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The two best performing pulses of the so called ITER-Baseline scenario ( I _p = 3.5 MA and P _in ≈ 35 MW) of JET-ITER like wall, one in deuterium (D) the other in deuterium–tritium (D–T) plasma are examined and compared in this study. Generally, the D–T Baseline pulses exhibit an electron density level higher than the D pulses and the plasma energy is higher than in the comparable D pulses by up to 20%, reaching about 12 MJ in the pulse studied here. In contrast with the D pulses, the D–T pulses are often characterised by the increase in time of the radiated power in the mantle region (0.70 < ρ < 0.95), which may lead to the loss of the edge localised mode activity when the threshold H–L transition power is approached and to the subsequent plasma disruption due to excessive radiation. In this study we try to identify the physical mechanisms responsible for this behaviour using the available experimental data (principally the total radiated power from the bolometry) and the results of the fluid COREDIV model (1D in the core, 2D in the scrape-off-layer (SOL)), self-consistent with respect to core-SOL and also to main plasma-impurities. In fact, the loss of power caused by impurity radiation affects the temperature profile and finally the power to the divertor plate. The electron density and temperature profiles are numerically reconstructed as well as the radiated power density profiles, indicating no major difference in impurity transport in D and D–T. Indeed, the impurity transport coefficients used in COREDIV to match the experimental radiated power profiles are similar in the two pulses. The computed tungsten sources and densities are lower in the D–T pulse and the divertor impurity retention capability is a little better in the D–T pulse, indicatinga stronger collisional drag force in the SOL. The higher electron density and the broadening of its profile are the main cause of the observed increase of the radiated power in the D–T pulse.

ITER is of key importance in the European fusion roadmap as it aims to prove the scientific and technological feasibility of fusion as a future energy source. The EUROfusion consortium of labs within Europe is contributing to the preparation of ITER scientific exploitation and operation and aspires to exploit ITER outcomes in view of DEMO. The paper provides an overview of the major progress obtained recently, carried out in the frame of the new (initiated in 2021) EUROfusion work-package called ‘ Pr eparation of I TER O peration’ (PrIO). The overview paper is directly supported by the eleven EUROfusion PrIO contributions given at the 29th Fusion Energy Conference (16–21 October 2023) London, UK [ https://www.iaea.org/events/fec2023 ]. The paper covers the following topics: (i) development and validation of tools in support to ITER operation (plasma breakdown/burn-through with evolving plasma volume, new infra-red synthetic diagnostic for off-line analysis and wall monitoring using Artificial Intelligence techniques, synthetic diagnostics development, development and exploitation of multi-machine databases); (ii) R&D for the radio-frequency ITER neutral beam sources leading to long duration of negative deuterium/hydrogen ions current extraction at ELISE and participation in the neutral beam test facility with progress on the ITER source SPIDER, and, the commissioning of the 1 MV high voltage accelerator (MITICA) with lessons learned for ITER; (iii) validation of neutronic tools for ITER nuclear operation following the second JET deuterium–tritium experimental campaigns carried out in 2021 and in 2023 (neutron streaming and shutdown dose rate calculation, water activation and activated corrosion products with advanced fluid dynamic simulation; irradiation of several materials under 14.1 MeV neutron flux etc).

The JET-ILW pure tritium and deuterium–tritium (DTE2) experimental campaigns took place in 2021–2022. Tritium (T) and deuterium–tritium (D–T) operations present challenges not encountered in present day tokamaks (Horton et al 2016 Fusion Eng. Des. 109–111 925–36). This contribution focuses on ion cyclotron resonance heating (ICRH) operations in tritium and deuterium–tritium plasmas, starting with a summary of the program of improvements to the ICRH system which spanned a few years prior to these experiments. Procedures were implemented to address specific constraints from tritium and deuterium–tritium operations (tritium safety and reduced access to the RF generator area) and increase the system reliability and power availability during plasma pulses. Operation of the upgraded real time RF power control system that maximises the launched power while taking into account limitations from the system or antenna coupling is described. We also report on the result from dedicated pulses performed to assess the potential harmful impact of the 2nd harmonic tritium resonance in the plasma, close to the inner wall, when using the standard central hydrogen minority ICRH scheme. During DTE2, the ITER-like antenna was not used because water leaked from an in-vessel capacitor into the vessel on day-2 of the experimental campaign. The lessons learnt from this incident are highlighted. Finally, the ICRH plant adjustments required to safely perform ion cyclotron wall cleaning discharges are described.

A baseline scenario of deuterium–tritium (D–T) plasma with peripheral high-field-side fuelling pellets has been produced in JET in order to mimic the situation in ITER. The isotope mix ratio is controlled in order to target the value of 50%–50% by a combination of tritium gas puffing and deuterium pellet injection. Multiple factors controlling the fuelling efficiency of individual pellets are analysed, with the following findings: (1) prompt particle losses due to pellet-triggered edge-localised modes (ELMs) are detected, (2) the plasmoid drift velocity might be smaller than that predicted by simulation, (3) post-pellet particle loss is controlled by transient phases with ELMs.The overall pellet particle flux normalised to the heat flux is similar to that in previous pellet fuelling experiments in AUG and JET.

The fusion reaction between deuterium and tritium, D ( T,n ) ^4 He is the main source of energy in future thermonuclear reactors. Alpha-particles ( ^4 He -ions) born with an average energy of 3.5 MeV transferring energy to the thermal plasma during their slowing down, should provide the self-sustained D–T plasma burn. The adequate confinement of α -particles is essential to provide efficient heating of the bulk plasma and steady burning of a reactor plasma. That is why the fusion-born α -particle studies have been a priority task in the second D–T experiments (DTE2) on the Joint European Torus (JET) to understand the main mechanisms of their slowing down, redistribution and losses and to develop optimal plasma scenarios. JET with Be -wall and W -divertor, enhanced auxiliary heating systems and improved energetic-particle diagnostic capabilities, producing significant population of α -particles, provided the possibility for comprehensive studying of the α -particle behaviour. Selected results of the confined and lost α -particle measurements, evidence of α -particle self-heating and assessments of the fusion performance are presented in this paper giving an opportunity for further modelling and extrapolation to the International Thermonuclear Experimental Reactor and burning plasma reactors.

The JET Deuterium-Tritium-Experiment Campaign 2 (DTE2) has demonstrated the highest-ever fusion energy production. To forecast the transport dynamics within these discharges, the TGLF and NEO models within the TGYRO transport code were employed. A critical development in this study is the new quasilinear transport model, TGLF-SAT2, specifically designed to resolve discrepancies identified in JET deuterium discharges. This model accurately describes the saturated three-dimensional (3D) fluctuation spectrum, aligning closely with a database of nonlinear CGYRO turbulence simulations, thereby enhancing the predictive accuracy of TGYRO simulations. In validating against the JET DTE2 discharges across two primary operating scenarios, TGYRO effectively predicted the temperature profiles within a broad radial window ( ρ ∼ 0.2–0.85), though with minor ion temperature discrepancies near the core. However, a consistent underprediction of electron density profiles by 20% across the simulation domain was noted, indicating areas for future refinement. To achieve a self-consistent steady-state solution based on the JET DTE2 discharges, an integrated modeling workflow TGYRO-STEP within the OMFIT framework was introduced. This workflow iterates among the core transport, the pedestal pressure and the MHD equilibrium, ultimately yielding a converged solution that significantly reduces dependence on experimental boundary conditions for temperature and density profiles. The integrated simulation results show negligible differences in electron density and temperature profiles compared to standalone TGYRO modeling, while the ion temperature profile is lower due to the updated boundary condition in TGYRO-STEP. The application of the TGYRO-STEP workflow to JET DTE2 discharges serves as a crucial test to validate its robustness and highlights its limitations, providing valuable insights for its potential future application in ITER and Fusion Power Plant deuterium and tritium prediction modeling.

A new H-mode regime at low density and low edge safety factor ( q _95 = 3.2, with $I_\mathrm{p}$ = 3 MA) that combines high energy confinement, stationary conditions for density and radiation and small Edge Localized Modes (ELMs) have been found in JET with Be/W wall. Such a regime is achieved by operating without external gas puffing, leading to a decrease in the edge density and a substantial increase in rotation and ion temperature in both the pedestal and the core region. Transport modelling shows a reduction of the turbulence, which starts in the pedestal region and extends into the plasma core, and outward impurity convection, consistent with the improved energy confinement and the lack of W accumulation observed in those conditions. In addition, large type I ELMs, typically found in gas-fuelled plasmas, are replaced by smaller and more frequent ELMs, whose appearance is correlated with a substantial reduction of the pedestal density and its gradient. Pedestals in this operating regime are stable to peeling–ballooning modes, consistent with the lack of large ELMs. This is in contrast to results in unfuelled JET-C plasmas that typically operated at higher pedestal densities and developed low frequency, large type I ELMs, thus pointing to the low density as one of the critical parameters for accessing this small ELMs regime in JET. This small ELMs regime exhibits the same low pedestal collisionality ( $\nu_{\mathrm{e},\mathrm{ped}}^*\sim0.1$ ) expected in ITER and operates at low q _95 , thus making it different from other small ELMs regimes that are typically obtained at higher q _95 and higher pedestal collisionality. These features make this newly developed H-mode regime in JET with Be/W wall a valuable tool for exploring the underlying transport, the different mechanisms of turbulence stabilization, as well as the physics associated with the appearance of small ELMs in high-temperature plasmas at ITER relevant pedestal collisionality.

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).

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.

The JET hybrid scenario has been developed from low plasma current carbon wall discharges to the record-breaking Deuterium-Tritium plasmas obtained in 2021 with the ITER-like Be/W wall. The development started in pure Deuterium with refinement of the plasma current, and toroidal magnetic field choices and succeeded in solving the heat load challenges arising from 37 MW of injected power in the ITER like wall environment, keeping the radiation in the edge and core controlled, avoiding MHD instabilities and reaching high neutron rates. The Deuterium hybrid plasmas have been re-run in Tritium and methods have been found to keep the radiation controlled but not at high fusion performance probably due to time constraints. For the first time this scenario has been run in Deuterium-Tritium (50:50). These plasmas were re-optimised to have a radiation-stable H-mode entry phase, good impurity control through edge T _i gradient screening and optimised performance with fusion power exceeding 10 MW for longer than three alpha particle slow down times, 8.3 MW averaged over 5 s and fusion energy of 45.8 MJ.

We present the JETTO-QuaLiKiz-SANCO fully predictive modelling of two JET-ILW high-performance baseline plasmas, a Ne seeded shot and an equivalent unseeded one. The motivation of the work lies in the experimental observation of a slightly higher confinement and performance of the Ne seeded shot with respect to the unseeded one, despite sharing the same main plasma parameters and heating powers. Moreover, the neon seeded shot shows a lower pedestal electron density and a higher core ion temperature with respect to the unseeded one. Integrated modelling is performed in order to understand if the cause of the improved confinement has to be ascribed to the improved pedestal parameters with neon seeding or if an impurity-induced turbulence stabilization is at play. The QuaLiKiz transport model is used for predicting the electron density, electron and ion temperatures and rotation in the core up to the pedestal top, while the pedestal is empirically modelled to reproduce the experimental kinetic profiles. The thermal diffusivities of the two shots, computed by QuaLiKiz, are compared, as well as the turbulence spectra, suggesting that the reduced transport found in the neon seeded shot is due in part to the stabilization of ion temperature gradient and electron temperature gradient modes. Further modelling is performed in order to disentangle the neon seeding effects, which are a direct effect on the turbulence stabilization and an indirect effect on the pedestal parameters. The results suggest that the improved performance with neon is due to a combination of turbulence stabilization and improved pedestal parameters.

In JET deuterium-tritium (D-T) plasmas, the fusion power is produced through thermonuclear reactions and reactions between thermal ions and fast particles generated by neutral beam injection (NBI) heating or accelerated by electromagnetic wave heating in the ion cyclotron range of frequencies (ICRFs). To complement the experiments with 50/50 D/T mixtures maximizing thermonuclear reactivity, a scenario with dominant non-thermal reactivity has been developed and successfully demonstrated during the second JET deuterium-tritium campaign DTE2, as it was predicted to generate the highest fusion power in JET with a Be/W wall. It was performed in a 15/85 D/T mixture with pure D-NBI heating combined with ICRF heating at the fundamental deuterium resonance. In steady plasma conditions, a record 59 MJ of fusion energy has been achieved in a single pulse, of which 50.5 MJ were produced in a 5 s time window ( P _fus = 10.1 MW) with average Q = 0.33, confirming predictive modelling in preparation of the experiment. The highest fusion power in these experiments, P _fus = 12.5 MW with average Q = 0.38, was achieved over a shorter 2 s time window, with the period of sustainment limited by high-Z impurity accumulation. This scenario provides unique data for the validation of physics-based models used to predict D-T fusion power.

The reference ion cyclotron resonance frequency (ICRF) heating schemes for ITER deuterium–tritium (D-T) plasmas at the full magnetic field of 5.3 T are second harmonic heating of T and ^3 He minority heating. The wave-particle resonance location for these schemes coincide and are central at a wave frequency of 53 MHz at 5.3 T. Experiments have been carried out in the second major D-T campaign (DTE2) at JET, and in its prior D campaigns, to integrate these ICRF scenarios in JET high-performance plasmas and to compare their performance with the commonly used hydrogen (H) minority heating. In 50:50 D:T plasmas, up to 35% and 5% larger fusion power and diamagnetic energy content, respectively, were obtained with second harmonic heating of T as compared to H minority heating at comparable total input powers and gas injection rates. The core ion temperature was up to 30% and 20% higher with second harmonic T and ^3 He minority heating, respectively, with respect to H minority heating. These are favourable results for the use of these scenarios in ITER and future fusion reactors. According to modelling, adding ICRF heating to neutral beam injection using D and T beams resulted in a 10%–20% increase of on-axis bulk ion heating in the D-T plasmas due to its localisation in the plasma core. Central power deposition was confirmed with the break-in-slope and fast Fourier transform analysis of ion and electron temperature in response to ICRF modulation. The tail temperature of fast ICRF-accelerated tritons, their enhancement of the fusion yield and time behaviour as measured by an upgraded magnetic proton recoil spectrometer and neutral particle analyser were found in agreement with theoretical predictions. No losses of ICRF-accelerated ions were observed by fast ion detectors, which was as expected given the high plasma density of n _e ≈ 7–8 × 10 ^19 m ^−3 in the main heating phase that limited the formation of ICRF-accelerated fast ion tails. ^3 He was introduced in the machine by ^3 He gas injection, and the ^3 He concentration was measured by a high-resolution optical penning gauge in the sub-divertor region. The DTE2 experiments with ^3 He minority heating were carried with a low ^3 He concentration in the range of 2%–4% given the fact that the highest neutron rates with ^3 He minority heating in D plasmas were obtained at low ^3 He concentrations of ∼2%, which also coincided with the highest plasma diamagnetic energy content. In addition to ^3 He introduced by ^3 He gas injection, an intrinsic concentration of ^3 He of the order of 0.2%–0.4% was measured in D-T plasmas before ^3 He was introduced in the device, which is attributed to the radioactive decay of tritium to ^3 He. According to modelling, even such low intrinsic concentrations of ^3 He lead to significant changes in ICRF power partitioning during second harmonic heating of T due to absorption of up to 30% of the wave power by ^3 He.

The need for a model able to provide the dynamic evolution of the intrinsic magnetic error fields amplitude in tokamaks is of a significant importance in order to develop strategies to correct its destabilizing effect. The present paper specifically aims to deliver this kind of model and adjacent code. A previously built direct perturbations interpretative model (Miron (JET Contributors) 2021 Nucl. Fusion 61 106016) is used to calculate the plasma instabilities amplitude and frequency aiming to match the same quantities provided by the device diagnostics data analysis specific code for various discharges at JET. The mentioned good match ensures our model validity against the experimental results. Based on its proven structural validity, our model considering the error field quantities is inverted. This time the initial direct model output results, namely the plasma perturbations amplitude and frequency, become input data for our inverse model aiming to solve the corresponding error field modes system of equations searched as unknowns. The error fields basically satisfy the same outer plasma perturbed equations as the plasma perturbations do (vacuum/wall/plasma column external structures). Obviously they do not satisfy the perturbed plasma equations. A clear and explicit dynamic error field solution is finally provided. It has been demonstrated that whenever the error fields are responsible for the mode locking effect and plasma rotation damping, the calculated error fields shows its expected disruptive resonant behavior during various discharges at JET.

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 $\left( {13.4 \pm 0.7} \right) \times {10^{22}}$ atoms or $\left( {0.67 \pm 0.03} \right)$ 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 midplane electron separatrix density, n _e,sep , in JET-ILW L-mode and H-mode low triangularity deuterium fuelled plasmas exhibits a strong explicit dependence on the averaged outer divertor target electron temperature, n _e,sep ∼ T _e,ot ^−1/2 . This dependence is reproduced by analytic reversed two point model (rev-2PM), and arises from parallel pressure balance, as well as the ratio of the power and momentum volumetric loss factors, (1 − f _cooling )/(1 − f _mom-loss ). Quantifying the influence of the (1 − f _cooling ) and (1 − f _mom-loss ) loss factors on n _e,sep has been enabled by measurement estimates of these quantities from L-mode density (fueling) ramps in the outer horizontal, VH(C), and vertical target, VV, divertor configurations. Rev-2PM n _e,sep estimates from the extended H-mode and more limited L-mode datasets are recovered to within ±25% of the measurements, with a scaling factor applied to account for use of $\left\langle {{T_{{\text{e,ot}}}}} \right\rangle $ , an averaged quantity, rather than flux tube resolved target values. Both the (1 − f _cooling ) and (1 − f _mom-loss ) trends and recovery of n _e,sep using the rev-2PM formatting are reproduced in EDGE2D-EIRENE L-mode-like and H-mode-like density scan simulations. The general lack of a divertor configuration effect in the JET-ILW n _e,sep trends can be attributed to a significant influence of main chamber recycling, which has been shown in the EDGE2D-EIRENE results to moderate n _e,sep with respect to changes in divertor neutral leakage imposed by changes in the divertor configuration. The unified n _e,sep vs $\left\langle {{T_{{\text{e,ot}}}}} \right\rangle $ trends can, however, be broken if large modifications to the divertor geometry (e.g. complete removal of the outer divertor baffle structure) are introduced in the model. The more pronounced high-field side high density region formation in the VH(C) configuration with reduced clearance to the separatrix does not appear to have a significant influence on the outer midplane separatrix and pedestal parameters when mapped to $\left\langle {{T_{{\text{e,ot}}}}} \right\rangle $ , although conditions at the inner midplane could not be assessed.