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Removal of Tritium from Fusion Reactor Blankets. Annual Report, FY 1977

机译:从聚变反应堆毯中去除氚。年度报告,1977财年

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The program is concerned with investigating the chemistry of those compounds and alloys of lithium which have acceptable physical and neutronic properties. In particular, information is needed on whether tritium can be removed rapidly, and how removal rates are affected by particle size, temperature, dose rate, etc.; how stable the solids are under long irradiation; and whether they are chemically compatible with the other materials that will be present in a blanket. The ultimate objective is to specify an optimum compound or alloy; to describe methods of synthesizing it and incorporating it, with or without diluents, into blanket modules; and to be able to predict its long-term behavior under operating conditions. (ERA citation 03:034283)

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