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.

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).

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.

The Joint European Torus (JET) fusion reactor was upgraded to the metallic wall configuration in 2011 which consisted of bulk beryllium (Be) tiles in the main chamber and bulk tungsten (W) and W-coated CFC tiles in the divertor (Matthews G.F. et al 2011 Phys. Scr. T148 014001). During each campaign, a series of wall damages were observed; on the upper dump plates (UDP) positioned to the top part of the vessel walls and on the inner wall—mainly affecting the inner wall guard limiters (IWGL). In both cases, it was concluded that the causes of these damages were unmitigated plasma disruptions. In the case of JET with the metallic wall configuration, most of these plasma disruptions were intentionally provoked. The overall objective was to study the behaviour of these phenomena, in order to assess their impact on the wall, improve understanding of morphological material changes, and—based on that—to develop, implement and test mitigation techniques for their prospective use on ITER. The current results bring additional information on the effects of the unmitigated plasma disruptions on the UDPs and are a significant extension of work presented in (Jepu et al 2019 Nucl. Fusion 59 086009) where the scale of the damage after three operational campaigns on the Be top limiters of JET was highlighted. In addition, new data is presented on the damaging effect that the high energetic runaway electrons had on the Be IWGL in JET.

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).

Divertor tiles after Joint European Torus-ITER like wall (JET-ILW) campaigns and dust collected after JET-C and JET-ILW operation were examined by a set of complementary techniques (full combustion and radiography) to determine the total, specific and areal tritium activities, poloidal tritium distribution in the divertor and the presence of that isotope in individual dust particles. In the divertor tiles, the majority of tritium is detected in the surface region and, the areal activities in the ILW divertor are in the 0.5-12 kBq cm-2 range. The activity in the ILW dust is associated mainly with the presence of carbon particles being a legacy from the JET-C operation. The total tritium activities show significant differences between the JET operation with ILW and the earlier phase with the carbon wall (JET-C) indicating that tritium retention has been significantly decreased in the operation with ILW.

A brief overview of ion beam analysis methods and procedures in studies of materials exposed to fusion plasmas in controlled fusion devices with magnetic confinement is presented. The role of accelerator techniques in the examination and testing of materials for fusion applications is emphasised. Quantitative results are based on robust nuclear data sets, i.e. stopping powers and reaction cross-sections. Therefore, the work has three major strands: (i) assessment of fuel inventory and modification of wall materials by erosion and deposition processes; (ii) equipment development to perform cutting-edge research; (iii) determination of nuclear data for selected ion-target combinations. Advantages and limitations of methods are addressed. A note is also given on research facilities with capabilities of handling radioactive and beryllium-contaminated materials.

This work was carried out to identify sources of errors, uncertainties and discrepancies in studies of fuel retention in wall components from the JET tokamak using methods based on thermal desorption. Parallel aims were to establish good practices in measurements and to unify procedures in data handling. A comprehensive program designed for deuterium quantification comprised the definition and preparation of two types of materials (samples of JET limiter Be tiles and deuterium-containing targets produced in the laboratory by magnetron-assisted deposition), their pre-characterization, quantitative analyses of the desorption products in three different thermal desorption spectroscopy systems and a detailed critical comparison of the results. Tritium levels were also determined by several techniques in samples from JET and in tritiated targets manufactured specifically for this research program. Facilities available for studies of Be- and tritium-contaminated materials from JET are presented. Apparatus development, future research options and challenges are discussed.

Polycrystalline (PC) and single crystal (SC) molybdenum mirrors were irradiated with 98Mo+, 1H+, 4He+, 11B+ and 184W+. Energies were chosen to impact the optically active region (up to 30 nm deep) of Mo mirrors. Some surfaces were coated by magnetron sputtering either with B or W films 4–65 nm thick. The overall objective was to simulate the neutron-induced damage and transmutation (H, He), and the impact of H, He, B, W on the optical performance of test mirrors, and on fuel retention. In parallel, a set of PC Mo mirrors irradiated with 1.6 MeV 98Mo3+ to a damage of 2 dpa and 20 dpa was installed in the JET tokamak for exposure during deuterium-tritium campaigns. Data from spectrophotometric, ion beam and microscopy techniques reveal: (i) the irradiation decreased specular reflectivity, whereby the differences between PC and SC in reflectivity are very small, (ii) He is retained in bubbles within 25–30 nm of the subsurface layer in all irradiated materials, (iii) W, either deposited or implanted, decreases reflectivity, but the strongest reflectivity degradation is caused by B deposition. Laboratory studies show the correlation of damage and H retention. Several cycles of W deposition and its removal from SC-Mo mirrors by plasma-assisted methods were also performed.