Published 2009 | Version v1
Miscellaneous

Thermal analysis of a uranium silicide miniplate irradiation experiment

  • 1. Idaho National Laboratory, Idaho Falls, ID (United States)

Description

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 Advanced Test Reactor. 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. The inclusion of machining tolerances in the numerical model has a large effect on heat transfer. 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 accident 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 elapsed time after reactor shutdown when the emergency pumps can be turned off and the experiment can be adequately cooled by natural convection cooling was determined using a system thermal hydraulic model. A canal draining scenario was investigated to determine the necessary additional cooling time in water prior to experiment handling and removal. (author)

Part of:
NURETH-13: Proceedings of the 13th international topical meeting on nuclear reactor thermal hydraulics

Additional details

Publishing Information

Imprint Title
NURETH-13: Proceedings of the 13th international topical meeting on nuclear reactor thermal hydraulics
Imprint Pagination
[4617 p.]
Journal Page Range
[12 p.]

Conference

Title
13. international topical meeting on nuclear reactor thermal hydraulics
Acronym
NURETH-13
Dates
27 Sep - 2 Oct 2009
Place
Kanazawa, Ishikawa (Japan)

Optional Information

Notes
Available as CD-ROM Data in PDF format, Folder Name: FullPaper, Paper ID: N13P1005.pdf; 14 refs., 12 figs., 5 tabs.