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Evidence that density shoulder broadening is dependent on high main-chamber neutral density is presented. Shoulder broadening does notoccur when the sources for main-chamber neutrals are minimized using divertor baffles and wide gaps to the first wall (similar to 3x the densitydecay length). Removing the baffles or reducing the gap to the inner wall both act to increase the density shoulder amplitude in otherwiseidentical TCV discharges. Radial turbulent transport is correlated with shoulder amplitude. (C) Published under an exclusive license by AIP Publishing

Multi-spectral imaging of helium atomic emission (HeMSI) has been used to create 2D poloidal maps of T e and n e 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 T e and n e . 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 n e measurements were compared with co-local Thomson scattering measurements. The two sets of measurements exhibited good agreement for ionizing plasmas: ( 5 eV ⩽ T e ⩽ 60 eV , and 2 × 10 18 m − 3 ⩽ n e ⩽ 3 × 10 19 m − 3 ) in the case of majority helium plasmas, and ( 10 eV ⩽ T e ⩽ 40 eV , 2 × 10 18 m − 3 ⩽ n e ⩽ 3 × 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 ⩽ 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 n e . Multi-spectral imaging of Helium emission has been demonstrated to produce accurate 2D poloidal maps of T e and n e within the divertor of a tokamak for plasma conditions relevant to contemporary divertor studies.

The Tokamak a Configuration Variable (TCV) has recently been equipped with gas baffles to increase its divertor closure for a broad range of divertor magnetic geometries. First experimental results reported in Reimerdes et al. (2021) demonstrated compatibility with a broad range of divertor magnetic geometries and confirmed the main design constraints of the baffles, in particular an increased divertor neutral pressure. The present article presents a more in-depths analysis and extended experiments of this first baffle assessment on the TCV boundary plasma. It is shown that the divertor neutral pressure increased following the installation of the baffles, as predicted by SOLPS-ITER and SolEdge2D-EIRENE simulations. Varying the divertor closure by changing the position of the plasma showed that the plasma equilibrium designed to assess the baffle effect was not far from the optimal trade-off between plasma plugging and recycling on the baffles. The baffles facilitate access to a partially detached regime in both L- and H-modes. In L-Mode, with the ion ∇ B -drift directed from the X-Point to the plasma core, a reduction of the line-averaged density detachment threshold by approximately 20% is observed at the outer target, while inner strike point detachment is only achieved in the presence of baffles. Multispectral imaging shows that the CIII front moves from the outer target towards the X-Point at a lower line-averaged plasma density, indicating a colder outer leg. In H-mode, the CIII front is generally located near the X-Point between the ELMs, while without baffles, N 2 -seeding is required to move the front up to that location.

This contribution presents preliminary results of the validation of state-of-the-art turbulence codes against well-diagnosed experiments in the TCV tokamak. The simulations were carried out by GBS, GRILLIX, and TOKAM3X codes for a Lower Single-Null (LSN) L-mode attached TCV plasma and, for the first time, in realistic size. The scenario depicted in the simulations was developed in the TCV tokamak and was optimized for this validation exercise. In the following, the level-of-agreement and comparisons of selected profiles are discussed.

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 q 95 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 q 95 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.

Recent experiments on Tokamak à Configuration Variable have made significant progress toward partial detachment of the outer divertor in neutral beam heated H-mode plasmas in conventional and alternative divertor configurations. The heating power required to enter H-mode was measured in a range of divertor configurations. It is found that at the core densities most favourable for H-mode access, the L–H threshold power is largely independent of the poloidal flux expansion and major radius of the outer divertor, and in the snowflake minus configuration. A factor 2 reduction in the outer divertor power load was achieved in ELM-free (using a fuelling and nitrogen seeding) and ELMy (using nitrogen seeding) H-mode plasmas. No significant reduction in the outer divertor particle flux was achieved in the ELM-free scenarios, compared with ~30% reduction in the most strongly detached ELMy cases. The poloidal flux expansion at the outer divertor was not found to significantly alter the cooling of the divertor in the ELM-free scenarios.

A gas baffle is being installed in the vessel of the tokamak à configuration variable (TCV) [1], in order to improve the closure of the divertor region. This upgrade has been envisaged, along with a foreseen increase in the available input power, in order to facilitate the access to detached divertor regimes at lower plasma collisionality, namely in more ITER-relevant conditions. It is necessary, in this framework, to be able to predict the impact of gas baffles, at the same time validating the current numerical tools available for the simulations of the edge plasma. The design of the gas baffle tiles has been supported by SOLPS-ITER simulations of TCV edge plasmas [2]. The optimized parameter is the neutral compression ratio, namely the ratio of neutral density in the divertor region and the one in the main chamber, at a given upstream electron density. The performance of the gas baffles in terms of neutral confinement has been predicted for an ideal case with intermediate plasma current, as a function of plasma density and input power [3]. These simulations predict an improvement of the compression ratio, with the current baffles, of a factor of approximately five for attached low-density plasmas, which increases to up to 10-20 at higher upstream densities. Envisaging a future modification of TCV gas baffles, a further numerical investigation on the optimal baffle extension has been carried out, by means of the SolEdge2D-EIRENE code [4]. The main advantage of this code is the possibility to simulate realistic shapes for plasma-facing components, thus coherently describing plasma parallel fluxes impinging on gas baffles. This analysis shows that the neutral compression factor in detached conditions could be further improved by almost a factor two by extending the Low-Field Side baffle by a few cm. With such a solution, the recycling on the baffle tip would still be acceptable. Both the mentioned numerical tools have been tested against experiments, SOLPS-ITER simulating an average current, Lower Single-Null scenario, and SolEdge2D-EIRENE a case at lower plasma current. Numerical results have been compared to experiments with baffle-compatible plasmas, in absence of the baffle. The density ramp imposed in experiments is reproduced numerically, and it allows the investigation of different divertor conditions. Transport coefficients are chosen to match experimental upstream profiles: numerical results are shown to reproduce, almost in a quantitative way, the experimental results at divertor targets. Keeping transport coefficients fixed, two additional wall geometries have been simulated, one including the High-Field Side baffle, and one with the full gas baffle. With the first experiments in baffled operation, both mentioned tools are tested against experiments: the results of the comparison will be discussed, shedding light on the capability of 2D transport codes of predicting the edge plasma behaviour in presence of gas baffles.

The effect of magnetic geometry on scrape-off layer (SOL) transport and detachment behaviour is investigated on the TCV tokamak with the goal of assessing the potential of alternative divertor geometries and for the validation of theoretical models. L-mode experiments reveal that increasing connection length and hence divertor volume by either increasing poloidal flux expansion or divertor leg length have different effects on the boundary plasma. In attached conditions, the SOL heat flux width q inferred from target infrared thermography measurements is weakly dependent on poloidal flux expansion but increases approximately with the square root of the divertor leg length. The divertor spreading factor S shows no clear trend with leg length but decreases with flux expansion. TOKAM3X turbulence simulations of the leg length scan are in qualitative agreement with the experiment and can explain observations by a strongly asymmetric (ballooning) transport at and below the X-point. Evidence for increased transport in the region of low poloidal field is obtained in the Snowflake minus geometry. The presence of an additional X-point in the low-field side SOL increases the effective SOL width by approximately a factor two. Increasing flux expansion and leg length both result in enhanced divertor radiation levels, with the effect being much larger in the latter case. This behaviour, together with the observed trend in q, is consistent with a substantial drop in the density threshold for divertor detachment with increasing leg length and a weak variation with flux expansion. Novel spectroscopic techniques reveal that the drop in target ion current and access to detachment is caused by a reduction of the divertor ionization source due to power starvation, while volume recombination is only a small contributor. This interpretation is confirmed by SOLPS modelling. TCV alternative divertor studies are being extended to neutral beam heated H-mode plasmas. The H-mode power threshold is found to vary weakly between standard, X-, and Super- X geometries. In all cases, ELMy H-mode is obtained at intermediate current, while the discharges are ELM-free at high current. Signs of detachment have so far only been observed in the latter case. Ongoing experiments further investigate H-mode detachment in these plasmas and will be extended to Snowflake configurations.

Detachment is predicted to be of paramount importance in handling the power exhaust for future fusion devices, such as ITER. However, a direct experimental quantification of the role and spatial profile of the various atomic processes controlling the loss of divertor ion current during detachment has, until now, not been available due to lack of ionisation measurements. The physics of the target ion current loss, a defining and important feature of detachment, is studied in TCV using density ramp, or N2 seeded, L-mode discharges with various plasma currents. Novel spectroscopic analysis techniques utilising the hydrogen Balmer series have been developed to infer the ionisation source magnitude and distribution. This provides, together with the ion sink in the plasma (recombination), electron density and divertor power balance measurements, a detailed picture of particle and power balance along the outer divertor leg. The results for a conventional single-null topology show the divertor ion source tracks the ion target flux both in magnitude and in time: both the ion target current and ion source decrease together at detachment. Surprisingly, the volumetric recombination ion sink - commonly thought to be the primary detachment ion loss mechanism - is only a small (sometimes negligible) portion of the ion current losses. New evidence from TCV shows the driver of the divertor ion source decrease is a decrease in power flowing into the ionisation region combined with an increase in the energy required per ionisation - essentially 'starving' the ionisation region of power. SOLPS modelling of a TCV density ramp discharge with a conventional divertor configuration has reproduced the general characteristics of the ion source reduction. 'Power starvation' of the ionization source, therefore, appears to be central to loss of divertor target ion current. The divertor plasma sink for ions, recombination, is maximised at the highest divertor densities, which are achieved at the highest core densities. Nitrogen seeding enables detachment by 'power starving' the ionisation region at lower core densities and hence lower recombination sink. The ratio between the recombination ion sink and the ion target flux increases when poloidal flux expansion is increased (x-divertor) under constant core conditions. Understanding these changes may be key to understanding the role of magnetic geometry on detachment and possibly the detachment process itself.

TCV experiments show evidence that the cross-field transport in the low-field side (LFS) Scrape-Off Layer (SOL) of the conventional Single-Null (SN) configuration is strongly enhanced when placing a secondary x-point in the common flux region of the primary separatrix. Such a magnetic configuration is known as the Snowflake Minus (SF-) [1]. The exhaust properties of the SF- configuration were investigated on TCV in Ohmically heated, L-mode attached plasmas with a range of x-point separations, magnetic field directions and locations of the secondary x-point (low-field side, LFS, or high-field side, HFS, SOL). The target heat flux profiles at all strike points are simultaneously measured with an Infrared Thermography (IR) system [2] and the main SOL plasma kinetic profiles with a fast reciprocating probe (RCP) [3]. The measured power repartition between the two branches of the SOL diverted by the secondary x-point yields an effective heat flux width ?q,u eff for the SOL. It is found that the LFS SOL cross-field transport is strongly enhanced by the presence of the secondary x-point, for both magnetic field directions, with ?q,u eff being a factor three larger than the value measured by the RCP at the outer mid-plane or by IR at the outer target of a comparable SN. The RCP inspection of the low poloidal field region reveals that the profiles of SOL density and ion saturation current develop steep gradients in proximity of the secondary separatrix, while the profiles flatten in between the two separatrixes, compared to the SN divertor. In contrast, the cross-field transport in the HFS SOL is not significantly affected by the presence of the secondary x-point. However, similar to comparable SN discharges, it is sensitive to the magnetic field direction, with a broader SOL width for reversed field. Placing the secondary x-point on one side of the SOL also affects the power balance between inner and outer divertor. This is interpreted as the effect of changed parallel transport in the SOL due to the changes in the geometry of the flux tubes caused by the presence of the secondary x-point, and was already seen in fluid calculations assuming constant cross-field transport coefficients.

This paper investigates the power loads at the inner and outer divertor targets of attached, Ohmic L-mode, deuterium plasmas in the TCV tokamak, in various experimental situations using an Infrared thermography system. The study comprises variations of the outer divertor leg length and target flux expansion, the plasma current and a reversal of the magnetic field direction. The direct impact of the divertor magnetic geometry on scrape-off layer (SOL) transport -parameterised by the SOL power fall-off length lambda_{q, u}, the divertor spreading factor S u and the in-out power asymmetry- is reported for constant core properties. The in-out power asymmetry increases, either with the divertor leg length, or the target flux expansion. The SOL width lambda_{q, u} scales positively with divertor leg length, with a strength that depends on the field direction and differs between the inner and outer divertor. This implies a parametric dependence of lambda_{q, u} that is not explicitly included in current multi-machine scaling laws. The divertor spreading factor at the target S = S_u f_x , where f_x is the target flux expansion, appears unaffected by changes in the divertor geometry and in the plasma current, is independent of the magnetic field direction and is similar between inner and outer divertor. Possible interpretations of these observations using an ad-hoc analytical purely conductive model for the SOL, by ion drifts or by asymmetric turbulent cross-field transport in the divertor are presented. The observed values of lambda_{q, u} are related to existing L-mode and H-mode scaling laws and to similar studies in other tokamaks. Finally, potential implications of these findings for future larger fusion machines are discussed.

Reciprocating probe data taken in the edge and SOL of the TCV tokamak shows ubiquitous intermittent events, caused by coherent filamentary objects moving convectively. The filament motion, which is due to Grad - B charge separa tion can result in significant broadening of the SOL and the formation of a density shoulder, which results in increased plasma - wall contact. Predictive modelling for future fusion devices requires testing of the theoretical models before reliable estimates of the SOL width can be generated. Conditio nal averaging techniques were applied to extract the filaments from the background turbulence and determine filament sizes and velocities. The database we obtained includes density scans in both forward and reverse Bt, with discharges from sheath - limited t o detached regimes. The filaments in TCV were found to range in size from a few mm ( due to the minimum resolution from pin separatio n) to the size of the SOL width and r adial velocities varied from 0 to 6 km/s . Near the LCFS, the poloidal velocity determined using poloidal cross - correlation were typically 5 - 10 km/s , significantly affecting the size determination . T he filaments span the range 0.05< ?

We present a fully developed, small-size multi-diagnostic probe head for the characterization of near-wall (scrape-off layer and divertor region) electric and magnetic plasma properties in Medium-Size Tokamaks (MST). The probe head can be mounted on a number of reciprocating probe manipulators of MST devices. It contains one high electron emissive probe (EEP) [1, 2], two cold Langmuir probes (CLP), two retarding field analyzers (RFA) and two miniaturized 3-axial magnetic pick-up coils (MPC) for magnetic field fluctuations measurements. The EEP and the two CLP are placed on almost the same radial position (shadowing effects considered) making various Langmuir probe measurement techniques (e.g. fast electron temperature measurements) applicable. The CLPs can further simply be replaced by two pins, geometrically parallel to the magnetic field, capable of determining the electron energy distribution function (EEDF) in the plasma [3]. Combining these measurements with the ion energy distribution function (IEDF) of the plasma, gained from the RFAs, the real local heat flux can be determined [4]. Here we present preliminary measurement results obtained in the linear magnetic device located at the Jozef Stefan Institute, Slovenia. Soon the probe head will be integrated into the Tokamak a Configuration Variable (TCV) for first scrape-off layer measurements.

A simple analytic model for the repartition of the Scrape-Off Layer (SOL) exhaust power between the inner and outer divertors in a diverted low-density tokamak plasma is introduced. Electron heat conduction is assumed to dominate the heat transport, from the outboard mid-plane to the divertor targets, with no heat sinks or sources in the SOL. Both divertor channels are in the attached, high-recycling regime. The model is in reasonable qualitative agreement with recent TCV experimental data and EMC3-Eirene simulations. For the Single Null divertor, it reproduces the experimentally observed increase in the power ratio between the inner and outer divertor plates of TCV, with increasing the outer divertor leg length or the outer target flux expansion. For the Snowflake Minus configuration, it reproduces the observed variation of with X-point separation, although only for the reversed magnetic field direction. Within the model limitations, it provides a basic understanding of the power sharing in alternative divertor geometries.

The aim of this work is to provide an understanding of detachment at TCV with emphasis on analysis of the Balmer line emission. A new Divertor Spectroscopy System has been developed for this purpose. Further development of Balmer line analysis techniques has allowed detailed information to be extracted from the three-body recombination contribution to the n = 7 Balmer line intensity. During density ramps, the plasma at the target detaches as inferred from a drop in ion current to the target. At the same time the Balmer 6 → 2 and 7 → 2 line emission near the target is dominated by recombination. As the core density increases further, the density and recombination rate are rising all along the outer leg to the x-point while remaining highest at the target. Even at the highest core densities accessed (Greenwald fraction 0.7) the peaks in recombination and density may have moved not more than a few cm poloidally away from the target which is different to other, higher density tokamaks, where both the peak in recombination and density continue to move towards the x-point as the core density is increased. The inferred magnitude of recombination is small compared to the target ion current at the time detachment (particle flux drop) starts at the target. However, recombination may be having more localized effects (to a flux tube) which we cannot discern at this time. Later, at the highest densities achieved, the total recombination does reach levels similar to the particle flux.