In recent years, irradiation-assisted stress-corrosion cracking (IASCC) of austenitic stainless steels for core internal component materials became a subject of concern for in light water reactors (LWRs). Intergranular cracking of baffle former bolts has been found in pressurized water reactors (PWRs) after long operating periods. Therefore, the authors investigated the possibility of IASCC in austenitic stainless steels used for core internal materials of PWRs in order to estimate the degradation of PWR plants up to the end of their lifetime. In this study, in order to verify the hypothesis that the IASCC in PWRs can be caused by the primary water stress-corrosion cracking (PWSCC) as a result of radiation-induced segregation (RIS) at grain boundaries, the authors melted materials whose bulk compositions simulated the grain boundary compositions of irradiated austenitic stainless steels. The effects of chromium, nickel and silicon content on PWSCC susceptibility was studied by slow strain-rate tensile (SSRT) tests. In order to improve the IASCC resistance of austenitic stainless steels for PWRs, authors developed modified 316CW(cold worked) and high chromium stainless steels. The former steel has high chromium content and ultra low impurity elements within the specification of chemical composition for ASTM A193 B8M Type 316 stainless steel. The latter steel has high chromium content of up to 30% chromium and ultra low impurity elements. Both materials are aged after solution annealing, in order to precipitate the M23C6 carbides coherent with the austenitic matrices along the grain boundaries and to recover the sensitization.
|ジャーナル||ASTM Special Technical Publication|
|出版ステータス||Published - 2001 1月 1|
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