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

    Published: Jan 1980

      Format Pages Price  
    PDF (160K) 10 $25   ADD TO CART
    Complete Source PDF (6.7M) 10 $55   ADD TO CART


    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.


    fracture mechanics, ASME Code, warm prestressing, crack propagation

    Author Information:

    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.

    Committee/Subcommittee: E08.08

    DOI: 10.1520/STP27460S

    CrossRef ASTM International is a member of CrossRef.