Tritium profile in plasma-facing components following D-D operation

QC 20100525 QC 20110926. 11th International Conference on Fusion Reactor Materials (ICFRM). Kyoto, JAPAN. DEC 07-12, 2003
We have investigated the tritium depth profile near the surface of the limiter/divertor tiles used in the deuterium fueled machines, such as TEXTOR, TFTR and JT-60U by means of the imaging plate technique and a tritium survey monitor. Tritium depth profiles near the surface of the sample tiles were estimated by comparing the experimental results to a calculation using a 3-D Monte-Carlo code. In every sample tile, there was little tritium in the range from the surface to 1 mum depth. In contrast, tritium density tended to increase beyond 1 mum depth. These results indicate that the tritium retained near the surface was easily removed by isotope exchange with a deuterium plasma or various other tritium removal operations. On the other hand, such operations did not remove tritium retained beyond 1 mum depth, and this could be a potential issue in a next D-T machine.

Ingår i:Journal of Nuclear Materials: Elsevier BV329-33, s. 874-8790022-31151873-4820

Av författaren/redakt...: Sugiyama, K.; Tanabe, T.; Miyasaka, K.; Masaki, K.; Tobita, K.; Miya, N.; visa fler...; Philipps, V.; Rubel, Marek J.; Skinner, C. H.; Gentile, C. A.; Saze, T.; Nishizawa, K.; visa färre...

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