Post-disruption runaway electron (RE) kinetic energy K and pitch angle sin ϑ are critical parameters for determining resulting first wall material damage during wall strikes, but are very challenging to measure experimentally. During the final loss instability, confined RE K and sin ϑ are reconstructed during center-post wall strikes for both high impurity (high-Z) and low impurity (low-Z) plasmas by combining soft x-ray, hard x-ray, synchrotron emission, and total radiated power measurements. Deconfined (wall impacting) RE sin ϑ is then reconstructed for these shots by using time-decay analysis of infra-red imaging. Additionally, deconfined RE K and sin ϑ are reconstructed for a low-Z downward loss shot by analyzing resulting damage to a sacrificial graphite dome limiter. The damage analysis uses multi-step modeling simulating plasma instability, RE loss orbits, energy deposition, and finally material expansion (MARS-F, KORC, GEANT-4, and finally COMSOL). Overall, mean kinetic energies are found to be in the range ⟨ K ⟩ ≈ 3 − 4 MeV for confined REs. KORC simulations indicate that the final loss instability process does not change individual RE kinetic energy K. Confined RE pitch angles are found to be fairly low initially pre-instability, ⟨ sin ϑ ⟩ ≈ 0.1 − 0.2 , but appear to increase roughly 2 × , to ⟨ sin ϑ ⟩ ≈ 0.3 − 0.4 for both confined and deconfined REs during instability onset in the low-Z case; this increase is not observed in the high-Z case.

The thermo-mechanical response of an ATJ graphite sample to controlled runaway electron (RE) dissipation, realized in DIII-D, is modelled with a novel work-flow that features the RE orbit code KORC, the Monte Carlo particle transport code Geant4 and the finite element multiphysics software COMSOL. KORC provides the RE striking positions and momenta, Geant4 calculates the volumetric energy deposition and COMSOL simulates the thermoelastic response. Brittle failure is predicted according to the maximum normal stress criterion, which is suitable for ATJ graphite owing to its linear elastic behavior up to fracture and its isotropic mechanical properties. Measurements of the conducted energy, damage topology, explosion timing and blown-off material volume, impose a number of empirical constraints that suffice to distinguish between different RE impact scenarios and to identify RE parameters which provide the best match to the observations.

To mitigate the impact of technical delays, provide a more rationalized approach to the safety demonstration and move forward as rapidly as possible to a reactor relevant materials choice, the ITER Organization embarked in 2023 on a significant re-baselining exercise. Central to this strategy is the elimination of beryllium (Be) first wall (FW) armour in favour of tungsten (W), placing plasma-wall interaction (PWI) centre stage of this new proposal. The switch to W comes with a modified Research Plan in which a first “Start of Research Operation” (SRO) campaign will use an inertially cooled, temporary FW, allowing experience to be gained with disruption mitigation without risking damage to the complex water-cooled panels to be installed for later DT operation. Conservative assessments of the W wall source, coupled with integrated modelling of W pedestal and core transport, demonstrate that the elimination of Be presents only a low risk to the achievement of the principal ITER Q = 10 DT burning plasma target. Primarily to reduce oxygen contamination in the limiter start-up phase, known to be a potential issue for current ramp-up on W surfaces, a conventional diborane-based glow discharge boronization system is included in the re-baseline. First-of-a-kind modelling of the boronization glow is used to provide the physics specification for this system. Erosion simulations accounting for the 3D wall geometry provide estimates both of the lifetime of boron (B) wall coatings and the subsequent B migration to remote areas, providing support to a simple evaluation which concludes that boronization, if it were to be used frequently, would dominate fuel retention in an all-W ITER. Boundary plasma (SOLPS-ITER) and integrated core–edge (JINTRAC) simulations, including W erosion and transport, clearly indicate the tendency for a self-regulating W sputter source in limiter configurations and highlight the importance of on-axis electron cyclotron power deposition to prevent W core accumulation in the early current ramp phase. These predicted trends are found experimentally in dedicated W limiter start-up experiments on the EAST tokamak. The SOLPS-ITER runs are used to formulate W source boundary conditions for 1.5D DINA code scenario design simulations which demonstrate that flattop durations of ∼100 s should be possible in hydrogen L-modes at nominal field and current (Ip = 15 MA, BT = 5.3 T) which are one of the principal SRO targets. Runaway electrons (RE) are considered to be a key threat to the integrity of the final, actively cooled FW panels. New simulations of RE deposition and subsequent thermal transport in W under conservative assumptions for the impact energy and spatial distribution, conclude that there is a strong argument to increase the W armour thickness in key FW areas to improve margins against cooling channel interface damage in the early DT operation phases when new RE seeds will be experienced for the first time.

An integral approach to plasma-wall interaction (PWI) modelling for DEMO is presented, which is part of the EUROfusion Theory and Advanced Simulation Coordination activities that were established to advance the understanding and predictive capabilities for the modelling of existing and future fusion devices using a modern advanced computing approach. In view of the DEMO design, the aim of PWI modelling activities is to assess safety-relevant information regarding the erosion of plasma-facing components (PFCs), including its impact on plasma contamination, dust production, fuel inventory, and material response to transient events. This is achieved using a set of powerful and validated computer codes that deal with particular PWI aspects and interact with each other by means of relevant data exchange. Steady state erosion of tungsten PFC and subsequent transport and re-deposition of eroded material are simulated with the ERO2.0 code using a DEMO plasma background produced by dedicated SOLPS-ITER simulations. Dust transport simulations in steady state plasma also rely on the respective SOLPS-ITER solutions and are performed with the MIGRAINe code. In order to improve simulations of tungsten erosion in the divertor of DEMO, relevant high density sheath models are being developed based on particle-in-cell (PIC) simulations with the state-of-the-art BIT code family. PIC codes of the SPICE code family, in turn, provide relevant information on multi-emissive sheath physics, such as semi-empirical scaling laws for field-assisted thermionic emission. These scaling laws are essential for simulations of material melting under transient heat loads that are performed with the recently developed MEMENTO code, the successor of MEMOS-U. Fuel retention simulations assess tritium retention in tungsten and structural materials, as well as fuel permeation to the coolant, accounting for neutron damage. Simulations for divertor monoblocks of different sizes are performed using the FESTIM code, while for the first wall the TESSIM code is applied. Respective code-code dependencies and interactions, as well as modelling results achieved to date are discussed in this contribution.

Owing to the high stored energy of ITER plasmas, the heat pulses due to uncontrolled Type I edge localized modes (ELMs) can be sufficient to melt the top surface of several poloidal rows of tungsten monoblocks in the divertor strike point regions. Coupled with the melt motion associated with tungsten in the strong tokamak magnetic fields, the resulting surface damage after even a comparatively small number of such repetitive transients may have a significant impact on long-term stationary power handling capability. The permissible numbers set important boundaries on operation and on the performance required from the plasma control system. Modelling is carried out with the recently updated MEMENTO melt dynamics code, which is tailored to tackle melt motion problems characterized by a vast spatio-temporal scale separation. The crucial role of coupling between surface deformation and shallow angle heat loading in aggravating melt damage is highlighted. As a consequence, the allowable operational space in terms of ELM-induced transient heat loads is history-dependent and once deformation has occurred, weaker heat loads, incapable of melting a pristine surface, can further extend the damage.

Multiphase Navier–Stokes simulations of liquid metal droplets colliding with solid plasma-facing components are carried out in conditions representative of magnetic confinement fusion devices. The flow dynamics of the spreading liquid are examined to assess the relative importance of various physical processes in the impact energy budget. Contributions from the initial droplet surface energy and the solidification-induced momentum sink are shown to be of great importance in determining the final geometry of the frozen spatter. Semi-empirical scaling laws available in the literature are adapted to provide robust predictions of the flattening ratio that can be extrapolated to general fusion-relevant impact scenarios.

In future fusion reactors, extended melt pools in combination with strong plasma-induced accelerations, suggest that the metallic melt could reach the gaps between castellated plasma-facing components, potentially accompanied by profound changes in their mechanical response. The first results of a combined experimental and modelling effort to elucidate the physics of melt transport across gaps are presented. Transient melting of specially designed tungsten samples featuring toroidal gaps has been achieved in ASDEX Upgrade providing direct evidence of gap bridging. Detailed modelling with the MEMENTO melt dynamics code is reported. Empirical evidence and simulations reveal that the presence of gaps can be safely ignored in macroscopic melt motion predictions as well as that the re-solidification limited melt spreading facilitates gap bridging and leads to poor melt attachment. The findings are discussed in the context of ITER and DEMO.

An overview of the modeling approaches, validation methods and recent main results of analysis and modeling activities related to the plasma-surface interaction (PSI) in JET-ILW experiments, including the recent H/D/T campaigns, is presented in this paper. Code applications to JET experiments improve general erosion/migration/retention prediction capabilities as well as various physics extensions, for instance a treatment of dust particles transport and a detailed description of melting and splashing of PFC induced by transient events at JET. 2D plasma edge transport codes like the SOLPS-ITER code as well as PSI codes are key to realistic description of relevant physical processes in power and particle exhaust. Validation of the PSI and edge transport models across JET experiments considering various effects (isotope effects, first wall geometry, including detailed 3D shaping of plasma-facing components, self-sputtering, thermo-forces, physical and chemically assisted physical sputtering formation of W and Be hydrides) is very important for predictive simulations of W and Be erosion and migration in ITER as well as for increasing quantitative credibility of the models. JET also presents a perfect test-bed for the investigation and modeling of melt material dynamics and its splashing and droplet ejection mechanisms. We attribute the second group of processes rather to transient events as for the steady state and, thus, treat those as independent additions outside the interplay with the first group.

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm's law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98(p,y) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios ('hybrid' and 'elevated' q-profile) were established on AUG and characterised concerning their plasma performance.