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    STP560

    Fracture Mechanics Evaluation of the Integrity of an Inlet Nozzle of a Pressurized Water Reactor Vessel Following a Postulated Loss of Coolant

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    Abstract

    A fracture mechanics analysis was performed to evaluate the effects of a postulated loss of coolant on an inlet nozzle of a pressurized water reactor vessel.

    Following rupture of one of the reactor main coolant pipes, ambient temperature water is introduced in the reactor vessel, through the primary inlet nozzles by the safety injection system (in the analysis, the temperature of the water was conservatively assumed to be equal to 32°F). Prior to the transient, the vessel is at high temperature (550°F) and the cold water injection produces a thermal shock resulting in high thermal stresses in the vessel's wall. A continuous circumferential surface crack is assumed to be present at the inside surface of one of the primary inlet nozzles at the time the transient occurs. The postulated crack is subjected to thermal stresses calculated using a finite element technique.

    The stress intensity factor relative to the continuous crack under the actual stress profile normal to the section of the crack was calculated at discrete time intervals during the transient, and compared to the fracture toughness of the material, for different values of the crack depth. The critical crack depths are obtained when the stress intensity factor equals the fracture toughness.

    Various locations of the nozzle were evaluated. The critical region is at the nozzle reinforcement where a 2-in.-deep crack would become unstable 200 s after the beginning of the transient. However, such a crack is readily detectable by current preservice and inservice inspection techniques, and thus the analysis demonstrates that the integrity of the reactor vessel nozzles would be maintained following a postulated loss of coolant accident.

    Keywords:

    fracture properties, nuclear reactors, thermal shock, structural design, stress analysis, cracks, fatigue (materials)


    Author Information:

    Buchalet, CB
    Senior Engineer, Westinghouse Electric Corporation, Nuclear Energy Systems, Pittsburgh, Pa.


    Committee/Subcommittee: E08.08

    DOI: 10.1520/STP33144S