Thermal‐Hydraulic Characterization Preliminary Analysis of Pool‐Type Natural Circulation With Liquid Lead

Lead‐cooled fast reactors are an important potential reactor type in the fourth‐generation nuclear energy system. And the clarity of the thermal‐hydraulic characteristics of pool‐type natural circulation is crucial for the design of lead‐based fast reactors. This study focuses on liquid lead and conducts experimental research on the characteristics of pool‐type natural circulation. Based on the pool‐type structure, megawatt‐scale natural circulation tests were carried out, achieving stable natural circulation at a final input power of 503 kW, which demonstrates the feasibility of the natural circulation design in a liquid lead pool. Meanwhile, experimental modeling using the redeveloped system code RELAP5/MOD4.0 was performed, and the verification analysis demonstrates that the redeveloped code can accurately simulate the experiment. The comparison shows that the experimental results and calculations exhibit the same trend, and the errors in core inlet temperature, outlet temperature, and inlet pressure are 2.5%, 2.2%, and 5.8%, respectively. This study experimentally verified the feasibility of the pool‐type natural circulation design and further redeveloped a system analysis code suitable for numerical simulation. The verification with experiments also demonstrated the accuracy of the new code for the simulation of lead‐cooled fast reactors. Considering the scarcity of research reports on natural circulation in pure liquid lead pools, the research outcomes can provide support for the design of pool‐type natural circulation reactors using liquid lead.