Development of a test task for coupled neutronic and thermal-hydraulic analysis of LFR fuel cell

Abstract

In the absence of operational experience with Generation IV reactors and limited experimental data, developing test tasks is crucial for ensuring the safety and reliability of novel reactor designs. This study presents a test task for coupled neutronic and thermal-hydraulic analysis of a lead-cooled fast reactor (LFR) fuel unit cell in the BREST-OD-300 core. Neutronic simulations using the Monte Carlo-based MCU-FR code identified an optimal axial discretization of 44 layers to reduce uncertainties in power density distribution. The resulting axial power density profiles were used in CFD thermal-hydraulic modeling with LOGOS software to assess the impact of different turbulent Prandtl number correlations on axial temperature profiles. The study shows that the choice of turbulent Prandtl number introduces up to 1% model-uncertainty in temperature predictions due to differences in turbulence modeling assumptions. In addition, the application of fully coupled neutronic and thermal-hydraulic feedback results in a deviation of above 2% in average fuel temperature compared to the non-coupled solution. This test task is aimed at enhancing the accuracy of reactor simulations and safety assessments while contributing to the development of a benchmark suite for LFR reactors, supporting cross-verification of computational tools and methods.