In-vessel tritium retention and removal in ITER-FEAT

Erosion of the divertor and first-wall plasma-facing components, tritium up take in the re-deposited films, and direct implantation in the armour mater ial surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the d erivation, and discuss the results, of current predictions of tritium inven tory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are ber yllium on the first-wall, carbon-fibre-composites on the divertor plate nea r the separatrix strike points. to withstand the high thermal loads expecte d during off-normal events, e.g.. disruptions, and tungsten elsewhere in th e divertor. Tritium co-deposition with chemically eroded carbon in the dive rtor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEA T. This demands efficient in-situ methods of mitigation and retrieval to av oid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., similar to 350g of in-vessel co-deposit ed tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly dis cussed.