Validation of the TWOPORFLOW code for the core analysis of liquid metal-cooled reactor with selected experiments

Abstract

TWOPORFLOW (TPF) is a thermal–hydraulic code that utilizes a porous medium approach in three-dimensional Cartesian coordinates. It can simulate the flow phenomenon in fuel assemblies or reactor cores by solving three conservation equations (single-phase) or six conservation equations in a two-fluid liquid–vapor approach. Many constitutive correlations e.g. pressure loss, interfacial friction, and heat transfer (wall, liquid–vapor interface), are included to close the system of equations. The code was recently extended to simulate pre-Critical Heat Flux (CHF) and CHF heat transfer relevant for Boiling Water Reactor (BWR). In this paper, the code's simulation capability for liquid metal-cooled reactors core is validated using three experimental data sets: pressure drop of the 19-pin rod in THEADES LBE loop, the BREST-type reactor benchmark problem, and the NACIE-UP Block Fuel Pin Simulator (BFPS) test. The results show that TPF can be used not only for square rod bundle arrangements but also for hexagonal arrangements; the code can reasonably predict the thermal–hydraulic behavior of liquid metal-cooled reactors under normal and off-normal conditions.