Corrosion of candidate materials for supercritical water cooled reactors

Full text: SCWRs are one of the more promising GEN IV nuclear concepts due to enhanced thermal efficiencies and relative compactness as compared with current LWR technology. The current research and development programs are focused on evaluation of SCWR viability. The key to demonstrate the viability of the SCWR design is to identify and develop materials that can assure safe and reliable operation at temperatures and pressures identified for this reactor concept. The high temperature and pressure along with the generation of oxygen and oxidizing species (OH, H2O2, HO2/O2) generated by radiolytic decomposition may result in higher corrosion and SCC rates than experienced with other reactor design. Thus, it must be developed candidate materials with good mechanical properties, microstructural stability and adequate corrosion resistance to high temperature and high pressure supercritical water. The existing data base on the corrosion and SCC of austenitic stainless steels, ferritic-martensitic steel, and nickel based alloys in supercritical water in this moment, is very sparse and data on the behavior of irradiated alloys is non-existent. Developing, testing and selecting suitable materials for internal components of SCWRs present the challenge to the long term performance of engineering materials. Thus, it is necessary to identify the most promising alloys, to investigate degradation of these materials in the supercritical water environments and finally to establish the recommendation of prime candidate materials for further in reactor testing. (author)