Actions for Thermal Analysis of a Uranium Silicide Miniplate Irradiation Experiment [electronic resource].

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Thermal Analysis of a Uranium Silicide Miniplate Irradiation Experiment [electronic resource].

Published
Washington, D.C. : United States. Office of the Assistant Secretary for Nuclear Energy, 2009.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
Additional Creators
Idaho National Laboratory, United States. Office of the Assistant Secretary for Nuclear Energy, and United States. Department of Energy. Office of Scientific and Technical Information
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Summary
This paper outlines the thermal analysis for the irradiation of high density uranium-silicide (U3Si2 dispersed in an aluminum matrix and clad in aluminum) booster fuel for a Boosted Fast Flux Loop designed to provide fast neutron flux test capability in the ATR. The purpose of this experiment (designated as Gas Test Loop-1 [GTL-1]) is two-fold: (1) to assess the adequacy of the U3Si2/Al dispersion fuel and the aluminum alloy 6061 cladding, and (2) to verify stability of the fuel cladding boehmite pre-treatment at nominal power levels in the 430 to 615 W/cm2 (2.63 to 3.76 Btu/s•in2) range. The GTL-1 experiment relies on a difficult balance between achieving a high heat flux, yet keeping fuel centerline temperature below a specified maximum value throughout an entire operating cycle of the reactor. A detailed finite element model was constructed to calculate temperatures and heat flux levels and to reveal which experiment parameters place constraints on reactor operations. Analyses were performed to determine the bounding lobe power level at which the experiment could be safely irradiated, yet still provide meaningful data under nominal operating conditions. Then, simulations were conducted for nominal and bounding lobe power levels under steady-state and transient conditions with the experiment in the reactor. Reactivity changes due to a loss of commercial power with pump coast-down to emergency flow or a standard in-pile tube pump discharge break were evaluated. The time after shutdown for which the experiment can be adequately cooled by natural convection cooling was determined using a system thermal hydraulic model. An analysis was performed to establish the required in-reactor cooling time prior to removal of the experiment from the reactor. The inclusion of machining tolerances in the numerical model has a large effect on heat transfer.
Report Numbers
E 1.99:inl/con-08-14660
inl/con-08-14660
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Note
Published through SciTech Connect.
09/01/2009.
"inl/con-08-14660"
NURETH-13,Kanazawa, Japan,09/27/2009,10/02/2009.
Donna Post Guillen.
Funding Information
DE-AC07-05ID14517
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