The retention of two neon isotopes, 20Ne and 22Ne, was studied by ion beam analysis (IBA) for thin-films of mixed W and B as well as for pure W and B layers grown on silicon-and tungsten-substrates by means of magnetron sputter deposition. Each isotope was implanted to a fluence of 3 × 1016 at./cm2 but at different energies (35–190 keV) to obtain deposition profiles closer to the surface and deeper into the film, depending on isotope and thin-film composition. Thermal annealing in combination with IBA was used to investigate the Ne-retention in a range of temperatures between RT and 1000 °C. Time-of-flight elastic recoil detection analysis was employed to monitor the retention and depth profiles of the Ne isotopes. Both Ne-isotopes remain at their original implantation depth, thus not indicating diffusion, intermixing or desorption for the full range of temperatures and for all studied compositions.

Boronization in tokamak devices with tungsten (W) plasma facing components (PFC) may lead to the formation of mixed layers of W and boron (B) that can affect wall retention of plasma fuel species. In this study, deuterium (D) retention was investigated in W-B thin films with different stoichiometries as well as in pure W and B, grown on silicon (Si) substrates by means of magnetron sputter deposition. After pre-characterization, the layers were implanted with 1 key D2+ ions to a fluence of 7 x 1017 D/cm2, followed by in-situ ion beam analysis. The samples were annealed to temperatures between 400-600 degrees C and in-situ ion beam analysis measurements were performed before, during and after the annealing process by simultaneous Elastic Recoil Detection Analysis and Rutherford Backscattering Spectrometry. The different B concentrations in the films led to significant differences in D retention, where higher boron concentrations resulted in higher deuterium retention immediately after implantation. After annealing, the lowest amount of retained D was observed for a W-to-B ratio of 2:1, with an areal density of 8 x 1013 D/cm2, about three times lower than for pure W. The highest retention of around 5 x 1016 D/cm2 after annealing to 600 degrees C was found for the pure B-film. Ex-situ electron microscopy techniques revealed significant morphological modifications due to implantation and/or annealing, including bubble formation (W film), W surface enrichment (B-rich film) and crack formation (W-rich film).

On the basis of several recent breakthroughs in fusion research, many activities have been launched around the world to develop fusion power plants on the fastest possible time scale. In this context, high-fidelity simulations of the plasma behavior on large supercomputers provide one of the main pathways to accelerating progress by guiding crucial design decisions. When it comes to determining the energy confinement time of a magnetic confinement fusion device, which is a key quantity of interest, gyrokinetic turbulence simulations are considered the approach of choice – but the question, whether they are really able to reliably predict the plasma behavior is still open. The present study addresses this important issue by means of careful comparisons between state-of-the-art gyrokinetic turbulence simulations with the GENE code and experimental observations in the ASDEX Upgrade tokamak for an unprecedented number of simultaneous plasma observables.

Three tungsten Langmuir probes retrieved from the JET tokamak with the ITER-Like Wall (JET-ILW) from the bulk tungsten Tile 5 have been studied. Nano-indentation, microscopy, ion beam analysis (IBA), and X-ray diffraction were used to assess changes in their mechanical properties, microstructure, and phase composition. Four regions of the probes were studied - the tip and the base, at two sides: front and back. The hardness value of one of the probes (no. 5, Stack B) in the tip area was reduced when compared to the value measured on the base section: 5.4 GPa versus 8.8 GPa, respectively. On the two other probes, the hardness was similar to that of the reference material. At the protrusion of probe 5, the recrystallized zone was observed. The IBA analysis revealed that the probes' surfaces below the tips were covered by a thin layer of deposit composed primarily of beryllium, oxygen, carbon, and hydrogen isotopes, along with smaller amounts of nickel, nitrogen, and helium at some locations. The presence of tungsten carbide W2C was revealed on the tip of probe 5, in the area where IBA measurements indicated elevated carbon content in the material, demonstrated by analysis of the XRD records.

Deuterium retention was measured in beryllium samples from the JET ITER-like wall limiter tiles that were in the JET vessel for one and three campaigns (in vessel during 2015-2016 and 2011-2016, respectively), using thermal desorption spectroscopy, ion beam analysis and secondary ion mass spectrometry. It was found that overall retention increases with time non-linearly but somewhat slower than a square root of plasma exposure time. Depth distribution of retained deuterium was observed to change with time, with near-surface content being variable and dependent on recent plasma exposure conditions, and bulk contribution progressively increasing. Desorption peaks were observed to shift to higher temperatures with time. Experimental evidence suggests that long-term deuterium accumulation in the Be limiter components in JET is diffusion-dominated, with observed changes as function of time being consistent with the correspondingly deeper diffusion due to the propagation of the diffusion front. Cleaning interventions are found to only slow down this propagation and not stop it.

The objective of thermonuclear fusion consists of producing electricity from the coalescence of light nuclei in high temperature plasmas. The most promising route to fusion envisages the confinement of such plasmas with magnetic fields, whose most studied configuration is the tokamak. Disruptions are catastrophic collapses affecting all tokamak devices and one of the main potential showstoppers on the route to a commercial reactor. In this work we report how, deploying innovative analysis methods on thousands of JET experiments covering the isotopic compositions from hydrogen to full tritium and including the major D-T campaign, the nature of the various forms of collapse is investigated in all phases of the discharges. An original approach to proximity detection has been developed, which allows determining both the probability of and the time interval remaining before an incoming disruption, with adaptive, from scratch, real time compatible techniques. The results indicate that physics based prediction and control tools can be developed, to deploy realistic strategies of disruption avoidance and prevention, meeting the requirements of the next generation of devices.

Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011–2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, “homogeneous” and “directional”, have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition.

JET, the world's largest operating tokamak with unique Be/W wall and tritium handling capability, completed a Deuterium-Tritium (D-T) campaign in 2021 (Maggi et al 29th Fusion Energy Conf.) following a decade of preparatory experiments, dedicated enhancements, technical rehearsals and training (Horton et al 2016 Fusion Eng. Des. 109-111 925). Operation with tritium raises significant technical, safety and scientific challenges not encountered in standard protium or deuterium operation. This contribution describes the tritium operational requirements, pulses and technical preparations, new operating procedures, lessons learned and details on the achieved operational availability and performance. The preparation and execution of the recent JET tritium experiments benefitted from the previous experience in 1991 (Preliminary Tritium Experiment), 1997 (DTE1 campaign) and 2003 (Trace Tritium Campaigns) and consisted of the following five phases: technical rehearsals and scenario preparation, tritium commissioning, 100% tritium campaign, D-T campaign (DTE2), tritium clean-up. Following the clean-up JET resumed normal operation and is currently undertaking a further D-T campaign (DTE3).

In tokamaks, a leading platform for fusion energy, periodic filamentary plasma eruptions known as edge-localized modes occur in plasmas with high-energy confinement and steep pressure profiles at the plasma edge. These edge-localized modes could damage the tokamak wall but can be suppressed using small three-dimensional magnetic perturbations. Here we demonstrate that these magnetic perturbations can change the magnetic topology just inside the steep gradient region of the plasma edge. We identify signatures of a magnetic island, and their observation is linked to the suppression of edge-localized modes. We compare high-resolution measurements of perturbed magnetic surfaces with predictions from ideal magnetohydrodynamic theory where the magnetic topology is preserved. Although ideal magnetohydrodynamics adequately describes the measurements in plasmas exhibiting edge-localized modes, it proves insufficient for plasmas where these modes are suppressed. Nonlinear resistive magnetohydrodynamic modelling supports this observation. Our study experimentally confirms the predicted role of magnetic islands in inhibiting the occurrence of edge-localized modes. This will be beneficial for physics-based predictions in future fusion devices to control these modes.

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.