Evaluation of a spent fuel as a waste form in a salt repository

Conference · Sat Oct 01 04:00:00 EDT 1983

OSTI ID:5696791

Gray, W J; McVay, G L; Barner, J O; Shade, J W; Cote, R W

Leach tests have been performed on spent fuel in sythetic Permian Basin salt brine at 25 and 75/sup 0/C. Complementary tests on unirradiated UO/sub 2/ pellets have been conducted in both salt brine and deionized water in the range 25 to 150/sup 0/C. Iron and/or oxidized zircaloy coupons were included in some of the tests. Uranium release from spent fuel was more than 100 times greater than from UO/sub 2/. In brine, iron had no significant effect on the total uranium release but substantially reduced the amount in solution by causing the uranium to plate out on the iron coupon and container walls and to precipitate as filterable particles.

• Research Organization: Pacific Northwest Lab., Richland, WA (USA); Battelle Memorial Inst., Columbus, OH (USA). Office of Nuclear Waste Isolation
• DOE Contract Number: AC06-76RL01830
• OSTI ID: 5696791
• Report Number(s): PNL-SA-11770; CONF-831174-25; ON: DE84003343
• Country of Publication: United States
• Language: English

Related Subjects

052002* -- Nuclear Fuels-- Waste Disposal & Storage
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
ACTINIDE COMPOUNDS
ACTINIDES
ALLOYS
BRINES
CHALCOGENIDES
CORROSIVE EFFECTS
DATA
DISSOLUTION
ELEMENTS
ENERGY SOURCES
EVALUATION
EXPERIMENTAL DATA
FEDERAL REGION VI
FUELS
GEOLOGIC DEPOSITS
HIGH TEMPERATURE
HYDROGEN COMPOUNDS
INFORMATION
IRON
LEACHING
MANAGEMENT
MATERIALS
MEDIUM TEMPERATURE
METALS
NORTH AMERICA
NUCLEAR FUELS
NUMERICAL DATA
OXIDES
OXYGEN COMPOUNDS
RADIOACTIVE WASTE DISPOSAL
REACTOR MATERIALS
SALT DEPOSITS
SEPARATION PROCESSES
SPENT FUELS
TEXAS
TIN ALLOYS
TRANSITION ELEMENTS
URANIUM
URANIUM COMPOUNDS
URANIUM DIOXIDE
URANIUM OXIDES
USA
WASTE DISPOSAL
WASTE MANAGEMENT
WATER
ZIRCALOY
ZIRCALOY 4
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS