Coupled analysis of oxidation corrosion and heat transfer in lead-cooled fast reactors

Abstract

The coupled code LETHAC-Oxide is developed for analysis of thermal–hydraulic and safety characteristics in lead-cooled fast reactors, considering the impact of oxidation corrosion during prolonged operation. Based on experimental data from CORRIDA, Tsu-2M, and SM-1 facility, the oxidation model is well verified. The reactor concepts LESMOR and BREST-OD-300 are modeled, and the results show that the oxide layer significantly influences heat transfer, particularly at higher temperatures. A comparison between LESMOR and BREST-OD-300 demonstrates that a 95 °C difference in average system temperature will cause 14 times increase in oxide layer thickness and 7 times decrease in steam generator heat exchange capability. Conclusively, LESMOR forms a protective oxide film after a refueling cycle, offering structural material protection without major heat transfer impact. In contrast, BREST-OD-300 shows a substantial increase in cladding temperature and decrease in heat transfer capacity. This result underscores the necessity of oxygen control technology to mitigate risks associated with oxidation corrosion, providing valuable insights for optimal reactor performance and safety.