Search

First principle modelling of edge plasma turbulence including neutrals and plasma recycling on the wall remains a challenge, in particular due to the long time scales necessary to simulate to reach particle balance. In this contribution, we propose a strategy to address these long time scales with the fluid code SOLEDGE, resorting to 2D reduced models for turbulence as well as 3D coarse grid simulations. The approach is applied to simulate TCV-X21 reference plasma scenario for edge turbulence modelling validation.

The fluid-drift code SOLEDGE3X, developed by CEA/IRFM in collaboration with Aix-Marseille University, is a powerful tool for simulating transport and turbulence in tokamak edge plasmas with axisymmetric magnetic configurations. In tokamaks such as WEST, the pronounced toroidal magnetic ripple significantly affects plasma confinement and power exhaust, modulating both the poloidal and toroidal components of the equilibrium field. Using a discrete Biot–Savart law, the ripple field is calculated as a magnetic perturbation on the SOLEDGE3X mesh. The transport model and parallel gradient solvers have been enhanced to incorporate the new radial magnetic field component. Preliminary simulations of a WEST scenario reveal a heat deposition pattern in the divertor region consistent with observations from infrared camera experiments.

Boundary plasma simulations are essential to estimate expected divertor and first wall (FW) heat and particle loads on ITER during burning plasma operation. A key missing feature of existing SOLPS simulations (Pitts et al., 2019) is the absence of a plasma solution out to the main chamber walls, essential to self-consistently estimate the gross sputtering of wall material. Here, SOLEDGE3X is applied for the first time to obtain up-to-the wall burning plasma solutions of the ITER boundary plasma at the nominal PSOL = 100 MW of the main SOLPS database simulations, including He ash, Ne seeding but without fluid drifts. Compared with the most recent SOLPS-ITER simulations, our simulations show differences in the exact impurity distribution, but the key results for divertor and wall heat flux remain consistent. In the context of the ITER re-baselining exercise (Pitts, 2024), in which the Be FW armour is proposed to be exchanged for tungsten (W), estimates of W wall sources are key to the assessment of likely core contamination and hence impact on fusion gain. We compare the W gross erosion rates due to the different species excluding W self-sputtering. For the cases simulated spanning 0.27%–0.47% separatrix-averaged Ne concentration and 7.5×1022s−1−1.95×1023s−1 D fuelling, Ne8+ remains the largest contributor to the sputtering flux with the largest source being the outer divertor and baffle. The species-wise contribution to W sputtering changes with fuelling with sputtering due to lower Ne charge states being significant at low D fuelling. In general, the gross W sputtering source is found to decrease with increase in D fuelling and increase with increased Ne seeding.

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.

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.

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>0to δ

To understand the beneficial effects of impurity boron (B) on the plasma, the transport of B species in both atomic and ionic states in the edge plasma and the ensuing evolution of the background plasma are simulated during real-time B powder injection by using two-dimensional fluid code SOLPS-ITER. The initial background plasma profiles for the simulation are reconstructed based on an EAST upper-single null discharge (shot No. 70601). With the injection rate of B atoms, 4.0 × 1021 atoms/s, the simulation shows that effective charge number is less than 1.1 at the inner boundary R - Rsep = −5.2 cm (background plasma density, 5.1 × 1019 m−3) at the outer mid-plane indicating that the B impurity does not have much influence on the energy confinement. The present work finds that the impurity B exhibits very specific features in the edge plasma, mostly present in the narrow region just outside the separatrix. The particular distribution pattern can reduce the heat flux greatly, leading to the divertor detachment. The divertor detachment can be achieved easily under an injection rate of B powder in a reasonable range. This study suggests that the impurity B has a promising potential application in future tokamaks.