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Title: Recovery of plutonium from nitric acid waste

Abstract

Seven candidate materials, each contained in a static-bed column, have been evaluated for removing plutonium from nitric acid waste. Three materials have insufficient capacity for plutonium: TBP (tri-n-butylphosphate) sorbed on Amberlite XAD-4 resin, O phi D(IB)CMPO (octylphenyl-N, N-diisobutylcarbamoylmethylphosphine oxide) sorbed on XAD-4, and Amberlite IRA-938 anion exchange resin. The remaining four materials reduced 10/sup -3/ g/l plutonium in 7.2M HNO/sub 3/ to low 10/sup -5/ g/l for 80 bed volumes (BV). The 10% breakthrough capacities for 3 x 10/sup -2/ g/l plutonium are: TOPO (tri-n-octylphosphine oxide) on XAD-4 - 1800 BV, DHDECMP (dihexyl-N, N-diethylcarbamoylmethylphosphonate) on XAD-4 - 960 BV, Dowex 1 x 4 - 840 BV, and DHDECMP + TBP - 640 BV. Based on these results and generally poor elution of all materials, TOPO on XAD-4 is recommended as the best candidate for recovery of plutonium followed by acid digestion or combustion of the TOPO to recover the concentrated plutonium.

Authors:
; ;
Publication Date:
Research Org.:
Rockwell International Corp., Golden, CO (USA). Rocky Flats Plant
OSTI Identifier:
6606080
Report Number(s):
RFP-4009
ON: DE87009471
DOE Contract Number:
AC04-76DP03533
Resource Type:
Technical Report
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
Subject:
37 INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL CHEMISTRY; PLUTONIUM; RECOVERY; DHDECMP; EXTRACTION CHROMATOGRAPHY; NITRIC ACID; RESINS; ROCKY FLATS PLANT; TBP; TOPO; ACTINIDES; BUTYL PHOSPHATES; CHROMATOGRAPHY; ELEMENTS; ESTERS; HYDROGEN COMPOUNDS; INORGANIC ACIDS; METALS; NATIONAL ORGANIZATIONS; ORGANIC COMPOUNDS; ORGANIC PHOSPHORUS COMPOUNDS; ORGANIC POLYMERS; PETROCHEMICALS; PETROLEUM PRODUCTS; PHOSPHONIC ACID ESTERS; PHOSPHORIC ACID ESTERS; POLYMERS; SEPARATION PROCESSES; TRANSURANIUM ELEMENTS; US AEC; US DOE; US ERDA; US ORGANIZATIONS 400105* -- Separation Procedures

Citation Formats

Muscatello, A.C., Saba, M.T., and Navratil, J.D. Recovery of plutonium from nitric acid waste. United States: N. p., 1986. Web.
Muscatello, A.C., Saba, M.T., & Navratil, J.D. Recovery of plutonium from nitric acid waste. United States.
Muscatello, A.C., Saba, M.T., and Navratil, J.D. 1986. "Recovery of plutonium from nitric acid waste". United States. doi:.
@article{osti_6606080,
title = {Recovery of plutonium from nitric acid waste},
author = {Muscatello, A.C. and Saba, M.T. and Navratil, J.D.},
abstractNote = {Seven candidate materials, each contained in a static-bed column, have been evaluated for removing plutonium from nitric acid waste. Three materials have insufficient capacity for plutonium: TBP (tri-n-butylphosphate) sorbed on Amberlite XAD-4 resin, O phi D(IB)CMPO (octylphenyl-N, N-diisobutylcarbamoylmethylphosphine oxide) sorbed on XAD-4, and Amberlite IRA-938 anion exchange resin. The remaining four materials reduced 10/sup -3/ g/l plutonium in 7.2M HNO/sub 3/ to low 10/sup -5/ g/l for 80 bed volumes (BV). The 10% breakthrough capacities for 3 x 10/sup -2/ g/l plutonium are: TOPO (tri-n-octylphosphine oxide) on XAD-4 - 1800 BV, DHDECMP (dihexyl-N, N-diethylcarbamoylmethylphosphonate) on XAD-4 - 960 BV, Dowex 1 x 4 - 840 BV, and DHDECMP + TBP - 640 BV. Based on these results and generally poor elution of all materials, TOPO on XAD-4 is recommended as the best candidate for recovery of plutonium followed by acid digestion or combustion of the TOPO to recover the concentrated plutonium.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 1986,
month =
}

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  • 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 distributionmore » 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.« less
  • Contaminated high-efficiency particulate air (HEPA) filter media, containing PuO/sub 2/ powder which had been calcined at 700/sup 0/C, were treated with concentrated H/sub 2/SO/sub 4/-HNO/sub 3/ at 190 to 200/sup 0/C for periods ranging from 0.5 to 2.0 hours. When followed by a dilute HNO/sub 3/ rinse, this treatment was shown to be very effective as a plutonium recovery operation (approximately greater than 97% of the plutonium was solubilized). A proposed treatment scheme is given which could provide both a plutonium recovery option for HEPA filters and a reduction in overall waste volume.
  • A detailed experimental study has shown absorption of plutenium(IV) from 7 M HNO/sub 3/ solutions by strongbase (quaternary amine) anion exchange resins to be a very powerful and versatile processing method for plutonium. The simple expedient of operating at 50 to 60 deg C relieves most of the objections of former anion exchange processes for plutonium in that it permits adequate processing rates and efficiency to be obtained with four per cent cross-linked resins. Commercially available strong-base anion exchange resins Dowex-1, X-4 (50 to 100 mesh), among 50 to 100 mesh resins, and Permutit SK (20 to 50 mesh), amongmore » 20 to 50 mesh resins, are markedly superior for most process applications. The anion exchange process consists of an absorption or loading step in which plutonium is absorbed out of a feed solution by the resin, washing step in which impurities are removed by washing the plutonium-laden resin with an appropriate wash solution, and an elution step in which plutonium is stripped off the resin in an appropriate aqueous solution The effects of the pertinent variables in each of these steps have been studied in some detail, yielding informatlon which can be readily extrapolated to new processing applications. (auth)« less