The experiments carried out in hydrogen at the TOMAS facility show the possibility of controlling plasma parameters such as temperature and electron density in a combined electron cyclotron resonance and radio frequency (ECR+RF) discharge. A maximum plasma density of up to ≈6 × 1016 m−3 and electron temperature of up to 35 eV are observed in the combined ECR+RF discharge. The propagation of RF waves in hydrogen plasma under a weak magnetic field is analyzed. Depending on RF frequency and experimental conditions, such as radial distribution of plasma density and magnetic field, there can be several cases: only the slow wave can propagate, simultaneously slow and fast waves can propagate, or only the fast wave can propagate. The injection of additional RF power into the ECR discharge allows us to change the flux of neutral particles and their distribution function. Even the injection of small RF power of ≈ 0.26 kW relative to microwave power of ≈ 1.7 kW leads to an increase in the hydrogen flux by a factor of ∼2.5. At RF power PRF ≈ 1.57 kW, the H0 flux increases by a factor of ∼9.3. The ability to control the fluxes and energies of particles leaving the plasma volume is important to approach the conditions necessary to study plasma-surface interactions in wall conditioning and fusion edge plasmas.

On the basis of several recent breakthroughs in fusion research, many activities have been launched around the world to develop fusion power plants on the fastest possible time scale. In this context, high-fidelity simulations of the plasma behavior on large supercomputers provide one of the main pathways to accelerating progress by guiding crucial design decisions. When it comes to determining the energy confinement time of a magnetic confinement fusion device, which is a key quantity of interest, gyrokinetic turbulence simulations are considered the approach of choice – but the question, whether they are really able to reliably predict the plasma behavior is still open. The present study addresses this important issue by means of careful comparisons between state-of-the-art gyrokinetic turbulence simulations with the GENE code and experimental observations in the ASDEX Upgrade tokamak for an unprecedented number of simultaneous plasma observables.

To improve the plasma performance and control the density and plasma quality during the flat top phase, wall conditioning techniques are used in large fusion devices like W7-X and in JT60-SA. To study the performance of electron cyclotron wall conditioning, numerous experiments were performed on the TOroidally MAgnetized System, which is operated by LPP-ERM/KMS at the FZ-Jülich. It is a facility designed to study plasma production, wall conditioning, and plasma-surface interactions. The produced electron cyclotron resonance heating plasmas are characterized in various conditions by density and temperature measurements using a movable triple Langmuir probe in the horizontal and the vertical direction, complemented by video and spectroscopic data, to obtain a 2D extrapolation of the plasma parameters in the machine. A way to calibrate the triple Langmuir probe measurements is also investigated. These data can be used to determine the direction of the plasma drift in the vessel and identify the power absorption mechanisms. This will give more insight in the plasma behavior and improve the efficiency of wall conditioning and sample exposure experiments.

A characterization of plasma parameters and neutral particle energies and fluxes has been performed for radio frequency and microwave discharges in the Toroidal Magnetized System (TOMAS). A movable triple Langmuir probe was used to study the electron densities and temperatures, and a time-of-flight neutral particle analyzer was used to measure the energy and fluxes of neutral particles, as a function of the total injected power and the antenna frequency used to generate the plasma. The experimental results can provide information on the behavior of neutral particles at low energies in wall conditioning plasmas.

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm's law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98(p,y) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios ('hybrid' and 'elevated' q-profile) were established on AUG and characterised concerning their plasma performance.

The Joint European Torus (JET) fusion reactor was upgraded to the metallic wall configuration in 2011 which consisted of bulk beryllium (Be) tiles in the main chamber and bulk tungsten (W) and W-coated CFC tiles in the divertor (Matthews G.F. et al 2011 Phys. Scr. T148 014001). During each campaign, a series of wall damages were observed; on the upper dump plates (UDP) positioned to the top part of the vessel walls and on the inner wall—mainly affecting the inner wall guard limiters (IWGL). In both cases, it was concluded that the causes of these damages were unmitigated plasma disruptions. In the case of JET with the metallic wall configuration, most of these plasma disruptions were intentionally provoked. The overall objective was to study the behaviour of these phenomena, in order to assess their impact on the wall, improve understanding of morphological material changes, and—based on that—to develop, implement and test mitigation techniques for their prospective use on ITER. The current results bring additional information on the effects of the unmitigated plasma disruptions on the UDPs and are a significant extension of work presented in (Jepu et al 2019 Nucl. Fusion 59 086009) where the scale of the damage after three operational campaigns on the Be top limiters of JET was highlighted. In addition, new data is presented on the damaging effect that the high energetic runaway electrons had on the Be IWGL in JET.

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).

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.

Efficient Electron Cyclotron Resonance Heating (ECRH) breakdown and pre-ionization can be achieved with fundamental X-mode, while higher harmonics can introduce excessive stray radiation. Fundamental heating however is characterized by a low cut-off density, introducing additional power absorption mechanisms in the plasma. A good knowledge of these mechanisms is necessary to use fundamental X-mode as an efficient pre-ionization method. Numerous experiments were performed on the TOroidally MAgnetized System (TOMAS) to study the power deposition for ECRH in helium. It is a facility designed to study plasma production, wall conditioning and plasma-surface interactions and is operated by LPP-ERM/KMS at the FZ-J & uuml;lich. The influence of the injected power P EC and the magnetic field B 0 on the absorption mechanisms is examined, in order to reduce stray radiation and improve the absorption efficiency. This will allow to determine the best scenarios for plasma start-up and pre-ionization.

A brief overview of ion beam analysis methods and procedures in studies of materials exposed to fusion plasmas in controlled fusion devices with magnetic confinement is presented. The role of accelerator techniques in the examination and testing of materials for fusion applications is emphasised. Quantitative results are based on robust nuclear data sets, i.e. stopping powers and reaction cross-sections. Therefore, the work has three major strands: (i) assessment of fuel inventory and modification of wall materials by erosion and deposition processes; (ii) equipment development to perform cutting-edge research; (iii) determination of nuclear data for selected ion-target combinations. Advantages and limitations of methods are addressed. A note is also given on research facilities with capabilities of handling radioactive and beryllium-contaminated materials.