Neutronic Design and Fuel Cycle Analysis of a Commercial-Scale Fluoride Salt-cooled High-Temperature Reactor (FHR)

All dates for this event occur in the past.

Scott Lab E435
201 W. 19th Avenue Columbus, OH 43210
Columbus, OH 43210
United States

Speaker: Zuolong Zhu

Abstract:
The fluoride salt-cooled high-temperature reactor (FHR) is a new reactor concept, which combines the fuel designed for high temperature gas-cooled reactors (HTGRs) and the liquid salt coolant developed for molten salt reactors (MSRs), creating a reactor with high output temperatures and the ability to operate at atmospheric pressures, which makes it a leading candidate for the next generation of nuclear power plants. This study began by exploring and characterizing the available design space for a commercial-scale FHR. In order to determine which configurations offered acceptable neutronic performance and met all the design goals, a large number of detailed core parameters were evaluated, such as assembly size, fuel channel pitch, reflector layers, composition of burnable poison, etc. Several core designs with different power levels were developed to meet the needs of a variety of markets. Among the design options, the 165 MWth small core design’s fuel cycle was evaluated in detail (cycle length, fuel burnup, power distribution, temperature coefficient, etc.). It was concluded that erbium-167 in burnable poison significantly contributed to the improvement of the moderator temperature coefficient (MTC). Moreover, the multiplication factor does not have a linear relationship with moderator temperature but shows different degrees of curvature due to thermal scattering law. The magnitude of the temperature perturbation should be carefully selected when calculating the MTC. This phase of work provides an acceptable reference for applying design alternatives to improve feasible designs in the next stage. In the second stage of the present work, a novel core with the movable moderator was proposed based on the prismatic “fuel inside radial moderator” (FIRM) assembly design to further improve fuel cycle economics. The FIRM’s fuel-bearing region is designed to be physically separable from the central fuel cluster region. It is therefore possible to move the central fuel cluster or the fuel-bearing region independently, which is required by the movable moderator design. Furthermore, only the central fuel region would be removed during the refueling, the amount of graphite in the spent nuclear fuel can be reduced. The movable moderator was demonstrated as an innovative concept that not only incorporates the advantages of FIRM, but also improves all aspects of the neutronic performance of the reference FHR design. As an example, movable moderator can effectively extend the fuel cycle length by 45 days for an 18-month fuel cycle. In addition, various design alternatives, including coolant, moderator, and fuel form, were tested in this study. Based on the analyses, each design alternative has its own specific advantages. However, the original design, which with Li2BeF4 (FLiBe) as the coolant, graphite as the moderator, and tri-structural isotropic (TRISO) as the fuel, still performs the best with regard to neutronic performance.

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Committee Members
Associate Professor Dean Wang
Professor Tunc Aldemir
Associate Professor Marat Khafizov
Assistant Professor Richard Vasques

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