Corrosion Behavior and Mechanism of Irradiated 304 Nuclear Grade Stainless Steel in High-Temperature Water  

作　　者：Ping Deng En-Hou Han Qunjia Peng Chen Sun 

机构地区：[1]CAS Key Laboratory of Nuclear Materials and Safety Assessment,Institute of Metal Research,Chinese Academy of Sciences,Shenyang 110016,China [2]Nuclear Power Institute of China,Chengdu 610213,China [3]Suzhou Nuclear Power Research Institute,Suzhou 215004,China [4]State Power Investment Corporation Research Institute,Beijing 102209,China

出　　处：《Acta Metallurgica Sinica(English Letters)》2021年第2期174-186,共13页金属学报（英文版）

基　　金：financially supported by the International Science&Technology Cooperation Program of China(No.2014DFA50800);partly supported by the Essential Research Fundby SNPTC(No.2015 SN010-007)。

摘　　要：Corrosion behavior and mechanism of irradiated 304 nuclear grade stainless steel were studied in simulated pressurized water reactor primary water.The microstructure of the oxide formed on the steel irradiated to different doses over an exposure period range of 25–1500 h was analyzed and compared.It was found that the general and intergranular corrosion rates of the steel were increased with irradiation dose,in correspondence with an evolution of the general oxide and the oxide formed at the grain boundary.Correlation of the oxide evolution with the corrosion kinetics and mechanism has been discussed in detail.

关 键 词：Stainless steel IRRADIATION High-temperature corrosion Intergranular corrosion 

分 类 号：TG172.1[金属学及工艺—金属表面处理]