Please wait a minute...
金属学报  2020, Vol. 56 Issue (2): 221-230    DOI: 10.11900/0412.1961.2019.00191
  研究论文 本期目录 | 过刊浏览 |
Zr-0.75Sn-0.35Fe-0.15Cr合金在250 ℃去离子水中的初期腐蚀行为
姚美意1,2(),张兴旺1,2,侯可可1,2,张金龙1,2,胡鹏飞1,2,彭剑超1,2,周邦新1,2
1. 上海大学材料研究所 上海 200072
2. 上海大学微结构重点实验室 上海 200444
The Initial Corrosion Behavior of Zr-0.75Sn-0.35Fe-0.15Cr Alloy in Deionized Water at 250 ℃
YAO Meiyi1,2(),ZHANG Xingwang1,2,HOU Keke1,2,ZHANG Jinlong1,2,HU Pengfei1,2,PENG Jianchao1,2,ZHOU Bangxin1,2
1. Institute of Materials, Shanghai University, Shanghai 200072, China
2. Laboratory for Microstructures, Shanghai University, Shanghai 200444, China
全文: PDF(18461 KB)   HTML
摘要: 

为研究锆合金从开始氧化至生成ZrO2的相组成及其晶体结构变化,采用锆合金大晶粒TEM薄样品在250 ℃、3 MPa去离子水中短时腐蚀的方法,利用距离TEM薄样品穿孔周围不同距离处样品厚度差别造成的O含量差别,采用HRTEM研究了Zr-0.75Sn-0.35Fe-0.15Cr合金的初期腐蚀行为以及早期形成的氧化膜晶体结构演化过程。结果表明:从开始氧化至ZrO2形成前,α-Zr的晶格点阵随着样品中O含量增加而不断演变;在Zr/O原子比为5~7时,基体晶格中原本无序的O原子有序地固溶在α-Zr中,形成有公度的长周期超点阵,其晶格常数(ac)与基本晶格α-Zr的晶格常数(a0c0)之间的关系为a=9a0c=2c0,称其为9a0-2H结构;当Zr/O原子比为3时,形成具有hcp超结构的Zr3O亚氧化物;当Zr/O原子比为1时,转变为具有fcc超结构的ZrO亚氧化物;Zr/O原子比为0.85时,形成单斜结构ZrO2

关键词 锆合金晶体结构转变初期腐蚀亚氧化物    
Abstract

Zirconium alloys are important structural materials in pressurized water reactors. During actual operation, the corrosion resistance of water side is the most important factor affecting its service life. The oxide film of zirconium alloys formed during the corrosion process will reduce the heat transfer performance, mechanical properties and service life of the cladding material, thus becoming a factor restricting the development of nuclear power. The initial phase composition and the defect state in the crystal affect the microstructural evolution of the oxide film during the corrosion process, which in turn determines the late growth of the oxide film. In order to study the phase composition and crystal structure evolution of zirconium alloys from the initial oxidation to the formation of ZrO2, the initial corrosion behavior of Zr-0.75Sn-0.35Fe-0.15Cr alloy was studied by using TEM thin foil specimens with coarse grains. The oxygen content varied due to the change of sample thickness at different distances along the perforation of TEM thin foil specimens with coarse grains, which could be investigated the crystal structure evolution of oxide film with the variation of oxygen content. Corrosion tests of these TEM specimens were conducted in an autoclave at 250 ℃ and 3 MPa in deionized water for short time exposure. The results showed a variation of the crystal structure along with the increase of oxygen contents at the initial oxidation stage. When the Zr/O atomic ratio reached 5~7, a commensurable long period super-lattice structure was formed. The lattice constants of the super-lattice (a, c) and α-Zr matrix (a0, c0) satisfied the relationship of a=9a0 and c=2c0, which was called 9a0-2H structure. When the Zr/O atomic ratio reached 3 and 1, sub-oxides Zr3O with hcp and ZrO with fcc ordered structures were formed, respectively. When the Zr/O atomic ratio was 0.85, monoclinic ZrO2 was detected.

Key wordszirconium alloy    crystal structural evolution    initial oxidation    sub-oxide
收稿日期: 2019-06-11     
ZTFLH:  TG146.4  
基金资助:国家自然科学基金项目(51471102);国家自然科学基金项目(51871141)
通讯作者: 姚美意     E-mail: yaomeiyi@shu.edu.cn
Corresponding author: Meiyi YAO     E-mail: yaomeiyi@shu.edu.cn
作者简介: 姚美意,女,1973年生,研究员,博士

引用本文:

姚美意,张兴旺,侯可可,张金龙,胡鹏飞,彭剑超,周邦新. Zr-0.75Sn-0.35Fe-0.15Cr合金在250 ℃去离子水中的初期腐蚀行为[J]. 金属学报, 2020, 56(2): 221-230.
Meiyi YAO, Xingwang ZHANG, Keke HOU, Jinlong ZHANG, Pengfei HU, Jianchao PENG, Bangxin ZHOU. The Initial Corrosion Behavior of Zr-0.75Sn-0.35Fe-0.15Cr Alloy in Deionized Water at 250 ℃. Acta Metall Sin, 2020, 56(2): 221-230.

链接本文:

https://www.ams.org.cn/CN/10.11900/0412.1961.2019.00191      或      https://www.ams.org.cn/CN/Y2020/V56/I2/221

图1  Zr-0.75Sn-0.35Fe-0.15Cr合金大晶粒样品腐蚀前显微组织的OM像和TEM像
图2  Zr-0.75Sn-0.35Fe-0.15Cr合金大晶粒TEM样品经过250 ℃、3 MPa去离子水腐蚀后沿着孔洞直径方向从厚到薄每隔500 nm拍摄的TEM像
图3  分别对应图2a~i中点1~9的SAED花样
PointAtomic fraction of Zr / %Atomic fraction of O / %Zr/O ratio
191.206.9613.11
285.5614.016.11
383.5616.215.15
482.9817.354.78
579.4820.463.88
674.6424.113.10
766.3432.282.06
856.2543.751.29
945.3453.040.85
表1  图2中点1~9处EDS分析
图4  图2i中点9的TEM像、SAED花样、HRTEM像和FFT图
图5  Zr/O比为5~7时形成的超结构点阵分析
图6  Zr/O比为3.10处(图2f中点6)的TEM像和超结构SAED花样
图7  Zr/O比为1.29处(图2h中点8)的TEM像和形成的超结构SAED花样
[1] Cox B. Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys [J]. J. Nucl. Mater., 2005, 336: 331
[2] Zhou B X, Li Q, Liu W Q, et al. The effects of water chemistry and composition on the microstructure evolution of oxide films on zirconium alloys during autoclave tests [J]. Rare Met. Mater. Eng., 2006, 35: 1009
[2] (周邦新, 李 强, 刘文庆等. 水化学及合金成分对锆合金腐蚀时氧化膜显微组织演化的影响 [J]. 稀有金属材料与工程, 2006, 35: 1009)
[3] Yilmazbayhan A, Breval E, Motta A T, et al. Transmission electron microscopy examination of oxide layers formed on Zr alloys [J]. J. Nucl. Mater., 2006, 349: 265
[4] Ni N, Lozano-Perez S, Sykes J, et al. Quantitative EELS analysis of zirconium alloy metal/oxide interfaces [J]. Ultramicroscopy, 2011, 111: 123
[5] Motta A T, Yilmazbayhan A, da Silva M J G, et al. Zirconium alloys for supercritical water reactor applications: Challenges and possibilities [J]. J. Nucl. Mater., 2007, 371: 61
[6] Iltis X, Lefebvre F, Lemaignan C. Microstructure evolutions and iron redistribution in zircaloy oxide layers: Comparative effects of neutron irradiation flux and irradiation damages [A]. Zirconium in the Nuclear Industry: 11th International Symposium [C]. West Conshohocken. PA: ASTM International Press, 1996: 242
[7] Vermaak N, Parry G, Estevez R, et al. New insight into crack formation during corrosion of zirconium-based metal-oxide systems [J]. Acta Mater., 2013, 61: 4374
[8] Garzarolli E, Seidel H, Tricot R, et al. Oxide growth mechanism on zirconium alloys [A]. Zirconium in the Nuclear Industry: 9th International Symposium [C]. West Conshohocken. PA: ASTM International Press, 1991: 395
[9] Gong W J, Zhang H L, Qiao Y, et al. Grain morphology and crystal structure of pre-transition oxides formed on zircaloy-4 [J]. Corros. Sci., 2013, 74: 323
[10] Pétigny N, Barberis P, Lemaignan C, et al. In situ XRD analysis of the oxide layers formed by oxidation at 743 K on zircaloy 4 and Zr-1NbO [J]. J. Nucl. Mater., 2000, 280: 318
[11] Ploc R A. Transmission electron microscopy of αZrO2 films formed in 573 K oxygen [J]. J. Nucl. Mater., 1976, 61: 79
[12] Warr B D, Elmoselhi M B, Newcomb S B, et al. Oxide characteristics and their relationship to hydrogen uptake in zirconium alloys [A]. Zirconium in the Nuclear Industry: Ninth International Symposium [C]. West Conshohocken. PA: ASTM International Press, 1991: 740
[13] Gou S Q, Zhou B X, Chen C M, et al. Investigation of oxide layers formed on zircaloy-4 coarse-grained specimens corroded at 360 ℃in lithiated aqueous solution[J]. Corros. Sci., 2015, 92: 237
[14] 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
[15] Bossis P, Lelièvre G, Barberis P, et al. Multi-scale characterization of the metal-oxide interface of zirconium alloys [A]. Zirconium in the Nuclear Industry: 12th International Symposium [C]. West Conshohocken. PA: ASTM International Press, 2000: 918
[16] Wei J, Frankel P, Polatidis E, et al. The effect of Sn on autoclave corrosion performance and corrosion mechanisms in Zr-Sn-Nb alloys [J]. Acta Mater., 2013, 61: 4200
[17] Qiu J, Zhao W J, Guilbert T, et al. High temperature oxidation behaviours of three zirconium alloys [J]. Acta Metall. Sin., 2011, 47: 1216
[17] (邱 军, 赵文金, Guilbert T等. 3种锆合金的高温氧化行为 [J]. 金属学报, 2011, 47: 1216)
[18] Zhou B X, Peng J C, Yao M Y, et al. Study of the initial stage and anisotropic growth of oxide layers formed on Zircaloy-4 [A]. Zirconium in the Nuclear Industry: 16th International Symposium [C]. West Conshohocken. PA: ASTM International Press, 2012: 620
[19] Sun G C, Zhou B X, Yao M Y, et al. Study of anisotropic behavior for zirconium alloys corroded in lithiated water [J]. Acta Metall. Sin., 2012, 48: 1103
[19] (孙国成, 周邦新, 姚美意等. 锆合金在LiOH水溶液中腐蚀的各向异性研究 [J]. 金属学报, 2012, 48: 1103)
[20] Wang Z, Zhou B X, Wang B Y, et al. Oxide layers formed earlier on zircaloy-4 alloy corroded at 300 ℃ in deionized water [J]. Rare Met. Mater. Eng., 2017, 46: 1602
[20] (王 桢, 周邦新, 王波阳等. Zr-4合金在300 ℃去离子水中腐蚀初期氧化膜的研究 [J]. 稀有金属材料与工程, 2017, 46: 1602)
[21] Anada H, Takeda K. Microstructure of oxides on zircaloy-4, 1.0Nb zircaloy-4, and zircaloy-2 formed in 10.3 MPa steam at 673 K [A]. Zirconium in the Nuclear Industry: 11th International Symposium [C]. West Conshohocken. PA: ASTM International Press, 1996: 35
[22] Dong Y, Motta A T, Marquis E A. Atom probe tomography study of alloying element distributions in Zr alloys and their oxides [J]. J. Nucl. Mater., 2013, 442: 270
[23] Zhou B X, Li Q, Yao M Y, et al. Microstructure of oxide films formed on zircaloy-4 [J]. Corros. Prot., 2009, 30: 589
[23] (周邦新, 李 强, 姚美意等. Zr-4合金氧化膜的显微组织研究 [J]. 腐蚀与防护, 2009, 30: 589)
[24] Zhou B X, Li Q, Yao M Y, et al. The grains morphology of oxide films for zircaloy-4 [J]. Rare Met. Mater. Eng., 2003, 32: 417
[24] (周邦新, 李 强, 姚美意等. 锆-4合金氧化膜中的晶粒形貌观察 [J]. 稀有金属材料与工程, 2003, 32: 417)
[25] Hiraga K, Shindo D, Hirabayashi M. High-voltage, high-resolution electron microscopy of Au-Cd alloys. III. One-dimensional long-period superstructure of D023 type [J]. J. Appl. Crystallogr., 1981, 14: 185
[26] Hirabayash M, Hiraga K, Shindo O. High-Voltage, high-resolution electron microscopy of Au-Cd alloys. I. Hexagonal long-period superstructures near 30 at.% Cd [J]. J. Appl. Crystallogr., 1981, 14: 169
[27] Rong Y H. Introduction to Analytical Electron Microscopy [M]. 2nd Ed., Beijing: Higher Education Press, 2015: 60
[27] (戎咏华. 分析电子显微学导论 [M]. 第2版. 北京: 高等教育出版社, 2015: 60
[1] 姚美意, 林雨晨, 侯可可, 梁雪, 胡鹏飞, 张金龙, 周邦新. Sn对锆合金在280 LiOH水溶液中初期腐蚀行为的影响[J]. 金属学报, 2019, 55(12): 1551-1560.
[2] 陈兵,高长源,黄娇,毛亚婧,姚美意,张金龙,周邦新,李强. β-(Nb, Zr)第二相合金在360 ℃去离子水中的腐蚀行为[J]. 金属学报, 2017, 53(4): 447-454.
[3] 任伊宾, 李俊, 王青川, 杨柯. MRI磁兼容合金研究[J]. 金属学报, 2017, 53(10): 1323-1330.
[4] 韦天国,林建康,龙冲生,陈洪生. 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.
[5] 张诚,宋西平,刘敬茹,杨云,尤力. 氘含量对Zr-4合金显微组织和力学性能的影响*[J]. 金属学报, 2016, 52(12): 1572-1578.
[6] 张骏,姚美意,冯炫凯,王志刚,黄娇,戴训,张金龙,周邦新. Zr-Sn-Fe-Cr-(Nb)合金在500 ℃过热蒸汽中的腐蚀各向异性研究*[J]. 金属学报, 2016, 52(12): 1565-1571.
[7] 王桢,周邦新,王波阳,黄娇,姚美意,张金龙. Zr-0.72Sn-0.32Fe-0.15Cr-0.97Nb合金中的第二相及其腐蚀行为*[J]. 金属学报, 2016, 52(1): 78-84.
[8] 王波阳,周邦新,王桢,黄娇,姚美意,周军. Zr-0.72Sn-0.32Fe-0.14Cr-xNb合金在500 ℃过热蒸汽中的耐腐蚀性能*[J]. 金属学报, 2015, 51(12): 1545-1552.
[9] 章海霞, 李中奎, 周廉, 许并社, 王永祯. 氧化膜结构及内应力对新锆合金腐蚀机理的影响[J]. 金属学报, 2014, 50(12): 1529-1537.
[10] 李烨, 张龙, 朱正旺, 李宏, 王爱民, 张海峰. 热处理对一种高强Zr-Ti合金组织和力学性能的影响*[J]. 金属学报, 2014, 50(1): 19-24.
[11] 韦天国,龙冲生,苗志,刘云明,栾佰峰. Zr-0.4Fe-1.0Cr-x Mo合金在500℃和10.3 MPa水蒸汽中的腐蚀行为[J]. 金属学报, 2013, 49(6): 717-724.
[12] 张金龙,谢兴飞,姚美意,周邦新,彭剑超,梁雪. Zr-1Nb-0.7Sn-0.03Fe-xGe合金在360 ℃ LiOH水溶液中耐腐蚀性能的研究[J]. 金属学报, 2013, 29(4): 443-450.
[13] 姚美意 邹玲红 谢兴飞 张金龙 彭剑超 周邦新. 添加Bi对Zr-4合金在400 ℃/10.3 MPa过热蒸汽中耐腐蚀性能的影响[J]. 金属学报, 2012, 48(9): 1097-1102.
[14] 孙国成 周邦新 姚美意 谢世敬 李强. 锆合金在LiOH水溶液中腐蚀的各向异性研究[J]. 金属学报, 2012, 48(9): 1103-1108.
[15] 谢兴飞 张金龙 朱莉 姚美意 周邦新 彭剑超. Zr-0.7Sn-0.35Nb-0.3Fe-xGe合金在高温高压LiOH 水溶液中耐腐蚀性能的研究[J]. 金属学报, 2012, 48(12): 1487-1494.