| In order to improve the corrosion resistance of nuclear power equipment，austeniticstainless steel and nickel alloy are mainly selected as structural materials, whose corrosionresistance is formed due to the steel surface of the chromium-rich oxide film (passivationfilm). However, austenitic stainless steel and nickel alloy long-term serve in high temperatureand high pressure and irradiation environments of nuclear pressure vessel, and the stresscorrosion cracking (SCC) would be induced by the residual stress and work loading, whichmight birng the nuclear power equipment to an extremely dangerous edge as SCC crackgrowing. Therefore, stress corrosion cracking (SCC), as a representative of environmentallyassisted cracking (EAC) in austenitic stainless steel and nickel-based alloy is one of the keyissues in long term safe service of equipment and structure in nuclear power plants.Because of the importance of this issue, many research institutes of the majorindustiralized countires performed a large number of EAC crack growth experiments of keynuclear materials based on standard rfacture mechanics specimens and simulatedhigh-temperature water environments, and obtained a lot of experimental data. Currently, theEAC was detected in many in-service nuclear structures, which urgently need to assess andpredict. Due to the complexity of the crack geometry and mechanical state in the actualnuclear structure, it is obviously not enough to evaluate EAC expanding situation in thenuclear structures only by using the EAC experimental data of related material.Li order to improve the interface between EAC growth rate data of the mateiralsobtained in laboratory and EAC growth of actual components in nuclear power plants, theresearches done in this dissertation are as follows. The EAC mechanism of austenitic stainlesssteel and nickel-based alloy in high temperature water environment of nuclear pressure vesseland piping was studied, and current prediction models of EAC growth rate proposed by scientists were analyzed and compared, and the oxidation mechanism and applicablecondition of sliding model was specially investigated. Based on discussions of EACmechanism of austenitic stainless steel and nickel based alloys in high temperature waterenvironments, the main factors affecting EAC growth of the nuclear structure wereinvestigated to establish a foundation for evaluating EAC growth situation of the actualcomponents in light water reactors. Based on experimental result analysis，numeircalsimulation and theoretical analysis of an EAC expeirment by using a tube-shape specimenwith inner surface cracks，an EAC prediction model was established incorporated by thecrack tip plastic strain rate and elastic-plastic finite element combining, which can be used topredict EAC growth statue of the actual component with complex crack shape and themechanical state in light-water reactors, and it was verified by expeirment. Through atheoretical analysis and ifnite element simulation on related electric ifeld, an in-situmeasurement approach of crack growing length (depth) by current potential drop (DCPD)was investigated, which provides a theoreitcal basis of developing new crack monitor andcalibrating parameters in DCPD. It was detail investigated that the distirbution of the stressand strain and the rfacture mechanics parameters along the crack rfont of a tube~shapespecimen with inner surface cracks, which establishes a foundation for obtaining the mainmechanical parameters nearby crack region of the actual components with complex geometryand mechanical state in light water reactors. By using a two dimension CDCB specimen anda three dimension tube-shape specimen with inner surface cracks, the effect of a single tensileoverload on EAC crack growth rate was investigated in detail，which establishes a foundationto understand the effect of load diversification on EAC experiments, and the effect of seismicand other overloading on EAC growth status. |
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