ABSTRACT
Austenitic stainless steel (SS) used as reactor core component materials becomes susceptible to SCC in high temperature water when irradiated to high neutron fluences. This phenomenon, termed as irradiation assisted stress corrosion cracking (IASCC), has been distinguished from intergranular stress corrosion cracking (IGSCC) which usually occurs of thermally sensitized SS in high temperature oxygenated water. Although the IASCC is phenomenologically quite similar to the IGSCC, its mechanism still remains unknown.
In this work both solution annealed and thermally sensitized Type 304 SSs were irradiated to 3x1019 n/cm2(E>1MeV) at 290 °C in the Japan Material Test Reactor (JMTR). The irradiated as well as the unirradiated specimen materials were tensile tested in argon gas and in high temperature water with different dissolved oxygen (DO) concentrations by slow strain rate tensile (SSRT) SCC tests at 290°C.
The increase in the yield stress of Type 304 SS due to neutron irradiation was within the range of dispersion of published data for the solution annealed Type 304 SS irradiated to higher neutron fluences in the ATR and commercial BWR. The irradiated sensitized material was more susceptible to SCC at 290 °C than the unirradiated sensitized material in high temperature water with more than 0.2 ppm DO. When at <0.01 ppm DO and in argon gas, the irradiated sensitized SS revealed a small fraction of intergranular cracking (IG), but the unirradiated sensitized SS did not. A comparison of the result in this work with published data indicated that the sensitized SS irradiated to 3x1019 n/cm2 was more susceptible to SCC than the solution annealed SS irradiated to over 1022 n/cm2.