Abstract
In this paper, an Unprotected Loss of Flow (ULOF) assessment has been performed on the European Sodium Fast Reactor developed in the ESFR-SMART EU project. To conduct the analysis, a simplified 42 channel thermal–hydraulic model has been established in TRACE system code, using a point kinetics model accounting for various reactivity feedback effects. The assessment reveals the core behavior of a commercial size, 3600 MWth, sodium fast reactor using a state-of-the-art low void effect reactor core design. The study focuses on the sodium boiling phenomenon and sodium reactivity feedback effect evolution during the accident with the reference subassembly (SA) design. Following this analysis, a study has been performed with a modified SA design. The boiling progression and phenomenology within the reference and the modified core have been compared, and the impact of the SA modification was described.
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