Advanced Test Reactor LEU Fuel Conversion Feasibility Study (2006 Annual Report)

Technical Report · Fri Dec 01 00:00:00 EST 2006

DOI:https://doi.org/10.2172/911863· OSTI ID:911863

Chang, Gray S; Ambrosek, Richard G; Lillo, Misti A

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of these operating parameters, and the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat profile between the HEU and LEU core can be minimized for operation at 125 EFPD with a total core power of 115 MW. The depletion methodology, Monte-Carlo coupled with ORIGEN2 (MCWO), was used to calculate K-eff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar in shape to the reference ATR HEU case. The LEU core conversion feasibility study can also be used to optimize the U-235 content of each fuel plate, so that the relative radial fission heat flux profile is bounded by the reference ATR HEU case. The detailed radial, axial, and azimuthal heat flux profiles of the HEU and optimized LEU cases have been investigated. However, to demonstrate that the LEU core fuel cycle performance can meet the UFSAR safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders (OSCCs), safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.

• Research Organization: Idaho National Laboratory (INL)
• Sponsoring Organization: DOE - NE
• DOE Contract Number: AC07-99ID13727
• OSTI ID: 911863
• Report Number(s): INL/EXT-06-11887
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
BURNUP
depletion analysis
DOPPLER COEFFICIENT
FISSION
FUEL CYCLE
FUEL PLATES
HEAT FLUX
HIGHLY ENRICHED URANIUM
highly enriched uranium (HEU)
low enriched uranium (LEU)
MCWO
MODIFICATIONS
NEUTRON FLUX
PERFORMANCE
POWER DENSITY
RESEARCH REACTORS
SAFETY
SCRAM RODS
SHAPE
SHUTDOWN
TEST REACTORS
THERMAL NEUTRONS
URANIUM