堆内构件用不锈钢应力腐蚀开裂的影响因素
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摘要
采用高温高压水慢应变速率拉伸测试(SSRT)及断口形貌观察研究了pH值和辐照对国产核反应堆堆内构件用不锈钢的应力腐蚀开裂(SCC)行为的影响。结果表明,不锈钢的SCC敏感性在pH值为7.0溶液中较小,在pH值为6.4与7.5水溶液条件下SCC敏感性显著增加。带电粒子辐照后出现辐照加速SCC(IASCC)现象,主要是由于辐照缺陷与局域形变对裂纹起裂的影响导致的,但由于离子辐照损伤深度的限制,不能从SSRT试样断口形貌观察到离子辐照对SCC的影响。
The effects of pH value,strain rate and irradiation on the stress corrosion cracking(SCC)in the stainless steel for reactor internals were studied by slow strain rate tension(SSRT)tests in high temperature and high press water environment and the observations of fracture morphology.The results indicated that the SCC susceptibility of stainless steel was tiny at pH=7.0but remarkably increased at pH=6.4and 7.5.Due to the effect of irradiation defects and localized deformation on the initiation of crack,the irradiation by charged particles caused the irradiation assisted SCC in stainless steel.But the influence of the ion irradiation on SCC cannot be observed from the SSRT fracture morphology because of the low irradiation penetration depth.
The effects of pH value,strain rate and irradiation on the stress corrosion cracking(SCC)in the stainless steel for reactor internals were studied by slow strain rate tension(SSRT)tests in high temperature and high press water environment and the observations of fracture morphology.The results indicated that the SCC susceptibility of stainless steel was tiny at pH=7.0but remarkably increased at pH=6.4and 7.5.Due to the effect of irradiation defects and localized deformation on the initiation of crack,the irradiation by charged particles caused the irradiation assisted SCC in stainless steel.But the influence of the ion irradiation on SCC cannot be observed from the SSRT fracture morphology because of the low irradiation penetration depth.
引文
[1]Gary S Was.Fundamentals of Radiation Materials ScienceMetals and Alloys[M].Berlin:Springer,2007.
[2]IAEA.Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety:Pwr Vessel Internals[R].Vienna:IAEA,2007.
[3]杨武.辐照促进应力腐蚀破裂研究的进展[J].材料保护,1994(27):1.
[4]李红梅,杨武,蔡珣,等.304不锈钢在含硼和锂的高温水中的应力腐蚀破裂和断口分析[J].中国腐蚀与防护学报,2004,24(1):16.
[5]Biersack J P,Haggmark L G.A Monte carlo computer program for the transport of energetic ions in amorphous targets[J].Nucl Instrum Methods,1980(174):257.
[6]ASTM International.Standard Practice for Neutron Radiation Damage Simulation by Charged-Particle Irradiation[S].West Conshohocken:ASTM 521-96,2003.
[7]Pathania R.The Use of Proton Irradiation to Determine IASCC Mechanisms in Light Water Reactors-Phase 3:Deformation Studies[R].USA:EPRI,2006.
[8]董超芳,关矞心,程学群,等.pH值对高温高压水中304L不锈钢应力腐蚀开裂的影响[J].北京科技大学学报,2010,32(12):1569.
[9]Congleton J,Shoji T,Parkins R N.The SCC of reactor pressure vessel in high temperature water[J].Corros Sci,1985,25(8/9):633.
[10]Pathania R S,Nelson J L.The Use of Proton Irradiation to Understand IASCC in LWR Cores[R].USA:EPRI,2001.
[11]徐超亮,王荣山,黄平,等.不锈钢中子辐照加速应力腐蚀开裂的带电粒子辐照模拟[J].材料导报,2012(26):150.
[12]Cookson J M,Was G S,Andresen P L.Oxide-induced initiation of stress corrosion cracking in irradiated stainless steel[J].Corrosion,1998,54(4):299.
[13]Baker M A,Castle J E.The initiation of pitting corrosion of stainless steels at oxide inclusions[J].Corros Sci,1992(33):1295.
[14]Kurt E Sickafus,Eugene A Kotomin,Blas P Uberuaga.Radiation Effects in Solids[M].Italy:Springer,2004.
[15]WANG Rong-shan,XU Chao-liang,LIU Xiang-bing,et al.The studies of irradiation assisted stress corrosion cracking on reactor internals stainless steel under Xe irradiation[J].Journal of Nuclear Materials,2015(457):130.
[16]Ann Arbor.The Use of Proton Irradiation to Determine IASCC Mechanisms in Light Water Reactors-Phase 3:Deformation Studies[R].USA:EPRI,2006.
[17]Fyfitch S,Xu H,Moore K,et al.Materials Reliability Program:PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values(MRP-175)[R].USA:EPRI,2005.
[18]Faza Sefta.Proposed Revision of the IASCC Sensitivity Criterion[R].France:Materials Ageing Institute,2015.
[2]IAEA.Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety:Pwr Vessel Internals[R].Vienna:IAEA,2007.
[3]杨武.辐照促进应力腐蚀破裂研究的进展[J].材料保护,1994(27):1.
[4]李红梅,杨武,蔡珣,等.304不锈钢在含硼和锂的高温水中的应力腐蚀破裂和断口分析[J].中国腐蚀与防护学报,2004,24(1):16.
[5]Biersack J P,Haggmark L G.A Monte carlo computer program for the transport of energetic ions in amorphous targets[J].Nucl Instrum Methods,1980(174):257.
[6]ASTM International.Standard Practice for Neutron Radiation Damage Simulation by Charged-Particle Irradiation[S].West Conshohocken:ASTM 521-96,2003.
[7]Pathania R.The Use of Proton Irradiation to Determine IASCC Mechanisms in Light Water Reactors-Phase 3:Deformation Studies[R].USA:EPRI,2006.
[8]董超芳,关矞心,程学群,等.pH值对高温高压水中304L不锈钢应力腐蚀开裂的影响[J].北京科技大学学报,2010,32(12):1569.
[9]Congleton J,Shoji T,Parkins R N.The SCC of reactor pressure vessel in high temperature water[J].Corros Sci,1985,25(8/9):633.
[10]Pathania R S,Nelson J L.The Use of Proton Irradiation to Understand IASCC in LWR Cores[R].USA:EPRI,2001.
[11]徐超亮,王荣山,黄平,等.不锈钢中子辐照加速应力腐蚀开裂的带电粒子辐照模拟[J].材料导报,2012(26):150.
[12]Cookson J M,Was G S,Andresen P L.Oxide-induced initiation of stress corrosion cracking in irradiated stainless steel[J].Corrosion,1998,54(4):299.
[13]Baker M A,Castle J E.The initiation of pitting corrosion of stainless steels at oxide inclusions[J].Corros Sci,1992(33):1295.
[14]Kurt E Sickafus,Eugene A Kotomin,Blas P Uberuaga.Radiation Effects in Solids[M].Italy:Springer,2004.
[15]WANG Rong-shan,XU Chao-liang,LIU Xiang-bing,et al.The studies of irradiation assisted stress corrosion cracking on reactor internals stainless steel under Xe irradiation[J].Journal of Nuclear Materials,2015(457):130.
[16]Ann Arbor.The Use of Proton Irradiation to Determine IASCC Mechanisms in Light Water Reactors-Phase 3:Deformation Studies[R].USA:EPRI,2006.
[17]Fyfitch S,Xu H,Moore K,et al.Materials Reliability Program:PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values(MRP-175)[R].USA:EPRI,2005.
[18]Faza Sefta.Proposed Revision of the IASCC Sensitivity Criterion[R].France:Materials Ageing Institute,2015.