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    SUN Yao, ZHANG Le-fu, LI Li. Stress Corrosion Behavior of Candidate Materials for Supercritical Water-cooled Reactor[J]. Corrosion & Protection, 2013, 34(12): 1053-1057.
    Citation: SUN Yao, ZHANG Le-fu, LI Li. Stress Corrosion Behavior of Candidate Materials for Supercritical Water-cooled Reactor[J]. Corrosion & Protection, 2013, 34(12): 1053-1057.
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    Stress Corrosion Behavior of Candidate Materials for Supercritical Water-cooled Reactor

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    • Stress corrosion cracking(SCC) behaviors of alloy 800H, alloy 825 and stainless steel HR3C in supercritical water at temperature of 550 and 650 ℃ and pressure of 25 MPa were studied by slow strain rate testing(SSRT). The results show that alloy 825 had the highest tensile strength and elongation. Fractography indicates that the failure modes of alloy 800H and alloy 825 at 550 ℃ and 650 ℃, and stainless steel HR3C at 550 ℃ were combination of ductile and brittle fracture. HR3C showed fully ductile fracture at 650 ℃.
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