Post-exposure analysis of C + SiC alloy-coated graphite armor tiles after high heat loadings in a plasma environment

Hsu, Chen-Sheng; Pétrie, T.W.; Trester, P.W.; Hopkins, G.R.; Hong, Rui; Johnson, W. R.

doi:10.1016/s0022-3115(86)80022-8

Cited by 2 publications

(1 citation statement)

References 4 publications

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“…Silicon carbide has long been identified as a promising low-Z refractory material for fusion reactors as a structural material since it is highly resistant to neutrondamage [43][44][45][46][47][48][49][50][67][68][69][70][71][72]. Although it has been less assessed with regard to its use as a plasma-facing material compatible with good plasma performance, SiC is considered to be promising for this application as well [27,28,47,[73][74][75]:…”

Section: Candidate Low-z Refractory Materials For the Main Wallsmentioning

confidence: 99%

Developing solid-surface plasma facing components for pilot plants and reactors with replenishable wall claddings and continuous surface conditioning. Part A: concepts and questions

Stangeby

Unterberg

Davis

et al. 2022

Plasma Phys. Control. Fusion

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It is estimated that pilot plants and reactors will experience rates of net erosion and deposition of solid PFC, Plasma Facing Component, material of 103 – 105 kg/year. Even if the net erosion (wear) problem can be solved, the redeposition of so much material has the potential for major interference with operation, including disruptions due to so-called ‘UFOs’ and unsafe dust levels. The potential implications appear to be as serious as for plasma contact with the divertor target: a dust explosion or a major UFO-disruption could be as damaging for an actively-cooled DT tokamak as target failure. It will therefore be necessary to manage material deposits to prevent their fouling operation. This situation appears to require a fundamental paradigm shift with regard to meeting the challenge of taming the plasma-material interface: it appears that any acceptable solid PFC material will in effect be flow-through, like liquid-metal PFCs, although at far lower mass flow rates. Solid PFC material will have to be treated as a consumable like brake pads in cars. It will be as essential to be able to in situ remove from the vessel the deposited PFC material as it will be to introduce new material to the vessel for refurbishing the claddings of the PFCs. ITER will use high-Z (tungsten) armor on the divertor targets and low-Z (beryllium) on the main walls. The ARIES-AT reactor design calls for a similar arrangement, but with SiC cladding of the main walls. Non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc.) used as in situ replenishable, relatively thin - of order mm - claddings on a substrate which is resistant to neutron damage could provide a potential solution for the main walls, while reducing risk of degrading the confined plasma. Separately, wall conditioning has proven essential for achieving high performance [1], even more so today with the change from C to W PFCs in many tokamaks. For DT devices, however, standard methods appear to be unworkable, but recently powder droppers injecting low-Z material ~continuously into discharges have been quite effective [2-5] and may be usable in DT devices as well. The resulting massive generation of low-Z debris, however, has the same potential to seriously disrupt operation as noted above. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in all current tokamaks, independent of whether their protection tiles are low-Z or high-Z.

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