Study on critical irradiation dose of ODS ferritic steel for fast reactor fuel cladding at high temperature

Tanno, Takashi ; Otsuka, Satoshi; Yamashita, Shinichiro; Yano, Yasuhide; Kaito, Takeji; Okubo, Nariaki; Jitsukawa, Shiro*; Sawai, Tomotsugu

9Cr-ODS ferritic steel has been developed as the candidate material for fast reactor fuel cladding in JAEA. It is necessary to evaluate the stability of the dispersoid under heavy irradiation at high temperature because the cladding will be irradiated up to 250 dpa at 700C. From the perspective of corrosion resistance, 11Cr-ODS steel also has been developed. Increase of Cr can cause irradiation hardening and ductility decrease. This behavior should be evaluated. Thus we started the irradiation up to 250 dpa with Fe ion irradiation which can achieve high dose in respectively short time. Nano-indentation test showed that 9Cr-ODS steel was hardened due to 60 dpa irradiation at 400C. Irradiation hardening of 11Cr-ODS steel was smaller than that of 9Cr-ODS steel. The reason for the difference is seemed that the hardness of unirradiated 11Cr-ODS steel was larger than that of 9Cr-ODS steel. The irradiation will be continue to obtain data at higher dose.