刘洋,赵亚欢,龚恒风,王占伟,廖业宏,任啟森

• 1:中广核研究院有限公司ATF研发项目部
• 2:核电站高安全性事故容错燃料技术工程实验室

摘要(Abstract)：

采用扫描电镜、X射线衍射仪及万能试验机等研究了微量氧环境下400、800和1200℃不同温度保温300 s对Zr-1%Nb合金力学性能的影响。结果表明:与未热处理的合金相比,400℃热处理后Zr-1%Nb合金的力学性能未发生明显变化;800℃热处理后,合金的塑性降低,压缩时外表面产生大量微裂纹;1200℃热处理后,压缩时合金出现了脆性断裂。热处理后合金样品的表面发生了氧化,氧化产物主要为ZrO_2相。随着热处理温度的升高,Zr-1%Nb合金的氧化程度逐步加深,并且1200℃热处理后试样氧化层的厚度显著增加,内部出现大量裂纹。

关键词(KeyWords)： Zr-1%Nb合金;热处理;塑性;氧化产物

Abstract:

Keywords:

基金项目(Foundation): 国家重点研发计划(2017YFB0702404)

作者(Author): 刘洋,赵亚欢,龚恒风,王占伟,廖业宏,任啟森

DOI: 10.13289/j.issn.1009-6264.2020-0269

参考文献(References)：

• [1] Udagawa Y,Mihara T,sudiyama T,et al.Simulation of the fracture behavior of zircaloy-4 cladding under reactivity-initiated accident conditions with a damage mechanics model combined with fuel performance codes FEMAXI-7 and RANNS[J].Journal of Nuclear Sicence and Technology,2014,51(2):208-219.
• [2] Leclercq S,Parrot A,Leroy M.Failure characteristics of cladding tubes under RIA conditions[J].Nuclear Engineering and Design,2008,238(9):2206-2218.
• [3] 许志安，王喆，杨萍，等.燃料组件边角栅元偏离泡核沸腾比分析[J].核安全，2016,15(2):84-88.XU Zhi-an,WANG Zhe,YANG Ping,et al.DNBR analysis of fuel assembly slide and corner channels[J].Nuclear Safety,2016,15(2):84-88.
• [4] Pain J,Melis J C,Lecomte C.Definition and status of the CABRI internation program for high burn-up fuel studies[C]//Proceedings of the 29th Water Reactor Safety Meeting,Washington,2001.
• [5] Fuketa T,Sugiyama T,Nakamura T,et al.Effects of pellet expansion and cladding hydrides on PCMI failure of high burnup LWR fuel during reactivity transients[C]//Proceedings of the 2003 Nuclear Safety Research Conference,Washington,2003.
• [6] Jun H K,Myoung H L,Byoung K C,et al.Effects of oxide and hydrogen on the circumferential mechanical properties of zircaloy-4 cladding[J].Nuclear Engineering and Design,2006,236:1867-1873.
• [7] 张文，邱长军，曾小安，等.表面镀Cr膜锆合金的抗高温氧化性能研究[J].稀有金属与硬质合金，2017,45(6):71-75.ZHANG Wen,QIU Chang-jun,ZENG Xiao-an,et al.Study on high temperature oxidation resistance of Cr-coated zirconium alloy[J].Rare Metals and Cemented Carbides,2017,45(6):71-75.
• [8] Kaddour D,Frechinet S,Gourgues A F,et al.Experimental determination of creep properties of zirconium alloys together with phase transformation[J].Scripta Materialia,2004,51:515-519.
• [9] Perez R J,Massih A R.Thermodynamic evaluation of the Nb-O-Zr system[J].Journal of Nuclear Materials,2007,360(3):242-254.
• [10] Leistikow S,Schanz G,Berg H V.Comprehensive presentation of extended zircaloy-4 steam oxidation results (600-1600 Deg.C)[C]//Specialists Meeting on Water Reactor Fuel Safety and Fission Product Release in Off-Normal and Accident Conditions,Denmark,1983.
• [11] 谢光善，张汝娴.快中子堆燃料元件[M].北京：化学工业出版社，2007.XIE Guang-shan,ZHANG Ru-xian.Fast Neutron Reactor Fuel Assembly[M].Beijing:Chemical Industry Press,2007.
• [12] Arima T,Miyata K,Inagaki Y,et al.Oxidation properties of Zr-Nb alloys at 500-600 ℃ under low oxygen potentials[J].Corrosion Science,2005,47:435-446.
• [13] Leech W J.Ductility testing of zircaloy-4 and zirlo cladding after high temperature oxidation in steam[C]//Topical Meeting on LOCA Fuel Safety Criteria,Aix-en-Provence,OECD/NEA,2001.
• [14] Westerman R E.High temperature oxidation of zirconium and Zircaloy-2 in oxygen or water vapour[J].Journal of the Electrochemical Society,1964,111(2):140-147.
• [15] Maigin A G,Markelov V A,Novikov V V.High temperature oxidation and residual ductility of fuel claddings from Zr-1Nb alloy having different contents of impurities[C]//16th International Symposium,Zirconium in Nuclear Industry,Chengdu,Sichuan Province,China,2010.
• [16] Michau A,Maury F,Schuster F,et al.High-temperature oxidation resistance of chromium-based coatings deposited by DLI-MOCVD for enhanced protection of the inner surface of long tubes[J].Surface and Coating Technology,2018,349:1048-1057.
• [17] Wei T G,Zhang R Q,Yang H Y,et al.Microstructure,corrosion resistance and oxidation behavior of Cr-coatings on zircaloy-4 prepared by vacuum arc plasma deposition[J].Corrosion Science,2019,158:1-12.

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