锆合金是一种新型的核反应堆用包壳材料.为测定锆合金经高频感应氧化后所获得的陶瓷膜导热系数及其膜厚,将锆合金包壳内装满蒸馏水后放入HS-4(B)型恒温浴槽,利用Advantech VisiDAQ软件纪录下包壳内水温由室温上升至60℃中水温变化的整个过程.然后对相关试验参数进行了分析,获得了氧化膜厚度与导热系数的关系曲线.最后,得出结论:锆合金包壳陶瓷膜厚度以2μm为宜.
参考文献
[1] | Barheris P;Frichet A .Characterization of Zircaloy-4 oxidation layers by impedance spectroscopy[J].Journal of Nuclear Materials,1999,273:182-191. |
[2] | Khatamian D.;Lalonde SD. .CRYSTAL STRUCTURE OF THIN OXIDE FILMS GROWN ON ZR-NB ALLOYS STUDIED BY RHEED[J].Journal of Nuclear Materials: Materials Aspects of Fission and Fusion,1997(1):10-16. |
[3] | Neogy S.;Srivastava D.;Tewari R.;Singh RN.;Dey GK.;Banerjee S. .Microstructural study of hydride formation in Zr-1Nb alloy[J].Journal of Nuclear Materials: Materials Aspects of Fission and Fusion,2003(2/3):195-203. |
[4] | Veshchnov M S;Berdyshev A V .Modelling of hydrogen absorption by zirconium alloys during high temperature oxidation in steam[J].Journal of Nuclear Materials,1998,255:250-262. |
[5] | 张俊才.核电站的反应堆及其安全性[J].四川电力技术,1994(06):45-48. |
[6] | 朱关仁;滕利军 .美国西屋公司先进的防碎片磨损燃料棒技术[J].核电工程与技术,1997,10(02):25-26. |
[7] | 钱翰城 .锆合金管表面处理技术[P].CN 2004100289840,2004-10-23. |
[8] | 赵文金.M5合金的堆内外性能概述[J].核动力工程,2001(01):60-64. |
[9] | 刘仁志.电镀新工艺和新技术的回顾与展望[J].表面技术,2001(05):7-9. |
[10] | 扎依莫夫斯基;姚敏智.核动力用锆合金[M].北京:原子能出版社,1988:62-63. |
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