核级锆及锆合金腐蚀性能研究现状
Study Status of Corrosion Properties of Zirconium and Zircaloy in Reactor
刘 鹏1, 杜忠泽1, 马林生2, 王快社1
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作者单位:1. 西安建筑科技大学 冶金工程学院, 陕西 西安 710055; 2. 国核宝钛锆业股份有限公司, 陕西 宝鸡 721000
中文关键字:锆合金; 腐蚀性能; 合金元素; 水化学; 热处理制度; 表面处理
英文关键字:zircaloy; corrosion properties; alloying element; reactor water chemistry; heat treatment; surface treatment
中文摘要:通过对已有成果的总结,对锆合金的工作环境做了简单的介绍,主要概括了锆合金的腐蚀性能。根据现有的数据,从添加合金元素、反应堆水化学、热处理制度及表面处理等方面对锆合金的腐蚀性能做以介绍,并对锆合金的发展提出建议。
英文摘要:Based on the summary of previous research, the working class environment of zirconium alloy was simply introduced. The corrosion properties of zircaloy were summarized mainly. According to the results of research, the corrosion properties of zircaloy from addition alloying element, reactor water chemistry, heat treatment regime and surface treatment were introduced. Finally some suggestions were proposed.