Dissolution of mixed oxide fuel as a function of fabrication variables

Conference · Wed Aug 01 04:00:00 EDT 1979

OSTI ID:5860288

Lerch, R. E.

Dissolution properties of mechanically blended mixed oxide fuel were very dependent on the six fuel fabrication variables studied. Fuel sintering temperature, source of PuO/sub 2/ and PuO/sub 2/ content of the fuel had major effects: (1) as the sintering temperature was increased from 1400 to 1700/sup 0/C, pellet dissolution was more complete; (2) pellets made from burned metal derived PuO/sub 2/ were more completely dissolved than pellets made from calcined nitrate derived PuO/sub 2/ which in turn were more completely dissolved than pellets made from calcined nitrate derived PuO/sub 2/; (3) as the PuO/sub 2/ content decreased from 25 to 15 wt % PuO/sub 2/, pellet dissolution was more complete. Preferential dissolution of uranium occurred in all the mechanically blended mixed oxide. Unirradiated mixed oxide fuel pellets made by the Sol Gel process were generally quite soluble in nitric acid. Unirradiated mixed oxide fuel pellets made by the coprecipitation process dissolved completely and rapidly in nitric acid. Fuel made by the coprecipitation process was more completely dissolved than fuel made by the Sol Gel process which, in turn, was more completely dissolved than fuel made by mechanically blending UO/sub 2/ and PuO/sub 2/ as shown below. Addition of uncomplexed fluoride to nitric acid during fuel dissolution generally rendered all fuel samples completely dissolvable. In boiling 12M nitric acid, 95 to 99% of the plutonium which was going to dissolve did so in the first hour. Irradiated mechanically blended mixed oxide fuel with known fuel fabrication conditions was also subjected to fuel dissolution tests. While irradiation was shown to increase completeness of plutonium dissolution, poor dissolubility due to adverse fabrication conditions (e.g., low sintering temperature) remained after irradiation.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (USA)
• DOE Contract Number: EY-76-C-14-2170
• OSTI ID: 5860288
• Report Number(s): HEDL-SA-1935; CONF-791086-3
• Country of Publication: United States
• Language: English

Related Subjects

050800* -- Nuclear Fuels-- Spent Fuels Reprocessing
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ACTINIDE COMPOUNDS
BREEDER REACTORS
CHALCOGENIDES
COPRECIPITATION
DISSOLUTION
ENERGY SOURCES
EPITHERMAL REACTORS
FABRICATION
FAST REACTORS
FBR TYPE REACTORS
FLUORIDES
FLUORINE COMPOUNDS
FUEL PELLETS
FUELS
HALIDES
HALOGEN COMPOUNDS
HYDROGEN COMPOUNDS
INORGANIC ACIDS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MIXING
NITRIC ACID
NUCLEAR FUELS
OXIDES
OXYGEN COMPOUNDS
PELLETS
PLUTONIUM COMPOUNDS
PLUTONIUM DIOXIDE
PLUTONIUM OXIDES
PRECIPITATION
REACTOR MATERIALS
REACTORS
SEPARATION PROCESSES
SOL-GEL PROCESS
SPENT FUELS
TRANSURANIUM COMPOUNDS
URANIUM COMPOUNDS
URANIUM DIOXIDE
URANIUM OXIDES