Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (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. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

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

Development of two-colour pyroreflectometry technique for temperature monitoring of tungsten plasma facing components

Disruptions are of significant concern to future devices, due to the large amount of energy released during the rapid quenching of the plasma. Disruption mitigation has been performed on MAST, to study the effect on heat loads and disruption time scales in a spherical tokamak. Massive gas injection is performed using a disruption mitigation valve capable of injecting between 0.6 and 1.5 × 1022 particles, corresponding to 10–150 times the plasma inventory. Noble gases are used for mitigation studies, specifically helium, argon, neon and a mixture of 10% argon and helium. The effect of mitigation is studied using a comparison between mitigated discharges and reference unmitigated disruptions.Mitigation has been seen to be effective at reducing peak divertor power loads, with reductions of up to 60% in the divertor power load being observed. The decrease in divertor power load is reflected by an increase in radiated power, mainly as a result of line radiation from the injected impurities. The largest reduction in the divertor power loads are seen for the impurities with Z > 4. The energy load to the divertor has been reduced by between 30–40% of the total stored plasma energy in ohmic, L mode and H mode discharges. Mitigation also affects the current quench times, accelerating the quench of the plasma with increasing injection quantity, with saturation observed at the highest injected quantities.

https://iopscience.iop.org/article/10.1088/0741-3335/54/12/125007/pdf

Characterization of disruption mitigation via massive gas injection on MAST

In nuclear fusion experiments, the plasma facing components are exposed to high heat fluxes and infrared (IR) imaging diagnostics are routinely used for surveying their surface temperature for preventing damages. However the future use of metallic components in the ITER tokamak adds complications in temperature estimation. Indeed, low and variable emissivity of the observed surface and the multiple reflections of the light coming from hot regions will have to be understood and then taken into account. In this paper, a realistic photonic modeling based on Monte Carlo ray-tracing codes is used to predict the global response of the complete IR survey system. This also includes the complex vessel geometry and the thermal and optical surface properties using the bidirectional reflectivity distribution function that models the photon-material interactions. The first results of this simulation applied to a reference torus are presented and are used as a benchmark to investigate the validity of the global model. Finally the most critical key model parameters in the reflected signals are identified and their contribution is discussed.

Development of a realistic photonic modeling for the evaluation of infrared reflections in the metallic environment of ITER.

In the majority of nuclear fusion experiments equipped with carbonaceous plasma-facing components (PFCs), infrared (IR) imaging diagnostics are routinely used for monitoring the surface temperature of the PFCs exposed to high heat fluxes. However, future fusion machines, such as ITER, will be equipped with metallic PFCs. As a consequence, due to their low emissivity, the evaluation of the surface temperature will have to take into account the multiple reflections of the light coming from hot regions. In order to assess the total IR flux collected by the sensor located at the upper port VIS/IR system of ITER, realistic simulations of the IR views of the ITER tokamak have been developed using a Monte Carlo ray-tracing code (SPEOS CAA V5 Based). The simulation includes the thermal emission, the bremsstrahlung radiation, and the reflections thereof inside the real complex 3-D geometry (from CATIA) of the vessel. The IR simulated images in conditions of plasma operation are presented, and the contribution of the reflected flux for two divertor configurations (carbon fiber composite and tungsten) is analyzed. Finally, the achievable measurement accuracy on the surface temperature of the PFCs from the upper port VIS/IR imaging system is discussed.

F. Moreau

Papers

Surface temperature measurements by means of pulsed photothermal effects in fusion devices