Exhausting power from the hot fusion core to the plasma-facing components is one fusion energy’s biggest challenges. The MAST Upgrade tokamak uniquely integrates strong containment of neutrals within the exhaust area (divertor) with extreme divertor shaping capability. By systematically altering the divertor shape, this study shows the strongest evidence to date to our knowledge that long-legged divertors with a high magnetic field gradient (total flux expansion) deliver key power exhaust benefits without adversely impacting the hot fusion core. These benefits are already achieved with relatively modest geometry adjustments that are more feasible to integrate in reactor designs. Benefits include reduced target heat loads and improved access to, and stability of, a neutral gas buffer that ‘shields’ the target and enhances power exhaust (detachment). Analysis and model comparisons shows these benefits are obtained by combining multiple shaping aspects: long-legged divertors have expanded plasma-neutral interaction volume that drive reductions in particle and power loads, while total flux expansion enhances detachment access and stability. Containing the neutrals in the exhaust area with physical structures further augments these shaping benefits. These results demonstrate strategic variation in the divertor geometry and magnetic topology is a potential solution to one of fusion’s power exhaust challenge. (Figure presented.)

A computational fluid dynamics model of ELM-induced tungsten melt flow across a gap between misaligned plasma-facing components is validated against data from dedicated leading-edge exposures in the ASDEX Upgrade tokamak. The macroscopic behavior of the simulated flow in terms of stability and attachment to the underlying solid surface agrees with experimental observations and is consistent with simplified dimensionless criteria based on the balance between fluid inertia and surface tension. Quantitative predictions of the total mass deposited on the downstream side of the gap, along with the characteristic extent of such deposits, are also shown to match the empirical evidence. Furthermore, the accumulation of re-solidified material due to consecutive melt events is found to progressively smooth the gap edge, which promotes better overall flow attachment as well as the growth of overhangs whose dimensions can eventually exceed the gap width.

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.

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.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

The dynamics of transient liquid beryllium flows induced on the ITER first wall during the current quench stage of unmitigated vertical displacement events are modelled by means of two-dimensional Navier-Stokes simulations. The study focuses on melt that is driven to the first wall panels' chamfered edges, where free-surface instabilities are the most likely to be seeded. Beyond their impact on plasma-facing component damage, these instabilities potentially result in material ejection in the form of droplets, which may ultimately solidify into dust and accumulate in the vessel. Based on prior integrated numerical predictions of quenching magnetic equilibria, wall energy deposition and melt-related damage in a concrete worst-case disruption scenario, the simulations suggest that, although the liquid layer is significantly destabilized, only 5% of the total melt mass created on the wall surface is lost through ejection. This result can serve as a basis to refine the estimates of the real transient-induced beryllium dust inventory expected in ITER.

Post-mortem and in situ evidence is presented in favor of the generation of high-velocity solid dust during the explosion-like interaction of runaway electrons with metallic plasma-facing components in FTU. The freshly-produced solid dust is the source of secondary de-localized wall damage through high-velocity impacts that lead to the formation of craters, which have been reproduced in dedicated light gas gun impact tests. This novel mechanism, of potential importance for ITER and DEMO, is further supported by surface analysis, multiple theoretical arguments and dust dynamics modeling.