Thermal hydraulic investigations on lead-bismuth cooled fuel assemblies with ducts

Abstract

Compared to sodium cooled fast reactors, heavy liquid metal (lead or lead-bismuth) cooled fast reactors use relatively open fuel lattices without wire spacers. In the present work, a subchannel analysis was performed for lead-bismuth cooled fuel assemblies with ducts and thermal hydraulic characteristics were investigated with emphasis on turbulent mixing between subchannels and interassembly heat transfer. The existing SLTHEN code, which had been originally developed for sodium cooled fast reactors, was modified and applied to typical lead-bismuth cooled fuel assemblies with ducts. The analysis started from a single fuel assembly to assess the effect of turbulent mixing and heat transfer between subchannels and was extended to 7 fuel assemblies to investigate interassembly heat transfer. The results of the modified SLTHEN show that the maximum cladding temperature is not largely affected by turbulent mixing and interassembly heat transfer and the simplified analysis such as single assembly consideration is very useful for thermal hydraulic analysis and design of lead-bismuth cooled fuel assemblies under conceptual design stages.