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.

Ni–30Cr model alloy was used to study chromia-scale formation behaviour. During its formation the mass variation signal is weak and it is important to take specific care to measure oxidation kinetics by thermogravimetric analysis. Symmetrical design allows the determination of very small mass changes in compensating buoyancy effects and limiting measurement drift. The mass variation measurement comprises noise coming from different sources that affects the electronic signal. It must be minimised to improve accuracy. This involves keeping the room-temperature constant, strictly balancing the beam and minimising buoyancy effects. By this way it was possible to acquire kinetics and to measured rate constants, kp, in a range from 3.10–5 to 3.10–10 mg2.cm−4.s−1 equivalent to 10–12 down to 10–17 cm2.s−1. Oxygen partial pressure ( P_O_2 ) was monitor and revealed an abnormal consumption of oxygen at the beginning of the thermal exposure. Experiments with an inert material showed parasitic reactions identified by mass spectrometry as combustion of impurities. As oxygen consumption is not only due to oxidation of the sample, corrosion kinetics can’t be deduced from it. Hence, to determine whether the oxygen supply from the gas is a limiting parameter, a model, which quantifies oxygen consumption by sample oxidation within the thermobalance, is proposed.

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

Reactor relevant fusion devices will use tungsten (W) for their plasma facing components (PFCs) due to its thermomechanical properties and low tritium retention. However, W introduces high-Z impurities into the plasma, degrading its performance. Different wall conditioning methods have been developed to address this issue, including coating of W PFCs with layers of low-Z material. Wall conditioning by boron (B) powder injection using an impurity powder dropper (IPD) is being studied in WEST. Two series of experiments were conducted since the installation of the new ITER grade full W divertor. During the first series in 2023 similar to 1 g of B powder was injected in total at a maximum rate of similar to 58 mg/s, both of which are three times greater than respective values in the initial WEST powder injection experiments. The second series of experiments included injection of B and BN powders for comparison of their effects on plasma performance. The presence of an instantaneous conditioning effect is suggested by visible spectroscopy measurements of low-Z impurity lines and a rollover of total radiated power past an injection rate of similar to 20 mg/s was observed. Presence of B coating layer formation is supported by the evolution of the average radiance of visible lines of B, W and oxygen (O). To understand B transport, an interpretative modeling workflow is employed, utilizing the SOLEDGE-EIRENE fluid boundary plasma code and the Dust Injection Simulator (DIS) code. Parameters like B perpendicular diffusivity and recycling coefficients are varied to match experimental results to see if the initial assumption of B sticking to the PFCs immediately after the contact with the wall is adequate for correctly modelling its distribution on the PFCs.

After 3 h of accumulated time from repeated plasma shots during C7 campaign performed in 2023, a deposited layer appeared on the ITER-grade W-monoblock of the lower divertor of WEST, mostly on the high field side. The growth of the deposit was observed during the campaign using infrared cameras, showing a large increase of the area covered by the deposit (x4) in the last two hours of cumulated plasma time. The deposit becomes problematic for the operation as it generates flakes which provoke radiative collapse when entering the plasma. A cleaning of the lower divertor is mandatory. A first cleaning was done using adhesive tape to remove all weakly adhered parts of the deposit. This method was chosen because it was easy to implement and did not generate dusts inside the tokamak. The cleaning enables partial removal of the more lightly adhered deposits but a large fraction remains stuck on the monoblock. A second cleaning was tried during 2024 operation by using the plasma as cleaner. A scenario was developed to put the inner strike line directly on the deposit to heat it and try to remove it by thermal stress. The deposit reaches temperature up to 1560°C but was not removed. The impurities generated were higher than normal operation and decreased during the cleaning session (−50% of light impurities observed at the end of the cleaning discharge session), showing an effect of cleaning by removing impurities from the deposit.

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.

Over a series of experiments performed on the WEST device we have demonstrated the ability to perform controlled impurity injections and improve overall wall conditions. Positive changes to overall machine conditions are evidenced by reduced native impurity content after injection as well as decreases in radiated power and reductions in recycling. These results are consistent with the formation of a gettering layer which provides a particle sink and a reduction of source terms. We also observe a reduction in overall Zeff at the conclusion of the powder injection period consistent with a reduced impurity burden within the plasma. Finally, we have demonstrated a minimal injection quantity required to affect a positive change in wall conditions. These results confirm that plasma assisted deposition of conditioning material through particulate injection shows substantial promise as a supplemental wall conditioning technique.

MxOy-Na2SO4 (with M = Ni, Cr, Fe, Al, Ti, Nb) and NiSO4-Na2SO4 mixtures were prepared and their reactivity was examined at 650 degrees C in air + SO2 (400 and 1000 ppm). The formation of a molten phase for NiSO4-Na2SO4 mixtures and the rapid sulfation of NiO in these conditions were demonstrated. NiO sulfation should be considered when explaining the Type II hot corrosion attacks. Cr2O3, Fe2O3 and Al2O3 also react with the gas phase showing these oxides dissolved according to an acidic-fluxing mechanism in air + SO2/SO3 (400 and 1000 ppm) environments. Al2O3 is more soluble than Cr2O3 in these conditions.

Boron (B) powder injection is a potential alternative to glow discharge boronization as a wall conditioning method for tokamaks. This technique is currently being studied in WEST experiments, during which B powder is injected by an Impurity Powder Dropper developed by PPPL. In order to interpret and analyse experimental trends, and to help develop future experiments, a modelling workflow using a boundary plasma simulation (SOLEDGE-EIRENE) and powder ablation simulation (Dust Injection Simulator) was developed and tested. The effect of adding a B neutral source to simulated deuterium + oxygen (D + O) plasmas was compared to experimental data from the WEST C5 campaign, where B powder was injected in a dedicated experiment. While the impact of B injection on radiated power Prad measurements at the upper divertor was similar, there were significant differences in measurements of Prad , outer strike point electron temperature TeOSP and O-II line intensity at the lower divertor between experiment and simulation. This discrepancy suggests that those parameters were affected by phenomena not present in the simulations, with the most likely candidates being reduced D recycling and a reduced O sourcing from the divertor.

We present new collisional-modeling calculations of tungsten plasmas at electron density of about 5 × 1013 cm−3 and for electron temperatures in the range 0.8–5 keV. These conditions are relevant to current tokamaks. In this temperature range, the modeling of the ionization balance and of spectra is a long-standing problem. Addressing this problem is also useful for plasmas that will be produced in the future tokamak ITER. In particular, we discuss the problem of ensuring completeness of the list of configurations included in the calculations. We also discuss comparisons of experimental measurements in the EUV range performed in the WEST tokamak with synthetized spectra based on the use of the unresolved transition array and of the spin–orbit split array formalisms. While this work does not rely on a precise identification of detailed lines, modeled spectra display emission features that looks quite similar to the experimental spectra. A conclusion is that standard calculation methods used for the evaluation of the configuration average collisional and radiative rates, are fine provided that a convenient list of configurations is used in the calculations.

The behavior of A718 and AD730TM was studied in air + SO2 gases with Na2SO4 deposit at 650 degrees C and 700 degrees C. The A718 alloy underwent pitting, whereas AD730TM was uniformly degraded. For a given alloy, the corrosion rate and morphology were similar at both temperatures but depended on the SO3 pressure. The sequence of oxides in the corrosion products demonstrates the presence of an aO2-gradient across the corrosion products. This indicates that the reductive and oxidative reactions were localized at the salt-gas and alloy-salt interfaces, respectively, questioning the need for diffusion of O2 and SO3 into the melt.

The VUV emission of tungsten in WEST is measured by an absolutely calibrated grazing incidence spectrometer of the Schwob-Fraenkel type, which can scan the lower half of the plasma. We have analysed the detected spectral lines in the range 120-140 angstrom and compared their behaviour with calculations and published information. We obtained an unambiguous identification of four intense and well-resolved spectral lines emitted by W42+-W45+ close to the magnetic axis in the analysed experiments. The measured spectral-line brightnesses are used to assess the Tungsten density in the emission region. In the case of a scanning line of sight, we investigate the possibility to calculate the Tungsten density profile from the angular brightness profiles. In a case of a fixed line of sight, we deduce from the measurements the core Tungsten density profile evolution during a radiative collapse.

High power experiments, up to 9.2 MW with LHCD and ICRH, have been carried out in the full tungsten tokamak WEST. Quasi non inductive discharges have been achieved allowing to extend the plasma duration to 53 s with stationary conditions in particular with respect to tungsten contamination. Transitions to H mode are observed, and H-modes lasting up to 4 s have been obtained. The increase in stored energy is weak since the power crossing the separatrix is close to the L–H threshold. Hot L mode plasmas (central temperature exceeding 3 keV) with a confinement time following the ITER L96 scaling law are routinely obtained. The weak aspect ratio dependence of this scaling law is confirmed. Tungsten accumulation is generally not an operational issue on WEST. Difficulty of burning through tungsten can prevent the discharge from accessing to a hot core plasma in the ramp-up phase, or can lead to rapid collapse of the central temperature when radiation is enhanced by a slight decrease of the temperature. Except a few pulses post-boronization, the plasma radiation is rather high ( P rad/ P tot ∼ 50%) and is dominated by tungsten. This fraction does not vary as the RF power is ramped up and is quite similar in ICRH and/or LHCD heated plasmas. An estimate of the contribution of the RF antennas to the plasma contamination in tungsten is given.

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.

Using a recently installed impurity powder dropper (IPD), boron powder (<150 μ m) was injected into lower single null (LSN) L-mode discharges in WEST. IPDs possibly enable real-time wall conditioning of the plasma-facing components and may help to facilitate H-mode access in the full-tungsten environment of WEST. The discharges in this experiment featured I p = 0.5 MA, B T = 3.7 T, q 95 = 4.3, t pulse = 12–30 s, n e,0 ∼ 4 × 10 19 m −2 , and P LHCD ∼ 4.5 MW. Estimates of the deuterium and impurity particle fluxes, derived from a combination of visible spectroscopy measurements and their corresponding S/XB coefficients, showed decreases of ∼50% in O + , N + , and C + populations during powder injection and a moderate reduction of these low- Z impurities (∼50%) and W (∼10%) in the discharges that followed powder injection. Along with the improved wall conditions, WEST discharges with B powder injection observed improved confinement, as the stored energy W MHD , neutron rate, and electron temperature T e increased significantly (10%–25% for W MHD and 60%–200% for the neutron rate) at constant input power. These increases in confinement scale up with the powder drop rate and are likely due to the suppression of ion temperature gradient (ITG) turbulence from changes in Z eff and/or modifications to the electron density profile.

Achieving a successful plasma current ramp-up in a full tungsten tokamak can be challenging due to the large core radiation (and resulting low core temperature) that can be faced with this heavy metallic impurity if its relative concentration is too high. Nitrogen injection during the plasma current ramp-up of WEST discharges greatly improves the core temperature and magnetohydrodynamic (MHD) stability. Experimental measurements and integrated simulations with the RAPTOR code, complemented with the QuaLiKiz neural network for computing turbulent transport, allow a detailed understanding of the mechanisms at play. Increased edge radiation during this transient phase is shown to improve confinement properties, driving higher core temperature and better MHD stability. This also leads to increased operation margins with respect to tungsten contamination.

Chromia forming Alloy 718 and AD730TM have been submitted to isothermal long-term air exposure up to 10 000 h at 700 degrees C. The resulting very low specific mass gain led to consider two alternatives to determine the oxidation kinetics. Measurements of the external layer thickness and of the intergranular oxide amount were carried out on SEM micrographs evidencing the intergranular oxidation contribute around 30% to the total specific mass change. AD730TM with a thicker external scale and a higher amount of intergranular oxide showed a twofold faster oxidation rate than Alloy718. The results demonstrated the need of even longer tests.

Contamination of core plasma by high-Z impurities, especially tungsten (W), is the main reason for the very high level of radiated power in WEST experiments. Intrinsic light impurities, mainly oxygen and carbon, play a dominant role in the sputtering of W on plasma facing components. In this contribution, we present a detailed analysis of WEST experiments supported by numerical modeling performed with the transport code SOLEDGE-EIRENE providing a clear picture of light impurities transport and poloidal distribution. Moreover, making use of SOLEDGE-ERO2.0 simulations, possible strategies to reduce core contamination due to W penetration are presented.

High energy ball milling of metallic powders leads to high reactivity in the milled mixture. The reaction is often faster and starts at a lower temperature. However, the mechanisms responsible for this high reactivity are not yet completely understood. The aim of this study is to evaluate one of the possible activating factors of this heightened reactivity: nano-scale mixing of the reagents. Molecular dynamics was used to analyze the role of an amorphous Ni–Al mixing layer, mimicking the powder microstructure after milling, between two Ni layers. The impact of temperature and stoichiometry was investigated in relation to the formation of the B2-NiAl intermetallic compound. At low temperatures, pre-mixing does not seem to slow down the diffusion of Ni atoms in an amorphous Al region. Homogeneous nucleation was observed in this peculiar milled microstructure. These two phenomena explain why the nano-scale mixing observed experimentally after high energy milling is indeed an activating factor in the reactivity of metallic systems such as Ni–Al.