CVTR FUEL CAPSULE IRRADIATIONS. Terminal Report

Technical Report · Wed Aug 01 04:00:00 EDT 1962

OSTI ID:4746695

Duncan, R N

Capsule irradiations of UO/sub 2/ fuel rods were performed to evaluate the effect of fuel rod power level on cladding dimensional changes and fission gas release. Two capsules were irradiated in the Westinghouse Test Reactor. One capsule contained three fuel rods with a 38 in. fuel length and was irradiated at peak fuel rod power levels of 17 to 19 kw/ft to a maximum fuel burnup of 3450 MWD/ TmU. The other capsule contained four fuel rods with 6 in. fuel lengths. Average fuel rod power levels of over 18 kw/ft were maintained during irradiation to 6250 MWD/TmU. The Zircaloy-2 fuel rods contained sintered UO/sup 2/ pellets, 0.430 in. in dia. 94 plus or minus 5% of theoretical density. He at 1 atm pressure was used to backfill all fuel rods except one of the 6 in. fuel rods which was initially filled with Ar. The temperature distribution within the UO/ sub 2/ fuel was estimated by post irradiation measurements of the radius at which various microstructural changes such as grain growth and columnar grain formation occurred. Comparison of the UO/sub 2/ microstructures from fuel rods with initial cold diametral gaps of 0.002, 0.006 and 0.012 in. indicated thnt the UO/ sub 2/ surface and center temperature increased with increasing gap. No evidence of UO/sub 2/ center melting was found. Diameter increases of up to 0.004 in. were measured on iuel rods with 0.002 inch initial diametral gaps. The deformation of the cladding is attributed to differential radial thermal expansion of the UO/sub 2/ fuel against the cladding. Estimates of the pellet-to- clad gap conductances indicnte that dilution of the He backfill gas in the pelletto-clad annulus occurs as the fission product gases, Xe and Kr, are released, substantially lowering the values obtained for the pellet-to-clad gap conductances. (auth)

• Research Organization: Westinghouse Electric Corp. Atomic Power Div., Pittsburgh
• NSA Number: NSA-17-006739
• OSTI ID: 4746695
• Report Number(s): CVNA-153
• Country of Publication: United States
• Language: English

Related Subjects

ARGON
BURNUP
CAPSULES
DEFORMATION
DISTRIBUTION
EXPANSION
FISSION PRODUCTS
FUEL CANS
FUELS
GRAIN SIZE
HELIUM
IRRADIATION
KRYPTON
MATERIALS TESTING
METALLOGRAPHY
METALS, CERAMICS, AND OTHER MATERIALS
RADIATION EFFECTS
REACTORS
RODS
SINTERED MATERIALS
TEMPERATURE
THERMAL CONDUCTIVITY
URANIUM DIOXIDE
WTR
XENON
ZIRCALOY