The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000°C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolants that are being evaluated have melting points between 350 and 500°C, values that imply minimum refueling temperatures between 400 and 550°C. At operating conditions, the liquid salts are transparent and have physical properties similar to those of water. A series of refueling studies have been initiated to (1) confirm the viability of refueling, (2) define methods for safe rapid refueling, and (3) aid the selection of the preferred AHTR design. Three reactor cores with different fuel element designs (prismatic, pebble bed, and pin-type fuel assembly) are being evaluated. Each is a liquid-salt-cooled variant of a graphite-moderated high-temperature reactor. The refueling studies examined applicable refueling experience from high-temperature reactors (similar fuel element designs) and sodium-cooled fast reactors (similar plant design with liquid coolant, high temperatures, and low pressures). The findings indicate that refueling is viable, and several approaches have been identified. The study results are described in this paper.
- Nuclear Engineering Division
Refueling Liquid-Salt-Cooled Very High-Temperature Reactors
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Forsberg, CW, Peterson, PF, Cahalan, JE, Enneking, JA, & MacDonald, P. "Refueling Liquid-Salt-Cooled Very High-Temperature Reactors." Proceedings of the 14th International Conference on Nuclear Engineering. Volume 3: Structural Integrity; Nuclear Engineering Advances; Next Generation Systems; Near Term Deployment and Promotion of Nuclear Energy. Miami, Florida, USA. July 17–20, 2006. pp. 603-612. ASME. https://doi.org/10.1115/ICONE14-89471
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