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.

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

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.

The effect of divertor closure and nitrogen seeding on the detachment process has been studied by performing 2D numerical simulations of tokamak a configuration variable (TCV) H-mode divertor scenarios with the SOLEDGE3X-EIRENE edge plasma transport code. The simulations reveal that, in the cases with only deuterium gas fuelling, detachment occurs at a similar level of divertor neutral pressure ( approximate to 0.76 Pa), despite the difference in divertor closure achieved by changing the length of the outer baffle. Nitrogen reduces the target temperature with little effect on the upstream density and momentum loss but drops the upstream pressure, leading to a decrease in the target particle flux and divertor neutral pressure. Furthermore, when the radiation front starts to move up from the outer target, the peak parallel heat flux level at the outer target remains approximately the same ( approximate to 2.3 MW m-2), regardless of whether it is a deuterium fuelling scan, a nitrogen seeding scan at a fixed fuelling rate, or a change in the length of the outer baffle. An empirical partial detachment qualifier calibrated on AUG experimental data was compared with the TCV simulations. The results show good agreement in detachment state prediction, indicating the potential of this detachment qualifier to be applied in devices of different sizes, emphasizing the combined influence of input power entering the divertor, neutral pressure, and the concentration and species of impurity in achieving divertor detachment.

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017-2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m(-2) were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

Estimates of plasma conditions in the far scrape-off layer (SOL) and of first wall (FW) fluxes in ITER are key input parameters to first wall erosion and impurity migration models, which are in turn involved in the assessment of FW panel lifetime and fuel retention studies. SOLEDGE3X up-to-the wall boundary plasma simulations are performed for ITER, based on an expected Pre-Fusion-Power-Operation (PFPO-1) scenario at PSOL= 20MW, including an impact study of enhanced far-SOL transport. This latter study concerns the possible formation of density shoulders, which are modelled here by applying an increase to the prescribed perpendicular particle and heat diffusivity coefficients maps in the far-SOL in the code, in order to flatten the density and temperature profiles there. Several kinds of such obtained “shoulders” are considered. A brief comparison with SOLPS-ITER is performed on the reference case with uniform coefficients, and shows good agreement. When far-SOL transport is increased, temperatures computed on the first wall rise to 20–30 eV for ions, and to 10–20 eV for electrons. It is also found that for first wall quantities of interest in the ITER machine, the assumed level of perpendicular transport in the far-SOL is the most relevant parameter, with the location at which transport is increased being much less important.

ITER Pre-Fusion Power Operation 1 (PFPO-1) phase low-power plasmas are simulated with the SOLEDGE3X-EIRENE code, a multi-fluid edge plasma solver that can describe plasma conditions up to the machine's first wall, including a kinetic description of neutrals. Here, SOLEDGE3X is used in 2D mean-field mode. A throughput scan is performed, reproducing a similar scan from the ITER SOLPS simulations database. The physics assumptions for these simulations are presented, in particular the plasma-neutral interaction model where molecule charge-exchange and ion-molecule elastic collisions have newly been added. Challenges encountered in such up-to-the-wall simulations are discussed, along with remaining open questions. Results are presented with two foci: on targets, and on the rest of the first wall. On targets, conditions transition from attached to partially detached as expected, and toroidally symmetric heat fluxes (i.e. without 3D effects) remain below 5 MW/m(2), within the divertor design limits. An increase in the parallel heat flux decay length lambda(q) with throughput is also observed. On the rest of the first wall, increasing puff rate does not only spread power deposited on the target but also increases heat load on the first wall. These fluxes remain however small (<100 kW/m(2)).

The multi-component fluid closure derived by Zhdanov (2002 Transport Processes in Multicomponent Plasma (London: Taylor and Francis)) is implemented in the fluid code SOLEDGE3X-EIRENE to deal with arbitrary edge plasma composition. The closure assumes no distinction between species such as light versus heavy species separation. The work of Zhdanov is rewritten in a matricial form in order to clearly link friction forces and heat fluxes to the different species velocities and temperature gradients.

In WEST experimental campaign C5, the divertor pumping capability has been improved by sealing the space between the divertor outer baffle and the vacuum vessel. It is expected that the degree of baffle leakage influences the transport of neutral particles inside the main chamber, which affects the detachment onset. Knowing the exact impacts of leakage and understanding the physical processes behind it are helpful for the study and control of plasma detachment. We investigate the impact of leakage by performing transport simulations through SOLEDGE-EIRENE code considering several cases with different leakage levels. Starting from the basic simulation case, non-constant radial transport coefficients obtained by the feedback control method are applied to achieve a better match with the experimental one (#54903) in L-mode. Based on the basic case, the evolution of plasma regimes from sheath limited regime to detached one in different wall geometries has been studied by ramping the upstream density. The numerical results show that the cases with closed or reduced leakage under the baffle have better performance in trapping the neutral particles and higher neutral pressure near the baffle. The neutral compression ratio is increased by a factor up to 4, leading to more significant momentum and power dissipation in the divertor, thus lowering the detachment threshold in ne,sep by up to 16%. At the same time, a much higher gas puff rate by a factor up to 22 is needed to maintain an equivalent ne,sep level in the case without or reduced leakage. For all the cases here, there exist characteristic parameters on which the baffle closure has no obvious influence on their value when plasma starts to detach. The evolution of radiator height as a function of target temperature shows no sensitivity to the leakage, which gives some insight into the stable detachment control strategy in the future. Finally, simulation results are compared with available neutral pressure measurements from WEST campaigns to verify the predictions from the simulation.

Contamination of core plasma by high-Z impurities, especially tungsten (W), is the main reason for the very high level of radiated power in WEST experiments. Intrinsic light impurities, mainly oxygen and carbon, play a dominant role in the sputtering of W on plasma facing components. In this contribution, we present a detailed analysis of WEST experiments supported by numerical modeling performed with the transport code SOLEDGE-EIRENE providing a clear picture of light impurities transport and poloidal distribution. Moreover, making use of SOLEDGE-ERO2.0 simulations, possible strategies to reduce core contamination due to W penetration are presented.

The influence of the input power (IP), puffing rate and neutral pressure on the fuel (hydrogen isotopes) inventory of the WEST and ITER divertors is investigated. For the chosen range of parameters (relatively low temperature at the strike points), the inventory of the WEST divertor evolves as the power 0.2 of the puffing rate and as the power 0.3 of the IP. The inventory at the strike points is highly dominated by ions whereas it is dominated by neutrals in the private zone. Increasing the fuelling rate increases the retention in the private zone and decreases slightly the retention at the strike points. Increasing the IP increases the inventory at the strike points and does not affect much the inventory at the private flux region. The inventory of the ITER divertor is not strongly dependent on the divertor neutral pressure. The inventory increases from 0 Pa to 7 Pa and then decreases slightly from 7 Pa to 10 Pa. After 107 s of continuous exposure, the maximum inventory in the ITER divertor was found to be 14 g. The inventory is not maximum at the strike points due to the high surface temperature of the monoblocks in this region. The maximum accumulation of H in the ITER divertor is below 5 mg per 400 s discharge and below 2 mg per 400 s discharge after 200 discharges.

This contribution presents the recent effort at CEA and French federation for fusion to simulate edge plasma transport with the new code SOLEDGE3X. The latter can be used both as a 2D transport code or as a 3D turbulence code. It makes possible simulating edge plasma up to the first wall including the complex wall geometry. It also includes neutral recycling and impurity sputtering, seeding and transport. In order to improve turbulence description in transport simulation, a reduced model for turbulence intensity prediction has been derived and implemented, based on 'k-epsilon' like models from the neutral fluid community. Applications to a JET L-mode detached plasma and to a WEST plasma are used as illustration of the code abilities.

The WEST experiment is currently operating with tungsten plasma-facing components and testing ITER-like divertor monoblocks. In order to support WEST experiments interpretation, numerical analyses were carried out. Starting from WEST experimental data, realistic background plasma conditions were reproduced through SolEdge-EIRENE and used as input for ERO2.0 simulations to investigate tungsten migration. Tungsten contamination due to the different plasma-facing components was modelled under different plasma conditions, highlighting a non-negligible contribution of tungsten coming from the tokamak main chamber. Tungsten penetration factor was computed and used as an indication for tungsten screening by the background plasma at the different tokamak plasma-facing components. Simulations showed the main chamber components to be very weakly screened. Light impurities charge was showed to influence not only tungsten sputtering, but also its probability to enter the confined plasma. Simulations results indicated that even when the tungsten source is not heavily influenced by self-sputtering, contamination of the confined plasma can be strongly impacted by it in low density background plasma conditions. Finally, a one-to-one comparison between tungsten visible spectroscopy at the lower divertor from experimental data and from synthetic diagnostics was performed, showing that it is possible to reproduce a realistic lower divertor signal following experimental evidence on light impurities asymmetry between the targets.

The impact of triangularity on edge plasma transport and turbulence is addressed from full 3D turbulence simulations performed with TOKAM3X. Flux driven fluid simulations are run on analytical magnetic equilibria generated with positive and negative triangularity δ in a bottom limiter configuration. The conservation of the energy is assured by the increase of the bottom limiter radial position from δ>0 to δ<0. Changing the triangularity impacts both the plasma equilibrium and the turbulence. In particular, negative triangularity leads to a reduction of the density and electron temperature decay lengths in agreement with the literature. Concerning the turbulence, in all the simulations, it remains ballooned with an enhanced level of fluctuations at low field side in comparison to the high field one. Moreover, no clear trend is visible on the relative level of fluctuations of both density and electron temperature in the CFR whereas an enhancement (resp. reduction) is visible in the scrape-off layer at the low field side midplane for the negative (resp. positive) triangularity simulations. This behaviour differs from TCV and DIII-D measurements which show the benefit of negative triangularity in terms of turbulence reduction and increased confinement. However, no conclusion is drawn from our preliminary study concerning the impact of triangularity on the turbulent transport. Change in triangularity impacts many simulation control parameters, as in the experiments, and that the analysis of its impact alone on the dynamics of the plasma is not obvious in this configuration.

The WEST (W-tungsten Environment in Steady-state Tokamak) is the transformation of the Tore Supra tokamak from a carbon limiter to a tungsten divertor configuration. In recent experiment campaign C5, the divertor pumping capability has been improved by the sealing of the space between the divertor baffle and the vacuum vessel. This modification leads to an interesting question about how the leakage below the outer baffle will influence the plasma in WEST. To investigate this question, we made simulations of three leak cases and one no leak case with the help of SOLEDGE3X-EIRENE transport code. Through the analysis of the simulation results, we work out the impact of leakage under the outer baffle and explain the physics behind some phenomena observed. We investigated the density regimes analyzing how parameters like density, temperature, particle flux, and neutral pressure evolve as functions of gas puff rate and electron density at the outer midplane separatrix comparing among the cases with and without leak. The results show that the case without leak has better performance than the cases with leak in trapping the neutral particles and has higher neutral pressure near the baffle by more than 35%, which can lead to greater power dissipation in the divertor, thus lower the detachment threshold in upstream separatrix density by more than 10%. At the same time, the operational gas puffing range becomes broader by a factor from 1 to 5 in the case without leak with respect to the three cases with leak. Simulation results are confronted with experimental data from the configurations with and without leak under the baffle in order to get further insight on the neutral particle circulation in WEST