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Advanced High-Temperature Sodium-Cooled Thermal Reactors Using Less Than 10% Enriched UO2 Fuel

Gilles Youinou

Nuclear Science and Engineering / Volume 199 / Number 4 / April 2025 / Pages 613-630

Research Article / dx.doi.org/10.1080/00295639.2024.2381387

Received:May 3, 2024
Accepted:July 3, 2024
Published:March 11, 2025

This paper presents a 1200-MW(thermal) advanced sodium-cooled thermal reactor concept that uses online refueling of 3.5% to 9.95% enriched UO2 fuel pin bundles; uses either graphite or beryllium oxide (BeO) as a neutron moderator; reaches outlet temperatures of 650°C enabling a thermal efficiency of at least 45%; has a high specific power of 133 W/g U; has average power densities of 16.4 and 43.2 W/cm3 with graphite and BeO, respectively; reaches an average discharge burnup of 100 MWd/kg U; and generates 52% less spent fuel volume, 28% less fission products, and 47% to 64% less transuranics than a typical large pressurized water reactor for the same amount of electricity produced.