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Thermal-hydraulic modeling of porous bed reactors

Conference · · Trans. Am. Nucl. Soc.; (United States)
OSTI ID:7100605

Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 ..mu..m in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures.

Research Organization:
Brookhaven National Lab., Upton, NY (USA)
OSTI ID:
7100605
Report Number(s):
CONF-8711195-
Journal Information:
Trans. Am. Nucl. Soc.; (United States), Journal Name: Trans. Am. Nucl. Soc.; (United States) Vol. 55; ISSN TANSA
Country of Publication:
United States
Language:
English