In a previous paper, a Self-Organizing Map had proven to be able to identify the regions of the plasma operative space characterizing the pre-disruptive phase at JET without relying on any a priori information. One of the strengths of this disruption predictor lies in its inherent self-organization capability. The Self-Organizing Map discovers non-trivial relationships and captures the complicated interplay of device diagnostics on the internal plasma states directly from the experimental data. Moreover, the provided model allows the visualization of high-dimensional plasma parameters and facilitates easy interrogation of the model to understand the reasons behind its correlations. In this paper, an additional step is taken towards the interpretability of models for predicting disruptions by training a Decision Tree to classify the plasma states according to the interpretation provided by the Self-Organizing Map (stable or at high risk of disruptions). The Decision tree provides a set of rules which describe the transition of the plasma towards the pre-disruptive phase as visualized in the Self-Organizing Map. The obtained rules for the database explored in the study identify four regions in the map, two of which are at risk of disruption. These regions correspond to partitions of a 3D space based on the peaking factors of the core and divertor radiation, as well as the Locked Mode. The agreement between the Self-Organizing Map answers and the rules supplied by the Decision Tree is confirmed by the comparison of the performance exhibited by the two models in the prediction of disruptions.

Commonwealth Fusion Systems (CFS) plans to operate a tokamak power plant called ARC in the early 2030s. Tokamak plasmas have stability limits that, if crossed, lead to a rapid termination of the plasma, referred to as a disruption. Disruptions pose a melt risk to the first wall resulting from thermal and non-thermal particle heat fluxes, and an electromagnetic loading risk on all metal components within the equilibrium coils. A comprehensive set of models is used herein to provide an assessment of both mitigated and unmitigated ARC disruption loads. A preliminary massive gas injection system is baselined and a runaway electron mitigation coil option is proposed to close possible gaps in the baseline. It is predicted that all ARC disruption loads are within a factor of 2 of the disruption loads in SPARC, a tokamak presently under construction by CFS, and therefore SPARC provides an opportunity to calibrate models, test solutions and inform the design of ARC. The goal for ARC is disruption-free operation, however, the pragmatic design target is to withstand one mitigated disruption per day, and to restart the plasma following mitigation in tens of seconds without interrupting the power output. Unmitigated disruptions must be rare, and experience with unmitigated disruption impacts in SPARC will better define what rare means. The implications of this strategy for plasma disruptivity and disruption prediction are discussed, and operating the ARC scenario on SPARC is expected to refine the ARC final design and operational plan.

nstrumented sacrificial limiter heads, both domed (proud) and flat (flush) are used in DIII-D runaway electron (RE) wall strikes to study the wall impact dynamics with high spatial and time resolution. The approximate structure of the RE wetted area and heating depth on the domed limiter heads were predicted qualitatively using orbit-tracking simulations, although a strong left–right asymmetry (about the magnetic field direction) was not captured well by the simulations. It is hypothesized that this difference is perhaps due to the local 3D magnetic field perturbation of the dome limiter head. The average kinetic energy K and pitch angle θ of REs striking the limiter head were estimated from the spatial distribution of local HXR emission and were estimated to be roughly K ≈ 4 MeV and θ ≈ 0.2 rad. These values are roughly consistent with in-plasma values estimated before the loss event, indicating that RE kinetic energy and pitch angle are not drastically altered when transporting to the wall. Large shot–shot variations (1–10 kJ) in energy deposition into the limiter head were observed and were explained by shot–shot variations in locked magneto-hydrodynamics mode toroidal phase. For the largest deposited energies (10 kJ), graphite material failure and explosive dust release was observed, and the depth of material failure at higher energy deposition was successfully reproduced using modelling of volumetric energy deposition and brittle failure. The presence of energetic (keV) level ion impact during the RE wall strike was confirmed by three different surface analysis techniques. The ratio of energetic ion to RE flux appears to be larger on flat surfaces within the RE wetted area, although the energetic ion flux and total energy flux due to energetic ions have not yet been quantified.

The runaway electron (RE) - induced damage on boron nitride (BN) tiles mounted on the inner bumpers of the WEST tokamak is modeled employing available empirical input and experimental constraints, concerning the post-mortem documentation of the damaged material topology and infra-red camera observations of the long-time decay of the surface temperature. A newly developed work-flow for the modeling of brittle failure due to RE impacts, recently validated against a controlled DIII-D experiment, is employed. Monte Carlo simulations of RE transport into BN provide volumetric heat source maps for finite-element simulations of the linear thermoelastic material response, while the brittle failure onset is predicted on the basis of the Rankine criterion. The physics of thermal stress driven failure and explosion are well captured by this model, which exhibits high sensitivity to RE impact parameters. Despite the accidental nature of the damage events, the workflow predicts failure in accordance with observations for realistic loading specifications expected in WEST disruptions.

After a 26-month vent ASDEX Upgrade (AUG) went back in operation with a newly designed upper W-divertor suitable for alternative divertor configurations (featuring in-vessel coils and cryo-pump). Parameter scans and an extensive set of measurements were obtained and their interpretation is ongoing. Prompted by the ITER wall change, dedicated experiments on non-boronized plasma startup were contrasted to that employing asymmetric and more symmetric boronizations. The asymmetric boronization proved to be as beneficial as the more symmetric one, which is in contrast to previous model calculations assuming perfect sticking of boron (measurements suggest sticking ≈ 0.3). In the startup phase also the impurity influxes at the outboard limiters were investigated contrasting the unboronized case featuring cold edges (low-Z radiation) to the boronized case, in which the lifetime of the boron layers could be estimated. Pedestal stability investigations revealed that the quasi-continuous exhaust (QCE) regime is obtained when ballooning modes are active in the vicinity of the separatrix and the global peeling-ballooning stability is high enough. Thus, at high enough shaping and high gas flux both can be achieved and QCE is a consequence. The closely related enhanced D-alpha (EDA) mode is not clearly distinguishable from QCE, e.g. the quasi-coherent mode characteristic for EDA also shows up in QCE. In QCE the impurity transport is behaving benign as could be measured for Ne with a novel analysis method making use of a comprehensive set of CXRS measurements. For high radiative fractions the regime of the X-point radiator (XPR) is accessible at AUG and the understanding of its access conditions and behaviour is further developed. Due to the localized radiative cooling at the X-point the XPR can be well diagnosed and thus controlled. For negative triangularity shapes, further experiments at increased shaping resulted in strongly heated L-mode plasmas avoiding ELMs. Two integrated modelling approaches towards ITER suggested that core W-accumulation will be no issue for ITER and that the fusion yield in ITER may be Q = 12 (i.e. ITPA20-IL scaling is too pessimistic). Further, investigations of the ITER ramp-down in AUG provide insights into maintaining position control. Various aspects of shattered pellet injection were investigated in AUG and one of the results show that with increasing Ne fraction the radiation during the current quench increases and the current decay becomes faster.

The electron current density that escapes from multi-emissive magnetized sheaths constitutes a critical quantity in modern evaluations of ITER plasma-facing component (PFCs) deformation due to macroscopic melt motion. During ITER edge-localized modes, electrons can be emitted from the PFCs through field assisted thermionic emission, secondary electron emission (SEE) and electron backscattering (EBS). In the case of space-charge limited sheaths, an analytic semi-empirical expression is available for the total escaping current density. Here, an analytic empirical model is proposed for the unexplored case of classical monotonic sheaths comprising (i) an existing semi-empirical description of prompt re-deposition in the presence of surface electric fields based on particle orbit simulations; (ii) existing analytic expressions for the SEE and EBS currents as a function of the electron temperature; (iii) a new semi-empirical relation between the total electron emission yield and surface electric field; (iv) a novel correlation for the magnitude of the surface electric field in the absence of electron emission. The new predictive model is valid for arbitrary magnetic field inclination angles and is benchmarked against systematic particle-in-cell simulations.

Runaway electron (RE) interaction with plasma-facing components (PFCs) has been documented to lead to deep volumetric melting and thermal shock driven material explosions followed by extensive wall cratering. This work reports a post-mortem FTU investigation that covers the primary localized RE-induced damage directly caused by beams striking poloidal or toroidal molybdenum (Mo)-based limiters and the subsequent secondary non-localized RE-induced damage inflicted on nearby limiter tiles by the mechanical impact of fast up to similar to 1 km/s solid debris violently ejected during the direct RE-PFC interaction. Early indications on the resilience of tin liquid limiters to RE incidence are also presented.

JET, the only tokamak capable of operating with deuterium–tritium (D–T) fuel (since TFTR was shutdown in 1999), has provided essential experimental data to support ITER and DEMO design and operation. Within the EUROfusion Tokamak Exploitation Work Package, JET completed its final campaigns (2022–2023), culminating in the third D–T campaign (DTE3). These experiments addressed key challenges in plasma scenarios, exhaust control, and tritium management under reactor-relevant conditions. Significant progress was achieved in demonstrating ITER-like integrated scenarios with impurity seeding, achieving partial divertor detachment and high confinement (H₉₈(y,2) ≈ 0.85) at 3 MA in D–T plasmas. Advanced exhaust regimes such as quasi-continuous exhaust (QCE) and X-point radiator (XPR) were successfully achieved first in D–D and then extended to D–T operation, confirming their relevance for mixed isotope operation. Operational milestones included a new world record of 69 MJ fusion energy in tritium-rich hybrid plasmas and long-pulse H-mode operation up to 60 s, contributing with unique data to the CICLOP database. Physics studies focused on peelinglimited pedestals in support of ITER and improved understanding of edge stability and impurity screening in metallic environments. Extensive usage of the shattered pellet injector (SPI) on JET provided critical information for the design of the ITER disruption mitigation system (DMS). Real-time control systems for D/T ratio control and plasma exhaust were deployed and demonstrated in D–D and D–T, while energetic particle physics investigations unfolded the role of fast ions in turbulence suppression mechanisms. Comprehensive tritium retention studies using gas balance method, post-mortem analysis, and ITER-relevant laser induced desorption spectroscopy (LIDS) diagnostics provided essential input for tritium accountancy strategies. These results are validating the ITER operational concepts, inform DEMO design, and deliver critical experience in nuclear operation and scenario integration.

Next step fusion devices will face unprecedented heat loads and particle fluence with thousands of hours of plasma exposure on plasma-facing components (PFCs) cumulated over the entire lifetime of the device. These components must guarantee an acceptable lifetime, reliable heat exhaust capabilities (10–15 MW m−2 power fluxes in steady state) and a high level of resilience after multiple thermal stresses generated by combined steady-state heat loads and transient events, such as edge localized modes (ELMs) or disruptions. An extensive tungsten (W) PFC testing work-programme has been conducted in the WEST (Tungsten Environment in Steady State Tokamak) and ASDEX Upgrade (AUG) tokamaks, taking advantage of key capabilities and strengths of the two machines. WEST is a superconducting tokamak with long pulse duration capabilities currently equipped with an ITER-grade actively cooled divertor, including shaped monoblocks with a 0.5 mm height toroidal bevel as foreseen for ITER, while AUG allows the exposure of dedicated tile-sized samples (with different geometries, gap sizes, slopes and materials) in ELMy H-mode discharges using its divertor manipulator DIM-II system. The first part of the paper reports on the 4 year long operation of the ITER-grade PFCs in WEST since the commissioning of the divertor in 2022. The second part of the paper presents dedicated experiments performed in AUG and WEST in order to study W failure modes, W melting across toroidal gaps (during sustained or transient melting) and the impact of the runaway electrons on W material. The results reported here provide new information on the W material response (e.g. heating, cracking or melting) of direct relevance to ITER.

The integrity of plasma-facing components (PFCs) in tokamaks is critically challenged by transient events such as runaway electron (RE) impacts. We report the first systematic analysis of the thermal damage to tungsten-based PFC tiles comprising the SPARC outboard off-midplane limiters that is induced by RE beams formed during vertical displacement events. Parametric scans in RE impacting characteristics as well as energy-pitch distribution functions from the Dream code are employed for calculations of the volumetric heat loads. A realistic panel design is adopted to enhance the fidelity of the thermal analysis. The PFC thermal responses are compared in terms of in-depth temperature profiles and damage characteristics, such as melt depth and vaporization losses.