STP725

    Fatigue Crack Growth Rates of Irradiated Pressure Vessel Steels in Simulated Nuclear Coolant Environment

    Published: Jan 1981


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    Abstract

    The results of the first fatigue crack growth rate (FCGR) tests of irradiated pressure vessel steels in a simulated reactor coolant environment are presented. These results are compared with data on the same and similar unirradiated steels fatigue-tested in high-temperature air, and in high-temperature pressurized reactor-grade water. In the aqueous environment, irradiated condition test results show a slight increase in FCGR over those for unirradiated materials, while both classes of data show a substantial increase above FCGR data for the unirradiated condition in a high-temperature air environment. For unirradiated FCGR data, this increase has been attributed to a hydrogen-assistance model, and this is also suggested to be the case for the irradiated results. Possible interactions of material and environment, irradiation, damage, and hydrogen embrittlement effects are discussed.

    Keywords:

    pressure vessel steels, light water reactors, corrosion fatigue, fatigue, hydrogen embrittlement, irradiation damage


    Author Information:

    Cullen, WH
    Research metallurgist, mechanical engineer, and physical science technician, Naval Research Laboratory, Washington, D.C.

    Watson, HE
    Research metallurgist, mechanical engineer, and physical science technician, Naval Research Laboratory, Washington, D.C.

    Taylor, RE
    Research metallurgist, mechanical engineer, and physical science technician, Naval Research Laboratory, Washington, D.C.

    Torronen, K
    Research metallurgist and group leader, Technical Research Centre of Finland, Espoo,


    Paper ID: STP28209S

    Committee/Subcommittee: E10.08

    DOI: 10.1520/STP28209S


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