Evidence that density shoulder broadening is dependent on high main-chamber neutral density is presented. Shoulder broadening does not occur when the sources for main-chamber neutrals are minimized using divertor baffles and wide gaps to the first wall (∼3× the density decay length). Removing the baffles or reducing the gap to the inner wall both act to increase the density shoulder amplitude in otherwise identical TCV discharges. Radial turbulent transport is correlated with shoulder amplitude.

A set of high density, highly shaped H-mode discharges has been performed in the TCV tokamak with the aim of assessing the effects of increasing divertor neutral recycling on the properties of upstream inter-ELM scrape-off layer (SOL) profiles and transport. An increase of divertor neutral pressure has been correlated with the evolution of separatrix properties and turbulence level. The latter has been quantified by means of the α t parameter introduced in (Eich 2020 Nucl. Fusion 60 056016), describing the contribution of resistive-interchange turbulence in the SOL relative to drift wave transport. The analysis reveals a general broadening of the upstream SOL profiles as α t increases, with the SOL power width measured by the vertical IR thermography system increasing significantly. In a similar way, the upstream density profile widens in the near SOL, whereas in the far SOL a density shoulder is observed to progressively form and increase in amplitude. This behaviour is associated with an enhancement of far SOL turbulent transport in the form of blob-filaments travelling radially faster across the far SOL and becoming bigger at higher α t. The detected filaments, evaluated from the fast reciprocating probe at the outer midplane, are determined to mostly belong to the resistive ballooning and resistive X-point regimes.

The tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e / T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

Fusion power plant designs based on magnetic confinement, such as the tokamak design, offer a promising route to sustainable fusion power but require robust exhaust solutions capable of tolerating intense heat and particle fluxes from the plasma at the core of the device. Turbulent plasma transport in the region where the interface between the plasma and the materials of the device is handled - called the divertor volume - is poorly understood, yet impacts several key factors ultimately affecting device performance. In this article a comprehensive study of the underlying physics of turbulence in the divertor volume is conducted using data collected in the final experimental campaign of the Mega Ampere Spherical Tokamak device, compared to high fidelity nonlinear simulations. The physics of the turbulence is shown to be strongly dependant on the geometry of the divertor volume - a potentially important result as the community looks to advanced divertor designs with complex geometry for future fusion power plants. These results lay the foundations of a first-principles physics basis for turbulent transport in the tokamak divertor, providing a critical step towards a predictive understanding of tokamak divertor plasma solutions.

Probe measurements of plasma fluctuations in the divertor region of ASDEX Upgrade reveal the existence of two quiescent regions close to the separatrix: one on the low field side scrape-off layer and one on the high field side of the private flux region. The X-point manipulator at ASDEX Upgrade provides the unique possibility of obtaining continuous measurements of the ion-saturation current from the low field side scrape-off layer through the private flux region and into the high field side scrape-off layer.

The here presented work studies the dynamics of filaments using 3D fluid simulations in the presence of detached background profiles. It was found that evolving the neutrals on the time-scale of the filament did not have a significant impact on the dynamics of the filament. In general a decreasing filament velocity with increasing plasma background density has been observed, with the exception of detachment onset, where a temporarily increase in radial velocity occurs. The decreasing trend with temporary increase was found for filaments around the critical size and larger, while smaller filaments where less affected by detachment. With detachment the critical filament size increased, as larger filaments were faster in detached conditions. This breaks the trend of attached conditions, where the critical size decreases with increasing density.

Interaction of coherent structures known as blobs in the scrape-off layer of magnetically confined plasmas is investigated. Isolated and interacting seeded blobs, as well as full plasma turbulence, are studied by two-dimensional numerical simulations. The features of the blobs (position, size, amplitude) are determined with a blob tracking algorithm, which identifies them as coherent structures with amplitudes above a chosen particle density threshold, and their velocities are compared to a conventional center of mass approach. We find that the theoretical velocity-size scaling dependence for isolated blobs is correctly resolved by the blob tracking method. The benchmarked approach is then extended to a population of interacting plasma blobs with statistically distributed amplitudes, sizes, and initial positions for different levels of blob interaction. We observe a correlation between the level of blob interaction and the number of blobs deviating from size–velocity scaling laws of perfectly isolated blobs. This is found to be caused by the interaction of blobs with the electrostatic potential of one another, leading to higher average blob velocities. We introduce a model specific intermittency parameter, quantifying the degree of blob interaction. For interacting blobs, we estimate the deviation from the picture of perfectly isolated blobs as a function of the intermittency parameter. For full plasma turbulence simulations, we observe a strong correlation between the blob amplitudes, sizes, and velocities estimated by the blob tracking algorithm.

Fusion energy is an area of active development and innovation worldwide, with many design concepts studied, each exhibiting a range of technical challenges. A significant portion of technical challenges will be unique for a given design concept; however, there are several overarching challenges that any design must address to some degree. These include tritium handling and the tritium cycle; materials and their survivability in the high-energy neutron environment of D-T fusion; neutronics and the validation of nuclear data; remote handling and maintenance activities; and integrated holistic approaches to fusion plant design. This paper provides an overview of these aspects for magnetic and inertial fusion approaches with a view to highlighting commonality and the benefits of shared knowledge that this may bring. This article is part of a discussion meeting issue 'Prospects for high gain inertial fusion energy (part 2)'.

The effect of plasma shaping on scrape-off layer (SOL) plasma turbulence is investigated through a rigorous validation exercise. Two- and three-dimensional simulations of the SOL plasma dynamics in three TCV limited discharges are carried out with the GBS code. These discharges realize an almost circular magnetic equilibrium, an elongated equilibrium, and an elongated equilibrium with negative triangularity. For the three plasma discharges, three simulations are performed, considering (i) a three-dimensional model with an explicit dependence on elongation, triangularity, and the inverse aspect ratio; (ii) a circular three-dimensional model in the infinite aspect ratio limit; and (iii) a two-dimensional model, which assumes a circular magnetic equilibrium in the infinite aspect ratio limit, cold ions, and interchange driven turbulence in the sheath limited regime. Ten validation observables common to simulations and experimental measurements from a reciprocating probe located at the TCV outer midplane are identified, and the agreement between experimental and numerical results relative to each observable is evaluated. The composite metric introduced by Ricci et al. [Phys. Plasmas 18, 032109 (2011)] is then used to assess the overall agreement between simulations and experimental measurements. It is found that the shaping model implemented in GBS improves the description of SOL plasma turbulence, taking into account the impact of elongation and triangularity.

This paper presents the effect of self-consistent plasma backgrounds including plasma-neutral interactions, on the dynamics of filament propagation. The principle focus is on the influence of the neutrals on the filament through both direct interactions and through their influence on the plasma background. Both direct and indirect interactions influence the motion of filaments. A monotonic increase of filament peak velocity with upstream electron temperature is observed, while a decrease with increasing electron density is observed. If ordered by the target temperature, the density dependence disappears and the filament velocity is only a function of the target temperature. Smaller filaments keep a density dependence, as a result of the density dependence of the plasma viscosity. The critical size δ^*, where filaments are fastest, is shifted to larger sizes for higher densities, due to the plasma viscosity. If the density dependence of the plasma viscosity is removed, δ^* has no temperature dependence, but rather a density dependence.

A detailed cross-device investigation on the role of filamentary dynamics in high-density regimes has been performed within the EUROfusion framework, comparing the ASDEX Upgrade (AUG) and TCV tokamaks. Both devices run density ramp experiments at different levels of plasma current, keeping the toroidal field or q95 constant in order to disentangle the role of the parallel connection length and the current. During the scan at a constant toroidal field, in both devices the scrape-off layer (SOL) profiles tend to develop a clear SOL density shoulder at a lower edge density whenever the current is reduced. Different current behaviour is substantially reconciled in terms of the edge density normalized to the Greenwald fraction. During the scan at constant q95 AUG exhibits similar behaviour, whereas in TCV no upstream profile modification signature has been observed at lower current levels. The latter behaviour has been ascribed to the lack of target density rollover. The relation between the upstream density profile modification and detachment condition has been investigated. For both devices the relation between blob size and the SOL density e-folding length is found independent of the plasma current, with the observation of a clear increase in blob size and the edge density normalized to a Greenwald fraction. ASDEX Upgrade has also explored filamentary behaviour in the H-mode. The experiments in AUG have focused on the role of neutrals, performing discharges with and without cryogenic pumps, highlighting how high neutral pressure, not only in the divertor but also at the midplane, is needed in order to develop an H-mode SOL profile shoulder in AUG.

A new tomographic inversion technique is presented for the identification of plasma filaments in wide-angle visible camera data. The technique works on the assumption that background subtracted images of filaments can be represented as a superposition of uniformly emitting magnetic equilibrium field lines. A large collection of equilibrium magnetic field lines is traced and projected onto the camera field of view and combined to form a geometry matrix describing the coordinate transformation from magnetic field aligned coordinates to image pixel coordinates. Inverting this matrix enables the reprojection of the emission in the camera images onto a field aligned basis, from which filaments are readily identifiable. The inversion is a poorly conditioned problem which is overcome using a least-squares approach with Laplacian regularization. Blobs are identified using the "watershed" algorithm and 2D Gaussians are fitted to get the positions, widths, and amplitudes of the filaments. A synthetic camera diagnostic generating images containing experimentally representative filaments is utilized to rigorously benchmark the accuracy and reliability of the technique. 74% of synthetic filaments above the detection amplitude threshold are successfully detected, with 98.8% of detected filaments being true positives. The accuracy with which filament properties and their probability density functions are recovered is discussed, along with sources of error and methods to minimize them.

In ITER and DEMO, various control objectives related to plasma control must be simultaneously achieved by the plasma control system (PCS), in both normal operation as well as off-normal conditions. The PCS must act on off-normal events and deviations from the target scenario, since certain sequences (chains) of events can precede disruptions. It is important that these decisions are made while maintaining a coherent prioritization between the real-time control tasks to ensure high-performance operation. In this paper, a generic architecture for task-based integrated plasma control is proposed. The architecture is characterized by the separation of state estimation, event detection, decisions and task execution among different algorithms, with standardized signal interfaces. Central to the architecture are a plasma state monitor and supervisory controller. In the plasma state monitor, discrete events in the continuous-valued plasma state are modeled using finite state machines. This provides a high-level representation of the plasma state. The supervisory controller coordinates the execution of multiple plasma control tasks by assigning task priorities, based on the finite states of the plasma and the pulse schedule. These algorithms were implemented on the TCV digital control system and integrated with actuator resource management and existing state estimation algorithms and controllers. The plasma state monitor on TCV can track a multitude of plasma events, related to plasma current, rotating and locked neoclassical tearing modes, and position displacements. In TCV experiments on simultaneous control of plasma pressure, safety factor profile and NTMs using electron cyclotron heating (ECH) and current drive (ECCD), the supervisory controller assigns priorities to the relevant control tasks. The tasks are then executed by feedback controllers and actuator allocation management. This work forms a significant step forward in the ongoing integration of control capabilities in experiments on TCV, in support of tokamak reactor operation.

In this paper we validate the finite element code RAPLICASOL, which models radiofrequency wave propagation in edge plasmas near ICRF antennas, against calculations with the TOPICA code. We compare the output of both codes for the ASDEX Upgrade 2-strap antenna, and for a 4-strap WEST-like antenna. Although RAPLICASOL requires considerably fewer computational resources than TOPICA, we find that the predicted quantities of experimental interest (including reflection coefficients, coupling resistances, - and -matrix entries, optimal matching settings, and even radiofrequency electric fields) are in good agreement provided we are careful to use the same geometry in both codes.

The role of magnetic perturbations generated by filaments in the scrape-off layer is investigated by performing simulations of 3D seeded filaments with an electromagnetic numerical code which includes sheath boundary conditions. Depending on the plasma β, three smoothly connecting regimes were identified: an electrostatic regime where the magnetic field is substantially unaffected by the motion of the filament, an intermediate regime where the magnetic field is excited but relaxes within the filament’s lifetime and an electromagnetic regime where the frozen-in condition is well satisfied. The electromagnetic effects are found to have a significant impact on the filament dynamics at values of β relevant for fusion devices, including increased midplane radial velocity, bending of the filament and braiding of magnetic field lines. Importantly, a finite Alfvén speed can allow electrical disconnection of the target and the upstream part of the filament, thus removing the slowing down effect associated with the sheath and producing faster filaments and more intense cross-field transport in perturbations that were affected by the sheath in the β = 0 limit.

This paper presents results of JET ITER-like wall L-mode experiments in hydrogen and deuterium (D) plasmas, dedicated to the study of the isotope dependence of ion heat transport by determination of the ion critical gradient and stiffness by varying the ion cyclotron resonance heating power deposition. When no strong role of fast ions in the plasma core is expected, the main difference between the two isotope plasmas is determined by the plasma edge and the core behavior is consistent with a gyro-Bohm scaling. When the heating power (and the fast ion pressure) is increased, in addition to the difference in the edge region, also the plasma core shows substantial changes. The stabilization of ion heat transport by fast ions, clearly visible in D plasmas, appears to be weaker in H plasmas, resulting in a higher ion heat flux in H with apparent anti-gyro-Bohm mass scaling. The difference is found to be caused by the different fast ion pressure between H and D plasmas, related to the heating power settings and to the different fast ion slowing down time, and is completely accounted for in non-linear gyrokinetic simulations. The application of the TGLF quasi-linear model to this set of data is also discussed.