After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the 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 experimental and theoretical analysis were focused on experiments conducted to assess the effect of plasma isotopes, protium (H), deuterium (D), and tritium (T) on ion cyclotron resonance heating (ICRH) related plasma wall interactions. Comparison of L-mode discharges with N = 1 3 He and N = 1 H minority ICRH heating scenarios were done for different isotopes. For the selected pulses, the behaviour of high-Z, mid-Z and low-Z intrinsic impurity and radiated power behaviour was investigated based on data from VUV, visible spectroscopy, and bolometry diagnostic at Joint European Torus. It was found that for N = 1 3 He scenario during radiofrequency antennas operation, core W, Ni content, Be source and the radiated power are higher for π /2 in comparison to dipole antenna phasing. Lowest core Ni, W content and radiated power is clearly observed for H plasmas in comparison to D and T, where for this ICRH scenario behaviour was similar. However, lower Be photon flux is observed for T in comparison to D plasmas. Be sputtering by He particles is responsible for such an effect. Additionally, several computer simulations were conducted using the COREDIV code. The difference in the electron temperature was due to the difference in the isotope masses. Increased temperature in the central plasma in the case of T plasmas leads to higher radiation in the central plasma in comparison to H plasmas. As a result, the power across separatrix is lower and the temperature on the divertor plate decreases with the increase of the isotope mass. At these temperatures on the divertor plate, W is not sputtered by the main plasma ions H, D and T and by He. For the N = 1 H ICRH scenario clear difference between D and T plasma was observed with higher metallic impurity content for T plasma in comparison to D. Impurity content in the plasmas is found to be sensitive to the power balance between the antenna straps. Its minimum is observed for the maximum of P cen / P tot .

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).

We present here the results of the spectroscopic analysis of the high-resolution visible spectra of beryllium hydride and its isotopologues (BeH, BeD, and BeT), produced during plasma–surface interactions during limiter and divertor JET-ILW (ITER-like Wall) pulses. The hydride production, being an important part of the wall erosion via chemical-assisted physical sputtering, shows visible dependence on plasma and wall conditions, and also on the isotope content of the plasma. This work shows that this dependence is also true for the molecular energy distributions, parameterized by rotational and vibrational temperatures. During the increase in the vessel wall temperature by 350 K, with constant plasma parameters, rotational temperature of the excited BeD molecules increases by 500 K, whereas the vibrational temperature decreases by 400 K. Another experiment was the scan of the integrated edge density, during which electron density increases with the increasing edge density, and electron temperature decreases. During that scan for BeD, Trot decreases by 300 K and Tvib by 400 K over the change in an integrated edge density of ΔNe=+8×1018 m−2. Both trends were compared with the trends in other isotopes, and the qualitative results are similar, but the limited data available restrict the possibility of a full quantitative comparison. To simplify the comparison between the results presented here and other experiments or modeling data, and also to check the comparability of the pulses performed in different isotopes, the Be ion line ratios are used as benchmarks of the edge plasma condition.

The work presents the recent outcome of the research on Joint European Torus with ITER-like wall (JET-ILW) concerning Be, mid-Z (Ni, Cu, Fe), and high-Z (W) impurities for a selection of Neutral Beam Injection-heated, ELMy H-mode pulses using visible and vacuum–ultraviolet spectroscopy together with bolometry diagnostic. The investigation is focused on the evaluation of the plasma triangularity (δ) impact on the impurity radiation in hydrogen (H), deuterium (D), tritium (T), and deuterium–tritium (DT) plasmas for pulses described in [P. A. Schneider, Nucl. Fusion 63, 112010 (2023)]. The variations of δ were in the range of 0.21–0.31 for different time windows with low magnetic toroidal field (Bt=1.7 T) and plasma current (Ip=1.4 MA) in the corner–corner configuration, that is with both the inner and outer strike points on the horizontal W-coated divertor target plates. The results confirm the rise in Be flux with plasma isotope and lower δ leading to higher plasma density close to the plasma wall. The dominant role of W as a source of plasma radiation has also been confirmed. For mid-Z impurities and W, the variations in their densities due to δ change are negligible and the rise in their densities is observed with higher isotope mass, although this effect is often masked by the dominant role of ELM frequency on the impurity level. The impurity radiation losses based on bolometry together with the W intensity are ruled by the changes in the electron density and are well correlated when the Te profiles are conserved.

The analysis of hydrogen isotope content is crucial for understanding the operation of fusion devices. Hydrogen isotopic analysis in the core and plasma edge is conducted through neutron and spectroscopic diagnostics. In the case of exhaust gas, mass spectrometry is employed using residual gas analyzers (RGAs), along with optical gas analysis (OGA) utilizing an optical Penning gauge. The use of traditional quadrupole mass spectrometers for RGA encounters challenges during discharges with tritium gas due to signal overlap of different hydrogen molecules at the same mass number. This paper introduces a novel technique to address this issue through a cross-related analysis of RGA data with OGA results. Another consideration in mass spectrometry analysis is the instrument’s varying sensitivity concerning the gas mass number. The paper includes gas calibration data results for all the quadrupoles used in the JET gas analysis. Hydrogen isotopic ratios are calculated from RGA detected currents using simple formulas. The results of this procedure are presented for selected DT JET discharges in relation to the JET pulse time, and they are compared with corresponding optical data. Time-averaged hydrogen isotopic ratio values are computed for numerous discharges during the DT and T campaigns.

In view of the ultimate goal of producing long-lasting quasi-stationary discharges required for future fusion power stations, the numerical study of steady-state solutions of equations describing the particle and energy balance rightfully gets ample attention. Transient states may, however, differ significantly from the steady state ultimately reached and will – in practice – impact on the actual fate of the discharge. Using brutally simple models, the present paper highlights a number of aspects to illustrate this dynamics. It e.g. shows the different signature of wave and beam heating, potentially giving room to transiently trigger desirable effects that may allow us to better steer a discharge.

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.

Sudden inflows of W particulate are routinely observed in correlation with plasma instabilities in JET-ITER-Like-Wall experimental campaigns. Dust acoustic wave destabilization by micrometer-size tungsten dust released in such events is simulated here using realistic parameters describing the JET scrape-off layer. Such low-frequency waves may serve as a monitorable signal for dust detection in the scrape-off layer on a pulse-by-pulse basis.

This paper reports first experiment with impurity seeding in Deuterium Tritium. The JET Neon seeded ITER-baseline scenario is being investigated in a D-T environment. The impact of the isotope mass effect on key aspects of the core-edge integration is being studied such as pedestal, ELMs, divertor detachment as well as the necessary adaption to the scenario itself.

High resolution Fulcher band spectroscopy was used in the MAST-U divertors during Super-X and elongated conventional divertor density ramps with $\text{D}_{2}$ fuelling from the mid-plane high-field side. In the Super-X case (density ramp from Greenwald fraction 0.12 to 0.24), the upper divertor showed ground state rotational temperatures of the $\text{D}_{2}$ molecules increasing from $\sim$6000 K, starting at the detachment onset, to $\sim$9000 K during deepening detachment. This was correlated with the movement of the Fulcher emission region, which is correlated with the ionisation source. The increase in rotational temperature did not occur near the divertor entrance, where the plasma was still ionising. Qualitative agreement was obtained between the lower and upper divertor. Similar rotational temperatures were obtained in the elongated divertor before the detachment onset, although the increase in rotational temperature during detachment was less clearly observed as less deep detachment was obtained. %In the elongated conventional divertor there was some qualitative agreement of this effect impeded by low signal. The measured vibrational distribution of the upper Fulcher state (first four bands) does not agree with a ground state Boltzmann distribution but shows a different characteristic with an elevated population especially in the $\nu = 2$ and $\nu = 3$ bands. The populations of the $\nu = 2$ and $\nu = 3$ band relative to the $\nu = 0$ band are roughly proportional to the $\textit{rotational}$ temperature.

The pre-fusion power operation (PFPO) phase of ITER, as described in the ITER research plan with Staged Approach2, includes both hydrogen (H) and helium (He) plasma operations. In preparation for PFPO, both WEST and JET ran He plasma campaigns to study plasma-wall interactions in a tungsten environment. The campaigns included a back-and-forth transition between H or deuterium (D) and He plasma operation allowing the assessment of the achievable plasma content as well as the accessible wall reservoirs for respective species. The WEST changeovers included tokamak pulses with a fixed divertor configuration. The JET changeovers applied ion cyclotron wall conditioning (ICWC) and tokamak pulses including limiter phases and four different divertor configurations. Glow discharge conditioning (GDC) was applied to complete the changeovers. The results are characterized by subdivertor optical and mass spectrometric gas analyzers and spatially resolved optical emission spectroscopy. A He content of 96-97% after H operations is achieved by tens of ICWC pulses (JET) and several dedicated diverted plasmas (WEST and JET), while a fivefold is estimated to be required for the back transition. Effective pumping of the wall released species is a key parameter for a fast changeover. Upon applying higher heating power, the relative content of the fueled plasma species decreases. The JET gas balance analysis indicates that He operation may increase H retention. WEST divertor spectroscopy indicates a larger He inventory near the inner divertor strike line. He GDC has a clear effect on the He recycling light at the WEST divertor while D GDC did not reduce the long lasting He content observed in D pulses after the JET He campaign.

Several cross-sections of the are plasma column were observed side-on at different distances from the cathode. The are was operated in argon at atmospheric pressure with an are current of 200 A. As a basic diagnostic method we used the Olsen-Richter diagrams. For determination of temperature the Larenz-Fowler-Milne and Boltzmann plot methods were applied. Electron density was measured fi om the Stark width of the ArI and ArII lines. The continuum intensity was also measured. The equilibrium state of the plasma near the cathode tip is found to be dependent of the axial and radial position. In the hot core, the plasma is in the LTE state. Toward outer zones, the plasma state deviates from LTE. It is caused after all by the ArI ground state overpopulation.

The recent deuterium–tritium campaign in JET-ILW (DTE2) has provided a unique opportunity to study the isotope dependence of the L-H power threshold in an ITER-like wall environment (Be wall and W divertor). Here we present results from dedicated L-H transition experiments at JET-ILW, documenting the power threshold in tritium and deuterium–tritium plasmas, comparing them with the matching deuterium and hydrogen datasets. From earlier experiments in JET-ILW it is known that as plasma isotopic composition changes from deuterium, through varying deuterium/hydrogen concentrations, to pure hydrogen, the value of the line averaged density at which the threshold is minimum, nˉe,min , increases, leading us to expect that nˉe,min (T) < nˉe,min (DT) < nˉe,min (D) < nˉe,min (H). The new power threshold data confirms these expectations in most cases, with the corresponding ordering of the minimum power thresholds. We present a comparison of this data to power threshold scalings, used for extrapolation to future devices such as ITER and DEMO.

Measurements of the ion currents to and plasma conditions at the low-field side (LFS) divertor target plate in low-confinement mode plasmas in the JET ITER-like wall materials configuration show that the core plasma density required to detach the LFS divertor plasma is independent of the hydrogenic species protium, deuterium and tritium, and a 40 %/60 % deuterium–tritium mixture. This observation applies to a divertor plasma configuration with the LFS strike line connected to the horizontal part of the LFS divertor chosen because of its superior diagnostic coverage. The finding is independent of the operational status of the JET cryogenic pump. The electron temperature (Te) at the LFS strike line was markedly reduced from 25 eV to 5 eV over a narrow range of increasing core plasma density, and observed to be between 2 eV and 3 eV at the onset of detachment. The electron density (ne) peaks across the LFS plasma when Te at the target plate is 1 eV, and spatially moves to the X-point for higher core densities. The density limit was found approximately 20 % higher in protium than in tritium and deuterium–tritium plasmas.

The required heating power, P L H , to access the high confinement regime (H-mode) in tritium containing plasmas is investigated in JET with ITER-like wall at a toroidal magnetic field of B t = 1.8 T and a plasma current of I p = 1.7 MA. P L H , also referred to as the L-H power threshold, is determined in plasmas of pure tritium as well as mixtures of hydrogen with tritium (H-T) and mixtures of deuterium with tritium (D-T), and is compared to the L-H power threshold in plasmas of pure hydrogen and pure deuterium. It is found that, for otherwise constant parameters, P L H is not the same in plasmas with the same effective isotope mass, A e f f , when they differ in their isotope composition. Thus, A e f f is not sufficient to describe the isotope effect of P L H in a consistent manner for all considered isotopes and isotope mixtures. The electron temperature profiles measured at the L-H transition in the outer half of the radius are very similar for all isotopes and isotope mixtures, despite the fact that the L-H power threshold varies by a factor of about six. This finding, together with the observation of an offset linear relation between the L-H power threshold, P L H , and an effective heat diffusivity, χ e f f , indicates that the composition-dependent heat transport in the low confinement mode (L-mode) determines, how much power is needed to reach the necessary electron temperatures at the edge, and hence P L H .

This paper explores the plasma parameters of helium and hydrogen Ion Cyclotron Wall Conditioning (ICWC) discharges performed in JET as part of a He/H fuelling changeover experiment. The conducted study shows that plasma with a higher density is formed in helium than in hydrogen. A distinct glow in the ion cyclotron resonance zone is observed throughout the discharge in He. In H-ICWC discharges, a lower radio-frequency coupling efficiency and coupled power was observed than in He-ICWC discharges. While the helium concentration decreased with the number of H-ICWC pulses and the same for hydrogen in He-ICWC, which is the intended result of the plasma wall interaction in the ICWC changeover procedure, the main features of hydrogen as well as the helium IC discharge do not change dramatically.

After the second Deuterium–Tritium Campaign (DTE2) in the JET tokamak with the ITER-Like Wall (ILW) and full tritium campaigns that preceded and followed after the DTE2, a sequence of fuel recovery methods was applied to promote tritium removal from wall components. The sequence started with several days of baking of the main chamber walls at 240 °C and at 320 °C. Subsequently, baking was superimposed with Ion-Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) cleaning cycles in deuterium. Diverted plasma operation in deuterium with different strike point configurations, including a Raised Inner Strike Point (RISP) configuration, and with different plasma heating—Ion Cyclotron Resonance Frequency (ICRF) and Neutral Beam Injection (NBI)—concluded the cleaning sequence. Tritium content in plasma and in the pumped gas was monitored throughout the experiment. The applied fuel recovery methods allowed reducing the residual tritium content in deuterium NBI-heated plasmas to about 0.1% as deduced from neutron rate measurements. This value is well below the requirement of 1% set by the maximum 14 MeV fusion neutron budget allocated in the ensuing deuterium plasma campaign. The quantified tritium removal over the course of the experiment was 13.4 ± 0.7 × 10 22 atoms or 0.67 ± 0.03 g with ∼58% attributed to baking, ∼12.5% to ICWC, ∼26% to GDC, and ∼3.5% to first low power RISP plasmas. The experimentally estimated amount of removed tritium is in good agreement with long-term tritium accounting by the JET tritium reprocessing plant, in which the unaccounted amount was reduced by 0.71 g after the cleaning experiment.

The properties of tungsten make it ideal for use as a plasma facing surface in the divertor of large plasma machines such as JET and ITER. However, the intense heat and particle fluxes that fall on the divertor surfaces lead to its release from these surfaces into the plasma and it is necessary to model its transport from the divertor and plasma edge into the plasma core. This requires measurement of spectral features over a wide temperature range. In large machines the W influx is often determined from W I line intensities, there being few measurements of discrete W lines from other low ionization stages. Their observation is highly desirable because the transport of neutral W differs markedly from that of the W ions. A change in the line-of-sight of a VUV survey spectrometer on JET to view directly into the divertor has led to the observation of numerous discrete low temperature W lines in the VUV spectral region. The spectrum of an intense influx in which W IV to W VIII features are observed has been analysed in order to provide spectral classifications so that these lines can be used for diagnostic purposes. The first observation of a VUV low temperature W magnetic dipole (M1) transition is reported for the W VIII ionization stage. The analysis shows where further line identifications are needed and that the provision of the highest quality atomic data for these ionization stages is desirable. W VI and W VII line intensity measurements are used to illustrate their use in determining the electron temperature of the emitting plasma region and the W concentration. Dependences of the W line intensities on plasma parameters shows the optimal conditions for the W release and suggests the site of its release.