Pierre Manas

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 beta β N ∼ 0.6 − 0.9 and confinement factor H 96 L = 1 − 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 thathas 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.

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 (FEC), significant progresses have been made in tokamaks and stellarators including very recent achievement in duration and/or performance. These progresses are reviewed by analysing the experimental data provided by 10 tokamaks and two stellarators. The published database (Litaudon et al 2024 Nucl. Fusion 64 015001), which initially included data up to January 2022, has been significantly updated for the 2025 IAEA FEC 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.

Understanding ion temperature dynamics is crucial for interpreting energy transport, heating and confinement in magnetically confined plasmas. In the WEST tokamak, ion temperature is typically inferred from neutron flux measurements, which provide reliable global estimates but lack spatial resolution. To overcome this limitation, we employ the X-ray Imaging Crystal Spectrometer (XICS), which allows ion temperature measurements with both temporal and spatial resolution. This paper presents the XICS diagnostic, methods, its capabilities and limitations. As an illustration an analysis of two representative WEST discharges are presented. The first features stationary plasma parameters, highlighting the temporal and spatial evolution of ion temperature profiles along with associated uncertainties. The second involves an X-Point Radiator (XPR) regime, where a sudden increase in ion temperature and profile modification are observed. A comparison with neutron flux measurements demonstrates good agreement, confirming the reliability of XICS and its potential for future experimental campaigns.

Tungsten peaking is predicted in the core of a WEST plasma with total-f gyrokinetic simulations, including both collisional and turbulent transport. This prediction is validated with a synthetic diagnostic of the bolometry. Although nitrogen impurities are shown to reduce the neoclassical peaking of tungsten on-axis, the overall tungsten peaking increases when nitrogen impurities are present, as they reduce the turbulence screening off-axis. This finding is important for the plasma current ramp-up phase of ITER, where light impurities seeding will be desirable to achieve low temperatures at the plasma-facing components and reduce tungsten sputtering. It provides further argument for applying early ECRH heating to maintain margins on the core power balance. The neoclassical peaking factor is cross-verified between XGC and FACIT. The heat flux at separatrix and the heat load width are modeled by XGC and compared to WEST data.

In electron heated plasmas, as the power increases, it is experimentally reported that the ion temperature (T_i ) saturates while the electron temperature (T_e ) increases [Beurskens NF 2022]. As on AUG, W7X and elsewhere, T_i saturates around 1.5 keV in WEST L-mode electron heated plasmas while T_e reaches 4 keV. Simulations within the integrated model METIS have been compared against a whole WEST campaign consisting mostly of L-mode plasmas with Lower Hybrid heating ranging from 1 to 5.5 MW. In METIS, the collisional equipartition is modeled as well as the turbulent heat transport using the neural network regression of the quasilinear gyrokinetic code QuaLiKiz. The observed T_i saturation is well captured by the modeling framework. The saturation correlates with a low ratio of the energy confinement time to the volume averaged electron-ion collisional heat exchange time. It is then shown that T_i saturation in electron heated plasma is due to an equipartition time higher than the energy confinement time. In larger devices, no T_i saturation is expected nor predicted by physics based integrated modeling used in this work, thanks to equipartition times sufficiently shorter than the energy confinement time.

The observation of radiative collapses during ion cyclotron resonance heating (ICRH) operation on the full tungsten WEST tokamak constitutes a unique opportunity to get a quantitative balance of the sources and sinks in the core region of the plasma. Experimental analysis and numerical modelling evidence a significant reduction of the effective electron heat source delivered by ICRH compared with expectations on the one hand, and a complex interplay of mechanisms acting on the collisional peaking of tungsten on the other hand. Besides providing an explanation for the observed radiative collapses, this work outlines the variety of phenomenon determining the tungsten profile in ICRH operation.

Using a recently installed impurity powder dropper (IPD), boron powder (μm) 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 × 10¹⁹ 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.