Large-leak sodium–water reaction accident analysis of sodium-cooled fast reactor

Abstract A sodium-cooled fast reactor (SFR) is one of Generation IV nuclear systems. The once-through steam generator is important for heat transfer and acts as a barrier to separate the sodium and water. If the heat transfer tube breaks, a sodium–water reaction (SWR) occurs, and the integrity of the secondary loop may be destroyed. Thus, a dynamic model to analyze the SWR must be developed, especially for a large-leak SWR. A large-leak SWR analysis model was developed for an SFR, consisting of four modules: water/steam leakage rate, hydrogen bubble growth, pressure wave propagation, and SWR protection system action. A large-leak SWR accident was simulated using this model, and the process was investigated. Sensitivity analysis was performed for the critical parameters, and the parameters with a key influence on the large-leak SWR were determined. The results are significant for the design of the steam generator and the SWR protection system.