反应堆冷却剂环境对690合金传热管疲劳性能影响研究  

Effect of Reactor Coolant Environment on Fatigue Performance of Alloy 690 Steam Generator Tubes

作　　者：刘晓强[1] 徐雪莲[1] 谭季波[2] 王媛[2] 吴欣强[2] 郑宇礼[2] 孟凡江[1] 韩恩厚[2] 

机构地区：[1]上海核工程研究设计院,上海200233 [2]中国科学院金属研究所中国科学院核用材料与安全评价重点实验室,沈阳110016

出　　处：《中国腐蚀与防护学报》2015年第3期213-219,共7页Journal of Chinese Society For Corrosion and Protection

基　　金：大型先进压水堆核电重大专项支撑项目(2010ZX06004-18)资助

摘　　要：针对核电站蒸汽发生器690合金传热管,通过在室温空气、高温空气以及模拟压水堆高温高压水环境下的疲劳性能测试,研究了环境介质对690合金传热管疲劳寿命的影响,并考察了溶解氧和应变速率等的影响规律,探讨了高温高压水环境下690合金传热管的腐蚀疲劳机理。结果表明,690合金传热管具有足够的疲劳设计安全裕度,且压水堆冷却剂环境对690合金传热管的疲劳寿命影响不明显;溶解氧和应变速率对690合金传热管的腐蚀疲劳寿命的影响也不敏感。推测690合金传热管在高温高压水中的腐蚀疲劳过程主要由膜破裂滑移/溶解机制控制。The low cycle fatigue behavior of alloy 690 steam generator tubes in three different environments was investigated, i.e. in air at ambient temperature and 325 ℃ respectively as well as in simulated PWR primary water at 325 ℃. Meanwhile, the influence of dissolved oxygen and strain rate on fatigue life was considered. The results show that the design related with the fatigue life for the alloy 690 steam generator tube is very conservative, and the effect of PWR coolant environment is limited; while the corrosion fatigue life of the alloy 690 is not susceptible to the effect of the dissolve oxygen and strain rate under this test condition. Therefore, it is conferred that the corrosion fatigue behavior may be controlled by film rupture slip/dissolution mechanism.

关 键 词：腐蚀疲劳 690合金 传热管 疲劳设计 

分 类 号：TM623[电气工程—电力系统及自动化] TL353[核科学技术—核技术及应用]