No abstract

Strong mitigation of edge-localized modes has been observed on Experimental Advanced Superconducting Tokamak, when lower hybrid waves (LHWs) are applied to H-mode plasmas with ion cyclotron resonant heating. This has been demonstrated to be due to the formation of helical current filaments flowing along field lines in the scrape-off layer induced by LHW. This leads to the splitting of the outer divertor strike points during LHWs similar to previous observations with resonant magnetic perturbations. The change in the magnetic topology has been qualitatively modeled by considering helical current filaments in a field-line-tracing code.

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.

The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of H IPB98(y,2) ∼ 1 has been obtained with about 1 MW lower hybrid wave power on the EAST superconducting tokamak. The first H-mode plasma appeared after wall conditioning by lithium (Li) evaporation before plasma breakdown and the real-time injection of fine Li powder into the plasma edge. The threshold power for H-mode access follows the international tokamak scaling even in the low density range and a threshold in density has been identified. With increasing accumulation of deposited Li the H-mode duration was gradually extended up to 3.6 s corresponding to ∼30 confinement times, limited only by currently attainable durations of the plasma current flat top. Finally, it was observed that neutral density near the lower X-point was progressively reduced by a factor of 4 with increasing Li accumulation, which is considered the main mechanism for the H-mode power threshold reduction by the Li wall coatings.