辐照后N18锆合金氢致延迟开裂临界温度预测
收稿日期: 2009-12-29
修回日期: 2010-03-31
网络出版日期: 2010-07-11
基金资助
中国核动力研究设计院青年科学基金资助项目JJXM--QNJJ--09--04--04
PREDICTION OF CRITICAL TEMPERATURE FOR DELAYED HYDRIDE CRACKING IN IRRADIATED N18 ZIRCONIUM ALLOY
Received date: 2009-12-29
Revised date: 2010-03-31
Online published: 2010-07-11
Supported by
Supported by Foundation for Young Scholars of NPIC (No.JJXM-QNJJ-09-04-04)
基于辐照后锆合金中H的固溶与析出浓度方程及应力诱导H扩散方程, 建立了辐照后锆合金氢致延迟开裂(DHC)临界温度的预测模型. 采用此模型对辐照后N18锆合金(Zr-Sn-Nb)的DHC临界温度进行预测, 并与未辐照N18锆合金临界温度实验数据进行对比分析. 结果表明, 中子辐照后N18锆合金屈服强度和H的固溶度增加, N18合金发生DHC的敏感性增加; 相同H含量条件下, 辐照后N18合金的临界温度最高上升幅度接近20 ℃; 辐照后N18锆合金临界温度变化规律与未辐照实验结果相似, 其临界温度介于相同H含量下完全固溶的温度与氢化物析出时的温度之间, 且最高开裂温度Tc低于最低止裂温度Th.
孙超 , 谭军 , 应诗浩 , 彭倩 , 李聪 , 赵素琼 . 辐照后N18锆合金氢致延迟开裂临界温度预测[J]. 金属学报, 2010 , 46(7) : 805 -809 . DOI: 10.3724/SP.J.1037.2009.00868
Delayed hydride cracking (DHC) has been recognized as a potential threat to the structural integrity of zirconium alloy components in water-cooled nuclear reactors since the early 1970 s. Although DHC has been studied for a long time, most of the investigations were focused on unirradiated Zr-Nb and Zr-Sn alloys, no information was found on Zr-Sn-Nb alloy. Currently, several new Zr-based alloys have been developed from the point of view of enhancement of corrosion resistance and hydrogen pickup properties. In China, N18 (Zr-Sn-Nb) alloy has been developed based on Zr-Sn and Zr-Nb systems to meet the requirements of higher fuel burn-up. Therefore, it is necessary to investigate the DHC behavior of N18 alloy systematically. The objective of this study is to predict the critical temperatures for initiating and arresting DHC in irradiated N18 alloy. A model used to forecast DHC critical temperature of irradiated zirconium alloys was established based on the terminal solid solubility of hydrogen in irradiated zirconium alloys and stress-induced hydrogen diffusion. By using the model, the critical temperatures for irradiated N18 alloy were predicted. The result showed that the irradiated N18 alloy has high sensitivity for DHC initiation, because neutron irradiation increased the solubility of hydrogen and yield strength of N18 alloy. Compared with unirradiated N18 alloy, the critical temperature of irradiated N18 alloy rise nearly 20℃. The variation of critical temperatures of irradiated N18 alloy is similar to the unirradiated one. Tc is slightly less than Th.
[1] Chu W Y. Hydrogen Damage and Delayed Fracture. Beijing: Metallurgical Industry Press, 1988: 189
(褚武扬. H损伤及滞后断裂. 北京: 冶金工业出版社, 1988: 189)
[2] Matsui H, Yoshikawa N, Koiwa M. Acta Metall, 1987; 35: 413
[3] Cann C D, Sexton E E. Acta Metall, 1980; 28: 1215
[4] Pan C, Li Z B, Liang D T, Tian Z L, Chu W Y, Qiao L J. Acta Metall Sin, 2001; 37: 296
(潘川, 李正邦, 梁东图, 田志凌, 褚武扬, 乔利杰. 金属学报, 2001; 37: 296)
[5] Gao K W, Wang Y B, Qiao L J, Chu W Y. Sci China, 1999; 29E: 289
(高克玮, 王燕斌, 乔利杰, 褚武扬. 中国科学, 1999; 29E: 289)
[6] Dutton R, Nuttall K, Puls M P, Simpson L A. Metall Trans, 1977; 8A: 1553
[7] Shi S Q, Puls M P. J Nucl Mater, 1994; 208: 232
[8] Shi S Q, Shek G K, Puls M P. J Nucl Mater, 1995; 218: 189
[9] Yang W D. Nuclear Reactor Material. Beijing: Atom Energy Press, 2000: 277
(杨文斗. 反应堆材料学. 北京: 原子能出版社, 2000: 277)
[10] Perryman E C W. Nucl Energy, 1978; 17: 95
[11] Efsing P, Pettersson K. Zirconium in the Nuclear Industry: 12th International Symposium, West Conshohocken: ASTM International, 2000: 340
[12] Viatcheslav G, Robert J. J ASTM Int, 2005; 2: 711
[13] Schofield J S, Edward C D, Gee C F. Zirconium in the Nuclear Industry: 13th International Symposium, West Conshohocken: ASTM International, 2002: 339
[14] Liu Y Z, Zu X T, Qiu S Y, Li C, Ma W G, Huang X Q. Surf Coat Technol, 2006; 200: 5631
[15] ZhaoWJ, Liu Y Z, Jiang H M, Peng Q. J Alloys Compd, 2008; 462: 103
[16] Tan J, Ying S H, Li C, Sun C. Scr Mater, 2006; 55: 513
[17] Sun C, Tan J, Ying S H, Li C, Peng Q, Zhao S Q. Acta Metall Sin, 2009; 5: 541
(孙超, 谭军, 应诗浩, 李聪, 彭倩, 赵素琼. 金属学报, 2009; 5: 541)
[18] Tang J R. Chin J Nuclear Sci Eng, 2003; 23: 266
(唐炯然. 核科学与工程, 2003; 23: 266)
[19] Liu J Z. Nuclear Structure Materials. Beijing: Chemistry Industry Press, 2007: 143
(刘建章. 核结构材料. 北京: 化学工业出版社, 2007: 143)
[20] Une K, Ishimoto S. J Nucl Mater, 2003; 322: 66
[21] McMinn A, Edward C D, Schofield J S. Zirconium in the Nuclear Industry: 12th International Symposium, West Conshohocken: ASTM International, 2000: 173
[22] Li J C M, Oriani R A, Darken L S. Z Phys Chem, 1966; 49: 271
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