690合金在核电一回路水环境的应力腐蚀行为  

Stress corrosion behavior of alloy 690in primary water of a nuclear power plant

作　　者：李成涛 宋利君 任爱 刘飞华 李岩 李晓刚[2] 程学群[2] 

机构地区：[1]苏州热工研究院电站寿命管理研究中心,江苏苏州215004 [2]北京科技大学新材料技术研究院,北京100083

出　　处：《哈尔滨工程大学学报》2013年第11期1465-1470,共6页Journal of Harbin Engineering University

基　　金：国家863计划资助项目(2012AA050901);国家自然科学基金资助项目(U1260201)

摘　　要：针对690合金应力腐蚀问题,在核电一回路模拟环境中,采用恒载荷和慢应变速率拉伸实验方法,研究了690合金的应力腐蚀开裂行为,并通过扫描电镜、能谱分析和X射线衍射方法对表面形貌、截面形貌和断口形貌进行观察与分析.扫描电镜结果表明,样品表面被颗粒状腐蚀产物所覆盖,成分主要为Fe、Ni、Cr的尖晶石氧化物;表面状态对表面腐蚀形貌影响较大;690合金恒载荷试样经历2 100 h腐蚀试验后未发生应力腐蚀开裂;拉伸试样表现出低的应力腐蚀敏感性,断口形貌为韧性断口和脆性断口的混合模式.The stress corrosion cracking （SCC） behavior of alloy 690 in primary water was investigated using the static load test and slow strain rate test. The surface, cross section and fracture morphology were analyzed by using scanning electron microscopy （SEM）, energy dispersive spectrometer （EDS） and X-ray diffraction. The results of the SEM showed that the surface is uniformly covered by corrosive oxide particles. The corrosion product mainly consists of spinel oxides of Fe, Ni and Cr. The surface condition has great effect on the corro- sion morphology. After a 2100 hours corrosion test in primary water, no significant SCC was observed in the static load specimens of alloy 690. The slow strain specimens have low susceptibility to SCC in primary water. The fracture morphology was the mixed mode of the ductile fracture and the brittle fracture.

关 键 词：690合金 应力腐蚀 一回路 核电 腐蚀 

分 类 号：TG172[金属学及工艺—金属表面处理]