Crack initiation of neutron-irradiated 304 L stainless steel in PWR primary water
Source abstract
Abstract
Abstract Stress corrosion crack (SCC) initiation of neutron-irradiated 304 L stainless steel was studied following irradiation to various doses. Tensile test results show that SCC susceptibility dependence of the alloy on dose is slight. Grain boundary (GB) oxidation was observed prior to crack initiation and DCs always proceeded before or accompanies crack initiation in the bend test. The DC-GB intersections were the predominant crack initiation sites due to the high local stress. Constant load tests resulted in greater GB oxidation and cracks initiated at applied stress as low as 50% of the irradiated yield strength due to the long exposure time.
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Research context
OpenAlex topics: Fusion materials and technologies · Nuclear Materials and Properties · Hydrogen embrittlement and corrosion behaviors in metals
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