We explore the theoretical capabilities of Electron Cyclotron Current Drive (ECCD) in a volumetric neutron source (VNS), a high-neutron-fluence tokamak for component testing and qualification that is being considered in the frame of the EUROfusion Consortium. Two selected applications are addressed, namely bulk current drive in the plasma centre and stabilization of Neoclassical Tearing Modes (NTMs). It is shown that a current drive efficiency above 50 kA/MW can be reached, close to typical values reported for DEMO central ECCD. Suppression of NTMs should require less than 10 MW, but this power might become marginal in case of significant beam broadening caused by density fluctuations. The optimum launcher parameters for the envisaged applications will be further iterated to find the sweet spot between physics and engineering constraints.

While ITER will demonstrate that long duration pulses producing significant fusion power is possible in a fusion reactor relevant environment, the aim of the volumetric neutron source (VNS) tokamak is to offer a testbed for studying the impact of long-time fusion neutron exposure to the plasma facing components. Presently, the auxiliary heating schemes foreseen for the VNS are electron cyclotron and neutral beam injection (NBI) heating while the main fusion power generation is intended to result from beam-target reactions between the fast deuterium NBI ions and the thermal Tritium ions in the plasma. The present paper discusses the potential of only using wave heating to maximise the fusion yield in modest scale fusion reactors. We focus here on ion cyclotron resonance heating (ICRH or IC) while electron cyclotron heating (ECRH or EC) provides the basic target plasma the IC waves couple to. More in particular, we exploit the idea that permitted the fusion performance to be boosted beyond what NBI alone could achieve in the JET DTE2 and DTE3 campaigns. Adopting IC heating at the fundamental cyclotron frequency allows to create a moderately fast ion tail that has a large fraction of particles of a high concentration minority tail near the energy where the $D-T$ cross section peaks, similar to the NBI beam-target fusion scheme. Rather than adopting balanced $D$ and $T$ concentrations and count on fusion reactions produced by 2 thermal fuel ion thermal populations, the steepness of the cross section curve is exploited to boost the nuclear reactions between a thermal population and an IC created fast particle ‘beam’. In the present wave-only exploratory study it is estimated that $20-25~\mathrm{MW}$ of fusion power can be achieved when adopting the 2025 reference parameters for the VNS device. A more comprehensive study (not just focusing on heating) is, however, needed to assess the potential in more detail.

We present the latest updates of the flight simulator of ASDEX Upgrade (AUG), 'Fenix AUG'. Fenix AUG contains simplified models for both the plasma physics and the device operation. The updated version is split into three independent repositories and has been integrated into Docker containers for better management and deployment. The device models can now directly import their configuration from the AUG control system. Although the physics models can be fine tuned to each case, the default settings have been set up to run a large variety of experiments without the need of case-to-case adaptation. Together, the enhancements make Fenix AUG a versatile tool that can be applied to a wide range of applications 'out-of-the-box', such as: to compare simple physics models to the experiment, to assess new controllers within the framework of the existing ones, or, to verify that a planned discharge will follow expected trajectories and will not run into machine or safety limits. An example of each application is showcased to highlight some strengths and limitations of the simulator.

In this paper, a feedback control algorithm to reduce the DEMO fusion power fluctuations due to the presence of possible disturbances and unpredicted events is proposed. Fluctuations can be limited controlling temperature and densities of the different plasma species by means of a set of actuators like electron cyclotron resonance heating, fuelling rate modulation and/or impurity injection. However, electron cyclotron resonance heating is expensive in terms of power cost and impurities are difficult to control because it is hard to remove them from the core. A possible alternative is to use plasma shape as a virtual actuator and in particular elongation through plasma shape magnetic control. This work exploits the possibility to use elongation modifications in a coordinated way with electron cyclotron resonance heating power, and electron density, for real time control. The integrated control scheme implements a supervised control allocation strategy and accounts for constraints coming from actuator limits as maximum elongation, maximum ECRH power or maximum electron density, but also on other key plasma parameters like plasma safety factor and Greenwald density limit. Numerical results of the overall control scheme to compensate the effect of uncontrolled argon injection and tungsten flake event are obtained within the DEMO flight simulator based on Fenix.

Early heating tokamak plasmas during the current ramp-up is a common approach to efficiently access advanced tokamak (AT) configurations. We have previously presented an analytical predictive model to aid in the AT scenario design, optimized for a single scenario (Schramm et al 2024 Nucl. Fusion 64 036013). We now show that this approach can be generalized to other scenarios and even other devices. To this end, we present results from an AUG scenario with higher current ( q 95 ∼ 4 ) and central counter-ECCD for elevated q -profiles, as well as from JET flux pumping experiments. For both cases, good agreement with the experiment could be shown with only minor changes to the model’s transport parameters. Comparisons with an alternative approach (Mulders et al 2024 Nucl. Fusion 64 026021) based on the RAPTOR code have also been carried out with good agreement.

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.

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 a shattered pellet injection (SPI) system the penetration and assimilation of the injected material depends on the speed and size distribution of the SPI fragments. ASDEX Upgrade (AUG) was recently equipped with a flexible SPI to study the effect of these parameters on disruption mitigation efficiency. In this paper we study the impact of different parameters on SPI assimilation with the 1.5D INDEX code. Scans of fragment sizes, speeds and different pellet compositions are carried out for single SPI into AUG H-mode plasmas. We use a semi-empirical global reconnection event (GRE) onset condition to study the material assimilation trends. For mixed deuterium-neon pellets, smaller/faster fragments start to assimilate quicker. However, at the expected onset of the GRE, larger/faster fragments end up assimilating more material. Variations in the injected neon content lead to a large difference in the assimilated neon for neon content below < 10(21) atoms. For larger injected neon content, a self-regulating mechanism limits the variation in the amount of assimilated neon. We use a back-averaging model to simulate the plasmoid drift during pure deuterium injections with the back-averaging parameter determined by a interpretative simulation of an experimental pure deuterium injection discharge. Again, larger and faster fragments are found to lead to higher assimilation with the material assimilation limited to the plasma edge in general, due to the plasmoid drift. The trends of assimilation for varying fragment sizes, speeds and pellet composition qualitatively agree with the previously reported experimental observations.

In this paper an extensive database of SPARC H-modes confinement predictions has been provided, to assess its variability with respect to few input assumptions. The simulations have been performed within the ASTRA framework, using the quasi-linear model TGLF SAT2, including electromagnetic effects, for the core transport, and a neural network trained on EPED simulations to predict the pedestal height and width self-consistently. The database has been developed starting from two SPARC H-mode discharges (12.2 T, i.e. Primary Reference Discharge or PRD, and 8 T, i.e. reduced field) and permuting 4 input parameters (W concentration, DT mixture concentration, temperature ratio at top of pedestal and deviation of pedestal pressure from the EPED prediction), to perform a sensitivity study. For the PRD a scan of auxiliary input power (ion cyclotron heating) has been performed up to 25MW, to keep highly radiative plasmas above the LH power threshold as predicted by Martin and Schmidtmayr power scalings. A scan of pedestal density has then been performed for both PRD and 8T databases. ptop/pEPED and Ti/Te at top of pedestal showed the biggest impact on the fusion gain. Significant variation is observed across the database, highlighting the importance of sensitivity studies. Below a certain W concentration, the 12T database shows that Q > 5 is consistently achieved for full-field H-modes with 11 MW of auxiliary power, and values of Q > 2 are assured when increasing the input power to keep the plasma in H-mode. The 8T database demonstrates that SPARC can access a Q > 1 operational window with low W concentration, making it a potentially interesting scenario for obtaining breakeven conditions.

State-of-the-art 1D transport solvers ASTRA and TRANSP are verified, then validated across a large database of semi-randomly selected, time-dependent DIII-D discharges. Various empirical models are provided as baselines to contextualize the validation figures of merit using statistical hypothesis tests. For predicting plasma temperature profiles, no statistically significant advantage is found for the ASTRA and TRANSP simulators over a baseline empirical (two-parameter) model. For predicting stored energy, a significant advantage is found for the simulators over a baseline empirical model based on confinement time scaling. Uncertainty in the results due to diagnostic and profile fitting uncertainties is approximated and determined to be insignificant due in part to the large quantity of discharges employed in the study. Advantages are discussed for validation methodologies like this one that employ (1) large databases and (2) baselines for comparison that are specific to the intended use-case of the model.

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).

The flight simulator predicts the dynamic behavior of a full plasma discharge (described in terms of one-dimensional profiles) by employing multiple control loops based on synthetic diagnostics, which could also emulate realistic sensor and actuator models. It serves as a valuable tool for designing and optimizing plasma scenarios, as well as for assessing the feasibility of controlling discharges. The Fenix flight simulator, originally developed for the ASDEX Upgrade, has been ported to EU-DEMO and is capable of modeling any tokamak.One of the essential elements in a flight simulator is the link between the co-simulated plasma physics and the control loops. This element is tightly coupled to the specifications of both the plasma model and the control algorithms to be implemented; but on the other hand, to ensure the portability and applicability of the flight simulator to different scenarios or devices, the coupling between plasma and control algorithms should be neutral to any concrete device and configuration. In addition, as a serial component of the control loop, data exchange takes place at every single step of the control simulation, therefore an efficient implementation is critical for the overall simulation performance. This paper summarizes the universal approach recently implemented in Fenix, which satisfies all the above requirements while remaining lightweight.

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.

Real-time reconstruction of the magnetic equilibrium provides fundamental control of plasma shape and position in a tokamak. Details of the implementation of the equilibrium reconstruction code developed for the ASDEX Upgrade tokamak (JANET++) are summarized. Cubic Hermite splines are introduced as current density basis functions for solving the Grad-Shafranov equation. The choice of the optimal Tikhonov regularization parameter is discussed. The code is validated by comparing the results of the equilibrium reconstruction with those of further equilibrium reconstructions available on ASDEX Upgrade (CLISTE and IDE). In a high time resolution study of a discharge with edge localized modes (ELM), the poloidal asymmetry of the fits and magnetic probe measurements suggest that the real-time equilibrium reconstruction captures the essential features of the current density redistribution in an ELMing edge plasma. An efficient algorithm to locate multiple X-points and identify the active one in advanced X-divertor and snowflake divertor configurations is presented.

The possibility of driving a large fraction of the plasma current in a tokamak reactor employing electron cyclotron waves is investigated for scenarios as envisaged in the European DEMO through beam tracing calculations performed with the TORBEAM code, which assumes a linear regime for power absorption and employs an adjoint method for the determination of the driven current. Comparatively high ECCD efficiency in the inner half of the plasma column can be achieved by injecting the wave from an elevated position. On the other hand, the efficiency deteriorates in the colder, outer part of the plasma, so that a prohibitive amount of power would be needed to sustain the plasma current non-inductively by ECCD only in the considered scenarios. As an alternative scheme, the injection of slow extraordinary wave below the fundamental resonance is considered. The basic physics features of this scenario are discussed.

Advanced tokamak (AT) scenarios applying additional heating during the current ramp (early-heating) usually require many iterations if developed fully empirically. To reduce the required experimental time, a model has been developed in the ASTRA framework, capable of doing predictive simulations of the relevant parameters. As scenario development requires fast iterations and inter-discharge runs, a sufficiently short run-time is required. While using a simplified transport model to achieve this, comparisons to experimental data from ASDEX-Upgrade (AUG) still show good agreement. Using this model, a new high performance early-heating AT scenario has been developed and successfully run on AUG with the results matching the predictions.

A long standing shortfall in the predicted L-mode edge energy transport by reduced quasi-linear models of gyrokinetic turbulent transport has been resolved. The improved model TGLF-SAT2 has higher fidelity to gyrokinetic simulations of the electron-scale contribution to the electron energy transport and the ion-scale flux surface shape dependence of energy transport. The success of TGLF-SAT2 in predicting the L-mode and Ohmic edge profiles is critical to whole pulse simulation and opens the door to prediction of the H-mode power threshold.

This study employs the established momentum transport analysis at ASDEX Upgrade [Zimmermann et al., Nucl. Fusion 63, 124003 (2023)] to investigate the parametric variations of the momentum transport coefficients in the core of H-mode plasmas. These experimental results are compared to a comprehensive database of gyrokinetic calculations. Generally, good agreement between predicted and measured diffusive and convective transport coefficients is found. The predicted and measured Prandtl numbers correlate most dominantly with the magnetically trapped particle fraction. The experimentally inferred pinch numbers strongly depend on the logarithmic density gradient and magnetic shear, consistent with the theoretical predictions of the Coriolis pinch. The intrinsic torque from residual stress in the inner core is small, scales with the local logarithmic density gradient, and the data indicate a possible sign reversal. In the outer periphery of the core, the intrinsic torque is always co-current-directed and scales with the pressure gradient. This is consistent with prior experimental findings and global, non-linear gyrokinetic predictions. It suggests that profile shearing effects generate the intrinsic torque in the inner core. Toward the outer core, most likely, effects from E×B-shearing become more influential. These results offer the first comprehensive picture of this transport channel in the core plasma and contribute to validating the corresponding theoretical understanding. The derived scaling laws are used to construct a reduced momentum transport model, which has been validated against an additional dataset. This demonstrates that the model captures the essential contributions to momentum transport in the core of H-mode plasmas.

An integrated framework that demonstrates multi-species, multi-channel modelling capabilities for the prediction of impurity density profiles and their feedback on the main plasma through radiative cooling and fuel dilution is presented. It combines all presently known theoretical elements in the local description of quasilinear turbulent and neoclassical impurity transport, using the models TGLF-SAT2 and FACIT. These are coupled to the STRAHL code for impurity sources and radiation inside the ASTRA transport solver. The workflow is shown to reproduce experimental results in full-radius L-mode modelling. In particular, a set of ASDEX Upgrade L-modes with differing heating power mixtures and plasma currents are simulated, including boron (B) and tungsten (W) as intrinsic impurities. The increase of predicted confinement with higher current and the reduction of core W peaking with higher central wave heating are demonstrated. Furthermore, a highly radiative L-mode scenario featuring an X-point radiator (XPR) with two intrinsic (B, W) and one seeded argon (Ar) species is simulated, and its measured radiated power and high confinement are recovered by the modelling. The stabilizing effect of impurities on turbulence is analysed and a simple model for the peripheral X-point radiation is introduced. A preliminary full-radius simulation of an H-mode phase of this same discharge, leveraging recent work on the role of the ExB shearing at the edge, shows promising results.

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.