The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017-2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m(-2) were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

Tore Supra is dedicated to long pulse operation at high power, with a record in injected energy of 1 GJ (2.8 MW x 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.

It has been demonstrated that lower hybrid current drive (LHCD) systems play a crucial role for steady-state tokamak operation, owing to their high current drive (CD) efficiency and hence their capability to reduce flux consumption. This paper describes the extensive technology programmes developed for the Tore Supra (France) and the KSTAR (Korea) tokamaks in order to bring continuous wave (CW) LHCD systems into operation. The Tore Supra LHCD generator at 3.7 GHz is fully CW compatible, with RF power P-RF = 9.2 MW available at the generator to feed two actively water-cooled launchers. On Tore Supra, the most recent and novel passive active multijunction (PAM) launcher has sustained 2.7 MW(corresponding to its design value of 25 MW m(-2) at the launcher mouth) for a 78 s flat-top discharge, with low reflected power even at large plasma-launcher gaps. The fully active multijunction (FAM) launcher has reached 3.8 MW of coupled power (24 MW m(-2) at the launcher mouth) with the new TH2103C klystrons. By combining both the PAM and FAM launchers, 950 MJ of energy, using 5.2 MW of LHCD and 1 MW of ICRH (ion cyclotron resonance heating), was injected for 160 s in 2011. The 3.7 GHz CW LHCD system will be a key element within the W (for tungsten) environment in steady-state Tokamak (WEST) project, where the aim is to test ITER technologies for high heat flux components in relevant heat flux density and particle fluence conditions. On KSTAR, a 2 MW LHCD system operating at 5 GHz is under development. Recently the 5 GHz prototype klystron has reached 500 kW/600 s on a matched load, and studies are ongoing to design a PAM launcher. In addition to the studies of technology, a combination of ray-tracing and Fokker-Planck calculations have been performed to evaluate the driven current and the power deposition due to LH waves, and to optimize the N-parallel to spectrum for the future launcher design. Furthermore, an LHCD system at 5 GHz is being considered for a future upgrade of the ITER Heating and Current Drive systems, with a power capability of 20 MW coupled to the plasma using a PAM launcher. An R&D programme is being conducted at CEA/IRFM to develop a BeO vacuum window which is a safety critical component of the transmission line. In addition, a mock-up of a TE10-TE30 mode converter at 5 GHz, designed for a rectangular transmission line, has been manufactured and successfully tested on Tore Supra at low RF power.

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.

The sustainment of steady-state plasmas in tokamaks requires efficient current drive systems. Lower hybrid current drive is currently the most efficient method to generate a continuous additional off-axis toroidal plasma current and to reduce the poloidal flux consumption during the plasma current ramp-up phase. The operation of the Tore Supra ITER-like lower hybrid (LH) launcher has demonstrated the capability to couple LH power at ITER-like power densities with very low reflected power during long pulses. In addition, the installation of eight 700 kW/CW klystrons at the LH transmitter has allowed increasing the total LH power in long-pulse scenarios. However, in order to achieve pure stationary LH-sustained plasmas, some R&D is needed to increase the reliability of all the systems and codes, from radio-frequency (RF) sources to plasma scenario prediction. The CEA/IRFM is addressing some of these issues by leading a R&D programme towards an ITER LH system and by the validation of an integrated LH modelling suite of codes. In 2011, the RF design of a mode converter was validated at a low power. A 500 kW/5 s RF window is currently under manufacture and will be tested at a high power in 2012 in collaboration with the National Fusion Research Institute. All of this work aims to reduce the operational risks associated with the ITER steady-state operations.

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.

The aim of the Tore Supra tokamak is to address physics and technology issues of long pulse discharges. For this purpose, Tore Supra is equipped with two actively cooled Lower Hybrid Current Drive (LHCD) antennas (f = 3.7GHz), designed to operate in 1000s long pulses. One of these is the recently installed passive active multijunction (PAM) antenna, whose design is chosen for an LHCD system for ITER. The first experiments with the PAM antenna in Tore Supra have shown extremely encouraging results in terms of reflection coefficient behaviour and power handling. The maximum power and energy reached on the PAM, after similar to 500 pulses on plasma, was 2.7MW during 78s (exceeding 200MJ injected energy). In addition, 2.7MW has been coupled at a plasma-antenna distance of 10cm. The coupling behaviour on the PAM, characterised by the fraction of reflected power (RC), shows good agreement with the predictions from the ALOHA coupling code. Full non-inductive discharges lasting 50s have been sustained with the PAM alone, exhibiting a current drive efficiency comparable to that of the full active multijunction antennas in Tore Supra, in similar experimental conditions.

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.

New real time tools have been developed for testing new 700kW/3.7GHz/CW klystrons and for the operations on very long plasma shots. After the commissioning of the 18 series tubes on the high power test bed facility, the installation of the first 8 klystrons in the Tore Supra transmitter and the adjustment tests on load, this upgrade work has been materialized during the last 2010 campaign by a successful operation on the Full Active Multijunction (FAM) C3 antenna, with new performances: 3.5MW/40s on plasma. The RF output power control in amplitude and phase has been improved for a better control of the wave spectrum launched into the plasma. The new klystrons have no modulating anode and the high cathode voltage must be adjusted with the RF input power in order to optimize the RF output power with a minimization of the thermal power losses in the collector. A new phase correction, depending on the 3 RF output power ranges used, has been introduced. The improvements made in 2009 and 2010 on the generic phase loop and the procedures used during the real time tests of the RF transfer functions in amplitude and phase are detailed below. All RF measurements systems, RF safety systems and the RF calibration procedures have been revised in order to have the best consistency, reproducibility and with a measurement error against the calorimetry measurement lower than 10%.

In the frame of the CIMES project [1], a collaborative effort between Association Euratom-CEA and Thales Electron Devices (TED) has led to the development of a high power CW klystron TH 2103 C, working at 3.7 GHz, for plasma heating and current drive for the Tokamak Tore Supra.

A one year-long major upgrade of the 3.7 GHz Lower Hybrid Current Drive (LHCD) generator for the Tore Supra (TS) tokamak has been performed. It consisted in installing a first series of eight Thales Electron Devices (TED) 700 kW CW klystrons, new CW components and auxiliaries, and in modifying the transmitter control and protection software. Modifications and calibration of the sensors and the RF subsystems were completed as well. Finally, the RF power available in the generator has been increased by 35% and the pulse duration could reach 1000 s. A complete validation and optimization of the klystrons have been performed in 2010 on matched load before the generator could enter into operation. The eight klystrons connected with the Full Active Multijunction (FAM) antenna delivered 3.5 MW/50s in December 2010. The upgrade of the generator and the steps to validate the modifications are described.

A new concept of multijunction-type antenna has been developed, the Passive Active Multijunction, which improves the cooling of the waveguides and the damping of the neutron energy (for ITER) compared to Full Active Multijunction. Due to the complexity of the structures, prototypes of the mode converters and of the Passive-Active-Multijunction launcher were fabricated and tested, in order to validate the different manufacturing processes and the manufacturer's capability to face this challenging project. This paper describes the manufacturing process, the tests of the various prototypes and the construction of the final Passive-Active-Multijunction launcher, which entered into operation in October 2009. It has been commissioned and is fully operational on the Tore-Supra tokamak, since design objectives were reached in March 2010: 2.75 MW - 78 s, power density of 25MW/m2 in active waveguides, steady-state apparent surface temperatures <; 350°C; 10 cm long distance coupling.

A new concept of multijunction-type antenna has been developed, the Passive–Active Multijunction, which improves the cooling of the waveguides and the damping of the neutron energy (for ITER) compared to Full Active Multijunction. Due to the complexity of the structures, prototypes of the mode converters and of the Passive–Active-Multijunction launcher were fabricated and tested, in order to validate the different manufacturing processes and the manufacturer's capability to face this challenging project. This paper describes the manufacturing process, the tests of the various prototypes and the construction of the final Passive–Active-Multijunction launcher, which entered into operation in October 2009. It has been commissioned and is fully operational on the Tore-Supra tokamak, since design objectives were reached in March 2010: 2.75MW – 78s, power density of 25MW/m2 in active waveguides, steady-state apparent surface temperatures <350°C; 10cm long distance coupling.

A new concept of lower hybrid antenna for current drive has been proposed for ITER (Bibet et al 1995 Nucl. Fusion 35 1213–23): the passive active multijunction (PAM) antenna that relies on a periodic combination of active and passive waveguides. An actively cooled PAM antenna at 3.7 GHz has recently been installed on the tokamak Tore Supra. This paper summarizes the comprehensive experimental characterization of the coupling properties of the PAM antenna to the Tore Supra plasmas. In this paper, the electromagnetic properties of the antenna are measured at a reduced power (<1 MW) to allow a systematic comparison with linear wave coupling theory and the associated modelling based on the linear ALOHA code. In a wide range of edge electron densities at the antenna aperture (spanning a factor 20 from 0.5 × n c to 10 × n c where n c is the slow wave density cut-off, n c = 1.7 × 1017 m−3 at 3.7 GHz) and antenna phasing, the ALOHA simulations reproduce the experimental results observed on Tore Supra. In addition, reduced power reflection coefficients (<5%) are measured at a low edge density, close to n c, i.e. in the range 0.5–3 × n c. Measurement and analysis with ALOHA of the antenna–plasma scattering matrices provide explanation of the good coupling properties of the PAM antenna close to n c by highlighting the crucial role of the slow wave intercoupling between active and passive waveguides through the plasma edge. This detailed validation of the coupling modelling is an important step towards the validation of the PAM concept in view of further optimizing the electromagnetic properties of the future ITER antenna.

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% is obtained at low density in front of the launcher, i.e. close to the cut-off density, and very good agreement between the experimental results and the coupling code predictions is obtained. Long pulse operation at ITER-relevant power density has been demonstrated. The maximum power and energy reached so far is 2.7 MW during 78 s, corresponding to a power density of 25 MW m−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 <2%. Finally, full non-inductive discharges have been sustained for 50 s with the PAM.

The ITER-relevant Lower Hybrid Current Drive (LHCD ) launcher, based on the Passive Active Multijunction (PAM) concept, was brought into operat ion on the Tore Supra tokamak in autumn 2009. The PAM launcher concept was designed in view of ITER to all w efficient cooling of the waveguides, as requir ed for long pulse operation. In addition, it offers low power reflection close to the cut-off density, which s an attractive feature for ITER, where the large distanc e between the plasma and the wall may bring the densi ty i front of the launcher to low values. The experiments with the PAM on Tore Supra have shown extremely encouraging results in terms of reflected power level and power handling. Power reflection coefficient < 2% is obtained at low density in front of the launcher, i. e. close to the cut-off density, and very good agreem ent between the experimental results and the coupling cod e prediction is obtained. Long pulse operation at ITE Rrelevant power density has been demonstrated. The ma ximum power and energy reached so far is 2.7MW during 78s, corresponding to a power density of 25MW/ m, i.e. its design value at f = 3.7GHz. In addition, 2.7MW has been coupled at a plasma-launcher distance o f 10cm, with a power reflection coefficient < 2%. Finally, full non-inductive discharges of 50s have bee n sustained with the PAM.

P.K. Sharma , M. Goniche, A. Ekedahl, V. Basiuk, J. Decker, D. Mazon, Y. Peysson, J. Achard, C. Balorin, G. Berger-By, S. Brémond, E. Co rbel, X. Courtois, E. Delmas , L. Delpech, D. Douai, C. Goletto, D. Guilhem, J.P. Gun n, P. Hertout, J. Hillairet, G.T. Hoang, F. Imbeaux, X. Litaudon, R. Magne, P. Mollard, P. More au, T. Oosako, S. Poli, M. Preynas, M. Prou, F. Saint Laurent, F. Samaille, B. Saoutic, J. Belo, C. Castaldo , S. Ceccuzzi , R. Cesario, F. Mirizzi, Y. Baranov, K.K. Kirov, J. Mailloux, V. Petrzilka, Y.S. Bae, J. Kim, S. Lee, X. Bai, X. Ding CEA, IRFM, 13108 Saint Paul-lez-Durance, France. 1 Permanent address: Institute for Plasma Research, Bhat, Gandhinagar, Gujarat, India. 2 Present address: ITER Organization, 13067 Saint Pa ul-lez-Durance, France. 3 Associaçao Euratom-IST, Centro de Fusao Nuclear 10 49-001 Lisboa, Portugal. 4 Associazione Euratom-ENEA sulla Fusione, CR Frasca ti, Roma, Italy. 5 Euratom/CCFE Fusion Association, Culham Science Ce ntre, Abingdon, OX14 3DB, UK. 6 Association Euratom-IPP.CR, Za Slovankou 3, 182 21 Praha 8, Czech Republic. 7 National Fusion Research Institute, Daejeon, Korea . 8 Southwestern Institute of Physics, Chengdu, P.R. Ch ina.