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.

This paper presents the current status of the WEST TWA antenna, its mock-up and a possible extrapolation to DEMO. The updated WEST TWA design has a reduced antenna length and features feeding and mechanical support from a single vessel port. A mock-up of the WEST TWA antenna was designed in 2019, manufactured during 2020 and installed in the TITAN test facility at the beginning of 2021. The results of the mock-up at low and high power, its diagnostic system and the prospects are explained. Extensions towards a TWA antenna for WEST and a possible TWA system for the future DEMO tokamak reactor are briefly discussed.

Three identical new WEST ion cyclotron resonance heating (ICRH) antennas have been designed, assembled then commissioned on plasma from 2013 to 2019. The WEST ICRH system is both load-resilient and compatible with long-pulse operations. The three antennas have been successfully operated together on plasma in 2019 and 2020, with up to 5.8 MW of coupled power. The load resilience capability has been demonstrated and the antenna feedback controls for phase and matching have been developed. The breakdown detection systems have been validated and successfully protected the antennas. The use of ICRH in combination with lower hybrid has triggered the first high confinement mode transitions identified on WEST.

Since the signature of the ITER treaty in 2006, a new research programme targeting the emergence of a new generation of neutral beam (NB) system for the future fusion reactor (DEMO Tokamak) has been underway between several laboratories in Europe. The specifications required to operate a NB system on DEMO are very demanding: the system has to provide plasma heating, current drive and plasma control at a very high level of power (up to 150 MW) and energy (1 or 2 MeV), including high performances in term of wall-plug efficiency (eta > 60%), high availability and reliability. To this aim, a novel NB concept based on the photodetachment of the energetic negative ion beam is under study. The keystone of this new concept is the achievement of a photoneutralizer where a high power photon flux (similar to 3 MW) generated within a Fabry-Perot cavity will overlap, cross and partially photodetach the intense negative ion beam accelerated at high energy (1 or 2 MeV). The aspect ratio of the beam-line (source, accelerator, etc) is specifically designed to maximize the overlap of the photon beam with the ion beam. It is shown that such a photoneutralized based NB system would have the capability to provide several tens of MW of D-0 per beam line with a wall-plug efficiency higher than 60%. A feasibility study of the concept has been launched between different laboratories to address the different physics aspects, i.e. negative ion source, plasma modelling, ion accelerator simulation, photoneutralization and high voltage holding under vacuum. The paper describes the present status of the project and the main achievements of the developments in laboratories.

The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, similar to 100MW of D degrees beam at 1MeV, including high wall-plug efficiency (eta > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D- beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R&D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.

Since the signature of the ITER treaty in 2006, a new research programme targeting the emergence of a new generation of NBI system for the future fusion reactor (DEMO) has been underway between several laboratories in Europe. The specifications required to operate a NBI system on a Fusion reactor (DEMO) are very demanding: the system has to provide plasma heating, current drive and plasma control at a very high level of power (up to 150 MW) and energy (1 or 2 MeV), including high performances in term of wall-plug efficiency (η > 60%), high availability and reliability. To this aim, a novel NBI concept based on the photo-detachment of the energetic negative ion beam is under study. The keystone of this new concept is the achievement of a photo-neutralizer where a high power photon flux (~3 MW) generated within a Fabry Perot cavity will overlap, cross and partially photo-detach the intense negative ion beam accelerated at high energy (1 or 2 MeV). The aspect ratio of the beam-line (source, accelerator, etc...) is specifically designed to maximize the overlap of the photon beam with the ion beam. It is shown that this new concept has the capability to provide several tens of MW of D 0 per beam line with a wall-plug efficiency higher than 60% leading to a significant reduction of the cost (a factor 2) of the NBI system. A feasibility study of the concept has been launched between different laboratories to address the different physics aspects, i.e., negative ion source, plasma modelling, ion accelerator simulation, photoneutralization and high voltage holding under vacuum. The paper describes the present status of the project and the main achievements of the developments in laboratories.

The main objective of the WEST project is to test the actively cooled tungsten plasma facing components under high flux conditions and to provide relevant information to ITER [1]. In this framework, two new power supplies will be installed in Tore Supra in order to feed the new upper and lower divertor coils. These new coils are installed inside the vacuum vessel to modify the magnetic configuration (X point) and to allow long pulse operation in H-mode.With a no load voltage of 405V, each power supply will be able to deliver up to 13kA during long pulses of 1000s every hour, or 20kA during short pulses of 10s every twenty minutes. The development of these power supplies is done through the frame of a partnership between CEA and SWIP (South Western Institute of Physics, China).This paper presents the validation of the power supply structure regarding both electrical rating and plasma position control. A model with coupled electrical circuit and the model for plasma position rebuilding, shows the efficiency of the plasma position control. The electrical design of power supplies including voltage filters and protective devices and the progress status of the power supplies manufacturing phase in China are also presented.

Summary form only given. Neutral beam injection (NBI) is an essential part of the initial plasma heating that is necessary for fusion reactors to enter in the burn phase. For ITER, the neutral beam will be generated by collisional detachment of negative ions produced in and extracted from an inductively coupled plasma, and accelerated to 1 MeV. The efficiency of negative ion neutralization by collisions with neutrals is modest (about 55%) and more efficient solutions are sought for future reactors (DEMO). The IRFM team a CEA Cadarache has proposed a more efficient concept of neutralization of the negative ion beam based on photodetachment 1 . This concept (called Siphore) requires that the ion source and accelerator provide a thin (~3 cm wide) ion beam sheet (30A of D - at 1 or 2 MeV) in the neutralization region.A new negative ion source (called Cybele) has been designed at IRFM in order to produce an intense ion beam sheet of negative ions1,2. This source is a magnetized plasma column (~1 m long) with a uniform axial magnetic field B. Extracting grids with a relatively thin aperture are distributed along the walls of the plasma column. In this paper we focus on the Cybele negative ion source. In a first experimental campaign at IRFM, the plasma of Cybele has been generated by heated filaments distributed along the axis of the plasma column. We present model results of the Cybele source obtained with Particle-In-Cell Monte Carlo Collision (PIC MCC) simulations compared with Langmuir probe measurements of the plasma properties. Good agreement between simulations and experiments is obtained for time averaged quantities such as plasma density and potential distributions. The simulations predict the development of rotating plasma instabilities that are consistent with the critical ionization concept3 and with the measurement of the time averaged radial electric field Er in the plasma (plasma rotation at a velocity Er/B).

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.

Following the completion in May 2011 of the shutdown for the installation of the beryllium wall and the tungsten divertor, the first set of JET campaigns have addressed the investigation of the retention properties and the development of operational scenarios with the new plasma-facing materials. The large reduction in the carbon content (more than a factor ten) led to a much lower Zeff (1.2–1.4) during L- and H-mode plasmas, and radiation during the burn-through phase of the plasma initiation with the consequence that breakdown failures are almost absent. Gas balance experiments have shown that the fuel retention rate with the new wall is substantially reduced with respect to the C wall. The re-establishment of the baseline H-mode and hybrid scenarios compatible with the new wall has required an optimization of the control of metallic impurity sources and heat loads. Stable type-I ELMy H-mode regimes with H98,y2 close to 1 and βN ∼ 1.6 have been achieved using gas injection. ELM frequency is a key factor for the control of the metallic impurity accumulation. Pedestal temperatures tend to be lower with the new wall, leading to reduced confinement, but nitrogen seeding restores high pedestal temperatures and confinement. Compared with the carbon wall, major disruptions with the new wall show a lower radiated power and a slower current quench. The higher heat loads on Be wall plasma-facing components due to lower radiation made the routine use of massive gas injection for disruption mitigation essential.

Negative ion based neutral beam injection (N-NBI) systems which produce powerful high energy hydrogen atom beams are currently used on fusion reactor devices for plasma heating and current drive. To produce sufficient negative ions (NIs), sources with a large surface area are needed for reactor relevant injectors. Plasma source non-uniformity is a recurring issue in the present N-NBI systems, it is mainly due to the transverse magnetic field (filter field) at the front of the source which generates plasma gradients. This paper presents experimental results on a large size arc driven ion source at CEA, called Cybele, the purpose of the experiment is to compare the plasma density distribution (uniformity) between two magnetic confinements, (1) the transverse (horizontal) magnetic filter which is implemented in the present negative ion (NI) sources and (2) a new concept based on the magnetic mirror principle. Experiments show that in Cybele the standard configuration with the horizontal filter field (1) leads to a significant vertical plasma drift with a measured plasma density seven times higher at the top of the source, while the source with the vertical field and mirror-like confinement (2) is vertically uniform within ±10% and can operate at very low filling pressure (0.1 Pa).

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.

Steady-state long pulse operation of tokamaks requires fully non-inductive current drive, thus an external current drive method. Lower Hybrid Current Drive is recognized as one of the most efficient technique used in the present day tokamaks. Progress of the conceptual design of the LHCD ITER relevant system is reported, as well as critical technology issues.

In order to assist the design of the future ITER Lower Hybrid launcher, coupling codes ALOHA, from CEA/IRFM, TOPLHA, from Politecnico di Torino, and GRILL3D, developed by Dr. Mikhail Irzak (A.F. Ioffe Physico-Technical Institute, St. Petersburg, Russia) and operated by ENEA Frascati, have been compared with the updated (six modules with four active waveguides per module) Passive-Active Multi-junction (PAM) Lower Hybrid antennas. Both ALOHA and GRILL3D formulate the problem in terms of rectangular waveguides modes, while TOPLHA is based on boundary-value problem with the adoption of a triangular cell-mesh to represent the relevant waveguides surfaces. Several plasma profiles, with varying edge density and density increase, have been adopted to provide a complete description of the simulated launcher in terms of reflection coefficient, computed at the beginning of each LH module, and of power spectra. Good agreement can be observed among codes for all the simulated profiles.

In the frame of the EFDA task HCD-08-03-01, a 5GHz Lower Hybrid system which should be able to deliver 20MW CW on ITER and sustain the expected high heat fluxes has been reviewed. The design and overall dimensions of the key RF elements of the launcher and its subsystem has been updated from the 2001 design in collaboration with ITER organization. Modeling of the LH wave propagation and absorption into the plasma shows that the optimal parallel index must be chosen between 1.9 and 2.0 for the ITER steady-state scenario. The present study has been made with n||=2.0 but can be adapted for n||=1.9. Individual components have been studied separately giving confidence on the global RF design of the whole antenna.

The use of rectangular oversized waveguides in the Main Transmission Lines of the Lower Hybrid Current Drive (LHCD) system of ITER, requires to investigate the problem of bends. The principal specifications that characterize the oversized bend design concern the minimization of the reflection of the fundamental mode and the maximization of its transmission, limiting at the same time its coupling to spurious modes that could propagate at the operational frequency. In this paper, the performances of bends with different geometries are compared. They are simulated using the commercial finite element software Ansoft HFSS. An innovative modified mitre-bend solution with trapezoidal-elements is proposed and analyzed.

In the framework of the Broader Approach Activities, the EU will deliver to Japan the 18 superconducting coils, which constitute the JT-60SA Toroidal field magnet. These 18 coils, manufactured by France and Italy, will be cold tested before shipping to Japan. For this purpose, the European Joint Undertaking for ITER, the Development of Fusion Energy ("Fusion for Energy", F4E) and the European Voluntary Contributors are collaborating to design and set-up a coil test facility (CTF) and to perform the acceptance test of the 18 JT-60SA Toroidal Field (TF) coils. The test facility is designed to test one coil at a time at nominal current and cryogenic temperature. The test of the first coil of each manufacturer includes a quench triggered by increasing the temperature. The project is presently in the detailed design phase. (C) 2011 Elsevier B.V. All rights reserved.

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.