Failure Analysis on Alloy 600 Components in Korean Nuclear Power Plants

Primary water stress corrosion cracking(PWSCC) of alloy 600 in a PWR has been reported in the control rod drive mechanism (CRDM), In original PWRs, the SCC was not considered appropriately. Beginning in the mid seventies the world's PWR plants suffered from a sequence of SCC events mostly confined to S/G tubes, initially ODSCC then PWSCC. In forged alloy 600 materials, PWSCC was first reported in the Bugey 3 vessel head penetration in September 1991. All reactor vessel heads(RVH) with alloy 600 penetrations (54 VH out of 58) were replaced in France. Other PWRs experienced cracking attributed to PWSCC of major primary side welds made from Alloy 182 at the end of the year 2000. The three events concerned dissimilar metal butt welds between the main austenitic stainless steel primary circuit piping and the outlet pressure vessel nozzles of Ringhals 4 and V. C. Summer and some J-groove welds of the CRDM of the RVH at Oconee 1. In addition to the RVH, PWSCC of Alloy 182/82 has been reported at Bottom Mounted Instrumentation (BMI) nozzle J-welds, Steam Generator (SG) drain Jwelds drain nozzle and SG tube sheet cladding. As of the year 2006, 344 PWSCC incidents of Alloy 600/82/182 were reported in RVH, SG, Pressurizer, and other primary side pipings. With regard to Alloy 600TT SG tubes which are known to be much more resistant to SCC than Alloy 600MA tubes, US plants (Seabrook and Vogtle 1) experienced ODSCC. The objective of the present work is to review the PWSCC events and some countermeasures to manage the degradation. Moreover this work deals with the ODSCC occurred in Alloy 600TT SG tubes