layer (SOL) determines the power deposition on the walls and especially on the antenna structure, which is a key point for long time clean discharges. Within independent flux tube models of rf sheath rectification the radial penetration of dc potentials is determined by the skin depth x0 = c/ωpe for the slow wave. When self-consistent exchange of transverse rf current is allowed between neighboring flux tubes, such a structure can be broadened radially up to a characteristic transverse length L. Broadening arises as soon as L>x0. A linear modeling of the process gives a first evaluation of the theoretical length L = (L∥ρci/2)1/2. Within the “flute assumption,” it scales with the length L∥ of open flux tubes and the ion Larmor radius Ωci. This trend has been confirmed by nonlinear fluid simulations using the SEM code taking into account nonlinearities of the sheath dynamics. Parametric regimes are outlined where broadening or nonlinearity arise. Langmuir probe measurements on Tore Supra suggest that the observed broadening is lower than predicted by the code. This suggests that actual rf current exchanges probably do not occur over the whole length of magnetic field lines but only on a fraction of it. This “effective parallel magnetic connection length” L∥eff is estimated from the measurements. The model is finally applied to several potential maps generated by an ITER antenna, with different plasma parameters depending on possible SOL scenarios in ITER, and “reasonable assumptions” about L∥eff. It comes out that L ranges between 1 and 10 cm depending on local L∥eff and on typical ITER plasma parameters.

Radio frequency (RF) sheaths are suspected of limiting the performance of present-day ion cyclotron range of frequencies (ICRFs) antennas over long pulses and should be minimized in future fusion devices. Within the simplest models, RF-sheath effects are quantified by the integral V RF = ∫ E ∥ · dl where the parallel RF field E ∥ is linked with the slow wave. On ‘long open field lines’ with large toroidal extension on both sides of the antenna it was shown that V RF is excited by parallel RF currents j ∥ flowing on the antenna structure. In this paper, the validity of this simple sheath theory is tested experimentally on the Tore Supra (TS) ITER-like antenna prototype (ILP), together with antenna simulation and post-processing codes developed to compute V RF. The predicted poloidal localization of high-|V RF| zones is confronted to that inferred from experimental data analysis. Surface temperature distribution on ILP front face, as well as ILP-induced modifications of RF coupling and hot spots on a magnetically connected lower hybrid current drive antenna, indicates local maxima of dc plasma potential in both the upper and lower parts of the ILP. This result, qualitatively conforming to V RF simulations, is interpreted in terms of j ∥ flowing on ILP frame. Once the validation is done, such reliable theoretical models and numerical codes are then employed to provide predictive results. Indeed, we propose two ways to reduce |V RF| by acting on j ∥ on the antenna front face. The first method, more adapted for protruding antennas, consists of avoiding the j ∥ circulation on the antenna structure, by slotting the antenna frame on its horizontal edges and by partially cutting the Faraday screen rods. The second method, well suited for recessed antennas, consists of compensating j ∥ of opposite signs along long flux tubes, with parallelepiped antennas aligned with (tilted) flux tubes. The different concepts are assessed numerically on a two-strap TS antenna phased [0, π] using near RF fields from the antenna code TOPICA. Simulations stress the need to suppress all current paths for j ∥ to substantially reduce |V RF| over the whole antenna height.

The main results of the Tore Supra experimental programme in the years 2007–2008 are reported. They document significant progress achieved in the domain of steady-state tokamak research, as well as in more general issues relevant for ITER and for fusion physics research. Three areas are covered: ITER relevant technology developments and tests in a real machine environment, tokamak operational issues for high power and long pulses, and fusion plasma physics. Results presented in this paper include test and validation of a new, load-resilient concept of ion cycotron resonance heating antenna and of an inspection robot operated under ultra-high vacuum and high temperature conditions; an extensive experimental campaign (5 h of plasma) aiming at deuterium inventory and carbon migration studies; real-time control of sawteeth by electron cyclotron current drive in the presence of fast ion tails; ECRH-assisted plasma start-up studies; dimensionless scalings of transport and turbulence; transport experiments using active perturbation methods; resistive and fast-particle driven MHD studies. The potential role of Tore Supra in the worldwide fusion programme before the start of ITER operation is also discussed.