Research Progress in Key Thermal-hydraulic Issue of Sodium-cooled Fast Reactor

Abstract

Sodium-cooled fast reactors are considered the preferred reactor type for fourth generation nuclear energy systems due to their high thermal efficiency, high fuel utilization efficiency, inherent safety, and unique advantages of achieving a closed fuel cycle to reduce long-lived radioactive waste. They are also a hot research topic in the international nuclear energy field. An overview of the development history and future directions of sodium-cooled fast reactors and thermal-hydraulic fields on an international scale was provides, and the main problems currently faced in development of sodium-cooled fast reactor thermal-hydraulic fields were pointed out. Based on the current development status and challenges in China, an in-depth analysis of three key core aspects in the field of thermal-hydraulic research was conducted, including the flow and heat transfer characteristics of coolant sodium, core thermal-hydraulic analysis methods, and design and capability verification of removing core decay heat through natural circulation. The research achievements on an international scale was summarized, the technological challenges encountered and the future development trends of research were analyzed, and the current research status and future research directions in these fields in China were explored. These analyses and summaries aim to provide guidance for the technical challenges of improving safety and economy in China’s sodium-cooled fast reactor technology, and also provide important reference for the thermal-hydraulic design and optimization of future sodium-cooled fast reactors, in order to promote the further development of China’s sodium-cooled fast reactor technology.

Keywords

• sodium-cooled fast reactor
• thermal-hydraulics
• natural circulation
• inter wrapper flow
• inherent safety
• passive safety
• inherent thermal-hydraulic safety
• turbulence model