Nuclear fusion could offer clean, abundant energy. However, managing the power exhausted from the core fusion plasma towards the reactor wall remains a major challenge. This is compounded in emerging compact reactor designs promising more cost-effective pathways towards commercial fusion energy. Alternative Divertor Configurations (ADCs) are a potential solution. In this work, we demonstrate exhaust control in ADCs, employing a novel method to diagnose the neutral gas buffer, which shields the target. Our work on the Mega Ampere Spherical Tokamak Upgrade shows that ADCs tackle key risks and uncertainties for fusion energy. Their highly reduced sensitivity to perturbations enables active exhaust control in otherwise unfeasible situations and facilitates an increased passive absorption of transients, which would otherwise damage the divertor. We observe a strong decoupling of each divertor from other reactor regions, enabling near-independent control of the divertors and core plasma. Our work showcases the real-world benefits of ADCs for effective heat load management in fusion power reactors.

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.)

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.

Magnetic confinement fusion represents one of the most promising pathways to achieving sustainable and clean energy production. In this approach, strong magnetic fields are used to confine hot plasma within a device preventing it from coming into direct contact with the vessel walls. However, plasma-wall interactions remain an unavoidable challenge, as some heat and particles inevitably escape confinement, particularly during energetic transient events. These interactions pose a significant threat to the integrity of plasma-facing components (PFCs), which are subjected to extreme thermal and particle loads. Among the various forms of damage caused by such loads, melt damage is particularly concerning due to its potential to severely degrade the performance and longevity of PFCs. 

To address these challenges, the MEMOS-U physics model was developed to simulate macroscopic melt motion in fusion environments. MEMOS-U simplifies the computational heavy thermoelectric magnetohydrodynamic equations by employing the shallow water approximation, which reduces the dimensionality of the problem. MEMOS-U has been validated against a series of dedicated tokamak experiments, demonstrating its ability to capture the essential features of melt motion in fusion environments.

Building on the MEMOS-U model, the MEMENTO code was developed as a modern numerical implementation designed to further enhance the predictive capabilities of melt motion simulations. MEMENTO leverages the AMReX framework to create and maintain a non-uniform, adaptive grid, enabling efficient simulations of large PFCs over long time scales. The code includes solvers for heat transfer, fluid dynamics, and current propagation, all of which are fully coupled to accurately model the interplay between thermal loading, melt motion, and electromagnetic effects. 

The MEMENTO code has been validated against experimental data from dedicated controlled melting experiments carried out in the ASDEX-Upgrade and WEST tokamaks. Predictive studies with MEMENTO have provided valuable insights into the potential melt damage in future tokamaks. In summary, MEMENTO represents a significant advancement in the modeling of macroscopic melt motion in fusion environments. By implementing the MEMOS-U physics model in a new code, MEMENTO provides a reliable and computationally efficient tool able to accurately predict melt damage in future reactors for regimes that could not be probed before. 

Magnetisk inneslutningsfusion representerar en av de mest lovande vägarna för att uppnå hållbar och ren energiproduktion. I detta tillvägagångssätt används starka magnetfält för att begränsa het plasma i en anordning som förhindrar att den kommer i direkt kontakt med kärlväggarna. Emellertid förblir plasma vägginteraktioner en oundviklig utmaning, eftersom en del värme och partiklar oundvikligen undkommer instängdhet, särskilt under energetiska övergående händelser. Dessa interaktioner utgör ett betydande problem mot integriteten hos plasmavända komponenter (PFC), som utsätts för extrema värme- och partikelbelastningar. Bland de olika former av skador som orsakas av sådana belastningar är smältskador särskilt oroande på grund av dess potential att allvarligt försämra prestandan och livslängden hos PFC.

För att möta dessa utmaningar utvecklades MEMOS-U-fysikmodellen för att simulera makroskopisk smältrörelse i fusionsmiljöer. MEMOS-U förenklar de beräkningsmässiga tunga termoelektriska magnetohydrodynamiska ekvationerna genom att använda den grunt vatten approximationen, vilket minskar dimensionaliteten av problemet. MEMOS-U har validerats mot en serie dedikerade tokamak-experiment, som visar dess förmåga att fånga de väsentliga egenskaperna hos smältrörelse i fusionsmiljöer.

Med utgångspunkt i MEMOS-U-modellen utvecklades MEMENTO-koden som en modern numerisk implementering utformad för att ytterligare förbättra de förutsägande kapaciteterna hos smältrörelsesimuleringar. MEMENTO utnyttjar AMReX-ramverket för att skapa och underhålla ett oenhetligt, adaptivt rutnät, vilket möjliggör effektiva simuleringar av stora PFC:er över långa tidsskalor. Koden inkluderar lösare för värmeöverföring, strömningsdynamik och strömspropagering, som alla är helt kopplade för att exakt modellera samspelet mellan termisk belastning, smältrörelse och elektromagnetiska effekter.

MEMENTO-koden har validerats mot experimentella data från dedikerade kontrollerade smältexperiment utförda i ASDEX-Upgrade och WEST tokamaks. Prediktiva studier med MEMENTO har gett värdefulla insikter om potentiella smältskador i framtida tokamaks. Sammanfattningsvis representerar MEMENTO ett betydande framsteg i modelleringen av makroskopisk smältrörelse i fusionsmiljöer. Genom att implementera MEMOS-U fysikmodellen i en ny kod tillhandahåller MEMENTO ett tillförlitligt och beräknings-effektivt verktyg som kan förutsäga smältskador i framtida reaktorer för regimer som inte kunde sonderas tidigare.

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.

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.

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).