J. P. Shukla

Extraction of plutonium(IV), uranium(VI) and some fission products by di-n-hexyl sulphoxide

The extraction of nitric acid, plutonium, uranium and fission products such as zirconium, ruthenium and europium has been investigated using di-n-hexyl sulphoxide in Solvesso-100. Results indicate that Pu(IV), U(VI), Zr(IV) and Ru NO(III) are extracted as disolvates, whereas Eu(III) is extracted as the trisolvate. The absorption spectra of the plutonium(IV) and uranium(VI) complexes extracted are similar to those of the species extracted by TBP which indicate the similarity of the species involved. Preliminary studies show that irradiated di-n-hexyl sulphoxide extracts zirconium to a smaller extent than irradiated TBP suggesting the use of long chain aliphatic sulphoxides as promising extractants for the recovery of plutonium in high radiation fields.

Macrocycle-Mediated Selective Transport of Plutonium(IV) Nitrate through Bulk Liquid and Supported Liquid Membranes Using Dicyclohexano-18-crown-6 as Mobile Carrier

Abstract Macrocycle-facilitated transport of Pu(IV) against its concentration gradient from aqueous nitric acid solutions across an organic bulk liquid membrane (BLM) and a thin-sheet supported liquid membrane (SLM) containing dicyclohexano-18-crown-6 as the mobile carrier and toluene as the membrane solvent was investigated. Extremely dilute and moderately concentrated plutonium nitrate solutions in about 3 mol/dm3 nitric acid generally constituted the source phase. Plutonium transport appreciably increased with stirring of both the source and recovery phases and also with the carrier concentration in the organic phase up to 0.3 mol/dm3 DC18C6 and then decreasing with higher concentrations of the carrier. Enhanced acidity of the aqueous source phase beyond 4 mol/dm3 HNO3 seriously affected the stripping of the cation. Among the several reagents tested, dilute sodium carbonate (0.5 mol/dm3) served efficiently as the strippant. Enka Accurel thin flat-sheet polypropylene (PP) films were used as the solid su...

Solvent Extraction of Plutonium(IV), Uranium(VI), and Some Fission Products with Di-n-octylsulfoxide

Abstract Extraction behavior of plutonium(IV), uranium(VI), and some fission products from aqueous nitric acid media with di-n-octylsulfoxide (DOSO) has been studied over a wide range of conditions. Both the actinides are extracted essentially completely, whereas fission product contaminants like Zr, Ru, Ce, Eu, and Sr show negligible extraction. The absorption spectra of sulfoxide extracts containing either Pu4+ or UO2 2+ indicate the species extracted from nitric acid into the organic phase to be Pu(NO3)4. 2DOSO and UO2(NO3)2. 2DOSO, respectively. Extraction of these actinides decreases with increasing temperature, indicating the extraction to be exothermic. DOSO extracts plutonium and uranium better than di-n-hexylsulfoxide (DHSO) under all condition and is also more soluble in aromatic diluents than the latter. The effect of gamma radiation on the extraction properties of DOSO is found to be similar to that of DHSO.

Tri-iso-amyl phosphate (TAP): An alternative extractant to tri-butyl phosphate (TBP) for reactor fuel reprocessing

Tri-iso-amyl phosphate (TAP), an indigenously prepared extractant was utilized for reactor fuel reprocessing and compared with tri-butyl phosphate (TBP) and tri-n-hexyl phosphate (THP). The potential of these extractants was found to be in the order TAP>THP>TBP by calculating the acid uptake value (KH). The effect of various parameters such as solvent degradation due to acid hydrolysis, radiation effect, decontamination factor and phase separation were investigated and it was found that TAP was always a better extractant in comparison to THP and TBP. In addition to this, the extraction of fission product contaminants such as 144Ce, 137Cs, 106Ru, 95Zr was almost negligible, even at very high nitric acid concentrations in the aqueous phase, indicating the potential application of TAP in actinide partitioning. Sodium carbonate solution or acidified distilled water was a good strippant for U(VI), similarly, uranium(IV) nitrate stripped Pu(IV) from the organic phase.

TETRA-BUTYL-MALONAMIDE AND TETRA-ISOBUTYL MALONAMIDE AS EXTRACTANTS FOR URANIUM(VI) AND PLUTONIUM(IV)

ABSTRACT Two new symmetrical diamides, namely straight-chain alkyl substituted neutral tetra-butyl-malonamide(TBMA) and sterlcally hindered branched-chain alkyl substituted tetra-isobutyl malonamide(TIBMA) were synthesised, characterised and used for the extraction of U(VI) and Pu(IV) from nitric acid media into n-dodecane. Both the cations were found to be extracted as their disolvates. Interestingly TBMA extracted more efficiently than TIBMA but afforded poor selectivity for Pu/U separation. The thermodynamic parameters involved in the extraction, determined by the temperature variation method, indicated the reactions in all cases to be enthalpy favoured. Entropy was found to be counteracting the extraction of U(VI) and favouring the extraction of Pu(IV). The recovery of diamides from the loaded actinides could be easily accomplished by using dilute oxalic acid or dilute U(IV) as the strippant for Pu(IV) and using dilute Na2C03 as that for U(VI).

Effect of Solvent Type on Neutral Macrocycle-Facilitated Transport of Uranyl Ions across Supported Liquid Membrane using Dicyclohexano-18-crown-6 as Carrier

The diffusion-Iimited transport of neutral macrocycle-facilitated uranyl ion permeation across thin sheet supported liquid membrane (SLM) has been investigated to quantify the membrane solvent effects on the important parameters such as cation flux and diffusion coefTicients. Also, results obtained with the aim of optimizing the permeation fluxes, physical stability (loss of carrier) and chemical stabihty of SLMs using various aromatic and aliphatic diluents are evaluated in detail. The uranyl ion transport is tested across SLM using dicyclohexano-18-crown-6 as membrane carrier. The diluents studied include Chloroform, 1,2-dichloroethane, dichloromethane, o-dichlorobenzene, toluene, benzene, xylene, and a mixture containing varying proportions of toluene and o-dichlorobenzene. Among the various aliphatic and aromatic diluents tested, the chlorinated hydrocarbons such as Chloroform, 1,2-dichloroethane, dichloromethane, afford much poor permeability compared to aromatic diluents like toluene and o-dichlorobenzene. Among the various mixtures tested for membrane impregnation, a good stability and efficient uranyl ion transport are achieved by using toluene + o-dichlorobenzene admixtures.

Thermodynamics of extraction of U(VI), Np(IV) and Pu(IV) with some long chain aliphatic sulphoxides

Abstract Distribution data for U(VI), Np(IV) and Pu(IV) from 2 M nitric acid medium with 0,2 M di-n-hexyl sulphoxide (DHSO) and di-n-octyl sulphoxide (DOSO) in Solvesso-100 have been obtained in the temperature range 20–50°C. From these data, the enthalpy, entropy and free energy changes associated with their extraction were evaluated. Extraction of Np(IV) and Pu(IV) with both sulphoxides is favoured by negative enthalpy and positive entropy changes whereas the extraction of U(VI) is favoured only by high negative enthalpy change. This behaviour has been explained as arising due to the higher hydration of Np4+ and Pu4+ ions as compared to the UO22+ ion.

Influence of the nature of organic diluents on the extraction of uranium(VI) by bis(2-ethylhexyl) sulfoxide from nitric acid solutions

The extraction of uranium(VI) from nitric acid solutions by bis(2-ethylhexyl) sulfoxide (BESO) has been examined using sixteen inert organic diluents in order to establish the correlation between its distribution coefficient and some physico-chemical properties of the diluents. The extracted solvated species is shown to be UO2(NO3)2·2BESO, irrespective of their nature. The extraction rate is lower in halogen substituted hydrocarbons as compared to the other diluents used. Among benzene derivatives, extraction is found to decrease with number of substituted methyl groups. Extraction efficiency decreases as the organic solvent is varied in the order: benzene>nitromethane>toluene> >nitrobenzene>cyclohexane>p-xylene>monochlorobenzene>dodecane>o-dichlorobenzene> >hexane>decalin>1,2-dichloroethane>1,1,1-trichloroethane>carbon tetrachloride> >tetrachloroethane>chloroform. Among the properties showing good correlations with distribution coefficient are Hansens three-dimensional solubility parameters and Dimroths empirical solvent polarity parameters [ET(30)]. Polarizability indices are most satisfactorily applicable to a wide variety of solvents.

Extraction of Uranium(VI) and Plutonium(IV) into Toluene by Crown Ethers from Nitric Acid Solution

Selective and effective extraction-separation of uranium(VI) and plutonium(IV) from aqueous nitric acid media into toluene by several crown ethers has been investigated in detail. The species extracted are of the ion-pair type, U0 2 (CE) 2 + • 2NOJ and Pu(CE)2 · 4NOJ. Recovery of macrocycles from loaded actinide is easily accomplished using dilute oxalic acid, perchloric acid, sulphuric acid or sodium carbonate as the strippants.

Extraction of uranium (VI), plutonium (IV) and some fission products by tri-iso-amyl phosphate

The extractive properties of tri-isoamyl-phosphate (TAP), an indigenously prepared extractant, and the loading capacity of extraction solvent containing TAP for U(VI) and Pu(IV) ions in nitric solution have been investigated. The dependence of the distribution ratio on the concentration of nitric acid showed that TAP has an ability to extract these actinides, while the fission product contaminants are poorly extracted. The distribution data revealed a quantitative extraction of both U(VI) and Pu(IV) from moderate nitric acidities in the range 2–7 mol · dm−3. Slope analysis proved predominant formation of the disolvated organic phase complex of the type UO2(NO3). 2TAP and Pu(NO3)4·2TAP with U(VI) and PU(IV), respectively. On the contrary, the extraction of fission product contaminants such as144Ce,137Cs,9Nb.,147Pr,106Ru,95Zr was almost negligible even at very high nitric acid concentrations in the aqueous phase indicating its potential application in actinide partitioning. The recovery of TAP from the loaded actinides could be easily accomplished by using a dilute sodium carbonate solution or acidified distiled water (≈0.01 mol · dm−3 HNO3) as the strippant for U(VI) and using uranous nitrate or ferrous sulphamate as that for Pu(IV). Radiation stability of TAP was adequate for most of the process applications.