EVALUATION OF AN ENGINEERING DEMONSTRATION OF THE MODIFIED ZIRFLEX AND NEUFLEX PROCESSES FOR THE PREPARATION OF SOLVENT EXTRACTION FEEDS FROM UNIRRADIATED ZIRCONIUM-BASE REACTOR FUELS

Technical Report · Sun Mar 01 04:00:00 EST 1964

DOI:https://doi.org/10.2172/4043434· OSTI ID:4043434

Kitts, F G

In order to recover uranium from zirconium-base reactor fuels by solvent extraction, the metailic fuel and cladding must first be dissolved and a suitable feed solution prepared. Such preparations of solvent extraction feeds were successfully accomplished batchwise using both the Modified Zirflex and Neuflex processes employing an NH/sub 4/F -- oxidant mixture to dissolve the fuel elements, and the feed. (The d Zirflex feed, and H/sub 2/O for the Neuflex feed.) In the Modified Zirflex process, a dissolvent about 6 M in NH/sub 4/F with an excess of H/sub 2/O/sub 2/ to oxidize uranium to the more-soluble U(VI) valence state. The off-gas, after NH/sub 3/ removal, is an H/sub 2/-O/sub 2/ mixture of small volume, which is diluted with air to a safe concentration. Then nitric acid-aluminum nitrate is added to the dissolution product, yielding a solvent extraction feed from which uranium is recovered by using TBP-Amsco as the extractant. In the Neuflex process, the dissolvent is NH/sub 4/F--H/sub 2/O/sub 2/, with less than a stoichiometric amount of NH/sub 4/NO/sub 3/. Without NH/sub 4/NO/sub 3/, the scrubbed off-gas is principally hydrogen, on the hydrogen-rich side of the flammable range of H/sub 2/-O/sub 2/ mixtures, Only water is added to this dissolution product, yielding a neutral fluoride feed from which uranium is extractable by use of Dapex reagents. ln both processes the F: Zr charge ratio, initial surface condition, and maximum section thickness of the fuel element were the principa1 determinants of total dissolution time. The zirconium loading as determined by the free fluoride - zirconium solubility relationship limited the capacity of fuels containing less than 2% U, while the free-fluoride-to-uranium ratio of about 100 required for solution stability was the limiting factor with alloys containing higher percentages of uranium, Hydrogen peroxide concentration was not an important factor in solution stability; the role of ammonla or NH/sub 4/OH was not studied. The feasibility of both processes was demonstrated by a series of batch dissolutions of kilogram quartities of various fuels containing 1 to 8% uranium. Continuous dissolution was demonstrated as was application to TRIGA fuel alloy (8% U-- ZrH). Stainless steel type 347 and a low-carbon nickel alloy were suitable materials of construction for the dissolution and the solvent extraction equipment. Since there were some discrepancies betweeq small-scale and engineering-scale work, especially in the prevention of precipitate formation near the end of the dissolution cycle, it is advised that some further investigation be made prior to attempted scaleup to plant operation. (auth)

• Research Organization: Oak Ridge National Lab., Tenn.
• Sponsoring Organization: USDOE
• DOE Contract Number: W-7405-ENG-26
• NSA Number: NSA-18-017932
• OSTI ID: 4043434
• Report Number(s): ORNL-3465
• Country of Publication: United States
• Language: English

Related Subjects

ACIDITY
AIR
ALLOYS
ALUMINUM NITRATES
AMMONIA
AMMONIUM COMPOUNDS
AMSCO
BUTYL PHOSPHATES
CARBON
CATIONS
CHEMICAL REACTIONS
CHEMICALS
CHEMISTRY
CONFIGURATION
DAPEX PROCESS
EXTRACTION COLUMNS
FLUORIDES
FUEL CANS
FUEL ELEMENTS
FUELS
GASES
HYDROGEN
HYDROGEN PEROXIDES
IMPURITIES
KEROSENE
METALS
MIXING
NICKEL ALLOYS
NITRATES
NITRIC ACID
OPERATION
OXIDATION
OXYGEN
PERFORMANCE
PLANNING
PRECIPITATION
PREPARATION
PRODUCTION
QUANTITY RATIO
REACTION KINETICS
RECOVERY
REPROCESSING
SAFETY
SAMPLING
SOLUBILITY
SOLUTIONS
SOLVENT EXTRACTION
STABILITY
STAINLESS STEELS
TBP
THICKNESS
TRACE AMOUNTS
UR
URANIUM
URANIUM ALLOYS