Tensile and stress corrosion cracking properties of type 304 stainless steel irradiated to a very high dose

摘要

Certain safety-related core internal structural components of light water reactor, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20-100 displacement per atom or dpa) by the end of life. Our databases and mechanistic understanding of the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high doses, i.e. is it purely mechanical failure or is it stress-corrosion cracking? In this work, hot-cell tests and microstructural characterization wee performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to ≈50 dpa at ≈370 deg. C.