Stress corrosion cracking of the AISI 316L stainless steel HAZ in a PWR nuclear reactor environment

Abstract

Low-carbon alloy steels and stainless steels are widely used in the primary circuits of pressurized water reactor (PWR)-type nuclear reactors. Nickel alloys are employed in welding these materials due to their characteristics such as high mechanical and corrosion resistance and suitable thermal expansion coefficients. Over the last 30 years, stress corrosion cracking (SCC) has mainly been observed in the regions of welds between dissimilar materials that exist in these reactors. The objective of this work is to evaluate, for comparison purposes, the susceptibility to SCC of the heat-affected zone of austenitic stainless steel AISI 316L when subject to an environment that is similar to the primary circuit of a PWR nuclear reactor at temperatures of 303 and 325°C. To carry out this evaluation, the slow strain rate test was used. The results indicate that the SCC is heat activated and that at 325°C, the most significant weak fractures arising from SCC process can be seen.