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    Reactor Pressure Vessel Structural Implications of Embrittlement to the Pressurized-Thermal-Shock Scenario


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    A deterministic fracture-mechanics parametric-type analysis of a generic pressurized-water reactor pressure vessel has been conducted for loading conditions imposed by a specific category of hypothetical pressurized-thermal-shock transients. The time in the life of the vessel for which the calculations were made corresponds to attainment of the limiting nil-ductility transition reference temperature specified by the U. S. Nuclear Regulatory Commission's pressurized thermal-shock-issue-related screening criteria.

    The transients considered were characterized by a constant pressure and an exponential decay of the downcomer coolant temperature. The decay constant, the final temperature of the coolant, and the fluid-film heat-transfer coefficient were the variable parameters. A search was performed to determine the critical pressure corresponding to incipient crack initiation for a range of crack depths up to 20% of the wall thickness. Results indicate that the critical pressure is greater than the normal operating pressure, if the coolant final temperature is greater than 150°C.

    The fracture mechanics model used in the study tends to be conservative in the sense that it ignores possible beneficial effects of warm prestressing and cladding.


    pressure vessel steels, radiation effects, screening criteria, reactor pressure vessels, over-cooling accident, pressurized thermal shock, neutron embrittlement, linear elastic fracture mechanics, crack initiation, reference temperature, flaws (materials), stress intensity factor

    Author Information:

    Iskander, SK
    Resident engineer, U. S. Nuclear Regulatory Commission,

    Universitaet Stuttgart (MPA),

    Sauter, AW
    Staff members, Universitaet Stuttgart (MPA),

    Föhl, J
    Staff members, Universitaet Stuttgart (MPA),

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP23035S