SZA-4和ZIRLO锆合金在360℃含氧水环境中的腐蚀行为
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摘要
通过动水循环高压釜回路,考察了国产新锆合金SZA-4(Zr-0.85n-0.25Nb-0.35Fe-0.1Cr-0.05Ge)和商用ZIRLO(Zr-1.0Nb-1.0Sn-0.1Fe)合金在含有约2.0 mg/L溶解氧的360℃/20.0 MPa高温高压水中的早期腐蚀行为,用透射电镜分析了两种合金基体和腐蚀30天后氧化膜的显微组织及成分分布。结果表明,SZA-4合金为完全再结晶晶粒和仅发生回复的等轴晶粒组成的"混晶"组织,主要含有富Nb的Zr(Fe,Cr)_2相及少量的Zr_3Fe相,而ZIRLO合金由均匀分布的短板条晶粒组成,主要以β-Nb和Zr(Nb,Fe)_2相为主。SZA-4合金在DO环境中的腐蚀增重明显低于商用ZIRLO合金,且随着时间的延长,增重差异逐渐增加。SZA-4合金的氧化膜厚度(1.0～1.2μm)明显低于ZIRLO合金(1.3～2.0μm),且含有较少的横向裂纹。SZA-4和ZIRLO合金中的第二相可延迟氧化并"镶嵌"至氧化膜外层等轴晶区,说明未充分氧化或溶解。SZA-4中的Cr能够更好地把Fe"束缚"在Zr(Fe,Cr)_2相中发生原位氧化,而ZIRLO合金中的Fe在Zr(Nb,Fe)_2相初始氧化时即扩散至周围氧化膜中,间接增加了Fe在氧化膜中的浓度。固溶原子Fe和Nb的不同可能是造成两种Zr合金早期腐蚀增重差异的主要原因。
Development of high-performance zirconium alloys with improved corrosion resistant is vital to meet the demands of higher fuel duty, increased cycle length and more aggressive water chemistries, such as potential dissolved oxygen(DO) in some advanced boil water reactor's(ABWR) and small module reactor's(SMR) environment. It is therefore necessary to consider the corrosion behavior of zirconium alloys in DO condition. In the present study, the corrosion behavior of a new-designed SZA-4 alloy(Zr-0.85 n-0.25 Nb-0.35 Fe-0.1 Cr-0.05 Ge, wt.%)(China) and a reference commercial ZIRLO alloy(Zr-1.0 Nb-1.0 Sn-0.1 Fe)(America)was estimated using an autoclave loop in 360 oC/20.0 MPa pure water with approximately 2.0 mg/L DO. TEM was employed to characterize the microstructure of the Zr metal and the oxide film and EDS to give corresponding composition analysis. The results show that, the partial-recrystallized annealing SZA-4 alloy mainly consists of large completely recrystallized grains and small recovered equiaxed grains, with dominant Nb-rich Zr(Fe,Cr)2 phase and a small amount of Zr3 Fe phase. The stress-relieved annealing ZIRLO alloy is composed of uniformly distributed short-lath grains, with dispersed Nb-containing β-Nb and Zr(Nb,Fe)2 phase. The weight gain of SZA-4 alloy is obviously lower than that of ZIRLO alloy, and the discrepancy in weight gain increases with prolonged corroded time, which suggests the SZA-4 alloy has better corrosion resistance. After corroded for 30 days, the oxide of both alloys are consists of loose quiaxed grains in outer layer and dense columnar grains in inner layer, which indicates the corrosion is at its pre-transition stage. The oxide of SZA-4 alloy(1.0-1.2μm) was significantly thinner than the ZIRLO alloy(1.3-2.0μm), and contains less transverse cracks. The presence of transverse cracks may be a result of the rapid release of stress in the oxide film during sample preparation. The second phase precipitates(SPPs) in SZA-4 and ZIRLO alloys was not sufficiently oxidized and retained into the oxide film at a certain distance from the oxide/metal boundary. In SZA-4 alloy, Cr tends to "bind" Fe into Zr(Fe,Cr)2 phase for in-situ oxidation, while in ZIRLO alloy, Fe is readily to diffuse into the surrounding oxide from the Zr(Nb,Fe)2 phase on its onset of oxidation. The different Fe and Nb content in solid solution is more possibly responsible for the difference in corrosion weight gain for the two Zr alloys at the early stage of corrosion.
Development of high-performance zirconium alloys with improved corrosion resistant is vital to meet the demands of higher fuel duty, increased cycle length and more aggressive water chemistries, such as potential dissolved oxygen(DO) in some advanced boil water reactor's(ABWR) and small module reactor's(SMR) environment. It is therefore necessary to consider the corrosion behavior of zirconium alloys in DO condition. In the present study, the corrosion behavior of a new-designed SZA-4 alloy(Zr-0.85 n-0.25 Nb-0.35 Fe-0.1 Cr-0.05 Ge, wt.%)(China) and a reference commercial ZIRLO alloy(Zr-1.0 Nb-1.0 Sn-0.1 Fe)(America)was estimated using an autoclave loop in 360 oC/20.0 MPa pure water with approximately 2.0 mg/L DO. TEM was employed to characterize the microstructure of the Zr metal and the oxide film and EDS to give corresponding composition analysis. The results show that, the partial-recrystallized annealing SZA-4 alloy mainly consists of large completely recrystallized grains and small recovered equiaxed grains, with dominant Nb-rich Zr(Fe,Cr)2 phase and a small amount of Zr3 Fe phase. The stress-relieved annealing ZIRLO alloy is composed of uniformly distributed short-lath grains, with dispersed Nb-containing β-Nb and Zr(Nb,Fe)2 phase. The weight gain of SZA-4 alloy is obviously lower than that of ZIRLO alloy, and the discrepancy in weight gain increases with prolonged corroded time, which suggests the SZA-4 alloy has better corrosion resistance. After corroded for 30 days, the oxide of both alloys are consists of loose quiaxed grains in outer layer and dense columnar grains in inner layer, which indicates the corrosion is at its pre-transition stage. The oxide of SZA-4 alloy(1.0-1.2μm) was significantly thinner than the ZIRLO alloy(1.3-2.0μm), and contains less transverse cracks. The presence of transverse cracks may be a result of the rapid release of stress in the oxide film during sample preparation. The second phase precipitates(SPPs) in SZA-4 and ZIRLO alloys was not sufficiently oxidized and retained into the oxide film at a certain distance from the oxide/metal boundary. In SZA-4 alloy, Cr tends to "bind" Fe into Zr(Fe,Cr)2 phase for in-situ oxidation, while in ZIRLO alloy, Fe is readily to diffuse into the surrounding oxide from the Zr(Nb,Fe)2 phase on its onset of oxidation. The different Fe and Nb content in solid solution is more possibly responsible for the difference in corrosion weight gain for the two Zr alloys at the early stage of corrosion.
引文
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[2]周汇东.水冷动力堆燃料包壳材料--锆合金[M].北京:原子能出版社,1979.ZHOU Huidong.Nuclear fuel cladding for water-cooled reactor:zirconium alloy[M].Beijing:Atomic Energy Press,1979.
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[4]周邦新,改善锆合金耐腐蚀性能的概述[J].金属热处理学报,1997,18(3):8-15.ZHOU Bangxin.The issues of improving corrosion resistance for zirconium alloys[J].Trans.Met.Heat Treatment,1997,18(3):8-15.
[5]杨忠波,赵文金.锆合金耐腐蚀性能及氧化特性概述[J].材料导报,2010,24(9):120-125.YANG Zhongbo,ZHAO Wenjin.Review of corrosion and oxide characterization for Zr alloys[J].Mater.Rev.,2010,24(9):120-125.
[6]MOTTA A T.Waterside corrosion in zirconium alloys[J].JOM,2011,63(8):59-63.
[7]MOTTA A T,COUET A,COMSTOCK R J.Corrosion of zirconium alloys used for nuclear fuel cladding[J].Annu.Rev.Mater.Res.,2015,45:311-343.
[8]YUEH H K,KESTERSON R L,COMSTOCK R J,et al.Improved ZIRLOTM cladding performance through chemistry and process modifications[J].J.ASTM Int.,2005,2:330-346.
[9]BOSSIS P,Pêcheur D,HANIFI K.et al.Comparison of the high burn-up corrosion on M5 and low tin zircaloy-4[J].J.ASTM Int.,2006,3:32.
[10]JEONG Y H,PARK S Y,LEE M H,et al.Out-of-pile and in-pile performance of advanced zirconium alloys(HANA)for high burn-up fuel[J].J.Nuc.Sci.Tech.,2006,43(9):977-983.
[11]ZHOU HANGXIN,YAO MEIYI.Optimization of N18zirconium alloy for fuel cladding of water reactors[J].Journal of Materials Science&Technology,2012,28(7):606-613.
[12]刘信荣,陈志奇,侯忠松.先进沸水堆(ABWR)的特性与可用性[J].山东电力技术,1996,91(5):46-52.LIU Xinrong,CHEN Zhiqi,HOU Zhongsong.The feature of ABWR and its availability[J].Shandong Electric Power,1996,91(5):46-52.
[13]周蓝宇,齐实,周涛.小型模块化反应堆发展趋势及前景[J].科技创新与应用,2017,21:195-196.ZHOU Lanyu,QI Shi,ZHOU Tao.Development and perspective of small module nuclear reactor[J].Tech.Innov.Appl.,2017,21:195-196.
[14]BRADHURST D H,SHIRVINGTON P J,HEUER P M.The effect of radiation and oxygen on the aqueous oxidation of zirconium and its alloys at 290oC[J].J.Nucl.Mater.,1973,46:53-76.
[15]KUMAR M K,AGGARWAL S,BENIWAL D,et al.Localized oxidation of zirconium alloys in high temperature and pressure oxidizing environments of nuclear reactors[J].Mater.Corros.,2014,65(3):244-249.
[16]KUMAR M K,AGGARWAL S,KAIN V,et al.Effect of dissolved oxygen on oxidation and hydrogen pick up behavior:Zircaloy vs Zr-Nb alloys[J].Nucl.Eng.Des.,2010,240:985-994.
[17]COX B.Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys[J].J.Nucl.Mater.,2005,336:331-368.
[18]韦天国,林建康,龙冲生,等.蒸汽中的溶解氧对锆合金腐蚀行为的影响[J].金属学报,2016,52(2):209-216.WEI Tianguo,LIN Jiankang,LONG Chongsheng,et al.Effect of dissolved oxygen in steam on the corrosion behaviors of zirconium alloys[J].Acta Metallur.Sin.,2016,52(2):209-216.
[19]SUN R R,YAO M Y,et al.The effect of dissolved oxygen on the corrosion behavior of Zr-0.85Sn-0.16Nb-0.37Fr-0.18Cr alloy in 500℃/10.3MPa supper-heated stream[C]//Proceedings of the 201725th International Conference on Nuclear Engineering,ICONE25-66486,2017:1-10.
[20]LUNDE L.Special features of external corrosion of fuel cladding in boiling water reactors[J].Nucl.Eng.Des.,1975;33:178-195.
[21]MATSUKAWA Y,KITAYAMA S,MURAKAMI K,et al.Reassessment of oxidation-induced amorphization and dissolution of Nb precipitates in Zr-Nb nuclear fuel cladding tubes[J].Acta Mater.,2017,127:153-146.
[22]TEJLAND P,THUVANDER M,ANDREN H O,et al.Detailed analysis of the microstructure of the metal/oxide interface region in zircaloy-2 after autoclave corrosion testing[M].Zirconium in the Nuclear Industry:16th International Symposium.2011:102956.
[23]HU J,GARNER A,NI N,et al.Identifying suboxide grains at the metal-oxide interface of a corroded Zr-1.0%Nb alloy using(S)TEM,transmission-EBSD and EELS[J].Micron,2015,69:35-42.
[24]YARDLEY S S,MOORE K L,NI N,et al.An investigation of the oxidation behaviour of zirconium alloys using isotopic tracers and high resolution SIMS[J].J.Nucl.Mater.,2013,443:443-436.
[25]NI N,HUDSON D,WEI J,et al.How the crystallography and nanoscale chemistry of the metal/oxide interface develops during the aqueous oxidation of zirconium cladding alloys[J].Acta Mater.,2012,60:7132-7149.
[26]GABORY B D,DONG Y,MOTTA A T,et al.EELS and atom probe tomography study of the evolution of the metal/oxide interface during zirconium alloy oxidation[J].J.Nucl.Mater.,2015,462:304-309.
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[28]ZENG Q F,ZHU L B,YUAN G H,et al.Microstructure and properties of new zirconium alloys for CAP1400 fuel assembly[C]//Proceedings of the 25th International Conference on Nuclear Engineering ICONE25-66951,2017:1-5.
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[30]程竹青,杨忠波,邱军,等.Zr-Sn-Nb-Fe锆合金耐腐蚀性能研究[J].核动力工程,2017,38(5):132-137.CHENG Zhuqing,YANG Zhongbo,QIU Jun et al.Study on corrosion resistance of Zr-Sn-Nb-Fe zirconium alloys[J].Nucl.Power Eng.,2017,38(5):132-137.
[31]FRANKEL P G,WEI J,FRANCIS E M,et al.Effect of Sn on corrosion mechanisms in advanced Zr-Cladding for pressurised water reactors[J].International Symposium on Zirconium in the Nuclear Industry,2015:404-438.
[32]GODLEWSKI J,BOUVIER P,LUCAZEAU G,et al.Stress distribution measured by Raman spectroscopy in zirconia films formed by oxidation of Zr-based alloys[J].Twelfth Int.Symp.Zr Nuclear Ind.,ASTM STP 1354(West Conshohocken,PA:ASTM,2000):877-900.
[33]PREUSS M,FRANKEL P,LOZANO-PEREZ S,et al.Studies regarding corrosion mechanisms in zirconium alloys[J].J.ASTM Inter.,2011,8(9):1-23.
[34]LY A,AMBARD A,BLAT-YRIEIX M,et al.Understanding crack formation at the metal/oxide interface during corrosion of zircaloy-4 using a simple mechanical model[J].J.ASTM Inter.,2011,8(9):1-18.
[35]PARK J Y,CHOI B K,YOO S J,et al.Corrosion and oxide properties of HANA alloys[J].J.ASTM Inter.,2008,5(5):471-485.
[36]LESURF J E,BRYANT P E C.Effect of water chemistry on the oxidation of zirconium alloys under reactor radiation[J].Global Governance,1968,9(3):301-324.
[37]H STEHLE W K R M.External corrosion of cladding in PWRs[J].Nucl.Eng.Des.,1975,33:155-168.
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