STRESS CORROSION CRACKING OF TYPE 304 STAINLESS STEEL IRRADIATED TO VERY HIGH DOSE

摘要

Certain safety-related core internal structural components of light water reactors, usuallyfabricated from Type 304 or 316 austenitic stainless steels (SSs], accumulate very high levelsof 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 itTadiatedcomponents, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high dose, i.e., is it purely mechanical failure or is itstress-corrosion cracking? In this work, hot-cell tests and microstructural characterizationwere performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-IIreactor after irradiation to =50 dpa at =370℃. Slow-strain-rate tensile tests were conducted at289℃ in air and in water at several levels of electrochemical potential (ECP), andmicrostructural characteristics were analyzed by scanning and transmission electronmicroscopies. The material deformed significantly by twinning and exhibited surprisingly highductility in air, b u t was susceptible to severe intergranular s t r e s s corrosion cracking (IGSCC)at high ECP. Low levels of dissolved O and ECP were effective in suppressing thesusceptibility of the heavily irradiated material to IGSCC, indicating t h a t the stress corrosionprocess associated with irradiation-induced grain-boundary Cr depletion, rather than purelymechanical separation of grain boundaries, plays the dominant role. However, althoughIGSCC was suppressed, the material was susceptible to dislocation channeling at low ECP,and this susceptibility led to poor work-hardening capability and low ductility.