This paper reviews recent results of the study of dust in magnetic fusion devices. Assessment of the role of dust in current fusion devices and ITER is presented. Dust diagnostics, main experimental results, different theoretical aspects of dust in fusion plasmas, as well as the comparison of theoretical estimates and numerical simulations with available experimental data are discussed. Some limitations of current theoretical models of dust–plasma interactions and the gaps in current experimental and theoretical approaches to dust study in fusion devices are considered. Possible directions for further advancements are suggested.

https://iopscience.iop.org/article/10.1088/0741-3335/53/8/083001/pdf

Dust in magnetic fusion devices

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.

/pdf/plasma-surface-interaction-scrape-off-layer-and-divertor-218wl02jsd.pdf

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

The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of HIPB98(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. (Some figures in this article are in colour only in the electronic version)

http://iopscience.iop.org/article/10.1088/0029-5515/51/7/072001/pdf

Study on H-mode access at low density with lower hybrid current drive and lithium-wall coatings on the EAST superconducting tokamak

To achieve desirable plasma density control, a supersonic molecular beam injection (SMBI) feedback control system has been developed recently for the EAST tokamak The performance of the SMBI and gas puffing (GP) feedback systems were used and compared The performance of pulse width mode is better than that of pulse amplitude mode when GP was used for density feedback control During one-day experiments, the variation of gas input and wall retention can be clarified into two stages In the first stage the retention ratio is as high as 80–90%, and the gas input is about an order of 1022 D2 However, in the second stage, the retention ratio is at a range of 50–70% The gas input of a single discharge is small and the net wall retention grows slowly The results of the SMBI feedback control experiment was analyzed The shorter delay time of SMBI makes it faster at feeding back control the plasma density The result showed that, compared with GP, the gas input of SMBI was decreased ~30% and the wall retention was reduced ~40% This shows SMBI's advantage for the long pulse high density discharges in EAST

https://iopscience.iop.org/article/10.1088/0741-3335/55/11/115010/pdf

Comparison between gas puffing and supersonic molecular beam injection in plasma density feedback experiments in EAST

Reductions in H content and particle recycling are important for the improvement of ion cyclotron range of frequency (ICRF) minority heating efficiency and the enhancement of plasma performance of the EAST superconducting tokamak. During recent years several techniques of surface conditioning such as baking, glow discharge cleaning/ICRF discharge cleaning, surface coatings, such as boronization, siliconization and lithium coating, have all been attempted in order to reduce the H/(H+D) ratio and particle recycling in EAST. Even though boronization and siliconization were both reasonably effective methods to improve plasma performance, lithium coatings were observed to reduce the H content and particle recycling to levels low enough to allow the attainment of enhanced plasma parameters and operating modes on EAST. For example, by accomplishing lithium coating using either vacuum evaporation or the real-time injection of fine lithium powder, the H/(H+D) ratio could be routinely decreased to about 5%, which significantly improved ICRF minority heating efficiency during the autumn campaign of 2010. Due to the reduced H/(H+D) ratio and lower particle recycling, and a reduced H-mode power threshold, improved plasma confinement and the first EAST H-mode plasma were obtained. Furthermore, with increasing accumulation of deposited lithium, several new milestones of EAST performance, such as a 6.4 s-long H-mode, a 100 s-long plasma duration and a 1 MA plasma current, were achieved in the 2010 autumn campaign.

https://iopscience.iop.org/article/10.1088/0741-3335/54/1/015014/pdf

Comparison of various wall conditionings on the reduction of H content and particle recycling in EAST

Vacuum and wall conditioning system on EAST

Oxidation for deposits removal and hydrogen release on HT-7

Design of actively cooled flat toroidal limiter with CuCr heat sink for the HT-7 superconducting tokamak

He–ICR cleanings on full metallic walls in EAST full superconducting tokamak

Oxygen removal with D2-ICR cleanings after oxidation experiment in HT-7

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