Research Progress on Corrosion of Nuclear-grade 316 Stainless Steel in High-temperature Liquid Sodium
DAI Yaonan, PAN Lingfeng, XU Yiheng, DING Peishan, ZHENG Xiaotao*
State Key Laboratory of Green and Efficient Development of Phosphorus Resources, School of Mechanical and Electrical Engineering, Wuhan Institute of Technology, Wuhan 430205, China
Abstract: As a preferred technology for Generation IV fast neutron reactors, the sodium-cooled fast reactor (SFR) primarily utilizes nuclear-grade 316 stainless steel for its core critical components. To achieve the ultra-long design life target of over 60 years for SFRs, in-depth investigations into the corrosion behavior mechanisms of nuclear-grade 316 stainless steel in high-temperature liquid sodium environments hold significant engineering importance. This summary systematically reviews the corrosion kinetic mechanisms of nuclear-grade 316 stainless steel in high-temperature liquid sodium, elucidating the influences of evolutionary behavior of the passive film and elemental segregation characteristics on the corrosion progression. Based on these findings, it quantitatively characterizes the degradation laws of mechanical properties under long-term service conditions, with a particular focus on summarizing the microscopic mechanisms of creep-fatigue interaction damage induced by high-temperature liquid sodium environments in the material, as well as developing a multi-factor coupled corrosion rate prediction model. In response to the limitations of existing experimental methods, this paper proposes a technical scheme for constructing an equivalent accelerated sodium corrosion experimental system. By coordinating the regulation of multi-parameters (temperature, time, stress), this system aims to achieve equivalent simulation of the accelerated degradation process. The research outcomes are expected to provide theoretical and data-driven foundations for establishing a so-dium corrosion evaluation standard system tailored to SFR core components.
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