TCV experiments and novel theoretical studies have addressed key questions regarding the operational space of negative triangularity (NT) plasmas in view of an L-mode NT reactor, in terms of stability, performance properties and core-edge integration. We show that reducing the top triangularity (?top) to more negative values induces an H-L back transition, confirming the direct dependence between H-mode existence and access to 2nd ballooning stability region. We also show that the X-point ? can prohibit H-mode access if sufficiently negative. Using these conditions to stay in L-mode, we show the sustainment of a stationary high ?N NT plasma, with H98y2 > 1, using real-time ? control with NBI as an actuator. This plasma has Ti > Te over the whole profile. Gyrokinetic simulations show that the benefit of NT might be lost for aspect ratio <2.5, in the TEM-dominated microturbulence regime and that the improved confinement is primarily due to the non X-point triangularity (top ? in single-null down divertor, SND). Global gyrokinetic studies also show that this improvement does not change with machine size (independent of ?*). These studies as well as other results presented at this conference allow us to propose in the conclusion a procedure for predicting and optimizing NT reactor plasmas which can be tested in present NT experiments.

Runaway electrons (REs) are a concern for fusion reactors from the startup to the termination of tokamak discharges. The sudden localized loss of a multi-MA RE beam can inflict severe damage to the first wall. The Tokamak à Configuration Variable (TCV) has recently explored various aspects of RE physics, designing and testing multiple strategies to prevent the formation of REs, reduce their number, or mitigate their effect. A density threshold above which RE generation can be avoided is identified and a corresponding effective critical field is determined. Resonant interaction between RE cyclotron motion and magnetic ripple oscillations is found to limit RE energy and illustrates how pitch-angle scattering can impede RE growth. RE confinement in low density TCV plasmas is good enough for the RE population to grow until it drives most of the toroidal current. The application of central electron cyclotron resonance heating under such conditions if found to enhance RE losses strongly enough for the entire RE population to be expelled in a few hundred milliseconds. RE beam formation can take place following disruptions triggered by massive gas injection. Full conversion to RE-driven current is observed. Successfully tested strategies to mitigate a post-disruption RE beam include safe ramp-down, regeneration of a healthy hot plasma, and most recently benign termination following massive injection of low-Z material. Combined achievements in avoiding startup RE generation, reducing an established flattop RE population, and mitigating a post-disruption RE beam, have furthered the prospects in safe operation for future tokamak reactors.

One of the crucial challenges on the path to a demonstration tokamak fusion power plant is the integration of long-pulse operation with high performance, compatible with low disruptivity and an acceptable power flux to the device wall. These issues are being addressed on several devices within the remit of the Tokamak Exploitation Work Package (WPTE) of EUROfusion. This contribution reports on progress on the TCV and MAST-U tokamaks in this area. These two devices share a number of features - they are similarly medium-sized, possess a carbon first wall, and are equipped with poloidal-field coil sets enabling advanced divertor configurations - but are complementary in their very dissimilar aspect ratios: conventional for TCV, tight for the spherical tokamak MAST-U. High-beta, potentially steady-state candidate scenarios for long-pulse operation are being studied in discharges heated with electron-cyclotron resonance heating (ECRH) and neutral beam injection (NBI) in TCV and by NBI only in MAST-U. On TCV, two avenues have been explored in parallel, with the ultimate aim of merging them into a single optimized scenario. The first avenue has added NBI to well-established, fully non-inductive, steady-state electron internal transport barriers (eITBs) originally sustained by ECRH alone; the other, conversely, starts with an equally well-established H-mode regime powered by NBI, to which ECRH is added to increase the non-inductive current fraction. Values of the normalized beta, ?N, up to 2.0 have been achieved transiently, and up to 1.8 in fully non-inductive conditions. On MAST-U, efforts have focused on systematic parameter scans in the baseline, up-down-symmetric double-null H-mode scenario, yielding thus far a maximum transient ?N~3.5. This work is due to continue in current and future campaigns.

Experiments, gyrokinetic simulations and transport predictions have been performed to investigate if a negative triangularity (NT) L-mode option for the DTT full power scenario would perform similarly to the positive triangularity (PT) H-mode reference scenario, avoiding the harmful edge localized modes (ELMs). The simulations show that if a beneficial effect of NT would not come from the edge-SOL region ?!"# > 0.9, one should renounce to part of the PT H-mode performance to avoid the ELMs. New dedicated experiments at TCV and AUG, with similar shapes and transport regimes, show that even with the small triangularity of the considered DTT scenarios, a beneficial effect of NT could come from ?!"# > 0.9, allowing NT L-modes to overperform corresponding PT L-modes, reaching central pressures comparable to larger power PT H-modes. For AUG, a similar central plasma performance is found with PT and NT for a PT/NT ELMy H-mode/H-mode couple of pulses and for a PT/NT non-ELMy H-mode/L-mode one, both with ECRH only. The beneficial effect of NT was shown by gyrokinetic runs to be due to an increase of the critical temperature logarithmic gradient when inverting the triangularity, with similar stiffness for positive and negative triangularity. The effect has been found in both ITG- and TEMdominant micro-instability regimes.

The tokamak `a configuration variable (TCV) is a small-sized tokamak, where finite size effects (often called 'rho-star' or 'global' effects) could significantly impact the heat and particle fluxes, leading to discrepancies between gyrokinetic flux-tube results and global ones (McMillan et al 2010 Phys. Rev. Lett. 105 155001). The impact of global effects on the radial profile of the plasma density has been investigated in a previous study for a particular TCV discharge with negligible particle source, satisfying the 'zero particle flux' (ZPF) condition. A radially local flux-tube analysis, reconstructing the dependence of the peaking of the density profile on the main physical parameters, invoking the ZPF constraint, was pursued close to mid-radius in (Mariani et al 2018 Phys. Plasmas 25 012313). This analysis was followed by a global one (Mariani et al 2019 Plasma Phys. Control. Fusion 61 064005), where local quasi-linear (QL) and nonlinear (NL) results were compared with global simulations, showing small global effects on the density peaking. However, these gradient-driven (GD) global runs considered Krook-type heat and particle sources to keep temperature and density profiles fixed on average, which differ from the experimental radially localized sources. To remove this possible bias on the results, a different evaluation of the density peaking for the same case is performed here, based on global NL hybrid simulations where the temperature profiles are [still] kept fixed with the Krook-type sources, however the density profile relaxes in a flux-driven way (with zero particle source). The new hybrid simulations show a good agreement with the old GD runs. A global QL model is also developed and applied using the output from linear global runs, to estimate ratios of fluxes, showing a good agreement with the flux-tube results of global NL GD simulations. The effect of collisions on the results is also investigated, in order to evaluate their impact on the radial variation of the density peaking.

One of the major problems for future tokamak devices are ELMs (Edge Localized Modes) as they can lead to large, uncontrolled heat fluxes at the machine targets. For this reason, different techniques and alternative magnetic configurations are under study to mitigate or avoid these phenomena. One of the most promising among these studies is the Negative Triangularity (NT) configuration, which exhibits a global confinement comparable with H-Mode operation and, staying in L-mode, could enable an easier power exhaust dissipation due to a possible bigger heat flux decay length with respect to the conventional Positive triangularity (PT) H-mode. In this work, the fluid code SOLEDGE2D-EIRENE is used to study edge transport. Studies are made on discharges performed in the TCV device (Tokamak a configuration variable), which can create a variety of different plasma geometries thanks to 16 independently powered poloidal field coils and its open vacuum vessel. In order to understand if and how power and particle exhaust in NT differs with respect to those of the PT shape, four discharges in single null magnetic divertor configuration with fixed lower triangularity (?bot = +0.5) but with different upper triangularity (from ?up = 0.28 to ?up = +0.45) have been modelled. All of them are ohmically heated, L-mode deuterium plasmas, and in the high recycling regime. Moreover, these discharges were previously used in [4] to measure the heat flux decay length by IRT (Infrared Thermography), allowing us to make comparisons with modelling results.

ITER will operate at low pedestal collisionality (?*eeped) and high separatrix density (n esep). The ITER pedestal collisionality is supposed to be sufficiently low (?*eeped<1) that the pedestal will be limited by the peeling instabilities, rather than ballooning instabilities. Most of the present days machines, in particular in Europe, tend to operate at higher pedestal collisionality, with the ELMs typically triggered by the balloning modes. While pedestal physics has been well studied at the ballooning boundary, so far information on the pedestal behaviour at the peeling boundary has been described mainly from DIII-D [1]. The aim of this work is to investigate the pedestal behaviour at the peeling boundary in TCV, with particular emphasis to the pedestal structure, pedestal stability and their links with pedestal density and separatrix density. The use of both neutral beam heating and ECRH heating in TCV at Ip=150kA, Bt=1.4T has allowed to reach a low collisionality regime characterized by ?*eeped?0.1-0.9, both at low and at high triangularity. Figure 1 shows the pedestal electron pressure (peped) and density (neped) for the low ?*eeped / high-? plasma #71472 (red square). The blue data show the corresponding pedestal pressure predicted with the Europed code [2] and the corresponding most unstable mode. The experimental pedestal is close to the peeling boundary, as shown by the fact that the most unstable mode has a rather low toroidal number (n=7). The work will investigate the change of the pedestal structure and of the pedestal stability with increasing n eped and with increasing n esep. The work will also highlight the difference at the peeling boundary between high and low triangularity. The investigation, in particular the role of n esep, is particularly useful to test the pedestal predictions in ITER-relevant regimes.

Runaway electrons (RE) pose a significant threat for ITER and a mitigation technique is yet to be validated. Preliminary experiments on DIII-D and JET showed that reduced damage to the vessel can be achieved by accessing MHD instabilities that lead to spreading of the RE beam impact area [Paz-Soldan 2019, Reux 2021]. These instabilities were attained by "flushing" impurities and recombining the companion plasma, thus reducing the resistance and electric field (E), and sustaining beams until a low edge safety factor (qa) was achieved. Regeneration of the REs did not occur during the current quench due to the low E. This approach was successfully applied on ASDEX Upgrade and TCV this year. On-going experiments have been exploring the possible operational space and the underling physics involved and this paper will report the preliminary findings and outlook of this program. RE beams were generated via massive gas injection (MGI) of argon on ASDEX Upgrade and neon on TCV. Secondary injections of D2 to flush impurities and recombine the companion plasma were executed with combinations of MGI, fueling pellets and fueling valves. Experiments on TCV showed that neutral pressure was more directly linked to plasma recombination than injected quantity. This was due to pumping effects that become non-negligible with beam durations in the order of seconds. It was found that neutral pressures of 0.10-0.15 Pa were required on both machines to recombine companion plasmas with RE currents of 120-600 kA. This study was further extended on TCV through variations in injected impurity gas quantity and the use of H2 instead of D2 as the secondary injection gas. It was found that higher neutral pressures were required to recombine the plasma when higher neon quantities were injected and no significant differences between H2 and D2 secondary injection were observed. Fueling valves were used to maintain the neutral pressure required to prevent re-ionisation of the companion plasma and access a low E. This led to a record RE beam duration of ~4s on ASDEX Upgrade and the ability to increase the RE current on both machines. Reduction of qa was achieved via plasma current ramps on TCV and compression of the plasma onto the central column on both machines. Beam currents, current ramp rates and compression rates were scanned. Most beams terminated at a qa of ~2, with the exception of a 200kA beam on AUG, compressed over 500 ms, which produced a benign termination at a qa of ~3. Experiments compressing partially recombined plasmas have also begun on both machines and differences in wetted area and RE beam energy for flushed and unflushed plasmas have been measured on both machines. Preliminary results show significantly reduced heat fluxes to the inner wall and a reduction in beam energy during the recombined companion plasma phase. Further experiments are planned on both machines for later this year. These experiments will focus on benign termination of partially recombined companion plasmas, scans of compression rates and variations in secondary injection gas species and concentration.

Understanding the plasma dynamics in the edge and Scrape-Off Layer (SOL) regions of a tokamak is mandatory for future exploitation of fusion power as a source of energy, since the SOL transport properties determine the peak particle and heat fluxes flowing towards the plasma-facing components (PFCs). The inherently turbulent nature of SOL transport makes this assessment particularly challenging. In the first part of this contribution, an investigation is reported on a set of H-mode discharges on the TCV tokamak, focused on the role of gas fueling in determining the properties of transport and kinetic profiles in the near and far SOL. In all discharges additional NBI heating was used for H-mode access, while different gas settings allowed to span a wide interval of divertor neutral pressure ... . The analysis has been conducted by first estimating the SOL power width ?... from a fit of the upstream-remapped parallel heat flux profile as measured by IR cameras. This allowed to estimate the separatrix temperature T...,?... under the assumption of conduction-dominated parallel transport (justified by the considerably lower temperatures at the target than upstream). After fitting the Thomson Scattering density and temperature profiles, the separatrix position and density n...,?... could be estimated. The level of turbulent transport has been quantified by the ?? ? Z... R...q...?...? n...,?...T...,?... ? parameter introduced in [1], this being a measure of the effect of the interchange instability drive on drift waves and thus effectively an estimate of the interchange turbulence level, being directly related to the separatrix collisionality. An increase in p...,?...? is seen to translate into an increase of n...,?... and ultimately of ??. Limited to the high-density part of the database, a p...,?...? scan between 20 and 120 mPa produces a variation of ?? between 0.4 and 0.95. This is accompanied by a broadening of ?...by up to a factor ~2.5 (from ~4 to ~10 mm), coherently with recent modelling results predicting a larger SOL power width in H-mode at high separatrix collisionality [2]. A similar increase of the near-SOL electron density, temperature and pressure e-folding lengths is observed as well. In the far-SOL the density profile shows a progressive flattening at increasing ??, leading to formation of the so-called density shoulder. This observation can be accounted for by a progressive increase of filamentary transport into the far SOL at higher ??, as observed by wall-mounted and reciprocating Langmuir Probes. Given that ?? shows an explicit dependence on q...?... , a scan in upper triangularity has been performed in H-mode on TCV as well, where ?... has been varied between 0 and ~0.45. Preliminary results show a transition from a Type-I ELMy regime towards a small ELM/QCE regime as ??... increases at similar fuelling levels, coherently with previous experimental observations [3, 4]. In the second part of this work the effect of such a ??... variation on ?? and consequently on SOL profile and turbulence properties will be investigated.

ITER will operate at low pedestal collisionality (?*eeped) and high separatrix density (n esep). The ITER pedestal collisionality is supposed to be sufficiently low (?*eeped<1) that the pedestal will be limited by the peeling instabilities, rather than ballooning instabilities. Most of the present days machines, in particular in Europe, tend to operate at higher pedestal collisionality, with the ELMs typically triggered by the balloning modes. While pedestal physics has been well studied at the ballooning boundary, so far information on the pedestal behaviour at the peeling boundary has been described mainly from DIII-D [1]. The aim of this work is to investigate the pedestal behaviour at the peeling boundary in TCV, with particular emphasis to the pedestal structure, pedestal stability and their links with pedestal density and separatrix density. The use of both neutral beam heating and ECRH heating in TCV at Ip=150kA, Bt=1.4T has allowed to reach a low collisionality regime characterized by ?*eeped?0.1-0.9, both at low and at high triangularity. Figure 1 shows the pedestal electron pressure (peped) and density (neped) for the low ?*eeped / high-? plasma #71472 (red square). The blue data show the corresponding pedestal pressure predicted with the Europed code [2] and the corresponding most unstable mode. The experimental pedestal is close to the peeling boundary, as shown by the fact that the most unstable mode has a rather low toroidal number (n=7). The work will investigate the change of the pedestal structure and of the pedestal stability with increasing n eped and with increasing n esep. The work will also highlight the difference at the peeling boundary between high and low triangularity. The investigation, in particular the role of n esep, is particularly useful to test the pedestal predictions in ITER-relevant regimes.

Disruptions represent one of the highest concerns for next-step fusion devices based on the tokamak principle. Active disruption avoidance and off-normal event handling strategies need to be envisaged and carefully designed in modern Plasma Control Systems (PCS) to monitor and predict when the plasma approaches operational boundaries. In the recent years, in the context of TCV internal and EUROfusion WPTE framework programs, several real-time control algorithms for disruption avoidance and prevention, such as for NTM stabilization and control as well as for H-mode density limit (HDL) active avoidance, have been successfully developed and embedded in the TCV real-time plasma supervision system (SAMONE). An NBI-heated scenario for the HDL has been newly developed and tested in a large number of experiments carried out for different plasma currents and divertor baffles configurations, allowing to reproduce the same physics characteristic phenomenology observed also in other devices. A first demonstration of the concept of portability across different devices has been achieved by transferring from AUG to TCV the same control algorithm based on the distance with respect to an empirically defined disruption boundary in the space of the H-Mode confinement factor (H98y,2) and a normalized line integrated edge electron density. Such a distance metrics, combined with a deep learning model for energy confinement-state detection, allows to react in real-time activating different control tasks regulating NBI power and gas flux with the objective of recovering from the strong confinement degradation observed when approaching the density limit. This contribution will present an overview of the experimental results, modelling activities supporting density limit scalings, advances in algorithms for detection of proximity to operational limits as well as the advances in the development of a generic control architecture enabling the integration of active disruption avoidance strategies and exception handling.

The tokamak a 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) similar to 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.

Experimental investigation and gyrokinetic simulations of multi-scale electron heat transport in JET, AUG and TCV.

A complete understanding of pedestal structure and limits remains an open question in fusion research. Research in this direction generally makes use of some generalised version of the EPED model[1]; transport mechanisms (which can have different drives for the particles and heat, and for ions and electrons) determine, in combination with the heat and particle sources, the profile gradients. In the "standard" case, the profiles continue to extend radially further inwards away from the separatrix until there is enough free energy in the pressure gradient (and associated driven edge current density) that a large magnetohydrodynamic (MHD) instability occurs, the so-called edge localised modes (ELMs). The onset of the ELM then determines the steepest and widest pedestal pressure profile.

Next-generation tokamaks will require integrated plasma control solutions beyond the present state-of-the art. Such solutions are being actively explored on the TCV and ASDEX Upgrade tokamaks. A summary of recent achievements is provided in this paper. We report on advances in plasma state reconstruction algorithms including improved density profile estimations on ASDEX Upgrade and a first implementation of kinetic equilibrium reconstruction constrained by real-time transport equation solutions on TCV. Recent progress in development of generic plasma supervisory control system on TCV, including a mix of continuous control and off-normal event handling is discussed, together with its application to high-density limit disruption avoidance.The RABBIT code for real-time NBI deposition calculations has been integrated in the ASDEX Upgrade control system and quantities computed by the code have been used for real-time feedback control of the power deposited to the ions.Finally, we report on a first demonstration of model-based shot-to-shot iterative improvement of actuator trajectories,applied to control of the central ion and electron temperature by an appropriate mix of ECH and NBI.

Tokamaks dominated by electron heating like ITER could possibly suffer from the consequences of an elec-tron temperature gradient (ETG) mode destabilisation, which could develop a turbulent electron heat flux ca-pable of setting an upper limit to the achievable electron temperature peaking, resulting in a degradation of thefusion performances. An effort is carried out in the paper to collect and compare the results of dedicated plasmadischarges performed during the last years at three of the major European tokamaks, TCV, AUG and JET, byanalysing the electron heat transport for cases presumably compatible with ETGs relevance given the actual the-oretical understanding of these instabilities. The response of the electron temperature profiles to electron heatflux changes is experimentally investigated by performing both steady state heat flux scans and perturbativeanalysis by radio frequency modulation. The experimental results are confronted with numerical simulations,ranging from simpler linear gyrokinetic or quasi-linear runs, to very computationally expensive nonlinear multi-scale gyrokinetic simulations, resolving ion and electron scales at the same time. The results collected so fartend to confirm the previously emerging picture indicating that only cases with a proper balance of electron andion heating, with similar electron and ion temperatures and sufficiently large electron temperature gradient, arecompatible with a non negligible impact of ETGs on the electron heat transport. The ion heating destabilisesETGs not only by increasing the ion temperature but also thanks to the stabilisation of ion-scale turbulence bya synergy of fast ions andE×Bshearing which are in some cases associated to it. The stabilising effect ofplasma impurities on ETGs is still under investigation by means of multi-scale gyrokinetic simulations, and alsodirect experimental measurements of density and temperature fluctuations at electron scales would be needed toultimately assess the impact of ETGs.