M5锆合金焊接封头组织结构与腐蚀性能研究
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摘要
M5锆合金以其优越力学性能和耐腐蚀性能在核电站包壳管材料中得到了广泛应用。然而随着反应堆技术向高燃耗、长换料周期方向的发展,对现有的包壳材料提出了更高的要求。本研究针对在M5锆合金焊接封头中含有板条马氏体组织以及马氏体组织在钢铁材料中的优异表现,开展对M5锆合金马氏体化的相关研究工作,以期通过改变M5锆合金的组织结构而达到提高其力学性能和耐腐蚀性能的目的,进而满足核反应堆技术的发展需要。
本文采用金相显微镜、透射电镜、维氏硬度试验、动态腐蚀试验和能谱分析等测试手段分别对M5锆合金焊接封头的组织结构和力学性能、Zr-4合金和M5锆合金的腐蚀动力学和耐腐蚀性能以及腐蚀产物及形貌进行了比较系统的研究,主要研究结果如下：
M5锆合金包壳管焊接封头金相和透射观察表明：在室温下M5锆合金母材的显微组织为αzr+βNb,,β相较均匀地分布于α相内；焊缝中心部位组织主要为晶界a-Zr (Nb)+棒针状马氏体,透射电镜下可见清晰而规则的板条α'马氏体束及板条束内大量位错分布；热影响区基本为a-Zr (Nb)和板条α'马氏体两相共存或a-Zr、βNb及板条α'马氏体三相共存的组织状态。α'马氏体板条束的宽度由焊缝中心至过渡区逐渐变小。由同心圆截点法对M5锆合金包壳管焊接封头的晶粒度研究结果显示,焊缝中心区域的晶粒度高达六级到七级,熔合线附近的晶粒度为八级到九级,热影响的晶粒度达到九级到十级,而母材晶粒相当细小,晶粒度仅为十三级到十四级。
M5锆合金包壳管焊接封头显微硬度试验结果表明,M5锆合金焊接封头焊缝中心硬度值高于热影响区和母材,由焊缝—热影响区—母材顺序依次呈抛物线降低。纵向显微硬度因为电子束的堆焊,使得峰值出现在热影响区。这和焊缝中心板条α'相马氏体及期内的大量位错分布密切相关。
对Zr-4合金和M5锆合金的380℃和400℃动态腐蚀试验结果表明,两种合金在不同温度下的腐蚀动力学曲线均符合抛物线规律,但M5锆合金的腐蚀速率要明显低于Zr-4合金。扫描电镜形貌观察显示,腐蚀后两种合金的表面颜色变成黯灰色,但仍具有金属光泽。其中Zr-4试样表面出现灰白色块状氧化膜,出现了轻微的疖状腐蚀,而M5表面则未发现疖状腐蚀。同时,Zr-4腐蚀层与包壳管基体金属的结合不如M5牢固。能谱分析则表明,两种合金的腐蚀产物均为Zr02和ZrO的固溶物。
As the cladding material of light-water nuclear reactor, M5 zirconium alloy is widely used in industrial applications with its excellent mechanical properties and corrosion resistance,. However, with the development of the nuclear reactor technology of high fuel consumption and long period of changing the cladding material, there is a new challenge for the exsiting cladding material. In view of the good performance of the martensite in steel, The comprehensively study on the martensite in M5 alloy welding joints to meet the needs of the development of the nucear power plant.
The microstrucuture, mechanical properties, corrosion dynamics, corrosion resistance and corrosion products of Zr-4 and M5 alloy were mainly investigated by using of OM, TEM, SEM, Vickers hardness test, dynamic corrosion test and energy spectrum analysis. The results were as follows.
The metallograph and microhardness test of M5 alloy cladding tube showed that the structure of base metal was azr+βNb with fineβNb homogeneously distributed in azr matrix. The soldered area consists of grain boundary a-Zr (Nb) and acicular martensite with high density of dislocations under TEM; The microstructure of HAZ presents the coexisting structures of a and a'duplex phases system, or the three-phase coexisting strucuture of a',α'andβ. The width of martensite became narrower from WM to HAZ. The Grain size of Center of the weld was 6-7, the fusion zone is 8-9 and the size of base metal was only 13-14.
Microhardness test indicated that the grain size of melted zone on the welding plug head was higher than that of other positions. For the different distributions of dislocations the highest microhardness value was on HAZ.
The dynamic corrosion test at 380℃and 400℃revealed that the corrosion dynamics of Zr-4 and M5 alloys abided by parabolic law and the corrosion speed of M5 alloys was obviously lower than that of Zr-4 alloys. The surface of the two alloys changed from bright to gray. There were some tiny gray oxidation film on the surface of Zr-4 rather than on M5 alloys, which showed that corrosion resistance of M5 was superior to that of Zr-4. The spectrum analysis showed the oxide products of M5 were probably ZrO2 and ZrO.
本文采用金相显微镜、透射电镜、维氏硬度试验、动态腐蚀试验和能谱分析等测试手段分别对M5锆合金焊接封头的组织结构和力学性能、Zr-4合金和M5锆合金的腐蚀动力学和耐腐蚀性能以及腐蚀产物及形貌进行了比较系统的研究,主要研究结果如下：
M5锆合金包壳管焊接封头金相和透射观察表明：在室温下M5锆合金母材的显微组织为αzr+βNb,,β相较均匀地分布于α相内；焊缝中心部位组织主要为晶界a-Zr (Nb)+棒针状马氏体,透射电镜下可见清晰而规则的板条α'马氏体束及板条束内大量位错分布；热影响区基本为a-Zr (Nb)和板条α'马氏体两相共存或a-Zr、βNb及板条α'马氏体三相共存的组织状态。α'马氏体板条束的宽度由焊缝中心至过渡区逐渐变小。由同心圆截点法对M5锆合金包壳管焊接封头的晶粒度研究结果显示,焊缝中心区域的晶粒度高达六级到七级,熔合线附近的晶粒度为八级到九级,热影响的晶粒度达到九级到十级,而母材晶粒相当细小,晶粒度仅为十三级到十四级。
M5锆合金包壳管焊接封头显微硬度试验结果表明,M5锆合金焊接封头焊缝中心硬度值高于热影响区和母材,由焊缝—热影响区—母材顺序依次呈抛物线降低。纵向显微硬度因为电子束的堆焊,使得峰值出现在热影响区。这和焊缝中心板条α'相马氏体及期内的大量位错分布密切相关。
对Zr-4合金和M5锆合金的380℃和400℃动态腐蚀试验结果表明,两种合金在不同温度下的腐蚀动力学曲线均符合抛物线规律,但M5锆合金的腐蚀速率要明显低于Zr-4合金。扫描电镜形貌观察显示,腐蚀后两种合金的表面颜色变成黯灰色,但仍具有金属光泽。其中Zr-4试样表面出现灰白色块状氧化膜,出现了轻微的疖状腐蚀,而M5表面则未发现疖状腐蚀。同时,Zr-4腐蚀层与包壳管基体金属的结合不如M5牢固。能谱分析则表明,两种合金的腐蚀产物均为Zr02和ZrO的固溶物。
As the cladding material of light-water nuclear reactor, M5 zirconium alloy is widely used in industrial applications with its excellent mechanical properties and corrosion resistance,. However, with the development of the nuclear reactor technology of high fuel consumption and long period of changing the cladding material, there is a new challenge for the exsiting cladding material. In view of the good performance of the martensite in steel, The comprehensively study on the martensite in M5 alloy welding joints to meet the needs of the development of the nucear power plant.
The microstrucuture, mechanical properties, corrosion dynamics, corrosion resistance and corrosion products of Zr-4 and M5 alloy were mainly investigated by using of OM, TEM, SEM, Vickers hardness test, dynamic corrosion test and energy spectrum analysis. The results were as follows.
The metallograph and microhardness test of M5 alloy cladding tube showed that the structure of base metal was azr+βNb with fineβNb homogeneously distributed in azr matrix. The soldered area consists of grain boundary a-Zr (Nb) and acicular martensite with high density of dislocations under TEM; The microstructure of HAZ presents the coexisting structures of a and a'duplex phases system, or the three-phase coexisting strucuture of a',α'andβ. The width of martensite became narrower from WM to HAZ. The Grain size of Center of the weld was 6-7, the fusion zone is 8-9 and the size of base metal was only 13-14.
Microhardness test indicated that the grain size of melted zone on the welding plug head was higher than that of other positions. For the different distributions of dislocations the highest microhardness value was on HAZ.
The dynamic corrosion test at 380℃and 400℃revealed that the corrosion dynamics of Zr-4 and M5 alloys abided by parabolic law and the corrosion speed of M5 alloys was obviously lower than that of Zr-4 alloys. The surface of the two alloys changed from bright to gray. There were some tiny gray oxidation film on the surface of Zr-4 rather than on M5 alloys, which showed that corrosion resistance of M5 was superior to that of Zr-4. The spectrum analysis showed the oxide products of M5 were probably ZrO2 and ZrO.
引文
[1]邹树梁.世界核电发展的历史、现状与新趋势[J].华南大学学报,2005,6(6)：38-39.
[2]孙家宁,孙剑彪.大亚湾核电站有关知识简介[J].物理教师,1996,6：36-37.
[3]张禄庆.压水堆技术发展趋势概述[J].中国核电,2008,1(1)：40-42.
[4]陈济东.大亚湾核电站系统及运行[M].北京：原子能出版社,1993.
[5]李自淳,宣自平.核电简介[J].上海大中型电机,2006,3：1-2.
[6]邹子和.核电站用管[J].钢管,1994,2(7).23-27.
[7]束国刚,陆念文.压水堆核电厂金属部件老化和寿命评估[J].我国电力,2006,39(5)：53-58.
[8]陈鹤鸣,马春来,白新德.核反应堆材料腐蚀及其防护[M].北京：原子能出版社,1984.
[9]杨文斗.反应堆材料学[M].北京：原子能出版社,2006.
[10]韩恩厚.超临界水环境中材料腐蚀研究现状[J].腐蚀科学和防护技术,1999,11(1)：53-57.
[11]刘建章.核结构材料[M].北京：化学工业出版社,2007.
[12]阎昌琪.核反应堆工程[M].哈尔滨：哈尔滨工程大学出版社,2004.
[13]吴玖.核电材料简介[M].北京：冶金工业出版社,1999.
[14]郑杰,牛惠玲.Zr-4合金工艺改进的研究[J].稀有金属,1995,19(4)：260-263.
[15]向德光,潘旦光,龙纯育,等.锆合金经历组织明场观察及测量的进项制备技术[G].国产Zr-4合金性能研究,1994,27(5)：156-159.
[16]周邦新,郑斯奎,汪顺新.真空电子束焊接对Zr-2合金熔区中成分、组织及腐蚀性能的影响[J].核科学与工程,1988,8(2)：130-137.
[17]周军,李中奎,刘荣臻.氢含量对新锆合金室温循环变形行为的影响[J].稀有金属,2008,37(5)：563-566.
[18]苟渊,李言荣,应诗浩.氢及氢化物对锆合金疲劳裂纹扩展速率的影响及估算[J].核动力工程,2005,23(4)：73-76.
[19].李献军.锆及锆合金概述[J].钛工业进展,2011,2(1)：28.
[20]周邦新,赵文金,蒋有荣,等.Zr-4合金元素表面偏聚[J].核动力工程,1998,19(6)：506-508.
[21]赵文金,苗志,蒋宏曼,等Zr-Sn-Nb合金的腐蚀行为研究[J].中国腐蚀与防护学报,2002,22(2)：124-128.
[22]李佩志,李中奎.合金元素对Zr-Nb合金耐蚀性能影响[J].稀有金属材料与工程,1998,27(6)：356-359.
[23]李中奎,刘建章.新锆合金在两种不同介质中的耐蚀性能[J].稀有金属材料与工程,1999,28(2)：101-104.
[24]朱梅生,刘建章.锆合金组织与耐蚀性能研究[J].稀有金属材料与工程,1996,25(4)：34-38.
[25]邝用庚Zr-Nb系合金耐蚀性能研究[J].稀有金属材料与工程,1995,24(3)：41-45.
[26]刘文庆.合金元素及水化学对锆合金耐腐蚀性能影响的研究[D].上海：上海大学,2002.
[27]李中奎,刘建章.合金元素对Zr-Sn-Fe-Cr-Nb合金性能的影响[J].稀有金属材料与工程,1996,25(5)：43-48.
[28]姚美意.合金成分及热处理对锆合金腐蚀和吸氢行为影响的研究[D].上海：上海大学,2007.
[29]袁改焕,白新德,刘振球.我国锆铪材加工技术的进步及发展[J].稀有金属快报,2004,2(3)：1-4.
[30]刘承新.锆合金在核工业中的应用现状及发展前景[J].稀有金属快报,2004,23(5)：21-23
[31]Grad G B, Pieres J J, Guillermet A F, et al. Lattice Parameter of the Zr-Nb Bcc Phase: Neutron Scattering Study and Assessment of Experimental Data[J]. Z. Met-allkd.1995, 86(6):395-400.
[32]Choo K N, Kang Y H, Pyum S I, et al. Effect of Composition and Heat Treatment on the Microstructure and Corrosion Behavior of Zr-Nb Alloys[J]. J Nucl Mat.,1994,20(9): 226-235.
[33]Lin Y P, Woo O T. Oxidation of β-Zr and Related Phases in Zr-Nb Alloy[J]. J Nucl Mat. 2000,27(7):11-27.
[34]刘文庆,李强,周邦新.锆锡合金腐蚀转折机理的讨论[J].稀有金属材料与工程,2001,30(2)：8-9.
[35]王文生.低铌新锆合金的抗蠕变性能[J].原子能科学与技术,2007,39(3)：14-17.
[36]陈桂辉.轻水堆核电站的原理与应用前景[J].福建建能源开发与节约,1996,27(1)：37-39.
[37]梁建烈,庄应烘,赵文金,等,新锆合金的相变温度[J].广西科学,2002,9(3)：181-185
[38]Yang Jung, M yung Ho Lee. Development of a Manufactuing Process for Zr Based Spacer Grid Materials [J]. J Nuclear Mater,2009,39(2):482-484.
[39]Yong Hwan Jeong. Effect of Phase, Precipitate and Nb Concentration in Matrix on Corrosion and Oxide Characteristics of Zr-Nb alloys[J].J Nuclear Mater,2003,31(7):1-5.
[40]Hyun Gil Kim, Yong Hawn Jeong. Effect of Isothermal Annealing on the Corrosion Behavior of Zr-xNb alloys[J]. J Nuclear Mater,2004,32(6):125-129.
[41]Caroline Toffolon Masclet.Study of Nb and Fe Precipitation in a Phase Temperature Range(400-550℃) in Zr Nb(Fe Sn)alloys[J].J A ST M Int,2005,2(5):81-83
[42]Hyun Gil Kim. Microstructure and Corrosion Characteristics of Zr-1.5Nb-0.4Sn-0.2Fe-0.1 Cr Alloy with an Nealing[J].J Nuclear Mater,2008,372(25):304-307.
[43]Jeong Yong Park, Seung Jo Yoo. Oxide Micro Structures of Advanced Zr Alloys Corroded in 360℃ Water Loop[J].J Alloys Compd,2007,43(7):274-275.
[44]Aylin Yilmazbay han.Structure of Zirconium Alloy Oxides for Med in Pure Water Studied with Synchro Tronradiation and Optical Microscopy[J].J Nuclear Mater,2004,32(4):6-11.
[45]Hyun Gil Kim, Jeong Yong Park. Evaluation of Pretransition Oxide on Zr-0.4Nb Alloy by Using the HVEM[J].J Nuclear Mater,2008,37(4):204-206.
[46]Aylin Yilmazbay han, Else Breva. Transmissio on Micro Scopy Examination of Oxide Layers Formed on Zr Alloys[J].J Nuclear Mater,2006,34(9):265-259.
[47]Jeong Yong Park, Hyun Gil Kim.Crystal Structure and Grain Size of Zr Oxide Characterized by Synchrotr on Radiation Micro Diffraction[J].J Nuclear Mater,2004, 33(5):433-437.
[48]Moyal J S.Zirconium Oxide Film for Mation on Zircaloy by Water Corrosion [J].Acta Mater,2000,48(4):47-49.
[49]Motta A T,Y ilmazbayhan A. Revealed that All the Oxides had a Layer Structure[J]. J AST M Int,2005,8(2):8-10.
[50]Bossis P,T homazet J. Study of the Mechanisms Controlling the Oxide Growth Under Irradiation:Characterization of Irradiated Zr-4 and Zr-lNb-O Oxide Scales[J].J ASTM Int, 2002,6(4):190-193.
[51]Anada H, Her b B J. Zirconium in the Nuclear Industry[J].J ASTM Int,1996,8(2):183.
[52]陈名浩.无标样X射线衍射定量相分析的回归求解[J].金属学报,1988,24(3)：21-25.
[53]丘利,胡玉和.X射线衍射技术及设备[M].北京：冶金工业出版社,1998.
[54]蒋艳平,肖刘.射线衍射法测Zn电镀层中的残余应力[J].湘潭大学自然科学学报,2001,23(4)：42-44.
[55]郭常霖.精确测定低对称晶系多晶材料点阵常数的X射线衍射方法[J].无机材料学报,1996,11(4)：597-601.
[56]黄孝瑛.透射电子显微学[M].上海：上海科学技术出版社,1987.
[57]李中奎,刘建章.新锆合金在两种不同介质中的耐蚀性能[J].稀有金属材料与工程,1999,28(2)：101-104.
[58]朱梅生,刘建章.8#新锆合金的组织与耐蚀性能的研究[J].稀有金属材料与工程,1996,25(4)：34-38.
[59]邝用庚Zr-Nb系合金耐蚀性能研究[J].稀有金属材料与工程,1995,24(3)：41-45.
[60]刘文庆.合金元素及水化学对锆合金耐腐蚀性能影响的研究[D].上海：上海大学,2002.
[61]王均,龙冲生,熊计,等.新型高温锆合金在过热蒸汽中的腐蚀性能[J].核动力工程,2009,35(5)：58-61.
[62]张代东.机械工程材料应用基础[M].北京：机械工业出版社,2001.
[63]湛继明.钒合金在聚变堆环境中的使用脆性[J].腐蚀科学与防护技术,2002,14(6)：315-321.
[64]王文生.低铌新锆合金的抗蠕变性能[J].原子能科学与技术,2007(39)：14-17.
[65]张平.吸氢与堆内辐照对Zr-2包壳管拉伸性能影响的研究[D].北京：中国原子能科学研 究院,2006.
[66]崔旭梅.新锆合金的碘致应力腐蚀行为研究[D].成都：四川大学,2002.
[67]周邦新.Zr-4合金渗氢方法的研究[J].核动力工程,1988,23(1)：43-47.
[68]赵文金,苗志,蒋宏曼,等Zr-Sn-Nb合金的腐蚀行为研究[J].中国腐蚀与防护学报,2002,22(2)：124-128.
[69]周邦新,李强,黄强.水化学对锆合金腐蚀性能的影响[J].核动力工程,2002,9(5)：439-447.
[70]杨芳林,张建军,宋启忠.锆合金管材在氢氧化锂溶液中的吸氢研究[J].稀有金属,1999,2(3)：237-239.
[71]耿迅,刘庆冬,刘文庆,等Zr-1Sn-1Nb-0.1Fe合金第二相的研究[J].稀有金属材料与工程,2008,27(4)：717-720.
[72]李强,周邦新,姚美意,等.锆合金在550℃,25MPa超临界水中的腐蚀行为[J].稀有金属材料与工程,2007,36(8)：1358-1360.
[73]徐健,白新德,范毓殿,等.中子辐照对锆合金氧化性能的影响[J].清华大学学报,1998,37(6)：79-81
[74]李聪.锆合金的低周疲劳行为研究[D].成都：四川大学,2003.
[75]成海涛,郭元蓉.核电用管现状及国产化进展[J].钢管,2008,37(4)：1-5.
[2]孙家宁,孙剑彪.大亚湾核电站有关知识简介[J].物理教师,1996,6：36-37.
[3]张禄庆.压水堆技术发展趋势概述[J].中国核电,2008,1(1)：40-42.
[4]陈济东.大亚湾核电站系统及运行[M].北京：原子能出版社,1993.
[5]李自淳,宣自平.核电简介[J].上海大中型电机,2006,3：1-2.
[6]邹子和.核电站用管[J].钢管,1994,2(7).23-27.
[7]束国刚,陆念文.压水堆核电厂金属部件老化和寿命评估[J].我国电力,2006,39(5)：53-58.
[8]陈鹤鸣,马春来,白新德.核反应堆材料腐蚀及其防护[M].北京：原子能出版社,1984.
[9]杨文斗.反应堆材料学[M].北京：原子能出版社,2006.
[10]韩恩厚.超临界水环境中材料腐蚀研究现状[J].腐蚀科学和防护技术,1999,11(1)：53-57.
[11]刘建章.核结构材料[M].北京：化学工业出版社,2007.
[12]阎昌琪.核反应堆工程[M].哈尔滨：哈尔滨工程大学出版社,2004.
[13]吴玖.核电材料简介[M].北京：冶金工业出版社,1999.
[14]郑杰,牛惠玲.Zr-4合金工艺改进的研究[J].稀有金属,1995,19(4)：260-263.
[15]向德光,潘旦光,龙纯育,等.锆合金经历组织明场观察及测量的进项制备技术[G].国产Zr-4合金性能研究,1994,27(5)：156-159.
[16]周邦新,郑斯奎,汪顺新.真空电子束焊接对Zr-2合金熔区中成分、组织及腐蚀性能的影响[J].核科学与工程,1988,8(2)：130-137.
[17]周军,李中奎,刘荣臻.氢含量对新锆合金室温循环变形行为的影响[J].稀有金属,2008,37(5)：563-566.
[18]苟渊,李言荣,应诗浩.氢及氢化物对锆合金疲劳裂纹扩展速率的影响及估算[J].核动力工程,2005,23(4)：73-76.
[19].李献军.锆及锆合金概述[J].钛工业进展,2011,2(1)：28.
[20]周邦新,赵文金,蒋有荣,等.Zr-4合金元素表面偏聚[J].核动力工程,1998,19(6)：506-508.
[21]赵文金,苗志,蒋宏曼,等Zr-Sn-Nb合金的腐蚀行为研究[J].中国腐蚀与防护学报,2002,22(2)：124-128.
[22]李佩志,李中奎.合金元素对Zr-Nb合金耐蚀性能影响[J].稀有金属材料与工程,1998,27(6)：356-359.
[23]李中奎,刘建章.新锆合金在两种不同介质中的耐蚀性能[J].稀有金属材料与工程,1999,28(2)：101-104.
[24]朱梅生,刘建章.锆合金组织与耐蚀性能研究[J].稀有金属材料与工程,1996,25(4)：34-38.
[25]邝用庚Zr-Nb系合金耐蚀性能研究[J].稀有金属材料与工程,1995,24(3)：41-45.
[26]刘文庆.合金元素及水化学对锆合金耐腐蚀性能影响的研究[D].上海：上海大学,2002.
[27]李中奎,刘建章.合金元素对Zr-Sn-Fe-Cr-Nb合金性能的影响[J].稀有金属材料与工程,1996,25(5)：43-48.
[28]姚美意.合金成分及热处理对锆合金腐蚀和吸氢行为影响的研究[D].上海：上海大学,2007.
[29]袁改焕,白新德,刘振球.我国锆铪材加工技术的进步及发展[J].稀有金属快报,2004,2(3)：1-4.
[30]刘承新.锆合金在核工业中的应用现状及发展前景[J].稀有金属快报,2004,23(5)：21-23
[31]Grad G B, Pieres J J, Guillermet A F, et al. Lattice Parameter of the Zr-Nb Bcc Phase: Neutron Scattering Study and Assessment of Experimental Data[J]. Z. Met-allkd.1995, 86(6):395-400.
[32]Choo K N, Kang Y H, Pyum S I, et al. Effect of Composition and Heat Treatment on the Microstructure and Corrosion Behavior of Zr-Nb Alloys[J]. J Nucl Mat.,1994,20(9): 226-235.
[33]Lin Y P, Woo O T. Oxidation of β-Zr and Related Phases in Zr-Nb Alloy[J]. J Nucl Mat. 2000,27(7):11-27.
[34]刘文庆,李强,周邦新.锆锡合金腐蚀转折机理的讨论[J].稀有金属材料与工程,2001,30(2)：8-9.
[35]王文生.低铌新锆合金的抗蠕变性能[J].原子能科学与技术,2007,39(3)：14-17.
[36]陈桂辉.轻水堆核电站的原理与应用前景[J].福建建能源开发与节约,1996,27(1)：37-39.
[37]梁建烈,庄应烘,赵文金,等,新锆合金的相变温度[J].广西科学,2002,9(3)：181-185
[38]Yang Jung, M yung Ho Lee. Development of a Manufactuing Process for Zr Based Spacer Grid Materials [J]. J Nuclear Mater,2009,39(2):482-484.
[39]Yong Hwan Jeong. Effect of Phase, Precipitate and Nb Concentration in Matrix on Corrosion and Oxide Characteristics of Zr-Nb alloys[J].J Nuclear Mater,2003,31(7):1-5.
[40]Hyun Gil Kim, Yong Hawn Jeong. Effect of Isothermal Annealing on the Corrosion Behavior of Zr-xNb alloys[J]. J Nuclear Mater,2004,32(6):125-129.
[41]Caroline Toffolon Masclet.Study of Nb and Fe Precipitation in a Phase Temperature Range(400-550℃) in Zr Nb(Fe Sn)alloys[J].J A ST M Int,2005,2(5):81-83
[42]Hyun Gil Kim. Microstructure and Corrosion Characteristics of Zr-1.5Nb-0.4Sn-0.2Fe-0.1 Cr Alloy with an Nealing[J].J Nuclear Mater,2008,372(25):304-307.
[43]Jeong Yong Park, Seung Jo Yoo. Oxide Micro Structures of Advanced Zr Alloys Corroded in 360℃ Water Loop[J].J Alloys Compd,2007,43(7):274-275.
[44]Aylin Yilmazbay han.Structure of Zirconium Alloy Oxides for Med in Pure Water Studied with Synchro Tronradiation and Optical Microscopy[J].J Nuclear Mater,2004,32(4):6-11.
[45]Hyun Gil Kim, Jeong Yong Park. Evaluation of Pretransition Oxide on Zr-0.4Nb Alloy by Using the HVEM[J].J Nuclear Mater,2008,37(4):204-206.
[46]Aylin Yilmazbay han, Else Breva. Transmissio on Micro Scopy Examination of Oxide Layers Formed on Zr Alloys[J].J Nuclear Mater,2006,34(9):265-259.
[47]Jeong Yong Park, Hyun Gil Kim.Crystal Structure and Grain Size of Zr Oxide Characterized by Synchrotr on Radiation Micro Diffraction[J].J Nuclear Mater,2004, 33(5):433-437.
[48]Moyal J S.Zirconium Oxide Film for Mation on Zircaloy by Water Corrosion [J].Acta Mater,2000,48(4):47-49.
[49]Motta A T,Y ilmazbayhan A. Revealed that All the Oxides had a Layer Structure[J]. J AST M Int,2005,8(2):8-10.
[50]Bossis P,T homazet J. Study of the Mechanisms Controlling the Oxide Growth Under Irradiation:Characterization of Irradiated Zr-4 and Zr-lNb-O Oxide Scales[J].J ASTM Int, 2002,6(4):190-193.
[51]Anada H, Her b B J. Zirconium in the Nuclear Industry[J].J ASTM Int,1996,8(2):183.
[52]陈名浩.无标样X射线衍射定量相分析的回归求解[J].金属学报,1988,24(3)：21-25.
[53]丘利,胡玉和.X射线衍射技术及设备[M].北京：冶金工业出版社,1998.
[54]蒋艳平,肖刘.射线衍射法测Zn电镀层中的残余应力[J].湘潭大学自然科学学报,2001,23(4)：42-44.
[55]郭常霖.精确测定低对称晶系多晶材料点阵常数的X射线衍射方法[J].无机材料学报,1996,11(4)：597-601.
[56]黄孝瑛.透射电子显微学[M].上海：上海科学技术出版社,1987.
[57]李中奎,刘建章.新锆合金在两种不同介质中的耐蚀性能[J].稀有金属材料与工程,1999,28(2)：101-104.
[58]朱梅生,刘建章.8#新锆合金的组织与耐蚀性能的研究[J].稀有金属材料与工程,1996,25(4)：34-38.
[59]邝用庚Zr-Nb系合金耐蚀性能研究[J].稀有金属材料与工程,1995,24(3)：41-45.
[60]刘文庆.合金元素及水化学对锆合金耐腐蚀性能影响的研究[D].上海：上海大学,2002.
[61]王均,龙冲生,熊计,等.新型高温锆合金在过热蒸汽中的腐蚀性能[J].核动力工程,2009,35(5)：58-61.
[62]张代东.机械工程材料应用基础[M].北京：机械工业出版社,2001.
[63]湛继明.钒合金在聚变堆环境中的使用脆性[J].腐蚀科学与防护技术,2002,14(6)：315-321.
[64]王文生.低铌新锆合金的抗蠕变性能[J].原子能科学与技术,2007(39)：14-17.
[65]张平.吸氢与堆内辐照对Zr-2包壳管拉伸性能影响的研究[D].北京：中国原子能科学研 究院,2006.
[66]崔旭梅.新锆合金的碘致应力腐蚀行为研究[D].成都：四川大学,2002.
[67]周邦新.Zr-4合金渗氢方法的研究[J].核动力工程,1988,23(1)：43-47.
[68]赵文金,苗志,蒋宏曼,等Zr-Sn-Nb合金的腐蚀行为研究[J].中国腐蚀与防护学报,2002,22(2)：124-128.
[69]周邦新,李强,黄强.水化学对锆合金腐蚀性能的影响[J].核动力工程,2002,9(5)：439-447.
[70]杨芳林,张建军,宋启忠.锆合金管材在氢氧化锂溶液中的吸氢研究[J].稀有金属,1999,2(3)：237-239.
[71]耿迅,刘庆冬,刘文庆,等Zr-1Sn-1Nb-0.1Fe合金第二相的研究[J].稀有金属材料与工程,2008,27(4)：717-720.
[72]李强,周邦新,姚美意,等.锆合金在550℃,25MPa超临界水中的腐蚀行为[J].稀有金属材料与工程,2007,36(8)：1358-1360.
[73]徐健,白新德,范毓殿,等.中子辐照对锆合金氧化性能的影响[J].清华大学学报,1998,37(6)：79-81
[74]李聪.锆合金的低周疲劳行为研究[D].成都：四川大学,2003.
[75]成海涛,郭元蓉.核电用管现状及国产化进展[J].钢管,2008,37(4)：1-5.