Reagents and methods that have been developed during the past 20 years for hydrometallurgical partitioning of actinides from different types of transuranium (TRU) wastes and dissolved fuels are reviewed. Emphasis is placed on the extraction performance of the fully-optimized reagents rather than on the structural iterations that were undertaken (and in some cases are still being conducted) to identify the optimum species. Particular attention is paid to separation processes that have been demonstrated in batch and counter-current solvent extraction, and batch and column mode extraction chromatography. The salient features of the various techniques and reagents for actinide recycle are compared. Sections of the review focus on neptunium behavior in hydrometallurgy and on characterization of those reagents best suited to the separation of trivalent actinides from fission product lanthanides. Selected flowsheets that have been reported for the separation and recovery of actinides from TRU wastes are presented.

Actinide partitioning—a review

A generic transurantc (TRU) element extraction/recovery process was developed based on the use of octyl(phenyl)-N,N-diiso-butylcarbamoylmetliylphosphine oxide, 0φD(iB)CMPO, dissolved in PUREX process solvent (tribntyl phosphate, TBP, in normal paraffluic hydrocarbon, NPH). The process (called TRUEX) is capable of reducing the TRU concentration by many orders of magnitude In waste solutions containing a wide range of nitric acid, salt, and fission product concentrations. A major feature of the process is that it is readily adaptable for waste processing in existing fuel reprocessing facilities.

THE TRUEX PROCESS - A PROCESS FOR THE EXTRACTION OF THE TKANSURANIC ELEMENTS EROM NITRIC AC In WASTES UTILIZING MODIFIED PUREX SOLVENT*

Actinide partitioning is a proposed strategy for effective mitigation of the long-term hazards associated with high-level waste (HLW). Octyl-(phenyl)–N,N-diisobutyl carbamoyl methyl phosphine oxide (CMPO) and diphenyl–N,N-diisobutyl carbamoyl methyl phosphine oxide (DφCMPO) are amongst the promising extractants extensively studied since the 1980s for actinide partitioning from wastes of different origin. During the last two decades, substituted malonamide extractants such as N,N'-dimethyl-N,N'-dibutyl tetradecyl malonamide (DMDBTDMA) and N,N'-dimethyl-N,N'-dioctyl hexylethoxy malonamide (DMDOHEMA) have emerged as viable green alternatives to phosphine oxides. During the last decade, diglycolamide-based extractants such as N,N,N′,N′-tetraoctyl diglycolamide (TODGA) and N,N,N′,N′-tetra-2-ethylhexyl diglycolamide (TEHDGA) have received considerable attention due to overwhelmingly favourable extraction and stripping efficiencies of minor actinides from different types of transuranium (TRU) wastes. The focus o...

Aqueous Partitioning of Minor Actinides by Different Processes

A number of neutral extractants containing the P(0)(CH2)nC(0)N raolety were evaluated for their ability to extract Am from nitric acid and their selectivity for Am over Fe and selected fission products. Extractants containing alkoxy, alkyl, and aryl substltuents were evaluated. Tetrachloroethylene was used as a diluent. Fission products selected for study were Y, Zr, Mo, Tc, Ru, Rh, Pd, La, Ce, Pr, Nd, Sra, and Eu. The conclusions drawn were that the most efficient and selective Am extractant contains a single carbon bridging group, one or two phenyl groups attached to the phosphorus atom and l9obutyl groups attached to the amide nitrogen.

EXTRACTION OF Am FROM NITRIC ACID BY CARBAMOYL-PHOSPHORYL EXTRACTANTS: THE INFLUENCE OF SUBSTITUENTS ON THE SELECTIVITY OF Am OVER Fe AND SELECTED FISSION PRODUCTS

N,N,N  ′,N  ′-Tetraoctyl-3-oxapentanediamide (TODGA) in n-dodecane solvent was studied to clarify the characteristics of third phase formation in extraction of lanthanides(III) and to modify the solvent. The loading capacity of the solvent was found to be dependent not only on the aqueous acidity and the temperature but also on molecular size of alkane solvent and kinds of aqueous anions. The loading capacity of 0.1 M TODGA-n-dodecane was 0.008 M Nd(III) with an aqueous phase of 3 M HNO3. Addition of N,N,-dihexyloctanamide (DHOA) of more than 0.5 M to the solvent eliminated the third phase in the extraction of Nd(III). The modified solvent of 0.1 M or 0.2 M TODGA with 1 M DHOA exhibited a satisfactorily high metal loading and slightly lower extractability for Nd(III) and moderately higher extraction of HNO3 than neat TODGA.

MODIFICATION OF TODGA- n -DODECANE SOLVENT WITH A MONOAMIDE FOR HIGH LOADING OF LANTHANIDES(III) AND ACTINIDES(III)

A combined anion exchange-extraction chromatography process for the recovery and purification of americium from molten salt extraction residues has been successfully demonstrated on a laboratory and pilot plant scale. The extraction chromatography process uses dihexyl- N , N -diethylcarbamoylmethylenephosphonate sorbed on Amberlite ® XAD-4 resin. The process effectively separates and purifies americium from impurities such as aluminum, calcium, chloride, copper, fluoride, iron, lead, magnesium, plutonium, potassium, sodium and zinc. A total of 100 g of americium oxide was produced during pilot plant testing. The product oxide contained 96.5 wt.% AmO 2 , with 0.085 wt.% Pu and less than 0.15 wt.% of any other individual impurity element.

Americium recovery and purification using a combined anion exchange‐extraction chromatography process

The US Department of Energy awarded Oak Ridge National Laboratory a program to develop a cost-risk-benefit analysis of partitioning long-lived nuclides from waste and transmuting them to shorter lived or stable nuclides. Two subtasks of this program were investigated at Rocky Flats. In the first subtask, methods for solubilizing actinides in incinerator ash were tested. Two methods appear to be preferable: reaction with ceric ion in nitric acid or carbonate-nitrate fusion. The ceric-nitric acid system solubilizes 95% of the actinides in ash; this can be increased by 2 to 4% by pretreating ash with sodium hydroxide to solubilize silica. The carbonate-nitrate fusion method solubilizes greater than or equal to 98% of the actinides, but requires sodium hydroxide pretreatment. Two additional disadvantages are that it is a high-temperature process, and that it generates a lot of salt waste. The second subtask comprises removing actinides from salt wastes likely to be produced during reactor fuel fabrication and reprocessing. A preliminary feasibility study of solvent extraction methods has been completed. The use of a two-step solvent extraction system - tributyl phosphate (TBP) followed by extraction with a bidentate organophosphorous extractant (DHDECMP) - appears to be the most efficient for removing actinides from saltmore » waste. The TBP step would remove most of the plutonium and > 99.99% of the uranium. The second step using DHDECMP would remove > 99.91% of the americium and the remaining plutonium (> 99.98%) and other actinides from the acidified salt waste. 8 figures, 11 tables.« less

Removal of actinides from selected nuclear fuel reprocessing wastes

The past 20 years have seen significant growth in the development and application of pyrochemical technology for the processing of plutonium. Non-aqueous high temperature processes offer key advantages over conventional hydrometallurgical systems for particular feed stocks and specific applications. Major processes in use today include 1. (1) direct oxide reduction for the conversion of PuO2 to metal 2. (2) molten salt extraction for americium removal from plutonium metal 3. (3) molten salt electrorefining for plutonium metal purification 4. (4) hydriding to remove plutonium from host substrates. Current major pyrochemical processes ranging from the classical calcination-fluorination-“bomb reduction” sequence to new techniques under development are reviewed. Each process is presented and brief descriptions of the production equipment are given.

Review of major plutonium pyrochemical technology

The chemistry of the mixed solvent extractants, dibutyl-N, N-diethylcarbamoylmethylphosphonate (DBDECMP)-tri -£-butyl phosphate (TBP), using carbon tetrachloride diluent, was studied for extracting americium (III) from nitric acid at various concentrations and temperatures. Synergistic extraction of americium was observed and an extraction mechanism is proposed based on slope analysis, infrared studies, and temperature variation studies. Similar studies were performed with dibutyl-N, N-diethylcarbamoylphosphonate (DBDECP)

The extraction of americ1um (iii) by mixtures of tributylphosphate with dibutyl-n,n-diethylcarbamoyl- phosphonate and -methylphosphonate (dbdecp and dbdecmp)

Process for the removal of plutonium polymer and ionic actinides from aqueous solutions by absorption onto a solid extractant loaded on a solid inert support such as polystyrenedivinylbenzene. The absorbed actinides can then be recovered by incineration, by stripping with organic solvents, or by acid digestion. Preferred solid extractants are trioctylphosphine oxide and octylphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide and the like.

James D. Navratil

Papers

Americium recovery and purification using a combined anion exchange‐extraction chromatography process

Removal of actinides from selected nuclear fuel reprocessing wastes

Review of major plutonium pyrochemical technology

The extraction of americ1um (iii) by mixtures of tributylphosphate with dibutyl-n,n-diethylcarbamoyl- phosphonate and -methylphosphonate (dbdecp and dbdecmp)

Actinide recovery process