Impact of magnetohydrodynamic on access and dynamics of X-point radiator regimes (XPR)
ITER and future fusion powerplants will need to operate without degrading too much the plasma facing components (PFC) in the divertor, the peripheral element with is dedicated to heat and particle exhaust in tokamaks. In this context, two key factors must be considered: heat fluxes must stay below engineering limits both in stationary conditions and during violent transient events. An operational regime recently developed can satisfy those two constraints: the X-point Radiator (XPR). Experiments on many tokamaks, in particular WEST which has the record plasma duration in this regime (> 40 seconds), have shown that it allowed to drastically reduce heat fluxes on PFCs by converting most of the plasma energy into photons and neutral particles, and that it also was able to mitigate – or even suppress – deleterious magnetohydrodynamic (MHD) edge instabilities known as ELMs (edge localised modes). The mechanisms governing these mitigation and suppression are still poorly understood. Additionally, the XPR itself can become unstable and trigger a disruption, i.e., a sudden loss of plasma confinement cause by global MHD instabilities.
The objectives of this PhD are: (i) understand the physics at play during the interaction XPR-ELMs, and (ii) optimise the access and stability of the XPR regime. To do so, the student will use the 3D non linear MHD code JOREK, the European reference code in the field. The goal is to define the operational limits of a stable XPR with small or no ELMs, and identify the main actuators (quantity and species of injected impurities, plasma geometry).
A participation to experimental campaigns of the WEST tokamak (operated by IRFM at CEA Cadarache) – and of the MAST-U tokamak operated by UKAEA – is also envisaged to confront numerical results and predictions to experimental measurements.
Study of impurity transport in negative and positive triangularity plasmas
Nuclear fusion in a tokamak is a promising source of energy. However, a question arises: which plasma configuration is most likely to produce net energy? In order to contribute to answering this, during this PhD, we will study the impact of magnetic geometry (comparison between positive and negative triangularity) on the collisional and turbulent transport of tungsten (W). The performance of a tokamak strongly depends on the energy confinement it can achieve. The latter degrades significantly due to turbulent transport and radiation (primarily from W). On ITER, the tolerated amount of W in the core of the plasma is about 0.3 micrograms. Experiments have shown that the plasma geometry with negative triangularity (NT) is beneficial for confinement as it significantly reduces turbulent transport. With this geometry, it is possible to reach confinement levels similar to those of the ITER configuration (H-mode in positive triangularity), without the need for a minimum power threshold and without the associated plasma edge relaxations. However, questions remain: what level of W transport is found in NT compared to a positive geometry? What level of radiation can be predicted in future NT reactors? To contribute to answering these questions, during this PhD, we will evaluate the role of triangularity on impurity transport in different scenarios in WEST. The first phase of the work is experimental. Subsequently, the modeling of impurity transport will be carried out using collisional and turbulent models. Collaboration is planned with international plasma experts in NT configurations, with UCSD (United States) and EPFL (Switzerland).
Control of trapped electron mode turbulence with an electron cyclotron resonant source
The performance of a tokamak-type fusion power plant in term of energy gain will be limited by turbulent transport. The instability of trapped electron modes is one of the main instabilities causing turbulence in tokamaks. Furthermore, electron cyclotron resonance heating (ECRH) is the generic heating system in current and future tokamaks. Both physical processes are based on resonant interactions with electrons, in space and velocity. Since heating has the effect of depopulating the resonant interaction zone of its electrons, superimposing its resonance on that of the instability can theoretically lead to a stabilisation of the trapped electron modes.
The objective of the thesis is twofold: (i) to construct scenarios where this mechanism exists and validate it using linear simulations, then (ii) to characterise its effect and quantify its effectiveness in non-linear regimes where linear effects compete with the self-organisation of turbulence, with collisional processes and with the dynamics of average profiles. Potentially, this entirely new control technique could improve the performance of tokamaks at no additional cost. The PhD thesis will require a detailed theoretical understanding of the two resonant processes and their various control parameters. It will be based on the use of the high performance computing gyrokinetic code GYSELA dedicated to the study of transport and turbulence in tokamak plasmas, which has recently been enhanced with an ECRH heating module. An experimental component is also planned on the WEST and/or TCV tokamaks to validate the identified most promising turbulence control scenario(s).
Plasma real time control by calorimetry
Inside thermonuclear fusion devices, plasma facing components are subject to intense heat fluxes. The WEST tokamak has water cooled plasma facing components to limit their heating. Calorimetric measurement on these components allows for the measurement of the power received by each component. This makes it possible to control the plasma position or the additional plasma heating in function of the power distribution.
During this PhD, a simulation of plasma control using calorimetry will be performed, simulating the heat fluxes received by the components as a function of the plasma position and the associated calorimetric response. In-situ calorimetric measurements will be carried out on the components at the top and bottom of the machine during dedicated plasma experiments to refine the simulations and the control of the WEST plasma position based on calorimetric measurements will finally be implemented and validated during dedicated experiments, for plasma-facing components protection and plasma physics purposes.
Real-Time control of MHD instabilities during WEST long pulses
In magnetically confined plasmas, low-frequency (typ. 1-10 kHz) large-scale magnetohydrodynamic (MHD) instabilities represent a risk for performance and plasma stability. During long pulses in the WEST tokamak, deleterious MHD modes appear frequently inducing a drop of central temperature and a higher plasma resistivity that result in lower performances and shorter discharge duration. The real-time detection of such instabilities and the application of mitigation strategies is therefore of great importance for plasma control in WEST but also for future devices like ITER.
These MHD instabilities induce coherent temperature/density perturbations. Instruments like Electron Cyclotron Emission (ECE) radiometer or reflectometrer provide localized, high time resolution of temperature or density fluctuations. However, MHD analysis is currently performed offline, after the discharge. Real-time capability is crucial for control applications. The modes must first be identified before applying a mitigation strategy based on the knowledge of the MHD stability criteria. MHD stability is strongly affected by local heating and current drive, for which Electron Cyclotron Resonance Heating and Current Drive systems (ECRH/ECCD) are especially well suited.
The objective of this PhD is to develop a control strategy for WEST long pulse operation. The first step is the real-time detection of low frequency MHD instabilities using first ECE radiometer, then adding instruments like ECE-imaging or reflectometry to enhance reliability and accuracy. Integrated plasma modelling will then be performed to explore MHD mitigation strategies. ECCD is an obvious actuator, but other tools such as a temporary change of the plasma parameters (current, density or temperature) will also be evaluated. The mitigation strategy will be integrated in WEST Plasma Control System. Initial strategy will rely on simple control loop, then Neural Network or deep-leaning algorithms will be tested.