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n-dodecane-supported liquid membrane
Hydrometallurgy 87 (2007) 190 – 196 www.elsevier.com/locate/hydromet
Carrier-mediated transport of uranium from phosphoric acid medium across TOPO/n-dodecane-supported liquid membrane Suman Kumar Singh a , S.K. Misra a,⁎, M. Sudersanan b , A. Dakshinamoorthy a , S.K. Munshi a , P.K. Dey a Fuel Reprocessing Division, Bhabha Atomic Research Centre, Trombay, Mumbai – 400085, India Analytical Chemistry Division, Bhabha Atomic Research Centre, Trombay, Mumbai – 400085, India
a b
Received 10 January 2007; received in revised form 20 March 2007; accepted 20 March 2007 Available online 13 April 2007
Abstract Present studies deals with the application of supported liquid membrane (SLM) technique for the separation of uranium (VI) from phosphoric acid medium. Tri-n-octyl phosphine oxide (TOPO)/n-dodecane is used as a carrier and ammonium carbonate as a receiving phase for the separation of uranium (VI) from the phosphoric acid medium. Throughout the study PTFE membranes are used as a support. The studies involve the investigation of process controlling parameters like feed acidity of phosphoric acid, carrier concentration and stripping agents. The effect of nitric acid and sodium nitrate in feed is also studied. It is found that there is negligible transport of uranium (VI) from pure phosphoric acid medium but it increases to very significant amount if 2 M nitric acid is added to feed phase. More than 90% uranium (VI) is recovered in 360 min using 0.5 M TOPO/n-dodecane as carrier and 1.89 M ammonium carbonate as stripping phase from the mixture of 0.001 M H3PO4 and 2 M of HNO3 as a feed. The flux and permeability coefficient are found to be 9.21 × 10− 6 mol/m2 s and 18.26 × 10− 5 m/s, respectively. Lower concentration of phosphoric acid with 2 M HNO3 and higher concentration of carrier is found to be the most suitable condition for maximum transport of uranium (VI) from its low-level sources like commercial phosphoric acid. © 2007 Elsevier B.V. All rights reserved. Keywords: Supported liquid membrane; Uranium; Phosphoric acid; TOPO; Separation
1. Introduction The nuclear power has an important short-term and long-term role in the present energy scenario in the world (Seaborg and Katz, 1954; Schulz et al., 1990). Emphasis is put on the secondary resources of uranium which can be proved to be a dependable source of uranium (Ninger, 2001). Uranium in low concentration is found in commercial phosphoric acid which is produced by the action of sulphuric acid on calcium phosphate rocks (Hurst ⁎ Corresponding author. E-mail address: [email protected] (S.K. Misra). 0304-386X/$ - see front matter © 2007 Elsevier B.V. All rights reserved. doi:10.1016/j.hydromet.2007.03.004
and Crouse, 1974; Singh et al., 1989; Sivaprakash, 1989). Uranium present in rock phosphate generally reports to the phosphoric acid. Since phosphoric acid is employed both for the manufacture of super phosphate as a fertilizer and for food applications, it is of interest to remove uranium from wet process phosphoric acid of different origin as well as from commercial merchant grade phosphoric acid (Singh et al., 1989, 2001; Sivaprakash, 1989; Rawajfeh and Al Matar Ali, 2000). It is also of interest to separate uranium since it can serve as an additional source of uranium for nuclear applications. In the radiochemical laboratories, we generate a good amount of analytical waste in phosphoric acid medium while analyzing uranium (VI) by Devis–Gray
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method (Davies and Gray, 1964). Before disposing this waste, we have to recover the uranium from the waste for safe disposal. Hence there is considerable interest on the development of a suitable separation scheme for the removal of uranium from phosphoric acid medium. The separation of an element or compound containing the element of interest may take effect using different separation techniques like solvent extraction, ion exchange, precipitation etc. (Preuss and Kunin, 1958; Mukherjee and Singh, 2003; Dietz et al., 2001).These processes have their own limitations and drawbacks such as solvent degradation, third phase formation, crud formation etc. To overcome these drawbacks, liquid membrane seems to be a good alternative. Supported liquid membrane (SLM) technique is very significant because of its potential application for industrial-scale separation and enrichment of metal salt species (Danesi, 1984). Their use in environmental applications, separation and recovery of important metals including nuclear material from its low-level sources and in cleaning up of effluent streams has also received attention (Danesi et al., 1981; Sastri et al., 1998; Nash and Choppin, 1997; Musikas et al., 1982). Consequently, substantial research is being carried out on optimizing the parameters for membrane processes for the removal of toxic or valuable metal ions from lean sources (Sastri et al., 1998; Hayworth et al., 1983; Barnse and Marshall, 1995). The success of the process depends on the permeation rate of the metal ion and this is decided based on the nature of the extractant used in the membrane and the kinetics and equilibrium constants of the exchange process (Visser, 1994). It is therefore necessary to obtain a good understanding of the chemistry of the membrane process in order to use this method effectively for the desired separation process (Danesi et al., 1981). Hence studies are carried out to optimize the various process parameters. The advantages of SLMs are: (1) High feed to strip volume ratios can be achieved, which leads to large concentration factors of the transported species. (2) Very low extractant (carrier) inventory is required. (3) No phase separation problem occurs because the organic and aqueous phases are never mixed. (4) Negligible organic phase entrainment occurs in the feed and strip aqueous solutions. (5) Modules have no moving parts and are simple to operate. 2. Experimental 2.1. Reagents All the reagents used are of analytical grade. Weighed amounts of UO2(NO3)2·6H2O (B.D.H.) are
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dissolved in phosphoric acid of the desired molarity. At the end, 2 M nitric acid is also added to the feed as and when required. 233 U radio tracer is used as the spike throughout this study. 1.89 M (NH4)2CO3 AR grade supplied by S.D. Fine Chemical Ltd, Mumbai, is used as stripping agent in the receiver compartment. 233U is estimated at different intervals of time in both feed and receiver compartments using α-scintillation counter. PTFE membrane is procured from Millipore (India) Pvt. Ltd Mumbai. 2.2. Membrane cell Details of glass SLM transport cell is described elsewhere (Shukla and Misra, 1991). Single-stage SLM measurements is carried out with two compartment permeation cell in which a source aqueous solution (10 mL) is separated from the aqueous receiving solution (10 mL) by a liquid membrane with an effective membrane area of 1.141 × 10− 4 m2. The source and receiving solutions are mechanically stirred using magnetic stirrer at room temperature to avoid concentration polarization conditions at the membrane interfaces and in the bulk solutions. Membrane permeabilities are determined by monitoring the uranium concentration radiometrically, primarily in the receiving phase, as a function of time. 2.3. Membrane supports Throughout this study, flat-sheet type PTFE (polytetra-fluoro-ethylene) hydrophobic microporous polymeric membranes are used. The porosity of the membrane is about 84%. The membrane had an average pore diameter and thickness of 0.45 and 160 μm, respectively. Filling the pores of these dry support polymers with the carrier solution is accomplished by immersing the membrane in the organic phase for at least 6–8 h before use. The pores are immediately and apparently quantitatively filled with the carrier solution by capillary action. This type of SLM polymeric support eliminates the transport of water through the membrane and is free from osmotic effects. The U (VI)–TOPO complex can diffuse across the membrane supported through the microporous structure. The concentration of uranium (VI) solution used in the present study is deliberately kept very dilute. The organic membrane phase is prepared by dissolving a weighed amount of TOPO (Fluka) in dodecane (Aldrich) to provide carrier solutions of varying concentrations which are subsequently equilibrated with acid solutions of the desired molarity before use. Various membrane
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experiments give the permeation results reproducible to approximately ± 10%. All batch experiments are carried out at room temperature i.e. 25 °C ± 1 °C. 2.4. Flux and permeability coefficient The flux is defined as the rate of mass transport of the solute through the SLM, and it is the criteria often used to evaluate a SLM. According to the Flick's first law of diffusion, the rate of diffusion (dc/dt) of a solute across an area (A), known as flux (J), is calculated as J ¼ CU; receiving V =ð A t Þ ð1Þ Where CU,receiving=U concentration in the receiving phase, g mol/dm3; V=volume of receiving phase, m3; A=effective area of the membrane, m2 t=time elapsed, s. Permeability coefficient is calculated using Eq. (2) lnðCt =C0 Þ ¼ ð A P=V Þ t
ð2Þ
Where P=permeability coefficient, m/s; C0=concentration of uranium at time 0; Ct=concentration of uranium at the time t; A=effective area of the membrane, m2; V=volume of receiving phase, m3. 2.5. Liquid–liquid distribution measurements
H2SO4, (NH4)2CO3, citric acid, urea, disodium salt of EDTA and H2O employed, (NH4)2CO3 found to be best among all. Fig. 1 shows that near quantitative recovery of uranium is possible with ammonium carbonate and around 90% uranium is recovered by using sulphuric acid. All other strippants are able to strip back 25% of uranium (VI) or less. 3.2. Optimization of feed conditions Single-ion transport of uranium (VI) across TOPO/ndodecane SLM from an aqueous feed adjusted to 0.5 M phosphoric acid is tested. In the present study, the concentration of 0.1 M TOPO in dodecane is used as membrane phase and 1.89 M ammonium carbonate is used as strippant. The concentration of (NH4)2CO3 is kept high to avoid any neutralization of the strippant through an acid transport from the feed to the receiving phase. The results as indicated in Table 1 shows that only 0.87% of uranium (VI) could be transported in 360 min from pure phosphoric acid medium. This may be due to the strong complex of uranium–phosphoric acid which is not extractable by TOPO (Singh et al., 2004). To overcome this problem, 2 M nitric acid is added to the feed (Divakar, 2001). The other conditions are kept same. The transport of uranium (VI) increases drastically to 38%. The effect of nitrate ion concentration on percentage transport of uranium (VI) from
15 mL of U (VI) std prepared in 0.5 M phosphoric acid is taken in a separating funnel. 1 mL of 233U tracer is added to separating funnel. 16 mL of 0.1 M TOPO in n-dodecane is added to the separating funnel. The separating funnel is equilibrated for 30 min and both the layers are allowed to settle. Once the phases clearly separated, the organic and aqueous phases are collected in different beakers. 1 mL of the extracted organic layer is taken and 1 mL of various stripping agents is added to the respective extraction vials. These vials are equilibrated for 30 min and centrifuged for 5 min. Material balance of both the layers is checked by monitoring the concentration of 233 U by α-scintillation counter. 3. Results and discussion 3.1. Effect of strippants on uranium permeation The transport of uranyl ions across TOPO /dodecane membrane is strongly dependent upon the nature of a strippant present on the recovery phase of the membrane. Hence detailed solvent extraction experiments are carried out to select the most suitable strippant. It is apparent that out of several aqueous strippants such as HNO3, HCl,
Fig. 1. Percentage back extraction of uranium using different stripping agents.
S.K. Singh et al. / Hydrometallurgy 87 (2007) 190–196 Table 1 Percentage transport of uranium with time at different feed conditions Time (min)
Percentage transport in different feed conditions (%) 0.5 M H3PO4
0.5 M H3PO4 + 2 M NaNO3
0.5 M H3PO4 + 2 M HNO3
15 30 60 120 240 360
0.26 0.60 0.73 0.77 0.82 0.87
2.03 3.27 5.38 8.37 10.78 12.58
2.71 6.51 15.25 26.11 33.70 37.99
Initial feed concentration: 1.28 × 10− 3 mol/L uranium. Carrier (TOPO) concentration: 0.1 M TOPO/n-dodecane. Strippant: 1.89 M ammonium carbonate.
phosphoric acid medium is studied. In place of 2 M nitric acid, 2 M NaNO3 is added to the feed solution keeping other conditions same. Around 13% percentage transport of uranium (VI) took place after 360 min of experiment. The results are shown in Table 1. From the results it is clear that the transport of uranium does not depends only upon the nitrate ion concentration but also depends on the concentration of H+ ion. The effect of various factors affecting the membrane transport can be qualitatively evaluated on the basis of the chemistry of the metal ion i.e. the complexation of metal with carrier, the solubility of complex in diluent and the diffusion of this complex in membrane phase as well as decomplexation from the carrier to the strippants used. The flux depends upon the concentration gradient of diffusing species inside the membrane just adjacent to the aqueous solutions interfacing it. This depends on the distribution coefficient of metal ions between the membrane and aqueous phases. Uranium (VI) forms a fairly strong complex with phosphate. The strength of the complex will be lower at lower concentration of phosphoric acid. Hence, assuming other factors to be the same, the transport of uranium (VI) can be expected to be more at lower concentration of the phosphoric acid in feed. This, however, assumes the formation of a complex between uranium (VI) and phosphate which is not formally covalent and neutral as in the case of uranyl nitrates (Preuss et al., 1958). The addition of nitric acid suppresses the dissociation of phosphoric acid because of the common ion (H+) present in nitric acid. Thus the availability of free nitrates enhances the complex formation with uranium which is finally transported by TOPO/n-dodecane membrane. This is confirmed by replacing the nitric acid by taking the same amount of sodium nitrate. The results indicated that the transport of uranium does not improve. Hence it is established that not only nitrate ions but H+ ions also
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plays the important role for uranium transport in phosphoric acid medium. Therefore, while recovering uranium from phosphoric acid medium the feed should always contain at least 2 M nitric acid along with the phosphoric acid for better recovery of uranium (Divakar et al., 2001). 3.3. Effect of phosphoric acid concentration in feed on uranium transport Single ion transport of uranium (VI) across TOPO/ndodecane SLM from an aqueous feed adjusted to 0.001– 3.5 M phosphoric acid is tested. In addition to the phosphoric acid, 2 M nitric acid is also added to the feed. In the present study, the concentration of 0.1 M TOPO in n-dodecane is used as membrane phase and 1.89 M ammonium carbonate is used as strippant. The uranium (VI) transport increases with the decrease in phosphoric acid concentration. The maximum transport of around 80% (Fig. 2) could be achieved by using 0.001 M phosphoric acid in feed (Table 2) after about 360 min across TOPO/n-dodecane SLM. Only 12% uranium (VI) could be transported while using the 3.5 M phosphoric acid. The permeation coefficients and average flux are found to be 18.26 × 10− 5 m/s, 9.21 × 10− 6 mol/m2 s and 0.10 × 10− 5 m/s, 0.663 × 10 − 6 mol/m 2 s for 0.001 M and 3.5 M phosphoric acid, respectively, which are tabulated in Table 2.
Fig. 2. Percentage of transport of uranium as a function of acidity of mixture of different concentration of H3PO4 and 2 M HNO3. [U] In feed = 1.28 × 10− 3 mol/L, Strippant = 1.89 M (NH4)2CO3, Carrier concentration = 0.1 M TOPO/n-dodecane.
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This is partly in agreement with the expected trend since the flux of a cation varies with H+/NO3− ion concentration following the relationship n n JM ¼ AT =g NO 3 aq ½Carrierorg CM;feed A=area of membrane, m2; T=absolute temperature, K; η=viscosity, cP; CM=concentration of metal in feed, mol/L. And, hence, there would be an increase in permeability with increase in proton or nitric acid concentration. The increase in phosphoric acid concentration requires more and more H+ ions from HNO3 to keep phosphoric acid undissociated. The remaining available NO3− ions are only responsible for complexing uranium (VI) for final transport through TOPO/n-dodecane SLM membrane transport. Fig. 3 depicts the trend of average flux of uranyl ions after 60 min and 360 min at various concentration of H3PO4 in the feed solution. 3.4. Effect of TOPO (Carrier) concentration on uranium transport Composition of the organic solution has a marked effect on a cation flux. When transport across a membrane occurs via a carrier, as in facilitated transport, the flux is generally expected to increase with increasing carrier concentration. The effect of TOPO/n-dodecane concentration on the transport of uranium (VI) is studied. The studies are made over a concentration range of 0.01 M to 0.5 M TOPO/n-dodecane. Table 3 summarizes data on uranium flux, permeability coefficient vs. TOPO concentration in the membrane. Table 2 Permeation of uranium as a function of source phase phosphoric acid molarity [H3PO4] [Uranium] Uranium Flux (M) receiving phase permeation (mol/m2 s) (mol/L) (%) 0.001 0.01 0.1 0.2 0.5 0.75 1.0 1.5 2.0 3.5
9.97 × 10− 4 8.13 × 10− 4 6.58 × 10− 4 5.80 × 10− 4 4.88 × 10− 4 4.09 × 10− 4 3.50 × 10− 4 2.73 × 10− 4 2.04 × 10− 4 1.53 × 10− 4
77.67 63.33 51.24 45.23 37.99 31.86 27.28 21.24 15.92 11.93
Permeability coefficient (m/s)
9.212 × 10− 6 18.26 × 10− 5 7.810 × 10− 6 14.61 × 10− 5 4.893 × 10− 6 8.76 × 10− 5 4.228 × 10− 6 7.79 × 10− 5 3.490 × 10− 6 6.33 × 10− 5 2.804 × 10− 6 3.90 × 10− 5 2.152 × 10− 6 3.41 × 10− 5 1.512 × 10− 6 0.24 × 10− 5 0.965 × 10− 6 0.15 × 10− 5 0.663 × 10− 6 0.10 × 10− 5
Initial feed concentration: 1.28 × 10− 3 mol/L uranium in varied amount of phosphoric acid along with 2 M nitric acid. Carrier (TOPO) concentration: 0.1 M TOPO/n-dodecane. Strippant: 1.89 M ammonium carbonate.
Fig. 3. Average flux vs. acidity of a mixture of varying concentration of phosphoric acid and 2 M HNO3 after 60 and 360 min. [U] In feed = 1.28 × 10− 3 mol/L, Strippant = 1.89 M (NH4)2CO3, Carrier concentration = 0.1 M TOPO/n-dodecane.
Generally an increase in carrier concentration produces an increase in cation flux; however, the concurrent increase in viscosity results in a steady decrease in the diffusivity of the carrier as well as of the metal–carrier complex. Eventually the cation flux will increase as the carrier concentration is increased, as observed in the present study (Fig. 4). The curve shown in Fig. 4 indicates that initially the flux increases and reaches maxima and then decreases. This is because metal ion flux is associated with its concentration present in the feed phase. Initially, the membrane phase starts saturating with the carrier–metal complex and very small amount of metal ion transport to the receiving side hence we get low values of flux. Once the membrane phase is saturated, because of the concentration gradient the maximum metal ion transported to the receiving phase Table 3 Permeation of uranium as a function of carrier concentration (TOPO) [TOPO] [Uranium] Uranium Flux (M) receiving phase permeation (mol/m2 s) (mol/L) (%) 0.01 0.02 0.05 0.1 0.2 0.3 0.5
2.74 × 10− 4 3.70 × 10− 4 4.19 × 10− 4 4.75 × 10− 4 7.91 × 10− 4 9.68 × 10− 4 1.17 × 10− 3
21.23 28.79 32.59 37.00 61.59 75.36 91.11
1.651 × 10− 6 2.280 × 10− 6 2.863 × 10− 6 3.480 × 10− 6 5.317 × 10− 6 6.614 × 10− 6 8.684 × 10− 6
Permeability coefficient (m/s) 1.34 × 10− 5 1.95 × 10− 5 2.31 × 10− 5 3.04 × 10− 5 4.50 × 10− 5 5.97 × 10− 5 8.89 × 10− 5
Initial feed concentration: 1.28 × 10− 3 mol/L uranium in 0.5 M phosphoric acid along with 2 M nitric acid. Carrier (TOPO) concentration: different concentration of TOPO/ n-dodecane. Strippant: 1.89 M ammonium carbonate.
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therefore the flux reaches maxima. Then finally decreases as the concentration of metal ion decreases at feed side or concentration gradient decreases. It is observed that with increasing carrier concentration in an inert diluent like dodecane, uranium (VI) transport gradually increases (Fig. 5). The permeation coefficients and average flux are found to be 8.89 × 10− 5 m/s, and 8.68 × 10− 6 mol/m2 s for 0.5 M TOPO in dodecane (Table 3). It is also obvious that the transport of uranium (VI) in such a system should be a function of both the distribution coefficient and diffusion coefficient because the transfer of uranyl ions through the membrane may be considered diffusive in nature. The difference in permeability between experiments with varying feed acidity, carrier concentration and uranium molarity can be understood by considering the probable expression for the rate of formation of the diffusing species at the feed interface: d UO2 ðNO3 Þ2 d2TOPO =dt 2 2þ ¼ k ⁎ ½TOPO2 NO UO2 3
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Fig. 5. Percentage of transport of uranium as a function of concentration of TOPO/n-dodecane. Feed acidity=0.5 M H3PO4 +2 M HNO3, [U] In feed =1.28×10−3 mol/L, Strippant=1.89 M (NH4)2CO3.
Where k⁎ is rate constant for the formation of the UO22+– TOPO complex. This equation is based on the stoichiometry of uranium (VI) extraction established earlier (Mansingh et al., 1996). Rate of diffusion of metal species will thus depend upon any changes in NO3−, TOPO and UO22+ concentration in the feed side. It is important to note that the trend for transport of uranyl ions through the liquid membrane is nearly alike as that encountered with this cation in liquid–liquid extraction system (Singh et al., 2004). Fig. 5 indicates
that at all the TOPO concentrations; the trend in variation of percentage transport of metal ions is the same, i.e. increasing with the increase in the concentration of TOPO.
Fig. 4. Flux vs. time for different concentration of TOPO in n-dodecane. Feed acidity=0.5 M H3PO4 +2 M HNO3, [U] In feed=1.28×10−3 mol/L, Strippant=1.89 M (NH4)2CO3.
Fig. 6. Concentration of uranium in feed and receiver at 0.001 M H3PO4 using 0.1 M TOPO in n-dodecane and 1.89 M (NH4)2CO3 as strippant.
4. Conclusion The transport of uranium (VI) by TOPO/n-dodecane SLM is studied and found that uranium (VI) can be very well separated from phosphoric acid medium. It is established that maximum uranium permeation through SLM membranes is attained with feed acidity around
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0.001 M phosphoric acid with 2 M HNO3 and 0.5 M TOPO in n-dodecane as the carrier, 1.89 M ammonium carbonate has been found to be the best strippants for uranium (VI). Fig. 6 shows that more than 90% uranium can be recovered in 360 min from feed to the receiving phase. The driving force for the separation is not only the nitrate ions but also the H + ions which are responsible for the transport of uranyl ions. Acknowledgements The authors wish to thanks Sri S.D. Misra, Director, Nuclear recycle group, Bhabha Atomic Research centre, Mumbai for his keen interest to carry out this work and for his constant encouragement during the entire work. References Barnse, D.E., Marshall, G.D., 1995. Rapid optimization of chemical parameters affecting supported liquid membranes. Separation Science Technology 30, 751–776. Danesi, P.R., 1984. Separation of metal species by supported liquid membranes. Separation Science Technology 19, 857–894. Danesi, P.R., Horwitz, E.P., Vandeghraff, G.F., 1981. Mass transfer rate through liquid membranes: interfacial chemical reaction and diffusion as simultaneous permeability controlling factors. Separation Science and Technology 16 (2), 201–211. Davies, W., Gray, W., 1964. A rapid and specific titremetric method for the precise determination of uranium using iron(II) sulphate as reductant. Talenta 11, 1203–1211. Dietz, Mark L., Horwitz, E. Philip, Sajdak, Larry R., Chiarzia, Renato, 2001. An improved extraction chromatographic resin for the separation of uranium from acidic nitrate media. Talanta 54, 1173–1184. Divakar, D.S., Parikh, K.J., Pandit, S.S., Singh, Suman Kumar, Jambunathan, U., Ramanujam, A., 2001. Sensitive Method for the Determination of Microquantities of 233U in the Presence of Large Amounts of Thorium. INSAC, CAT, Indore, pp. 295–298. Hayworth, H.C., Ho, W.S., Burns Jr., W.A., Norman, N.Li., 1983. Extraction of uranium from wet phosphoric acid by liquid membranes. Hydrometallurgy 493–521. Hurst, F.J., Crouse, D., 1974. Recovery of uranium from wet process phosphoric acid by extraction with octylphenylphosphoric acid. Industrial Engineering Chemistry Process Development 11, 122–131. Mansingh, P.S., Veeraraghavan, R., Mahapatra, P.K., Manchanda, V.K., Dash, K.C., 1996. Adduct formation of a uranyl isoxazolonate with organophillic neutral oxodonors. Radiochemica Acta 72, 127–132.
Mukherjee, T.K., Singh, H., 2003. Recovery of Uranium and Thorium from Secondary Resources. INSAC, Kalpakkam IT/1. Musikas, C., Benjelloun, N., Lours, S., 1982. Actinide Recovery from Waste and Low Grade Sources. Marcel Decker, New York, USA, pp. 245–251. Nash, K.L., Choppin, G.R., 1997. Separation chemistry for actinides elements: recent developments and historical perspective. Separation Science Technology 32, 255–274. Ninger, D.E., 2001. Uranium exploration policy, economics, and future prospects, U.S. Atomic Energy Commission, Division of Production and Material Management, USA, IAEA-PL-490/7. Preuss, A., Kunin, A., 1958. Uranium recovery by ion exchange. In: Clegg, J.W., Foley, D.D. (Eds.), Uranium Ore Processing. Addison-Wesley, Reading, pp. 191–236. Chap. 9. Rawajfeh, Khaled M., Al Matar Ali, Kh., 2000. Uranium extraction from purified wet process Jordanian phosphoric acid: a development study. Hydrometallurgy 56, 309–322. Sastri, A.M., Kumar, A., Shukla, J.P., Singh, R.K., 1998. Improved techniques in liquid membrane separation: an overview. Separation and Purification Methods 27, 213–219. Schulz, W.W., Burger, L.L., Navratil, J.D., Bender, K.P., 1990. Application of tributyl phosphate in nuclear fuel processing. Science and Technology of Tributyl Phosphate, III. CRC Press Inc, Boca Raton, FL USA. Seaborg, G.T., Katz, J.J., 1954. The Actinide Elements, 1st ed. McGraw-Hill Book Company, Inc., New York. Shukla, J.P., Misra, S.K., 1991. Carrier mediated transport of uranyl ions across tri-butyl phosphate/dodecane liquid membranes. Journal of Membrane Science 64, 93. Singh, H., Nagle, R.A., Gariyankar, A.B., Fonseca, M.F., Koppikar, K.S., 1989. On site tests for recovery of uranium from wet process phosphoric acid at FACT. International Symposium on Uranium Technology 621–634. Singh, H., Vijyalakshmi, R., Misra, S.L., Gupta, C.K., 2001. Studies on uranium extraction from phosphoric acid using di-nonyl phenyl phosphoric acid-based synergistic mixtures. Hydrometallurgy 59, 69–76. Singh, Suman Kumar, Dhami, P.S., Vijyan, K., Jumbunathan, U., Dey, P.K., 2004. Studies on the Behaviour of Diluents on Extraction of Uranium From Phosphoric Acid Medium Using Tri-n-Octyl Phosphine Oxide as Extractant (RCEAS–2004). National seminar on role of chemistry in the emerging areas of applied science, S.V. University, Thirupathi, India, pp. 189–287. Sivaprakash, G., 1989. Uranium recovery from phosphoric acid. International Symposium on Uranium Technology, Mumbai, pp. 592–604. Visser, H.C., 1994. Supported Liquid Membranes with Improved Stability; Kinetics and Mechanism of Carrier Mediated Salt Transport. Chap. 1,2.
Carrier-mediated transport of uranium from phosphoric acid medium across TOPO/n-dodecane-supported liquid membrane Suman Kumar Singh a , S.K. Misra a,⁎, M. Sudersanan b , A. Dakshinamoorthy a , S.K. Munshi a , P.K. Dey a Fuel Reprocessing Division, Bhabha Atomic Research Centre, Trombay, Mumbai – 400085, India Analytical Chemistry Division, Bhabha Atomic Research Centre, Trombay, Mumbai – 400085, India
a b
Received 10 January 2007; received in revised form 20 March 2007; accepted 20 March 2007 Available online 13 April 2007
Abstract Present studies deals with the application of supported liquid membrane (SLM) technique for the separation of uranium (VI) from phosphoric acid medium. Tri-n-octyl phosphine oxide (TOPO)/n-dodecane is used as a carrier and ammonium carbonate as a receiving phase for the separation of uranium (VI) from the phosphoric acid medium. Throughout the study PTFE membranes are used as a support. The studies involve the investigation of process controlling parameters like feed acidity of phosphoric acid, carrier concentration and stripping agents. The effect of nitric acid and sodium nitrate in feed is also studied. It is found that there is negligible transport of uranium (VI) from pure phosphoric acid medium but it increases to very significant amount if 2 M nitric acid is added to feed phase. More than 90% uranium (VI) is recovered in 360 min using 0.5 M TOPO/n-dodecane as carrier and 1.89 M ammonium carbonate as stripping phase from the mixture of 0.001 M H3PO4 and 2 M of HNO3 as a feed. The flux and permeability coefficient are found to be 9.21 × 10− 6 mol/m2 s and 18.26 × 10− 5 m/s, respectively. Lower concentration of phosphoric acid with 2 M HNO3 and higher concentration of carrier is found to be the most suitable condition for maximum transport of uranium (VI) from its low-level sources like commercial phosphoric acid. © 2007 Elsevier B.V. All rights reserved. Keywords: Supported liquid membrane; Uranium; Phosphoric acid; TOPO; Separation
1. Introduction The nuclear power has an important short-term and long-term role in the present energy scenario in the world (Seaborg and Katz, 1954; Schulz et al., 1990). Emphasis is put on the secondary resources of uranium which can be proved to be a dependable source of uranium (Ninger, 2001). Uranium in low concentration is found in commercial phosphoric acid which is produced by the action of sulphuric acid on calcium phosphate rocks (Hurst ⁎ Corresponding author. E-mail address: [email protected] (S.K. Misra). 0304-386X/$ - see front matter © 2007 Elsevier B.V. All rights reserved. doi:10.1016/j.hydromet.2007.03.004
and Crouse, 1974; Singh et al., 1989; Sivaprakash, 1989). Uranium present in rock phosphate generally reports to the phosphoric acid. Since phosphoric acid is employed both for the manufacture of super phosphate as a fertilizer and for food applications, it is of interest to remove uranium from wet process phosphoric acid of different origin as well as from commercial merchant grade phosphoric acid (Singh et al., 1989, 2001; Sivaprakash, 1989; Rawajfeh and Al Matar Ali, 2000). It is also of interest to separate uranium since it can serve as an additional source of uranium for nuclear applications. In the radiochemical laboratories, we generate a good amount of analytical waste in phosphoric acid medium while analyzing uranium (VI) by Devis–Gray
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method (Davies and Gray, 1964). Before disposing this waste, we have to recover the uranium from the waste for safe disposal. Hence there is considerable interest on the development of a suitable separation scheme for the removal of uranium from phosphoric acid medium. The separation of an element or compound containing the element of interest may take effect using different separation techniques like solvent extraction, ion exchange, precipitation etc. (Preuss and Kunin, 1958; Mukherjee and Singh, 2003; Dietz et al., 2001).These processes have their own limitations and drawbacks such as solvent degradation, third phase formation, crud formation etc. To overcome these drawbacks, liquid membrane seems to be a good alternative. Supported liquid membrane (SLM) technique is very significant because of its potential application for industrial-scale separation and enrichment of metal salt species (Danesi, 1984). Their use in environmental applications, separation and recovery of important metals including nuclear material from its low-level sources and in cleaning up of effluent streams has also received attention (Danesi et al., 1981; Sastri et al., 1998; Nash and Choppin, 1997; Musikas et al., 1982). Consequently, substantial research is being carried out on optimizing the parameters for membrane processes for the removal of toxic or valuable metal ions from lean sources (Sastri et al., 1998; Hayworth et al., 1983; Barnse and Marshall, 1995). The success of the process depends on the permeation rate of the metal ion and this is decided based on the nature of the extractant used in the membrane and the kinetics and equilibrium constants of the exchange process (Visser, 1994). It is therefore necessary to obtain a good understanding of the chemistry of the membrane process in order to use this method effectively for the desired separation process (Danesi et al., 1981). Hence studies are carried out to optimize the various process parameters. The advantages of SLMs are: (1) High feed to strip volume ratios can be achieved, which leads to large concentration factors of the transported species. (2) Very low extractant (carrier) inventory is required. (3) No phase separation problem occurs because the organic and aqueous phases are never mixed. (4) Negligible organic phase entrainment occurs in the feed and strip aqueous solutions. (5) Modules have no moving parts and are simple to operate. 2. Experimental 2.1. Reagents All the reagents used are of analytical grade. Weighed amounts of UO2(NO3)2·6H2O (B.D.H.) are
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dissolved in phosphoric acid of the desired molarity. At the end, 2 M nitric acid is also added to the feed as and when required. 233 U radio tracer is used as the spike throughout this study. 1.89 M (NH4)2CO3 AR grade supplied by S.D. Fine Chemical Ltd, Mumbai, is used as stripping agent in the receiver compartment. 233U is estimated at different intervals of time in both feed and receiver compartments using α-scintillation counter. PTFE membrane is procured from Millipore (India) Pvt. Ltd Mumbai. 2.2. Membrane cell Details of glass SLM transport cell is described elsewhere (Shukla and Misra, 1991). Single-stage SLM measurements is carried out with two compartment permeation cell in which a source aqueous solution (10 mL) is separated from the aqueous receiving solution (10 mL) by a liquid membrane with an effective membrane area of 1.141 × 10− 4 m2. The source and receiving solutions are mechanically stirred using magnetic stirrer at room temperature to avoid concentration polarization conditions at the membrane interfaces and in the bulk solutions. Membrane permeabilities are determined by monitoring the uranium concentration radiometrically, primarily in the receiving phase, as a function of time. 2.3. Membrane supports Throughout this study, flat-sheet type PTFE (polytetra-fluoro-ethylene) hydrophobic microporous polymeric membranes are used. The porosity of the membrane is about 84%. The membrane had an average pore diameter and thickness of 0.45 and 160 μm, respectively. Filling the pores of these dry support polymers with the carrier solution is accomplished by immersing the membrane in the organic phase for at least 6–8 h before use. The pores are immediately and apparently quantitatively filled with the carrier solution by capillary action. This type of SLM polymeric support eliminates the transport of water through the membrane and is free from osmotic effects. The U (VI)–TOPO complex can diffuse across the membrane supported through the microporous structure. The concentration of uranium (VI) solution used in the present study is deliberately kept very dilute. The organic membrane phase is prepared by dissolving a weighed amount of TOPO (Fluka) in dodecane (Aldrich) to provide carrier solutions of varying concentrations which are subsequently equilibrated with acid solutions of the desired molarity before use. Various membrane
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experiments give the permeation results reproducible to approximately ± 10%. All batch experiments are carried out at room temperature i.e. 25 °C ± 1 °C. 2.4. Flux and permeability coefficient The flux is defined as the rate of mass transport of the solute through the SLM, and it is the criteria often used to evaluate a SLM. According to the Flick's first law of diffusion, the rate of diffusion (dc/dt) of a solute across an area (A), known as flux (J), is calculated as J ¼ CU; receiving V =ð A t Þ ð1Þ Where CU,receiving=U concentration in the receiving phase, g mol/dm3; V=volume of receiving phase, m3; A=effective area of the membrane, m2 t=time elapsed, s. Permeability coefficient is calculated using Eq. (2) lnðCt =C0 Þ ¼ ð A P=V Þ t
ð2Þ
Where P=permeability coefficient, m/s; C0=concentration of uranium at time 0; Ct=concentration of uranium at the time t; A=effective area of the membrane, m2; V=volume of receiving phase, m3. 2.5. Liquid–liquid distribution measurements
H2SO4, (NH4)2CO3, citric acid, urea, disodium salt of EDTA and H2O employed, (NH4)2CO3 found to be best among all. Fig. 1 shows that near quantitative recovery of uranium is possible with ammonium carbonate and around 90% uranium is recovered by using sulphuric acid. All other strippants are able to strip back 25% of uranium (VI) or less. 3.2. Optimization of feed conditions Single-ion transport of uranium (VI) across TOPO/ndodecane SLM from an aqueous feed adjusted to 0.5 M phosphoric acid is tested. In the present study, the concentration of 0.1 M TOPO in dodecane is used as membrane phase and 1.89 M ammonium carbonate is used as strippant. The concentration of (NH4)2CO3 is kept high to avoid any neutralization of the strippant through an acid transport from the feed to the receiving phase. The results as indicated in Table 1 shows that only 0.87% of uranium (VI) could be transported in 360 min from pure phosphoric acid medium. This may be due to the strong complex of uranium–phosphoric acid which is not extractable by TOPO (Singh et al., 2004). To overcome this problem, 2 M nitric acid is added to the feed (Divakar, 2001). The other conditions are kept same. The transport of uranium (VI) increases drastically to 38%. The effect of nitrate ion concentration on percentage transport of uranium (VI) from
15 mL of U (VI) std prepared in 0.5 M phosphoric acid is taken in a separating funnel. 1 mL of 233U tracer is added to separating funnel. 16 mL of 0.1 M TOPO in n-dodecane is added to the separating funnel. The separating funnel is equilibrated for 30 min and both the layers are allowed to settle. Once the phases clearly separated, the organic and aqueous phases are collected in different beakers. 1 mL of the extracted organic layer is taken and 1 mL of various stripping agents is added to the respective extraction vials. These vials are equilibrated for 30 min and centrifuged for 5 min. Material balance of both the layers is checked by monitoring the concentration of 233 U by α-scintillation counter. 3. Results and discussion 3.1. Effect of strippants on uranium permeation The transport of uranyl ions across TOPO /dodecane membrane is strongly dependent upon the nature of a strippant present on the recovery phase of the membrane. Hence detailed solvent extraction experiments are carried out to select the most suitable strippant. It is apparent that out of several aqueous strippants such as HNO3, HCl,
Fig. 1. Percentage back extraction of uranium using different stripping agents.
S.K. Singh et al. / Hydrometallurgy 87 (2007) 190–196 Table 1 Percentage transport of uranium with time at different feed conditions Time (min)
Percentage transport in different feed conditions (%) 0.5 M H3PO4
0.5 M H3PO4 + 2 M NaNO3
0.5 M H3PO4 + 2 M HNO3
15 30 60 120 240 360
0.26 0.60 0.73 0.77 0.82 0.87
2.03 3.27 5.38 8.37 10.78 12.58
2.71 6.51 15.25 26.11 33.70 37.99
Initial feed concentration: 1.28 × 10− 3 mol/L uranium. Carrier (TOPO) concentration: 0.1 M TOPO/n-dodecane. Strippant: 1.89 M ammonium carbonate.
phosphoric acid medium is studied. In place of 2 M nitric acid, 2 M NaNO3 is added to the feed solution keeping other conditions same. Around 13% percentage transport of uranium (VI) took place after 360 min of experiment. The results are shown in Table 1. From the results it is clear that the transport of uranium does not depends only upon the nitrate ion concentration but also depends on the concentration of H+ ion. The effect of various factors affecting the membrane transport can be qualitatively evaluated on the basis of the chemistry of the metal ion i.e. the complexation of metal with carrier, the solubility of complex in diluent and the diffusion of this complex in membrane phase as well as decomplexation from the carrier to the strippants used. The flux depends upon the concentration gradient of diffusing species inside the membrane just adjacent to the aqueous solutions interfacing it. This depends on the distribution coefficient of metal ions between the membrane and aqueous phases. Uranium (VI) forms a fairly strong complex with phosphate. The strength of the complex will be lower at lower concentration of phosphoric acid. Hence, assuming other factors to be the same, the transport of uranium (VI) can be expected to be more at lower concentration of the phosphoric acid in feed. This, however, assumes the formation of a complex between uranium (VI) and phosphate which is not formally covalent and neutral as in the case of uranyl nitrates (Preuss et al., 1958). The addition of nitric acid suppresses the dissociation of phosphoric acid because of the common ion (H+) present in nitric acid. Thus the availability of free nitrates enhances the complex formation with uranium which is finally transported by TOPO/n-dodecane membrane. This is confirmed by replacing the nitric acid by taking the same amount of sodium nitrate. The results indicated that the transport of uranium does not improve. Hence it is established that not only nitrate ions but H+ ions also
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plays the important role for uranium transport in phosphoric acid medium. Therefore, while recovering uranium from phosphoric acid medium the feed should always contain at least 2 M nitric acid along with the phosphoric acid for better recovery of uranium (Divakar et al., 2001). 3.3. Effect of phosphoric acid concentration in feed on uranium transport Single ion transport of uranium (VI) across TOPO/ndodecane SLM from an aqueous feed adjusted to 0.001– 3.5 M phosphoric acid is tested. In addition to the phosphoric acid, 2 M nitric acid is also added to the feed. In the present study, the concentration of 0.1 M TOPO in n-dodecane is used as membrane phase and 1.89 M ammonium carbonate is used as strippant. The uranium (VI) transport increases with the decrease in phosphoric acid concentration. The maximum transport of around 80% (Fig. 2) could be achieved by using 0.001 M phosphoric acid in feed (Table 2) after about 360 min across TOPO/n-dodecane SLM. Only 12% uranium (VI) could be transported while using the 3.5 M phosphoric acid. The permeation coefficients and average flux are found to be 18.26 × 10− 5 m/s, 9.21 × 10− 6 mol/m2 s and 0.10 × 10− 5 m/s, 0.663 × 10 − 6 mol/m 2 s for 0.001 M and 3.5 M phosphoric acid, respectively, which are tabulated in Table 2.
Fig. 2. Percentage of transport of uranium as a function of acidity of mixture of different concentration of H3PO4 and 2 M HNO3. [U] In feed = 1.28 × 10− 3 mol/L, Strippant = 1.89 M (NH4)2CO3, Carrier concentration = 0.1 M TOPO/n-dodecane.
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This is partly in agreement with the expected trend since the flux of a cation varies with H+/NO3− ion concentration following the relationship n n JM ¼ AT =g NO 3 aq ½Carrierorg CM;feed A=area of membrane, m2; T=absolute temperature, K; η=viscosity, cP; CM=concentration of metal in feed, mol/L. And, hence, there would be an increase in permeability with increase in proton or nitric acid concentration. The increase in phosphoric acid concentration requires more and more H+ ions from HNO3 to keep phosphoric acid undissociated. The remaining available NO3− ions are only responsible for complexing uranium (VI) for final transport through TOPO/n-dodecane SLM membrane transport. Fig. 3 depicts the trend of average flux of uranyl ions after 60 min and 360 min at various concentration of H3PO4 in the feed solution. 3.4. Effect of TOPO (Carrier) concentration on uranium transport Composition of the organic solution has a marked effect on a cation flux. When transport across a membrane occurs via a carrier, as in facilitated transport, the flux is generally expected to increase with increasing carrier concentration. The effect of TOPO/n-dodecane concentration on the transport of uranium (VI) is studied. The studies are made over a concentration range of 0.01 M to 0.5 M TOPO/n-dodecane. Table 3 summarizes data on uranium flux, permeability coefficient vs. TOPO concentration in the membrane. Table 2 Permeation of uranium as a function of source phase phosphoric acid molarity [H3PO4] [Uranium] Uranium Flux (M) receiving phase permeation (mol/m2 s) (mol/L) (%) 0.001 0.01 0.1 0.2 0.5 0.75 1.0 1.5 2.0 3.5
9.97 × 10− 4 8.13 × 10− 4 6.58 × 10− 4 5.80 × 10− 4 4.88 × 10− 4 4.09 × 10− 4 3.50 × 10− 4 2.73 × 10− 4 2.04 × 10− 4 1.53 × 10− 4
77.67 63.33 51.24 45.23 37.99 31.86 27.28 21.24 15.92 11.93
Permeability coefficient (m/s)
9.212 × 10− 6 18.26 × 10− 5 7.810 × 10− 6 14.61 × 10− 5 4.893 × 10− 6 8.76 × 10− 5 4.228 × 10− 6 7.79 × 10− 5 3.490 × 10− 6 6.33 × 10− 5 2.804 × 10− 6 3.90 × 10− 5 2.152 × 10− 6 3.41 × 10− 5 1.512 × 10− 6 0.24 × 10− 5 0.965 × 10− 6 0.15 × 10− 5 0.663 × 10− 6 0.10 × 10− 5
Initial feed concentration: 1.28 × 10− 3 mol/L uranium in varied amount of phosphoric acid along with 2 M nitric acid. Carrier (TOPO) concentration: 0.1 M TOPO/n-dodecane. Strippant: 1.89 M ammonium carbonate.
Fig. 3. Average flux vs. acidity of a mixture of varying concentration of phosphoric acid and 2 M HNO3 after 60 and 360 min. [U] In feed = 1.28 × 10− 3 mol/L, Strippant = 1.89 M (NH4)2CO3, Carrier concentration = 0.1 M TOPO/n-dodecane.
Generally an increase in carrier concentration produces an increase in cation flux; however, the concurrent increase in viscosity results in a steady decrease in the diffusivity of the carrier as well as of the metal–carrier complex. Eventually the cation flux will increase as the carrier concentration is increased, as observed in the present study (Fig. 4). The curve shown in Fig. 4 indicates that initially the flux increases and reaches maxima and then decreases. This is because metal ion flux is associated with its concentration present in the feed phase. Initially, the membrane phase starts saturating with the carrier–metal complex and very small amount of metal ion transport to the receiving side hence we get low values of flux. Once the membrane phase is saturated, because of the concentration gradient the maximum metal ion transported to the receiving phase Table 3 Permeation of uranium as a function of carrier concentration (TOPO) [TOPO] [Uranium] Uranium Flux (M) receiving phase permeation (mol/m2 s) (mol/L) (%) 0.01 0.02 0.05 0.1 0.2 0.3 0.5
2.74 × 10− 4 3.70 × 10− 4 4.19 × 10− 4 4.75 × 10− 4 7.91 × 10− 4 9.68 × 10− 4 1.17 × 10− 3
21.23 28.79 32.59 37.00 61.59 75.36 91.11
1.651 × 10− 6 2.280 × 10− 6 2.863 × 10− 6 3.480 × 10− 6 5.317 × 10− 6 6.614 × 10− 6 8.684 × 10− 6
Permeability coefficient (m/s) 1.34 × 10− 5 1.95 × 10− 5 2.31 × 10− 5 3.04 × 10− 5 4.50 × 10− 5 5.97 × 10− 5 8.89 × 10− 5
Initial feed concentration: 1.28 × 10− 3 mol/L uranium in 0.5 M phosphoric acid along with 2 M nitric acid. Carrier (TOPO) concentration: different concentration of TOPO/ n-dodecane. Strippant: 1.89 M ammonium carbonate.
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therefore the flux reaches maxima. Then finally decreases as the concentration of metal ion decreases at feed side or concentration gradient decreases. It is observed that with increasing carrier concentration in an inert diluent like dodecane, uranium (VI) transport gradually increases (Fig. 5). The permeation coefficients and average flux are found to be 8.89 × 10− 5 m/s, and 8.68 × 10− 6 mol/m2 s for 0.5 M TOPO in dodecane (Table 3). It is also obvious that the transport of uranium (VI) in such a system should be a function of both the distribution coefficient and diffusion coefficient because the transfer of uranyl ions through the membrane may be considered diffusive in nature. The difference in permeability between experiments with varying feed acidity, carrier concentration and uranium molarity can be understood by considering the probable expression for the rate of formation of the diffusing species at the feed interface: d UO2 ðNO3 Þ2 d2TOPO =dt 2 2þ ¼ k ⁎ ½TOPO2 NO UO2 3
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Fig. 5. Percentage of transport of uranium as a function of concentration of TOPO/n-dodecane. Feed acidity=0.5 M H3PO4 +2 M HNO3, [U] In feed =1.28×10−3 mol/L, Strippant=1.89 M (NH4)2CO3.
Where k⁎ is rate constant for the formation of the UO22+– TOPO complex. This equation is based on the stoichiometry of uranium (VI) extraction established earlier (Mansingh et al., 1996). Rate of diffusion of metal species will thus depend upon any changes in NO3−, TOPO and UO22+ concentration in the feed side. It is important to note that the trend for transport of uranyl ions through the liquid membrane is nearly alike as that encountered with this cation in liquid–liquid extraction system (Singh et al., 2004). Fig. 5 indicates
that at all the TOPO concentrations; the trend in variation of percentage transport of metal ions is the same, i.e. increasing with the increase in the concentration of TOPO.
Fig. 4. Flux vs. time for different concentration of TOPO in n-dodecane. Feed acidity=0.5 M H3PO4 +2 M HNO3, [U] In feed=1.28×10−3 mol/L, Strippant=1.89 M (NH4)2CO3.
Fig. 6. Concentration of uranium in feed and receiver at 0.001 M H3PO4 using 0.1 M TOPO in n-dodecane and 1.89 M (NH4)2CO3 as strippant.
4. Conclusion The transport of uranium (VI) by TOPO/n-dodecane SLM is studied and found that uranium (VI) can be very well separated from phosphoric acid medium. It is established that maximum uranium permeation through SLM membranes is attained with feed acidity around
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0.001 M phosphoric acid with 2 M HNO3 and 0.5 M TOPO in n-dodecane as the carrier, 1.89 M ammonium carbonate has been found to be the best strippants for uranium (VI). Fig. 6 shows that more than 90% uranium can be recovered in 360 min from feed to the receiving phase. The driving force for the separation is not only the nitrate ions but also the H + ions which are responsible for the transport of uranyl ions. Acknowledgements The authors wish to thanks Sri S.D. Misra, Director, Nuclear recycle group, Bhabha Atomic Research centre, Mumbai for his keen interest to carry out this work and for his constant encouragement during the entire work. References Barnse, D.E., Marshall, G.D., 1995. Rapid optimization of chemical parameters affecting supported liquid membranes. Separation Science Technology 30, 751–776. Danesi, P.R., 1984. Separation of metal species by supported liquid membranes. Separation Science Technology 19, 857–894. Danesi, P.R., Horwitz, E.P., Vandeghraff, G.F., 1981. Mass transfer rate through liquid membranes: interfacial chemical reaction and diffusion as simultaneous permeability controlling factors. Separation Science and Technology 16 (2), 201–211. Davies, W., Gray, W., 1964. A rapid and specific titremetric method for the precise determination of uranium using iron(II) sulphate as reductant. Talenta 11, 1203–1211. Dietz, Mark L., Horwitz, E. Philip, Sajdak, Larry R., Chiarzia, Renato, 2001. An improved extraction chromatographic resin for the separation of uranium from acidic nitrate media. Talanta 54, 1173–1184. Divakar, D.S., Parikh, K.J., Pandit, S.S., Singh, Suman Kumar, Jambunathan, U., Ramanujam, A., 2001. Sensitive Method for the Determination of Microquantities of 233U in the Presence of Large Amounts of Thorium. INSAC, CAT, Indore, pp. 295–298. Hayworth, H.C., Ho, W.S., Burns Jr., W.A., Norman, N.Li., 1983. Extraction of uranium from wet phosphoric acid by liquid membranes. Hydrometallurgy 493–521. Hurst, F.J., Crouse, D., 1974. Recovery of uranium from wet process phosphoric acid by extraction with octylphenylphosphoric acid. Industrial Engineering Chemistry Process Development 11, 122–131. Mansingh, P.S., Veeraraghavan, R., Mahapatra, P.K., Manchanda, V.K., Dash, K.C., 1996. Adduct formation of a uranyl isoxazolonate with organophillic neutral oxodonors. Radiochemica Acta 72, 127–132.
Mukherjee, T.K., Singh, H., 2003. Recovery of Uranium and Thorium from Secondary Resources. INSAC, Kalpakkam IT/1. Musikas, C., Benjelloun, N., Lours, S., 1982. Actinide Recovery from Waste and Low Grade Sources. Marcel Decker, New York, USA, pp. 245–251. Nash, K.L., Choppin, G.R., 1997. Separation chemistry for actinides elements: recent developments and historical perspective. Separation Science Technology 32, 255–274. Ninger, D.E., 2001. Uranium exploration policy, economics, and future prospects, U.S. Atomic Energy Commission, Division of Production and Material Management, USA, IAEA-PL-490/7. Preuss, A., Kunin, A., 1958. Uranium recovery by ion exchange. In: Clegg, J.W., Foley, D.D. (Eds.), Uranium Ore Processing. Addison-Wesley, Reading, pp. 191–236. Chap. 9. Rawajfeh, Khaled M., Al Matar Ali, Kh., 2000. Uranium extraction from purified wet process Jordanian phosphoric acid: a development study. Hydrometallurgy 56, 309–322. Sastri, A.M., Kumar, A., Shukla, J.P., Singh, R.K., 1998. Improved techniques in liquid membrane separation: an overview. Separation and Purification Methods 27, 213–219. Schulz, W.W., Burger, L.L., Navratil, J.D., Bender, K.P., 1990. Application of tributyl phosphate in nuclear fuel processing. Science and Technology of Tributyl Phosphate, III. CRC Press Inc, Boca Raton, FL USA. Seaborg, G.T., Katz, J.J., 1954. The Actinide Elements, 1st ed. McGraw-Hill Book Company, Inc., New York. Shukla, J.P., Misra, S.K., 1991. Carrier mediated transport of uranyl ions across tri-butyl phosphate/dodecane liquid membranes. Journal of Membrane Science 64, 93. Singh, H., Nagle, R.A., Gariyankar, A.B., Fonseca, M.F., Koppikar, K.S., 1989. On site tests for recovery of uranium from wet process phosphoric acid at FACT. International Symposium on Uranium Technology 621–634. Singh, H., Vijyalakshmi, R., Misra, S.L., Gupta, C.K., 2001. Studies on uranium extraction from phosphoric acid using di-nonyl phenyl phosphoric acid-based synergistic mixtures. Hydrometallurgy 59, 69–76. Singh, Suman Kumar, Dhami, P.S., Vijyan, K., Jumbunathan, U., Dey, P.K., 2004. Studies on the Behaviour of Diluents on Extraction of Uranium From Phosphoric Acid Medium Using Tri-n-Octyl Phosphine Oxide as Extractant (RCEAS–2004). National seminar on role of chemistry in the emerging areas of applied science, S.V. University, Thirupathi, India, pp. 189–287. Sivaprakash, G., 1989. Uranium recovery from phosphoric acid. International Symposium on Uranium Technology, Mumbai, pp. 592–604. Visser, H.C., 1994. Supported Liquid Membranes with Improved Stability; Kinetics and Mechanism of Carrier Mediated Salt Transport. Chap. 1,2.