Table of contents

Controlling the heat transport profile is important for high performance in magnetically confined fusion plasmas. In this study, improved electron heat transport was achieved in neutral beam injection plasmas by applying high-intensity gas puffing (HIGP) on a stellarator/heliotron device called Heliotron J. Compared with conventional gas puffing (GP) fueling discharge, a higher and more peaked electron temperature profile was obtained, and the core ion temperature was slightly higher but similarly shaped. Using similar parameters, the electron density profile for HIGP remained similar and differed from the hollow density profile observed in electron cyclotron heating-eIBT plasma. Transport analysis using the FIT3D and TR-snap codes showed a clear reduction in the effective electron heat transport coefficient in the plasma core region. However, more detailed experiments are required to understand the mechanisms underlying this improvement fully.

At the Joint European Torus, the reference diagnostic to measure electron density is Thomson scattering. However, this diagnostic has a low sampling rate, which makes it impractical to study the temporal dynamics of fast processes, such as edge localized modes. In this work, we use machine learning to predict the density profile based on data from another diagnostic, namely reflectometry. By learning to transform reflectometry data into Thomson scattering profiles, the model is able to generate the density profile at a much higher sampling rate than Thomson scattering, and more accurately than reflectometry alone. This enables the study of pedestal dynamics, by analyzing the time evolution of the pedestal height, width, position and gradient. We also discuss the accuracy of the model when applied on experimental campaigns that are different from the one it was trained on.

When designing a fusion power plant, many first-of-a-kind components are required. This presents a large potential design space across as many dimensions as the component's parameters. In addition, multiphysics, multiscale, high-fidelity simulations are required to reliably capture a component's performance under given boundary conditions. Even with high performance computing (HPC) resources, it is not possible to fully explore a component's design space. Thus, effective interpolation between data points via machine learning (ML) techniques is essential. With sequential learning engineering optimisation, ML techniques inform the selection of simulation parameters which give the highest expected improvement for the model: balancing exploitation of the current best design with exploration of uncertain areas in the design space. In this paper, the application of an ML-driven design of experiment procedure for the sequential learning engineering design optimisation of a fusion component is shown. A parameterised divertor monoblock is taken as a typical example of a fusion component requiring HPC simulation to model. The component's geometry is then optimised using Bayesian optimisation, seeking the design which minimises the stress experienced by the component under operational conditions.

This study is motivated by experiments on Tore Supra and WEST tokamaks where a deepening of the E × B velocity—governed by the radial electric field E_r—near the edge is observed when the safety factor decreases in L-mode plasmas. Flux-driven global simulations of ion temperature gradient turbulence recover qualitatively the trend observed in the experiments, i.e. the E × B velocity increases when decreasing the safety factor. From these simulations, multiple clues point out the role of turbulence in the establishment of the radial electric field even though the turbulent intensity increases with the safety factor. The proposed mechanism to elucidate this phenomenon, backed up by a reduced model, is that the damping of the poloidal flow, governed by the neoclassical friction, increases more strongly with the safety factor than the turbulent drive for E_r, due to the $(r,\theta)$ component of the Reynolds stress.

In recent years a strong effort has been made to investigate disruption avoidance schemes in order to aid the development of integrated operational scenarios for ITER. Within the EUROfusion programme the disruptive H-mode density limit (HDL) has been studied on the WPTE (Work Package Tokamak Exploitation) devices ASDEX Upgrade, TCV and JET. Advanced real-time control coupled with improved real-time diagnostics has enabled the routine disruption avoidance of the HDL. This allowed the systematic study of the influence of various plasma parameters on the onset and behavior of the HDL in regimes not easily accessible otherwise. The upper triangularity $\delta_\mathrm{top}$ is found to have a significant influence on the x-point radiator (XPR), which plays a major role for the evolution of the disruptive HDL. At high $\delta_\mathrm{top}$ the gas flow rate at which the onset of the XPR occurs is strongly reduced compared to low $\delta_\mathrm{top}$. The reduction of $\delta_\mathrm{top}$ has proven to be an effective actuator for the HDL disruption avoidance on ASDEX Upgrade for highly shaped scenarios ($\delta_\mathrm{top}\gt0.25$). It is observed that the occurrence of the XPR and the H–L transition at the density limit are two separate events, the order of which depends on the applied auxiliary heating power. At sufficiently high heating power the XPR occurs before the H–L transition. Impurity seeding, used for divertor detachment, influences the onset and the dynamics of the XPR and the behavior of the HDL. The stable existence of the XPR, which is thought to be a requirement for detachment control in future devices, has also been observed without impurity seeding. The implementation of a robust and sustainable operational scenario, e.g. for ITER, requires the combination of continuous control and exception handling. For each disruption path the appropriate observers and actuators have to be validated in present devices. Automation of the dynamic pulse schedule has proven successful to scan the operational space of the HDL without disruption. Applying such a technique to ITER could reduce the machine risk induced by disruptions during commissioning. The methodology to develop physics-based observers, which indicate the entry into a disruption path well in time, and applying the appropriate action before the discharge becomes unstable has proven successful.

Low convergence ratio implosions (where wetted-foam layers are used to limit capsule convergence, achieving improved robustness to instability growth) and auxiliary heating (where electron beams are used to provide collisionless heating of a hotspot) are two promising techniques that are being explored for inertial fusion energy applications. In this paper, a new analytic study is presented to understand and predict the performance of these implosions. Firstly, conventional gain models are adapted to produce gain curves for fixed convergence ratios, which are shown to well-describe previously simulated results. Secondly, auxiliary heating is demonstrated to be well understood and interpreted through the burn-up fraction of the deuterium-tritium fuel, with the gradient of burn-up with respect to burn-averaged temperature shown to provide good qualitative predictions of the effectiveness of this technique for a given implosion. Simulations of auxiliary heating for a range of implosions are presented in support of this and demonstrate that this heating can have significant benefit for high gain implosions, being most effective when the burn-averaged temperature is between 5 and 20 keV.

We discuss how the combination of experimental observations and rapid modeling has enabled to improve understanding of the tokamak ramp-down phase in ASDEX Upgrade. A series of dedicated experiments has been performed, to disentangle the effect of individual actuators like plasma current, auxiliary heating and plasma shaping. Optimized discharge termination strategies with increased margin with respect to radiative and vertical stability limits are proposed and tested in experiment. Radiative collapse of the edge T_e profile after the HL back-transition is avoided by initially maintaining auxiliary heating during L-mode, showing beneficial effects even after the auxiliary heating is turned off. The capability of the RAPTOR code to model the time evolution of the internal inductance $\ell_{i3}$ has been validated, including the effect of a change in the I_p ramp-down rate and the HL transition timing. The reduction of $\ell_{i3}$ caused by rapid compression of the plasma cross-section has been quantitatively recovered in simulations. Successful modeling of the $\ell_{i3}$ time evolution is essential to optimize ramp-down scenarios for future fusion reactors, for which vertical stability and power balance control will be more challenging.

An optimized plasma current ramp-down strategy is critical for safe and fast termination of plasma discharges in a tokamak demonstration fusion reactor (DEMO), both in planned and emergency scenarios, avoiding plasma disruptions and excessive heat loads to the first wall. Plasma stability limits and machine-specific technical requirements constrain the stable envelope through which the plasma must be navigated. Large amounts of auxiliary heating are required throughout the ramp-down phase, to avoid a radiative collapse in the presence of intrinsic tungsten and seeded xenon impurities, as quantitatively estimated in this work. As the plasma current is reduced, the current density becomes increasingly peaked, reflected by a growing value of the internal inductance $\ell_{i3}$, resulting in reduced controllability of the vertical position of the plasma. The feasibility of different plasma current ramp-down rates is tested by applying an automated optimization framework embedding the RAPTOR core transport solver. Optimal time traces for plasma current $I_p(t)$ and plasma elongation $\kappa(t)$ are proposed, to satisfy an I_p-dependent upper limit on the plasma internal inductance, as obtained from vertical stability studies using the CREATE-NL code, as well as a constraint on the time evolution of q₉₅, to avoid an ideal MHD mode. A negative current density near the plasma edge is observed in our simulations, even for the most conservative I_p ramp-down rate, indicating significant transient dynamics due to a large resistive time.

High resolution $\textrm{D}^{*}_{2}$ Fulcher band spectroscopy was used in the MAST-U divertors during Super-X and (shorter-legged) elongated divertor density ramps with $\textrm{D}_{2}$ fuelling from the mid-plane high-field side. In the Super-X case, the upper divertor showed ground state rotational temperatures of the $\textrm{D}_{2}$ molecules increasing from ~6000 K, starting at the detachment onset, to ~9000 K during deepening detachment. This was correlated with the movement of the Fulcher emission region towards the X-point, which is in turn correlated with the movement of the ionisation source. The increase in rotational temperature occurred throughout the divertor except near the divertor entrance, where ionisation was still the dominant process. Qualitative agreement was obtained between the lower and upper divertor. Similar rotational temperatures were obtained in the elongated divertor before the detachment onset, although the increase in rotational temperature during detachment was less clearly observed as less deep detachment was obtained. The measured vibrational distribution of the upper Fulcher state does not agree with a ground state Boltzmann distribution but shows a characteristically elevated population in the ν = 2 and ν = 3 bands in particular; which is strongly correlated to the rotational temperature.

Simulation of the impact of charge-exchange (CX) reactions on beam ions in the Mega Amp Spherical Tokamak (MAST) Upgrade was compared to measurements carried out with a fission chamber (neutron fluxes) and a fast ion deuterium-alpha (FIDA) diagnostic. A simple model was developed to reconstruct the outer-midplane neutral density based on measurements of deuterium-alpha emission from edge neutrals, and on Thomson scattering measurements of electron density and temperature. The main computational tools used were the ASCOT orbit-following code and the FIDASIM code for producing synthetic FIDA signals. The neutral density reconstruction agrees qualitatively with SOLPS-ITER modelling and yields a synthetic passive FIDA signal that is consistent with measurement. When CX losses of beam ions are accounted for, predicted neutron emission rates are quantitatively more consistent with measurement. It was necessary to account for CX losses of beam ions in simulations to reproduce the measured passive FIDA signal quantitatively and qualitatively. The results suggest that the neutral density reconstruction is a good approximation, that CX with edge neutrals causes significant beam-ion losses in MAST Upgrade, typically 20% of beam power, and that the ASCOT fast-ion CX model can be used to accurately predict the redistribution and loss of beam ions due to CX.

A new method for estimating the global erosion of beryllium (Be) in ITER is proposed. The method uses ray tracing-aided tomography to reconstruct the three-dimensional (3D) profile of beryllium visible-light emissivity in boundary plasma from images captured with filtered cameras of VIS/IR wide angle viewing system, H-alpha (and Visible) Spectroscopy diagnostics and signals collected with divertor impurity monitor. The light reflected into the detectors from metallic plasma-facing components (PFCs) is filtered out in the process. The reconstructed Be emissivity is then used to assess the Be influx density distribution along all Be PFCs by integrating the product of the emissivity and the S/XB coefficient along the normal to the PFC surface. The accuracy of this method is evaluated by a comparison with synthetic emissivity data produced by recent simulation of global Be erosion and migration in ITER using the ERO2.0 code. The impact of the uncertainty of PFC light reflection properties on the error in reconstructing the 3D Be emissivity profile and Be influx density is analyzed. The method allows to recover with good accuracy the Be influx density in plasma-wetted areas under the conditions of H-mode fusion power operation with high plasma density in far scrape-off layer (SOL). Under the conditions of lower far-SOL plasma density and L-mode operation, only the total Be influx integrated over the area of the first wall panels with relatively high Be erosion can be reconstructed with a high accuracy. It is shown that neglecting the effects of light reflection may lead to a twofold overestimation of the total Be influx.

A novel experimental method is applied to localize the initial suppression of turbulence, in the form of density fluctuations, at the transition from the low (L-) to the high (H-) confinement mode in toroidal magnetic fusion plasmas. The high radial and temporal resolution, combined with the unprecedented statistical significance, provided the awaited information on a possible dominant $\mathbf{E}\times\mathbf{B}$ shear layer in L-H transition physics. We show, for the first time, that the H-mode turbulence suppression is initiated at the inner $\mathbf{E}\times\mathbf{B}$ shear layer in the ASDEX Upgrade tokamak possibly shedding light on the causality behind the L-H transition process.

Recently, the geodesic acoustic modes (GAMs) induced by negative electrode biasing (EB) are first observed in J-TEXT tokamak, and the electrostatic and magnetic fluctuation characteristics of GAMs and ambient turbulence (AT) are measured by combined Langmuir-magnetic probe. Experimental results reveal that with the ramped up negative biasing, the mean flow shear gradually strengthens, and exciting a high-frequency GAM (H-GAM) and a low-frequency GAM (L-GAM) with frequencies of 24 kHz and 17 kHz respectively. With the combined effects of ${E_r} \times B$ shear by EB and ${\tilde E_r} \times B$ shear by GAMs, the amplitude of AT is effectively suppressed, leading to an improvement in plasma confinement. This work investigates the relationship between AT, GAM zonal flows and ${E_r} \times B$ mean flow shear, which provides proper mean flow shear can enhance the Reynolds stress via enhancement of symmetry breaking and thus enhance the amplitude of GAMs, but excessive flow shear can simultaneously suppress AT and GAM zonal flows.

High repetition rates and efficient energy transfer to the accelerating beam are important for a future linear collider based on the beam-driven plasma wakefield acceleration scheme (PWFA-LC). This paper reports the first results from the Plasma Wakefield Acceleration Collaboration (E300) that are beginning to address both of these issues using the recently commissioned FACET-II facility at SLAC national accelerator laboratory. We have generated meter-scale hydrogen plasmas using time-structured 10 GeV electron bunches from FACET-II, which hold the promise of dramatically increasing the repetition rate of PWFA by rapidly replenishing the gas between each shot compared to the hitherto used lithium plasmas that operate at 1–10 Hz. Furthermore, we have excited wakes in such plasmas that are suitable for high gradient particle acceleration with high drive-bunch to wake energy transfer efficiency- a first step in achieving a high overall energy transfer efficiency. We have done this by using time-structured electron drive bunches that typically have one or more ultra-high current ($\gt$30 kA) femtosecond spike(s) superimposed on a longer (∼0.4 ps) lower current ($\lt$10 kA) bunch structure. The first spike effectively field-ionizes the gas and produces a meter-scale (30–160 cm) plasma, whereas the subsequent beam charge creates a wake. The length and amplitude of the wake depends on the longitudinal current profile of the bunch and plasma density. We find that the onset of pump depletion, when some of the drive beam electrons are nearly fully depleted of their energy, occurs for hydrogen pressure $\unicode{x2A7E}$1.5 Torr. We also show that some electrons in the rear of the bunch can gain several GeV energies from the wake. These results are reproduced by particle-in-cell simulations using the QPAD code. At a pressure of ∼2 Torr, simulation results and experimental data show that the beam transfers about 60% of its energy to the wake.

The production efficiencies of organic light emitting diode (OLED) displays and semiconductor manufacturing have been dramatically improving with the help of plasma physics and engineering technology by utilizing a process monitoring methodology based on physical domain knowledge. This domain knowledge consists of plasma-heating and sheath physics, plasma chemistry, and plasma-material surface reaction kinetics. They were applied to the plasma information based virtual metrology (PI-VM) algorithm with the plasma diagnostics and noticeably enhanced process prediction performance by parameterizing plasma information (PI) in various processes of OLED display and semiconductor manufacturing fabs. PI-VM has shown superior process prediction accuracy, which can trace the states of processing plasmas as an application of data-driven plasma science compared to the classical statistics and machine learning-based virtual metrologies; thus, various plasma processes have been managed and controlled with the help of the PI-VM models. More than this, we have adopted the PI-VM model to optimize the patterning architecture and plasma processes simultaneously. The best combination of the etching pattern structure and plasma condition was adjustable based on the detailed understanding of the angular distribution of sputtered atoms from the etching target surface and their interaction with the plasma sheath based on the PI-VM modeling for etching profile failure prediction.

A precise estimate of the local energy fluxes and erosion profiles at the divertor monoblocks of a fusion reactor requires a kinetic modeling of the plasma–wall interaction. Here, a two-dimensional Particle-in-Cell code is used to quantify the particle and energy fluxes and ion impact distribution functions across the divertor monoblocks of the 'Divertor Tokamak Test' reactor, focusing on poloidal gaps with toroidal bevelling. The considered critical locations are close to the strike points at both Inner and Outer Vertical Targets. A worst-case scenario for particle fluxes corresponding to attached plasma conditions and featuring a single-null magnetic configuration is assumed. The separate and cumulative effects of including electron wall emission and ions/electrons collisions with a background neutral gas (recycled at the walls) are also assessed. It is found that a non-negligible energy flux affects the shadowed regions of the monoblocks, especially when accounting for collisions, and that the ion impact distribution functions are strongly influenced by the considered kinetic effects, with important implications on the induced sputtering yield.

Under the auspices of EUROfusion, the ITER baseline (IBL) scenario has been jointly investigated on AUG and TCV in the past years and this paper reports on the developments on TCV. Three ITER shapes, namely the JET, AUG and ITER IBL have been reproduced in TCV, illustrating that the higher the triangularity the larger the ELM perturbation and the more difficult it is to reach stationary states with q₉₅< 3.6. It is found that the performance of TCV IBL is mainly limited by (neoclassical) tearing modes, in particular 2/1 modes which are triggered after a large ELM. It is demonstrated that the shorter the ELM period the larger β_N at the NTM onset. We show that these modes can be avoided with central X3 EC heating at relatively high q₉₅ and moderate β_N. However, the lack of significant ECH at the high central densities obtained in TCV IBL scenario limits the duration of low q₉₅ cases to about four confinement times. During this time, density usually keeps peaking until (neoclassical) tearing modes are triggered. Nevertheless, the TCV IBL database covers the ITER target values ($H_{98y2}\sim$1, $\beta_N\sim$1.8 at $q_{95}\sim$3) and a slightly better confinement than requested for ITER is reported. Integrated modelling results show that ITG modes are the dominant instabilities, and show that, in TCV, fuelling also plays a role to sustain peaked density profiles. The role of profiles, sawteeth and ELMs regarding MHD stability are also discussed.

Bulk perturbations (voids or crystalline structure) inside the ablator of a capsule used for inertial confinement fusion are seeds for instabilities that can hinder the ignition. The study of these defects and their evolution during the implosion is one of the steps needed to achieve fusion. The current methods used by the field are to infer these effects indirectly with measurements of implosion velocity and neutron yield, among others. Observing them directly with an x-ray imaging diagnostic is difficult due to the small scale length of these defects. In this work we study the feasibility of a new diagnostic based on x-ray phase-contrast imaging. This technique has been demonstrated to perform better than standard x-ray absorption techniques in critical situations like this. By using a synthetic diagnostic we show the capabilities of this new possible approach and the limits in relation to the parameters of currently available laser facilities.

The separation between the last closed flux surface of a plasma and the external coils that magnetically confine it is a limiting factor in the construction of fusion-capable plasma devices. This plasma-coil separation must be large enough so that components such as a breeding blanket and neutron shielding can fit between the plasma and the coils. Plasma-coil separation affects reactor size, engineering complexity, and particle loss due to field ripple. For some plasmas it can be difficult to produce the desired flux surface shaping with distant coils, and for other plasmas it is infeasible altogether. Here, we seek to understand the underlying physics that limits plasma-coil separation and explain why some configurations require close external coils. In this paper, we explore the hypothesis that the limiting plasma-coil separation is set by the shortest scale length of the magnetic field as expressed by the $\nabla \mathbf{B}$ tensor. We tested this hypothesis on a database of $\gt$40 stellarator and tokamak configurations. Within this database, the coil-to-plasma distance compared to the minor radius varies by over an order of magnitude. The magnetic scale length is well correlated to the coil-to-plasma distance of actual coil designs generated using the REGCOIL method (Landreman 2017 Nucl. Fusion57 046003). Additionally, this correlation reveals a general trend that larger plasma-coil separation is possible with a small number of field periods.