Irradiation assisted stress corrosion cracking (IASCC) is one of the degradation phenomena of reactor internals in boiling water reactors (BWRs) and it is necessary to evaluate the crack growth rates (CGRs) of IASCC for their safety assessment. Many CGR testing of neutron irradiated austenitic stainless steels (SSs) have been conducted and CGR data for IASCC have been accumulated. However, there are only few data in the stress intensity factor (K) region lower than 10 MPam^0.5^ and higher than 30MPam^0.5^. In the last few years, a methodology has been developed to describe the effect of neutron irradiation by modifying the input parameters of Hashimoto-Koshiishi model, which is a theoretical prediction model for CGRs based on the slip dissolution/oxidation mechanism of stress corrosion cracking in BWRs. In addition, Eason et al. also developed the disposition curves for IASCC of SSs in light water reactors. Although their predictions for CGRs well describe experimental CGR data at various K levels, it is expected to confirm that these models can reproduce CGRs in K regions where there are currently few experimental data. This study expanded CGR data at high (more than 30 MPam^0.5^) K levels for neutron irradiated austenitic SSs. CGRs were obtained in simulated BWR normal water chemistry (NWC) under constant K conditions for Type 316NG irradiated up to 2.0 dpa. The paper will show the CGRs at several K values in the K regions higher than 30MPam^0.5^, and discuss the capability of predictions at high K levels, comparing the experimental CGRs with the predictions.


Evaluation of Crack Growth Rate of Neutron Irradiated Austenitic Stainless Steels at High Stress Intensity Factor Levels

Division / Sponsors: 
Pressure Vessels & Piping (PVP)


Paper Proceedings: https://asmedigitalcollection.asme.org/PVP/PVP2020/volume/83815

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