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.

Estimating the single-pass absorption (SPA) of a given Ion Cyclotron Resonance Heating (ICRH) scenario as function of the plasma and the radio-frequency (RF) wave properties requires a multispecies hot plasma wave solver. Even in 1D and using truncated finite Larmor radius expansion for the plasma dielectric response, the simulations cannot be done much faster than in a couple of seconds. In ITER, it is envisaged to estimate the single-pass absorption of the ICRF waves in real-time to take preventive actions in case of poor absorption for avoiding increased plasma-wall interaction and enhanced heat loads on the plasma facing components. This work proposes a potential solution based on pre-calculated look-up tables of the ICRF single-pass absorption that can be interpolated in real-time by the Advanced Protection System (APS) and by the Plasma Control System (PCS) in ITER [1,2,3] during the discharge evolution. The tables can also be used as input for the Pulse Design Simulator (PDS) [4] and other modelling tools that require the ICRF absorption information for control and/or protection. The SPA tables are produced as function of the dominant parameters governing the RF wave propagation and absorption using the 1D TOMCAT code [5] for some ITER-relevant ICRH scenarios.

The present paper shows a number of numerical methods under development at LPP-ERM/KMS to model wave propagation and damping in the RF domain of frequencies. The philosophy is to adopt a maximum of physics tools already available in the literature (i.e. no new derivations are proposed; at most mild, intuitive upgrades of the existing analytical description are introduced) and extend them to make the description more realistic while keeping the CPU needs as modest as possible.

In the 2021 and 2023 D-T campaigns in JET various scenarios with potential for application in fusion reactors have been studied. The mandate of the “Baseline” experiments was to explore the possibility to operate at high density, magnetic field and current. Although extremely promising results were obtained in D plasmas in the running-up to the actual D-T campaign and up to 8MW of fusion power was produced when adopting this scenario in D-T [1],[2], it was - in contrast to the record T-rich scenario [3] - not possible to sustain D-T shots for the envisaged 5 seconds while also steadily producing more than 10MW of fusion power. In view of the Baseline being considered as a prime candidate for maintaining a high-density plasma in future machines, the underlying reasons are still being explored to enable offering perspectives for possible cures for next-generation experiments. The present paper contributes to that: It concentrates on the detailed modelling of auxiliary (RF & NBI) heating aspects and on the synergy between them, allowing a better understanding of the key role of the auxiliary heating in these high-performance shots. It complements papers that concentrated on key - interrelated - aspects such as transport (see e.g. [4]), MHD (see e.g. [5]), impurities [6, 7], pedestal dynamics [8] and control [9]. One aspect setting the scenarios tested in D-T apart and which has several repercussions is that the Baseline plasma current and hence the density is higher and flatter. This allows to profit optimally from the fact that the neutron rate is proportional to the densities of the fusion fuel ions. However, the higher density affects the beam penetration and modifies the collisionality as well as the beam power deposition profiles. This has nonnegligible implications, some of which will be discussed here.

The capability of the global, electromagnetic gyrokinetic GENE code interfaced with the transport Tango solver is exploited to address the impact of fusion alpha particles (in their dual role of fast particles and heating source) on plasma profiles and performance at JET in the discharges with the highest quasi-stationary peak fusion power during the DTE2 experimental campaigns. Employing radially global nonlinear electromagnetic GENE-Tango simulations, we compare results with/without alpha particles and alpha heating. Our findings reveal that alpha particles have a negligible impact on turbulent transport, with GENE-Tango converging to similar plasma profiles regardless of their inclusion as a kinetic species in GENE. On the other hand, alpha heating is found to contribute to the peaking of the electron temperature profiles, leading to a 1keV drop on the on-axis electron temperature when alpha heating is neglected in Tango. The minimal impact of alpha particles on turbulent transport in this JET discharge - despite this being the shot with the highest fusion output - is attributed to the low content of fusion alpha in this discharge. To assess the potential impact of alpha particles on turbulent transport in regimes with higher alpha particle density, as expected in ITER and fusion reactors, we artificially increased the alpha particle concentration to levels expected for ITER. By performing global nonlinear GENE standalone simulations, we found that increasing the alpha particle density beyond five times the nominal value lead to significant overall turbulence destabilization.

The paper provides an assessment of the ion-cyclotron resonance heating (ICRH) system performance on JET since the year 2000. The vast amount of collected data offer an insight into the historical challenges and trends in the ICRH system performance encompassing the transition from carbon (JET-C) to beryllium & tungsten ITER-like wall (JET-ILW) operations, the deuterium-tritium experiments (DTE2 & DTE3) and introduction of new RF antenna & matching systems. The best achieved operational parameters are reported and statistics on the RF plant reliability and performance is analysed. Antenna-plasma coupling is identified as the dominant factor critical to all the aspects of the ICRH system behaviour; parametric dependencies of coupling resistance on plasma parameters and the RF plant settings are discussed and the key role of local electron concentration profiles close to the antennas is highlighted. Following confident antenna performance at high RF voltages over the recent decade, observations are presented suggesting improved electrical strength of the RF vacuum components after the JET-C to JET-ILW transition; this is tentatively attributed to the reduction of dust levels in the JET vessel. Statistics on application rate and typical origins of the RF amplifier failures and protection power limits is presented indicating that the amplifier issues noticeably affected the high-power ICRH operations. Performance comparison is provided for different RF antenna & matching systems installed at JET since 2000 including the original system, two load-tolerant systems based on the 3 dB hybrid and external conjugate-T power-splitters, and the ITER-like antenna. The paper could be of interest both as a summary of technical challenges, constraints and achievements related to the ICRH application on JET and as a reference for design and operations of high-power RF systems in future fusion devices.

An important step on the way to future fusion power plants was the 2021 deuterium-tritium experimental campaign (DTE2) at the Joint European Torus (JET), in which crucial DT physics was investigated. In this study, we have reconstructed the fast-ion deuterium distribution function in JET discharge 99971 which broke the former fusion energy record. It is the first time that the fast-ion distribution has been reconstructed from experimental data in a DT discharge. The reconstruction shows that the fast-ion deuterium distribution is anisotropic, with a bias towards co-going ions (p > 0). The fast-ion deuterium distribution likely peaks in energy (E) at around E similar to 60-70 keV and has a marginal high-energy tail ( E greater than or similar to 180 keV). Furthermore, an orbit analysis shows that the fast-ion distribution is composed of mostly co-passing orbits ( 50%), trapped orbits ( 21%) and counter-passing orbits ( 27%), as well as a small population of potato orbits ( 1.7%) and counter-stagnation orbits ( 0.3%). The orbit-type constituents of the neutron measurements are distributed in similar fractions.

This paper presents a study on the dependence of the ion temperature stiffness on the plasma main ion isotope mass in JET ITER-like wall and C wall discharges. To this aim, a database of H, D and T shots is analyzed, including new dedicated shots, comparing experiments with lower and higher power injected by the NBI system. In order to characterize the turbulence dependence on the isotope mass, three of these discharges (two in T and one in D) with same external heating scheme are studied in detail and interpreted with gyrokinetic linear and nonlinear simulations. The analysis is performed at fixed radius rho tor=0.33, selected in order to maximize the electromagnetic stabilizing effects on turbulence, both from thermal and suprathermal particles. The experimental results show a clear ion temperature stiffness reduction when heavier isotopes are considered, thus moving from H to D to T, which is attributed to an increasing thermal electromagnetic stabilization with increasing main isotope mass.

The quasi-continuous exhaust (QCE) regime is a regime that is naturally type-I ELM-free. It combines the high density at the plasma edge needed for power exhaust with the high normalised energy confinement typical for H-mode operation. In the QCE regime large-scale ELMs are avoided and high-frequency, low-amplitude filaments are present leading to the name-giving quasi-continuous edge transport of particles and energy. This contribution reports that for the first time the QCE regime was successfully achieved in JET with a metal wall. Moreover, it was demonstrated in the recent JET deuterium-tritium campaign DTE3 that the regime is compatible with D-T operation. Porting the QCE regime to JET strongly benefited from the experimental and modelling efforts at the medium sized tokamaks ASDEX Upgrade and TCV. Using the physics picture developed from the ASDEX Upgrade experimental results, the route to the QCE regime in JET reported here is following closely the approach that was successful in ASDEX Upgrade. First, strong plasma shaping-large elongation and triangularity and the highly correlated closeness to double null-is developed. Second, sufficient fuelling to achieve high enough density at the pedestal foot, close to the separatrix, is applied. In addition, neon seeding proved to be very beneficial to avoid type-I ELMs when reducing the main ion fuelling.

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.

Experimental campaigns at the Joint European Torus with an ITER-like Be/W wall with pure deuterium (D), tritium (T), and deuterium–tritium (DT) were a unique opportunity to explore various aspects related to the ITER operation. One of the most important challenges in recent years was the development of the hybrid scenario for D-T, based on reference deuterium and tritium plasmas. This kind of scenario, one of the foreseen for ITER, is characterized by a low current plasma and a high normalized beta βN factor compared to the parallel optimized baseline scenario [Hobirk et al., Plasma Phys. Controlled Fusion 54, 095001 (2012)]. As the experiments have shown, controlling the plasma edge in the different phases of the hybrid scenario becomes more difficult with higher isotope mass, and therefore, are also in risk of impurity accumulation [Hobirk et al., Nucl. Fusion 63, 112001 (2023)]. For this reason, investigation of the impurity behavior, as well as their control, constituted the crucial issue. The present contribution aims to compare mid-Z and high-Z impurities behavior within H-mode hybrid discharges in D and T plasmas, as well as D and DT plasmas. Detailed analysis shows that in the H-mode regime in the hybrid scenario, higher impurity radiation is observed for DT in comparison to D plasmas, as well as for T compared to D plasmas. Additionally, it was noticed that the most significant contribution to the plasma radiated power comes from W and to a lesser extent from Ni (∼10%). Moreover, it was found that an earlier transition from small edge localized modes (ELMs) to ELM-free phase can result in the earlier increase in impurities.

Tokamak a configuration variable (TCV), recently celebrating 30 years of near-continual operation, continues in its missions to advance outstanding key physics and operational scenario issues for ITER and the design of future power plants such as DEMO. The main machine heating systems and operational changes are first described. Then follow five sections: plasma scenarios. ITER Base-Line (IBL) discharges, triangularity studies together with X3 heating and N2 seeding. Edge localised mode suppression, with a high radiation region near the X-point is reported with N-2 injection with and without divertor baffles in a snowflake configuration. Negative triangularity (NT) discharges attained record, albeit transient, beta(N) similar to 3 with lower turbulence, higher low-Z impurity transport, vertical stability and density limits and core transport better than the IBL. Positive triangularity L-Mode linear and saturated ohmic confinement confinement saturation, often-correlated with intrinsic toroidal rotation reversals, was probed for D, H and He working gases. H-mode confinement and pedestal studies were extended to low collisionality with electron cyclotron heating obtaining steady state electron iternal transport barrier with neutral beam heating (NBH), and NBH driven H-mode configurations with off-axis co-electron cyclotron current drive. Fast particle physics. The physics of disruptions, runaway electrons and fast ions (FIs) was developed using near-full current conversion at disruption with recombination thresholds characterised for impurity species (Ne, Ar, Kr). Different flushing gases (D2, H2) and pathways to trigger a benign disruption were explored. The 55 kV NBH II generated a rich Alfvenic spectrum modulating the FI fas ion loss detector signal. NT configurations showed less toroidal Alfven excitation activity preferentially affecting higher FI pitch angles. Scrape-off layer and edge physics. gas puff imaging systems characterised turbulent plasma ejection for several advanced divertor configurations, including NT. Combined diagnostic array divertor state analysis in detachment conditions was compared to modelling revealing an importance for molecular processes. Divertor physics. Internal gas baffles diversified to include shorter/longer structures on the high and/or low field side to probe compressive efficiency. Divertor studies concentrated upon mitigating target power, facilitating detachment and increasing the radiated power fraction employing alternative divertor geometries, optimised X-point radiator regimes and long-legged configurations. Smaller-than-expected improvements with total flux expansion were better modelled when including parallel flows. Peak outer target heat flux reduction was achieved (>50%) for high flux-expansion geometries, maintaining core performance (H-98 > 1). A reduction in target heat loads and facilitated detachment access at lower core densities is reported. Real-time control. TCV's real-time control upgrades employed MIMO gas injector control of stable, robust, partial detachment and plasma beta feedback control avoiding neoclassical tearing modes with plasma confinement changes. Machine-learning enhancements include trajectory tracking disruption proximity and avoidance as well as a first-of-its-kind reinforcement learning-based controller for the plasma equilibrium trained entirely on a free-boundary simulator. Finally, a short description of TCV's immediate future plans will be given.

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.

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

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.

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.

This work presents a study of the interaction between radio frequency (RF) waves used for ion cyclotron resonance heating and the fast deuterium (D) and tritium (T) neutral Beam injected (NBI) ions in DT plasma. The focus is on the effects of this interaction, also referred to as synergistic effects, on the fusion performance in the recent JET DTE2 campaign. Experimental data from dedicated pulses at 3.43 T/2.3 MA heated at (i) 51.4 MHz, giving the central minority H and n = 2 D, and at (ii) 32.2 MHz for the central minority 3 He and n = 2 T. Resonances are analysed and conclusions are drawn and supported by modelling of the synergistic effects. Modelling with transport code TRANSP runs with and without the RF kick operator predict a moderate increase, of about 10%, in DT rates for the case of the RF wave—fast D NBI ion interactions at the n = 2 harmonic of ion cyclotron resonance, and a negligible impact due to synergistic interaction between fast T NBI ions and RF waves. JETTO modelling gives a 29% enhancement in fusion rates due to the interction between RF waves and fast D NBI ions, and an 18% enhancement in fast T NBI ions. Analysis of experimental neutron rates compared to TRANSP predictions without synergistic effects and magnetic proton recoil neutron spectrometer indicate an enhancement of approximately 25%–28% in fusion rates due to RF interaction with fast D ions, and an enhancement of approximately 5%–8% when RF waves and fast T NBI ions are interacting. The contributions of various heating and fast ion sources are assessed and discussed.

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.

Accurate measurements of the ion and electron temperatures in tokamaks are essential for understanding the heating and transport properties of the plasma. While electron temperature measurements are readily available in most devices (e.g. with electron cyclotron emission or Thomson scattering diagnostics), ion temperature estimates usually require more sophisticated procedures such as charge-exchange recombination spectroscopy [1] or crystal spectroscopy [2]. In the absence of such dedicated ion temperature measurements, a technique often adopted in tokamak plasmas is to estimate the ion temperature by reverse engineering of the total neutron yield, assuming Maxwellian distributions for the ions. In the absence of external ion heating, this procedure is satisfactory since it mainly depends on the nuclear reaction cross sections and the bulk ion density, usually inferred from a combination of the electron density and the plasma dilution measurements. On the other hand, when Ion-Cyclotron Resonance Heating (ICRH) or Neutral-Beam Injection (NBI) is applied, the ion distribution functions are distorted and the Maxwellian representation used for the ion temperature estimate is not necessarily justified, typically leading to an overestimate of the thermal component of the ion temperature. By using a coupled wave / Fokker-Planck numerical solver to calculate the actual ion distribution functions under the influence of external heating and the resulting neutron yield, the fast and thermal components of the bulk ion distributions can be disentangled and a ‘thermal-only’ ion temperature consistent with the neutron measurements can be extracted. An example of this procedure for the WEST tokamak with ICRF heating will be presented and the differences between the results obtained with respect to the simplified Maxwellian-based procedure will be discussed.