Dissolution of COPRECAL mixed-oxide fuel
Conclusions based on this investigation of continuous coprecipitation-calcination (COPRECAL) mixed oxide (MOX) dissolution are as follows: (1) A large fraction of the COPRECAL MOX dissolved very rapidly in nitric acid alone. Appeoximately 95% of the powder tested dissolved in the first 5 to 6 minutes in boiling 8 to 12M nitric acid. The remainder dissolved much more slowly, with greater than 99% dissolved after 8 hours. The dissolution rate increased slightly as the nitric concentration was increased from 8 to 12M. (2) Of the fuel fabrication operations investigated, only the sintering operation had a significant effect on the dissolution rate. In tests with samples from a fuel fabrication run, the sintered pellet samples had approximately one-tenth as much residue remaining at the end of 6- and 12-hour tests in 12M nitric acid compared to the MOX powder starting material. (3) The presence of Pu and U in the nitric acid dissolvent had very little effect on the dissolution rate. In tests with Pu-U nitrate-nitric acid solution as the dissolvent, the final 5% of the COPRECAL sample dissolved slightly slower than it had in tests with nitric acid only as the dissolvent. (4) The plutonium-uranium ratio was higher in the final residue than in the starting material. At the end of a 6-hour dissolution test in 10M nitric acid the Pu-U ratio in the residue was 4 vs 0.25 in the starting material. (5) A final residue of undissolved oxide can be expected in the Wet Scrap Development Laboratory (WSDL) dissolver. As a way to minimize the buildup of undissolved MOX powder in the continuous dissolver, all scrap material could be subjected to high temperature sintering (e.g., approx. 1700/sup 0/C) prior to processing through WSDL.
- Research Organization:
- Hanford Engineering Development Lab., Richland, WA (USA)
- DOE Contract Number:
- AC14-76FF02170
- OSTI ID:
- 6076060
- Report Number(s):
- HEDL-SA-2327; CONF-810606-93; ON: DE81028205; TRN: 81-015868
- Resource Relation:
- Conference: American Nuclear Society's annual meeting, Miami Beach, FL, USA, 7 Jun 1981
- Country of Publication:
- United States
- Language:
- English
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MIXED OXIDE FUELS
DISSOLUTION
FABRICATION
CALCINATION
COPRECIPITATION
NITRIC ACID
PLUTONIUM OXIDES
REPROCESSING
SPENT FUELS
URANIUM OXIDES
ACTINIDE COMPOUNDS
CHALCOGENIDES
CHEMICAL REACTIONS
DECOMPOSITION
ENERGY SOURCES
FUELS
HYDROGEN COMPOUNDS
INORGANIC ACIDS
MATERIALS
NUCLEAR FUELS
OXIDES
OXYGEN COMPOUNDS
PLUTONIUM COMPOUNDS
PRECIPITATION
PYROLYSIS
REACTOR MATERIALS
SEPARATION PROCESSES
SOLID FUELS
THERMOCHEMICAL PROCESSES
TRANSURANIUM COMPOUNDS
URANIUM COMPOUNDS
050800* - Nuclear Fuels- Spent Fuels Reprocessing
050700 - Nuclear Fuels- Fuels Production & Properties