STP1405

    Comparison of Transition Temperature Shifts Between Static Fracture Toughness and Charpy-v Impact Properties Due to Irradiation and Post-Irradiation Annealing for Japanese A533B-1 Steels

    Published: Jan 2001


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    Abstract

    It is assumed in the integrity analysis of reactor pressure vessel (RPV) that the irradiation-induced shift of fracture toughness in the ductile-brittle transition region is the same as the Charpy transition temperature shift. To confirm this assumption, both shifts are compared using irradiated and post-irradiation annealed RPV steels made by Japanese manufacturers. Five ASTM A533B class 1 plates having high- and low-level impurities, which correspond to first generation and modern Japanese reactors are used in this study. Neutron irradiation of precracked Charpy-v (PCCv) specimens as well as standard Charpy-v specimens was carried out at the Japan Materials Testing Reactor (JMTR). The values of fast neutron fluence for this study are 2 to 13×1019 (n/cm2, E>1MeV) considering typical fluence values at the EOL and extended operation of Japanese PWRs. The irradiation temperature was controlled within the range of 290±10°C. Thermal annealing treatments at 350°C and 450°C for 100 hours were performed for irradiated PCCv and Charpy-v specimens. The master curve approach according to ASTM Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range (E1921) was applied using PCCv specimens. The irradiation-induced shifts of the reference temperature, ΔT0, obtained in this study were spread in the range of 10 to 200°C. The annealing recoveries of the reference temperature were compared with those of the Charpy transition temperature. Although large scatter was seen in the relation between ΔT0 values and Charpy 41J shifts, both shifts were almost the same in the average.

    Keywords:

    reactor pressure vessel, surveillance, structural integrity, irradiation embrittlement, fracture toughness, precracked Charpy, master curve, lower bound


    Author Information:

    Onizawa, K
    Senior Engineer and Head, Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken

    Suzuki, M
    Senior Engineer and Head, Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken


    Paper ID: STP10527S

    Committee/Subcommittee: E10

    DOI: 10.1520/STP10527S


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