Experimental and numerical investigation on the flow and heat transfer characteristics of lead-bismuth eutectic through a helical single rod and four-rod bundle

The flow and heat transfer performance of nuclear fuel are critical factors influencing the safety and economic efficiency of nuclear reactor systems. This study investigates a novel helical fuel design for lead-bismuth eutectic-cooled reactors. Experiments are performed in three different test sections to evaluate flow resistance, convective heat transfer, transverse mixing, and flow distribution. Key experimental parameters, including temperature, pressure drop, and volumetric flow rate, are measured. The results demonstrate that, at a constant Reynolds number (Re) and constant Péclet number (Pe), the inlet temperature has negligible influence on both the resistance coefficient and Nusselt number (Nu). The flow resistance correlation is found to be consistent across fuel rods similar to helical cruciform geometry. Besides, transverse mixing experiments in the four-rod test section demonstrate that transverse mixing efficiency across subchannels is influenced by experimental parameters. Moreover, flow distribution results indicate that the flow distribution ratio (average = 1.89) between inside and outside channels remains independent of Re in turbulent regions. These experiments provide a comprehensive dataset for validating numerical models. Based on these data, computational fluid dynamics (CFD) simulations are performed. The predictions using the Realizable two-layer k−ε turbulence model show excellent agreement with experimental data.