Application of fully ceramic microencapsulated fuels in light water reactors

Conference · Sun Jul 01 04:00:00 EDT 2012

OSTI ID:22107763

Gentry, C.; George, N.; Maldonado, I. ^[1]; Godfrey, A.; Terrani, K.; Gehin, J. ^[2]

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1. Dept. of Nuclear Engineering, Univ. of Tennessee-Knoxville, Knoxville, TN 37996-2300 (United States)
2. Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22107763
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
CERAMICS
COMPARATIVE EVALUATIONS
DESIGN
ENRICHMENT
FISSILE MATERIALS
FUEL ASSEMBLIES
FUEL PARTICLES
FUEL PINS
NEPTUNIUM
NUCLEAR FUELS
NUCLEAR POWER PLANTS
PARAMETRIC ANALYSIS
PEAK LOAD
PLUTONIUM
PWR TYPE REACTORS
REACTOR SIMULATORS
THREE-DIMENSIONAL CALCULATIONS
URANIUM
URANIUM DIOXIDE
URANIUM NITRIDES