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X-Point Radiator (XPR) regimes have been obtained in WEST tokamak experiments with nitrogen seeding during the experimental campaigns of 2023 and 2024. These experiments showed the formation of a stable toroidal radiating ring near the X-point, similar to observations in other devices such as JET, ASDEX-Upgrade, TCV, and COMPASS. In WEST, the onset of this regime is associated with a sharp transition of the divertor plasma from hot to cold and dense conditions, with increased particle fluxes, indicating that the plasma is not detached. At the same time, core conditions are significantly improved. These scenarios were successfully controlled in WEST using an interferometry line-of-sight passing through the X-point. Interpretative modeling of these discharges with the SOLEDGE3X-EIRENE code reveals that a physics mechanism needed to stabilize WEST nitrogen XPRs is not present when driving the simulations at constant power. On the contrary, a stable XPR can be obtained by increasing the power injected at the time of the XPR onset to represent the reduction of W contamination, highlighting the need to describe the plasma dynamics and go toward integrated core–edge simulations.

Two reduced models for predicting detachment onset and divertor reattachment times are validated on MAST Upgrade (MAST-U). These models are essential for future tokamak reactor design, providing rapid calculations based primarily on engineering parameters. The first model predicts detachment onset using a qualifier developed on ASDEX Upgrade (AUG) and later tested on JET, while the second model provides an estimate for the time required for a given transient to burn through the neutral particles in the divertor. Experiments in H-mode plasma scenarios were conducted on MAST-U with double-null and single-null configurations, which involved D2 fuelling ramps and N2 seeding. The detachment onset was determined by monitoring divertor parameters, including the target heat flux profile, electron temperature, and electron density, with measurements showing consistency with AUG-derived predictions. Reattachment times were assessed during dynamic vertical shifts of the plasma centroid position, with observations indicating reattachment within milliseconds, consistent with model predictions. Overall, the results confirm the applicability of both reduced models to MAST-U, extending their validation beyond AUG and JET.

Tokamak à 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, β _N ∼ 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 Alfvénic spectrum modulating the FI fas ion loss detector signal. NT configurations showed less toroidal Alfvén 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 β 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.

The impact of plasma shaping on the properties of high density H-mode scrape-off layer (SOL) profiles and transport at the outer midplane has been investigated on Tokamakà configuration variable. The experimental dataset has been acquired by evolving the upper triangularity while keeping the other parameters constant. The scan comprises $\delta_\mathrm{up}$ values between 0.0 and 0.6, excluding negative triangularity scenarios. Within this study, a transition from type-I edge localised modes to the quasi-continuous exhaust regime takes place from low to high $\delta_\mathrm{up}$ . The modification of the upstream SOL profiles has been assessed, in terms of separatrix quantities, within the $\alpha_\mathrm{t}$ turbulence control parameter theoretical framework (Eich et al 2020 Nucl. Fusion 60 056016). The target parallel heat load and the upstream near-SOL density profiles have been shown to broaden significantly for increasing $\alpha_\mathrm{t}$ . Correspondingly, in the far SOL a density shoulder formation is observed when moving from low to high $\delta_\mathrm{up}$ . These behaviours have been correlated with an enhancement of the SOL fluctuation level, as registered by wall-mounted Langmuir probes as well as the thermal helium beam diagnostic. Specifically, both the background and the filamentary-induced fluctuating parts of the first wall ion saturation current signal are larger at higher $\delta_\mathrm{up}$ , with filaments being ejected more frequently into the SOL. Comparison of two pulses at the extremes of the $\delta_\mathrm{up}$ scan range, but with otherwise same input parameters, shows that the midplane neutral pressure does not change much during the H-mode phase of the discharge. This indicates that indirect effects of the change in geometry, linked to first wall recycling sources, should not play a significant role. The total core radiation increases at high $\delta_\mathrm{up}$ , on account of a stronger plasma–wall interaction and resulting larger carbon impurity intake from the first wall. This is likely associated to the enhanced first wall fluctuations, as well as a smaller outer gap and the close-to-double-null magnetic topology at high shaping.

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

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 2021 JET exploited its unique capabilities to operate with T and D–T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D–T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements—new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors—as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D–T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D–T plasmas to date and expanding our understanding of isotopes and D–T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D–T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D–T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D–T operations is presented in (King et al Nucl. Fusion submitted).

This paper presents a quantitative validation of Scrape-Off Layer Plasma Simulation-ITER (SOLPS-ITER) simulations against the TCV-X21 reference case and provides insights into the neutral dynamics and ionization source distribution in this scenario. TCV-X21 is a well-diagnosed diverted L-mode sheath-limited plasma scenario in both toroidal field directions, designed specifically for the validation of turbulence codes (Oliveira, Body et al 2022 Nucl. Fusion 62 096001). Five new, neutrals-related observables are added here to the extensive, publicly available TCV-X21 dataset. These are three deuterium Balmer lines in the divertor and neutral pressure measurements in the common and private flux regions. The quantitative agreement metric used in the validation is combined with the conjugate gradient method to approach the SOLPS-ITER input parameters that return the best overall agreement with the experiment. A proof-of-principle test of this method results in a modest improvement in the level-of-agreement; the shortcomings impacting the result and how to improve the methodology are discussed. Alternatively, a scan of the particle and heat diffusion coefficients shows an improvement of 10.4% in the level-of-agreement, approximately twice as high as that achieved by the gradient method. This result is found for an increased transport coefficient compared to what is usually used for TCV L-mode plasmas. The simulations further indicate that ∼65% of the total ionization occurs in the SOL, from which ∼70% in the divertor regions, despite being a sheath-limited regime, motivating the inclusion of self-consistent neutral models in future turbulence simulations on the path towards improved agreement with the experiment.

Building on prior analysis of ASDEX Upgrade (AUG) experiments (Henderson et al 2023 Nucl. Fusion 63 086024), this study compares simple analytical formula predictions for divertor detachment onset and reattachment timescales in JET experiments. Detachment onset primarily scales with divertor neutral pressure, impurity concentration, power directed to the targets, machine size, and integral perpendicular power decay length. JET experiments, focusing on seeding mixtures of Ne and Ar, align with the detachment onset predictions. Radiation efficiencies among the impurities show good agreement with the model predictions, contrasting with AUG observations which suggested higher efficiency for Ar and lower efficiency for Ne. The time taken to re-ionise the neutral volume in front of the outer target in fully detached divertor conditions was measured following both abrupt increases in injected neutral beam power and, separately, cutting of the impurity gas flow. Re-ionisation of the neutrals occurs within approximately 1 s on JET, which aligns with the simple model prediction derived from AUG data. While the AUG results are not new, their comparison with the JET results enhances understanding, reinforcing confidence in using simple models to predict future reactor scenarios.

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.

Future fusion reactors require a safe, steady-state divertor operation. With deep divertor detachment, which is typically induced by impurity seeding, the radiation concentrates in a small region at the X-point or on closed flux surfaces above the X-point. This so-called X-point radiator (XPR) moves further inside the confined region with increasing seeding and the location can be actively controlled.At AUG, the parameter space for operation with an XPR was significantly extended, using active feedback on the XPR location. The XPR is observed in nearly the whole operational space of AUG in the high-densities or high collisionality regime. ELM suppression is consistently observed in all cases where the XPR was moved to a significant height above the X-point.Direct measurements of density and temperature from the region around the XPR using the new divertor Thomson scattering system at AUG indicate that the temperature at the location of the XPR remains high (>30eV) and only the region towards the X-point cools down further. In this cold XPR core, the temperature reduces to about 1eV.An XPR is also observed in TCV by the injection of nitrogen as extrinsic impurity. This highlights that the wall material (W for AUG, C for TCV) or machine size does not play a significant role for the existence of the regime. However, the scenario appears to be less stable in TCV. First experiments show the necessity of an active control for the XPR: Depending on the wall conditions and the nitrogen wall storage, the required nitrogen seeding level to achieve an XPR changes.Both, the low temperatures measured radially outside of the radiation zone at AUG, and the lower stability of the XPR regime at TCV with the presence of carbon are consistent with the predictions of a one-dimensional model of the XPR. However, the model would predict the development of the cold XPR core, and significant radiation at the X-point might already exist before reaching this cold temperature solution.

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.

Multi-spectral imaging of helium atomic emission (HeMSI) has been used to create 2D poloidal maps of ${{T}}_{\mathrm{e}}$ and ${{n}}_{\mathrm{e}}$ 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 ${{T}}_{\mathrm{e}}$ and ${{n}}_{\mathrm{e}}$ . 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 ${\textrm{He}}^{+}$ . The HeMSI ${{T}}_{\mathrm {e}}$ and ${{n}}_{\mathrm{e}}$ measurements were compared with co-local Thomson scattering measurements. The two sets of measurements exhibited good agreement for ionizing plasmas: $({5 \,{\textrm{eV}} \leqslant {{T}}_{\mathrm{e}} \leqslant 60 \,{\textrm{eV}}}$ , and ${2 \times 10^{18}\, {\textrm{m}}^{-3} \leqslant{{n}}_{\mathrm{e}} \leqslant 3 \times 10^{19}\,{\textrm{m}}^{-3}})$ in the case of majority helium plasmas, and $({10 \,{\textrm{eV}} \leqslant {{T}}_{\mathrm{e}} \leqslant 40 \,{\textrm{eV}}},{2 \times 10^{18}\,{\textrm{m}}^{-3} \leqslant{{n}}_{\mathrm{e}} \leqslant 3 \times 10^{19} \,{\textrm{m}}^{-3}})$ in the case of majority deuterium plasmas. However, there were instances where HeMSI measurements diverged from Thomson scattering. When ${{T}}_{\mathrm{e}} \leqslant 10 \,{\textrm{eV}}$ in majority deuterium plasmas, HeMSI deduced inaccurately high values of ${{T}}_{\mathrm{e}}{}$ . 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 ${{n}}_{\mathrm{e}}$ . Multi-spectral imaging of Helium emission has been demonstrated to produce accurate 2D poloidal maps of ${{T}}_{\mathrm{e}}$ and ${{n}}_{\mathrm{e}}$ within the divertor of a tokamak for plasma conditions relevant to contemporary divertor studies.

Plasma-molecular interactions generate molecular ions which react with the plasma and contribute to detachment through molecular activated recombination (MAR), reducing the ion target flux, and molecular activated dissociation (MAD), both of which create excited atoms. Hydrogenic emission from these atoms has been detected experimentally in detached TCV, JET and MAST-U deuterium plasmas. The TCV findings, however, were in disagreement with SOLPS-ITER simulations for deuterium, indicating a molecular ion density ( $D_2^+$ ) that was insufficient to lead to significant hydrogenic emission, which was attributed to underestimates of the molecular charge exchange rate ( $D_2 + D^+ \rightarrow D_2^+ + D$ ) for deuterium (obtained by rescaling the hydrogen rates by their isotope mass). In this work, we have performed new SOLPS-ITER simulations with the default rate setup and a modified rate setup where ion isotope mass rescaling was disabled. This increased the $D_2^+$ content by ${\gt}{\times} 100$ . By disabling ion isotope mass rescaling: (1) the total ion sinks are more than doubled due to the inclusion of MAR; (2) the additional MAR causes the ion target flux to roll-over during detachment; (3) the total $D\alpha$ emission in the divertor increases during deep detachment by roughly a factor of four; (4) the neutral atom density in the divertor is doubled due to MAD, leading to a 50% increase in neutral pressure; (5) total hydrogenic power loss is increased by up to 60% due to MAD. These differences result in an improved agreement between the experiment and the simulations in terms of spectroscopic measurements, ion source/sink inferences and the occurrence of an ion target flux roll-over. Extrapolating simplified scalings of divertor molecular densities (TCV & MAST-U) to reactor-relevant densities suggests the underestimation of molecular charge exchange could strongly impact divertor physics (neutral atom density, ions sinks) and hydrogen emission (which has implications for detachment control) in deeply detached conditions, warranting further study.

The Tokamak à 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, N2-seeding is required to move the front up to that location.

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, Pin/Pout, with increasing the outer divertor leg length or the outer target flux expansion. For the Snowflake Minus configuration, it reproduces the observed variation of Pin/Pout 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. Keywords: Heat exhaust, SOL transport, Long-legged Single Null Divertor, Snowflake Divertor, In-out power sharing, TCV

Plasma emission of hydrogen and impurities is measured with use of filtered camera imaging in TCV divertor. The measurement technique is discussed in detail including: selection choice of filters, the absolute calibration procedure and the tomographic inversion. The process is applied to an unseeded L-mode density ramp discharge in an unfavorable for H-mode access magnetic field direction with divertor baffles. Obtained poloidal emission profiles are used to split Balmer emission into excitation and recombination parts revealing a recombination emission maximum on the high field side. The optical emission is further studied by selecting radial profiles in cross sections of interest finding an opposite radial shift direction for high and low n Balmer line emission maxima in detached conditions.

A double-null configuration is being considered for the EU-DEMO, due to its potential benefits for power exhaust arising from the use of two active divertors and magnetically disconnected low- and high-field sides. Using systematic parameter scans in fluid simulations, we have investigated the divertor power exhaust in the EU-DEMO in a connected double-null configuration, and compared the edge plasma properties to those obtained in a single-null configuration under detached conditions anticipated for reactor operation. Neglecting drift effects and kinetic behaviour of the neutrals, no clear benefits of the double-null configuration could yet be identified for the radiation pattern and power mitigation on open field lines. Future work should address the aforementioned physics as well as the effect of the additional X-point on core radiation.

A physics and engineering analysis of alternative divertor configurations is carried out by examining benefits and problems by comparing the baseline single null solution with a Snowflake, an X- and a Super-X divertor. It is observed that alternative configurations can provide margin and resilience against large power fluctuations, but their engineering has intrinsic difficulties, especially in the balance between structural solidity and accessibility of the components and when the specific poloidal field coil positioning poses further constraints. A hybrid between the X- and Super-X divertor is proposed as a possible solution to the integration challenge.

A new set of carbon tiles, neutral beam heating optics and gas baffles were installed on TCV during the baffled divertor upgrade in early 2019. The installation of the baffles allows a deconvolution of the roles of main chamber and divertor neutral pressure on the H-mode pedestal structure. This physical barrier allows relatively high neutral pressures to be constrained to the divertor, thus preventing neutrals from entering the main chamber and potentially degrading core confinement. This study presents the experimentally measured and modelled pedestal heights and structure for a series of H-mode discharges prior to and after this upgrade.Increased pedestal performance at high divertor neutral pressure was observed after the baffled divertor upgrade. This was consistent across all triangularities and outer target locations investigated and is attributed to higher pedestal top temperatures being maintained at high gas injection rates. ASTRA simulations indicated beam heating power coupled to the plasma did not significantly vary after the baffled divertor upgrade or as a function of divertor neutral gas pressure. Analysis of the pedestal structure exposed a strong correlation between pedestal performance and the density pedestal position prior to and after the baffled divertor upgrade. The baffled divertor upgrade limited the outward shift of the density pedestal, thus maintaining higher pedestal performance at high divertor neutral pressures. Stability analysis indicated the majority of discharges studied were within 25% of the stability boundary. No correlation was found between the distance from the stability boundary and pedestal performance or structure. Comparison with the EPED1 model indicated that TCV discharges do not have a fixed dependence between pedestal βθand pedestal width. A large variation in the EPED1 relating parameter was observed and found to vary with the density pedestal position.

Recent TCV experiments confirm the predicted formation of an electric potential well, below the magnetic X-point, in configurations with unfavorable Btfield direction (ion ∇Bdrift away from the divertor), that substantially reshapes the typical divertor E×Bflow pattern. The local charge balance ∇⋅jin the private flux region (PFR) of diverted tokamak plasmas has been previously argued to be dominated by parallel and diamagnetic currents. This hypothesis is tested herein in TCV discharges by comparison with SOLPS-ITER simulations, fully accounting for drifts and currents. Simulated parallel currents correctly capture measured current profile characteristics for both targets and both Bt-directions, whereas those omitting drifts fail. It is shown that the resulting parallel currents dictate the electric fields in the PFR for low temperature (detached divertor) conditions resulting in locally negative electric plasma potential in configurations with unfavorable H-mode access.

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. Keywords: Detachment, Divertor spectroscopy, Volumetric recombination, Tokamak power exhaust, TCV tokamak, Balmer line spectroscopy

To address the risk that, in a fusion reactor, the conventional single-null divertor (SND) configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD), are investigated in TCV. The expected benefits of the SFD-minus in terms of power load and peak heat flux are discussed and compared to experimental measurements. In addition, key results obtained during the last years are summarized. Keywords: Tokamaks, Snowflake divertor, Detachment, Heat load mitigation

The Swiss Plasma Center (SPC) is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

Divertor detachment in the TCV tokamak has been investigated through experiments and modelling. Density ramp experiments were carried out in ohmic heated L-mode pulses with the ion ∇B drift directed away from the primary X-point, similar to previous studies [1]. Before the roll-over in the ion current to the outer strike point, C III and Dα emission from the outer leg recede slowly from the strike point toward the X-point, at a rate of ∼2.0 × 10−19m/m−3 along the magnetic field as the electron temperature along the leg reduces with increasing density. Around the onset of detachment, the upstream density profile and outer target Dα profiles broaden, possibly leading to an increase in radiation in the SOL by increased interaction between the SOL and the carbon tiles lining the outer wall. The plasma conditions upstream and at various locations along the detached outer divertor leg have been characterised, and the consistency of this data has been checked with the interpretive OSM-EIRENE-DIVIMP suite of codes [2] and are broadly found to be consistent with measured Dγ/Dα emissivity profiles along the detached outer divertor leg.

Safe ITER operations will rely on power spreading to keep the peak heat flux within divertor material constraints. A solid understanding and parameterization of heat flux profiles is therefore mandatory. This paper focuses on the impact of plasma geometry on the power decay length (λq) and the spreading factor (S). Numerical heat flux profiles, obtained with the simple SOL transport model MONALISA, agree with theoretical predictions for purely diffusive cylindrical plasmas: λq does not depend on the machine-specific divertor geometry but only on transport parameters and global geometry (a, R, k). A dedicated experiment on TCV was designed to further test this assumption in L-mode plasmas with similar control parameters and upstream shape but different divertor leg length (Zmag=−14, 0, 28 cm). Characterization of OSP q∥ profiles with Langmuir probes and infrared thermography enlightens unexpected behavior with the divertor leg length: λq increases, while S shows no clear trend. These findings suggest that the link between heat flux profiles, plasma geometry and transport is currently not fully understood.