After 40 years of successful operation and multiple Deuterium-Tritium (D-T) campaigns, the Joint European Torus (JET) concluded its operations in December 2023. From the outset, Ion Cyclotron Range of Frequency (ICRF) heating was identified as a key auxiliary heating system, with the first ICRF system becoming operational in 1985. This system underwent significant development, culminating in a high-power (32 MW installed), wideband (23-57 MHz range), and highly versatile system. This contribution provides an overview of the JET RF system’s evolution over its lifetime, highlighting selected operational challenges and technological achievements. It addresses the rationale behind the development and operation of the ICRF system, including the implementation and operation of load-resilient systems (3 dB hybrid coupler, External Conjugate-T, ILA), as well as the operational limits and system protection measures, such as the specific mode of operation for Ion Cyclotron Wall Conditioning (ICWC). The impact of transitioning from a carbon wall to the JET-ILW metal wall (tungsten divertor and beryllium first wall) on the system’s operation and the challenges posed by D-T operations are also emphasized. Finally, key lessons learned from the technical challenges, constraints, and achievements of the JET experience are highlighted, with an eye towards the design and operation of ICRF systems in future machines.

This study presents the validation of the finite element code ERMES 20.0, benchmarked against the well-established method-of-moments code TOPICA. The simulations focus on Ion Cyclotron Resonance Heating (ICRH) coupling for the JET A2 antennas. Validation is performed by comparing two key metrics: the scattering parameter matrix (S-matrix) and the electric field distribution in front of the antenna. These parameters are critical for assessing ICRH antenna-plasma coupling efficiency and understanding interactions with plasma-facing components. The results show a strong agreement between ERMES 20.0 and TOPICA, confirming the accuracy and reliability of the finite element approach. This comparative analysis highlights the capability of ERMES 20.0 to support advanced modeling of sheath rectification and wave–edge plasma interactions, which are central phenomena in ICRH system design and optimization.

JET returned to deuterium-tritium operations in 2023 (DTE3 campaign), approximately two years after DTE2. DTE3 was designed as an extension of JET's 2022-2023 deuterium campaigns, which focused on developing scenarios for ITER and DEMO, integrating in-depth physics understanding and control schemes. These scenarios were evaluated with mixed D-T fuel, using the only remaining tritium-capable tokamak until its closure in 2023. A core-edge-SOL integrated H-mode scenario was developed and tested in D-T, showing good confinement and partial divertor detachment with Ne-seeding. Stationary pulses with good performance, no tungsten accumulation, and even without ELMs were achieved in D-T. Plasmas with pedestals limited by peeling modes were studied with D, T-rich, and D-T fuel, revealing a positive correlation between pedestal electron pressure and pedestal electron density. The Quasi-Continuous Exhaust regime was successfully achieved with D-T fuel, with access criteria similar to those in D plasmas. A scenario with full detachment, the X-point radiator regime, was established in D-T, aided by the real-time control of the radiator's position. The crucial characterisation of tritium retention continued in DTE3, using gas balance measurements and the new LID-QMS diagnostic. Nuclear technology studies were advanced during the DTE3 campaign, addressing issues such as the activation of water in cooling loops and single event effects on electronics. Building on the previous D, T and DTE2 campaigns and the lessons learned from them, DTE3 extended our understanding of D-T plasmas, particularly in scenarios relevant to next-generation devices such as ITER and DEMO.

Here we report the progress of the development and optimization of operational scenarios for ITER and beyond, focusing upon baseline, hybrid, and steady-state scenarios since 2007. This includes advancements made by the integrated operation scenarios (IOS) topical group of the international tokamak physical activity as well as contributions from the broader tokamak community. The key area of research involves developing IOSs that encompass tokamak physics, operation, and technology by utilizing integrated modeling and control strategies. This requires leveraging available actuators to simultaneously control plasma position and shape, MHD activities that could lead to disruptions, transport, plasma-wall interaction and power exhaust, fuel cycle, fusion burn, and tritium breeding. The control extends from the plasma initiation phase, through the current ramp-up, flattop, start and end of the fusion burn, and current ramp-down, to the plasma termination phase. A review of the currently developed scenarios and modeling is provided in terms of (i) optimizing plasma initiation in ITER, (ii) preparing for the low activation phase to fully commission all tokamak systems and establish and validate physics and scenario conditions in preparation for deuterim-tritium (DT) operation, (iii) developing and preparing baseline and hybrid scenarios to demonstrate the feasibility of achieving these regimes within device constraints, (iv) exploring steady-state scenarios to meet ITER’s steady-state goals, (v) evaluating and preparing actuators for ITER, (vi) developing integrated control solutions using shared actuators. The most notable achievements include; (i) the development of ITER demonstration discharges by matching various dimensionless parameters, (ii) the development of scenarios in an ITER-like tungsten environment and DT operation, and (iii) the development of scenarios in superconducting tokamaks, enabling long-pulse operations with similar coil constraints to ITER. Along with these significant achievements, outstanding issues and recommendations for further research and development are provided. Importantly, this study goes beyond simply updating the ITER Physics Basis; it carries profound implications for the broader field of burning plasma research, offering valuable insights and guidance for the next generation of fusion experiments and devices.

The typical pulse on the JET tokamak is ∼10 s during the main phase of the discharge, however long discharge operation (>30 s) is possible with sufficient preparation and care. During the last period of JET operation in 2023 long pulses in deuterium plasmas were developed to assess the sustainment of the plasma performance over several times the current resistive time scale and to address plasma-wall interaction physics in a full metallic environment with the ITER-like wall, with a W divertor and a Be first wall. To prepare for the long pulse operation an analysis of heatloads was required to ensure the pulse was safe for the machine, this defined a number of choices on toroidal field and plasma configuration. While the 30 s pulse was within the control and protection systems commissioned operating envelope the target 60 s pulse was beyond the normal operation of the control and protection systems. These systems were adapted and tested as far as possible to ensure they would work in the real pulse and a number of issues resolved over a series of tests. Significant modifications were required to carry out the experiment which had to be reversed before going back to standard operations. Even with these extensive preparations issues were found and resolved leading to the success of the 60 s pulse. The technical details of these preparations and their implementation will be discussed in detail.

This work studies the influence of radio frequency (RF) waves in the ion cyclotron resonance heating (ICRH) range of frequencies on fusion alphas during the recent JET D-T campaign. Fusion alphas from D-T reactions are created with energies of about 3.5 MeV and therefore have significant Doppler shifts enabling synergistic interactions between them and RF waves at a broad range of frequencies, including the ones foreseen for future fusion machines in ITER (Schneider et al 2021 Nucl. Fusion 61 126058) and SPARC (Creely et al 2020 J. Plasma Phys. 86 865860502). Resonant interactions between RF waves and alphas, also called synergistic effects, will modify the alpha distribution and ultimately will have an impact on alpha orbit losses and heating. Data from JET 3.43 T/2.3 MA pulses based on the hybrid scenario (Hobirk et al 2023 Nucl. Fusion ; Hobirk et al 29th IAEA FEC23 Conf. ( 16–21 October 2023 ); Challis et al 48th EPS Conf. on Plasma Physics ( 27 June–1 July 2022 ) during the DTE2 campaign (Maggi et al 2023 Nucl. Fusion )) were used for the analysis in this study. The impact of synergistic effects on alpha orbit losses and alpha heating are assessed. The conclusions are based on the analysis of experimental data for fast alpha losses, i.e. measurements from neutral particle analyser (NPA), fast ion losses scintillator detector, Faraday cups (FCs), and TRANSP (Hawryluk et al 1980 Physics of Plasmas Close to Thermonuclear Conditions vol 1 (CEC) pp 19–46) simulations. Experimental data and TRANSP analysis indicates that there are indeed changes in the alpha distribution function (DF) due to interaction with RF waves. Data from the NPA show increased 4 He flux in the range from a few hundred keV up to 800 keV for pulses with RF power, while TRANSP clearly shows modifications in the fast alpha DF for these energies. Data from the scintillator detector and the FCs were compared for pulses with and without ICRH power and versus cases with enhanced alpha losses due to MHD activity. The trends from these diagnostics consistently show no additional alpha losses due to interaction with RF waves. TRANSP predictions for the impact of ynergistic effects on alpha heating show up to a 42% increase in alpha electron heating and up to a 25% increase in alpha ion heating. These effects, however, become negligibly small, less than 1%, when alpha heating is compared to the total auxiliary heating power in the investigated JET pulses.

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).

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 alpha-particles is essential to provide efficient heating of the bulk plasma and steady burning of a reactor plasma. That is why the fusion-born alpha-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 alpha-particles, provided the possibility for comprehensive studying of the alpha-particle behaviour. Selected results of the confined and lost alpha-particle measurements, evidence of alpha-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.

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.

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.

Building on prior analysis of ASDEX Upgrade (AUG) experiments (Henderson et al 2023 Nucl. Fusion 63 086024), this study compares simple analytical formula predictions for divertor detachment onset and reattachment timescales in JET experiments. Detachment onset primarily scales with divertor neutral pressure, impurity concentration, power directed to the targets, machine size, and integral perpendicular power decay length. JET experiments, focusing on seeding mixtures of Ne and Ar, align with the detachment onset predictions. Radiation efficiencies among the impurities show good agreement with the model predictions, contrasting with AUG observations which suggested higher efficiency for Ar and lower efficiency for Ne. The time taken to re-ionise the neutral volume in front of the outer target in fully detached divertor conditions was measured following both abrupt increases in injected neutral beam power and, separately, cutting of the impurity gas flow. Re-ionisation of the neutrals occurs within approximately 1 s on JET, which aligns with the simple model prediction derived from AUG data. While the AUG results are not new, their comparison with the JET results enhances understanding, reinforcing confidence in using simple models to predict future reactor scenarios.

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.

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).

Unusually high power loads on the beryllium limiter caused by neutral beam re-ionisation, and much cooler divertor target surfaces were observed during the recent JET tokamak tritium campaign. As both phenomena are driven by scrape-off layer (SOL) physics, the SOL features of 72 tritium H-mode discharges and their deuterium references have been studied. The majority (70) of tritium H-mode discharges had exponentially decaying SOL profiles. The tritium plasmas are observed to have increased separatrix density and collisionality compared to their deuterium references. This is associated with ≈ 2 − 3 times broader SOL width for both density and temperature profiles. This is consistent with previous observations in highly collisional deuterium H-mode plasma on the ASDEX Upgrade tokamak (Sun et al 2015 Plasma Phys. Control. Fusion 57 125011) and interpreted as high collisionality enhancing cross-field transport across the separatrix and resulting in the broadening of near SOL above a critical value. The other two tritium H-mode discharges had near flat SOL density profiles, similar to the so-called ‘ density shoulder formation ’ observed in L-mode plasma. The SOL collisionality of these two pulses lies within the range of T pulses without density shoulder formation. This supports the conclusion of previous studies (Vianello et al 2017 Nucl. Fusion 57 116014; Wynn et al 2018 Nucl. Fusion 58 056001) that increased collisionality is not sufficient for the formation of a ‘ density shoulder ’ and additional factors, likely divertor condition or interaction with neutrals, are required. JET tritium plasma provides evidence of favourable and unfavourable effects of enhanced cross-field SOL transport on machine operation. The larger limiter power loads due to re-ionisation of neutral beam injection observed in the T pulses relative to their D references has been shown to be consistent with the combined effects of the broadening of the SOL profile and larger beam ion Larmor radius. The enhanced cross-field particle transport and the resulting broader SOL width provides more particles to ionize the fast Beam neutrals, causing the unfavourable power load issue on the beryllium limiter. The broader near SOL profiles of the T plasma spreads the heat load over a larger area and, together with the increased separatrix density, results in a favourably cooler divertor target surface.

M. MASLOV, E. LERCHE, F. AURIEMMA, E. BELLI, C. BOURDELLE, C. D. CHALLIS, A. CHOMICZEWSKA, A. DAL MOLIN, J. ERIKSSON, J. GARCIA, J. HOBIRK, I. IVANOVA-STANIK, PH. JACQUET, A. KAPPATOU, Y. KAZAKOV, D. KEELING, D.B. KING, V. KIPTILY, K. KIROV, D. KOS, R. LORENZINI, E. DE LA LUNA, C.F. MAGGI, J. MAILLOUX, P. MANTICA, M. MARIN, G. MATTHEWS, I. MONAKHOV, M. NOCENTE, G PUCELLA, D. RIGAMONTI, F. RIMINI, S. SAARELMA, M SALEWSKI, E.R. SOLANO, Z. STANCAR, G. STANKUNAS, H. SUN, M. TARDOCCHI, D. VAN EESTER and JET CONTRIBUTORS*

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.

The success of magnetic confinement fusion as an energy source relies crucially on reaching the necessary high temperatures for the fuel D and T ions.In a fusion reactor, plasma heating with waves in the ion cyclotron range of frequencies (ICRF) is the only system capable of providing a large fraction of bulk ion heating.Furthermore, in view of a better understanding of the non-linear physics of alpha heating in ITER and future reactors, generating MeV-range ions and studying the impact of fast ions on plasma stability and confinement becomes progressively more important.This paper summarizes recent theoretical progress and experimental demonstrations of the so-called three-ion ICRF scenario on the tokamaks Alcator C-Mod, ASDEX Upgrade and JET.In particular, the paper highlights several key results relevant for ITER and future fusion reactors.

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 Ti 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.

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.

We present an overview of results from a series of L–H transition experiments undertaken at JET since the installation of the ITER-like-wall (JET-ILW), with beryllium wall tiles and a tungsten divertor. Tritium, helium and deuterium plasmas have been investigated. Initial results in tritium show ohmic L–H transitions at low density and the power threshold for the L–H transition (P LH) is lower in tritium plasmas than in deuterium ones at low densities, while we still lack contrasted data to provide a scaling at high densities. In helium plasmas there is a notable shift of the density at which the power threshold is minimum ( n¯e,min ) to higher values relative to deuterium and hydrogen references. Above n¯e,min (He) the L–H power threshold at high densities is similar for D and He plasmas. Transport modelling in slab geometry shows that in helium neoclassical transport competes with interchange-driven transport, unlike in hydrogen isotopes. Measurements of the radial electric field in deuterium plasmas show that E r shear is not a good indicator of proximity to the L–H transition. Transport analysis of ion heat flux in deuterium plasmas show a non-linearity as density is decreased below n¯e,min . Lastly, a regression of the JET-ILW deuterium data is compared to the 2008 ITPA scaling law.