High dose neutron irradiation damage in beryllium as blanket material

Abstract The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8×10 22 and 8.0×10 22 cm −2 ( E >1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70–150 °C and neutron fluence of 2.8×10 22 cm −2 ( E >1 MeV) makes up 0.8–1.5%, at 400 °C and fluence of 8×10 22 cm −2 ( E >1 MeV) – 3.2–5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.