Recent Publications

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.

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).

Dimensionless analysis of H-mode plasmas for energy confinement and transport on EAST tokamak

Can Su, Shouxin Wang, Haiqing Liu, Xiao Lan Zou, Shiying Su, Zhuoyang Chen, Zheng Sun, Chen Cheng, Zichao Lin, Guangle Lin, et al.

This work presents an investigation of dimensionless parameter scaling laws for H-mode energy confinement and transport on EAST tokamak, through dedicated experiments designed to isolate the individual effects of normalized plasma pressure ( β ) and collisionality ( ν * ). The β scaling experiment revealed a weak dependence of energy confinement time on β, resulting in a scaling exponent of α β ≈ 0.24±0.20. This weak positive dependence, consistent with experiment observations and linear simulations, reflects an electrostatic turbulence dominated confinement regime in EAST H-mode plasmas. In contrast, a strong, negative dependence of confinement time on ν * was obtained, following the scaling B τ E ∝ ν * -0.71±0.32, indicating confinement improvement with decreasing ν * . Local transport analysis indicates that this strong ν* dependence is primarily associated with electron heat transport, while ion heat transport exhibits a comparatively weak sensitivity. While the linear instability spectrum is dominated by ion temperature gradient (ITG) modes across the explored parameter range, ν * strongly regulates turbulence characteristics relevant to electron heat transport, leading to a pronounced ν* dependence of global confinement. These distinct behaviors of electron and ion heat transport are associated with different physical mechanisms governing the confinement scaling with dimensionless parameters.The present results suggest that the confinement scaling exponents observed on EAST are closely tied to the turbulence regimes accessed in the dedicated β and ν* scans, providing physical insight into the differences with respect to global multi-machine scaling trends.

Overview of Achievements and Outlook of the IFMIF/EVEDA Project

Kazuo Hasegawa, Atsushi Kasugai, Keitaro Kondo, Kai Masuda, Satoshi Sato, Kentaro Ochiai, Hervé Dzitko, Fabio Cismondi, Yann Carin, Dominique Gex, et al.

The Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) project have been conducted as one of the three projects (IFMIF/EVEDA, IFERC and JT60SA) within the Broader Approach (BA) agreement between EURATOM and the Japanese government since 2007. The IFMIF is intended to deliver accelerator-based deuterium-lithium (D-Li) neutrons at energies and intensities to sufficient to enable the qualification of candidate materials for future fusion energy reactors, such as DEMO. The primary objective of the IFMIF/EVEDA project is twofold: (i) to develop a detailed engineering design of the IFMIF and (ii) to validate its major components, namely the Accelerator Facility, the Lithium Target Facility and the Test Facility. During Phase I of the BA, which concluded in March 2020, the Engineering Validation Activity (EVA) for the Lithium Target Facility and the Test Facility were successfully completed through the construction and testing of prototypes. In contrast, the EVA for the Accelerator Facility, implemented through the Linear IFMIF Prototype Accelerator (LIPAc), remains on-going. The current phase (Phase II) focuses on the continued commissioning of the LIPAc and the enhancement of some sub-systems to support the development of the Fusion Neutron Source Design (FNSD). This article presents an overview of the progress achieved in the LIPAc commissioning and FNSD activities and outlines the future directions of the activities.

Collision of two radial rarefaction waves in unmagnetized ambient plasmas: Effects of the ambient plasma density

M. François, M. E. Dieckmann, L. Romagnani, X. Ribeyre, E. d'Humières

The expansion of two circular rarefaction waves in vacuum or in a thin ambient plasma is examined with particle-in-cell simulations that resolve two spatial dimensions. In the simulation with no ambient plasma, the rarefaction waves interpenetrate near the symmetry line between both rarefaction wave centers. The exponential density decrease in rarefaction waves with distance implies that the sum of their density does not lead to a density maximum near the symmetry line. The absence of a density maximum, which would yield a repelling electric potential for the inflowing rarefaction wave ions near the symmetry line, and the high interpenetration speed of the ion beams lead to ion–ion instabilities rather than shocks in the overlap layer. The simulations with ambient plasma show that the rarefaction waves pileup the ions of the ambient plasma near the symmetry line. A localized piston of hot ambient ions forms. If its density is large enough, its thermoelectric field allows reverse shocks to grow in the rarefaction waves. These reverse shocks move slowly in the simulation frame and enclose a slab of downstream plasma. A decrease in the speed of the rarefaction wave ions upstream of the shocks with time leads to their collapse.

A structural beam-based approach for preliminary structural sizing of toroidal coil system: A supplementary module to MADE algorithm

Lorenzo Giannini, Laura Pittaluga, Giuseppe Ruta

Analysis of neutron single event effect experiment on electronics during JET DTE-3 campaigns

Martin Felix Dentan, Soilihi Moindjie, Matteo Cecchetto, Jean-Luc Autran, Ruben Garcia Alia, Richard Naish, John Waterhouse, Alan R Horton, Xavier L Litaudon, Fernanda Rimini, et al.

We measured SEEs in 40 nm and 65 nm SRAMs exposed to the D–T fusion neutron environment of the JET tokamak during the DTE-3 campaign (September–October 2023), while also characterizing the neutron spectrum seen by the devices. Using consolidated JET neutron production data, the measured SEE rates show good agreement with predictions based on the measured spectra and independently determined SRAM sensitivity parameters from thermal and monoenergetic neutron tests. These results extend the validation of our spectrum-based SEE prediction methodology, previously demonstrated in a D–D fusion neutron environment in 2021, to a D–T fusion neutron field. The novelty of this work is therefore the first experimental validation of this complete prediction approach in a D–T fusion environment, using real-time neutron spectrometry and independent device sensitivity data, rather than a new SEE model. The experiments also show that a local B₄C shield reduces the bit-flip rate by absorbing thermal neutrons, and we discuss the conditions under which such shielding can improve electronics reliability in tokamak and accelerator neutron environments.

High-charge collimated and energy-selected laser-driven MeV electron beams produced by magnetic selection

I. Cohen, I. Slabu, Q. Peysson, S. Dorard, Y. Abe, J. Béard, T. Moraine, S. N. Chen, A. Chessa, K. Iida, et al.

We have developed a compact passive energy-selector for MeV-range electrons produced by irradiating solid targets by ultra-intense short-pulse lasers. The device allows for generating electron beams with a variable energy spread over a broad range of energies, from tens of keV to tens of MeV. Here we have demonstrated its use by producing electrons from solid targets in the MeV range and with a ~10% bandwidth, thereby compensating the intrinsic broadband nature of the electrons produced from such source. Coupled with a pulsed magnetic field to further compensate the intrinsic large divergence of this source, it allows to produce a highly-collimated beam of narrow-band and ultra-fast electrons, suitable for a wide range of applications, e.g. radiation therapy or time-resolved electron probing.

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.

Measurements and analysis of short-term activation of ITER samples exposed in irradiation-end during DT operations at JET

Ewa Łaszyńska, Jakub Piotr Włodarczyk, Sabahattin Akbas, Barbara Bieńkowska, Zamir Ghani, Callum Grove, Xavier L Litaudon, Lee William Packer, Chantal Shand, Rosaria Villari, et al.

Short-lived activation products in fusion materials are critical for assessing radiological safety and maintenance planning in future fusion reactors like ITER. These products influence shutdown dose rates, impact maintenance procedures, and provide essential validation for nuclear data and activation models in the short term. Understanding their formation and decay behavior under deuterium-tritium (DT) plasma conditions allows for more accurate predictions of material performance and safety constraints. This study analyzes the short-term activation of ITER materials exposed at the KN2 6U irradiation end in the JET (Joint European Torus) tokamak during the DTE2 and DTE3 campaigns. CuCrZr, IWS A286, and SS316L(N) samples were activated during DTE2, while tungsten monoblock, a key ITER divertor material, was irradiated during DTE3. Activated samples were promptly transported via a pneumatic post system for gamma spectrometry measurements using UKAEA’s 190% relative efficiency HPGe (High-Purity Germanium) detector. The efficiency calibration was performed with the MCNP 6.1 code, employing an MCNP detector model validated by a certified multi-gamma calibration source. The experimental data were compared with FISPACT-II activation calculations, enabling the determination of calculation-to-experiment (C/E) ratios. This study provides key validation data for nuclear activation models and improves the accuracy of computational predictions for material behavior in fusion reactors.

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.

Analysis of background plasma behavior under external fields in the low energy beam transport section of LIPAC

Tomonobu Itagaki, Tomoya Akagi, Kai Masuda, Nicolas Chauvin, Luca Bellan, Fabio Cismondi, Hervé Dzitko, Yann Carin

Background plasma behaviour in the low energy transport section of the LIPAc is analysed with 3-dimensional particle-in-cell simulation. The simulation showed some characteristic effect under external field elements in the LEBT: a positive biased chopper absorbed the electron plasma in the surrounding drift region. Such absorption by the chopper is limited spatially by the solenoid lenses. However, some extent of electron plasma flows through the solenoid. Analytically estimated flux of the electron leakages through solenoid is compared to the simulation results. Results of experiments still show mysterious transient motion of beam pulse shape that is not shown in the simulations, while saturated state is roughly reproduced in it. Additionally, we developed a one-dimensional electron fluid simulation to investigate effects by thermalization of the background electron plasma, but this also could not reproduce the transient motion observed in the experiment. As a secondary outcome, the fluid simulation showed qualitatively close results to that of the particle in cell simulation in saturated states with much smaller computational resource.

Microcoulomb-level electron beam and multi-Joule hard X-rays driven by a high-efficiency laser-plasma accelerator

B. Mahieu, L. Ribotte, W. Cayzac, G. Boutoux, R. Parreault, J. Gastineau, E. Lamoine, F. Audo, R. Babjak, D. Batani, et al.

We report on the production of ultrahigh-charge relativistic electron beams and the development of a laser-wakefield acceleration platform at the LMJ facility. Making use of the kilojoule-class, sub-picosecond PETAL laser pulse focused onto a supersonic helium gas jet, electron beams carrying a total charge beyond 1 $μ$C were generated, with energies up to $\sim$500 MeV. Given the ps-scale laser pulse duration, an on-target intensity approaching $10^{19}~\mathrm{W/cm^2}$, and a plasma density reaching 2% of the critical density, electron energisation arises from a combination of self-modulated laser wakefield acceleration (SMLWFA) and direct laser acceleration (DLA). The resulting electron spectrum exhibits a Maxwellian-like distribution, characteristic of this mixed SMLWFA/DLA regime. The total energy carried by the electron beam is estimated to be up to 17 J, within a sub-ps duration. A broadband Joule-level photon beam was also produced by Bremsstrahlung, demonstrating the potential for future applications. Experimental results are supported by start-to-end numerical simulations, including 3-D particle-in-cell and Monte-Carlo particle transport calculations. These findings pave the way for applications requiring high-charge electron beams, including the generation of high-power secondary radiation or particle sources. The use of these beams to probe matter in high-energy density states driven by the nanosecond-duration LMJ beams represents another promising avenue.

Progress of LHCD experiment at 4.6 GHz towards long-pulse and high-power operation on EAST

Miaohui Li, Weiwei Zhang, Junlin Chen, Chenbin Wu, Bojiang Ding, Mao Wang, Liang Liu, Lianmin Zhao, Wendong Ma, Yong Yang, et al.

Recent achievements of lower hybrid current drive (LHCD) experiment at 4.6 GHz towards long-pulse and high-power operation on EAST tokamak are reported. The duration of long-pulse plasmas has been extended to 1056 s with 1.1 MW LH power in I-mode and 1066 s with 0.92 MW in H-mode. The operational domains in plasma current and line-averaged density for fully non-inductive discharges with LH alone and with the combined LH and electron cyclotron (EC) waves are presented. The dependence of LHCD efficiency on plasma density in both L- and H-mode discharges characterized by residual loop voltage Vloop = 0 is quantified. It is found that the LHCD efficiency is improved significantly by EC heating due to the increase of electron temperature. The dominant issues in long-pulse and high-power operation are summarized and discussed, including the power coupling imbalance, the hot spot and arc events in front of the antenna, and the deteriorated plasma heating effect with high LH power. Finally, prospects with a new 4 MW LHCD system at 4.6 GHz which is under development are given.

Direct comparison of 3D non-linear JOREK simulations of shattered pellet injection with ASDEX Upgrade experiments

Weikang Tang, Matthias Hoelzl, Paul Heinrich, Di Hu, F J Artola, Pascal de Marne, Mathias Dibon, Mike G Dunne, Ondřej Ficker, Peter Halldestam, et al.

Shattered pellet injection (SPI) as primary mitigation method for major disruptions in ITER has a large parameter space available for optimization including the total amount of injected material, the size of the individual pellet fragments, the material composition, and the timing of multiple injections. This flexibility needs to be exploited to simultaneously minimize thermal heat loads, electromagnetic vessel forces, and formation of relativistic electrons and their impacts on plasma facing components. In this article, we apply 3D non-linear magnetohydrodynamic modelling to SPI experiments in the ASDEX Upgrade tokamak, going beyond our previous work [Tang et al Nucl. Fusion 65 116003 (2025)] by resolving some discrepancies between simulations and experiment and carrying out direct qualitative and quantitative comparisons to experimental measurements. The key element that enables the transition is the incorporation of the parallel heat-flux limit, which is done here in a simplified form. The work increases the confidence of reproducing key processes of disruption mitigation in direct 3D non-linear simulations in view of future predictive studies for ITER.

Activation analysis on diagnostic windows and coating materials for HCPB and WCLL blankets in EU DEMO

Sabahattin Akbas, Barbara Bieńkowska, Ewa Łaszyńska, Jakub Piotr Włodarczyk, Matthew Lukacs, Sandrine Rosanvallon, Joelle Elbez-Uzan

Plasma diagnostics and reactor control systems in future fusion power plants, such as DEMO, will rely on optical windows to monitor key plasma characteristics under intense irradiation conditions. Selecting suitable window and coating materials such as quartz, fused silica, sapphire, HfO₂, and MgF₂ is therefore critical to ensuring component reliability and safety. This study investigates these candidate materials’ activation calculations in the DEMO environment. Using the MCNP code, neutron spectra have been calculated at window-relevant locations in the outboard equatorial port and upper port limiters, considering both Helium-Cooled Pebble Bed and Water-Cooled Lithium Lead breeding blankets. Activation analyses have been subsequently performed with the FISPACT-II inventory code to evaluate the activity, decay heat, contact dose rates, and dominant contributing nuclides under the initial DEMO irradiation scenario. The results provide insights into neutron irradiation effects relevant to the activation performance of window materials and coatings, and contribute valuable input to component design and safety evaluations within the EUROfusion Safety & Environment Work Package framework.

Development of tungsten actively cooled divertor target plasma facing components for W7-X and JT-60SA fusion devices

Marianne Richou, Mehdi Firdaouss, Joris Fellinger, Thierry BAFFIE, Bernd Böswirth, Mariano Di Bartolomeo, Diogo Dias Alexio, Daniel Dickes, Daniel Dorow-Gerspach, Martin Draksler, et al.

The present study describes the ongoing developments for plasma-facing components (PFCs) intended for future fusion devices, requiring PFCs to handle high heat loads of at least 10 MW/m² in the divertor region. tungsten (W) is chosen as the armor material due to its resistance to plasma-wall interactions and high heat fluxes (HHF). Significant efforts are underway on various fusion devices (WEST, EAST, KSTAR…) to test and optimize water-cooled W-based PFCs. For W7-X, new PFCs are being developed, using W armor material placed at the divertor target region. The design aims to simplify manufacturing, inspection, and installation processes compared to the currently operated actively cooled carbon fiber composite based divertor, while meeting thermal and mechanical constraints. The developed concept and related manufacturing choices are promising since, we show that it is able to sustain the required heat loads. Also, the JT-60SA tokamak plans a transition to metallic PFCs after 2029. The W monoblock concept, successfully used in WEST, EAST and KSTAR and planned for ITER, is being adapted for JT-60SA, and is currently the reference concept. However, some advanced designs exploring cost-saving measures and enhanced heat transfer capabilities are also under development. Manufacturing processes, such as laser powder bed fusion (LPBF) and hot isostatic pressing (HIP), are optimized to ensure reliable cooling channels for the development of the proposed enhanced designs. For these developments, thermo-hydraulic analyses and HHF tests have shown promising results, with surface temperatures being in agreement with models and material temperature limits. These developments lead to show a successful use of combined advanced manufacturing processes (additive manufacturing…), design hypotheses and adapted modelling tools to propose components with relevant performances.

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.

HL-3 research towards high-performance plasma and power exhaust solution

Wulyu Zhong, Xiao Quan Ji, Wei Chen, Xingyu Bai, Alain Becoulet, Jerome B Bucalossi, Zhe Chen, Shaoyong Chen, Zhipeng Chen, Yihang Chen, et al.

The HL-3 tokamak program addresses critical challenges in developing integrated high-performance scenarios compatible with power exhaust demands for ITER and future reactors. Through systematic facility enhancements including auxiliary heating upgrades to 19.5 MW and AI-enabled control systems achieving 95.5% disruption avoidance, HL-3 finished a new round of exploration and validation for high-performance operation and power exhaust solution. Experiments at mega-ampere plasma currents demonstrated a reactor-relevant hot-ion regime with core ion temperatures exceeding 10 keV and a fusion triple product of . High normalized beta scenarios featuring internal and double transport barriers were successfully established. To address the compatibility of the plasma boundary, various small- or no-edge localized mode (ELM) regimes were achieved, including the enhanced D-alpha (EDA) H-mode, quiescent H-mode (QH-mode), and the quasi continuous exhaust (QCE) regime. Investigations into power exhaust integration highlighted the efficacy of advanced divertor configurations, specifically snowflake and tripod geometries, in significantly reducing peak heat fluxes. Active ELM control was demonstrated via resonant magnetic perturbations (RMPs), lower hybrid waves (LHW), and impurity supersonic molecular beam injection (SMBI), complemented by real-time feedback control of divertor detachment. These developments establish the fundamental physics basis and technical foundations necessary for demonstrating and extrapolating high-performance operations to ITER and next-step devices.

Evolution of SPI-induced disruptions in ASDEX Upgrade

Paul Heinrich, Gergely Papp, Stefan Jachmich, F J Artola, Matthias Bernert, Pascal de Marne, Mathias Dibon, Ralph Dux, Thomas Eberl, Ondřej Ficker, et al.

Disruptions are a major concern for future fusion reactors based on the tokamak principle. To ensure machine protection, the thermal loads and vessel forces that arise during disruptions have to be mitigated reliably. For the ITER disruption mitigation system (DMS), the shattered pellet injection (SPI) technology has been selected. It can provide a prompt delivery of the injection material into the plasma core, with the mitigation efficiency depending on fragment size and velocity. A highly flexible SPI system was built and installed at the tokamak ASDEX Upgrade (AUG) to aid the finalization process of the ITER DMS and provide crucial input for modeling. The SPI-induced disruptions in the 2022 AUG experiments follow a typical chain of events, which are discussed in this paper: The first light (FL), main fragment arrival (MFA), plasma movement event (PME), MARFE, thermal quench (TQ)/plasma current spike (IP-spike), current quench (CQ), and vertical displacement event (VDE) phase. Depending on the injection parameters, these phases may vary significantly or some might not be present at all. In this paper, we will focus on the characterization of these disruption phases and figures of merit for the mitigation efficiency, depending on the SPI configuration. With increasing amount of assimilated neon in the plasma - primarily influenced by the neon content in the pellet but also the shattering parameters - the disruptions exhibit different behaviors. This disruption evolution seems to be a continuous process, with the most prominent feature being the changing disruption time scales and plasma current time trace shape during the CQ from convex (poorly or unmitigated) → concave (well mitigated/radiation dominated). Depending on the injection, pre-TQ durations between 15 - 0.5 ms and early CQ durations (Δt CQ 100 → 80 ) between 13.3 - 8.2 ms had been achieved at AUG.