Abstract

AbstractAustenitic stainless steels have been widely used for fabricating reactor core-internal components in PWRs due to its high strength, ductility and fracture toughness. The accelerated failure or degradation of austenitic stainless steel represented by IASCC has become one of the key problems affecting the safe and efficient operation of reaction core-internal in PWR nuclear power plants. IASCC is generally divided into three stages: crack initiation, crack propagation and instable fracture. Among the three stages, the crack initiation stage would occupy the major service time, the crack growth stage is featured by quasi-steady crack propagation at a certain rate, and the instable fracture stage should be avoided. Stress intensity factor K at the crack tip is often used to represent the mechanical driving force for SCC as well as IASCC.In this paper, SCC crack growth rate (CGR) data of austenitic stainless steels irradiated in high temperature water were compiled and reanalyzed to evaluate the influence of key parameters such as radiation dose and mechanical properties on IASCC sensitivity and crack growth rate of these materials in PWR nuclear power plant environment. The CGR-K curves of the irradiated materials were also analyzed. The effects of low, medium and high doses of neutron irradiation are compared, and the analysis process is illustrated with examples. In the research process, abnormal CGR and K of materials under a specific irradiation dose was found, so this phenomenon was analyzed. The CGR data and irradiation dose of austenitic stainless steel in different K range were analyzed. And proposed a way to judge the type of change:type I, type II and type III. Finally, the yield strength of the material under the same irradiation dose was found, and combined with other research data, it was further demonstrated that the neutron irradiation dose had a significant effect on the crack growth rate.

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