Plasma–surface interaction, scrape-off layer and divertor physics: implications for ITER

Abstract

The work of the ITPA SOL/divertor group is reviewed and implications for ITER discussed. Studies of near SOL gradients have revealed a connection to underlying turbulence models. Analysis of a multi-machine database shows that parallel conduction gradients near the separatrix scale as major radius. New SOL measurements have implicated low-field side transport as driving parallel flows to the inboard side. The high-n nature of ELMs has been elucidated and new measurements have determined that they carry ~10-20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. Analysis of ELM measurements imply that the ELM continuously loses energy as it travels across the SOL-larger gaps should reduce surface loads. The predicted divertor power loads for ITER disruptions has been reduced as a result of finding that the divertor footprint broadens during the thermal quench and that the plasma can lose up to 80% of its thermal energy before the thermal quench (not true for VDEs or ITBs). On the other hand predictions of power loading to surfaces outside the divertor have increased. Disruption mitigation through massive gas puffing has been successful at reducing divertor heat loads but estimates of the effect on the main chamber walls indicate 10s of kG of Be could be melted/mitigation. Estimates of ITER tritium retention have reduced the amount retained/discharge although the uncertainties are large and tritium cleanup may be necessary every few days to weeks. Long-pulse studies have shown that the fraction of injected gas that can be recovered after a discharge decreases with discharge length. The retention rate on the sides of tiles appears to ~ 1-3% of the ion flux to the front surface for C tiles and ~100x less for Mo tiles. T removal techniques are being developed based on surface heating and surface ablation although ITER mixed materials will make T removal more difficult. The use of mixed materials gives rise to a number of potential processes-e.g. reduction of surface melting temperatures (formation of alloys) and reduction of chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Non-boronized compared with boronized operation of ASDEX Upgrade with full-tungsten plasma facing components

“…In the stable part of the operational domain of AUG (see figure 7), no impact of the boronization state on energy confinement is observed. This is in contrast to results from the molybdenum covered Alcator C-Mod, where a clear reduction of the confinement with unboronized conditions was reported [20]. The confinement degradation in C-Mod has been attributed to high core molybdenum concentrations and corresponding radiative cooling.…”

Convective transport by intermittent blob-filaments: Comparison of theory and experiment

“…Several recent review articles have already described measurements of edge plasma turbulence in toroidal magnetic fusion devices, 1 theoretical progress in describing turbulent edge and SOL transport, [2][3][4] and the implications of these and other empirical results for SOL and divertor physics for ITER. 5 The purpose of the present review is to specifically compare theory, simulation and experiment in the area of blob transport, as a guide for future work. Our review will cover both toroidal fusion and basic plasma devices.…”

Advanced divertor configurations with large flux expansion

“…Divertor geometry, in particular, has been known as a key factor for divertor performance optimization [1,3,4]. The ITER divertor design, a closed divertor with tilted vertical targets and a partial radiative detachment of the strike points, represents an optimized standard divertor geometry based on experimental tokamak studies, theory and modeling developments over the last two decades [1,2,5]. However, for the proposed advanced tokamak and spherical tokamak (ST) based fusion nuclear science facilities [6,7] and for the DEMO [8], the standard divertor solution is insufficient since the expected heat fluxes would exceed the presently allowed steady-state limit of 5-10 MW/m 2 and ELM-like transients 0.1-0.5 MJ/m 2 .…”

“…In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50 % increase in divertor radiation and a peak divertor heat flux reduction from [3][4][5][6][7] MW/m 2 to 0.5-1 MW/m 2 was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region.…”