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.

ITER will be equipped with two types of diagnostics for fusion power measurement in deuterium-tritium plasmas: neutron flux monitors and neutron cameras, both absolutely counting the 14 MeV neutrons emitted by the D(T, 4He)n reaction. The radial gamma-ray spectrometer (RGRS), though, has recently been considered as an additional fusion power diagnostic, detecting 17 MeV gamma-rays emitted by the radiative D(T, 5He)gamma channel of the DT reaction. This work describes the RGRS performance by determining the upper limit of its operational domain, strictly correlated with the expected background intensity and resulting between 2.6 & sdot;1017 neutrons per second and 3.2 & sdot;1018 neutrons per second. Based on this assessment, design optimization are introduced to fit the RGRS performances to the ITER requirements for safety-relevant fusion power measurements, for which the required operational range is between 2.4 & sdot;1019 neutrons per second and 3.2 & sdot;1020 neutrons per second. The expected different energy ranges covered by signal and background are exploited with the employment of a proper gamma-ray attenuator, allowing to widen and shift the RGRS operational domain toward the most relevant, high power plasma scenarios.

In the initial stages of ITER operation, ELM mitigation systems need to be commissioned. This requires controlled flat-top operation in type-I ELMy H-mode regimes. Hydrogen or helium plasma discharges are used exclusively in these stages to ensure negligible production of neutrons from fusion reactions. With the expected higher L-H power threshold of hydrogen and helium plasmas compared to corresponding D and D/T plasmas, it is uncertain whether available auxiliary power systems are sufficient to operate in stable type-I ELMy H-mode. This has been investigated using integrated core and edge/SOL/divertor modelling with JINTRAC. Assuming that the L-H power threshold is well captured by the Martin08 scaling law, the presented simulations have found that 30 MW of ECRH power is likely required for the investigated hydrogen plasma scenarios, rather than the originally planned 20 MW in the 2016 Staged Approach ITER Baseline. However, past experiments have shown that a small helium fraction (similar to 10%) can considerably reduce the hydrogen plasma L-H power threshold. Assuming that these results extrapolate to ITER operation regimes, the 7.5 MA/2.65 T hydrogen plasma scenario is likely to access stable type-I ELMy H-mode operation also at 20 MW of ECRH.

This paper explores the operational boundaries and power availability of the neutral beam injection (NBI) system in ITER, with a specific focus on shine-through (ST) loss prevention. ST, a phenomenon where part of the injected neutral beam remains un-ionized in the plasma and directly impacts the first wall components, poses a significant risk to the lifetime of ITER's plasma-facing components (PFCs). The operational window for NBI is consequently constrained by these losses, which are influenced by factors such as plasma density, beam energy, and injection geometry. Leveraging advanced numerical simulations, we investigate these dependencies across various ITER plasma scenarios, particularly for the DT-1 phase, which will mark the first NBI operations. In light of recent ITER blanket design changes, our analysis refines previous estimates of the maximum acceptable ST power on PFCs. We then present a new heuristic formula which permits the calculation of the ST fraction and the minimum plasma density that permits ITER NBI operations as a function of global variables. This allows for establishing operational limits for Hydrogen and Deuterium NBI in Hydrogen, Deuterium, and Deuterium-Tritium plasmas. Additionally, we compare commonly used beam ionisation codes for ITER and tokamak simulations, evaluating their reliability in the investigated parameter space. The findings of this study are crucial for ensuring the efficient operation of the NBI system during ITER's experimental phases. They define the conditions under which beam power can be fully utilised without compromising operational lifetime, thereby informing future plasma operation plans and contributing to the success of ITER's scientific objectives.

Gyrokinetic and kinetic-MHD simulations are performed for the fishbone instability in the DIII-D discharge #178631, chosen for validation of first-principles simulations to predict the energetic particle (EP) transport in an ITER prefusion baseline scenario. Fishbone modes are found to generate zonal flows, which dominate the fishbone saturation. The underlying mechanisms of the two-way fishbone-zonal flows nonlinear interplay are discussed in details. Numerical and analytical analyses identify the fishbone-induced EP redistribution as the dominant generation mechanism for zonal flows. The zonal flows modify the nonlinear dynamics of phase space zonal structures, which reduces the amount of EPs able to resonate with the mode, leading to an early fishbone saturation. Simulation results including zonal flows agree quantitatively with DIII-D experimental measurements of the fishbone saturation amplitude and EP transport, supporting this novel saturation mechanism by self-generated zonal flows. Moreover, the wave-particle mode-locking mechanism is shown to determine quantitatively the fishbone frequency down-chirping, as evident in GTC simulation results in agreement with predictions from analytical theory. Finally, the fishbone-induced zonal flows are possibly responsible for the formation of an ion-ITB in the DIII-D discharge. Based on the low EP transport and the large zonal flow shearing rates associated with the fishbone instability in gyrokinetic simulations of the ITER scenario, it is conjectured that high performance scenarios could be designed in ITER burning plasmas through fishbone-induced ITBs.

A wide variety of ITER plasmas are investigated using a recently-developed integrated modelling framework, with particular emphasis on the transport and effects of tungsten (W) motivated by ITER's new full-W walls. This workflow is entirely based on theoretical transport models and it has been recently validated against experimental data. Simulations in L-mode and H-mode are performed at different plasma currents and heating powers to assess the maximum W concentrations that allow access and sustainment of H-mode operation, as well as dynamical simulations of the ramp-up, from the limiter to the diverted phase, finding maximum tolerable W concentrations to avoid a radiative collapse of the plasma. A simple physical parameter is shown to order the ratio of W neoclassical to turbulent transport magnitudes for all considered ITER plasmas as well as previously simulated ASDEX Upgrade experiments.

In magnetic fusion devices, error field (EF) sources, spurious magnetic field perturbations, need to be identified and corrected for safe and stable (disruption-free) tokamak operation. Within Work Package Tokamak Exploitation RT04, a series of studies have been carried out to test the portability of the novel non-disruptive method, designed and tested in DIII-D (Paz-Soldan et al 2022 Nucl. Fusion 62 126007), and to perform an assessment of model-based EF control strategies towards their applicability in ITER. In this paper, the lessons learned, the physical mechanism behind the magnetic island healing, which relies on enhanced viscous torque that acts against the static electro-magnetic torque, and the main control achievements are reported, together with the first design of the asynchronous EF correction current/density controller for ITER.

The ITER Radial Gamma-Ray Spectrometer (RGRS) consists of three gamma-ray detectors observing the plasma through three collimated, coplanar, radial lines of sight (LoS). The system was initially designed to monitor the runaway electron emission and the alpha-particle density profile [Nocente et al., Nucl. Fusion 57, 076016 (2017)]. This work presents a novel technique for measuring the fusion power during D-T operation using the RGRS. This method is based on the absolute measurement of the 17 MeV fusion gamma-rays and a semi-analytical computation of their transport from the plasma source to the detectors. This approach was initially developed and tested at JET during the second D-T campaign (DTE2) on a single LoS diagnostic [Dal Molin et al., Phys. Rev. Lett. (submitted) (2024); Rebai et al., Phys. Rev. C (submitted) (2024); and Marcer et al., Nucl. Fusion (unpublished) (2024)]. This work exploits the multiple LoS of the RGRS to create a combined virtual diagnostic whose detected fraction of the total plasma emission is less affected by variations in the plasma emission profile, reducing systematic uncertainties on the estimated total emission, compared to the individual detectors.

The as-built stiffness in the ITER central solenoid (CS) modules (CSM1 thorough to CSM4 are currently manufactured) determines the range of vertical compression forces that can be tolerated by the CS modules during ITER operation. Since the as-built stiffness of the CS modules manufactured (∼32 GPa and ∼34 GPa for CSM1 and CSM2, respectively and similar for the other modules) has been reduced from the design value (53 GPa), the CS axial (vertical) force criteria have been updated assuming a conservative stiffness (25 GPa) with margins for all six CS modules. Initial analysis using the updated CS force criteria has revealed that this reduction affects only the plasma initiation with fully charged CS in the ITER 15 MA Baseline DT scenario, resulting in a slight reduction of poloidal magnetic flux, from 117.5 Wb to 116.2 Wb at initial CS magnetization. Therefore, the 15 MA Baseline scenario has been re-developed with an updated plasma start-up, and then the entire evolution of the CS and poloidal field coil parameters has been validated against all the coil currents, fields and forces criteria. To explore potential risks and opportunities for further optimization of scenarios, the equilibrium operational space (the plasma internal inductance versus the poloidal magnetic flux produced by the coils) at flat-top burn has been analyzed using the CORSICA and DINA codes. The three major ITER reference DT operation scenarios, 15 MA Q = 10 Baseline, 12.5 MA Q > 5 Hybrid and 10 MA Q ∼ 5 Steady-State, satisfy all the coil criteria including the CS force updated reflecting the as-built stiffness. The evolution of the plasma discharge parameters within the equilibrium operational spaces provided a guidance for potential optimization with margins.

Designing a robust termination scenario for a burning ITER plasma is a challenge that requires extensive core plasma and divertor modelling. The presented work consists of coupled core/edge/SOL/divertor simulations, performed with the JINTRAC code, to study the Q = 10 flat-top phase and exit phase of the ITER 15 MA/5.3 T DT scenario. The modelling utilizes the recently implemented option to treat deuterium and tritium separately in the SOL/divertor, enabling a consistent treatment of deuterium and tritium in the whole plasma volume, which is a unique capability of JINTRAC. In addition, these are the first JINTRAC simulations of this scenario that use a first-principles transport model to self-consistently model the ECRH power deposition and to include tungsten while keeping track of tungsten sputtering and accumulation. The flat-top simulations demonstrate the possibility of sustaining a steady state fusion Q of 10 using pure deuterium gas puffs together with DT mixed pellets, which is an option to make a more effective use of tritium. Simulations of the exit phase are set up sequentially, with each phase providing initial conditions for the next, starting with a density decay at full current and auxiliary power, and demonstrate the possibility of reducing the density robustly within a few seconds. Following the density decay, a subsequent auxiliary power ramp-down in H-mode is performed with a late H-L transition at low auxiliary power, which may provide an option for the optimization of the plasma termination. The final ramp-down phase consists of a current ramp-down in L-mode to 3.75 MA.

Gyrokinetic simulations of the fishbone instability in DIII-D tokamak plasmas find that self-generated zonal flows can dominate the nonlinear saturation by preventing coherent structures from persisting or drifting in the energetic particle phase space when the mode frequency down-chirps. Results from the simulation with zonal flows agree quantitatively, for the first time, with experimental measurements of the fishbone saturation amplitude and energetic particle transport. Moreover, the fishbone-induced zonal flows are likely responsible for the formation of an internal transport barrier that was observed after fishbone bursts in this DIII-D experiment. Finally, gyrokinetic simulations of a related ITER baseline scenario show that the fishbone induces insignificant energetic particle redistribution and may enable high performance scenarios in ITER burning plasma experiments.

Long Pulse Scenarios (LPS) in ITER foreseen during the Pre-Fusion Power Operation (PFPO) phase of the ITER Research Plan (IRP) are assessed using 1.5D transport simulations within the ASTRA framework. Such assessment is required to predict the operational space for LPS operation in PFPO, as well as to evaluate which physics processes for LPS operation during Fusion Power Operation (FPO) could be studied during PFPO. An important aspect in the development of LPSs in PFPO is to minimize lifetime consumption of the Central Solenoid (CS) for these scenarios. The maximum pulse length achievable for LPSs in PFPO with no consumption of CS lifetime (currents in CS coils ⩽30 kA per turn) has been assessed for a range of heating schemes and heating mixes, confinement regimes (L-mode and H-mode) and for helium and hydrogen plasmas. The operational space of LPS and pulse length has been explored through density scans with the Heating and Current Drive mix required for the FPO Q ⩾ 5 steady-state plasma scenario (namely Neutral Beam Injection and Electron Cyclotron Heating) including acceptable shine through losses on the first wall for both helium and hydrogen plasmas. Fast particle physics aspects that are common between FPO plasmas and LPS PFPO H-mode plasmas at low densities are studied including MHD stability analysis with the KINX code and non-perturbative critical gradient model based on high-n Toroidal Alfven Eigenmodes (TAE) stability kinetic ballooning code HINST calculations.

The success of magnetic confinement fusion as an energy source relies crucially on reaching the necessary high temperatures for the fuel D and T ions.In a fusion reactor, plasma heating with waves in the ion cyclotron range of frequencies (ICRF) is the only system capable of providing a large fraction of bulk ion heating.Furthermore, in view of a better understanding of the non-linear physics of alpha heating in ITER and future reactors, generating MeV-range ions and studying the impact of fast ions on plasma stability and confinement becomes progressively more important.This paper summarizes recent theoretical progress and experimental demonstrations of the so-called three-ion ICRF scenario on the tokamaks Alcator C-Mod, ASDEX Upgrade and JET.In particular, the paper highlights several key results relevant for ITER and future fusion reactors.

Pellet injection is the most promising technique to achieve efficient plasma core fuelling, key for attaining stationary scenarios in large magnetic confinement fusion devices. In this paper, the injection of pellets with different volumes and speeds into standard plasma scenarios in ITER (tokamak) and Wendelstein 7-X (stellarator) is studied by modeling the pellet ablation and particle deposition, focusing on the evaluation of the expected differences in pellet plasmoid drifts in tokamaks and stellarators. Since the efficiency of the damping-drift mechanisms is predicted to depend on the magnetic configuration, device-specific characteristics are expected for the temporal evolution of the plasmoid drift acceleration. For instance, plasmoid-internal Pfirsch–Schlüter currents dominate the drift damping process for stellarators, while plasmoid-external currents are more relevant for tokamaks. Also, relatively larger drifts are in principle expected for W7-X due to higher field gradients in relation to machine dimensions. However, shorter plasmoid-internal charge reconnection lengths result in the drift damping due to internal Pfirsch–Schlüter currents being more effective than in a tokamak. Therefore, the average relative drift displacement during the whole plasmoid homogenization may a priori be comparable in both magnetic configurations. Moreover, High Field Side (HFS) injection is expected to be highly advantageous to maximize pellet particle deposition in ITER, whereas it may only be beneficial in medium to high β environments in W7-X. Finally, there may be means for the optimization of pellet injection configurations in both ITER and W7-X for the considered plasma scenarios despite the sizeable differences in the relative importance of the mechanisms of plasmoid drift acceleration and deceleration in play.

The pre–fusion power operation 1 phase of ITER is planned to be characterized by electron cyclotron resonance heating only. Under the assumption that the access to H–mode is determined by a critical ion heat flux at the plasma edge, full–radius ASTRA simulations with the TGLF–SAT2 transport model are performed in order to compute the ion heat flux produced by the thermal exchange between electrons and ions in different operational conditions. Both hydrogen and deuterium plasmas at 5 MA are considered, respectively at 1.8 T and 2.65 T, corresponding to one third and half of the nominal maximum magnetic field. Different levels of electron cyclotron heating power are considered in sets of simulations with increasing values of the electron line averaged density. The predictions are compared with the currently available scaling of the critical ion heat flux. In hydrogen, 20 MW of electron heating power are predicted to allow H–mode access in a vanishingly small density window, whereas 30 MW and 40 MW would allow more substantial H–mode operational windows. Despite the fact that in deuterium plasmas the thermal exchange between electrons and ions is smaller by the hydrogen to deuterium mass ratio compared to hydrogen plasmas, the lower H–mode power threshold in deuterium leads to the prediction that an even broader and more robust domain to access H–mode is obtained at half field at 40 MW in deuterium as compared to operation in hydrogen at one third of the maximum magnetic field, even at the same power.

Edge localized modes (ELMs) suppression by resonant magnetic perturbations (RMPs) was studied with the non-linear magneto-hydro-dynamic (MHD) code JOREK for the ITER H-mode scenarios at 15 MA, 12.5 MA, 10 MA/5.3 T. The main aim of this work was to demonstrate that ELMs can be suppressed by RMPs while the divertor 3D footprints of heat and particle fluxes remain within divertor material limits. The unstable peeling–ballooning modes responsible for ELMs without RMPs were modelled first for each scenario using numerically accessible parameters for ITER. Then the stabilization of ELMs by RMPs was modelled with the same parameters. RMP spectra, optimized by the linear MHD MARS-F code, with main toroidal harmonics N = 2, N = 3, N = 4 have been used as boundary conditions of the computational domain of JOREK, including realistic RMP coils, main plasma, scrape off layer (SOL) divertor and realistic first wall. The model includes all relevant plasma flows: toroidal rotation, two fluid diamagnetic effects and neoclassical poloidal friction. With RMPs, the main toroidal harmonic and the non-linearly coupled harmonics remain dominant at the plasma edge, producing saturated modes and a continuous MHD turbulent transport thereby avoiding ELM crashes in all scenarios considered here. The threshold for ELM suppression was found at a maximum RMP coils current of 45 kAt–60 kAt compared to the coils maximum capability of 90 kAt. In the high beta poloidal steady-state 10 MA/5.3 T scenario, a rotating QH-mode without ELMs was observed even without RMPs. In this scenario with RMPs N = 3, N = 4 at 20 kAt maximum current in RMP coils, similar QH-mode behaviour was observed however with dominant edge harmonic corresponding to the main toroidal number of RMPs. The present MHD modelling was limited in time by few tens of ms after RMPs were switched on until the magnetic energy of the modes saturates. As a consequence the thermal energy was still evolving on this time scale, far from the ITER confinement time scale and hence only the form of 3D footprints on the divertor targets can be indicated within this set-up. Also note, that the divertor physics was missing in this model, so realistic values of fluxes are out of reach in this modelling. However the stationary 3D divertor and particle fluxes could be simply extrapolated from these results to the stationary situation considering that a large power fraction should be radiated in the core and SOL and only about 50 MW power is going to the divertor, which is an arbitrary, but reasonable number used here. The 3D footprints with RMPs show the characteristic splitting with the main RMP toroidal symmetry. The maximum radial extension of the footprints typically was ∼20 cm in inner divertor and ∼40 cm in outer divertor with stationary heat fluxes decreasing further out from the initial strike point from ∼5 MW m−2 to ∼1 MW m−2 assuming a total power in the divertor and walls is 50 MW. The heat fluxes remain within the divertor target and baffle areas, however with rather small margin in the outer divertor which could be an issue for the first wall especially in transient regimes when part of the plasma thermal energy is released due to switching on the RMP coils. This fact should be considered when RMPs are applied with a more favorable application before or soon after the L–H transition, although optimization is required to avoid increasing the L–H power threshold with RMPs.

Both linear and quasi-linear aspects of the plasma response to the resonant magnetic perturbation (RMP) field are numerically investigated for various H-mode scenarios in ITER, covering the pre-fusion power operation and the fusion power operation phases. Linear response computations for eight ITER scenarios, with varying plasma current and toroidal magnetic field, reveal that the best coil current phasing for controlling the type-I edge localized modes (ELMs) scales roughly linearly with the edge safety factor. The coil phasing is defined as the relative toroidal phase of the coil currents between different rows, for a given toroidal harmonic. Quasi-linear initial value simulation, which is the focus of the present study, shows that application of the n = 3 (n is the toroidal mode number) RMP field has a minimum side effect on the plasma core momentum confinement but potentially a large effect on the global particle transport. Generally, the RMP field with the best (worst) coil phasing for ELM control produces the strongest (weakest) effect on the plasma edge flow and the overall density. This robustly holds for all eight ITER scenarios. Consequently, in order to minimize the RMP induced side effects while achieving ELM control (suppression) in ITER, a compromise is necessary in choosing the coil current configuration.

The graphics processing unit (GPU) version of the Lorentz-orbit code for use in stellarators and tokamaks ( LOCUST ) has been applied to study the fast-ion transport and loss caused by resonant magnetic perturbations in the high-performance Q = 10 ITER baseline scenario. The unique computational efficiency of the code is exploited to calculate the impact of the application of ITER's edge-localised mode (ELM) control coil system on neutral beam heating efficiency, as well as producing detailed predictions of the resulting plasma-facing component power loads, for a variety of operational parameters—the applied fundamental toroidal mode number n 0 , mode spectrum and absolute toroidal phase of the imposed perturbation. The feasibility of continually rotating the perturbations is assessed and shown to be effective at reducing the time-averaged power loads. Through careful adjustment of the relative phase of the applied perturbation in the three rows of coils, peak power loads are found to correlate with reductions in neutral beam injection (NBI) heating efficiency for n 0 = 3 fields. Adjusting the phase this way can increase total NBI system efficiency by approximately 2%–3% and reduce peak power loads by up to 0.43 MW m −2 . From the point of view of fast-ion confinement, n 0 = 3 ELM control fields are preferred overall to n 0 = 4 fields. In addition, the implementation of 3D magnetic fields in LOCUST is also verified by comparison with the SPIRAL code for a DIII-D discharge with ITER-similar shaping and n 0 = 3 perturbation.

Effects of an unstable n = 1 (n is the toroidal mode number) resistive wall mode (RWM) on the energetic particle (EP) confinement and loss are numerically investigated, for an ITER steady state scenario with 10 MA plasma current and 5.3 T toroidal field. The eigenfunction of the RWM is computed, with the associated three-dimensional magnetic field perturbation superposed with the 2D equilibrium field for tracing the EP drift orbits. Considered are mono-energetic EPs at 0.5 MeV and 1 MeV for deuterium ions, and 3.5 MeV for fusion-born alphas, with a range of distribution in the particle pitch angle. Modeling finds that less than 20% of EPs can be lost to the limiting surface in ITER assuming a source distribution uniform in minor radius, due to an unstable RWM that produces 100 Gauss poloidal field perturbation at the outboard mid-plane just inside the (effective) resistive wall surface. On top of the initial prompt drift orbit loss for counter-current EPs, the RWM induced particle loss occurs on a one second time scale, which is comparable to the RWM growth time in ITER. The ‘wetted’ area, due to the lost EPs striking the limiting surface, is generally found to be large due to the RWM. This is a favorable prediction for ITER. The loss distribution in the poloidal angle is more uniform for co-current EPs. Counter-current EPs experience outward orbit drift when launched from the low-field side, and tend to more often hit the bottom region of the limiting surface.

Predicting the impact of heating and current drive (H&CD) sources is essential to evaluate the performance of ITER plasmas and to subsequently optimise the scenarios for the four stages of the ITER research plan. This should be done in the context of global transport calculations of complete plasma discharges. For this purpose, a dedicated workflow has been developed in the ITER integrated modelling and analysis suite as a modular component to be used together with transport solvers to quantify the dynamics of H&CD sources for the different phases of a plasma discharge, including possible synergetic effects between the heating sources. This paper presents the results of the combined modelling of H&CD sources for the ITER DT baseline 15 MA/5.3 T scenario including the synergy between neutral beam injection (NBI) of deuterium, fusion-born alpha particles and ion cyclotron resonance heating (ICRH) at the fundamental frequency of deuterium, showing modest synergetic effects. The results of the combined H&CD sources for an ITER 7.5 MA/2.65 T helium plasma of the second pre-fusion power operation phase (PFPO-2) are also shown, exhibiting more significant synergetic effects between the fundamental ICRH minority hydrogen heating and NBI hydrogen beams. Finally, a study of electron cyclotron heating absorption for an ITER helium PFPO scenario at 7.5 MA/2.65 T is also presented with a discussion on the edge parasitic absorption that arises under specific conditions.