核电厂蒸汽发生器管材800与690合金应力腐蚀性能  

Stress Corrosion Behavior of Steam Generator Material Alloy 800 and Alloy 690

作　　者：吴恋恋 张乐福[1] 潘向烽 李明星[1] 

机构地区：[1]上海交通大学核科学与工程学院,上海200240

出　　处：《腐蚀与防护》2014年第1期47-51,共5页Corrosion & Protection

摘　　要：采用慢应变速率拉伸方法,研究了模拟PWR一回路水环境中国产进口蒸汽发生器管材800合金与690合金的拉伸性能;并辅以SEM断口形貌观察,分析了800合金与690合金的应力腐蚀敏感性。结果表明,国产管800合金具有一定应力腐蚀开裂倾向,690合金具有很好的抗应力腐蚀性能。SSRT （Slow Strain Rate Test） was used to study and compare the tensile properties of domestic and imported steam generator materials alloy 800 and alloy 690 in simulated PWR primary loop water environment. Together with SEM fracphotographs, SSRT were used to study the SCC resistance and behavior of alloy 800 and alloy 690. The results showed that alloy 800 was susceptible to SCC in PWR primary water, while alloy 690 exhibited very good corrosion resistance.

关 键 词：PWR-回路水环境 慢应变速率试验 拉伸性能 应力腐蚀开裂倾向 

分 类 号：TG174[金属学及工艺—金属表面处理]