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Application of the Floating Curve Model for Estimation of Re-Irradiation Embrittlement of VVER-440 RPV Steels Pages: 11 Published: Jan 2000
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View License Agreement Source: STP1366-EB Abstract Radiation embrittlement and its mitigation by annealing of VVER-440 reactor pressure vessel steels are studied. The Russian regulatory approach for prediction of radiation embrittlement of VVER-440 steels is considered. Results of an investigation of materials cut out of operating nuclear power plants are discussed. Different models of re-irradiation embrittlement are compared. A new model for assessment of embrittlement under re-irradiation is developed. Keywords: reactor pressure vessel, radiation embrittlement, post-irradiation annealing, re-irradiation, ductile-to-brittle transition temperature Paper ID: STP12408S Committee/Subcommittee: E10.12 DOI: 10.1520/STP12408S ASTM International is a member of CrossRef. | ||