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Recovery of Retained Tritium from Graphite Tile of JT-60U

Toshiharu Takeishi, Kazunari Katayama, Masabumi Nishikawa, Naoyuki Miya, Kei Masaki

Fusion Science and Technology / Volume 48 / Number 1 / Pages 565-568

July 2005

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Tritium thermal release and full combustion with oxygen were performed on isotropic graphite tiles used for plasma facing material of JT-60U. Approximately 50-80 % of tritium was released by dry argon gas purge and 20-50 % of tritium was released by humid argon gas purge up to 800-1200 °C within one day, respectively. Further several percent of tritium was released by full combustion with oxygen. It was experimentally confirmed that all retained tritium is not released by thermal dry gas purge and by use of isotope exchange reaction at high temperature in such a short period. In the full combustion operation, isotropic graphite begins to combust at higher temperature than 650 °C, but effective combustion temperature was higher than 700 °C. Since it is very difficult to heat the graphite tile attached on the wall of vacuum vessel at higher than 700 °C, it is considered to be not easy to recover all the tritium retained in the graphite while in the vacuum vessel.

 
 
 
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