Subchannel analysis of fuel assemblies of a lead–alloy cooled fast reactor

• Flow distribution model of natural circulation systems has been introduced.
• Subchannel analysis of fuel assemblies of a natural circulation lead–alloy cooled fast reactor was conducted.
• Subchannel analysis results verified the thermal hydraulic design of a natural circulation lead–alloy cooled reactors.

Lead–alloy cooled fast reactor is one of the six Gen-IV reactors. It has many attractive features such as excellent natural circulation performance, better shielding against gamma rays or energetic neutrons and potentially reduced capital costs. A natural circulation lead–alloy cooled fast reactor with 10 MWth is under design in China (hereafter called LFR-10MW). Fuel assemblies thermal hydraulic analysis is of vital importance for a successful design. A subchannel analysis code with flow distribution model was used to carry out the thermal hydraulic analysis. This work briefly gave the thermal-hydraulic design for the LFR-10MW and analyzed the thermal-hydraulic characteristics under steady-state condition using the subchannel analysis code. Whole core analysis was performed to locate the hottest fuel assembly using the code. The hottest fuel assembly was analyzed to obtain the cladding temperature, fuel temperature and coolant velocity. The maximum cladding temperature, the maximum fuel center temperature and the maximum coolant velocity are all below the design constraints. These results imply that the thermal-hydraulic design of LFR-10MW is feasible.