Thermal-Hydraulic Safety Analysis of Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100 Core Based on Porous Medium Approach

Abstract

In this study, a thermal-hydraulic analysis is carried out for the core of natural circulation lead-cooled fast reactor SNCLFR-100. Steady-state and transient analysis are performed with porous medium approach-based code TWOPORFLOW. In the steady-state analysis, mass flow distribution and temperature distributions of the assemblies are analyzed in assembly-wise mode. At the same time, the hottest assembly is analyzed in pin level and the safety performance is investigated. In the transient analysis, a typical Design Extension Conditions unprotected over-power transient is simulated with one-way coupling method.