TCV is a medium sized tokamak equipped with a large suite of plasma diagnostics, a versatile array of poloidal shaping coils, an electron cyclotron heating system, and a two source neutral beam injection (NBI) system. The NBI system is capable of a simultaneous injection of energetic neutrals in the co-and counter-current directions. The resulting fast ion (FI) populations are used to study a multitude of FI-driven instabilities. While plasma instabilities in the below 1 MHz frequency range can be observed via standard diagnostics such as soft x-ray detectors, magnetics, fast ion loss detectors, and reflectometers, fluctuations in the range of >10 MHz require a specialized diagnostic. For this purpose the TCV tokamak has been equipped with a dedicated ion cyclotron emission (ICE) diagnostic, following a similar approach to AUG [1] and W7-X [2]. The diagnostic consists of a pair of magnetic coils (8 turns, 177 nH, 16 mm long, 7 mm in diameter), oriented orthogonally to each other to detect magnetic field fluctuations in the toroidal and poloidal directions. The coils are housed in a stainless steel electrostatic shield with a slit and are installed on the torus low field side at the midplane location, behind graphite protection tiles. The coil electric outputs are routed to the vacuum feedthroughs via a pair of coaxial cables, at which point one of the outputs is grounded at the feedthrough, from the airside. The second output is then routed to a rectifying radio frequency wave detector and a fast digitizer in the diagnostics rack. The rectifying detector is sensitive to signals in the 10-100 MHz frequency range and can measure the signal amplitude and the phase difference between the two probes, while the frequency information of the signal is lost. The benefit of this detection method is that the output signal can be digitized at a “slow” speed (200 kHz in the case of TCV) with a low cost digitizer and the data volume per plasma discharge is low. The fast digitizer, on the other hand, preserves the frequency information, albeit with a larger data volume. For the case of TCV, a 250 MHz sampling rate digitizer has been temporarily used, while a dedicated 1 GHz digitizer is currently being implemented. First results of high frequency (>10 MHz) instabilities detected in the presence of energetic ions will be presented.

A multispectral camera setup is used to infer a 2D map of plasma parameters in a tokamak from spectral emissions. However, the light measured by these cameras is line integrated in the toroidal direction, whereas emissivities on the poloidal plane are necessary for the inference. The poloidal plasma emissivity can be obtained by tomographic reconstruction, but classical techniques are too slow to use these emissivities for real-time control. We present two machine learning based approaches to accelerate the reconstruction of the poloidal emissivities from line integrated data measured by the camera setup. Both approaches yield more accurate results on synthetic data than the iterative approach while being, with the right implementation, fast enough for real-time control applications.

Plasma edge simulations with the SOLPS-ITER code are performed to study the influence of divertor baffles on nitrogen-seeded detachment in TCV single-null, L-mode discharges. Scans of nitrogen seeding rate are conducted in both baffled and unbaffled TCV divertors, where the nitrogen seeding with baffles is found to yield lower target temperatures and heat fluxes than with baffles-only and with seeding-only. The cumulative effects of baffles and seeding on target parameters are explained by the two-point model. The divertor neutral density and neutral compression increase with baffles, due to lower divertor to main chamber neutral conductance, as explained by a schematic neutral transport model with baffles. The nitrogen retention, defined as the ratio of average nitrogen nuclei density in divertor and main chamber, increases with the seeding rate if baffled, and remains constant if unbaffled. At the same outboard mid-plane separatrix plasma density, the nitrogen retention with baffles is lower than the unbaffled retention at low seeding levels and is higher at high seeding levels, which is explained by the changes of nitrogen ion and neutral transport with baffles and seeding. The baffled carbon retention is higher than the unbaffled retention due to lower divertor to main chamber carbon neutral conductance. Baffles increase the divertor radiation. The predicted trends of target parameters, the distribution of neutrals and radiations are well supported by TCV experiments, though discrepancies in the absolute values remain. The simulations yield an overall colder and denser divertor, consistent with previous SOLPS-ITER simulations of Ohmically heated L-modes in TCV. The successful comparison of simulation and experiment, together with the understanding gained from the neutral transport model, increases the confidence in the SOLPS simulations for the next TCV divertor upgrade.

To reliably predict the distribution of heat and particle fluxes at the target plates of tokamaks, a comprehensive understanding of turbulence throughout the entire Scrape-Off-Layer (SOL) is imperative. This study examines divertor turbulence systematically across a broad parameter range on the TCV tokamak, including variations in magnetic field direction, plasma current I-p is an element of [140, 320] kA, edge safety factor q(95) is an element of [2.6,4.7] and Greenwald fraction f(G) is an element of [0.18,0.6]. The TCV X-point Gas Puff Imaging (GPI) system is used to measure 2D filament properties in the inner and outer divertor region. The fluctuation levels in the divertor are found to strongly increase with density (to 80% over most of the SOL) while remaining insensitive to I-p . The previously identified divertor-localized filaments (DLF), located on the bad curvature side of the outer divertor leg, are found to be a common feature on TCV, while no filaments are observed in the PFR. DLFs are present over most of the parameter space and in both field directions. However, they are absent, or appear only closer to the target, for sufficiently large Lambda(div) greater than or similar to 10 or q(95) greater than or similar to 3.7. Across both I-p and f(G) scans, some clear trends with Lambda(div) are found for divertor filament sizes and velocities, and with target fall-off lengths of density and heat flux profiles at the outer target. This study provides important experimental insights to turbulent transport in the divertor also for comparison with self-consistent, turbulence simulations and extrapolation to future reactor conditions.

The paper describes the operation and upgrade of the Heating Neutral Beam Injection (HNBI) system as well as the related infrastructure at the Tokamak Configuration Variable (TCV) during the last years. In 2023, the installation of new TCV neutron shielding was completed, enabling the operation of two deuterium heating beams at their full power for 30 (instead of 5) high-performance TCV plasma shots. The automation control capabilities of HNBIs have been enhanced through the integration of beam power scenario design into TCV discharge programming, along with real-time control of beams power based on real time plasma measurements of plasma during plasma discharge. The beam ducts were re-designed for minimization of power loads impact. The integration of thermocouples was modified for better monitoring of the ducts heating. Development of alternative ion optics for the first heating beam, designed for intermediate particle energies (up to similar to 38 keV), has commenced. The first results of a beamlet geometry numerical optimization, based on beam formation IBSimu calculations, beam transportation through the duct, are considered respective the beam in-duct losses. The prototype of spectral multichannel measurement setup for beam profile scanning and species composition evaluation was tested. The first images of the beam's spectral profile, providing fractional divergence consideration, have been analyzed.

Tokamak a configuration variable (TCV), recently celebrating 30 years of near-continual operation, continues in its missions to advance outstanding key physics and operational scenario issues for ITER and the design of future power plants such as DEMO. The main machine heating systems and operational changes are first described. Then follow five sections: plasma scenarios. ITER Base-Line (IBL) discharges, triangularity studies together with X3 heating and N2 seeding. Edge localised mode suppression, with a high radiation region near the X-point is reported with N-2 injection with and without divertor baffles in a snowflake configuration. Negative triangularity (NT) discharges attained record, albeit transient, beta(N) similar to 3 with lower turbulence, higher low-Z impurity transport, vertical stability and density limits and core transport better than the IBL. Positive triangularity L-Mode linear and saturated ohmic confinement confinement saturation, often-correlated with intrinsic toroidal rotation reversals, was probed for D, H and He working gases. H-mode confinement and pedestal studies were extended to low collisionality with electron cyclotron heating obtaining steady state electron iternal transport barrier with neutral beam heating (NBH), and NBH driven H-mode configurations with off-axis co-electron cyclotron current drive. Fast particle physics. The physics of disruptions, runaway electrons and fast ions (FIs) was developed using near-full current conversion at disruption with recombination thresholds characterised for impurity species (Ne, Ar, Kr). Different flushing gases (D2, H2) and pathways to trigger a benign disruption were explored. The 55 kV NBH II generated a rich Alfvenic spectrum modulating the FI fas ion loss detector signal. NT configurations showed less toroidal Alfven excitation activity preferentially affecting higher FI pitch angles. Scrape-off layer and edge physics. gas puff imaging systems characterised turbulent plasma ejection for several advanced divertor configurations, including NT. Combined diagnostic array divertor state analysis in detachment conditions was compared to modelling revealing an importance for molecular processes. Divertor physics. Internal gas baffles diversified to include shorter/longer structures on the high and/or low field side to probe compressive efficiency. Divertor studies concentrated upon mitigating target power, facilitating detachment and increasing the radiated power fraction employing alternative divertor geometries, optimised X-point radiator regimes and long-legged configurations. Smaller-than-expected improvements with total flux expansion were better modelled when including parallel flows. Peak outer target heat flux reduction was achieved (>50%) for high flux-expansion geometries, maintaining core performance (H-98 > 1). A reduction in target heat loads and facilitated detachment access at lower core densities is reported. Real-time control. TCV's real-time control upgrades employed MIMO gas injector control of stable, robust, partial detachment and plasma beta feedback control avoiding neoclassical tearing modes with plasma confinement changes. Machine-learning enhancements include trajectory tracking disruption proximity and avoidance as well as a first-of-its-kind reinforcement learning-based controller for the plasma equilibrium trained entirely on a free-boundary simulator. Finally, a short description of TCV's immediate future plans will be given.

Recent results from MAST Upgrade are presented, emphasising understanding the capabilities of this new device and deepening understanding of key physics issues for the operation of ITER and the design of future fusion power plants. The impact of MHD instabilities on fast ion confinement have been studied, including the first observation of fast ion losses correlated with Compressional and Global Alfvén Eigenmodes. High-performance plasma scenarios have been developed by tailoring the early plasma current ramp phase to avoid internal reconnection events, resulting in a more monotonic q profile with low central shear. The impact of m / n = 3/2, 2/1 and 1/1 modes on thermal plasma confinement and rotation profiles has been quantified, and scenarios optimised to avoid them have transiently reached values of normalised beta approaching 4.2. In pedestal and ELM physics, a maximum pedestal top temperature of ∼350 eV has been achieved, exceeding the value achieved on MAST at similar heating power. Mitigation of type-I ELMs with n = 1 RMPs has been observed. Studies of plasma exhaust have concentrated on comparing conventional and Super-X divertor configurations, while X-point target, X-divertor and snowflake configurations have been developed and studied in parallel. In L-mode discharges, the separatrix density required to detach the outer divertors is approximately a factor 2 lower in the Super-X than the conventional configuration, in agreement with simulations. Detailed analysis of spectroscopy data from studies of the Super-X configuration reveal the importance of including plasma-molecule interactions and D _2 Fulcher band emission to properly quantify the rates of ionisation, plasma-molecule interactions and volumetric recombination processes governing divertor detachment. In H-mode with conventional and Super-X configurations, the outer divertors are attached in the former and detached in the latter with no impact on core or pedestal confinement.

The Super-X Divertor (SXD) is an alternative divertor configuration leveraging total flux expansion at the Outer Strike Point (OSP). While the extended 2-Point Model (2PM) predicts facilitated detachment access and control in the SXD configuration, these attractive features are not always retrieved experimentally. These discrepancies are at least partially explained by the effect of parallel flows which, when self-consistently included in the 2PM, reveal the role of total flux expansion on the pressure balance and weaken the total flux expansion effect on detachment access and control, compared to the original predictions. This new model can partially explain the discrepancies between the 2PM and experiments performed on tokamak à configuration variable (TCV), in ohmic L-mode scenarios, which are particularly apparent when scanning the OSP major radius Rt. In core density ramps in lower Single-Null (SN) configuration, the impact of Rt on the CIII emission front movement in the divertor outer leg—used as a proxy for the plasma temperature in the divertor—is substantially weaker than 2PM predictions. Furthermore, in OSP radial sweeps in lower and upper SN configurations, in ohmic L-mode scenarios with a constant core density, the peak parallel particle flux density at the OSP is almost independent of Rt, while the 2PM predicts a linear dependence. Finally, analytical and numerical modeling of parallel flows in the divertor is presented. It is shown that an increase in total flux expansion can favour supersonic flows at the OSP. Parallel flows are also shown to be relevant by analysing SOLPS-ITER simulations of TCV.

In recent years, negative triangularity (NT) has emerged as a potential high-confinement L-mode reactor solution. In this work, detachment is investigated using core density ramps in lower single null Ohmic L-mode plasmas across a wide range of upper, lower, and average triangularity (the mean of upper and lower triangularity: delta) in the TCV tokamak. It is universally found that detachment is more difficult to access for NT shaping. The outer divertor leg of discharges with delta approximate to - 0.3 could not be cooled to below 5 eV through core density ramps alone. The behavior of the upstream plasma and geometrical divertor effects (e.g. a reduced connection length with negative lower triangularity) do not fully explain the challenges in detaching NT plasmas. Langmuir probe measurements of the target heat flux widths (lambda q ) were constant to within 30% across an upper triangularity scan, while the spreading factor S was lower by up to 50% for NT, indicating a generally lower integral scrape-off layer width, lambda int. The line-averaged core density was typically higher for NT discharges for a given fuelling rate, possibly linked to higher particle confinement in NT. Conversely, the divertor neutral pressure and integrated particle fluxes to the targets were typically lower for the same line-averaged density, indicating that NT configurations may be closer to the sheath-limited regime than their PT counterparts, which may explain why NT is more challenging to detach.

This work explains the anomalously high runaway electron (RE) pitch angles inferred in the flat-top of dedicated Tokamak à Configuration Variable (TCV) experiments. Kinetic modelling shows that the resonant interaction between the gyromotion of the electrons and the toroidal magnetic field ripple will give rise to strong pitch angle scattering in TCV. The resulting increase in synchrotron radiation power losses acts as a RE energy barrier. These observations are tested experimentally by a magnetic field ramp-down, which gradually reduces the resonant parallel momentum at which the REs interact with the ripple. Resulting changes in synchrotron emission geometry and intensity are observed using three multi-spectral camera imaging systems, viewing the RE beam at distinct spatial angles in multiple wavelength ranges. Experimental reconstructions of the RE distribution in momentum- and real-space are consistent with kinetic model predictions.

Novel power exhaust solutions are being developed to address the challenge of integrating a high performance fusion core plasma with a well-protected divertor, if the single null configuration does not scale to a reactor device. This work aims to elucidate the physics mechanisms responsible for the reduction in peak target heat flux in configurations with multiple X-points. Experimental studies on tokamak à configuration variable in the Snowflake Minus configuration are extended to a novel configuration with three nearby divertor X-points, termed a Jellyfish, allowing us to enhance the expected effects of an additional divertor X-point. These studies are complemented by simplified 1D scrape-off layer (SOL) modeling with the SPLEND1D code and by interpretative modeling with the edge transport code EMC3-EIRENE applied to the Snowflake Minus, to further elucidate some of the key underlying processes. We find that configurations with multiple nearby X-points, and increased near-SOL connection length, exhibit reductions in peak target heat flux and an earlier detachment onset compared to a reference single null configuration, consistent with expectations from SPLEND1D. A strong correlation is experimentally observed between the radially localized radiated power and connection length. While this does not necessarily map to higher total divertor radiative losses for configurations with multiple X-points, it can, at least, provide some control over the radial position of the spatial radiation distribution. Experiments are shown to exhibit radial striations in the emissivity of multiple spectral lines in the inter-null region in these configurations. Although comparisons with EMC3-EIRENE simulations support enhanced cross field transport in the inter-null region, additional transport physics is required in the model to obtain a quantitative match with experiment. No significant differences in divertor-core compatibility are attributed to the presence of additional divertor X-points. However, impurity source optimization is required in such geometries to ensure a low core impurity content is maintained.

The Tokamak a Configuration variable (TCV) is equipped with two neutral beam injection (NBI) systems delivering up to 1.2 MW each for pulse durations of up to 2 s. The first system (NBI1), designed for an injection energy in the range of 25 to 30 keV has been operational since 2016. The existing concrete neutron shielding of the experimental hall proved insufficient for fully protecting human accessible areas, limiting the number of daily plasma pulses using NBI1. The recently commissioned second system (NBI2) is designed for injection synergies in the range 50 to 60 keV. Both systems are tangentially oriented in opposite directions in order to permit experiments with low or no net torque. Calculations with the TRANSP and ORBIS heating codes show that neutron rates from deuterium-deuterium fusion reactions may be as high as 10(14) n/s, up to 10 times higher than with the lower energy beam only. This is due both to the similar to five times larger beam-plasma neutron rates from the higher energy beam and to an exceptionally high contribution from beam-beam reactions between the opposing beams. The radiation protection policy at the Swiss Plasma Center is that all staff members be considered as members of the general public, limiting the daily personal dose to 4 mu Sv. This is also the maximum admissible daily dose in any publicly accessible zone, whether occupied or not. Currently, with only the lower energy beam, this limit can be attained in the control room adjacent to the device hall after only five NBI pulses out of a possible 30 daily pulses. To allow for exploitation of the two beams at full specifications, the source side of the existing barite concrete walls of the 15 x 20 x 8 m large TCV hall will be covered with 20-cm-thick polythene (PE) cladding and a ceiling made of 35-cm-thick PE will be added. The total mass of PE will be 200 tons. The usage of PE at this scale for neutron shielding is unprecedented at any fusion research facility.

This study employs the established momentum transport analysis at ASDEX Upgrade [Zimmermann et al., Nucl. Fusion 63, 124003 (2023)] to investigate the parametric variations of the momentum transport coefficients in the core of H-mode plasmas. These experimental results are compared to a comprehensive database of gyrokinetic calculations. Generally, good agreement between predicted and measured diffusive and convective transport coefficients is found. The predicted and measured Prandtl numbers correlate most dominantly with the magnetically trapped particle fraction. The experimentally inferred pinch numbers strongly depend on the logarithmic density gradient and magnetic shear, consistent with the theoretical predictions of the Coriolis pinch. The intrinsic torque from residual stress in the inner core is small, scales with the local logarithmic density gradient, and the data indicate a possible sign reversal. In the outer periphery of the core, the intrinsic torque is always co-current-directed and scales with the pressure gradient. This is consistent with prior experimental findings and global, non-linear gyrokinetic predictions. It suggests that profile shearing effects generate the intrinsic torque in the inner core. Toward the outer core, most likely, effects from E×B-shearing become more influential. These results offer the first comprehensive picture of this transport channel in the core plasma and contribute to validating the corresponding theoretical understanding. The derived scaling laws are used to construct a reduced momentum transport model, which has been validated against an additional dataset. This demonstrates that the model captures the essential contributions to momentum transport in the core of H-mode plasmas.

Runaway electrons (REs) are a concern for tokamak fusion reactors from discharge startup to termination. A sudden localized loss of a multi-megaampere RE beam can inflict severe damage to the first wall. Should a disruption occur, the existence of a RE seed may play a significant role in the formation of a RE beam and the magnitude of its current. The application of central electron cyclotron resonance heating (ECRH) in the Tokamak & agrave; Configuration Variable (TCV) reduces an existing RE seed population by up to three orders of magnitude within only a few hundred milliseconds. Applying ECRH before a disruption can also prevent the formation of a post-disruption RE beam in TCV where it would otherwise be expected. The RE expulsion rate and consequent RE current reduction are found to increase with applied ECRH power. Whereas central ECRH is effective in expelling REs, off-axis ECRH has a comparatively limited effect. A simple 0-D model for the evolution of the RE population is presented that explains how the effective ECRH-induced RE expulsion results from the combined effects of increased electron temperature and enhanced RE transport.

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm’s law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98( p,y ) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios (‘hybrid’ and ‘elevated’ q -profile) were established on AUG and characterised concerning their plasma performance.

TCV’s operating regime with an X-point radiator (XPR) has been broadened by changing the magnetic geometry. XPRs have properties that could make them an attractive power exhaust solution for fusion reactors. These include the conversion of a high fraction of exhaust power into radiation. TCV had previously accessed the XPR regime only with difficulties, as predicted for plasmas where radiative losses are dominated by carbon impurities, that are ubiquitous in TCV. Guided by this theoretical model of the XPR, recent experiments employed TCV’s configurational versatility to demonstrate that XPR access can be facilitated by introducing a second X-point in the vicinity of the separatrix. This configuration, which has a snowflake-minus topology, features a particularly long magnetic connection length from the region just above the X-point to the outer midplane together with a wide geometrical interface with the private flux region that reaches high neutral pressures. Transitioning to this configuration in a high-power H-mode leads to a shift in the radiating region across the separatrix from the divertor to a volume above the X-point, i.e. within the last closed flux surface (LCFS). This displacement of the radiating region is co-incident with the disappearance of edge localised modes (ELMs), while retaining H-mode confinement, a behaviour only, to date, observed in devices with metallic walls. In contrast to observations in these other devices, on TCV, the primary strike points in these configurations remain attached. Detailed measurements of the plasma kinetic parameters inside and outside of the separatrix now challenge the models for access and stability of the XPR and ELMs alike.

Carbon impurity transport is studied in the TCV tokamak using a charge exchange recombination diagnostic. TCVs flexible shaping capabilities were exploited to extend previous impurity transport studies to negative triangularity (delta < 0). A practical way of studying light impurity transport (like C, TCVs main impurity species due to graphite tiled walls) is to investigate the correlations between the impurity ion gradients that, in this study, highlighted significant differences between positive (PT) and negative delta (NT) plasma configurations. delta scans ( - 0.6 < delta < + 0.6 ) were performed in limited configurations, but displayed little correlation between C temperature, rotation and density gradients for positive delta. This stiff response for delta > 0 changes for negative delta, where the evolution of del(vtor) was accompanied by variations of del(nC) over a range of negative delta, showing that transport, in NT, is affected by velocity gradients. Similar delta scans were performed with additional NBH (Neutral Beam Heating), with power steps ranging from 0.25 MW to 1.25 MW, highlighting increased momentum confinement in negative delta. Finally, the evolution of intrinsic plasma toroidal rotation across linear to saturated ohmic confinement regime (LOC/SOC) transitions was explored at delta < 0, expanding previous studies performed in TCV for delta > 0 (Bagnato et al 2023 Nucl. Fusion 63 056006). Toroidal rotation reversal was not observed for delta < 0, despite clear LOC/SOC transitions, confirming that these two phenomena occur concomitantly only in a restricted number of cases and under specific conditions.

Plasma start-up is typically achieved manipulating poloidal magnetic fields, gas injection and possibly auxiliary heating. Model-based design techniques have been gaining increasing attention in view of future large tokamaks which have more stringent constraints and less room for trial-and-error. In this paper, we formulate the tokamak start-up scenario design problem as a constrained optimization problem and introduce a novel shot-to-shot correction algorithm, based on the Iterative Learning Control concept, to compensate for unavoidable modeling errors based on experimental data. The effectiveness of the approach is demonstrated in experiments on the TCV tokamak showing that the target ramp-up scenario could be obtained in a small number of shots with a rough electromagnetic model.

Multi-spectral imaging of helium atomic emission (HeMSI) has been used to create 2D poloidal maps of T-e and ne in TCV's divertor. To achieve these measurements, TCV's MANTIS multispectral cameras (Perek et al 2019 Rev. Sci. Instrum. 90 123514) simultaneously imaged four He I lines (two singlet and two triplet) and a He II line (468 nm) from passively present He and He+. The images, which were absolutely calibrated and covered the whole divertor region, were inverted through the assumption of toroidal symmetry to create emissivity profiles and, consequently, line-ratio profiles. A collisional-radiative model (CRM) was applied to the line-ratio profiles to produce 2D poloidal maps of Te and ne. The collisional-radiative modeling was accomplished with the Goto helium CRM code (Zholobenko et al 2018 Nucl. Fusion 58 126006, Zholobenko et al 2018 Technical Report, Goto 2003 J. Quant. Spectrosc. Radiat. Transfer 76 331-44) which accounts for electron-impact excitation (EIE) and deexcitation, and electron-ion recombination (EIR) with He+. The HeMSI T-e and ne measurements were compared with co-local Thomson scattering measurements. The two sets of measurements exhibited good agreement for ionizing plasmas: (5 eV 5 T-e 5 60 eV, and 2 x 10(18)m(-3) 5 ne 5 3 x 1019m(-3)) in the case of majority helium plasmas, and (10 eV 5 T-e 5 40 eV, 2 x 1018m-3 5 ne 5 3 x 10(19)m(-3)) in the case of majority deuterium plasmas. However, there were instances where HeMSI measurements diverged from Thomson scattering. When T-e 5 10 eV in majority deuterium plasmas, HeMSI deduced inaccurately high values of T-e. This disagreement cannot be rectified within the CRM's EIE and EIR framework. Second, on sporadic occasions within the private flux region, HeMSI produced erroneously high measurements of ne. Multi-spectral imaging of Helium emission has been demonstrated to produce accurate 2D poloidal maps of Te and ne within the divertor of a tokamak for plasma conditions relevant to contemporary divertor studies.

Numerical simulations explore a possible tightly baffled, long-legged divertor (TBLLD) concept in a future upgrade of the tokamak à configuration variable (TCV). The SOLPS-ITER code package is used to compare the exhaust performance of several TBLLD configurations with results from unbaffled and baffled TCV configurations. The investigated TBLLDs feature a range of radial gaps between the separatrix and the divertor baffles, with a smaller gap resulting in tighter baffling. All modeled TBLLDs are predicted to lead to a denser and colder plasma in front of the targets and increase the power handling by factors of 2–3 compared to the present, baffled, divertor and by up to a factor of 12 compared to the original, unbaffled, configuration. This improved TBLLD performance is attributed to an increased neutral confinement with more plasma–neutral interactions in the divertor region. Both power handling capability and neutral confinement increase with tighter baffling. The core compatibility of TBLLDs with nitrogen seeding is also evaluated and the detachment window, with acceptable core pollution, for these TBLLDs is explored, showing a reduction of the required upstream impurity concentration to achieve detachment by up to 18% with tighter baffling.