Tritium and Helium Behavior in Irradiated Beryllium
Fusion Science and Technology / Volume 19 / Number 3P2B / May 1991 / Pages 1707-1714
Material and Tritium / Proceedings of the Ninth Topical Meeting on the Technology of Fusion Energy (Oak Brook, Illinois, October 7-11, 1990) / dx.doi.org/10.13182/FST91-A29588
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Large quantities of Be (>100 metric tons) are planned for use in the ITER blanket design to enhance tritium breeding and to act as a thermal barrier between coolant and breeder. Tritium retention/release and He-induced swelling are important issues in blanket design. The data base on tritium and helium behavior in Be is reviewed. New data on tritium retention/release and He bubble growth are presented for Be irradiated to 5 × 1022 n(E > 1 MeV)/cm2 at ∼75°C and postirradiation-annealed for 700 hours at 500°C. A model (diffusion/desorption) is proposed and tested against the data base to determine tritium diffusivity and the desorption rate constant. Similarly a model for He-induced swelling is developed and tested against the data base. The dependence of tritium retention and release on He content and impurities (e.g. BeO) is also explored.
