Supercritical water-cooled reactors (SCWR) are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current light water reactors, LWRs) and considerable plant simplification. Materials for SCWR must have adequate corrosion and stress corrosion cracking response. Initial examinations indicate ferritic-martensitic steels may have excessive oxidation for thin-walled reactor components and austenitc alloys are susceptible to stress corrosion cracking, with the problem exacerbated by radiation. Two novel techniques, grain boundary engineering and surface modification, have shown initial promise in improving performance of both alloy classes in supercritical water. The promise of these techniques is demonstrated for an austenitic alloy, Alloy 800H, and a ferritic-martensitic alloy, HT9.

Subject
Ions, Stainless alloys, Oxides, Nickel based alloys, Weight gain, Irradiation assisted stress corrosion cracking, Oxide surfaces, Steel, Oxygen, Grain boundary, Alloys, Stress corrosion cracking, Oxidation
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2007
Association for Materials Protection and Performance (AMPP)
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