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.

Since the last IAEA-FEC in 2021, significant progress on the development of long pulse steady state scenario and its related key physics and technologies have been achieved, including the reproducible 403 s long-pulse steady-state H-mode plasma with pure radio frequency (RF) power heating. A thousand-second time scale (similar to 1056 s) fully non-inductive plasma with high injected energy up to 1.73 GJ has also been achieved. The EAST operational regime of high beta(P) has been significantly extended (H-98y2 > 1.3, beta(P) similar to 4.0, beta(N) similar to 2.4 and n(e)/n(GW) similar to 1.0) using RF and neutral beam injection (NBI). The full edge localized mode suppression using the n = 4 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with q(95) approximate to 3.1 on EAST, extrapolating favorably to the ITER baseline scenario. The sustained large ELM control and stable partial detachment have been achieved with Ne seeding. The underlying physics of plasma-beta effect for error field penetration, where toroidal effect dominates, is disclosed by comparing the results in cylindrical theory and MARS-Q simulation in EAST. Breakdown and plasma initiation at low toroidal electric fields (<0.3 V m(-1)) with EC pre-ionization is developed. A beneficial role on the lower hybrid wave injection to control the tungsten concentration in the NBI discharge is observed for the first time in EAST suggesting a potential way toward steady-state H-mode NBI operation.

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/).

The European fusion research activities have, over recent decades, generated a vast and varied set of data. The volume and diversity of the data that need to be catalogued and annotated make the task of organising and making the data available within a broader environment very challenging. Nevertheless, there are strong scientific drivers as well as incentives and mandates from national research agencies suggesting that a more coherent approach to data referencing, dissemination and sharing would provide strong benefits to the fusion research community and beyond. Here, we discuss the technical requirements and developments needed to transition the current, and future, range of fusion research data to an open and Findable, Accessible, Interoperable, and Reusable data sharing structure guided by the principle 'as open as possible, as closed as necessary'. Here we propose a set of recommendations and technical implementations needed to form a European data sharing environment for the fusion research programmes. Consistency with the emerging IMAS (ITER Integrated Modelling and Analysis Suite) infrastructure is considered to facilitate future deployments.

Self-consistent full-size turbulent-transport simulations of the divertor and scrape-off-layer (SOL) of existing tokamaks have recently become feasible. This enables the direct comparison of turbulence simulations against experimental measurements. In this work, we perform a series of diverted ohmic L-mode discharges on the tokamak a configuration variable (TCV) tokamak, building a first-of-a-kind dataset for the validation of edge turbulence models. This dataset, referred to as TCV-X21, contains measurements from five diagnostic systems from the outboard midplane (OMP) to the divertor targets-giving a total of 45 one- and two-dimensional comparison observables in two toroidal magnetic field directions. The experimental dataset is used to validate three flux-driven 3D fluid-turbulence models-GBS, GRILLIX and TOKAM3X. With each model, we perform simulations of the TCV-X21 scenario, individually tuning the particle and power source rates to achieve a reasonable match of the upstream separatrix value of density and electron temperature. We find that the simulations match the experimental profiles for most observables at the OMP-both in terms of profile shape and absolute magnitude-while a comparatively poorer agreement is found towards the divertor targets. The match between simulation and experiment is seen to be sensitive to the value of the resistivity, the heat conductivities, the power injection rate and the choice of sheath boundary conditions. Additionally, despite targeting a sheath-limited regime, the discrepancy between simulations and experiment also suggests that the neutral dynamics should be included. The results of this validation show that turbulence models are able to perform simulations of existing devices and achieve reasonable agreement with experimental measurements. Where disagreement is found, the validation helps to identify how the models can be improved. By publicly releasing the experimental dataset and validation analysis, this work should help to guide and accelerate the development of predictive turbulence simulations of the edge and SOL.

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017-2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m(-2) were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

The concept of FAIR data is being increasingly adopted at national and global levels as a way of maximising the impact and transparency of publicly-funded research outcomes and data. In this paper we introduce the FAIR requirements of the Fusion community as well as technological directions proposed by the Fair4Fusion project aiming at increasing the accessibility to fusion data.

The EUROfusion JET-ILW pedestal database is described, with emphasis on three main issues. First, the technical aspects are introduced, including a description of the data selection, the datasets, the diagnostics used, the experimental and theoretical methods implemented and the main definitions. Second, the JET-ILW pedestal structure and stability are described. In particular, the work describes the links between the engineering parameters (power, gas and divertor configuration) and the disagreement with the peeling-ballooning (PB) model implemented with ideal magnetohydrodynamics equations. Specifically, the work clarifies why the JET-ILW pedestal tends to be far from the PB boundary at high gas and high power, showing that a universal threshold in power and gas cannot be found but that the relative shift (the distance between the position of the pedestal density and of the pedestal temperature) plays a key role. These links are then used to achieve an empirical explanation of the behavior of the JET-ILW pedestal pressure with gas, power and divertor configuration. Third, the pedestal database is used to revise the scaling law of the pedestal stored energy. The work shows a reasonable agreement with the earlier Cordey scaling in terms of plasma current and triangularity dependence, but highlights some differences in terms of power and isotope mass dependence.

The next step after ITER is the demonstration of stable electricity production with a fusion reactor. Key design performance will have to be met by the corresponding power plant demonstrator (DEMO), fulfilling a large number of constraints. System codes such as SYCOMORE, by simulating all the fusion power plant sub-systems, address these questions. To be able to perform design optimizations, simplified models relying on physical and technological assumptions have to be used, resulting in a large number of input parameters. As these parameters are not always exactly known, the impact of their associated uncertainties on final design performance has to be evaluated. Sensitivity methods, by measuring the relative influence of inputs on the figures of merit of the design, allow the dominant parameters to be selected. This information helps the search for optimal working points, guides the priority for technical improvements and finally allows meaningful inputs for uncertainty propagation to be selected. A full set of sensitivity methods and their application to an ITER and DEMO design will be presented, discussing both the statistical method behaviours and the physical results. Plasma shape parameters (minor radius and plasma elongations) share half of the net electricity power sensitivity for the DEMO 2015 design, while the toroidal magnetic field and the 95% safety factor are responsible for 23% and 17% of the electric power sensitivity, respectively. The plasma minor radius is responsible for 45% of the pulse duration sensitivity for the DEMO 2015 design, while plasma physics parameters drive % of the pulse duration sensitivity.

Remote experiment with WEST tokamak in CEA Cadarache from ITER Remote Experimentation Centre (REC) in Rokkasho had been successfully carried out under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach (BA) activities in the field of fusion energy research. The WEST and the REC sites were connected with the international broadband network for science and education of SINET, GEANT and RENATER. Due to a long distance over 10,000 km of two sites, the latency was about 200 ms (round trip time) in total via the direct line between EU and Japan of SINET5. Based on the communication between local and remote participants, plasma discharges were successfully executed repeatedly according to the pulse schedule produced by remote Session Leader after the approval by Engineer in Charge on WEST site. Intershot analyses were successfully executed during the pulse interval on an analysis server prepared in Partners network in the WEST for external collaborator. Full function required for the remote experiments were successfully demonstrated in the large size tokamak facility, WEST. For remote participation of ITER, however, it is expected that the amount of experimental data and the number of remote participations will become very large. Distributed database system based on data replication using high-speed data transferring method and distributed computing resources in local and remote sites are proposed as a possible solution in ITER.

The ITER Integrated Modelling and Analysis System (IMAS) has been adopted by the EUROfusion Consortium as a platform to facilitate the analysis and verification of data from multiple tokamaks for the integration of physics codes and the validation of physics models for fusion plasma simulations. Data mapping tools have been developed to translate the tokamaks' native data format into IMAS. The mapping required the adoption of standard coordinates, conventions on direction of vectors, signs of fields, and harmonization of physics units. The mapped data have been verified by running integrated simulations using Kepler workflows. Results of the test using IMAS data are reported here along with an assessment of the system for present and future fusion applications.

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.

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.

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.

Auxiliary heating is essential to initiate fusion in future tokamaks. In particular, ion heating tends to maximise the alpha power generation by increasing the thermal ion temperature. In order to simulate the plasma heating by ion cyclotron radio frequency waves, the EVE code, a full wave code for IC wave propagation, and SPOT, an orbit following Monte Carlo code combined with the RFOF library which calculates the absorption of wave by ions, have been coupled together. This new package is used for simulating JET plasmas with strong interplay between ion cyclotron resonant heating and neutral beam injection. Simulations shows that up to 20% of the neutron rate generated in recent JET D plasmas is due to the synergy between both heating mechanisms. However, the H concentration plays a critical role on such interplay, because the synergy efficiency weakens with the H concentration. Therefore, the control of the H concentration is mandatory for optimising the fusion reaction rate generation at JET.