Recent Publications

Singularities of the cold plasma theory: Modeling challenges for ICRF operation in low-density edge plasma

Wouter Tierens, Chris Klepper, Raymond Diab, Guillaume Urbanczyk

Sustained ICRF operation in a fusion power plant may require low edge densities to mitigate plasma-wall interactions, a regime which was recently achieved in WEST with very little impurity sputtering. Cold plasma theory, however, predicts singular radiofrequency electric fields in this regime, both at the lower hybrid resonance and along the resonance cones, raising the question of whether standard collisional cold plasma models suffice to describe low-density edge ICRF at all. Collisions in principle remove these singularities, replacing them with finite but sharply peaked fields. We derive these peak length scales analytically and confirm them with a 2D finite-element simulation using exponential mesh refinement, achieving micrometer resolution where needed. We conclude that edge collisions in cold plasma do not remove the need to resolve length scales ordinarily associated with hot-plasma and Bernstein-wave physics.

WEST advanced wall protection achievements toward long pulse operation

Raphael Mitteau, Marie-Helene Aumeunier, Leo Dubus, Jonathan Gerardin, Valentin Gorse, Erwan Grelier, Victor Moncada, Sébastien Vives, Xavier L Litaudon, Marcin Jakubowski, et al.

Long pulse operation in magnetic fusion devices requires well controlled plasma power exhaust to the divertor & wall, and avoidance of wall hot spots that could evolve in wall damage. At WEST, 10 major plasma facing components are monitored using 10 series of temperature/power indicators, based on multiple diagnostic systems, among which the infrared viewing system is especially relevant. These indicators span from the most basic ones (temperatures, power and energy from deterministic models) to advanced processes using artificial intelligence acquired through machine learning. Some advanced processes do operate in real time, and feedback on power actuators through the plasma control system, providing active control toward remaining within the safe operational domain. Other advanced processes intervene as forensic tools post discharge to identify possible dangerous situation regarding the power loading to the wall, so that the discharge run plan is adjusted to avoid running into aggravating wall events. No critical wall power event happened during the campaigns C9 to C11 (2024-2025), totalling about 13h of plasma, that would have affected the campaign plan. While it cannot be demonstrated that the active & intelligent wall protection enabled the new plasma duration record of 1337 seconds, the wall protection system as a whole plausibly contributed to obtaining these record durations by preventing wall hot spots to become critical during the campaigns.

Ripple-induced fast-ion losses modeling in the WEST tokamak

Hugo Corvoysier, J Morales, Jonathan Dufour, Guido Huijsmans, Samuele Mazzi, Remi Dumont

The confinement of fast ions is of crucial importance in present and future tokamaks. In typical ICRH-heated scenarios in WEST, fast-ion losses can represent more than 20 % of the total injected power. To investigate the origin and underlying mechanisms of these losses, we perform simulations of fast hydrogen ion trajectories in a realistic three-dimensional magnetic field. A weighting method is proposed as a post-process to study a particle distribution function with anisotropic temperatures, while preserving uniform statistics in term of energy. The influence of the initial and final particle position in phase space is studied in details. The simulation results show that ripple losses fall into two categories: deeply trapped particles moving vertically towards the baffle, and drifting bananas striking the antenna along the direction of the magnetic field. If fast particle losses to the baffle can be directly compared with infrared (IR) measurements, diagnosing particles striking the antennas is more challenging, as IR cameras detect a combination of fast-ion losses, plasma heat convection, and radiation. A net reduction of baffle strikes is observed for particles accelerated on the high field side, whereas antenna strikes, in contrast, may be enhanced. Hence, this study suggests that optimizing the ICRH absorption layer position involves a trade-off: reducing deeply trapped losses at the cost of increasing drifting banana losses.

Experimental investigation of ICRF antenna operation and plasma-wall interaction in the regime of propagating slow waves in front of the antenna

Raymond Diab, Laurent Colas, Seung Gyou Baek, Nicolas Fedorczak, Benoit Guillermin, James Paul Gunn, Julien Hillairet, Curtis A Johnson, Ernesto A Lerche, Guillaume Urbanczyk

Ion cyclotron range of frequencies (ICRF) antenna operation and plasma-wall interaction were investigated on the WEST tokamak in the regime where the density at the antenna limiters was sufficiently low for the slow wave (SW) to propagate in front of the antenna. Using a reciprocating emissive probe magnetically connected to the antenna, we measured for the first time the DC plasma potential, VDC, during a radial scan of the LH resonance layer across the antenna limiter. VDC peaks when the density at the antenna limiter edge approaches the LH resonance density, but never exceeds typical values of a few hundred volts. From the plasma-wall interaction standpoint, this regime is highly favorable: because particle fluxes are reduced while sheath potentials remain comparable to standard operating conditions at the same antenna voltage, local tungsten sources at the ICRF antenna and other outer-wall components become nearly undetectable. In general, tungsten sputtering from active WEST ICRF antennas is dominated by the particle flux rather than the sputtering yield; that is, it follows variations in the local density rather than the plasma potential. By contrast, sputtering in the divertor is primarily governed by the sputtering yield. Core impurity contamination is likewise significantly reduced when the antennas are positioned far from the separatrix, both with and without ICRF power, and the radiated power fraction decreases accordingly. Despite the modest coupled powers obtained at large antenna-plasma clearance, satisfactory ICRF heating is maintained as the density at the antenna limiter edge falls below the LH resonance, and no deleterious effects are observed in any key core plasma metrics when operating the antenna from this low-density region. Overall, the experimental results point toward minimal coupling to the SW by the fast wave antenna despite it being located in a region where the SW can propagate.

WEST Operation - Reliability and availability of a long pulse fusion tokamak

Valérie Lamaison, Cyril Brun, Elodie Corbel, Annika Ekedahl, Laurent Gargiulo, Sebastien Hacquin, Michael Houry, Lionel Meunier, Philippe Moreau, Lionel Toulouse

Since 2016, the WEST tokamak has demonstrated its capability to perform long plasma discharges approaching 1000 seconds in a fully metallic environment. It operates with a permanent magnetic field of up to 3.65T generated by 18 superconducting toroidal field coils cooled with helium at 1.8K produced by a cryogenic system. Since 2021, all plasma-facing components (PFCs), including the tungsten ITER-grade divertor, are actively cooled by pressurized water, making WEST representative of future superconducting fusion devices. Between 2022 and 2024, experimental campaigns achieved significant improvements in performance. The number of long-duration discharges (>100s) increased threefold thanks to the non-inductive current drive from the Lower Hybrid Current Drive (LHCD) system, culminating in a world record plasma duration of 22 minutes with 2.6 GJ injected energy. Total plasma time exceeded five hours per year, with over 70% of successful pulses. These results are enabled thanks to the availability of the WEST machine and all subsystems, higher than 70%. The study of downtimes recorded during the last three years experimental campaigns, shows four main elements/systems impacting WEST operation: the water and air leaks in vacuum vessel, the poloidal field system, the CODAC (Control, Data Access and Communication) system and the cryogenic system. Key lessons for future fusion devices to achieve plasma long pulses include a high availability of the machine based on a targeted maintenance plan to maximize reliability of all sub-systems and an effective responsiveness in incident diagnosis and repair, particularly in water leak detection on actively cooled Plasma-Facing Components (PFC).

Overview of plasma wall interactions in the new high particle fluence campaign of WEST operated with cold divertor plasma conditions

J. Gaspar, Y. Corre, N. Fedorczak, J.P. Gunn, P. Maget, N. Rivals, E. Tsitrone, T. Barbui, B.De Martino, C. Desgranges, et al.

WEST long-pulse achievements in support of next-step fusion devices

Remi Dumont, Theo Fonghetti, Patrick Maget, Pierre Manas, Jean-Francois Artaud, Tullio Barbui, Clarisse Bourdelle, Laurent Colas, Guido Ciraolo, Yann Corre, et al.

The WEST tokamak is equipped with a superconducting toroidal magnetic field system, a multi-megawatt radiofrequency auxiliary power system, and an actively cooled ITER-grade tungsten divertor. As such, it is well adapted to explore experimental aspects related to the long pulse operation of next-step devices. Supported by predict-first integrated modeling, bespoke scenario development has allowed zero-loop voltage pulses to be achieved. The resulting discharges, with plasma currents in the range I p ∼0.22-0.28 MA exclusively sustained by the Lower Hybrid Current Drive (LHCD) system as an auxiliary power source, have achieved durations in excess of 22 min and injected/extracted energies up to 2.61 GJ. Plasma performance is characterized by ranges of poloidal beta β p ∼1.6-2.0, normalized toroidal β N ∼0.6-0.9 and confinement factor H 96L ∼1.0-1.3. Mild MHD activity, identified as resulting from the interaction of 3/1 and 4/1 tearing modes, is occasionally present, depending on the LHCD antenna combination used. This article describes the predict-first approach that has been employed in the context of this long-pulse scenario development endeavor. The main achievements and the physics analyses performed are reviewed, including post-experiment integrated modeling aspects. Prospects for further long-pulse developments are drawn.

Tungsten limiter start-up experiments on ASDEX Upgrade and WEST in different boronization states in support of ITER

Joerg Hobirk, Richard A Pitts, Pierre Manas, Clemente Angioni, Matthias Bernert, Dominik Brida, Guido Ciraolo, Laurent Colas, Corinne Desgranges, Ralph Dux, et al.

Understanding the performance of limiter plasmas in the ITER start-up phase is important for the whole pulse and a possible challenge if performed on tungsten, as will now be the case following the switch to a W first wall in the new 2024 ITER Baseline. Experiments were performed on ASDEX Upgrade and WEST to characterise limiter plasmas using boronizations with different degree of boron surface coverage and toroidal asymmetries. Non-boronized start-up is shown to be slow and laborious, and, on both machines, was aborted in favour of a non-homogeneous boronization, performed also in support of the ITER re- baseline to study the impact of spatially non-uniform boron coatings. This allows normal start- up and short (few 100 ms, ITER will run ≈ 10s long limiter phases) limiter phases can be run without problems. Even with a full boronization, the limiter can de-condition and long limiter plasmas of several seconds suffer from high densities and radiation. Short limiter plasmas for plasma current ramp-up remain possible. The conditioning effect on limiter plasmas is documented, but also how the start-up is affected by an ageing boronization.

Investigating long-duration plasma operation with the international multi-machine CICLOP database

Xavier L Litaudon, Ernesto A Lerche, Olaf Grulke, Christopher Thomas Holcomb, Juan Huang, Marcin Jakubowski, Hyun-Seok Kim, Pierre Manas, Tomohiro Morisaki, Francesca Turco, et al.

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Since the previous 2023 IAEA Fusion Energy Conference, significant progresses have been made in tokamaks and stellarators including very recent achievement in duration and/or performance. These progresses are reviewed by analyzing the experimental data provided by 10 tokamaks and two stellarators. The published database [Litaudon X. et al 2024 Nucl. Fusion 64 015001], which initially included data up to January 2022, has been significantly updated for the 2025 IAEA Fusion Energy Conference to incorporate the latest 2023-2025 experiments (up to May 2025) including recent records performance with new entries provided by DIII-D, EAST, JET, KSTAR, WEST, and W7-X. The update dataset has been gathered and coordination have been provided by the IEA-IAEA international CICLOP group (Coordination on International Challenges on Long duration OPeration). An overview of the recent progress toward long pulse operation analysing the CICLOP database is provided in this publication.

Impact of ICRH resonant layer position on core plasma performances in WEST H-minority plasmas

Samuele Mazzi, J Morales, Ernesto Lerche, Laurent Colas, Julien Hillairet, Hugo Corvoysier, Remi Dumont, Nicolas Fedorczak, Jeronimo Garcia, Philippe Huynh, et al.

A systematic experimental study of hydrogen-minority ion cyclotron resonance heating (ICRH) has been performed in deuterium plasmas in WEST to optimize the power deposition. Dedicated scans of the radial position of the ion cyclotron resonant layer and of the minority concentration were carried out while keeping global plasma parameters and antenna coupling nearly constant. The maximum performance is obtained when the resonant layer is shifted towards the high-field side (HFS) by around 7 cm with a hydrogen minority concentration between 5 and 8%. A pronounced degradation occurs for larger inward displacements, whereas the performances are almost halved when the resonance is located in the low-field side. No clear dependence of turbulence characteristics on the resonance position is observed, indicating that the confinement variations are primarily governed by wave absorption and fast-ion confinement modifications. Infrared thermography and calorimetry measurements show that fast-ion ripple-induced losses are minimized for resonance positions beyond a certain value on the HFS, revealing a trade-off between optimal absorption and wall load mitigation. These results demonstrate that coordinated control of the resonant layer position and minority concentration allows simultaneous optimization of ICRH efficiency and fast-ion wall loads in a fully metallic reactor-relevant environment.

Integrated full pulse modeling for pellet injection in tokamaks: HPI2 model improvement and validation in WEST.

Alex Panera Alvarez, Florian Koechl, Jean-Francois Artaud, Eleonore Geulin, Bernard Pegourie, Enzo Vergnaud, Clarisse Bourdelle, Sven Wiesen

Reliable modeling and control of core density is essential for reactor-relevant magnetic confinement fusion operation, motivating cryogenic pellet injection as a primary fueling actuator and the need for predictive pellet source models in integrated modeling. Here we present an upgrade of the physics-based pellet code HPI2 in which the plasmoid release spatial step is determined self-consistently from ablation physics, dx var =v pel t exit (optionally rescaled to trade accuracy for computational cost), removing an ad-hoc discretization parameter and improving numerical robustness across injection conditions. The upgraded model is first validated in stand-alone against a high-field-side pellet-fueled, ohmic, WEST discharge (#58656) by comparing synthetic and measured interferometry line-integrated density increments, obtaining a mean error of ∽10%. We then perform full-radius, time-dependent integrated modeling validation by coupling the new HPI2 within the High Fidelity Pulse Simulator (HFPS) workflow (JINTRAC/IMAS), combining JETTO with SANCO for the impurity/radiation evolution and TGLF-SAT2 for the turbulent transport. The coupled simulations reproduce the main density rise and relaxation after pellet injection and the associated electron-temperature transient, while taking into account the strong influence of tungsten radiation in WEST, supporting the consistency of HPI2 as a predictive pellet particle source in integrated modeling frameworks. Ultimately, this validation study supports the use of pellet modeling tools in integrated modeling studies for larger devices such as ITER.

Overview of material migration and erosion experiments in the full-tungsten WEST tokamak during Phase 1 and Phase 2 operations

A. Hakola, M. Diez, N. Fedorczak, J. Gaspar, E. Tsitrone, M. Balden, Y. Corre, S. Di Genova, A. Huart, C. Martin, et al.

This paper gives an overview of erosion and migration studies of tungsten (W) in the WEST tokamak during its Phase 1 (2016–2021) and Phase 2 (from 2022) experimental campaigns with a focus on plasma-facing components (PFCs) at the divertor. In Phase 1, gross erosion of PFCs is in line with observations from other major fusion devices and attributed to low- Z impurities in the plasma. In addition, a strong asymmetry is observed between the high- (inner) and low-field (outer) side divertor targets, in favour of the inner side. Net erosion at rates of <0.5 nm s −1 is measured around the strike points while the remaining areas are dominated by net deposition. The thickest deposited layers (up to 50 μ m) with the most complex structures result from a cumulated plasma exposure of ∼7 h. The overall erosion-deposition pattern is further influenced by the strong magnetic ripple of WEST, which can result in almost an order of magnitude difference between the maxima and minima of the ripple. In Phase 2, increasing plasma fluence leads to the deposits growing to hundreds of micrometres in thickness. At the same time, erosion proceeds at a constant rate and can reach values up to 30 µ m in ∼18 h of plasma time. In the main chamber, erosion is weaker than at the divertor but especially at low densities it can result in notable transport of W into the core. In addition, upon switching on the ICRF antennas, W sputtering on the close-by limiter structures can increase by a factor of more than 10. Modelling is able to catch many of the observed phenomena in Phase 1, with the exception of the inner–outer asymmetry and the formation of the thick deposits. In contrast, the patterns during the high-fluence operations in Phase 2 require more work to be reproduced.

Overview of the WEST contributions to the new ITER baseline and fusion power plants

Jerôme Bucalossi, E. Joffrin, A. Ekedahl

In 2023–2025, WEST has developed robust long pulse discharges and reached duration exceeding 1000 s (record 1337 s), supported by a predict first approach. Reliable machine operation was key to achieving those performances. WEST is equipped with a fully actively cooled ITER grade divertor and has produced ITER relevant heat fluxes (up to 12 MW m −2 ) and particle fluence (over 10 27 D m −2 ). Cracks have been observed on the top surface of the mono-blocks, but do not hamper the divertor power exhaust capability so far. Significant erosion has been observed in the strike point area while thick deposited layers built up on the high field side of the divertor. Those deposited layers are prone to flaking, leading to impurity ingress into the plasma hampering plasma operation. A laser cleaning procedure has been developed to remove those deposited layers during shutdown. With the objective to mitigate tungsten erosion, a cold divertor X -point radiator (XPR) regime has been developed and controlled for more than 30 s. The transition into an XPR was modelled with the SOLEDGE code. WEST has also addressed urgent R&D issues in support of the tungsten first wall (FW) planned in the new ITER baseline, to inform the design of the ITER boronization system and characterize runaway impact. It was confirmed that glow discharge boronization (GDB) is required to ensure a robust plasma start-up. It was also shown that a non-uniform GDB is sufficiently efficient, and is less asymmetric than anticipated from modelling. WEST has produced dedicated run-away beam crash on the inboard limiter to validate the models of run-away damage in ITER. In the next two years, 3 MW of ECRH (electron cyclotron resonance heating) will be made available to extend the operational domain to H-mode and control tungsten transport. Beyond 2027, WEST plans to produce long pulses at higher power by upgrading its FW to improve power exhaust and operate with an optimized ITER like configuration.