Extraction and recovery of plutonium and americium from nitric acid waste solutions by the TRUEX process - continuing development studies

Technical Report · Sun Sep 01 04:00:00 EDT 1985

DOI:https://doi.org/10.2172/5160292· OSTI ID:5160292

Leonard, R A; Vandegrift, G F; Kalina, D G; Fischer, D F; Bane, R W; Burris, L; Horwitz, E P; Chiarisia, R; Diamond, H

This report summarizes the work done to date on the application of the TRUEX solvent extraction process for removing and separately recovering plutonium and americium from a nitric acid waste solution containing these elements, uranium, and a complement of inert metal ions. This simulated waste stream is typical of a raffinate from a tributyl phosphate (TBP)-based solvent extraction process for removing uranium and plutonium from dissolved plutonium-containing metallurgical scrap. The TRUEX process solvent in these experiments was a solution of TBP and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) dissolved in carbon tetrachloride. A flowsheet was designed on the basis of measured batch distribution ratios to reduce the TRU content of the solidified raffinate to less than or equal to 10 nCi/g and was tested in a countercurrent experiment performed in a 14-stage Argonne-model centrifugal contractor. The process solvent was recycled without cleanup. An unexpectedly high evaporative loss of CCl/sub 4/ resulted in concentration of the active extractant, CMPO, to nearly 0.30M in the solvent. Results are consistent with this higher CMPO concentration. The raffinate contained only 2 nCi/g of TRU, but the higher CMPO concentration resulted in reduced effectiveness in the stripping of americium from the solvent. Conditions can be easily adjusted to give high yields and good separation of americium and plutonium. Experimental studies of the hydrolytic and gamma-radiolytic degradation of the TRUEX-CCl/sub 4/ showed that solvent degradation would be (1) minimal for a year of processing this typical feed, which contained no fission products, and (2) could be explained almost entirely by hydrolytic and radiolytic damage to TBP. Even for gross amounts of solvent damage, scrubbing with aqueous sodium carbonate solution restored the original americium extraction and stripping capability of the solvent. 43 refs., 5 figs., 36 tabs.

• Research Organization: Argonne National Lab., IL (USA)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 5160292
• Report Number(s): ANL-85-45; ON: DE86000826
• Country of Publication: United States
• Language: English

Related Subjects

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ACTINIDES
ALKALI METAL COMPOUNDS
AMERICIUM
BUTYL PHOSPHATES
CARBON COMPOUNDS
CARBON TETRACHLORIDE
CARBONATES
CHEMICAL RADIATION EFFECTS
CHEMICAL REACTIONS
CHEMISTRY
CHLORINATED ALIPHATIC HYDROCARBONS
DATA
DECOMPOSITION
DIAGRAMS
ELEMENTS
ESTERS
EXPERIMENTAL DATA
EXTRACTION
FLOWSHEETS
HALOGENATED ALIPHATIC HYDROCARBONS
HYDROGEN COMPOUNDS
INFORMATION
INORGANIC ACIDS
MATERIALS
METALS
NITRIC ACID
NUMERICAL DATA
ORGANIC CHLORINE COMPOUNDS
ORGANIC COMPOUNDS
ORGANIC HALOGEN COMPOUNDS
ORGANIC PHOSPHORUS COMPOUNDS
OXYGEN COMPOUNDS
PHOSPHORIC ACID ESTERS
PLUTONIUM
RADIATION CHEMISTRY
RADIATION EFFECTS
RADIOACTIVE MATERIALS
RADIOACTIVE WASTES
RADIOLYSIS
RECOVERY
SCRUBBING
SEPARATION PROCESSES
SODIUM CARBONATES
SODIUM COMPOUNDS
SOLVENT EXTRACTION
TBP
TRANSPLUTONIUM ELEMENTS
TRANSURANIUM ELEMENTS
WASTES