Chemical interactions between UO/sub 2/ and Zircaloy-4 from 1000 to 2000/sup 0/C
The chemical interaction between solid and liquid Zircaloy-4 and solid UO/sub 2/ was examined in the temperature region 1000 to 2000/sup 0/C in argon. The solid/ solid reaction experiments were performed with short light water reactor fuel rod sections with an external pressure of 1 to 80 bar. The annealing times varied between 60 and 9000s. The reaction experiments with liquid Zircaloy were performed in UO/sub 2/ crucibles between 1800 and 2000/sup 0/C. In addition, the wetting behavior between liquid Zircaloy and UO/sub 2/ was also examined. The extent of the chemical interaction below the melting point of Zircaloy depends decisively on the solid/solid contact between fuel and cladding. If good contact exists, Zircaloy reduces UO/sub 2/ to form oxygenstabilized ..cap alpha..-Zr(O) and metallic uranium. The uranium reacts with zirconium to form a (U,Zr) alloy, which lies between two ..cap alpha..-Zr(O) layers. The UO/sub 2//Zircaloy-4 reaction obeys a parabolic rate law. The rate-determining step in the reaction is the diffusion of oxygen into Zircaloy. The growth of the different reaction zones can be represented in an Arrhenius diagram. The extent of the reaction between liquid Zircaloy and UO/sub 2/ depends on the wetting behavior. A Zircaloy melt rich in oxygen wets UO/sub 2/ better than a melt poor in oxygen. Molten Zircaloy containing little or no oxygen reacts with UO/sub 2/ to form a homogeneous (U,Zr,O) melt. As the oxygen content of the melt increases, solid (U,Zr)O /SUB 2-x/ particles precipitate. The technical significance of these out-of-pile UO/sub 2//Zircaloy reaction experiments is that Zircaloy cladding can be oxidized by UO/sub 2/ fuel as quickly as by steam, and that UO/sub 2/, far below its melting point, can be ''liquefied'' by molten Zircaloy. As a consequence, release of fission gas and volatile fission products is enhanced.
- Research Organization:
- Kernforschungszentrum Karlsruhe, Institut fur Material- und Festkorperforschung Karlsruhe
- OSTI ID:
- 5279599
- Journal Information:
- Nucl. Technol.; (United States), Vol. 65:1
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
URANIUM DIOXIDE
FUEL-CLADDING INTERACTIONS
WETTABILITY
ZIRCALOY 4
ARGON
CHEMICAL REACTION KINETICS
DIFFUSION
FISSION PRODUCT RELEASE
FUEL RODS
MELTING POINTS
OXIDATION
OXYGEN
PRECIPITATION
QUANTITY RATIO
URANIUM ALLOYS
VERY HIGH TEMPERATURE
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCONIUM ALLOYS
ZIRCONIUM OXIDES
ACTINIDE ALLOYS
ACTINIDE COMPOUNDS
ALLOYS
CHALCOGENIDES
CHEMICAL REACTIONS
ELEMENTS
FLUIDS
FUEL ELEMENTS
GASES
KINETICS
NONMETALS
OXIDES
OXYGEN COMPOUNDS
PHYSICAL PROPERTIES
RARE GASES
REACTION KINETICS
REACTOR COMPONENTS
REACTORS
SEPARATION PROCESSES
THERMODYNAMIC PROPERTIES
TIN ALLOYS
TRANSITION ELEMENT COMPOUNDS
TRANSITION TEMPERATURE
URANIUM COMPOUNDS
URANIUM OXIDES
ZIRCALOY
ZIRCONIUM BASE ALLOYS
ZIRCONIUM COMPOUNDS
210100* - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled