高温水中溶解氧、氯离子和处理方式对316L不锈钢应力腐蚀性能的影响
Effects of Dissolved Oxygen,Chloride and Treatment on Stress Corrosion Behavior of 316L SS in High Temperature Pure Water
摘 要
采用慢应变速率试验(SSRT),在模拟压水堆核电厂高温高压水环境(325 ℃、15.5 MPa)中,研究了溶解氧(DO)、氯离子和处理方式对316L不锈钢(SS)的应力腐蚀开裂(SCC)倾向的影响规律,并观察了试样断口形貌.结果表明,冷变形和固溶处理的316L不锈钢分别在含0 μg/kg DO+0 μg/kg Cl-,200 μg/kg DO+0 μg/kg Cl-,0 μg/kg DO+30 μg/kg Cl-及200 μg/kg DO+30 μg/kg Cl-的高温水环境中都呈现出了不同程度的SCC倾向,且200 μg/kg DO和30 μg/kg Cl-使SCC倾向更为明显.同时,冷变形可大大提高316L不锈钢的屈服强度,但是也提高了SCC敏感性.
高温高压水
SCC
SSRT
316L SS
high temperature and high pressure water
SCC
SSRT
Abstract
The effects of DO,Cl- and treatment on the SCC behavior of 316L SS were srudied by SSRT in high temperature (325 ℃) and high pressure (15.5 MPa) environments and the fracture morphology of specimens after testing was observed carefully.The results showed that both cold worked and solution treated 316L SS had SCC susceptibility in high temperature water environments containing 0 μg/kg DO+0 μg/kg Cl-,200 μg/kg DO+0 μg/kg Cl-,0 μg/kg DO+30 μg/kg Cl- or 200 μg/kg DO+30 μg/kg Cl-.And the effect of 200 μg/kg DO and 30 μg/kg Cl- on SCC susceptibility was very remarkable.Although cold work could enhance the yield strength of 316L SS,it increased the SCC susceptibility of 316L SS.
中图分类号 TG174 TL341 DOI 10.11973/fsyfh-201507005
所属栏目 试验研究
基金项目 压水堆核电材料环境相容性研究重大专项(2011ZX0600400908)
收稿日期 2014/7/14
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备注张乐福(1967-),副教授,博士,从事核电材料研究.
引用该论文: LU Hui,SHEN Zhao,DU Dong-hai,ZHANG Le-fu. Effects of Dissolved Oxygen,Chloride and Treatment on Stress Corrosion Behavior of 316L SS in High Temperature Pure Water[J]. Corrosion & Protection, 2015, 36(7): 620
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参考文献
【1】赵彦芬,文新,汪小龙,等.核电站用钢管材料及其国产化[J].钢管,2007,36(2):11-14.
【2】MANFREDI C,OTEGUI J L.Failures by SCC in buried pipelines[J].Engineering Failure Analysis,2002,9(5):495-509.
【3】LJUNGBERG L G.SCC testing of pipe materials in BWR environment[J].Nuclear Engineering and Design,1984,81(1):121-125.
【4】YANG G,YOON K B,MOON Y C.Stress corrosion cracking of stainless steel pipes for Methyl-Methacrylate process plants[J].Engineering Failure Analysis,2013,29:45-55.
【5】KANG S S,HWANG S S,KIM H P,et al.The experience and analysis of vent pipe PWSCC (primary water stress corrosion cracking) in PWR vessel head penetration[J].Nuclear Engineering and Design,2014,269:291-298.
【6】SHOJI T.Progress in the mechanistic understanding of BWR SCC and its implication to the prediction of SCC growth behavior in plants[C]//La Grange Park IL:American Nuclear Society,Proc.11th Int.Symp.on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors,[S.l]:[s.n],2003.
【7】SUZUKI S,KUMAGAI K,SHITARA C,et al.Damage evaluation of SCC in primary loop recirculation system pipe[J].Maintenology,2004,3(2):65-70.
【8】FORD F P,POVICH M J.The effect of oxygen temperature combinations on the stress corrosion susceptibility of sensitized type 304 stainless steel in high purity water[J].Corrosion,1979,35(12):569-574.
【9】JONES R L,GORDON G M,NEILS G H.Environmental degradation of materials in boiling water reactors[C]//New Port Beach:Proc.Fourth Int.Symp.Environ.Degradation of Materials in Nuclear Power Systems-Water Reactors,1989:1-10.
【10】ANDRESEN P L.Perspective and direction of stress corrosion cracking in hot water[C]//NACE Proc 10th Int.Symp.on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors,Houston:[s.n.],2001.
【11】KANE,RUSSELL D.Slow strain rate testing for the evaluation of environmentally induced cracking:research and engineering applications[M].Astm International,1993.
【12】BERTINI L,SANTUS C,VALENTINI R,et al.New high concentration-high temperature hydrogenation method for slow strain rate tensile tests[J].Materials Letters,2007,61(11):2509-2513.
【13】MTHIS K,PRCHAL D,NOVOTNYR,et al.Acoustic emission monitoring of slow strain rate tensile tests of 304L stainless steel in supercritical water environment[J].Corrosion Science,2011,53(1):59-63.
【14】LONG B,DAI Y,BALUC N.Investigation of liquid LBE embrittlement effects on irradiated ferritic/martensitic steels by slow-strain-rate tensile tests[J].Journal of Nuclear Materials,2012,431(1):85-90.
【15】李红梅,蔡珣,吕战鹏,等.304不锈钢在含硼和锂的高温水中的应力腐蚀破裂和表层氧化膜分析[J].材料工程,2004(4):7-10.
【16】ANDRESEN P L,FORD F P.Life prediction by mechanistic modeling and system monitoring of environmental cracking of Fe and Ni alloys in aqueous systems[J].Materials Science and Engineering,1988,103:167-183.
【17】ANDRESEN P L.Critical processes to model in predicting SCC response in hot water[C]//Corrosion/05,Housten:NACE,2005:Paper 05470.
【2】MANFREDI C,OTEGUI J L.Failures by SCC in buried pipelines[J].Engineering Failure Analysis,2002,9(5):495-509.
【3】LJUNGBERG L G.SCC testing of pipe materials in BWR environment[J].Nuclear Engineering and Design,1984,81(1):121-125.
【4】YANG G,YOON K B,MOON Y C.Stress corrosion cracking of stainless steel pipes for Methyl-Methacrylate process plants[J].Engineering Failure Analysis,2013,29:45-55.
【5】KANG S S,HWANG S S,KIM H P,et al.The experience and analysis of vent pipe PWSCC (primary water stress corrosion cracking) in PWR vessel head penetration[J].Nuclear Engineering and Design,2014,269:291-298.
【6】SHOJI T.Progress in the mechanistic understanding of BWR SCC and its implication to the prediction of SCC growth behavior in plants[C]//La Grange Park IL:American Nuclear Society,Proc.11th Int.Symp.on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors,[S.l]:[s.n],2003.
【7】SUZUKI S,KUMAGAI K,SHITARA C,et al.Damage evaluation of SCC in primary loop recirculation system pipe[J].Maintenology,2004,3(2):65-70.
【8】FORD F P,POVICH M J.The effect of oxygen temperature combinations on the stress corrosion susceptibility of sensitized type 304 stainless steel in high purity water[J].Corrosion,1979,35(12):569-574.
【9】JONES R L,GORDON G M,NEILS G H.Environmental degradation of materials in boiling water reactors[C]//New Port Beach:Proc.Fourth Int.Symp.Environ.Degradation of Materials in Nuclear Power Systems-Water Reactors,1989:1-10.
【10】ANDRESEN P L.Perspective and direction of stress corrosion cracking in hot water[C]//NACE Proc 10th Int.Symp.on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors,Houston:[s.n.],2001.
【11】KANE,RUSSELL D.Slow strain rate testing for the evaluation of environmentally induced cracking:research and engineering applications[M].Astm International,1993.
【12】BERTINI L,SANTUS C,VALENTINI R,et al.New high concentration-high temperature hydrogenation method for slow strain rate tensile tests[J].Materials Letters,2007,61(11):2509-2513.
【13】MTHIS K,PRCHAL D,NOVOTNYR,et al.Acoustic emission monitoring of slow strain rate tensile tests of 304L stainless steel in supercritical water environment[J].Corrosion Science,2011,53(1):59-63.
【14】LONG B,DAI Y,BALUC N.Investigation of liquid LBE embrittlement effects on irradiated ferritic/martensitic steels by slow-strain-rate tensile tests[J].Journal of Nuclear Materials,2012,431(1):85-90.
【15】李红梅,蔡珣,吕战鹏,等.304不锈钢在含硼和锂的高温水中的应力腐蚀破裂和表层氧化膜分析[J].材料工程,2004(4):7-10.
【16】ANDRESEN P L,FORD F P.Life prediction by mechanistic modeling and system monitoring of environmental cracking of Fe and Ni alloys in aqueous systems[J].Materials Science and Engineering,1988,103:167-183.
【17】ANDRESEN P L.Critical processes to model in predicting SCC response in hot water[C]//Corrosion/05,Housten:NACE,2005:Paper 05470.
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