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    Application of the Floating Curve Model for Estimation of Re-Irradiation Embrittlement of VVER-440 RPV Steels


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    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.


    reactor pressure vessel, radiation embrittlement, post-irradiation annealing, re-irradiation, ductile-to-brittle transition temperature

    Author Information:

    Nikolaev, YA
    Leading research scientists, Russian Research Center “Kurchatov Institute”, ORM, Moscow,

    Nikolaeva, AV
    Senior research scientist, Nuclear Safety Institute, Russian Academy of Sciences, Moscow,

    Committee/Subcommittee: E10.12

    DOI: 10.1520/STP12408S