STP513: Fatigue Crack Growth Analysis of Pressurized Water Reactor Vessels

    Riccardella, PC
    Westinghouse Electric Co., Pittsburgh, Pa.

    Mager, TR
    Westinghouse Electric Co., Pittsburgh, Pa.

    Pages: 20    Published: Jan 1972


    Abstract

    The concepts of linear elastic fracture mechanics are applied to perform a fatigue evaluation of a pressurized water reactor vessel. The results of this work indicate that, for the types of flaws which could exist in a reactor vessel, fatigue crack growth due to normal operating conditions during the design life of the vessel is negligible. We have previously presented a simplified technique to approximate the potential for fatigue crack growth in the beltline region of a typical pressurized water reactor vessel during the forty-year design life of the vessel. The beltline region is the most critical from the standpoint of irradiation embrittlement of the material, and thus was the first to be considered. However, there are several areas in a reactor vessel which are more critically stressed during operation than the beltline region, and the crack growth in these regions must be considered before any general conclusions are drawn concerning the vessel.

    In this paper, the following four regions of the reactor vessel are considered separately: (1) the closure head, (2) the nozzle shell course, (3) the beltline region, and (4) the lower head. A fatigue crack growth analysis is performed on the most critically stressed location in each region. Initial flaw sizes are assumed which are conservative upper bounds of the types of flaws which could exist in the hardware. Crack tip stress intensity factors are estimated using continuous surface flaw expressions, after linearizing the thermal and pressure stress distributions through the thickness of the vessel. An estimate of the degree of conservatism introduced by these two major approximations is included. Fatigue crack growth data from various sources were used. The sensitivity of the data to such variables as temperature, irradiation, and biaxiality is discussed.

    Keywords:

    fractures (materials), fatigue (materials), failure, stresses, nuclear reactors, pressurized water reactors, cyclic loads, crack propagation, stress analysis


    Paper ID: STP34124S

    Committee/Subcommittee: E08.06

    DOI: 10.1520/STP34124S


    CrossRef ASTM International is a member of CrossRef.