skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: RECENT DEVELOPMENTS IN FEED PREPARATION AND SOLVENT EXTRACTION

Abstract

For presentation at the 5th Nuclear Congress, Clevelands Apr. 7, 1959. Increasing emphasis has been placed recently on the application of solvent extraction to the recovery of uranium and plutonium from spent power reactor fuels. Zircaloy-2 jackets were removed from PWR blankettype fuels by dissolution with the Zirflex Process, and the UO/sub 2/ cores were dissolved in 10 M HNO/sub 3/. Zirflex treatment of prototype samples irradiated to 2500 Mwd/ ton resulted in satisfactory dissolution rates and losses to the dejacketing solution generally less than 0.2% for U and Pu. U-Zr alloy fuels were dissolved in 6 M NH/sub 4/F and adjusted for solvent extraction by the addition of ride was recycled by metathesis and precipitation. Stainless steel jackets were removed from Consolidated Edisontype fuels by dissolution in 6 M H/sub 2/SO/sub 4/ (Sulfex Process), and the ThO/sub 2/-UO/sub 2/ core was dissolved in 13 M HNO/sub 3/0.04 M F-0.04 M Al. Dejacketing losses in unirradiated samples were about 0.02%. Use of the ORNL Reference Darex flowsheet for APPR processing resulted in solvent extraction feed containing 30 ppm chloride. Mechanical equipment was designed to declad SRE fuels and chop and leach techniques are being developed to treat stainless and Zrmore » clad ceramic fuel. A solvent extraction flowsheet was developed for Foreign Research Reactor Fuels, (Al-20% enriched U alloy) using a revised tributyl phosphate extraction system for the separation of Pu from U. Feasibility tests were carried out on the coupling of Redox solvent extraction with Darex and Niflex head-end treatments. New solvents for the reprocessing of power reactor fuels are being studied. Among these, the amines show an order of magnitude greater radiation stability than does tributyl phosphate. A primary amine was proposed for the recovery of U and Pu from Sulfex decladding wastes. (auth)« less

Authors:
; ;
Publication Date:
Research Org.:
Oak Ridge National Lab., Tenn.
OSTI Identifier:
4245829
Report Number(s):
CF-58-11-91
NSA Number:
NSA-13-014385
DOE Contract Number:
W-7405-ENG-26
Resource Type:
Technical Report
Resource Relation:
Other Information: For presentation at the 5th Nuclear Congress, Cleveland, Apr. 7, 1959. Orig. Receipt Date: 31-DEC-59
Country of Publication:
United States
Language:
English
Subject:
CHEMISTRY; ACIDITY- BLANKETS- CHEMICAL REACTIONS- FUEL CANS- FUEL ELEMENTS- FUELS- IRRADIATION- NITRIC ACID- PLUTONIUM- PWR- QUANTITY RATIO- RADIATION DOSES- REACTION KINETICS- REACTORS- RECOVERY- REPROCESSING- SAMPLING- SOLUTIONS- SOLVENT EXTRACTION- SOLVENTS- SULFEX PROCESS- TESTING- URANIUM- URANIUM DIOXIDE- ZIRCALOY- ZIRFLEX PROCESS; ALUMINUM ALLOYS- AMINES- BUTYL PHOSPHATES- CERAMICS- DECOMPOSITION- ORGANIC NITROGEN COMPOUNDS- - OXIDATION- PLANNING- POWER PLANTS- RADIATION CHEMISTRY- RADIATION EFFECTS- REDUCTION- REPROCESSING- SOLVENT EXTRACTION- STABILITY- URANIUM ALLOYS- WASTE SOLUTIONS; ALUMINUM FLUORIDES-

Citation Formats

Bruce, F.R., Blanco, R.E., and Bresee, J.C. RECENT DEVELOPMENTS IN FEED PREPARATION AND SOLVENT EXTRACTION. United States: N. p., 1959. Web. doi:10.2172/4245829.
Copy to clipboard
Bruce, F.R., Blanco, R.E., & Bresee, J.C. RECENT DEVELOPMENTS IN FEED PREPARATION AND SOLVENT EXTRACTION. United States. doi:10.2172/4245829.
Copy to clipboard
Bruce, F.R., Blanco, R.E., and Bresee, J.C. Fri . "RECENT DEVELOPMENTS IN FEED PREPARATION AND SOLVENT EXTRACTION". United States. doi:10.2172/4245829. https://www.osti.gov/servlets/purl/4245829.
Copy to clipboard
@article{osti_4245829,
title = {RECENT DEVELOPMENTS IN FEED PREPARATION AND SOLVENT EXTRACTION},
author = {Bruce, F.R. and Blanco, R.E. and Bresee, J.C.},
abstractNote = {For presentation at the 5th Nuclear Congress, Clevelands Apr. 7, 1959. Increasing emphasis has been placed recently on the application of solvent extraction to the recovery of uranium and plutonium from spent power reactor fuels. Zircaloy-2 jackets were removed from PWR blankettype fuels by dissolution with the Zirflex Process, and the UO/sub 2/ cores were dissolved in 10 M HNO/sub 3/. Zirflex treatment of prototype samples irradiated to 2500 Mwd/ ton resulted in satisfactory dissolution rates and losses to the dejacketing solution generally less than 0.2% for U and Pu. U-Zr alloy fuels were dissolved in 6 M NH/sub 4/F and adjusted for solvent extraction by the addition of ride was recycled by metathesis and precipitation. Stainless steel jackets were removed from Consolidated Edisontype fuels by dissolution in 6 M H/sub 2/SO/sub 4/ (Sulfex Process), and the ThO/sub 2/-UO/sub 2/ core was dissolved in 13 M HNO/sub 3/0.04 M F-0.04 M Al. Dejacketing losses in unirradiated samples were about 0.02%. Use of the ORNL Reference Darex flowsheet for APPR processing resulted in solvent extraction feed containing 30 ppm chloride. Mechanical equipment was designed to declad SRE fuels and chop and leach techniques are being developed to treat stainless and Zr clad ceramic fuel. A solvent extraction flowsheet was developed for Foreign Research Reactor Fuels, (Al-20% enriched U alloy) using a revised tributyl phosphate extraction system for the separation of Pu from U. Feasibility tests were carried out on the coupling of Redox solvent extraction with Darex and Niflex head-end treatments. New solvents for the reprocessing of power reactor fuels are being studied. Among these, the amines show an order of magnitude greater radiation stability than does tributyl phosphate. A primary amine was proposed for the recovery of U and Pu from Sulfex decladding wastes. (auth)},
doi = {10.2172/4245829},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Mar 20 00:00:00 EST 1959},
month = {Fri Mar 20 00:00:00 EST 1959}
}
Copy to clipboard
Technical Report:
View Technical Report (6.48 MB)
DOI:  10.2172/4245829
Save / Share:
Export Metadata
Save to My Library
You must Sign In or Create an Account in order to save documents to your library.

  • ; ;
    Increasing emphasis has been placed recently on the application of solvent extraction to the recovery of uranium and plutonium from spent power- reactor fuels. Zircaloy2 jackets were removed from PWR blanket type fuels by dissolution with the Zirflex process, and the UO/sub 2/ cores were dissolved in 10 M HNO/sub 3/. Zirflex treatment of prototype samples irradiated to 2500 Mwd/ ton resulted in satisfactory dissolution rates and losses to the dejacketing solution generally less than 0.2% for uranium and plutonium. Zr-U alloy fuels were dissolved in 6 M NH/sub 4/F and adjusted for solvent extraction by the addition of Al(NO/submore » 3/)/sub 3/ and HNO/sub 3/. In an alternative procedure, fluoride was recycled by metathesis and precipitation. Stainlesssteel jackets were removed from Consolidated Edison type fuels by dissolution in 6 M H/sub 2/SO/ sub 4/ (Sulfex process), and the ThC/sub 2/-UO/sub 2/ core was dissolved in l3 M HNO/sub 3/-0.04 M F0.04 M Al. Dejacketing losses in unirradiated samples were about 0.02%. Use of the ORNL Deference Darex flowsheet for APPR processing resulted in solventextraction feed containing 10 ppr chloride. Mechanical equipment was designed to declad SRE fuel, and chop and leach techniques are being developed to treat stainlessand zirconium-clad ceramic fuel. A solvent- extraction flowsheet was developed for Foreign Research Reactor Fuels, (Al-20% enriched U alloy) using a revised tributyl phosphate extraction system for the separation of plutonium from uranium. Feasibility tests were carried out on the coupling of Redox solvent-extraction with Darex and Niflex head-end treatments. New solvents for the reprocessing of power-reactor fuels are being studied. Among these, the amines show an order of magnitude greater radiation stability than does tributyl phosphate. A primary amine was proposed for the recovery of uranium and plutonium from Sulfex decladding wastes. (auth)« less

  • Reprocessing of HTGR fuel requires denitration of the highly acidic dissolver solution prior to solvent extraction. The results of testing a pilot plant batch process on unirradiated fuel are given in this interim development report.

  • ; ; ; ...
    Tributyl phosphate can be used for extraction in processing all current power reactor fuels. Nitric acid is the only salting agent required. Typical flowsheets are presented. In aluminum nitrate systems which are more than 0.1 M acid deficient, the uranium distribution coefficient is a function of pH and independent of aluminum concentration; the coefficient remains constant at one in fluoride systems when the nitrate to fluoride ratio is approximates 3.5. Many objectionable properties of degraded diluents are ascribed to nitroparaffins. Aliphatic diluents with the least branching are the most stable to nitration. The nitration stability of aromatic diluents varies withmore » structure, e.g., stabilities of diethylbenzenes decrease as meta >> ortho > para. Solvent purification by flash distillation appears superior to other methods. The stability of Amsco 125-82 was permanently improved by treatment with sulfuric acid. The radiation stability of TBP was approximates 2 times higher in an aromatic diluent than in Amsco 125-82. The G decomposition value for 1 M TBP in Amsco alone was approximates 0.9; whereas in 1 to 3 M HNO/sub 3/ it was 1 to 5 and G (--HNO/sub 3/ org phase) was 3 to 20. Variation of uranium--thorium separation factors with structure of some neutral organophosphorus reagents is presented. Basic studies include measurement of activities in multicomponent solutions and description of aqueous activity coefficients by an extended Debye- Huckel equation. (auth)« less

  • ;
    An investigation was made of the feasibility of producing pure Th compounds from monazite sand by a process involving the digestion of the sand with H/sub 2/SO/sub 4/ and separation of Th from the rare earths by solvent extraction. Direct extraction was not practicable. Precipitation of the sulfate from the system with lime and HNO/sub 3/ was tried and found to be partially successful but impracticable because of the large losses of Th by occlusion in the precipitate. A successful method for producing tributylphosphate-extractable solution was found, consisting of neutralizing the original solution to precipitate the heavy metals as phosphates,more » and, after filtering and washing, dissolving this precipitate in HNO/sub 3/. (auth)« less

  • The two-cycle Acid-Thorex solvent extraction process requires that the feed stream to each thorium cycle be processed to reduce its nitric acid concentration (feed adjustment). This interim development report presents the results of bench-scale and pilot-plant-scale feed adjustment experiments using a continuous mode of operation. An examination of formic acid denitration and fluoride ion volatilization is also included.