Development and Validation of a Sub-Channel Analysis Code for Lead-Based Wire-Wrapped Bundles

Abstract

Due to the intrinsic safety and good neutron economy, lead-based fast reactor (LFR) has been considered as one of the most promising concepts for Generation IV nuclear reactors. With a high calculating efficiency, sub-channel analysis is frequently used to obtain the distribution of coolant and cladding temperature. In this study, a high-resolution sub-channel analysis code, STAR-LFR, is developed for lead-based wire-wrapped bundles. A coolant heat conduction model, a distributed resistance model and a hot spot factor were applied to improve the prediction accuracy. Crossflow prediction was validated by high-fidelity large eddy simulation of a 7–pin bundle with wire spacers. Heat transfer prediction was validated by an experiment with a 19-pin wire-wrapped bundle. Overall, great ability on flow and heat transfer prediction of lead-based wire-wrapped bundles was well proved. This work provided a reliable tool that could support the thermal-hydraulic analysis as well as the design of LFR.

Keywords: LFR, sub-channel analysis, code validation