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.

Future fusion reactors like ITER and DEMO will have all-tungsten (W) walls and long pulses. These features will make wall conditioning more difficult than in most of the existing devices. The W Environment Steady-state Tokamak (WEST) is one of the few long pulse (364 s) fusion devices with actively cooled W plasma-facing components in the world. WEST is a unique test bed to study impurity migration and plasma density control via reactor relevant wall conditioning techniques. The phase II of WEST operations began in 2022, after the installation of a new lower divertor, now entirely equipped with actively cooled, ITER grade, W monoblocks. After pump down, we baked WEST between 90 degrees C and 170 degrees C for similar to 2 weeks. After 82.5 h at 90 degrees C and 33 h at 170 degrees C, vacuum conditions were stable with a vessel pressure of 6x10(-5) Pa and mass spectra dominated by H-2 molecules. While at 170 degrees C, we performed similar to 40 h of D-2 glow discharge cleaning (GDC) and similar to 5 h of glow discharge boronization (GDB), using a 15 %-85 % B2D6-He mix and a total boron mass of similar to 12 g. This was the very first GDB at such high temperature for WEST. The whole wall conditioning sequence led to a similar to 10 times reduction of the H2O signal as well as to a similar to 3 times reduction of the O-2 signal, according to mass spectra. Once back to 70 degrees C, the vessel pressure was 5.5x10(-6) Pa and plasma restart was seamless with similar to 30 s cumulated over the very first 5 pulses and an Ohmic radiated power fraction F-rad = 0.6, showing successful conditioning of the new ITER grade divertor. The effect of the first, 'hot' GDB faded with a characteristic cumulative injected energy of 2.45 GJ and saturation towards F-rad similar to 0.8. After 1.4 h and 7.5 GJ of cumulative plasma time and injected energy, we carried out a second GDB, this time at 70 degrees C. This 'cold' GDB initially led to a much lower Ohmic F-rad = 0.3-0.4 but the effect lasted similar to 7 times less, with a characteristic cumulative injected energy of 0.37 GJ. At the end of the campaign, we cumulated similar to 3h and similar to 30 GJ through repetitive, minute long pulses without any boronization. Throughout this 4-weeks-long experiment, F-rad in the 4 MW heating phase evolved only marginally (from 0.5 to 0.55). This increase is mostly due to the build-up of re/co-deposited layers on both lower divertor targets.

New fueling systems have been designed for WEST, a full tungsten wall tokamak with divertor configuration. They consist of standard gas injection, supersonic molecular beam injection and pellets injection systems, each being able to fuel continuously the plasma discharge up to a duration of 1000s. A massive gas injection system is also implemented to mitigate the impact of disruptions and runaway electrons. The standard gas system consists of 4 independent lines allowing the injection of up to 4 different gas species during a plasma discharge through 11 injections areas, 4 of them being located in the divertors and in the private flux region. The pellet injection system is able to inject up to 10 pellets per second through 4 tracks: one on the low field side, and the other 3 on the high field side. The supersonic molecular beam injection system consists of 3 injectors located close to the midplane, two on the high field side and one on the low field side. On trigger, the massive gas injection system can deliver in a single puff up 800 Pa.m3 of noble gas in order to stop the plasma discharge. All the fueling systems have been characterized and can be used routinely on plasma.

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.

In carbon dominated devices, the in vessel D inventory obtained from post-mortem analyses of plasma facing component samples is generally smaller by a factor of ∼4 than that estimated from gas balance measurements. However, for an accurate evaluation of the wall inventory, gas balance measurements must be done not only during discharges and conditioning procedures, but also in between discharges and during vents. From the analysis of the whole Tore Supra database for the 2002–2007 period, we show that long term outgassing during nights, weekends and vents is essential for evaluating the deuterium release. Taking these contributions into account reconciles the gas balance and post-mortem estimations of fuel retention.

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.

This paper presents an analysis of the carbon–deuterium circulation and the resulting balance in Tore Supra over the period 2002–2007. Carbon balance combines the estimation of carbon gross erosion from spectroscopy, net erosion and deposition using confocal microscopy, lock-in thermography and SEM, and a measure of the amount of deposits collected in the vacuum chamber. Fuel retention is determined from post-mortem (PM) analyses and gas balance (GB) measurements. Special attention was paid to the deuterium outgassed during the nights and weekends of the experimental campaign (vessel under vacuum, Plasma Facing Components at 120°C) and during vents (vessel at atmospheric pressure, PFCs at room temperature). It is shown that this outgassing is the main process reconciling the PM and GB estimations of fuel retention, closing the coupled carbon–deuterium balance. In particular, it explains why the deuterium concentration in deposits decreases with increasing depth.