Sodium-cooled fast reactor (SFR) is the preferred technology of the generation-IV fast neutron reactor, and its core body mainly uses nuclear-grade 316 stainless steel. In order to prolong the design life of SFRs to 60 years and more, it is necessary to summarize and analyze the anti-corrosion effect of nuclear grade 316 stainless steel in high temperature sodium environment. The research on sodium corrosion of nuclear grade 316 stainless steel is mainly composed of several important factors, including the microstructure of stainless steel (ferrite layer, degradation layer, etc.), the trace chemical elements of stainless steel (Cr, Ni and Mo, etc) and liquid impurity elements in sodium (O, C and N, etc), carburization and mechanical properties of stainless steel, etc. Through summarizing and constructing the sodium corrosion rate equations of nuclear grade 316 stainless steel, the stainless steel loss of thickness can be predicted. By analyzing the effects of temperature, oxygen content in sodium and velocity of sodium on corrosion rate, the basis for establishing integrity evaluation standard of SFR core components with sodium corrosion is provided.

钠冷快堆（SFR）是第四代快中子反应堆的首选技术，其堆芯主体主要采用核级316不锈钢。为了将SFRs的设计寿命延长至60年以上，有必要对核级316不锈钢在高温钠环境下的防腐效果进行总结分析。核级316不锈钢钠腐蚀的研究主要由几个重要因素组成，包括不锈钢的显微组织（铁素体层、降解层等）、不锈钢微量化学元素（Cr、Ni和Mo，等）和液态杂质元素（O、C、N等），不锈钢的渗碳和力学性能等。通过对核级316不锈钢钠腐蚀速率方程的总结和构建，可以预测不锈钢的厚度损失。通过分析温度、钠中氧含量和钠速度对腐蚀速率的影响，为钠腐蚀SFR核心部件完整性评价标准的建立提供依据。