Efficient and rapid tritium removal will be necessary for ITER to meet its physics and engineering goals and R&D is urgently needed to explore potential removal methods and assess their engineering requirements and collateral effects. We consider transiently heating plasma facing components by a cw CO2 laser as a potential means to remove co-deposited tritium. Tritium release from co-deposited layers on graphite is known to increase rapidly with temperatures above 350 deg C, however it is difficult to maintain ITER plasma facing components (PFC's) at this temperature due to excessive water pressure. Transiently heating the PFC surface to high temperature by a scanning CO2 laser would enable an elevated surface temperature to be attained without the complications of bulk heating. The laser would be external to the ITER vacuum vessel and the beam rastered across the PFCs by remotely controlled mirrors. Early work has shown substantial release of deuterium from graphite by 50 ns electron beam heating [1,2]. We will present modeling of the dependence of surface temperature and tritium release on laser power and exposure duration, review the present state of knowledge and discuss the further R&D needed to validate this technique for ITER.

*Work supported by US DOE Contracts: DE-AC02-76-CHO3073 and DE-AC04-94AL85000.

1. B. L. Doyle, F L Vook, J. Nucl. Mater. 85&86 1019 (1979)
2. S. T. Picraux, W R Wampler, J . Nucl. Mater. 93&94 853 (1980).