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

• 作 者: Ping Deng En-Hou Han Qunjia Peng Chen Sun
• 机 构: CAS Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research, Chinese
• 出 处: 金属学报（英文版）
• 发表日期: 2021年2期
• 关键词: Stainless steel Irradiation High-temperature corrosion Intergranular corrosion

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.