Formation of deuterium–carbon inventories in gaps of plasma facing components

Abstract

Plasma facing components for ITER will be manufactured as macro brush structures or with castellated surfaces. Material samples with gaps of similar geometry as intended for ITER were exposed to different plasma conditions in TEXTOR, DIII-D and ASDEX Upgrade. In all devices a decrease of both carbon and deuterium inventories at the side faces from the gap entrance into the gap with scale-lengths in the mm range is found. The fraction of D retained at the gap surfaces is in the range of 0.4-4% of the incident flux. Main parameters determining the retained D-fraction are the temperature of the respective surfaces and the carbon fraction in the incident flux. Extrapolation of tritium inventory growth rates to ITER dimensions assuming the measured retention fractions at T>200ºC and using a D/T-flux distribution with a carbon fraction of from B2/EIRENE simulations of an ITER H-mode discharge yields 1% ≈ 1 a contribution to the increase of the total in-vessel tritium inventory in the range of 0.5-5g T/discharge.

Dynamics of erosion and deposition in tokamaks

“…These remote locations are hard for currently available cleaning techniques to access [2] and can offer a reservoir for 2 tritium accumulation. Gaps of castellated PFCs are a special type of remote areas specific for ITER [3]. The crucial role of shadowed areas is illustrated by the deuterium-tritium experiment DTE1 in JET in 1997, after which the vast majority of retained tritium was found in the form of hydrogen rich carbon layers on the cooled louvre structures in the pumping duct of the inner divertor leg [4,5].…”

Castellated structures for ITER: Differences of impurity deposition and fuel accumulation in the toroidal and poloidal gaps

“…Given the difficulties in fuel removal [3][4][5], there is a concern that the radioactive fuel will be accumulated in the gaps of such a castellation [6][7][8][9]. Dedicated studies addressing the performance of castellated structures under plasma impact are ongoing on several tokamaks [10][11][12]. Series of experimental investigations with ITER-like castellated samples were performed in TEXTOR along with dedicated modeling to address the issues of carbon transport and fuel accumulation in the gaps [13][14][15] and on molten layer formation and its impact on the performance of castellation [16].…”

Fuel removal from tile gaps with oxygen discharges: reactivity of neutrals

“…Taking into account that recent experiments in DIII-D, Textor, ASDEX Upgrade, and JT-60-U [14,16,37] showed deposition profiles with 1/e decay lengths that are typically a few times the gap width, one can conclude from our experiments, that cleaning of remote areas with oxygen glow discharges can be successful at elevated temperatures but will be ineffective at room temperature as the surface loss probability of the species dominating the erosion process is too large. The optimal strategy to remove redeposited carbon films by oxygen glow discharges in future fusion devices would be to operate at the highest temperature possible.…”

“…Apart from deposition of thick layers on plasma-exposed surfaces [12,13] co-deposition in narrow gaps of castellated structures is a major concern [13][14][15][16]. Especially the side walls of gaps are expected to be the main deposition areas [12][13][14][15][16].…”

“…Especially the side walls of gaps are expected to be the main deposition areas [12][13][14][15][16]. Also deposition far away from the first wall in socalled remote locations may contribute significantly to the total fuel build-up (see [2] and references therein).…”