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.

Nuclear fusion could offer clean, abundant energy. However, managing the power exhausted from the core fusion plasma towards the reactor wall remains a major challenge. This is compounded in emerging compact reactor designs promising more cost-effective pathways towards commercial fusion energy. Alternative Divertor Configurations (ADCs) are a potential solution. In this work, we demonstrate exhaust control in ADCs, employing a novel method to diagnose the neutral gas buffer, which shields the target. Our work on the Mega Ampere Spherical Tokamak Upgrade shows that ADCs tackle key risks and uncertainties for fusion energy. Their highly reduced sensitivity to perturbations enables active exhaust control in otherwise unfeasible situations and facilitates an increased passive absorption of transients, which would otherwise damage the divertor. We observe a strong decoupling of each divertor from other reactor regions, enabling near-independent control of the divertors and core plasma. Our work showcases the real-world benefits of ADCs for effective heat load management in fusion power reactors.

Exhausting power from the hot fusion core to the plasma-facing components is one fusion energy’s biggest challenges. The MAST Upgrade tokamak uniquely integrates strong containment of neutrals within the exhaust area (divertor) with extreme divertor shaping capability. By systematically altering the divertor shape, this study shows the strongest evidence to date to our knowledge that long-legged divertors with a high magnetic field gradient (total flux expansion) deliver key power exhaust benefits without adversely impacting the hot fusion core. These benefits are already achieved with relatively modest geometry adjustments that are more feasible to integrate in reactor designs. Benefits include reduced target heat loads and improved access to, and stability of, a neutral gas buffer that ‘shields’ the target and enhances power exhaust (detachment). Analysis and model comparisons shows these benefits are obtained by combining multiple shaping aspects: long-legged divertors have expanded plasma-neutral interaction volume that drive reductions in particle and power loads, while total flux expansion enhances detachment access and stability. Containing the neutrals in the exhaust area with physical structures further augments these shaping benefits. These results demonstrate strategic variation in the divertor geometry and magnetic topology is a potential solution to one of fusion’s power exhaust challenge. (Figure presented.)

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.

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.

To complement wall conditioning research in TOMAS, a characterisation of radio-frequency hydrogen plasmas has been performed using a radially movable triple Langmuir probe. Experimental measurements of electron temperature and density radial profiles at different magnetic field on axis strengths and neutral pressures have been performed. First results of simulations of the radial profiles with the code TOMATOR-1D can qualitatively reproduce the measurements of the diagnostic and may be used to understand the behaviour of the waves inside the plasma.

To measure the local ion energy distribution functions for support of Ion Cyclotron Wall Conditioning (ICWC) application studies in the TOMAS device, a Semion Retarding Field Energy Analyzer (RFEA) by Impedans Ltd. has been installed in the sample load-lock system. The RFEA system has been commissioned in hydrogen Radio-Frequency (RF) plasmas. The first studies have been performed on the influence of the neutral hydrogen pressure, the input power, the magnetic field and the location on the average ion energy and ion flux density.