Assessing SCC and IASCC of Austenitic Alloys for Application to the SCWR Concept

摘要

From the standpoint of environmental degradation of material, the selection of alloys for use as structural material in a supercritical water-cooled reactor (SCWR) must include assessment of the corrosion and stress corrosion cracking susceptibility of the alloys in supercritical water. Moreover, as experience in current reactors showed that irradiation-assisted stress corrosion cracking (IASCC) is a major concern, a comprehensive study must include the assessment of the effect of irradiation on SCC in supercritical water. Therefore, such selection faces multiple obstacles. The first is the lack of data on the corrosion and SCC susceptibility of the candidate alloys in this environment. There is a need to produce basic data using complementary experimental techniques. The second is the difficulty to obtain material irradiated in conditions relevant for SCWR. Availability of such material is needed to determine the influence of irradiation and its influence on SCC. Techniques such as proton irradiation are appealing surrogates for neutron irradiation in assessing its effect of stress corrosion cracking initiation, and can be used for screening of various material and environmental conditions. However, neutron irradiation is required to confirm the role of in-core irradiation on crack growth and in performing final verification of the effect of alternative irradiation on candidate alloys. Another obstacle would be the lack of facilities for testing materials in the unirradiated and irradiated state in supercritical water. The University of Michigan has developed a comprehensive programme to assess the stress corrosion cracking susceptibility of austenitic alloys in supercritical water in unirradiated, proton-irradiated and neutron-irradiated state. The cracking susceptibility of unirradiated alloys has been evaluated by a set of constant extension rate tensile, CERT, experiments and by determination of the crack propagation rate by DCPD technique under constant K loading in pure de-aerated supercritical water at temperatures ranging from 400-600°C. The effect of irradiation on alloy microstructure and on stress corrosion cracking were evaluated after proton irradiation. In addition, the construction of a new laboratory allowed the evaluation of the cracking susceptibility of neutron-irradiated JPCA alloy in 400°C and 500°C supercritical water. Results indicate that stainless steel 316L and nickel-based alloy 690 were susceptible to cracking at all temperatures and that the crack propagation rate under constant K loading mode decreased with increasing temperature. Results also showed that irradiation greatly increased the cracking susceptibility of the alloys and that the extent of the increase depends upon the alloy considered. Finally, the neutron-irradiated JPCA alloy showed severe susceptibility to IASCC.