IASCC Properties of Cold Worked 316 Stainless Steel in PWR Primary Water

摘要

IASCC (Irradiation Assisted Stress Corrosion Cracking) is a matter of great concern as a degradation of core internal components in light water nuclear reactor. A Japanese national project (PWR) was aiming to establish a data base for plant life management (PLM) which contributes to lifetime evaluation of baffle former bolt (BFB)for operating period up to 60 years. In this project, IASCC tests had been conducted on BFBs irradiated up to maximum 20 dpa and flux thimble tubes (FTTs) irradiated up to maximum 70 dpa removed from actual plants, which were made of cold worked 316 austenitic stainless steel. Constant load SCC initiation tests were carried out in simulated PWR primary water using C-ring type specimens. And obtained mechanical and metallurgical data for irradiated materials. IASCC initiation time became shorter with increasing irradiation dose and increasing applied stress. IASCC initiation was observed even in conditions where applied stress was very lower (400MPa) in high irradiation dose. And IASCC initiation threshold stress became lower with increasing irradiation dose. The difference of the IASCC initiation threshold stress of BFBs and FTTs was caused by the hardening near the outer surface area of the C-ring specimens.