Pierre David

Optimising the performance of tokamak operation is not only a matter of tuning individual control loops, but needs a holistic approach with a focus on the effective interplay of all control system components. The ASDEX Upgrade Discharge Control System (DCS) has enabled outstanding results of physics effect studies and operation scenario development. This is facilitated by efficient and highly robust algorithms for measurements data processing and plasma state estimation, a versatile feedback control suite, a powerful actuator management approach that optimises the utilisation of the limited number of actuators, and pulse supervision control reacting to events and plasma state conditions to achieve a maximum exploitation of the pulse and protect device and investment. Moreover, Fenix, a fast ‘flight simulator’ with a full model of the control system and a control-oriented plasma and plant model, assists in scenario design, validation and control method development, thus accelerating the turnaround time from the design of emerging capabilities to their exploitation in plasma pulses. Larger and more complex fusion devices such as ITER and DEMO in particular, will benefit from such a control structure and active contributions are being made to the architectural design of their control systems.

Recent experiments conducted during the 2025 campaign in ASDEX Upgrade have pushed the limits of negative triangularity shaping, achieving top triangularity values of δ top ∼ − 0.55 while keeping bottom triangularity around 0 in diverted plasmas. Guided by ideal-MHD stability calculations using the linear MHD BALOO solver and supported by TCV results, these experiments have demonstrated a transition to an edge that is more stable against peeling-ballooning modes, leading to ELM-free plasmas. Those ELM-free plasmas also exhibit a dithering behavior similar to limit-cycle oscillations. Linear MHD modeling showed that the achieved shapes remained marginal with respect to the second stability region for ballooning modes suggesting that with minor further shaping, a robust NT edge could be achieved. The associated energy confinement is generally reduced, with clear signs of strong power degradation. In addition, a high fraction of fast-ion energy is observed, particularly in high-power/low density discharges. An approximately 15% improvement in normalized energy confinement is obtained when seeding nitrogen. As this effect comes from the edge, it highlights a path towards pedestal tailoring for further confinement optimization and supporting negative triangularity as a potentially viable reactor-relevant operational scenario in full-metal-wall devices.

Future fusion devices are complex systems and present-day experimental devices need additional versatility to test as large a parameter set as possible. To that end, more and more advanced control and monitoring strategies and actuators are used, such that although each element is well known, the system’s behaviour exceeds the predictability of its individual components. As a result, being able to run simple yet realistic enough simulations that include the plant systems, the plasma and their interactions became particularly important. Such simulation platforms, usually referred to as flight simulators, are designed to use the experimental discharge program as input, and to output the temporal evolution of both the plant and the plasma quantities. In addition, flight simulators are expected to be used iteratively to test and improve scenarios or models. Therefore, while any integrated model is technically eligible, only simplified models remain practically viable due to the time constraints. The tutorial paper is based on a lecture delivered at the 14th ITER International School focused on ‘Integrated Modelling of Magnetic Fusion Plasmas’. It describes the concept of flight simulators, with a focus on the control system elements surrounding the physics model. In particular, the requirements and benefits of closed-loop simulations are highlighted.

The goal of this work is to develop a simulator for the TCV tokamak that self-consistently includes transport and equilibrium while taking the plant control systems into account. To achieve this, we adapted the Fenix flight simulator that was originally developed for ASDEX Upgrade for use on TCV. The MEQ-FGE free boundary equilibrium code that is widely used for TCV analysis as well as the TCV magnetic controller were coupled to this framework. In addition, a reduced transport model drawing inspiration from earlier RAPTOR modelling for TCV was implemented. Applying this simulator to discharges in the TCV database shows a good agreement of both kinetic and magnetic results with experimental measurements. In particular, it is illustrated how the shape controller is expected to stabilise discharges and it is found that details of the kinetic profiles can be important for the shape and the stability of the equilibrium. This work paves the way to using Fenix for discharge preparation and validation as well as controller development and validation for TCV.

Tomographie d'une colonne de plasma magnétisé: Quel est le point commun entre des propulseurs à effet hall, des sources d'ions et les grandes machines de recherche sur la fusion magnétique ? Ils sont tous composés de plasmas interagissant avec des champs électrique et magnétique orthogonaux, et leurs tailles, sophistications et inaccessibilités rendent leur étude directe compliquée. Cette étude directe peut être menée à bien sur des machines plus simples, comme la colonne de plasma magnétisée Mistral utilisée dans ce travail, qui sont conçues pour l’étude de mécanismes fondamentaux. La tomographie, quant à elle, est couramment utilisée dans les tokamaks et stellarators, mais plus rarement sur des machines de laboratoire. Son intérêt majeur est de pouvoir reconstituer l’évolution temporelle de section 2D de plasma, et ce sans mesure intrusive. Dans le cadre de cette thèse un diagnostic de tomographie a été entièrement conçu, installé, calibré et testé. Les modèles existant de tomographie ont d'abord été adaptés à ce nouveau contexte, pour ensuite développer et valider le code complet d’inversion tomographique associé. Puis, une étude de faisabilité a été réalisée en mettant au point un diagnostic de tomographie utilisant un seul capteur avec un échantillonnage conditionnel sur des modes réguliers. L’attention est alors portée au développement, à la configuration et à l’application du diagnostic complet à 128 voies. Enfin, une étude paramétrique des modes réguliers a fait ressortir l'importance des paramètres de contrôle sur les modes (présence, fréquence et parité), et l'attention qui doit être portée à l'ensemble des paramètres expérimentaux, ainsi que l’évolution de leur forme et le comportement du plasma central.

The 3D fluid edge code EMC3-EIRENE has been used to simulate three ASDEX Upgrade discharges with the same equilibrium but different upstream collisionality ν ∗ . In this scan, the plasma transits from an ELMy state at low collisionality to being in the quasi-continuous exhaust regime (QCE) at high ν ∗ . Particular attention was devoted to matching the experimental plasma profiles in front of the ion cyclotron resonance heating (ICRH) limiter, aiming to obtain a reliable estimation of the ICRH limiter peak heat flux in the QCE regime discharge. As a result, we observed that while moving from ELMy to QCE, the limiter peak heat flux increases, reaching values in the order of the MW m − 2 for a limiter-separatrix distance of 6 cm . These values were compared with those at the outer divertor, measured by infrared thermography, revealing that the heat flux perpendicular to the limiter surface can reach up to 90% of the one impinging onto the outer target. Finally, an attempt was made to extend the obtained results towards ITER. This simple analysis suggests that a large wall clearance and a small incidence angle of the magnetic field line, not only on the divertor target but also on ITER’s first wall, might have a crucial role in preventing wall damage during QCE discharges.

The empirical scaling for the H-mode power threshold in tokamaks has been revisited using a database with threshold data from machines with a metallic first wall as part of International Tokamak Physics Activity (ITPA) task TC-26. The database contains discharges from ASDEX Upgrade (W), JET (Be/W) and Alcator C-Mod (Mo). This was motivated by reports that in like-for-like discharges the power threshold was reduced by approximately 30% after the change from carbon based to metallic first wall materials in AUG [F. Ryter et al., Nuclear Fusion, 53(11):113003 (2013)] and JET [C.F. Maggi et al., Nuclear Fusion, 54(2):023007 (2014)]. The database contains L-H transition data for all hydrogen isotopes and mixtures, including T and DT from the recent JET campaigns. compared to the ITPA 2008 scaling [Y R Martin et al., Journal of Physics: Conference Series, 123(1):012033 (2008)], the metal wall scaling has a smaller magnetic field exponent but a larger density exponent. We present an additional parameter to capture the strong dependence of the L-H power threshold (approx. factor 2) on the magnetic configuration in the divertor on JET. The scaling recovers the approximate inverse isotope mass scaling of the threshold power. Alternative scalings involving the plasma current and poloidal magnetic field are explored. Despite the reduction in threshold observed earlier, the scalings based on the metal wall database do not necessarily extrapolate to a lower threshold for ITER compared to the ITPA 2008 scaling, especially at high density. The divertor configuration effect induces the largest uncertainty in the extrapolation.

The required heating power, P LH , to access the high confinement regime (H-mode) in nitrogen seeded deuterium plasmas is investigated at the ASDEX Upgrade tokamak. It is shown that a stepwise increase of the nitrogen seeding level increases P LH monotonically by up to a factor of two with respect to the unseeded reference. At the highest seeding rates a stable plasma discharge from a detached low confinement mode (L-mode) to a detached H-mode without interrupting the divertor detachment during the L–H transition is achieved. The results suggest that the impurity seeding level required for continuous divertor detachment, as it is mandatory for a reactor-scale fusion plasma, inevitably leads to a substantial increase in P LH . Therefore, the impact of required impurity seeding for detachment purposes on P LH must be taken into account for future reactor-scale plasma scenarios.

In this work, the theory-based turbulence-driven transport model Trapped Gyro Landau Fluid (TGLF)-SAT2 (Staebler et al 2020 Plasma Phys. Control. Fusion 63 015013) (hereafter abbreviated as TGLF) is implemented and tested in the Fenix flight simulator (Fable et al 2022 Plasma Phys. Control. Fusion 64 044002) against the current ramp-up phases of ASDEX Upgrade discharges. Turbulence spectra for the measured kinetic profiles are calculated and analyzed using TGLF. Additionally, predictive simulations of the ramp-up are conducted using Fenix, with TGLF computing turbulent energy and particle transport, and compared with the measurements. In order to reproduce the evolution of the kinetic profiles in the presented modeling framework, the transport coefficients have to be limited in absolute value for ρ pol ≳ 0.975 . Moreover, an additional artificial particle diffusive component has to be added to avoid unphysical spikes in the density profile. Even without any modifications, the transport computed by TGLF for ρ pol ≲ 0.9 seems to be close to the transport in the experiment, whereby the kinetic profiles of the experiment can be reproduced.