STP1049: Fatigue Crack Growth Behavior of Different Stainless Steels in Pressurized Water Reactor Environments

    Amzallag, C
    Research engineer, UNIREC, Centre de Recherches, Firminy,

    Maillard, J-L
    Research engineer, E.C.A.N. INDRET, La Montagne,

    Pages: 32    Published: Jan 1990


    Abstract

    An experimental program has been conducted in order to determine the fatigue crack growth rate (FCGR) curves of different stainless steels for nuclear pressure vessels and pipings in pressurized water reactor environments and to define reference fatigue crack growth rate curves for these materials. The parameters studied are: the steel structure, the load ratio (R), and the water chemistry. A slight, but clear effect of environment was observed for the five steels studied. The paper presents the FCGR curves and the fractographic examinations for the different testing conditions and stresses the differences between FCGR behavior of stainless steels and low-alloy steels.

    Keywords:

    corrosion fatigue, fatigue crack growth, stainless steels, water chemistry, fatigue (materials), cracking, environmental effects


    Paper ID: STP24081S

    Committee/Subcommittee: G01.06

    DOI: 10.1520/STP24081S


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