压水堆燃料包壳新锆合金的发展(英文)  

Development of New Zirconium Alloys for PWR Fuel Rod Claddings

作　　者：赵文金[1] 周邦新[1] 苗志[1] 李聪[1] 蒋宏曼[1] 于晓卫[1] 蒋有荣[1] 黄强[1] 苟渊[1] 黄德诚 

机构地区：[1]中国核动力研究设计院核燃料及材料国家级重点实验室,成都610041

出　　处：《中国核科技报告》2000年第1期608-618,共11页China Nuclear Science and Technology Report

摘　　要：通过分析优化,确定了 Zr-Sn-Nb 系两种先进锆合金。研究了新锆合金的加工工艺、显微组织及腐蚀性能间的关系。通过在高压釜360 ℃、含 LiOH 的高温水以及在 500 ℃过热蒸气中的实验,评价了热处理和化学成分对腐蚀性能的影响。结果表明,两种新锆合金(N18,N36)在含 LiOH 的高温水中的腐蚀性能大大优于 Zr-4 合金,尤其是在 500 ℃过热蒸气中表现出极好的耐疖状腐蚀性能。电子显微镜分析证明含细小和均匀分布β-Nb 和 Zr(Fe,Nb)粒子的试样具有最好的堆外性能。此外,新锆合金的力学性能也优于 Zr-4合金,所有试样的吸氢量与氧化膜厚度成线性关系。An advanced zirconium alloys containing Sn, Nb, Fe and Cr have been developed. The relationships between manufacturing, microstructure and corrosion performance for the new alloys have been studied. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in lithiated water at 633 K and high-temperature steam at 773 K. Analytical electron microscopy demonstrated that the best out-of-pile corrosion performance was obtained for microstructure containing a fine and uniform distribution of β-Nb and Zr (Fe, Nb)2 particles. Autoclave testing in LiOH solution indicated that two kinds of alloys (N18, N36) showed the lower corrosion rate than the reference Zr-4 tested, and especially, the corrosion resistance in superheated steam at 773 K was much better. Moreover, the mechanical properties were superior to Zr-4. And the hydrogen absorption data for all of alloys from corrosion reactions under various corrosion conditions showed a linear increase with the oxide thickness.

关 键 词：新锆合金 燃料包壳 压水堆 动力研究 核燃料 英文 重点实验室 设计院 LiOH MATE 

分 类 号：TL[核科学技术]