Grain Boundary Modification During Neutron Irradiation at Intermediate Temperatures
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Grain Boundary Modification During Neutron Irradiation at Intermediate Temperatures S. M. Bruemmer, D. J. Edwards, V. Y. Gertsman and E. P. Simonen Pacific Northwest National Laboratory P.O. Box 999, Richland, WA 99532 ABSTRACT Grain boundary compositions and near-boundary microstructures have been measured in complex Fe-Cr-Ni alloys after neutron irradiation at intermediate temperatures where nanometer-scale damage promotes structural integrity problems in nuclear reactor systems. Radiation-induced segregation (RIS) and dislocation loop microstructures have been determined as a function of irradiation dose up to 13 dpa and at temperatures near 280oC. The most significant effect on RIS was the grain boundary structure (low-energy special boundaries versus high-energy random boundaries) and composition (enrichment of Cr and Mo) before irradiation. Grain boundary character distribution did not change with irradiation and only high-energy boundaries exhibited significant radiation-induced changes. The initial grain boundary composition in mill-annealed stainless steels was difficult to remove during subsequent irradiation and retarded the development of Cr- and Mo-depleted regions. The predominant microstructural feature produced at irradiation temperatures below ~300oC was faulted dislocation loops. A distinct denuded zone was observed at high-energy boundaries in materials irradiated at low-to-intermediate doses that disappeared at higher doses. Heat-to-heat differences were detected in denuded zone characteristics that could not be directly related to composition, RIS or matrix microstructure. Radiation-induced grain boundary changes are evaluated in relation to the current understanding of irradiation-assisted SCC in stainless steel core components. INTRODUCTION Degradation of core internal components in commercial nuclear power reactors has been a growing concern for electric power utilities worldwide. Failures have occurred in Fe- and Nibase austenitic stainless alloys after many years of service in both boiling water reactor (BWR) and pressurized water reactor (PWR) systems. Intergranular cracking is observed in materials exposed to a significant fluence of neutron radiation in the reactor coolant environment (oxygenated or hydrogenated water at about 290°C, but temperature can range from 270 to 370°C in specific locations). Since cracking susceptibility requires the combination of radiation, stress and a corrosive environment, the failure mechanism has been termed irradiation-assisted stress corrosion cracking (IASCC). Recent reviews [1-6] have described much of the current knowledge related to IASCC service experience and laboratory investigations. The importance of neutron fluence on IASCC has been well established. Intergranular (IG) SCC is promoted in austenitic stainless steels when a threshold fluence is reached although the threshold varies with stress, water chemistry, etc. The effect of neutron fluence on IG failure in high-temperature water environments is illustrated in Figure 1. Cracking is observed
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