Irradiation Embrittlement in Grade-300 Steel Internals: A Comprehensive Description and Evaluation Approach

Water-cooled nuclear reactors are equipped with internal structural elements made of grade 300 steels, which are subject to embrittlement over their operational lifetime. More specifically, typical internal components are simultaneously exposed to neutron irradiation and pressurized water at elevated temperatures, making them susceptible to intergranular crack initiation. Irradiation damage accumulation markedly affects plastic flow spreading, which is further coupled with strain localization effects. Concurrent exposure to the reactor coolant also contributes to grain-boundary configuration changes in terms of local chemical composition and the resulting tolerance to applied loading. Our approach for evaluating the material response integrates multiple physical mechanisms applicable to post-irradiation austenitic steels, including strain localization in the form of defect-depleted channels, dislocation pile-ups, and the loss of grain-boundary strength. The above time-dependent effects are evaluated and rationalized in terms of the evolution of the dimensionless factor «R», which depends on: (i) the applied loading level, (ii) the irradiation defect size and number density, and (iii) the local grain boundary cohesive energy evolution. The model predictions are consistent with an applied intergranular crack initiation stress of about 700 MPa after combined exposure to pressurized water and neutron irradiation up to 3 dpa at 300 °C.