CFD-DEM analysis of realistically heated pebble bed geometry
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Date
2019-08-18
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American Nuclear Society
Abstract
Pebble-bed, molten salt cooled reactors (PB-FHR) are a prominent reactor design and stepping stone between current generation reactors and Gen IV liquid fuelled molten salt reactor designs. They utilize spherical fuel elements as used in previous gas cooled pebble-bed reactors, while the coolant is a molten salt capable of much higher temperatures than conventional light-water reactors with the advantage of remaining at atmospheric pressure. This study implements a realistic power distribution, determined from a neutronic analysis, on to a packed bed of fuel pebbles. Utilizing coupled computational fluid dynamics (CFD) and discrete element methods (DEM) it is possible to observe the fluid and pebble flow and heat transfer characteristics in such PB-FHR systems. Previous work by the authors considered an isothermal case and a uniform power distribution across the pebble bed. The inclusion of a realistic power distribution allows for realization of accurate temperature profiles throughout the coolant fluid and fuel pebbles. The coupled nature of the method includes the pebble-pebble, pebble-fluid, and fluid-pebble exchange of momentum and energy. Current results are for a PB-FHR geometry at steady state adapted from a pebble recirculation experiment. The pebbles within the core are all assumed to be fuel pebbles, with no dummy graphite or absorber pebbles present. Analysis has demonstrated the strongly non-uniform power distribution present in the fuel pebbles in such a reactor. This non-uniformity is exacerbated at the near wall region due to the moderator effect where fuel pebbles have a marked increase in power. © 2019 The Authors.
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Keywords
Pebble bed reactors, Design, Gas cooled reactors, Atmospheric pressure, Heat transfer, Molten Salt cooled reactors
Citation
Mardus-Hall, R., Yeoh, G., & Ho, M. (2019). CFD-DEM analysis of realistically heated pebble bed geometry. Paper presented to the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, (NURETH 2019), 18-23 Aug 2019, Portland, OR. In Proceedings of the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH 2019). (pp. 1726-1739). Retrieved from: https://www.ans.org/pubs/proceedings/article-46184/