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

Certain safety-related core internal structural components of light water r eactors, usually fabricated from Type 304 or 316 austenitic stainless steel s (SSs), accumulate very high levels of irradiation damage (20-100 displace ment per atom or dpa) by the end of life. Our databases and mechanistic und erstanding of the degradation of such highly irradiated components, however , are not well established. A key question is the nature of irradiation-ass isted intergranular cracking at very high doses, i.e. is it purely mechanic al failure or is it stress-corrosion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from th e hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to approximate to 50 dpa at approximate to 370 degreesC. Slow-strain-rate tensile tests were conducted at 289 degreesC in air and in water at several levels of electrochemical potential (ECP), and microstructural characteris tics were analyzed by scanning and transmission electron microscopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corro sion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were e ffective in suppressing the susceptibility of the heavily irradiated materi al to IGSCC, indicating that the stress corrosion process associated with i rradiation-induced grain-boundary Cr depletion, rather than purely mechanic al separation of grain boundaries, plays the dominant role. However, althou gh IGSCC was suppressed, the material was susceptible to dislocation channe ling at a low ECP, and this susceptibility led to a poor work-hardening cap ability and low ductility. (C) 2001 Elsevier Science B.V. All rights reserv ed.