Tritium and Helium Retention in Neutron-Irradiated Beryllium

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 degreesC 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 degreesC, with a fast neutron fluence (En > 1 MeV) of about 2.1 x 10(25) 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 degreesC and reaches a maximum when the specimens first reach about 1000 degreesC. Although tritium is released between 600 degreesC and 900 degreesC, no helium release is observed in that temperature range. However, after several minutes heating at 1000 degreesC 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.