AbstractSince beryllium is considered one of the best neutron multipliers for the blanket of a fusion reactor, several studies have been initiated to evaluate its behaviour under irradiation for both typical operating and accidental conditions. In the present work the effects of neutron irradiation on the mechanical properties of beryllium were studied with the aid of a computer code describing the most important mechanisms of helium and tritium retention and release. The analysis of irradiated beryllium was carried out following a mechanistic scheme, whereby the microscopic properties of the atomic lattice, the metallographic structure of the material and the geometrical design parameters of the specimens are considered. A model describing the trapping effects on tritium due to chemical reactions with beryllium oxide and capture in helium bubbles was also included in the code. The results of the numerical calculations referring to gas distribution, helium-induced swelling and tritium release from beryllium are compared with corresponding experiments. The good agreement between calculated and experimental data obtained so far constitutes a significant step in validation of the predictive capability of the code. Moreover, preliminary analyses were performed concerning a proposed EC water cooled ceramic blanket for the ITER fusion reactor and for the European BOT DEMO solid breeder blanket with the aim of providing an estimation of the operational and off-normal behaviour of this component.
Investigation of γ lithium aluminate as tritium breeding material for a fusion reactor blanket
Helium content and induced swelling of neutron irradiated beryllium