Burst and Leak Behaviour of SCC Degraded SG Tubes of PWRs

Abstract

Steam generators of Pressurized Water Reactor (PWR) have suffered from many types of corrosion, such as pitting, wastage and stress corrosion cracking (SCC) in the primary and secondary sides (Kim 2003, MacDonald, 1996). Some failure events of steam generator tube have been reported in some nuclear power plants around the world(MacDonald, 1996). In order to prevent the primary coolant from leaking into the secondary side, the tubes are repaired by sleeving or plugging(Benson, 1999). It is important to establish the repair criteria to maintain the plugging ratio within a plugging limit which allows successful plant operation. In the international steam generator tube integrity program (ISG TIP) supported by the US NRC (Nuclear Regulatory Commission), tasks such as in-service inspection technology development, and studies on steam generator tube degradation modes had been undertaken. As a part of the cooperation work, leak and burst tests were carried out, and the burst behaviour of axial mechanical flaws was studied. Leak rate from archive tubes were measured under different water pressures in order to understand the leak behaviour under operating and accident conditions of PWRs.