Extraction and recovery of plutonium and americium from nitric acid waste solutions by the TRUEX process - continuing development studies
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. (ANL), Argonne, IL (United States)
- DOE Contract Number:
- W-31-109-ENG-38
- OSTI ID:
- 5160292
- Report Number(s):
- ANL-85-45; ON: DE86000826
- Resource Relation:
- Other Information: Portions of this document are illegible in microfiche products. Original copy available until stock is exhausted
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
ORGANIC
PHYSICAL AND ANALYTICAL CHEMISTRY
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
AMERICIUM
SOLVENT EXTRACTION
PLUTONIUM
TBP
RADIOLYSIS
CARBON TETRACHLORIDE
EXPERIMENTAL DATA
FLOWSHEETS
NITRIC ACID
ORGANIC PHOSPHORUS COMPOUNDS
RADIOACTIVE WASTES
RECOVERY
SCRUBBING
SODIUM CARBONATES
ACTINIDES
ALKALI METAL COMPOUNDS
BUTYL PHOSPHATES
CARBON COMPOUNDS
CARBONATES
CHEMICAL RADIATION EFFECTS
CHEMICAL REACTIONS
CHEMISTRY
CHLORINATED ALIPHATIC HYDROCARBONS
DATA
DECOMPOSITION
DIAGRAMS
ELEMENTS
ESTERS
EXTRACTION
HALOGENATED ALIPHATIC HYDROCARBONS
HYDROGEN COMPOUNDS
INFORMATION
INORGANIC ACIDS
MATERIALS
METALS
NUMERICAL DATA
ORGANIC CHLORINE COMPOUNDS
ORGANIC COMPOUNDS
ORGANIC HALOGEN COMPOUNDS
OXYGEN COMPOUNDS
PHOSPHORIC ACID ESTERS
RADIATION CHEMISTRY
RADIATION EFFECTS
RADIOACTIVE MATERIALS
SEPARATION PROCESSES
SODIUM COMPOUNDS
TRANSPLUTONIUM ELEMENTS
TRANSURANIUM ELEMENTS
WASTES
400105* - Separation Procedures
052002 - Nuclear Fuels- Waste Disposal & Storage