Irradiation assisted stress corrosion cracking is a well-known degradation mode for reactor core
structural materials in both PWRs and BWRs. While it is known that multiple factors such as
stress, a susceptible microstructure and corrosive environment are responsible for IASCC, a better
understanding of its underlying mechanisms has long eluded discovery. The goal of this study is
to provide direct evidence on the role of grain boundary oxidation in crack initiation. To determine
whether initiation of an intergranular crack is due to oxidation of grain boundaries exposed to high
temperature water, 304L stainless steel irradiated in reactor to 69 dpa was exposed to 320 ℃
simulated PWR primary water without application of load for a period of 210 h. The pre-oxidized
sample was then strained to 50% of its irradiated yield strength (i.e., 0.5YS) at 320 ℃ in an inert
argon gas atmosphere in a four-point bend test. Numerous IG cracks were found at the oxidized
grain boundaries. To substantiate the dependence of crack initiation on grain boundary oxidation,
a companion sample of the same dose level was strained to 0.8YS in purified argon without the
pre-oxidation step with no evidence of cracking. These results demonstrate that the irradiated state
is not inherently susceptible to IG cracking, rather it is the oxidation of the grain boundaries that
is responsible for crack initiation.