STP711: Crack Arrest in Water-Cooled Reactor Pressure Vessels During Loss-of-Coolant Accident Conditions

    Marston, TU
    Project manager and program manager, Electric Power Research Institute, Palo Alto, Calif.

    Smith, E
    Professor, University of Manchester, Manchester,

    Stahlkopf, KE
    Project manager and program manager, Electric Power Research Institute, Palo Alto, Calif.

    Pages: 10    Published: Jan 1980


    Abstract

    The current American Society of Mechanical Engineers Code procedure for crack arrest in a reactor vessel is based on a static linear elastic fracture mechanics analysis and the assumption that arrest occurs when the crack-tip stress intensification, KI, equals some critical value, KIa—the so-called crack arrest toughness. The paper argues that the conditions for crack arrest in a nuclear pressure vessel, when it is subjected to thermal stresses resulting from a hypothetical loss of coolant accident, can be demonstrated with the current Code method, provided that an appropriate value for KIa can be obtained from laboratory tests. The difficulties in selecting the appropriate value of KIa for this component analysis are highlighted, particular attention being given to the effect of test specimen geometry, crack jump length, and material variability. Against this background, the paper outlines a possible procedure for obtaining the appropriate KIa-value. It is emphasized that the viability of the approach may be highly dependent on the structure's characteristics, and an approach considering kinetic (dynamic) effects may be necessary for other crack arrest problems.

    Keywords:

    fracture mechanics, ASME Code, warm prestressing, crack propagation


    Paper ID: STP27460S

    Committee/Subcommittee: E08.08

    DOI: 10.1520/STP27460S


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