Influence of Stress Intensity and Loading Mode on Intergranular Stress Corrosion Cracking of Alloy 600 in Primary Waters of Pressurized Water Reactors Available to Purchase

The steam generator in a pressurized water reactor (PWR) of a nuclear power plant consists mainly of a shell made of carbon (C) steel and tubes made of alloy 600 (UNS N06600). However, alloy 600 suffers environmentally induced cracking with exposure to high-temperature primary water. The susceptibility of alloy 600 to intergranular stress corrosion cracking (IGSCC) was investigated as a function of the level of applied stresses and mode of loading. Constant load tests were conducted with specimens prepared from thin wall tubes, and constant deformation tests were conducted using specimens prepared from plates. With tubes exposed to primary water at 330°C, the crack propagation rate (CPR) was found to increase from 1 x 10^–11 m/s at a stress intensity (Ki) of 10 MPa√m to 1 x 10^–9 at Ki = 60 MPa√m. CPR obtained using compact specimens prepared from plates were 1 order of magnitude lower than values measured in tubes at the same temperature and in the same solution at each stress intensity. The corollary was that values of crack propagation and threshold stress intensities obtained using compact specimens could not be extrapolated to the behavior of thin wall tubes.