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LI Guan-jun, PENG Jun, LI Guang-fu, YANG Wu. Stress Corrosion Cracking and Mechanical Behaviors of Domestic Low Alloy Steel SA-508III in Simulated PWR Primary Water Environment[J]. Corrosion & Protection, 2011, 32(9): 673-676.
Citation: LI Guan-jun, PENG Jun, LI Guang-fu, YANG Wu. Stress Corrosion Cracking and Mechanical Behaviors of Domestic Low Alloy Steel SA-508III in Simulated PWR Primary Water Environment[J]. Corrosion & Protection, 2011, 32(9): 673-676.

Stress Corrosion Cracking and Mechanical Behaviors of Domestic Low Alloy Steel SA-508III in Simulated PWR Primary Water Environment

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  • Received Date: June 06, 2011
  • The stress corrosion cracking (SCC) and mechanical behaviors of homemade low alloy steel SA-508Ⅲ (S: 0.0025%) in simulated pressurized water reactor (PWR) primary water environment at 290 ℃ were studied by slow strain rate test (SSRT) technique. The tests were mainly performed in the water at various applied electrode potentials in the range from -720 mV to +400 mV (SHE) which simulated the electrochemical conditions of the steel in the water environment with different dissolved oxygen and hydrogen contents. A test was also performed in pure nitrogen gas for comparison. Results showed that the susceptibility to SCC of the steel increased with increasing electrode potential. No apparent SCC was found on the specimens tested at the potentials in the range from -720 mV to -200 mV (SHE). Some signs like tiny SCC were observed on the specimens tested at -50 mV and +200 mV (SHE). Significant SCC happened when the potential was raised to +300 mV and +400 mV (SHE). The cracks were nucleated at inclusions and propagated in fan-shaped quasi-cleavage transgranular mode. Results suggest that the steel has excellent resistance to SCC. All the tensile curves of the tests exhibited the character of dynamic strain ageing (DSA). The SCC mechanism and its relation to DSA are discussed.
  • [1]
    Saji G. Degradation of aged plants by corrosion:“Long cell action” in unresolved corrosion issues[J]. Nuclear Engineer and Design, 2009, 239:1591-1613.
    [2]
    Scott P M. A review of environment-sensitive fracture in water reactor materials[J]. Corrosion Science, 1985, 25(5):583-606.
    [3]
    Congleton J, Shoji T, Parkins R N. The stress corrosion cracking of reactor pressure vessel steel in high temperature water[J]. Corrosion Science, 1985, 25(5):633-650.
    [4]
    杨武, 张云柯, 杨鸿根. 压力容器用钢A533B在模拟压水堆高温水中的应力腐蚀破裂敏感性[J]. 腐蚀科学与防护技术, 1992, 4(4): 236-241.
    [5]
    Moshier W C, James L A. The effect of potential on the high-temperature fatigue crack growth response of low alloy steels, part 2:sulfide-potential interaction[J]. Corrosion Science, 1999, 41(4):401-415.
    [6]
    Atkinson J D, Yu J, Chen Z Y. An analysis of the effects of sulphur content and potential on corrosion fatigue crack growth in reactor pressure vessel steels[J]. Corrosion Science, 1996, 38(5):755-765.
    [7]
    Hnninen H, Cullen W, Kemppainen M. Effects of MnS inclusion dissolution on environmentally assisted cracking in low-alloy and carbon steels[J]. Corrosion, 1990, 46(7):563-575.
    [8]
    Ford F P. Prediction of corrosion-fatigue initiation in low-alloy steel and carbon-steel/water systems at 288 ℃[C]//Proc 6th Int Symp on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, San Diego CA:1993.
    [9]
    Wu X Q, Han E H, Ke W. Effects of loading factors on environmental fatigue behavior of low-alloy pressure vessel steels in simulated BWR water[J]. Nuclear Engineering and Design, 2007, 237:1452-1459.
    [10]
    Scott P M, Tice D R. Stress corrosion in low alloy steels[J]. Nuclear Engineering and Design, 1990, 119:399-413.
    [11]
    Ford P. Quantitative prediction of environmentally assisted cracking[J]. Corrosion, 1996, 52(5):375-395.
    [12]
    Peng Q J, Li G F, Shoji T. The crack tip solution chemistry in sensitized stainless steel in simulated boiling water reactor water studied using a micro-sampling technique[J]. Journal of Nuclear Science and Technology, 2003, 40(6):397-404.

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