Recent Publications

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.

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.

French Alternative Energies and Atomic Energy Commission, Chinese Academy of Sciences, Princeton University Plasma Physics Laboratory, IUSTI, Koninklijke Militaire School

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.

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.

CEA, LPP, ERM/KMS, Technical University of Denmark, Institut Jean Lamour

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.

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.

CEA, National Centre for Scientific Research ‘Demokritos’, VTT Technical Research Centre of Finland Ltd, Max-Planck-Institut für Plasmaphysik, Ruđer Bošković Institute

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.

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

CEA Cadarache, DIFFER—Dutch Institute for Fundamental Energy Research, Eindhoven University of Technology, ITER Organization

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.

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.

IRFM, Max-Planck-Institut für Plasmaphysik, Chinese Academy of Sciences, General Atomics, Culham Science Centre

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.

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

Jerôme Bucalossi, E. Joffrin, A. Ekedahl

CEA

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.