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Before the installation of the first ICRF launcher in WEST, dedicated RF, cold and hot leak tests have been performed in the TITAN test bed [ 1 , 2 ] (Bernard et al., 2011a,b). All the water circuits have been controlled in order to avoid any in-vessel leak. The impedance matching has been checked at low level (mW range) and the high RF voltage standoff has been validated at the nominal value of 27 kV peak. During the WEST C3a campaign, the commissioning of the first antenna has been successfully performed. About 0.6 MW of power has been coupled to the plasma, and the launcher’s load-resilience has been proved.

International audience; The paper discusses the characterization of the three high power steady-state and load-resilient ICRF launchers of WEST before their installation in the tokamak. These launchers have been characterized and validated in low-power experiments (milliwatt range) as well as in experiments at the nominal RF voltages and currents in the TITAN vacuum chamber ($\sim$30 kV and 915 A peak). The successful commissioning of two of the launchers during the WEST C3 campaign at $\sim$1 MW power level is illustrated. Manual and real-time controlled impedance-matching of the launchers are discussed, as well as the validation of their load-resilience. Furthermore, several redundant and complementary protection systems have been validated and are reviewed in the paper.

Embedded RF contacts are integrated within the ITER ICRH launcher to allow assembling, sliding and to lower the thermo-mechanical stress. They have to withstand a peak RF current up to 2.5 kA at 55 MHz in steady-state conditions, in the vacuum environment of themachine.The contacts have to sustain a temperature up to 250{\textdegree}Cduring several days in baking operations and have to be reliable during the whole life of the launcher without degradation. The RF contacts are critical components for the launcher performance and intensive R&D is therefore required, since no RF contactshave so far been qualified at these specifications. In order to test and validate the anticipated RF contacts in operational conditions, CEA has prepared a test platform consisting of a steady-state vacuum pumped RF resonator. In collaboration with ITER Organization and the CYCLE consortium (CYclotronCLuster for Europe), an R&D program has been conducted to develop RF contacts that meet the ITER ICRH launcher specifications. A design proposed by CYCLE consortium, using brazed lamellas supported by a spring to improve thermal exchange efficiency while guaranteeing high contact force, was tested successfully in the T-resonator up to 1.7 kA during 1200 s, but failed for larger current values due to a degradation of the contacts. Details concerning the manufacturing of the brazed contacts on its titanium holder, the RF tests results performed on the resonator and the non-destructive tests analysis of the contacts are given in this paper., Comment: Fusion Engineering and Design, Elsevier, 2015, Available online 21 April 2015

The WEST (Tungsten (W) Environment in Steady-state Tokamak) is an upgrade of Tore-Supra (TS) which aims it into an X -point magnetic configuration tokamak equipped with an actively cooled tungsten divertor. To be a platform of ITER technologies of high heat flux components testing, three sets of Ion Cyclotron Resonant Heating (ICRH) antennas have been designed to inject 9 MW during 30 s or 3 MW during 1000 s. The antenna design is based on a load resilient prototype successfully tested in Tore Supra in 2007. In order to allow continuous-wave (CW) operations, the mechanical design of the WEST ICRH antenna is emphasized on its cooling performances by designing fully active cooling structure. Two kinds of cooling water loops are used, with temperature and pressure of 70 °C/30 bar and 25 °C/5.2 bar, respectively. The hot water loop is used for the Faraday screen (FS) and the housing box (HB), while the cold water loop is used for the straps, the matching capacitors and the impedance transformer. To enhance the heat removal ability and control the pressure drop, the cooling channels in the FS and HB are drilled directly and parallel connected as much as possible. By performing the hydraulic–thermal analysis, the lack of cooling efficiency was found in the front face of lateral collector where 1 MW/m 2 is imposed and fluid dead zones were found in some of the bars. After optimization, the cooling performance of the cooling circuit increased significantly. With a mass flow rate of 2.5 kg/s, the total pressure drop is 3.1 bar, and the peak temperatures on the FS and HB are 500 °C and 261 °C, respectively. Besides, no cavitation is found in the total cooling circuit.

International audience; ITER is a large scale fusion experimental device under construction in Cadarache (France) intended to prove the viability of fusion as an energy source. Ion Cyclotron Resonance Heating (ICRH) system is one of the three heating systems which will supply total heating power of 20 MW (40-55 MHz) up to one hour of operation. Radio-Frequency (RF) contacts are integrated within the antennas for assembly and operation considerations, which will face extremely harsh service conditions, including neutron irradiation, heavy electrical loads (RF current reaches up to 2 kA with a linear current density of 4.8 kA/m) and high thermal loads. Based on the thermal analysis, the contact resistance is expected to be lower than 7 mΩ to keep the maximum temperature on the louvers lower than 250°C. Few weeks of vacuum (~10$^{-5}$ Pa) baking at 250°C for outgassing is expected before each plasma experimental campaign, under which the RF contact materials' mechanical properties change and diffusion phenomena between different materials are inevitable. CuCrZr and 316L are proper base materials for ITER RF contact louvers and conductors respectively. In order to improve the RF contact's wear and corrosion resistivity as well as to reduce the contact resistance, Au-Ni and Rh functional layers could be electroplated on CuCrZr and 316L accordingly. The application of the Au-Ni/Rh coating pairs is assessed through the thermal ageing and diffusion tests. Wear and electrical contact performances of the Au-Ni/Rh pairs are deeply studied on a dedicated tribometer operated at ITER relevant conditions.

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m2 range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program. WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.

This year Tore-Supra celebrated its 25 years of operation. During this long time, a number of technologies have been developed. First of all, it was mandatory to develop reliable superconducting magnets at ∼1.8 K, with superfluid helium as an efficient coolant. For the production of steady state discharge, three types of radio frequency (RF) additional heating systems have been developed: 1) lower hybrid current drive; 2) ion cyclotron resonance heating; and 3) electron cyclotron resonance heating. 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 were deployed in Tore-Supra for the first time in a tokamak environment. Tore-Supra is now being modified into a D-shape axisymmetric tokamak with AC tungsten main chamber walls and a divertor, the WEST project (W-for tungsten-environment in steady-state tokamak). This new facility has the objective to offer ITER a test bed for validating the relevant AC metallic technologies in D-shape H-mode plasmas. In contrast to other metallic devices, such as JET and ASDEX Upgrade, WEST will rely only on the RF additional power systems. A set of plasma scenarios have been identified, ranging from a high total RF power scenario up to 15 MW in 30 s, to a high fluence scenario of 1000 s with up to 10 MW of injected RF power. These scenarios are able to reproduce ITER relevant conditions of steady state heat loads of 10-20 MW/m2, to test tungsten AC divertor technologies with relevant power heat fluxes and particle fluence.

After the installation and commissioning of the TITAN [1] , [2] , test facility, a key element – the T resonator – was assembled in order to facilitate testing components at high RF voltages and currents. This work is within the CEA roadmap for testing ITER ICRH components in a relevant environment. Several components of the future ITER ICRH antenna have been targeted. The embedded RF contact within the ITER ICRH antenna appeared as a critical component for antenna performance, requiring extensive R&D. Therefore, CEA has proposed and subsequently prepared a platform to test and validate the anticipated RF contact. A steady state resonator with active water cooling has been manufactured and assembled within the TITAN facility, including a hot pressurized water loop. The program consists of testing the contact at 2.25 kA and 62 MHz in steady state conditions. Sliding tests are also performed at high temperature and vacuum to understand component aging, including wear. The equipment installed is consistent with that required to test an ITER ICRH extensively.

First CW test bed, devoted for ion cyclotron resonance heating (ICRH), is being built at CEA Cadarache. It has been designed for testing the ICRH antenna sub assemblies under ITER relevant conditions (vacuum, cooling and RF). This paper presents a technical overview of these facilities and discusses their future operations in the framework of the ITER ICRH European R&D program.

A new Passive Active Multijunction (PAM) Lower Hybrid heating and Current Drive (LHCD) launcher, has been successfully manufactured and tested on Tore-Supra (TS). The design and the fabrication of this new actively cooled launcher based on the PAM concept, as the present ITER LHCD design, is a major component of the TS CIMES project (Components for the Injection of Matter and Energy in Steady-state), and will play a key role in the TS near term program. To achieve 1000 s pulses with a power flux of 25 MW/m 2 the PAM launcher has been designed for steady state (CW) operation (active cooling) with the objective of coupling 2.7 MW of LHCD power to the plasma at 3.7 GHz with a parallel index N ∥ = 1.7 ± 0.2. The launcher has achieved its qualifications tests, i.e. low power Radio Frequency measurements, vacuum and hydraulic leak tests, and has been installed on Tore-Supra tokamak in September 2009. It is commissioning on plasma started a month later, quickly achieving its design performance of 2.7 MW on a 35 s pulse. After a technical description of the PAM, this paper presents an overview of the project phases (RF optimization, manufacturing and qualification) and concludes with the first experimental results of the PAM.

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

Ten years after the first ion cyclotron range of frequency (ICRF) heated Tore Supra (TS) plasma shot, and as TS facility, now equipped with the new ‘CIEL’ high power handing in-vessel components (This SOFT), is progressing towards long pulse performances, the main results and experience gained from the ICRF system are reviewed. An overview of the present status of TS ICRF facility is given, with highlights on the main modifications made to the initial plant and associated results. TS ICRF system achievements are then reported from a large ICRF heated pulses database covering ten experimental campaigns (1991–2001). The main technological issues encountered, especially those relevant to next step ICRF system design, are also discussed. Finally, in view of extended pulse length performances on TS, the present limitations of the ICRF facility and future plans are outlined.

Three new ion cyclotron resonance heating (ICRH) launchers have been designed for the WEST project (W-Tungsten Environment in Steady-state Tokamak) in order to operate at 3 MW/launcher for 30 s and 1 MW/launcher for 1000 s on H-mode plasmas. These new launchers will be to date the first ICRH launchers to offer the unique combination of continuous-wave (CW) operation at high power and load tolerance capabilities for coupling on H-mode edge. The radio-frequency (RF) design optimization process has been carried out using full-wave electromagnetic solvers combined with electric circuit calculations. Cavity modes occurring between the launchers structures and the vacuum vessel ports have been evaluated and cleared out.

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.

International audience; With WEST (Tungsten Environment in Steady State Tokamak) (Bucalossi et al 2014 Fusion Eng. Des. 89 [http://dx.doi.org/10.1016/j.fusengdes.2014.01.062] 907?12 ), the Tore Supra facility and team expertise (Dumont et al 2014 Plasma Phys. Control. Fusion 56 [http://dx.doi.org/10.1088/0741-3335/56/7/075020] 075020 ) is used to pave the way towards ITER divertor procurement and operation. It consists in implementing a divertor configuration and installing ITER-like actively cooled tungsten monoblocks in the Tore Supra tokamak, taking full benefit of its unique long-pulse capability. WEST is a user facility platform, open to all ITER partners. This paper describes the physics basis of WEST: the estimated heat flux on the divertor target, the planned heating schemes, the expected behaviour of the L?H threshold and of the pedestal and the potential W sources. A series of operating scenarios has been modelled, showing that ITER-relevant heat fluxes on the divertor can be achieved in WEST long pulse H-mode plasmas.

The longstanding expertise of the Tore Supra team in long pulse heating and current drive with radiofrequency (RF) systems will now be exploited in the WEST device (tungsten-W Environment in Steady-state Tokamak) [1]. WEST will allow an integrated long pulse tokamak programme for testing W-divertor components at ITER-relevant heat flux (10-20 MW/m2), while treating crucial aspects for ITER-operation, such as avoidance of W-accumulation in long discharges, monitoring and control of heat fluxes on the metallic plasma facing components (PFCs) and coupling of RF waves in H-mode plasmas. Scenario modelling using the METIS-code shows that ITER-relevant heat fluxes are compatible with the sustainment of long pulse H-mode discharges, at high power (up to 15 MW / 30 s at IP = 0.8 MA) or high fluence (up to 10 MW / 1000 s at IP = 0.6 MA) [2], all based on RF heating and current drive using Ion Cyclotron Resonance Heating (ICRH) and Lower Hybrid Current Drive (LHCD). This paper gives a description of the ICRH and LH...

International audience; The design of the WEST (Tungsten-W Environment in Steady-state Tokamak) Ion cyclotron resonance heating antennas is based on a previously tested conjugate-T Resonant Double Loops prototype equipped with internal vacuum matching capacitors. The design and construction of three new WEST ICRH antennas are being carried out in close collaboration with ASIPP, within the framework of the Associated Laboratory in the fusion field between IRFM and ASIPP. The coupling performance to the plasma and the load-tolerance have been improved, while adding Continuous Wave operation capability by introducing water cooling in the entire antenna. On the generator side, the operation class of the high power tetrodes is changed from AB to B in order to allow high power operation (up to 3 MW per antenna) under higher VSWR (up to 2:1). Reliability of the generators is also improved by increasing the cavity breakdown voltage. The control and data acquisition system is also upgraded in order to resolve and react on fast events, such as ELMs. A new optical arc detection system comes in reinforcement of the V r /V f and SHAD systems.

The tokamak Tore Supra (TS) has been operated since 1988 in Cadarache Centre. With the Composants Internes et Limiteur project, it will be the only machine to be able to exploit the Advanced Tokamak Regimes in steady state operation with three heating systems ICRH, LH, ECRH. The injected RF power in the Torus is a very important parameter of the total energy balance of TS, for this reason we need the best possible knowledge of these powers. The analysis and the explanation of the different RF techniques at low power which have been used in TS on the three RF systems will be described. Principles of the different measurement systems are given. The calibration methods of these systems are discussed and the present status of the calibration uncertainties of the RF measurements is analysed. Finally, when the generators are connected to a dummy load, RF measurements are compared with the calorimetric power measurements. This study allows preparation of improvements of these systems which will be used in these next steps: Composants pour l'Injection de Matiere et d'Energie Stationnaire project (B. Beaumont, Tore Supra steady state power and particles injection: the CIMES project, in: Proceedings of the 21st SOFT, Madrid, 2000).

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...

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.

A prototype of an ICRF antenna based on the load‐resilient electrical layout foreseen for ITER has been built at CEA‐Cadarache [1]. It consists of a two resonant double loop (RDL) based on the conjugate‐T concept proposed for the ITER ICRF array arranged toroidally in a close package. The antenna conditioning and commissioning have been successfully performed to validate the load resilience and the heating capability at the end of 2007. On Tore Supra despite unexciting ELMs, the load resilience studies have been conducted in L‐mode by mean of slow variations of the plasma position and fast transient variations with both pellet injection and Supersonic Molecular Beam Injection (SMBI). The capability to maintain the RF power on the ITER‐like antenna has been done successfully contrary to the classical ICRF antenna during the SMBI shots, in which the corresponding transient coupling resistance increased by a factor 3–4. In term of coupled power, a maximum power density of 6.6 MW/m2 is reached.

Since the release of the ITER ICRH system reference design report [ITER Final Design Report: DDD 5.1 –Ion Cyclotron and Current Drive System, July 2001], further design studies have been conducted. If the base of the reference design [Final Report on EFDA contract 04/1129, ITER ICRF antenna and Matching system design (Internalmatching), April 2005] is kept unchanged, several significant modifications have been proposed for a better efficiency and reliability. The increase of the poloidal order of the array and strong modifications of the matching system concept are the main changes. Technical aspects insufficiently covered in previous studies are also now worked out in detail, like the integration on a mid-plane port satisfying the constraints of the ITER environment.

The paper gives a picture of the present status and understanding of technology and physics of Lower Hybrid Current Drive for long pulse operation in tokamaks, including the development of continuous wave (CW) high power klystrons, and its evolutions towards ITER. 3.7 GH / 700 kW CW klystrons produced in series by Thales Electron Devices are now in operation on Tore Supra. First series of eight klystrons delivered more than 4 MW to sustain non-inductive plasmas during 50 s. Moreover, a prototype of 500 kW CW klystron operating at 5 GHz developed for KSTAR by Toshiba Electron Tubes and Devices, and foreseen for ITER, is able to produce RF output powers of 300 kW / 800 s and 450 kW / 20 s on matched load. The situation on wave coupling and antennas is reported, with the latest Tore Supra results of the new CW Passive-Active Multi-junction (PAM) launcher: the antenna concept foreseen for ITER. First experiments with the PAM antenna in Tore Supra have provided extremely encouraging results in terms of power handling and coupling. Relevant ITER power density of ∼25 MW/m 2 (2.7 MW of power injected into the plasma) has been maintained over ∼80 s. In addition, LH power of 2.7 MW has been coupled at a plasma-antenna distance of 10 cm.

A new ITER-relevant lower hybrid current drive (LHCD) launcher, based on the passive-active-multijunction (PAM) concept, was brought into operation on the Tore Supra tokamak in autumn 2009. The PAM launcher concept was designed in view of ITER to allow efficient cooling of the waveguides, as required for long pulse operation. In addition, it offers low power reflection close to the cut-off density, which is very attractive for ITER, where the large distance between the plasma and the wall may bring the density in front of the launcher to low values. The first experimental campaign on Tore Supra has shown extremely encouraging results in terms of reflected power level and power handling. Power reflection coefficient −2, i.e. its design value at f = 3.7 GHz. In addition, 2.7 MW has been coupled at a plasma–launcher distance of 10 cm, with a power reflection coefficient