Francisco Javier Artola

Tokamak disruptions and the resulting multi-megampere beams of relativistic runaway electrons (REs) present severe machine-protection risks for ITER. While shattered pellet injection serves as the primary line of defense for mitigation, a robust diagnostic system is mandatory to evaluate the RE current and energy. This paper provides the first assessment of the RE measurement capabilities of the radial gamma-ray spectrometers (RGRS) in ITER 15 MA deuterium–tritium (DT) scenarios. Under steady-state conditions, the combined three-detector RGRS system is sensitive to RE currents as low as IRE,min ≈ 60 kA for neon densities of ≈1020m−3, which is within the requirement to make measurements every 10 ms and for currents down to 5% of the plasma current, i.e., 750 kA for a 15 MA full-power DT scenario, as well as within the desired safety limit of 150 kA. However, when accounting for the highly dynamic, millisecond-scale growth of the RE beam in realistic DT disruption scenarios, the practical detection threshold shifts into the 500 kA–4 MA range, with most scenarios having a threshold between 1 and 2 MA. We identify the post-disruption background due to the residual activation of the first wall as the primary modeling uncertainty, which, if negligible, could lower IRE,min detected by the RGRS system by up to a factor of ≈10, as expected for early DT experiments, when tritium will be injected in ITER for the first time. We also show that a single, tangential detector with a design similar to those of the RGRS could further lower the sensitivity to IRE,min by a factor of ≈60 using the current assumption about the background.

A simplified model analysis is used for the investigation of the survival of runaway electrons following the break-up of the magnetic field lines and the reformation of the flux surfaces during the current-quench phase of tokamak disruptions in vertically unstable plasmas. It is found that even low levels of the runaway current ( ∼ a few—tens of kAs) when the plasma touches the wall might lead to runaway damage due to the strong runaway avalanche during scraping-off. Low enough runaway deconfinement times ( τ d < 1 ms ) in a sufficiently long stochastic phase ( τ / τ d > 5 ) would be required to avoid a large energy transfer to the runaway electrons and damage of the plasma facing components for low temperatures (a few eVs) of the residual ohmic plasma. The role played by the primary runaway generation mechanisms in regenerating the runaway beam after the magnetic surfaces have healed is also discussed.

Runaway electron (RE) generation represents a major concern for future tokamaks such as ITER, where high plasma temperatures and large plasma currents significantly enhance the risk of RE formation. In particular, the hot tail mechanism, triggered during the thermal quench (TQ), occurs when a rapid drop in plasma temperature prevents high-energy electrons from thermalizing, allowing them to become REs under the accelerating effect of a strong toroidal electric field. In ITER, this mechanism is expected to be the dominant primary generation before avalanching in the case of an unmitigated/poorly mitigated disruption. The recently developed hot tail tracker framework in JOREK is used here to perform the first 3D estimates of hot tail generation during disruptions of an ITER 15 MA D–T H-mode scenario mitigated by Shattered Pellet Injection (SPI). Two situations are considered, with and without pre-disruption thermal energy degradation. The non-degraded case, more pessimistic, results in strong RE generation driven by weak stochastic losses limited by the trapped population, strong helical cooling, and the migration of hot electrons from the core toward regions of high parallel electric field. In contrast, the degraded case shows successful mitigation within the assumptions of this simulation, with suppression of the core hot tail generation and the removal of the edge seed through stochastic transport. The difference between a critical and a non-critical scenario appears to be linked to localized overdensities of injected material produced by magneto-hydrodynamic (MHD) activity, which, counter-intuitively, significantly enhance local generation. This phenomenon is confirmed by a 0D two-fluid model of the bulk plasma during the TQ, coupled with a new 0D hot tail estimator based on backward tracking of a single particle in velocity space.

We assess runaway-electron (RE) generation in ITER disruptions mitigated by shattered pellet injection (SPI) using improved physics modelling in the 1D disruption simulation framework Dream . To this end, we extend Dream with four ITER-relevant physics models: (i) a reduced model for RE scrape-off associated with the vertical plasma motion, (ii) a semi-analytical plasmoid-drift model for material deposition, (iii) an adaptive hyper-resistive transport model to suppress unphysical thin-current channels during the current quench (CQ), and (iv) an updated Compton RE generation seed calculated for the new ITER tungsten first-wall design. We simulate full-current 15 MA L-mode (H26, non-nuclear) and H-mode (DTHmode24, nuclear) scenarios, and an intermediate-current 7.5 MA H-mode non-nuclear case, from realistic ITER inputs. Within the adopted reduced-model framework, complete avoidance of a multi-MA RE beam is found to require a long pre-thermal quench (TQ) duration to thermalize the hot-tail electrons, high deuterium assimilation with limited neon, and a representative seed current comparable to a single RE in ITER. As previously found with lower fidelity setups (Vallhagen et al 2024 Nucl. Fusion 64 ), these conditions are met by staggered or low-Ne injections in H26, but are typically violated in DT H-mode when nuclear seeds are present. In addition to analysing the effect of the new models, we investigate the role of the current spike associated with the TQ and the importance of radial transport of runaways in the CQ. After incorporating these additional physical effects into a comprehensive disruption model and analysing their impact, we present a representative ITER DT H-mode SPI scenario which, within the adopted reduced-model framework, yields a substantially mitigated representative RE current. This case illustrates a possible route toward reduced RE-current levels in ITER DT operation, although its quantitative tolerability remains dependent on scenario-specific impact and termination assumptions.

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 modeling to SPI experiments in the ASDEX Upgrade tokamak, going beyond our previous work (Tang et al 2025 Nucl. Fusion 65 116003) 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.

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 and 0.5 ms and early CQ durations ( Δ t CQ 100 → 80 ) between 13.3 and 8.2 ms had been achieved at AUG.

Predicting three-dimensional thermal loads during tokamak disruptions is essential for ITER yet remains weakly developed. We present a physics-based workflow that couples MHD simulations of vertical displacement events with field line tracing on a realistic 3D first wall model and a transient wall thermal response. The approach is validated against JET discharges with beryllium main chamber armour, reproducing key global dynamics, non-axisymmetric current features, and the occurrence (or absence) of melting, thereby building confidence in the methodology. We then apply the same workflow to ITER-relevant conditions with tungsten (W) armour, consistent with the new 2024 ITER re-baseline, to assess disruption heat loads and their 3D localization. The resulting analysis demonstrates the resilience of the ITER W first wall against these events and provides predictions for the energy deposition and current flow profiles. Beyond these studies, the workflow enables scenario-by-scenario estimates of disruption-induced thermal loading, allowing to assess the disruption-budget consumption for these events in future devices.

Transient phenomena and their control are of high relevance in magnetic confinement fusion plasmas to guarantee a stable and safe plasma operation. Interpretative simulations can maximize the insights gained from experiments on present machines and predictive simulations can help in the preparation of design, mitigation techniques and operational scenarios for future devices. In this article, we provide an overview of recent advances and novel scientific results obtained with the 3D non-linear hybrid fluid-kinetic code JOREK, covering physics of plasma transients from the core to the scrape-off layer (SOL) both for tokamak and stellarator devices. Substantial progress was made in the physics understanding, model validation with experiments and experiment interpretation, thus, giving confidence for predictions to devices like DTT, ITER and DEMO. The topics addressed comprise a wide range: the edge physics of new operation scenarios and edge localized mode suppression; major disruptions with a focus on runaway electrons and vertical displacement events as well as disruption mitigation by shattered pellet injection; the physics mechanisms and operational limits of the flux pumping regime for sawtooth control; MHD limits of stellarators and work towards incorporating advanced edge/SOL/exhaust dynamics; continuing improvements of the code for more efficient hybrid simulations on conventional and accelerated high performance computing architectures.

The role of plasmoid drifts towards the tokamak low field side and associated rocket motion of solid fragments in the material assimilation of shattered pellet injection (SPI) is investigated. In particular, a trace neon SPI scheme, where a small amount of neon is added to the pellet, envisaged to suppress plasmoid drifts via stronger radiation in the plasmoid, is examined combining experiments on JET and modelling with the non-linear MHD code JOREK. In the JET experiments, pure deuterium ( D 2 ) pellet or pellet with a neon atomic mixture ratio ranging from 0.016 % to 1 % is shattered and injected into 2.5 MA / 4 MJ H-mode plasmas. The experiments demonstrate evident rocket motion in D 2 SPI discharges, in contrast to trace neon SPI cases. However, this may be related to another key observation that the cooling time, defined as the duration between the SPI arrival and dip of plasma current before the current spike, drops drastically from about 70 ms with D 2 SPI to about 6 ms with 0.016 % neon SPI, which may not leave enough time for the rocket motion to occur. JOREK simulations show that plasmoid drifts play an important role in the D 2 SPI discharge and discover that the rocket motion in fact facilitates its core material assimilation since the fragments reached the high field side before moving outward. Simulations of trace neon SPI discharges demonstrate that the key limiting factor of their core material assimilation is their short cooling time rather than the level of plasmoid drifts in these discharges.

This Roadmap article addresses the critical and multifaceted challenge of plasma-facing component (PFC) damage caused by runaway electrons (REs) in tokamaks, a phenomenon that poses a significant threat to the viability and longevity of future fusion reactors such as ITER and DEMO. The dramatically increased RE production expected in future high-current tokamaks makes it very difficult to avoid or mitigate REs in such devices when a plasma discharge terminates abnormally. Preventing damage from the intense localized heat loads they can cause requires a holistic approach that considers plasma, REs and PFC damage. Despite decades of progress in understanding the physics of REs and the thermomechanical response of PFCs separately, their complex interplay remains poorly understood. This document aims to initiate a coordinated, interdisciplinary approach to bridge this gap by reviewing experimental evidence, advancing diagnostic capabilities, and improving modeling tools across different scales, dimensionalities, and fidelities. Key topics include RE beam formation and transport, damage mechanisms in both brittle and metallic PFCs, and observed effects in major facilities such as JET, DIII-D, WEST and EAST. The Roadmap emphasizes the urgency of predictive, high-fidelity modeling validated against well-diagnosed controlled experiments, particularly in the light of recent changes in ITER’s wall material strategy and the growing importance of private sector fusion initiatives. The modeling gaps include calculations of the magnetic equilibrium, RE distribution function and background plasma properties at the start of the beam termination, with a critical need to advance modeling of access conditions to benign terminations. This is to be followed by simulations of the termination itself which requires a self-consistent description of the 3D MHD fields together with the REs, which is now available thanks to recent developments. Finally, the damage assessment is enabled by modeling of the full thermo-mechanical response, encompassing the hydrodynamic and the deviatoric behavior, to RE impacts with characteristics as provided by the termination simulations. The full workflow is computationally heavy for scoping studies to assist design of future machines, calling for development of reduced or surrogate models. Each section of the Roadmap article is written to provide a concise overview of one area of this multidisciplinary subject, with an assessment of the status, a look at current and future challenges, and a brief summary. The ultimate goal of this initiative is to guide future mitigation strategies and design resilient components that can withstand the intense localized loads imposed by REs, thus ensuring the safe and sustainable operation of the next generation of fusion power plants.