Solubility of UO[sub 2] in molten Zircaloy-4 from 2000 to 2300[degrees]C

Conference · Fri Jan 01 04:00:00 EST 1993 · Transactions of the American Nuclear Society; (United States)

OSTI ID:6911615

Hayward, P J; George, I M ^[1]

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1. Atomic Energy of Canada Ltd., Manitoba (Canada)

This paper reports solubility measurements for unirradiated UO[sub 2] fuel at 2000, 2100, 2200, and 2300[degrees]C in initially oxygen-free Zircaloy-4 (Zr-4) and at 2100, 2200, and 2300[degrees]C in Zircaloy-4 with an initial 25 at.% oxygen content (Zr-4/25% oxygen); the latter alloy was used to represent the oxygen-saturated Zr-4 component of steam-oxidized cladding. The work was performed to determine the maximum extent of fuel dissolution that could occur at various temperatures during the high-temperature transient of a hypothetical reactor accident involving impaired cooling. These values could then be used to estimate the fractional release of the volatile fission product inventory, which, to a first approximation, would be proportional to the volume fraction of dissolved fuel.

• OSTI ID: 6911615
• Report Number(s): CONF-931160--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 69
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220300 -- Nuclear Reactor Technology-- Fuel Elements
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ACTINIDE COMPOUNDS
ALLOY-ZR98SN-4
ALLOYS
CHALCOGENIDES
CHEMICAL REACTIONS
CHROMIUM ADDITIONS
CHROMIUM ALLOYS
CORROSION RESISTANT ALLOYS
DISSOLUTION
ENERGY SOURCES
FUEL CANS
FUELS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
IRON ADDITIONS
IRON ALLOYS
MATERIALS
NUCLEAR FUELS
OXIDATION
OXIDES
OXYGEN COMPOUNDS
REACTOR ACCIDENTS
REACTOR CORE DISRUPTION
REACTOR MATERIALS
SOLUBILITY
TEMPERATURE RANGE
TEMPERATURE RANGE 1000-4000 K
TIN ALLOYS
URANIUM COMPOUNDS
URANIUM DIOXIDE
URANIUM OXIDES
ZIRCALOY
ZIRCALOY 4
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS