Alexander Huber

The JET DTE2 experimental campaign revealed an isotopic dependence in the baseline scenario, where the target stationary conditions achieved in deuterium plasmas could not be maintained in tritium and deuterium–tritium mixtures. This loss of stationarity was associated with a gradual increase in plasma density, reduced ELM activity, and rising core radiation. This work investigates whether isotope-dependent tungsten sputtering contributes to the loss of stationarity observed in JET DTE2 baseline plasmas. The analysis focuses on the tungsten sputtering source and on tungsten transport during inter- and intra-ELM phases, as a function of the isotopic species of the main plasma. The study is carried out using the integrated core-edge-SOL framework COCONUT/JINTRAC, which was used to evolve the plasma self-consistently across the entire domain. The simulations show a clear isotopic dependence of the tungsten sputtering source, which increases with the mass of the main plasma ion and reaches the highest levels in tritium plasmas. Despite the enhanced tungsten source, most of the sputtered tungsten remains confined to the SOL, and no substantial net increase of the tungsten content is observed in the plasma core over the ELM time window. Pedestal conditions affect tungsten transport across the separatrix and the transient tungsten response during the ELM cycle. The results also indicate that ELM dynamics, in particular the ELM frequency, can modify the transient tungsten content in the core. Overall, the simulations suggest that isotope composition and ELM activity affect the tungsten sputtering source and its short-time redistribution, but that ELM-induced sputtering alone does not explain the loss of stationarity observed experimentally.

This paper reports the first experiment carried out in deuterium–tritium addressing the integration of a radiative divertor for heat-load control with good confinement. Neon seeding was carried out for the first time in a D–T plasma as part of the second D–T campaign of JET with its Be/W wall environment. The technical difficulties linked to the re-ionisation heat load are reported in T and D–T. This paper compares the impact of neon seeding on D–T plasmas and their D counterpart on the divertor detachment, localisation of the radiation, scrape-off profiles, pedestal structure, edge localised modes and global confinement.

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.

The presented data refer to the Shattered Pellet Injector (SPI) experiments carried out at JET in 2019–2020. This paper is a full journal version of the data originally presented as posters at TMPDM_2020 and EPS_2021. This paper presents various aspects of the interaction of pellets with plasma and associated disruptions. The experiment was performed with I _p = (1.1–3.1) MA plasmas and mainly with Ne + D₂ pellet composition, but also with Ar pellets. The Current Quench (CQ) time, τ ₈₀₋₂₀, is the key characteristic of mitigation effectiveness. A pellet with a high content of Ne or Ar can reduce the CQ duration below the upper required JET threshold. Plasmas with high (thermal + internal poloidal magnetic) pre-disruptive plasma energy require a high content of Ne pellets to obtain a short CQ duration. Pellets with a small amount of Ne (and accordingly large amount of D), instead of causing a mitigated CQ, create the conditions for a ‘cold’ Vertical Displacement Events (VDE). The SPI was applied to plasma with different status: mainly to normal (‘healthy’) plasma, i.e. not prone to disruption, post-disruptive and VDE plasma. This study shows that SPI effectiveness in terms of CQ duration and, accordingly, EM loads does not depend on the state of the plasma, whether it is ‘healthy’ or post-disruptive plasma. SPI has been shown to reduce the axisymmetric vertical vessel reaction forces by about (30–40) % compared to unmitigated disruptions. On JET, the VDE, whether ‘hot’ or ‘cold’, always creates the conditions for a toroidal asymmetry in the plasma, so the VDE on the JET is referred to as Asymmetric VDE (AVDE). The interrupting of VDE and prevention of AVDE with SPI has been demonstrated. Thus, the effectiveness of disruption mitigation using SPI has been confirmed.

The paper reports the first demonstration of in situ laser-induced desorption — quadrupole mass spectrometry (LID-QMS) application on a large scale fusion device performed in summer 2023. LID-QMS allows direct measurements of the fuel inventory of plasma facing components without retrieving them from the fusion device. The diagnostic desorbs the retained gases by heating a 3 mm diameter spot on the wall using a 1 ms long laser pulse and detects them by QMS. Thus, it can measure the gas content at any wall position accessible to the laser. The successful LID-QMS application in laboratory scale and on medium size fusion devices has now been demonstrated on the larger scale and it is already foreseen as tritium monitor diagnostic in ITER. This in situ diagnostic gives direct access to retention physics on a short timescale instead of campaign-integrated measurements and can assess the space-resolvedefficacy of detritation methods. LID-QMS can be applied on many materials: on Be deposits like in JET, B deposits like in TEXTOR, C based materials or on bulk-W.

Predictions of the Huber–Chankin (HC) scaling for the upstream impurity fraction were verified in a series of EDGE2D-EIRENE (‘code’) runs for highly radiating plasmas with nitrogen injection. The main quantity extracted from the code was poloidally averaged, from X-point to X-point, separatrix impurity fraction c_Z in the main scrape-off layer (SOL). Variation of the main working gas (H, D and T) revealed a qualitative agreement between the model and code results owing to the very large isotope difference in the predicted c_Z values caused primarily by the inverse isotope mass dependence of the H-mode power threshold assumed in the HC model and implemented in the code. At the same time, the variation of the toroidal field and safety factor in deuterium cases yielded no correlation between the model predictions and code results. The code showed much higher local impurity fractions (f_Z ) in the divertor compared to the main SOL, as well as large case-to-case variations in the divertor to the main SOL ratio of impurity fractions. The analysis of code results has wide-ranging consequences not only for the HC model, but also for other similar 1D models which use simple geometry ignoring strong neutral recycling in the divertor/ Different topology makes plasma parameters in the divertor and main SOL very different, resulting in different impurity charge state composition. Missing mechanisms in 1D codes (e.g. friction and thermo-forces exerted on impurity ions by main working gas ions) lead to impurity density redistribution. Neglecting all above factors, 1D models assume a constant impurity fraction along field lines.

Divertor power load is a major challenge towards a burning plasma in a next-step tokamak. Here, the first results of a divertor power load characterisation in tritium plasmas in type-I ELMy H-mode, obtained in the JET deuterium-tritium campaign (DTE2) performed in 2021, are presented. It is demonstrated that both, transient loads due to type-I ELMs as well as the power fall-off length, do not exhibit an explicit ion mass dependence, with remarkably similar values in the tritium plasmas and in the deuterium references. This gives an improved credence to published scaling law predictions, solely based on deuterium plasma experiments. Moreover, the type-I ELM impact on the inner divertor target is studied in deuterium discharges. A slightly increased parallel energy fluence on the inner target with a factor of 1.08 compared to the outer target is observed. This is explained by the smaller major radius of the inner target.

A remote viewing system located outside the biological shield wall to image the divertor region in JET is presented here. This paper focuses on the optical and mechanical designs of the installation of a line of sight which is capable of imaging the divertor over an optical distance of approximately 31 m. In addition, it includes the experience obtained during operations and it demonstrates its successful performance. This line of sight was the second viewing system used during the D-T campaign and it follows from the paper dedicated to the remote wide angle view line of sight which is also compatible with D-T operations and includes common elements to the construction of the two lines of sight. The main novel feature is the design in house of two glass reinforced plastic kinematic mounts standing vertically to each accommodate a mirror of 500 mm in diameter.