Fast Reactor Physics, Liquid Metal Cooling, and Supercritical Carbon Dioxide Power Cycles
This community investigates the design, thermal-hydraulics, and materials compatibility of fast neutron reactors, with a specific focus on liquid lead-bismuth eutectic and sodium coolants, alongside supercritical carbon dioxide power cycles for high-temperature energy systems.
The research is dominated by neutron transport calculations, cross-section data validation, and Monte Carlo simulations for reactor core design. A substantial portion of the work addresses the corrosion and oxidation of structural steels in liquid lead-bismuth eutectic and sodium environments, as well as heat transfer optimization in fuel assemblies. Concurrently, the community examines supercritical carbon dioxide Brayton cycles, including their integration with nuclear heat sources and compressed air energy storage systems. These studies frequently involve computational fluid dynamics, thermal-hydraulic analysis, and the evaluation of nuclear data libraries such as JEFF and ENDF/B to support reactor safety and performance modeling.
The community represents 9.7% of all tracked research on plutonium and 4.2% of uranium research, with 3,566 and 4,719 papers respectively. It also accounts for 2.2% of bismuth research (2,019 papers) and 4.2% of thorium research (1,648 papers).
The community comprises 44,150 papers, published primarily in the Annals of Nuclear Energy, Journal of Nuclear Science and Technology, and Nuclear Science and Engineering.
Recent work continues to focus on the corrosion behavior of high-entropy alloys and ferritic steels in liquid lead-bismuth eutectic, the thermal stratification modeling of sodium fast reactors, and the integration of supercritical carbon dioxide energy storage with coal-fired power units.