Application of the FRI crack growth model for neutron-irradiated stainless steels in high-temperature water of a boiling water reactor environment

摘要

This study considered a methodology to reflect the effect on material properties by neutron irradiation onto the FRI model of the crack growth rate (CGR) of stress corrosion cracking developed at Tohoku University. Yield strength and strain hardening exponent were evaluated by a tensile test for irradiated stainless steels, and the relationship between mechanical property and strain to fracture of the oxide film was derived. Effect of the radiation-induced segregation on CGR was also discussed. The experimental tendencies for the CGR to increase with dose and almost saturate above 3 dpa were well replicated by the model calculations.