Role of Acetohydroxamic Acid in Selective Extraction of Technetium and Uranium Employing N,N-Dihexyloctanamide as Extractant

Straight chain N,N-dihexyloctanamide (DHOA) has been identified as a promising alternate extractant to tributyl phosphate (TBP) for the reprocessing of uranium based spent fuels. The present work compares extraction behavior of technetium using DHOA and TBP solutions in n-dodecane, under varying experimental conditions such as acidity (0.5–6 M HNO₃); extractant concentration (1.1 and 1.5 M), and uranium loading (50 g/L, relevant for Pu rich spent fuel feed solutions). The effect of acetohydroxamic acid concentration on U, Pu, Np, and Tc extraction behavior has also been investigated. Pu(IV)-AHA interaction and its influence on extraction using TBP and DHOA extractants has been studied spectrophotometrically. The experimental data suggest that 1.1 M DHOA is better than 1.1 M TBP with respect to co-extraction of Tc and U, and U decontamination with respect to Np/Pu.     

Comparison of Extraction Behavior of Neptunium from Nitric Acid Medium Employing Tri-n-Butyl Phosphate and N,N-dihexyl Octanamide as Extractants

Extraction behavior of neptunium has been compared for tri-n-butyl phosphate (TBP) and N,N-dihexyl octanamide (DHOA) extractants as a function of nitric acid concentration (0.5 ? 6 M HNO₃), uranium loading (50 and 300 g/L relevant to Pu rich fast reactor and Pressurized Heavy Water Reactor, PHWR spent fuels, respectively), and in the presence of oxidizing and reducing agents. These studies suggest the possibility of co-recovery of U(VI), Pu(IV) and Np(IV) from spent fuel dissolver solutions (of Pu rich fuels) employing DHOA as extractant.     

Radiolytic Degradation Studies on N,N‐dihexyloctanamide (DHOA) under PUREX Process Conditions

The radiolytic stability of DHOA, a high molecular weight N,N‐dialkyl amide has been investigated to evaluate its performance under PUREX process conditions vis–a–vis TBP. Gas chromatographic studies revealed the presence of caprylic acid, dihexylamine and dihexylketone in irradiated DHOA. Batch distribution studies of Pu, U, and fission products (¹⁴⁴Ce, ^103,106Ru, and ¹³⁷Cs) were carried out using the irradiated samples of 1.1 M DHOA and TBP in n‐dodecane, which showed significant retention of Pu, U, and fission products in the irradiated TBP as compared to that of DHOA even after successive contacts with the stripping solutions. Typically at 60 M Rad dose, the Pu content for DHOA was 1.4 mg/L after three contacts with 0.5 M HNO₃, and that for TBP was ～24 mg/L. White precipitate was observed at the interface during the stripping of Pu (with 0.5 M HNO₃) from the loaded irradiated TBP phase. The DHOA system, on the other hand showed no such problem during the stripping cycle but there was an increase in the density and viscosity for the irradiated DHOA.     

Comparison of Hydrometallurgical Parameters of N,N‐Dialkylamides and of Tri‐n‐Butylphosphate

Straight‐chain N,N‐dihexyloctanamide (DHOA) and branched‐chain N,N‐di(2‐ethylhexyl)isobutyramide (D2EHIBA) have been identified as promising alternatives to tri‐n‐butylphosphate (TBP) for the reprocessing of spent uranium based fuels, and selective extraction of ²³³U from irradiated thorium fuels, respectively. The present work deals with the effects of different hydrodynamic parameters such as viscosity, density, and interfacial tension (IFT) on the phase‐separation time (PST) under uranium and thorium loading conditions. The IFT values have been determined under varying experimental conditions such as the aqueous nitric acid concentration, n‐dodecane purity, ligand concentration, and thorium/uranium loading conditions. These studies have suggested that the quality of n‐dodecane affects the IFT values of different solutions. The IFT values of D2EHIBA changed marginally (23.3 ± 0.9 mNm^?1) against THOREX feed solution for the wide range of D2EHIBA concentration (0.1–1.0 M). However, IFT, viscosity, and PST values increased with uranium loading of 1.1 M DHOA. These studies suggested that a lower phase‐disengagement rate with increased uranium loading was mainly due to the increased viscosity of the loaded 1.1 M DHOA solution.     

Studies on the Development of a Flow-Sheet for AHWR Spent Fuel Reprocessing Using TBP

The present paper describes the results of solvent extraction studies carried out in batch mode to collect data on distribution of uranium, plutonium, and thorium using 5% TBP in n-dodecane. Extraction studies are carried out from feed solutions having bulk thorium containing aluminum and fluoride ions in ～3.00–4.00 M nitric acid at concentration levels anticipated in feed solutions during Advanced Heavy Water Reactor (AHWR) spent fuel reprocessing. Studies are carried out under varied experimental conditions. Parameters such as organic to aqueous phase ratio during extraction, concentration of nitric acid for scrubbing co-extracted thorium from loaded organic phase etc., are studied in detail. Hydroxylamine nitrate is selected for reductive stripping of plutonium in preliminary studies. Reagent mixture containing 0.30 M HAN + 0.60 M HNO₃ and 0.20 M N₂H₄ is found to be optimum for plutonium partitioning. This paper also describes the extraction and stripping of uranium and plutonium in co-current mode. The extraction behavior of relevant fission products is studied from a simulated feed solution. A preliminary study on a few commercially available reducing agents is also included. These data are useful in developing a flow-scheme for the recovery of uranium and plutonium from spent fuel originating from AHWR.     

Ion‐Exchange Separation of Pu from Macro Concentration of Th in HNO3 Medium

Abstract

Sorption behavior of Th and Pu from anion‐ as well as cation‐exchange resin was investigated from nitric acid medium by both batch and column methods. The anion‐exchange studies involved anionic nitrate complexes of Pu⁴⁺ and Th⁴⁺ sorbed onto DOWEX 1x4 resin (50–100 mesh), and the cation‐exchange studies involved the sorption of Pu³⁺ and Th⁴⁺ onto BIORAD AG 50Wx8 (50–100 mesh) or DOWEX 50Wx4 (50–100 mesh) resin. The batch data gave a separation factor (K _d,Pu/K _d,Th) of 22 for the anion‐exchange method and 0.017 for the cation‐exchange method at 3 and 2 M HNO₃, respectively. A two‐stage ion‐exchange separation method was developed for the quantitative separation of Pu (8 g/L) from a macro amount of Th (200 g/L) in nitric acid medium. The first step involved the quantitative sorption of plutonium from the mixture while about 90% of Th could be washed in 6 column volumes. The plutonium, eluted (as Pu³⁺) using 0.5 M HNO₃ + 0.2 M hydrazinium nitrate (HN) + 0.2 M hydroxyl ammonium nitrate (HAN), and the residual (～10%) Th were subsequently loaded onto a cation‐exchange column in the second step. Greater than 99% Pu was recovered with 2 M HNO₃ (in ～8 column volumes) containing 0.2 M HN + 0.2 M HAN. The final elution of thorium from the cation‐exchange column was achieved in about 6 column volumes of 1 M α‐hydroxy isobutyric acid. A (Pu, Th)O₂ fuel scrap sample was dissolved in 16 M HNO₃ containing 0.005 M HF and was used subsequently as the feed for the anion‐exchange column. The eluted Pu was subsequently loaded onto a cation‐exchange column for final purification. The recovery of plutonium and thorium was found to be >99% and >98%, respectively, while the respective decontamination factors were estimated to be 215 and 292.     

Extraction of Uranium(VI) and Plutonium(IV) with Tetra-Alkylcarbamides

ABSTRACT

The use of tetra-alkylcarbamides as novel extractants for the separation of uranium(VI) and plutonium(IV) by solvent extraction from spent nuclear fuels is investigated in this study. Batch extraction experiments show that tetra-alkylcarbamides strongly extract U(VI) with high distribution ratios. Plutonium(IV) can be co-extracted with U(VI) at high nitric acid concentration, while high U(VI)/Pu(IV) selectivities can be reached at lower acidity. Loading capacity experiments with high uranium concentrations show that alkyl chains longer than butyl are necessary to avoid third phase formation. Nevertheless, the viscosity of uranium-loaded solvents gets too high with alkyl chains longer than pentyl. Overall, this study shows that with TPU extractant (with four pentyl chains), an efficient co-extraction of uranium and plutonium can be reached (D_U,Pu > 1) for a concentration of nitric acid higher than 4 mol?L^?1, while the partition between uranium(VI) and plutonium(IV) could be operated even at 2 mol?L^?1 nitric acid without redox chemistry.     

Optimization Studies for the Recovery of Thorium from Advanced Heavy Water Reactor High Level Waste (AHWR-HLW) Solutions Using Green Solvents

An Advanced Heavy Water Reactor (AHWR) has been specifically designed to exploit Th/²³³U as fuels. The reprocessing is focused mainly on the recovery of ²³³U and Pu from the spent fuels leaving bulk of Th (～100 g/L) in the High Level Waste (HLW) solutions. No systematic attempts have been made so far to identify suitable solvents for the recovery of thorium from high level waste (HLW) solutions generated after AHWR spent fuel reprocessing. Tri-n-butyl phosphate (TBP), though the work horse for nuclear fuel reprocessing as an extractant, suffers from the serious limitation of third-phase formation during the extraction of macro concentrations of thorium. Two straight chain dialkyl amides such as N,N-dihexyl octanamide (DHOA), and N,N-dihexyl decanamide (DHDA) as well as TBP were evaluated for the recovery of the thorium from AHWR-HLW solutions. Attempts were made to identify suitable solvent (extractant + diluent) system and optimize the conditions for the recovery of thorium from HLW solutions. Selectivity of the solvents was examined for thorium extraction over fission products/structural materials under AHWR raffinate solutions. Counter-current centrifugal contactor runs were also carried out on simulated waste solutions to validate the optimized conditions for the recovery of thorium from the simulated AHWR waste solutions.     

EXTRACTION OF PLUTONIUM (III) BY TBP IN PRESENCE OF URANIUM

ABSTRACT

The distribution ratio (D) values for the extraction of plutonium (III) from nitric acid medium into 30% TBP in n-dodecane saturated with uranium(VI) (0% to 80%) were determined. For a fixed saturation of TBP with uranium, the D values for Pu(III) were found to increase with increase in nitric acid concentration (1M to 5M). At a fixed nitric acid concentration, the D values were found to decrease with increase in loading of TBP with uranium. The D values for the extraction of Pu(III) using 20% TBP in n-dodecane and 30% TBP in n-paraffin at 80% uranium saturation were also determined The distribution data was least squares analysed against concentration of HNO₃ as well as percentage saturation of TBP with uranium and the coefficients obtained are reported. For all these extraction systems, D values for U(VI) were also determined.     

Comparative Evaluation of Tri-n-butyl Phosphate (TBP) and Tris(2-ethylhexyl) Phosphate (TEHP) for the Recovery of Uranium from Monazite Leach Solution

Separation of U(VI) from Th(IV) and rare earth elements (REEs) present in monazite leach solution (nitric acid medium) has been studied using tris(2-ethylhexyl) phosphate (TEHP) and tri-n-butyl phosphate (TBP) dissolved in n-paraffin as solvents under varying experimental conditions such as nitric acid, extractant and metal ion concentrations etc. There is an increase in distribution ratio of U(VI) (D _U ) with increase in aqueous phase acidity up to 5 M HNO₃ beyond which a decrease is observed. Typically for 1 × 10^?3 M U(VI), the D_U values increase from 8 (0.5 M HNO₃) to 80 (5 M HNO₃) for 1.1 M TEHP, and from 2 (0.5 M HNO₃) to 43 (5 M HNO₃) for 1.1 M TBP in n-paraffin. The separation factors of U(VI) (β: D_U/D_M) over metal ions (M) such as Th(IV) and Y(III) (chosen as a representative of heavy REEs) are better for TEHP than TBP at all nitric acid concentrations. Batch solvent extraction data have been used to construct the McCabe-Thiele diagrams for the recovery of U(VI) employing TEHP as the extractant. A process flow sheet has been proposed with 0.2 M TEHP in n-paraffin as solvent for the recovery of U(VI) from simulated monazite leach solution in HNO₃ medium.     

Recovery of Neptunium from Highly Radioactive Waste Solutions of Purex Origin Using Tributyl Phosphate

Abstract

The present work deals with the extraction of neptunium into the TBP/dodecane phase under conditions relevant to highly radioactive waste solutions, along with uranium and plutonium, by oxidizing it to the hexavalent state using 0.01 M K₂Cr₂O₇ and subsequently recovering it by selective stripping. Three types of simulated HLW solutions, namely sulfate-bearing (SB, in ～0.3 M HNO₃) and non sulfate wastes originating from the reprocessing of fuels from pressurised heavy water reactors (PHWR) and fast breeder reactors (FBR) (both in 3.0 M HNO₃), have been employed in this study. Very high extraction of U(VI), Np(VI), and Pu(VI) was obtained from PHWR and FBR-HLW solutions, whereas extraction was less but reasonably high from the SB-HLW solution. The uptake of cerium at tracer level concentrations in the millimolar range (encountered in HLW solutions) and from the simulated HLW solutions containing 0.01 M K₂Cr₂O₇ by 30% TBP has shown that its extraction takes place only at tracer level concentrations and not at millimolar levels. The stripping of the metal ions from the loaded organic phase was done with a mixture of 0.01 M ascorbic acid and 0.1 M H₂C₂in 2.0 M HNO₃ at organic to aqueous phase ratios of 1:1, 2:1, and 4:1. Quantitative recovery of neptunium and plutonium was achieved. Based on these results, a scheme was formulated for the recovery of neptunium, and it was tested using the actual high level waste solution originating from the reprocessing of research reactor fuels.     

Uranium Permeation Studies from Nitric Acid Medium across Supported Liquid Membrane Impregnated with PC88A and its Mixtures with Neutral Oxodonors in n-paraffin as Carriers

The permeation of U(VI) from nitric acid medium using supported liquid membrane (SLM) technique has been studied employing varying compositions of feed (uranium concentration and acidity), carrier, and receiving phase. Microporous polytetrafluoroethylene (PTFE) membranes were used as a solid support and 2-ethylhexyl phosphonic acid mono-2-ethylhexyl ester (PC88A) either alone or as a mixture of neutral donors like tri-n-butyl phosphate (TBP), tris(2-ethylhexyl) phosphate (TEHP), and tri-n-octyl phosphine oxide (TOPO) dissolved in n-parrafin as the carrier. Oxalic acid/Na₂CO₃ solutions were used as the receiving phase. The permeability coefficient (P) of U(VI) decreased with increased nitric acid concentration up to 3 M HNO₃ and thereafter increased up to 5 M HNO₃. Uranium permeation was also investigated from its binary mixtures with other metal ions such as Zr(IV), Th(IV), and Y(III) at 2 M HNO₃ employing 0.1 M PC88A/n-paraffin as the carrier, and 0.5 M oxalic acid as the receiver phase. The presence of neutral donors in the carrier solution enhanced the permeation of U(VI) across the SLM in the following order: TEHP ～ TBP > TOPO using 0.1 M oxalic acid as receiver phase. There was significant enhancement in uranium transport for feed acidity ≤2 M HNO₃ employing 1 M Na₂CO₃ as the receiver phase. These studies suggested that 0.1 M PC88A and 0.5 M oxalic acid as carrier and receiver phases appear suitable for selective and faster transport of uranium from the uranyl nitrate raffinate (UNR) waste solutions.     

Pseudo-Emulsion Based Hollow Fiber Strip Dispersion Technique (PEHFSD): Optimization,Modelling and Application of PEHFSD for Recovery of U(VI) from Process Effluent

Abstract

Pseudo emulsion based hollow fiber strip dispersion technique (PEHFSD) is the first of its kind ever explored in radioactive environment for the extraction of uranium from acidic process streams. Permeation of U(VI) was investigated as a function of various experimental variables such as hydrodynamic conditions (flow rates of pseudo-emulsion and feed phase), concentration of U(VI) in the feed phase, concentration of tri-n-butylphosphate (TBP), HNO₃ concentration in feed phase, O/A ratio and 0.01 M HNO₃ as stripping agent in pseudo-emulsion phase. The mass transfer coefficient was calculated from the experimental results and a model has been presented for determining mass transfer characteristics. PEHFSD has been demonstrated for separation/recovery of uranium from oxalate supernatant waste generated during plutonium precipitation by oxalic acid. PEHFSD and HFSLM (hollow fiber supported liquid membrane) performance has been compared in order to analyze the efficiency of the technique.     

Studies Related to the Processing of U–Zr and U–Pu–Zr Metallic Fuels Using Tri-iso-amyl Phosphate (TiAP) as Extractant

Metallic fuel is reprocessed by a compact nonaqueous pyrochemical process. The present study explores the possible uses of an aqueous reprocessing-based method as an alternative to the pyro-processing of metallic fuels. Solvent extraction studies in the batch mode were carried out for both U–Zr- and U–Pu–Zr-based metal alloy systems. Earlier studies carried out in our laboratory have established that Tri-iso-amyl Phosphate (TiAP) is a promising extractant for the reprocessing of spent fuels. In the present study, the distribution data has been generated for the extraction of uranium and zirconium as a function of equilibrium aqueous phase metal ion and nitric acid concentration with TiAP and Tri-n-butyl Phosphate (TBP)-based solvents. These studies indicate the formation of a third phase with zirconium in the presence of uranium with TBP under certain experimental conditions whereas it was not encountered with the TiAP system. Flow sheet for the co-extraction and co-stripping of heavy metal ions by 1.1M TiAP and 1.1M TBP in n-dodecane from U–Zr as well as U–Pu–Zr feed solutions in stage-wise mode has been evaluated. Percentage extraction and stripping of metal ions were calculated stage-wise and the results are discussed.     

Tuning the Extractive Properties of Purex Solvent using Room Temperature Ionic Liquid

An Aliquat-336 based ionic liquid, namely, tri-n-octylmethylammonium bis(2-ethylhexyl)phosphate ([A3636]⁺[DEHP]^?) was prepared and studied for the extraction of U(VI), Pu(IV), and Am(III) from nitric acid medium. Since the ionic liquid, [A336]⁺[DEHP]^? was miscible in n-dodecane (n-DD), the extraction of these actinides in the PUREX solvent, 1.1 M tri-n-butylphosphate (TBP) in n-dodecane (n-DD), was investigated in the presence of small concentrations of ionic liquid. The distribution ratio of U(VI) and Am(III) in 0.03 M [A336]⁺[DEHP]^?/n-DD decreased with increase in the concentration of nitric acid; whereas the extraction of Pu(IV) initially increased, it reached a maximum at 4 M nitric acid followed by the decrease. The extraction of actinides in ionic liquid medium decreased in the order Pu(IV) > U(VI) >> Am(III), indicating the feasibility of modifying the extractive properties of TBP/n-DD to favor Pu(IV) extraction. Therefore, the extraction of Pu(IV) in a solution of TBP – [A336]⁺[DEHP]^? in n-DD was also studied. The distribution ratio of Pu(IV) increased with increase in the concentration of ionic liquid and decrease in the concentration of TBP in organic phase. The distribution ratio of Pu(IV) determined in the presence of small concentration of ionic liquid in 1.1 M TBP/n-DD was always much higher than that observed in 1.1 M TBP/n-DD. In contrast to this, the distribution ratio of U(VI) decreased by the addition of ionic liquid and Am(III) was inextractable even in the presence of ionic liquid.     

Separation and Recovery of Uranium,Neptunium, and Plutonium from High Level Waste Using Tributyl Phosphate: Countercurrent Studies with Simulated Waste Solution

ABSTRACT

The present work deals with countercurrent extraction studies on the partitioning of uranium, neptunium, and plutonium using 30% tributyl phosphate (TBP) from simulated high level waste solution generated during reprocessing of spent uranium fuel from pressurized heavy water reactors. The oxidation states of neptunium and plutonium were adjusted either by 0.01 M potassium dichromate or 0.01 M dioxovanadium ion. Neptunium and plutonium, extracted in the TBP phase, were stripped together using a mixture containing 0.05 M ascorbic acid and 0.25 M hydrogen peroxide in 2.0 M nitric acid solution. Although dioxovanadium ion is more effective for proper adjustment of the oxidation states of plutonium and neptunium, subsequent recovery of these actinides from loaded TBP is better if potassium dichromate is used for the valency adjustment. Results of the stagewise analysis of extraction and stripping of actinides using mixer-settlers are presented.     

Note: Effect of Temperature in the Extraction of Uranium and Plutonium by Triisoamyl Phosphate

Abstract

The extraction of uranium(VI) by triisoamyl phosphate (TiAP) has been studied to derive the thermodynamic parameters such as entropy change and the free-energy change. The extraction of U(VI) and Pu(IV) has also been studied with 1.1 M solutions of tri-n-butyl phosphate (TBP), tri-n-amyl phosphate (TAP), and TiAP as a function of temperature. While the enthalpy of U(VI) extraction was found to be exothermic, the enthalpy for the extraction of Pu(IV) was always found to be endothermic. The temperature at which the distribution ratios of U(VI) and Pu(IV) cross each other (the temperature of inversion) has been derived for TBP, TAP, and TiAP, and the results reveal the lowest temperature of inversion occurs for TiAP. The results indicate the advantage of TiAP as an extractant in avoiding plutonium reflux during the PUREX process involving high plutonium feed solutions, in addition to lower aqueous solubility, freedom from the third-phase formation problem, lower degradation, and better economics.     

1-Phenyl-3-methyl-4-benzoyl-pyrazolone-5: A Promising Extractant for Plutonium

Abstract

Pyrazolones and isoxazolones have been found to be promising extractants for metal ions, particularly from strong acidic media and in the presence of complexing anions. Extraction constants (log kex) in toluene medium at 25°C for PuX₄ species, where X = 1-phenyl-3-methyl-4-acetyl-pyrazolone-5 (HPMAP), 1-phenyl-3-methyl-4-benzoyl-pyrazolone-5 (HPMBP), or 1-phenyl-3-methyl-4-(3:5-dinitrobenzoyl)pyrazolone-5 (HPMDP), are determined as 11.35 ± 0.04, 12.89 ± 0.03, and 12.73 ± 0.02, respectively. These values are comparable to the corresponding value for 3-phenyl-4-benzoyl-5-isoxazolone (HPBI) and several order of magnitude larger than that for 2-thenoyltrifluoroacetone (HTTA). A systematic study is carried out to investigate the extraction behavior of these β-diketones toward plutonium present in the analytical waste solution obtained during the determination of uranium in a (U, Pu) fuel sample by the Davies Gray method. Whereas 0.3 M HPMBP extracts ≥85% of the plutonium present in a single step, maximum extraction observed with other reagents is ?0.1% HTTA, 0.3% HPMAP, and 2.5% HPBI. The extraction of plutonium increases with different diluents in the order n-dodecane < n-hexane < CHCl₃ < CCl₄ < toluene. Extracted plutonium is quantitatively stripped with either 10 M HNO₃ or 1:1 HCl + 0.1 M hydroquinone.     

Extraction of uranium(VI) from nitric acid solutions using N,N-dihexyloctanamide in ionic liquids: Solvent extraction and spectroscopic studies

Liquid-liquid extraction studies of uranium(VI) were carried out from nitric acid medium using di-n-hexyloctanamide (DHOA) in several room-temperature ionic liquids (RTIL). The extraction of the metal ion as a function of nitric acid concentration showed different trends based on the alkyl substituents of the RTIL. While the D_U values decreased with increasing HNO₃ concentration (up to ca. 0.5 M) with [C₄mim][NTf₂] and [C₆mim][NTf₂], almost no change was seen with [C₈mim][NTf₂]. This suggested that while a cation-exchange mechanism is operative with the former diluents, which was to a much lower extent for the latter. The extracted species were found to contain about 2 molecules of DHOA from the feed solutions containing either 0.01 M and 4 M HNO₃, which was arrived at from the ligand-concentration-variation experiments. Recycling studies were also performed by carrying out stripping and radiolytic stability studies. The nature of the extracted species as ascertained from the UV-visible spectrophotometry studies indicated similarity between the extracts obtained in RTIL medium, which were entirely different from that observed with n-dodecane as the diluent.     

Recovery of Plutonium from Analytical Laboratory Waste using Hollow Fiber Supported Liquid Membrane Technique

Plutonium from analytical laboratory waste was recovered on liters scale using Hollow Fiber Supported Liquid Membrane (HFSLM) technique using 30% TBP/n-dodecane as the carrier. The technique is faster, gives lower radiation exposure to the working personnel, and generates lower volume of secondary waste as compared to traditional precipitation/ion-exchange technique. The recovery of plutonium was carried out in two stages from waste containing a mixture of 3.22 g/L Pu, 110 g/L U, and 60.2 mg/L Am. In the first stage, >96% Pu(IV) and U(VI) were transported into the receiver phase in two hours. The Am(III) contamination in the Pu(IV) and U(VI) fraction was <0.1%. In the subsequent stage, plutonium was reduced to Pu(III) and U(VI) was selectively transported in to the receiver phase. In this method, a pure fraction of uranium was also obtained along with pure fraction of plutonium. The purity of plutonium fraction was confirmed by ICP-AES analysis.