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Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

Abstract

This dissertation intends to examine basic materials properties, identify optimized fabrication techniques, model behavior under relevant environments, and experimentally quantify kinetic phenomena associated with hydride nuclear fuels. Hydride fuels have been examined extensively for application in light water reactors (LWR) from the neutronics and thermal hydraulic standpoints, the benefits of this fuel have been underscored through such studies. This manuscript provides the background for understanding materials aspects of hydride fuel incorporation in LWR environments.

The proposed LWR hydride fuel concept consists of uranium-zirconium hydride pellets clad in Zircaloy and bonded with a lead-bismuth alloy. The fuel material consists of metallic uranium particles dispersed in a zirconium hydride matrix, although thorium and/or other minor actinide hydride matrices could be utilized. The eutectic lead-bismuth alloy is liquid during reactor operating temperatures and replaces the conventionally-used helium gas in the fuel-cladding gap, thereby providing a thermal conductivity increase of two orders of magnitude. Initially uranium-thorium-zirconium hydrides were fabricated and extensively characterized. This provided detailed insight into fuel properties and the influence of fabrication methodology. A modeling approach was undertaken to examine hydride fuel behavior under steady-state and transient-power conditions in a typical LWR. This study outlined the operating parameters and fuel-response characteristics under various reactor operating conditions that support the feasibility of hydride fuel incorporation into LWRs. The kinetics of hydrogen release from the fuel, associated with one of the most severe accident scenarios, was investigated in detail. Mechanisms were identified for hydrogen desorption from and adsorption on zirconium hydride and the rates associated with each process were quantified. Hydrogen diffusivity in the thorium-zirconium hydride matrix, which is one of the critical parameters affecting fabrication and in-reactor fuel behavior, was experimentally determined by the means of incoherent quasielastic neutron scattering. Finally experiments were conducted to examine compatibility of hydride fuel with Zircaloy cladding when bonded by liquid-metal. A thin oxide grown on the surface of the cladding coupled with liquid metal was tentatively identified as adequate to limit hydrogen transport form the fuel to the cladding. Recognizing the necessity of a shift from laboratory scale experiments to more relevant fuel-operating environments, an irradiation experiment was conceived to examine the liquid-metal-bonded LWR hydride fuel concept.

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