Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Review on sodium corrosion evolution of nuclear-grade 316 stainless steel for sodium-cooled fast reactor applications

  • Dai, Yaonan (Hubei Provincial Engineering Technology Research Center of Green Chemical Equipment, Wuhan Institute of Technology) ;
  • Zheng, Xiaotao (Hubei Provincial Engineering Technology Research Center of Green Chemical Equipment, Wuhan Institute of Technology) ;
  • Ding, Peishan (Hubei Provincial Engineering Technology Research Center of Green Chemical Equipment, Wuhan Institute of Technology)
  • Received : 2020.12.16
  • Accepted : 2021.05.17
  • Published : 2021.11.25
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Abstract

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.

Keywords

Acknowledgement

This work was financially supported by the National Natural Science Foundation of China (51975424).

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