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- W2041569783 abstract "Among the presently available low-Z materials beryllium represents one of the most promising candidate materials to be used as protection of the first wall and as neutron multiplier in the blanket of a next-step fusion reactor. Both sintered-product blocks and pebbles have been considered, and research and evaluations associated with safety, tritium release, heat transfer, thermal-mechanical and irradiation stability are underway to study the characteristics of several material grades. This paper presents the results of a series of out-of-pile annealing tests up to 1000 °C aimed at investigating both tritium and helium release kinetics from the S-65C beryllium grade irradiated in the BR2 reactor at temperatures of 235, 485 and 600 °C, with a fast neutron fluence (En > 1 MeV) of about 2.1 × 1025 m-2 and with a damage dose of 2.45, 2.1 and 2.3 dpa, respectively. In agreement with previous studies, all the beryllium samples show a tritium release which starts to increase above about 600–650 °C and reaches a maximum when the specimens first reach about 1000 °C. Although tritium is released between 600 °C and 900 °C, no helium release is observed in that temperature range. However, after several minutes heating at 1000 °C the samples showed a burst release leading to the release of essentially all retained tritium. Correspondingly, a peak of helium release was observed. This unambiguous and concurrent release of tritium and helium leads to the conclusion that T and He partially reside in common bubbles in the irradiated material." @default.
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- W2041569783 date "2001-01-01" @default.
- W2041569783 modified "2023-10-10" @default.
- W2041569783 title "Tritium and Helium Retention in Neutron-Irradiated Beryllium" @default.
- W2041569783 doi "https://doi.org/10.1238/physica.topical.094a00083" @default.
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