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    Fracture Mechanical Testing of In Service Thermally Aged Cast Stainless Steel

    Published: 01 April 2017

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    Embrittlement of duplex stainless steels by thermal aging shortens the service life of structural components in light water reactors (LWRs). This is an important issue when life-extension programs are aiming at 60 to 80 years in service because avoidance of brittle failure is a design prerequisite. Cast and welded austenitic stainless steels, which contain some ferrite, are known to be affected by thermal aging. Historically, many LWR components of complex geometry have been cast in the molybdenum-containing grade, CF8M. Aging is mainly attributed to two types of phase transformations occurring within the minor ferritic phase: (1) demixing of the ferrite by spinodal decomposition into chromium-rich α´ and iron rich α regions and (2) precipitation of G phase, carbides, and other secondary phases. Two in service aged pipe bend castings from the pressurized water reactor (PWR) Ringhals 2 steam generator are tested. These components are large castings of stainless steel quality CF8M. The manufacturing process produces a nonuniform microstructure with coarse ferrite and a high degree of directionality affecting properties as well as the methodology for testing. The materials are exposed to primary circuit PWR water for 72 kh at 291ºC and 325ºC, followed by 22 kh at a reduced service temperature. The testing approach and matrix, consisting of microstructural characterization, mechanical testing, and modeling, is discussed. Results of fracture mechanical evaluation using the J-integral resistance curve technique and instrumented Charpy tests are provided. Effects of large microstructural heterogeneity and anisotropy from the casting and heat-treating processes are seen. The effect on aging embrittlement and fracture mechanisms within each phase as well as phase interaction are also discussed.


    thermal aging, cast austenitic stainless steel (CASS), fracture mechanical testing, spinodal decomposition

    Author Information:

    Bjurman, Martin
    Studsvik Nuclear AB and Royal Institute of Technology, Stockholm,

    Forssgren, Björn
    Ringhals AB, Väröbacka,

    Efsing, Pål
    Ringhals AB and Royal Institute of Technology, Stockholm,

    Committee/Subcommittee: E08.03

    DOI: 10.1520/STP159820160086