A central unresolved question in fusion energy research is whether energetic alpha particles, the primary products of deuterium-tritium fusion reactions, enhance or degrade plasma confinement. In burning plasmas, the operating regime of future devices such as ITER and SPARC, alpha particles become the dominant heating source, yet their impact on confinement has remained uncertain. Here, we present self-consistent simulations of burning plasmas that simultaneously evolve microturbulence, alpha-particle heating, and macroscopic plasma profiles to steady state, and find that alpha particles can substantially improve confinement. Fusion-born alpha particles weakly destabilize toroidal Alfven eigenmodes (TAEs), which nonlinearly enhance zonal flows that shear apart and suppress ion-scale turbulence. The resulting reduction in turbulent heat transport drives stronger core profile peaking, increasing alpha heating by up to 25% and establishing a self-reinforcing feedback loop. This mechanism has no direct analogue in present-day experiments, where external heating dominates, and reveals an intrinsic pathway toward improved confinement in burning plasmas.

The codes TORIC and SSFPQL allow detailed selfconsistent simulations of wave propagation and absorption in the Ion Cyclotron (IC) range of frequencies in tokamak plasmas. We review the options presently available in the codes, and we illustrate their performances with two examples: a typical minority heating experiment in a medium size device, and a possible IC heating scenario in a ITER like plasma. The problem of ensuring charge neutrality when the ion distribution functions develop anisotropic suprathermal tails is discussed.

We present fast and accurate surrogate models that predict ion cyclotron resonance heating (ICRH) of a hydrogen minority in deuterium plasmas at ASDEX Upgrade (AUG) that can be used in fast transport modeling. Especially for medium-sized tokamaks with high power densities, such as AUG and SPARC, reliable ICRH modeling with full-wave codes like TORIC requires the coupling of a Fokker-Planck (FP) solver like SSFPQL, which is indispensable to account for the formation of a high-energy ion tail that can drastically alter the absorption behavior. For the first time, our surrogates reproduce the simulations provided by a coupled full-wave code TORIC, combined with a FP solver SSFPQL, demonstrating quantitative agreement not only on synthetic data but also on real AUG discharges that were entirely unseen during training. The models, implemented as feed-forward neural networks, capture collisional heating profiles for electrons and deuterium generated by the TORIC-SSFPQL codes, and reduce inference times from minutes to the order of 40 mu s. To this end, we developed a complete machine learning pipeline that includes volumetric weighting of the radial heating profiles to emphasize physically relevant plasma regions during training. Finally, we integrate our Python-trained models into a C++ and Fortran-compatible Open Neural Network Exchange framework, demonstrating their suitability for cross-platform, real-time deployment in integrated modeling workflows.

Abstract This work presents a stepwise validation of the evolution of the radial electric field E r and transport during the pre L–H transition phase in the ASDEX Upgrade (AUG) tokamak using edge and scrape-off layer full- f gyrokinetic simulations including X-point geometry. Several L-mode time slices up to the L–H transition from a dedicated hydrogen discharge, featuring stepwise increases in electron cyclotron resonance heating input power, are selected (Bonanomi et al 2024 Phys. Plasmas 31 072302) and simulated with the GENE-X code. As the edge boundary conditions are progressively increased between the time slices, particle and heat fluxes rise, and the radial electric field E r well deepens. A detailed validation of the E r profiles and of the E r well depth shows excellent agreement with experimental measurements at the successive time slices approaching the L–H transition. A force balance decomposition identifies turbulence-driven poloidal flows as the dominant contribution within the E r well. Edge turbulence is governed by a competition between electron drift waves and trapped-electron modes. The introduction of an edge density source, modeling neutral gas ionization, is shown to be essential to reproduce experimentally relevant density profiles, E r , and edge ion heat fluxes, which are dominated by both turbulent and diamagnetic contributions. This stepwise validation constitutes an important milestone toward predictive, first-principles gyrokinetic simulations of the L–H transition power threshold.

This paper complements the study reported in G. Urbanczyk et al 2025 Nucl. Fusion 65 046018 by investigating how near-field effects from Ion Cyclotron Range of Frequency (ICRF) antennas depend on the magnetic geometry, particularly focusing on the sensitivity of rectified sheath potentials and associated impurity production to variations in magnetic field tilt-angle. Simulations were conducted using Petra-M, employing a flat model of the ASDEX Upgrade (AUG) three-strap antenna, and COMSOL, utilizing a curved model of the WEST antenna. Results demonstrate indications of sensitivity of the rectified sheath potentials and impurity sputtering to the alignment of Faraday screen bars with the magnetic field lines, emphasizing the necessity for careful consideration of geometric alignment in antenna design for fusion devices. This trend is consistent with experimental data measured with various diagnostics.

This paper discusses the integration of the latest release of the FELICE code (2023) with the TOPICA code. A comparison is made between the current and previous versions of FELICE, starting from the plasma impedance matrix itself and, eventually, getting to the standard TOPICA outputs as the antenna input parameters, the coupled power and the radiated fields. The ADEX Upgrade (AUG) 3-strap antenna is adopted for this comparison, showing very similar results for the two versions of FELICE.

This work introduces a novel 1D hybrid modeling approach to investigate how density fluctuations localized in the edge and Scrape-Off Layer (SOL) regions of tokamaks affect ion-cyclotron wave propagation and absorption in a relevant Divertor Tokamak Test (DTT) plasma scenario. The wave-power transmission through the edge is obtained using the Invariant Imbedding Method in the fluctuating layer, while the absorbed power fractions in the smooth core are evaluated with finite-element and Wentzel–Kramers–Brillouin based approaches. Time-independent and spatially correlated density fluctuations are generated analytically via a correlation-matrix method with prescribed root-mean-square amplitude and correlation length, and are superposed on an analytical equilibrium profile. A Monte Carlo approach is then used to sample many realizations and perform statistical analysis. The resulting edge power-transmission distribution exhibits a clearly multimodal structure, with peaks at high, low, and intermediate transmission levels. The model is fully implemented in Python and is computationally lightweight, making it suitable for fast exploratory studies without the need for high-performance computing resources, which are typical of high-fidelity modeling.

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.

A scenario of ion cyclotron range of frequency (ICRF) wave injection from a top launcher is proposed as an efficient and direct heating method for thermal deuterium ions in deuterium–tritium tokamak plasmas. Positioned between the tritium cyclotron layer and ion–ion hybrid layer, the top launcher allows effective wave penetration to the ion–ion hybrid layer and enables significant power transfer to thermal deuterium. This is achieved through favorable wave polarization for fundamental cyclotron damping. There is a Doppler broadening around the cyclotron resonance and this overlaps with the ion–ion hybrid layer. Low toroidal mode numbers and ion temperature in the range of 5–20 keV are favorable for enhancing the main ion damping relative to electron damping. In contrast to the neutral beam injection, which penetration strongly depends on machine size and plasma density, the proposed ICRF-based direct ion heating scenario is shown to be scalable and applicable to both larger and smaller tokamak devices within practical constraints.

Advanced Tokamak (AT) scenarios are attractive candidates for future nuclear fusion power plants. These scenarios often feature peaked temperature profiles, suggesting a local reduction of turbulent transport. The mechanisms behind this are, as yet, not fully understood. Parameters that are thought to be connected to these transport reductions include the q-profile and the E x B-shear omega(ExB). Another parameter considered to be important is the fast ion content, which can reduce transport in multiple different ways. This work presents AT experiments performed at the ASDEX Upgrade tokamak (AUG) in which these three parameters are varied to investigate their respective effect on the observed peaked ion temperature profiles. To disentangle potentially competing effects, simulations using the codes GENE and TGLF have been performed. It is found that the E x B-shear does not play a role in the AT scenarios performed at AUG, which have relatively low rotation, with v(tor)=100-250km h(-1). Experimentally, a significant dependence of R/L-Ti on the electron cyclotron current drive settings was found, indicating that either the current profile (and, therefore, the q-profile) or T-e/T-i has a significant impact on the observed suppression of turbulent transport. In support of the first option, both GENE and TGLF show a q-profile dependence of R/L-Ti. Furthermore, according to GENE, electromagnetic (EM) fast ion effects are essential in reproducing the experimental results. Scans with TGLF (which does not include such effects), suggest that these fast ion effects become relevant above a threshold of R/L-Ti= 4-5. In the presence of ICRF, additional electrostatic (ES) fast ion effects seem to come into effect, according to GENE, albeit to a smaller degree than the EM effects.

The current response of a hot magnetized plasma to a radio-frequency wave is non-local, turning the electromagnetic wave equation into an integro-differential equation. Non-local physics gives rise to wave physics and absorption processes not observed in local media. Furthermore, non-local physics alters wave propagation and absorption properties of the plasma. In this work, an iterative method that accounts for parallel non-local effects in 2D axisymmetric tokamak plasmas is developed, implemented, and verified. The iterative method is based on the finite element method and Fourier decomposition, with the advantage that this numerical scheme can describe non-local effects while using a high-fidelity antenna and wall representation, as well as limiting memory usage. The proposed method is implemented in the existing full wave solver FEMIC and applied to a minority heating scenario in ITER to quantify how parallel non-local physics affect wave propagation and dissipation in the ion cyclotron range of frequencies (ICRF). The effects are then compared to a reduced local plane wave model, both verifying the physics implemented in the model, as well as estimating how well a local plane wave approximation performs in scenarios with high single pass damping. Finally, the new version of FEMIC is benchmarked against the ICRF code TORIC.

The Doppler-shifted resonance condition for high frequency Alfv & eacute;nic eigenmodes has been extensively studied on ASDEX Upgrade in the presence of one or a combination of two neutral beam injected (NBI) fast ion populations. In general, only centrally deposited NBI sources drive these modes, while off-axis sources globally stabilize the mode activity. For the case of a single central NBI source, the observed trend is: the highest frequency modes are driven by the lowest energy and lowest pitch angle NBI sources, in line with the expectation from the Doppler-shifted resonance condition. The expected mode frequencies are determined analytically from the two-fluid cold plasma dispersion relation and the most unstable frequency relation, while the mode growth rates are estimated using the fast ion slowing down distribution functions from the ASCOT code. The overall mode frequency trend in a source-to-source variation is tracked, although a systematic overestimate of similar to 1 MHz is observed. Possible causes of this overestimate include the finite size of the resonant fast ion drift orbit and non-linear effects such as mode sideband formation. Alternatively, the expected mode frequencies are determined by tracking the growth rate maxima trajectories, this method improves the agreement with the experimentally measured values. A combination of two central mode-driving NBI sources results in the suppression of the mode driven by the lowest energy and the lowest pitch angle NBI source. Computing the analytically expected mode frequency following the method outlined above, again, generally tracks the experimentally observed trend. The mode's Alfv & eacute;nic nature allows for a practical application to track the core hydrogen fraction by following the mode frequency changes in response to a varying ion mass density. Such application is demonstrated in a discharge where the average ion mass is varied from similar to 2m(p) to similar to 1.5m(p) (where m(p) is the proton mass) via a hydrogen puff in a deuterium plasma, in the presence of a strong mode activity. The expected mode frequency changes are computed from the existence of the resonance condition, and the values track the measured results with an offset of similar to 0.5 MHz. Overall, the results suggest an intriguing possibility to monitor and control the D-T ion fraction in the core of a fusion reactor in real time using a non-invasive diagnostic.

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).

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.

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.

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.

Combining the analysis of multiple diagnostics and well-chosen prior information in the framework of Bayesian probability theory, the Integrated Data Analysis code (IDA Fischer et al 2010 Fusion Sci. Technol. 58 675–84) can provide density and temperature radial profiles of fusion plasmas. These IDA-fitted measurements are then used for further analysis, such as discharge simulations and other experimental data analysis. Since IDA considers measurement data, which is frequently fragmentary, with statistical and systematic uncertainties, which are often difficult to quantify, from a heterogeneous set of diagnostics, the fitted profiles and their gradients may be in contradiction to well-established expectations from transport theory. Using the modeling suite ASTRA coupled with the quasi-linear transport solver TGLF, we have created a loop in which simulated profiles and their uncertainties are fed back into IDA as an additional prior, thus providing constraints about the physically reasonable parameter space. We apply this physics-motivated prior to several different plasma scenarios and find improved heat flux match, while still matching the experimental data. This work feeds into a broader effort to make IDA more robust against measurement uncertainties or lack of measurements by combining multiple transport solvers with different levels of complexity and computing costs in a multi-fidelity approach.

Experimental data from the 2022 ASDEX Upgrade (AUG) helium (He) campaign reveals that a lower tungsten (W) content can be achieved during operation of the 3-strap antennas with the W-coated limiters. By fine-tuning of electrical feeding of the 3-strap antennas, the core W content was reduced by about a factor of 2 compared to that during operation of the 2-strap antennas with boron coated limiters at the same power and target plasma. This is in contrast to what is observed in deuterium (D) plasmas, where both types of antennas perform similarly in terms of the W release. A higher significance of remote W sources in He than in D is one plausible explanation.Special AUG experiments in D plasmas, with a density in the SOL reduced to the level relevant for some of the predicted profiles for ITER, show similar characteristics of the near-fields and the sputtering even when the slow wave is propagative. Nevertheless, it is advised for ITER to tailor the SOL profiles by the foreseen local gas injection close to the antenna, to avoid the conditions when the lower hybrid resonance position is approximately aligned with the position the leading edges of the plasma facing components.Based on near-field calculations for the AUG 3-strap and the ITER ICRF antennas using the HFSS code under the same conditions, the experimentally validated RF-sheath rectified voltages of the AUG 3-strap antenna were scaled to the ITER ICRF antenna. Using conservative assumptions, the estimations of the ICRF-specific W sources for ITER were then made for the D-T case with neon seeding. The extrapolation shows that by carefully choosing the electrical feeding, the increment of the W sputtering rate during ITER ICRF operation at up to 20 MW, can be kept (in the worst case) below 10 % of the total W wall rate without ICRF, as well as below 25 % of the W rate at the outboard wall without ICRF.

In present-day fusion devices, central wave heating is crucial to avoid core tungsten (W) accumulation. In this work, we present an integrated modelling framework that reproduces the reduction of core W peaking in ASDEX Upgrade experiments when multiple transport channels are self-consistently evolved, emphasizing the effects of wave heating on turbulent and neoclassical W transport. Predictions for the ITER 15 MA baseline are then provided. We show that the core of a reactor is in a different transport regime for W as compared to present-day tokamaks. The challenges introduced by W arise from global radiation losses that can hinder operation in H-mode, instead of local central accumulation.

In this note we describe how Large Larmor Radius corrections can be incorporate in TORIC and other codes which solve the Finite Larmor Radius wave equations in toroidal axisymmetric geometry.