NST  >> Vol. 5 No. 2 (April 2017)

    锆合金中的氢化锆在400℃过热蒸汽中的氧化行为
    Oxidation Behavior of Zirconium Hydride in Zirconium Alloys in Superheated Steam at 400?C

  • 全文下载: PDF(1621KB) HTML    PP.77-83   DOI: 10.12677/NST.2017.52011  
  • 下载量: 424  浏览量: 774   国家自然科学基金支持

作者:  

毛亚婧,段文荣,黄 娇,姚美意,张金龙,周邦新:上海大学材料研究所,上海;上海大学微结构重点实验室,上海;
袁改焕,高 博,孙国成:国核宝钛锆业股份公司,陕西 宝鸡

关键词:
金属材料氢化锆氧化应力再取向Metallic Material Zirconium Hydride Oxidation Stress Reorientation

摘要:

采用气相渗氢法制备了SZA-4合金预渗氢样品,通过高压釜腐蚀实验研究了锆合金中氢化锆的氧化行为。结果表明:在氧化膜/金属(O/M)界面附近金属基体中较大张应力的作用下,基体中已存在的氢化锆发生应力再取向,近似平行于O/M界面的面心立方(fcc)结构的δ-ZrH1.66变为垂直于O/M界面的面心四方(fct)结构的ε-ZrH2。ε-ZrH2氧化速度比α-Zr基体快,氧化为单斜ZrO2(m-ZrO2)。

To investigate the oxidation of zirconium hydride, the corrosion test on pre-hydrided SZA-4 alloy prepared by gaseous hydrogen charging method was performed in an autoclave. Results showed that under high tensile stress produced in α-Zr matrix at the oxide/metal (O/M) interface, hydride reorientation took place and led to the transition from δ-ZrH1.66 (fcc structure) parallel to the O/M interface to ε-ZrH2 (fct structure) perpendicular to the O/M interface. ε-ZrH2 oxidized faster than α-Zr matrix, and was oxidized to m-ZrO2(monoclinic ZrO2).

文章引用:
毛亚婧, 段文荣, 袁改焕, 高博, 孙国成, 黄娇, 姚美意, 张金龙, 周邦新. 锆合金中的氢化锆在400℃过热蒸汽中的氧化行为[J]. 核科学与技术, 2017, 5(2): 77-83. https://doi.org/10.12677/NST.2017.52011

参考文献

[1] 刘建章, 赵文金, 薛祥义, 陆世英. 核结构材料[M]. 北京: 化学工业出版社, 2007: 142.
[2] Ivanova, S.V. (2002) Effect of Hydrogen on Serviceability of Zirconium Items in VVER and REMK-Type Reactors Fuel Assemblies. International Journal of Hydrogen Energy, 27, 819-824.
[3] Ivanova, S.V. (2006) Hydrogen Effected Defects Evolution in Zirconium Items of Light-water Reactors. International Journal of Hydrogen Energy, 31, 295-300.
[4] Cox, B. (1990) Environmentally-Induced Cracking of Zirconium Alloys. Journal of Nuclear Materials, 170, 1-23.
[5] Singh, R.N., Kumar, N., Kishore, R., Roychaudhury, S., Sinha, T.K. and Kashyap, B.P. (2002) Delayed Hydride Cracking in Zr-2.5Nb Pressure Tube Material. Journal of Nuclear Materials, 304, 189-203.
[6] Kim, S.J., Kim, K.H., Beak, J.H., Choi, B.K., Jeong, Y.H. and Jung, Y.H. (1998) The Effect of Hydride on the Corrosion of Zircaloy-4 in Aqueous LiOH Solution. Journal of Nuclear Materials, 256, 114-123.
[7] Kim, Y.S., Rheem, K.S. and Min, D.K. (1994) Phenomenological Study of In-Reactor Corrosion of Zircaloy-4 in Pressurized Water Reactor. In: Garde, A.M. and Bradley, E.R., Eds., Zirconium in the Nuclear Industry: 10th International Symposium, ASTM STP 1245, American Society for Testing and Materials, Philadelphia, 745-759.
https://doi.org/10.1520/stp15218s
[8] Blat, M. and Noel, D. (1996) Detrimental Role of Hydrogen on the Corrosion Rate of Zirconium Alloys. In: Bradley, E.R. and Sabol, G.P., Eds., Zirconium in the Nuclear Industry: 11th International Symposium, ASTM STP 1295, American Society for Testing and Materials, Philadelphia, 319-337.
https://doi.org/10.1520/stp16179s
[9] Zhou, B.X. (1989) Electron Microscopy Study of Oxide Films Formed on Zircaloy-2 in Superheated Steam. In: Van Swam, L.E P. and Eucken, C.M., Eds., Zirconium in the Nuclear Industry: 8th International Symposium, ASTM STP 1023, American Society for Testing and Materials, Philadelphia, 360-373.
[10] 耿建桥, 周邦新, 姚美意, 王锦红, 张欣, 李士炉, 杜晨曦. 水化学和腐蚀温度对锆合金氧化膜中压应力的影响[J]. 上海大学学报(自然科学版), 2011, 17: 293-296.
[11] Garzarolli, F., Seidsl, H., Tricot, R. and Gros, J.P. (1991) Oxide Growth Mechanism on Zirconium Alloys. In: Eucken, C.M. and Garde, A.M., Eds., Zirconium in the Nuclear Industry: 9th International Symposium, ASTM STP 1132, American Society for Testing and Materials, Philadelphia, 395-415.
https://doi.org/10.1520/stp25519s
[12] 周邦新, 蒋有荣. 锆-2合金在500-800℃空气中氧化过程的研究[J]. 核动力工程, 1990, 11(3): 233-239.
[13] Godlewski, J. (1994) How the Tetragonal Zirconia Is Stabilized in the Oxide Scale That Is Formed on a Zirconium Alloy Corroded at 400℃ in Steam. In: Garde, A.M. and Bradley, E.R., Eds., Zirconium in the Nuclear Industry: 10th International Symposium, ASTM STP 1245, American Society for Testing and Materials, Philadelphia, 663-686.
https://doi.org/10.1520/stp15214s
[14] Singh, R.N., Kishore, R.S., Singh, S., Sinha, T.K. and Kashyap, B.P. (2004) Stress-Reorientation of Hydrides and Hydride Embrittlement of Zr-2.5 Nbwt% Pressure Tube Alloy. Journal of Nuclear Materials, 325, 26-33.
[15] Singh, R.N., Mikin, R.L., Dey, G.K., Sah, D.N., Batra, I.S. and Ståhle, P. (2006) Influence of Temperature on Threshold Stress for Reorientation of Hydrides and Residual Stress Variation across Thickness of Zr-2.5Nb Alloy Pressure Tube. Journal of Nuclear Materials, 359, 208-219.
[16] Chu, H.C., Wu, S.K. and Kuo, R.C. (2008) Hydride Reorientation in Zircaloy-4 Cladding. Journal of Nuclear Materials, 373, 319-327.
[17] Masaki, A., Toshikazu, B., Toshiyasu, M., Katsuichiro, K., Takayoshi, Y., Yasunari, S. and Toru, T. (2008) Evaluation of Hydride Reorientation Behavior and Mechanical Properties for High-Burnup Fuel-Cladding Tubes in Interim Dry Storage. In: Kammenzind, B. and Limbäck, M., Eds., Zirconium in the Nuclear Industry: 15th International Symposium, ASTM STP 1505, ASTM International, Baltimore, 651-674.
[18] Min, S.J., Kim, M.S. and Kim, K.T. (2013) Cooling Rate- and Hydrogen Content-Dependent Hydride Reorientation and Mechanical Property Degradation of Zr-Nb Alloy Claddings. Journal of Nuclear Materials, 441, 306-314.
[19] Colas, K.B., Motta, A.T., Almer, J.D., Daymond, M.R., Kerr, M., Banchik, A.D., Vizcaino, P. and Santisteban, J.R. (2010) In Situ Study of Hydride Precipitation Kinetics and Re-orientation in Zircaloy Using Synchrotron Radiation. Acta Materialia, 58, 6575-6583.
[20] 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, 49(4): 443-450.
[21] Lim, B.H., Hong, H.S. and Lee, K.S. (2003) Measurement of Hydrogen Permeation and Absorption in Zirconium Oxide Scales. Journal of Nuclear Materials, 312, 134-140.
[22] Zhou, B.X., Yao, M.Y., Li, Z.K., Wang, X.M., Zhoua, J., Long, C.S., Liu, Q. and Luan, B.F. (2012) Optimization of N18 Zirconium Alloy for Fuel Cladding of Water Reactors. Journal of Materials Science & Technology, 28, 606-613.
[23] 王建伟, 黄倩, 陈伟东, 王力军. 氢化锆在高温水蒸气中的氧化行为[J]. 材料导报, 2011, 25(6): 4-6.
[24] Tupin, M., Bisor, C., Bossis, P., Chêne, J., Bechade, J.L. and Jomard, F. (2015) Mechanism of Corrosion of Zirconium Hydride and Impact of Precipitated Hydrides on the Zircaloy-4 Corrosion Behavior. Corrosion Science, 98, 478-493.