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 excitation of modes in the JET tokamak in the sub-cyclotronic range of frequencies (frequencies comprised between the Alfvén frequency and the cyclotron frequency) is for the first time reported. The modes were identified as compressional Alfvén eigenmodes and have characteristics similar to those of the sub-cyclotronic modes observed in other tokamaks, in particular those first reported in the NSTX tokamak. On the other hand, the modes observed in JET present some unique features and were observed to be excited by ion cyclotron resonance heating instead of by the injection of beams (neutral beam injection).

The JT-60SA device offers unique conditions before ITER for the study of the interaction of energetic particles with plasma waves. With similar dimensions to JET, e.g., a major radius but with a slightly more elongated plasma volume, JT-60SA is used as a high-power device where additional heating power (including 10 MW of the 500 keV Neutral Beam Injection) of up to 41 MW and the potential for high non-inductive plasma current operation pave the path for numerous challenges in physics on MHD stability, in particular, when considering the effects of energetic particles. Several operational scenarios with ITER and DEMO-relevant plasma regimes, in terms of non-dimensional plasma parameters, are anticipated. In this work, the stability of Alfvén eigenmodes (AEs) in variants of two of the most relevant operational scenarios with single null is analyzed: a full Ip inductive scenario at high density (1.1 × 1020 m−3 on-axis electron density) and 5.48MA/2.05T toroidal plasma current and magnetic field, and an advanced (hybrid) scenario with an ion energy transport barrier (ITB) and 3.5MA/2.28T toroidal plasma current and magnetic field. The workflow included the CRONOS code to establish the scenario, the ASCOT code to calculate the slowing-down energetic particle distributions for a positive/negative ion source-based neutral beam, and the MISHKA/CASTOR-K suite to calculate the MHD spectra of AEs and the associated drive/damping contributions from the NBI energetic ions, as well as the thermal ion landau damping. The systematic analysis, over a large Fourier space of the toroidal mode number/mode frequency, provides evidence that although a significant fraction of supra-Alfvénic particles stemming from the negative ion source-based neutral beam (500 keV) can, in some cases, drive to AEs in both scenarios, it is not enough to overcome the thermal ion landau damping. In addition, the advanced scenario with ITB is shown to be stable against AEs localized in the vicinity of the barrier as well, offering good prospects of sustainability of the plasma performance and of ITB. Finally, some sensitivity scan results are shown on the influence of fast ion density and q-profile on the AE mode spectra and stability.

A toroidal Alfvén eigenmode (TAE) has been observed to be driven by alpha particles in a JET deuterium-tritium internal transport barrier plasma. The observation occurred 50 ms after the removal of neutral beam heating (NBI). The mode is observed on magnetics, soft-xray, interferometry and reflectometry measurements. We present detailed stability calculations using a similar tool set validated during deuterium only discharges. These calculations strongly support the conclusion that the observed mode is a TAE, and that this mode was destabilized by alpha particles. Non-ideal effects from the bulk plasma are interpreted as responsible for suppressing the majority of TAEs which were also driven by alpha particles, but the modes that match the observations are predicted to be particularly weak for these non-ideal effects. This mode located far from the core on the outboard midplane is found to be driven by both trapped and passing particles despite alpha particles originating in the core.

This paper is a written summary of an overview oral presentation given at the 1st Spanish Fusion High Performance Computer (HPC) Workshop that took place on the 27 November 2020 as an online event. Given that over the next few years ITER 24 24 ITER (‘The Way’ in Latin) is the world’s largest tokamak under construction in the south of France: a magnetic fusion device that has been designed to prove the feasibility of fusion as a large-scale and carbon-free source of energy (https://www.iter.org/). will move to its operation phase and the European-DEMO design will be significantly advanced, the EUROfusion consortium has initiated a coordination effort in theory and advanced simulation to address some of the challenges of the fusion research in Horizon EUROPE (2021–2027), i.e. the next EU Framework Programme for Research and Technological Development. This initiative has been called E-TASC, which stands for EUROfusion-Theory and Advanced Simulation Coordination. The general and guiding principles of E-TASC are summarized in this paper. In addition, an overview of the scientific results obtained in the pilot phase (2019–2020) of E-TASC are provided while highlighting the importance of the required progress in computational methods and HPC techniques. In the initial phase, five pilot theory and simulation tasks were initiated: towards a validated predictive capability of the low to high transition and pedestal physics; runaway electrons in tokamak disruptions in the presence of massive material injection; fast code for the calculation of neoclassical toroidal viscosity in stellarators and tokamaks; development of a neutral gas kinetics modular code; European edge and boundary code for reactor-relevant devices. In this paper, we report on recent progress made by each of these projects. ITER (‘The Way’ in Latin) is the world’s largest tokamak under construction in the south of France: a magnetic fusion device that has been designed to prove the feasibility of fusion as a large-scale and carbon-free source of energy (https://www.iter.org/).

In this work, we use reduced and perturbative models to examine the stability of toroidal Alfvén eigenmodes (TAEs) during the internal transport barrier (ITB) afterglow in JET experiments designed for the observation of alpha driven TAEs. We demonstrate that in JET-like conditions, it is sufficient to use an incompressible cold plasma model for the TAE to reproduce the experimental adiabatic features such as frequency and position. When ion cyclotron resonant heating (ICRH) is used to destabilize TAEs, the core-localised modes that are predicted to be most strongly driven by minority ICRH fast ions correspond to the modes observed in the DD experiments, and conversely, modes that are predicted to not be driven are not observed. Linear damping rates due to a variety of mechanisms acting during the afterglow are calculated, with important contributions coming from the neutral beam and radiative damping. For DT equivalent extrapolations of discharges without ICRH heating, we find that for the majority of modes, alpha drive is not sufficient to overcome radiative damping.

The collisional broadening of resonance lines is a universal phenomenon in physics. For example, in atomic physics, collisions lead to abrupt changes in the phase and plane of vibration, thereby destroying phase coherence and leading to uncertainty in the associated photon energy. This leads to broadening of an atom emission/absorption profile [1,2]. In plasma physics, decoherence of the orbital motion of resonant particles allows the reduction of reversible equations of motion into a diffusive system of equations that governs the resonant particle dynamics without detailed tracking of the ballistic motion—as is the case in the widely used quasilinear (QL) kinetic plasma formulation. This work shows how to calculate, from first principles, the resonance function i.e., the envelope function that weights the strength of the resonant interaction, in QL theory in the presence of collisions. It is demonstrated that a quasilinear system that employs the calculated broadening functions (Fig. 1a) recently reported [3] automatically recovers the nonlinear growth rate and mode saturation levels for near-threshold plasmas previously calculated from a significantly more complex nonlinear kinetic theory based on solving a time delayed integro-differential equation [4]. Moreover, it is shown that a QL theory can be constructed for a single resonance, provided that it experiences enough background stochasticity. The results enable realistic reduced modeling of diffusive transport observed in fusion devices. A systematic methodology for establishing a transport theory is developed from first principles, which can be of interest over a range of disciplines that employ kinetic theory. The present formulation gives simple analytical answers to the shape of discrete-resonance collisional functions in quasilinear theory, for both Krook and Fokker-Planck scattering collisions. These functions, that integrate to unity, replace a simple delta function that appears in the diffusion coefficient for the case of no broadening. It is shown that the knowledge of these functions removes a major arbitrariness with respect to previous resonance broadening approaches, which consisted of tuning broadening parameters to match the expected saturation levels [5]. In addition, analytical forms for the modification of the distribution function (shown in Fig. 1b) can be naturally constructed. These forms are particularly useful for code verification akin to studies reported in [6]. This formulation is required to model and verify the fast ion relaxation in tokamaks upon interaction with Alfvénic eigenmodes being implemented in the Resonance Broadened Quasilinear (RBQ) code [7], where particle diffusion occurs along the canonical toroidal momentum and the energy of the beam ion distribution function.

Transport modelling of Joint European Torus (JET) dimensionless collisionality scaling experiments in various operational scenarios is presented. Interpretative simulations at a fixed radial position are combined with predictive JETTO simulations of temperatures and densities, using the TGLF transport model. The model includes electromagnetic effects and collisions as well as E ⃗ × B ⃗ shear in Miller geometry. Focus is on particle transport and the role of the neutral beam injection (NBI) particle source for the density peaking. The experimental 3-point collisionality scans include L-mode, and H-mode (D and H and higher beta D plasma) plasmas in a total of 12 discharges. Experimental results presented in (Tala et al 2017 44th EPS Conf. ) indicate that for the H-mode scans, the NBI particle source plays an important role for the density peaking, whereas for the L-mode scan, the influence of the particle source is small. In general, both the interpretative and predictive transport simulations support the experimental conclusions on the role of the NBI particle source for the 12 JET discharges.

Fusion performance in tokamaks hinges critically on the efficacy of the edge transport barrier (ETB) in suppressing energy losses. The new concept of 'fingerprints' is introduced to identify the instabilities that cause transport losses in the ETBs of many of today's experiments, from among widely posited candidates. Analysis of the gyrokinetic-Maxwell equations and gyrokinetic simulations of experiments reveals that each mode type produces characteristic ratios of transport in the various channels: density, heat, and impurities. This, together with experimental observations of transport in some channel or of the relative size of the driving sources of channels, can identify or determine the dominant modes causing energy transport. In multiple H-mode cases with edge-localized modes that are examined, these fingerprints indicate that magnetohydrodynamic (MHD)-like modes are apparently not the dominant agent of energy transport; rather, this role is played by micro-tearing modes (MTMs) and electron temperature gradient (ETG) modes, and in addition, possibly by ion temperature gradient/ trapped electron modes (ITG/TEM) on JET (Joint European 'Torus). MHD-like modes may dominate the electron particle losses. Fluctuation frequency can also be an important means of identification, and is often closely related to the transport fingerprint. The analytical arguments unify and explain previously disparate experimental observations on multiple devices, including DIII-D, JET, and ASDEX-U. Detailed simulations of two DIII-D ETBs also demonstrate and corroborate this.

In this paper, we analyze heat transport in the JET tokamak using data from its high resolution ECE diagnostic and analyses based on the transfer entropy (TE). The analysis reveals that heat transport is not smooth and continuous, but is characterized by 'trapping regions' separated by `minor transport barriers'. Meat may 'jump over' these barriers and when the heating power is raised, this 'jumping' behavior becomes more prominent. To check that our results are relevant for global heat transport, we deduced an effective diffusion coefficient from the TE results. Both its value and overall radial variation are consistent with heat diffusivities reported in literature. The detailed radial structure of the effective diffusion coefficient was shown to be linked to the mentioned minor transport barriers.

In this study, beryllium tiles from Joint European Torus (JET) vacuum vessel wall were analysed and compared regarding their position in the vacuum vessel and differences in the exploitation conditions during two campaigns of ITER-Like-Wall (ILW) in 2011-2012 (ILW1) and 2013-2014 (ILW2) Tritium content in beryllium samples were assessed. Two methods were used to measure tritium content in the samples - dissolution under controlled conditions and tritium thermal desorption. Prior to desorption and dissolution experiments, scanning electron microscopy and energy dispersive x-ray spectroscopy were used to study structure and chemical composition of plasma-facing-surfaces of the beryllium samples. Experimental results revealed that tritium content in the samples is in range of 2.10(11)-2.10(13) tritium atoms per square centimetre of the surface area with its highest content in the samples from the outer wall of the vacuum vessel (up to 1.9.10(13) atoms/cm(2) in ILW1 campaign and 2.4.10(13) atoms/cm(2) in ILW2). The lowest content of tritium was found in the upper part of the vacuum vessel (2.0.10(12) atoms/cm(2) and 2.0.10(11) atoms/cm(2) in ILW1 and ILW2, respectively). Results obtained from scanning electron microscopy has shown that surface morphology is different within single tile, however if to compare two campaigns main tendencies remains similar.

Results from two different sets of JET experiments are presented.In experiments in which toroidicity-induced Alfvén eigenmodes (TAEs) localized at different radial locations had the same frequencies and toroidal mode numbers, the occurrence of enhanced losses after the excitation of TAEs in the core of the plasma was observed.On the contrary, enhanced losses were not observed if the TAEs localized at different radial locations had different frequencies and toroidal mode numbers.Numerical modeling indicates that, in the first set of experiments, the enhanced losses were caused by a combined effect of the TAEs localized at different radial locations.The TAEs localized in the plasma core transported energetic ions from the core to outer regions of the plasma.Then, the TAEs localized in outer regions of the plasma interacted with these ions just transported by the core-localized TAEs causing a further radial displacement of the ions to the plasma edge.This process eventually ends up causing the loss of the resonant ions.In the second set of experiments, it was found that TAEs localized in the plasma core and in outer regions did not interact with the same ions and so no enhanced losses were measured.Sheared profiles of the safety factor combined with flat mass density profiles lead to larger differences on the frequencies of the TAEs localized at different radial locations, eventually avoiding loss of energetic ions through the described mechanism.

The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994-5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing C-13-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project.

Data on erosion and melting of beryllium upper limiter tiles, so-called dump plates (DP), are presented for all three campaigns in the JET tokamak with the ITER-like wall. High-resolution images of the upper wall of JET show clear signs of flash melting on the ridge of the roofshaped tiles. The melt layers move in the poloidal direction from the inboard to the outboard tile, ending on the last DP tile with an upward going waterfall-like melt structure. Melting was caused mainly by unmitigated plasma disruptions. During three ILW campaigns, around 15% of all 12376 plasma pulses were catalogued as disruptions. Thermocouple data from the upper dump plates tiles showed a reduction in energy delivered by disruptions with fewer extreme events in the third campaign, ILW-3, in comparison to ILW-1 and ILW-2. The total Be erosion assessed via precision weighing of tiles retrieved from JET during shutdowns indicated the increasing mass loss across campaigns of up to 0.6 g from a single tile. The mass of splashed melted Be on the upper walls was also estimated using the high-resolution images of wall components taken after each campaign. The results agree with the total material loss estimated by tile weighing (similar to 130 g). Morphological and structural analysis performed on Be melt layers revealed a multilayer structure of re-solidified material composed mainly of Be and BeO with some heavy metal impurities Ni, Fe, W. IBA analysis performed across the affected tile ridge in both poloidal and toroidal direction revealed a low D concentration, in the range 1-4 x 10(17) D atoms cm(-2).

EDGE2D-EIRENE (the ‘code’) simulations show that radial electric field, Er, in the near scrape-off layer (SOL) of tokamaks can have large variations leading to a strong local E × B shear greatly exceeding that in the core region. This was pointed out in simulations of JET plasmas with varying divertor geometry, where the magnetic configuration with larger predicted near SOL Er was found to have lower H-mode power threshold, suggesting that turbulence suppression in the SOL by local E × B shear can be a player in the L–H transition physics (Delabie et al 2015 42nd EPS Conf. on Plasma Physics (Lisbon, Portugal, 22–26 June 2015) paper O3.113 (http://ocs.ciemat.es/EPS2015PAP/pdf/O3.113.pdf), Chankin et al 2017 Nucl. Mater. Energy 12 273). Further code modeling of JET plasmas by changing hydrogen isotopes (H–D–T) showed that the magnitude of the near SOL Er is lower in H cases in which the H-mode threshold power is higher (Chankin et al 2017 Plasma Phys. Control. Fusion 59 045012). From the experiment it is also known that hydrogen plasmas have poorer particle and energy confinement than deuterium plasmas, consistent with the code simulation results showing larger particle diffusion coefficients at the plasma edge, including SOL, in hydrogen plasmas (Maggi et al 2018 Plasma Phys. Control. Fusion 60 014045). All these experimental observations and code results support the hypothesis that the near SOL E × B shear can have an impact on the plasma confinement. The present work analyzes neutral ionization patterns of JET plasmas with different hydrogen isotopes in L-mode cases with fixed input power and gas puffing rate, and its impact on target electron temperature, Te, and SOL Er. The possibility of a self-feeding mechanism for the increase in the SOL Er via the interplay between poloidal E × B drift and target Te is discussed. It is also shown that reducing anomalous turbulent transport coefficients, particle diffusion and electron and ion heat conductivities, leads to higher peak target Te and larger Er, suggesting the possibility of a positive feedback loop, under an implicitly made assumption that the E × B shear in the SOL is capable of suppressing turbulence.

Beryllium limiter marker tiles were exposed to plasma in the Joint European Torus to diagnose the erosion of main chamber wall materials. A limiter marker tile consists of a beryllium coating layer (7-9 mu m) on the top of bulk beryllium, with a nickel interlayer (2-3 mu m) between them. The thickness variation of the beryllium coating layer, after exposure to plasma, could indicate the erosion measured by ion beam analysis with backscattering spectrometry. However, interpretations from broad beam backscattering spectra were limited by the non-uniform surface structures. Therefore, micro-ion beam analysis (mu-IBA) with 3 MeV proton beam for Elastic back scattering spectrometry (EBS) and PIXE was used to scan samples. The spot size was in the range of 3-10 mu m. Scanned areas were analysed with scanning electron microscopy (SEM) as well. Combining results from mu-IBA and SEM, we obtained local spectra from carefully chosen areas on which the surface structures were relatively uniform. Local spectra suggested that the scanned area (approximate to 600 mu m x 1200 mu m) contained regions with serious erosion with only 2-3 mu m coating beryllium left, regions with intact marker tile, and droplets with 90% beryllium. The nonuniform erosion, droplets mainly formed by beryllium, and the possible mixture of beryllium and nickel were the major reasons that confused interpretation from broad beam EBS.

The impurity concentration in the tokamak divertor plasma is a necessary input for predictive scaling of divertor detachment, however direct measurements from existing tokamaks in different divertor plasma conditions are limited. To address this, we have applied a recently developed spectroscopic N II line ratio technique for measuring the N concentration in the divertor to a range of H-mode and L-mode plasma from the ASDEX Upgrade and JET tokamaks, respectively. The results from both devices show that as the power crossing the separatrix, Psep, is increased under otherwise similar core conditions (e.g. density), a higher N concentration is required to achieve the same detachment state. For example, the N concentrations at the start of detachment increase from ≈ 2% to ≈ 9% as Psep is increased from ≈ 2.5 MW to ≈ 7 MW. These results tentatively agree with scaling law predictions (e.g. Goldston et al.) motivating a further study examining the parameters which affect the N concentration required to reach detachment. Finally, the N concentrations from spectroscopy and the ratio of D and N gas valve fluxes agree within experimental uncertainty only when the vessel surfaces are fully-loaded with N.

Hot spots on the divertor tiles at JET result in overestimation of the tile surface temperature which causes unnecessary termination of pulses. However, the appearance of hot spots can also indicate the condition of the divertor tile surfaces. To analyse the behaviour of the hot spots in the outer divertor tiles of JET, a simple image processing algorithm is developed. The algorithm isolates areas of bright pixels in the camera image and compares them to previously identified hot spots. The activity of the hot spots is then linked to values of other signals and parameters in the same time intervals. The operation of the detection algorithm was studied in a limited pulse range with high hot spot activity on the divertor tiles 5, 6 and 7. This allowed us to optimise the values of the controlling parameters. Then, the wider applicability of the method has been demonstrated by the analysis of the hot spot behaviour in a whole experimental campaign.

The effect of fast ions on turbulent particle transport, driven by ion temperature gradient (ITG)/trapped electron mode turbulence, is studied. Two neutral beam injection (NBI) heated JET discharges in different regimes are analyzed at the radial position ρ t = 0.6, one of them an L-mode and the other one an H-mode discharge. Results obtained from the computationally efficient fluid model EDWM and the gyro-fluid model TGLF are compared to linear and nonlinear gyrokinetic GENE simulations as well as the experimentally obtained density peaking. In these models, the fast ions are treated as a dynamic species with a Maxwellian background distribution. The dependence of the zero particle flux density gradient (peaking factor) on fast ion density, temperature and corresponding gradients, is investigated. The simulations show that the inclusion of a fast ion species has a stabilizing influence on the ITG mode and reduces the peaking of the main ion and electron density profiles in the absence of sources. The models mostly reproduce the experimentally obtained density peaking for the L-mode discharge whereas the H-mode density peaking is significantly underpredicted, indicating the importance of the NBI particle source for the H-mode density profile.

Recent achievements in studies of the effects of ICRF (Ion Cyclotron Range of Frequencies) power on the SOL (Scrape-Off Layer) and PWI (Plasma Wall Interactions) in ASDEX Upgrade (AUG), Alcator C-Mod, and JET-ILW are reviewed. Capabilities to diagnose and model the effect of DC biasing and associated impurity production at active antennas and on magnetic field connections to antennas are described. The experiments show that ICRF near-fields can lead not only to E x B convection, but also to modifications of the SOL density, which for Alcator C-Mod are limited to a narrow region near antenna. On the other hand, the SOL density distribution along with impurity sources can be tailored using local gas injection in AUG and JET-ILW with a positive effect on reduction of impurity sources. The technique of RF image current cancellation at antenna limiters was successfully applied in AUG using the 3-strap AUG antenna and extended to the 4-strap Alcator C-Mod field-aligned antenna. Multiple observations confirmed the reduction of the impact of ICRF on the SOL and on total impurity production when the ratio of the power of the central straps to the total antenna power is in the range 0.6 < P-cen / P-total < 0.8. Near-field calculations indicate that this fairly robust technique can be applied to the ITER ICRF antenna, enabling the mode of operation with reduced PWI. On the contrary, for the A2 antenna in JET-ILW the technique is hindered by RF sheaths excited at the antenna septum. Thus, in order to reduce the effect of ICRF power on PWI in a future fusion reactor, the antenna design has to be optimized along with design of plasmafacing components.