The Stress Corrosion Cracking Behavior of Simulated Reactor Components Welded 304L Stainless Steel in Oxygenated and Chloride Ions Solution at Different Temperature
Welded 304L stainless steel is the weak link in pressurized water reactor (PWR) components, Stress corrosion cracking (SCC) of simulated reactor components welded 304L stainless steel in 5mg/L chloride ion and oxygenation atmosphere at different temperature was studied using slow strain rate testing (SSRT), the basic water chemistry is simulated in pressure water reactor(PWR) primary loop, high temperature and high pressure water containing boric and lithium ion.Results shown the SCC susceptibility of welded 304L at 285 ℃ is the lowest, and the SCC susceptibility of welded 304L increased with increasing temperature from 285℃ to 330℃.The mechanical properties of 270℃ sample declined a lot due to the welded porosity defects.The brittle fracture area in section increased with increasing the temperature, while the dimple area declined with the temperature.The oxygen content of brittle fracture areas is obviously higher than that of dimple areas, when the brittle fracture zone was coved with oxide particles, the oxygen content in this area is obviously higher than that in clean brittle fracture areas.The content of iron, chromium and nickel in dimple area is obviously higher than that in brittle fracture areas.
stress corrosion cracking (SCC) slow strain rate testing(SSRT) welded 304L stainless steel oxygenation atmosphere different temperature
Dequan Peng Shilin Hu Pingzhu Zhang Hui Wang
China Institute of Atomic Energy, P.O.Box 275-53, Bejing 102413, China
国内会议
沈阳
英文
204-215
2015-09-20（万方平台首次上网日期，不代表论文的发表时间）