You are being redirected because this document is part of your ASTM Compass® subscription.
    This document is part of your ASTM Compass® subscription.

    If you are an ASTM Compass Subscriber and this document is part of your subscription, you can access it for free at ASTM Compass

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

    Published: 01 January 1990

      Format Pages Price  
    PDF (908K) 32 $25   ADD TO CART
    Complete Source PDF (14M) 534 $123   ADD TO CART

    Cite this document

    X Add email address send
      .RIS For RefWorks, EndNote, ProCite, Reference Manager, Zoteo, and many others.   .DOCX For Microsoft Word


    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.


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

    Author Information:

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

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

    Committee/Subcommittee: G01.06

    DOI: 10.1520/STP24081S