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Uranium recovery from simulated molybdenum-99 production residue using non-dispersive membrane based solvent extraction
Hydrometallurgy 164 (2016) 330–333
Contents lists available at ScienceDirect
Hydrometallurgy journal homepage: www.elsevier.com/locate/hydromet
Technical note
Uranium recovery from simulated molybdenum-99 production residue using non-dispersive membrane based solvent extraction Maretha Fourie a,b,⁎, Willem Carl Meyer Hunter Meyer a,c, Derik Jacobus van der Westhuizen b, Henning Manfred Krieg b a b c
South African Nuclear Energy Corporation SOC Ltd (Necsa), Elias Motsoaledi Street, R104 Pelindaba, P.O. Box 582, Pretoria 0001, South Africa Chemical Resource Beneficiation, North-West University, Potchefstroom campus, Private Bag X6001, Potchefstroom 2531, South Africa Centre for Applied Radiation Science and Technology, North-West University, Mafikeng campus, Private Bag X2046, Mmabatho 2735, South Africa
a r t i c l e
i n f o
Article history: Received 15 March 2016 Received in revised form 27 June 2016 Accepted 4 July 2016 Available online 5 July 2016 Keywords: Uranium Tributyl phosphate Fission products Acetohydroxamic acid Membrane based solvent extraction Alkaline 99Mo production residue
a b s t r a c t Commercial alkaline dissolution molybdenum-99 (99Mo) producers are currently not recovering enriched uranium (U) from 99Mo production waste which has a unique actinide and fission product composition. During the development of a process to recover U from 99Mo production residue, the suitability of non-dispersive membrane based solvent extraction (MBSX) for this process was investigated. The residue was dissolved in nitric acid (HNO3) and extracted using tributyl phosphate (TBP). Acetohydroxamic acid (AHA) was added to the feed to prevent co-extraction of plutonium (Pu). Complete U extraction with MBSX was achieved at low U(VI) concentrations in the feed while more than one MBSX contacting step was required for high U(VI) concentrations. Ammonium carbonate ((NH4)2CO3) was investigated as an alternative stripping solution to traditionally used dilute HNO3. Almost complete U(VI) stripping (99%) from 7500 mg L−1 U(VI) in the organic phase was obtained in a single contacting step with 0.5 M (NH4)2CO3. Practically none of the characteristic fission products present in the 99Mo production residue were extracted with AHA and also neither influencing U(VI) nor fission product extraction. The study has shown that MBSX was suitable for selective extraction (with AHA added) and stripping (0.5 M (NH4)2CO3) of U(VI) from a simulated 99Mo production residue. © 2016 Elsevier B.V. All rights reserved.
1. Introduction U recovery from 99Mo production waste will decrease the residue volumes that have to be disposed of while the recovered enriched U can be used during isotope production, thereby reducing production costs. During alkaline dissolution of the target plates used in the 99Mo process, a residue precipitates that contain most of the enriched U. The U in this residue can be dissolved in a (NH4)2CO3/H2O2 solution, which can then be further purified using firstly ion exchange followed by solvent extraction (Stassen and Suthiram, 2015). Ammonia (NH3) and carbon dioxide (CO2) are released during steam stripping of the U containing (NH4)2CO3 solution which then forms uranium trioxide (UO3·2H2O) (Kweto et al., 2012), that can be dissolved in HNO3 similar to what is being done during spent fuel dissolution. While various different technologies such as dispersive solvent extraction, ion exchange and precipitation have been investigated and
⁎ Corresponding author at: South African Nuclear Energy Corporation SOC Ltd (Necsa), Elias Motsoaledi Street, R104 Pelindaba, P.O. Box 582, Pretoria 0001, South Africa. E-mail addresses: [email protected] (M. Fourie), [email protected] (W.C.M.H. Meyer), [email protected] (D.J. van der Westhuizen), [email protected] (H.M. Krieg).
http://dx.doi.org/10.1016/j.hydromet.2016.07.001 0304-386X/© 2016 Elsevier B.V. All rights reserved.
used for the recovery of U from liquid waste (Gupta et al., 2005), they pose specific disadvantages when used in the nuclear industry (Dixit et al., 2012; Gabelman and Hwang, 1999). A promising contacting method entails the use of a micro-porous hollow fiber membrane contactor for non-dispersive solvent extraction. Advantages of using membrane contactors include a constant and large contact surface for mass transfer which is independent on flow rates, being able to significantly vary phase ratios, leading to reduced secondary waste volumes (Rathore et al., 2001), high energy efficiencies, while the process is environmentally benign and easy to operate (Pabby and Sastre, 2013). In addition the processing and maintenance costs are low (Roy et al., 2010), the modular design of the membrane contactor ensures easy scale up possibilities (Dixit et al., 2012) and their small footprint makes them ideal to use in hot cells and glove boxes. In general, the nuclear industry uses the PUREX process with TBP (Tkac et al., 2008) to recover U from spent fuel. Additionally AHA can be added to the aqueous feed solution to reduce Pu(IV) to Pu(III) while also forming an unextractable hydrophilic Pu complex (Paulenova et al., 2009; Taylor et al., 1998). Traditionally (PUREX process) dilute HNO3 is used as the stripping solution (Gupta et al., 2007) however, more than one stripping contact is required for complete U(VI) recovery (Tkac et al., 2008) due to the
M. Fourie et al. / Hydrometallurgy 164 (2016) 330–333
unfavorable D of U(VI) in TBP and dilute HNO3 (Gupta et al., 2005). It has been shown that U(VI) forms stable uranyl-tricarbonate (UO2(CO3)−4 3 ) complexes when stripped from TBP using carbonate solutions at high pH values, which occurs naturally when using (NH4)2CO3 due to an ammonia-ammonium buffer at pH = 9 (Kweto et al., 2012). In addition, (NH4)2CO3 is a milder alkali than sodium carbonate (Na2CO3) or sodium bicarbonate (NaHCO3) leading to a reduced chance of U(VI) precipitation (Dixit et al., 2012). (NH4)2CO3 also decomposes above 60 °C to form NH3, CO2 and H2O (Kweto et al., 2012) which will reduce secondary waste volumes created during U(VI) recovery. While the extraction of U(VI) using TBP is well documented, it has, to our knowledge, never been investigated for the recovery of U(VI) from the residue formed during the alkaline dissolution of U/Al target plates during 99Mo production with its unique composition, specifically using MBSX technologies. It was accordingly the aim of this study to develop a MBSX based process to recover U(VI) from a simulated 99Mo production residue in the presence of AHA using TBP for the extraction and (NH4)2CO3 for the stripping. 2. Materials and methods All of the chemicals used were of A.R. grade and used without further purification. Uranyl nitrate (UO2(NO3)2) stock solutions were prepared by dissolving natural uranium(VI) oxide in 3 M HNO3 (70% sourced from Sigma Aldrich) and milli-Q water at 60 °C, stirred at 250 rpm for 60 min. The stock solution contained U(VI) (83,050 ± 0.5 mg L−1), K (12.5 ± 2.8 mg L−1), Na (8.0 ± 1.7 mg L−1), Al (20.2 ± 1.3 mg L−1), Mo (3.4 ± 0.3 mg L−1) and W (1.0 ± 0.2 mg L−1). TBP (97%) and kerosene (reagent grade) were sourced from Sigma-Aldrich. ICP (inductively coupled plasma) standard solutions (1000 mg L−1) for Co, Cs, Ru, Sb and Sr from Ultraspec were used as fission product surrogates. Dilute HNO3 and 0.5 M (NH4)2CO3 (sourced from Sigma Aldrich) stripping solutions were prepared using milli-Q water. AHA (sourced from Sigma Aldrich) was dissolved in the aqueous feed containing U(VI), HNO3 and milli-Q water. Isopropyl alcohol (C3H8O) from Merck and milli-Q water was used to clean the polypropylene hollow fiber membrane contactor. The MBSX set-up and procedure (Gupta et al., 2005) contained Cole Parmer gear pumps, pressure gauges from Wika, Aalborg PTFE flowmeters, a polypropylene Liqui-Cel™ Extra-Flow 2.5 × 8″ membrane contactor, polyethylene tubing and perfluoroalkoxy alkane (PFA) fittings. U(VI) concentrations were analysed spectrophotometrically with the carbonate (Rodden, 1950) and Bromo-Padap (Johnson and Florence, 1971) methods on a Perkin-Elmer Lambda 1050 High Performance UV/Vis/NIR spectrophotometer, using the “Wavelength Quant” method. In addition to ICP, U(VI) concentrations were also analysed with X-Ray fluorescence (XRF) on a Panalytical Axios instrument with the measurement done in the liquid samples under a helium medium. Analysis assurance was done using direct spikes of Rh in the samples.
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Table 1 U(VI) extraction from simulated 99Mo production residue containing 7500 mg L−1 U(VI) at 25 °C and A/O = 1:1 after 30 min. Time (min)
% Extraction
5 10 15 20 30
91.7 ± 7.0 95.8 ± 5.1 95.2 ± 5.6 95.0 ± 5.7 94.6 ± 5.5
saturated with the UO2(NO3)2·2TBP complex slowing the diffusion into the bulk organic phase resulting in an decreased mass transfer rate and extraction efficiency. This tendency was also observed in other studies for a variety of metals (Kumar et al., 2005; Kumar et al., 2011). While the saturation of TBP could be overcome by increasing the TBP concentration, it would i) lead to an increase in the organic phase's viscosity resulting in a lower diffusion coefficient for the UO2(NO3)2·2TBP complex (Davies and Gray, 1964; Roy et al., 2010), ii) result in increased HNO3-TBP complexation thereby reducing the HNO3 concentration in the feed, leading again to lower U(VI) extraction efficiencies (Gupta et al., 2007; Roy et al., 2010) and iii)increase the probability of extracting unwanted fission products such as Zr and Nb (Gupta et al., 2005). This confirms that complete U(VI) extraction from a simulated 99Mo production residue with a single MBSX contact (or membrane contactor) is only possible at lower U(VI) concentrations. Another approach to achieve complete U(VI) extraction from simulated 99Mo production residue with high U(VI) concentrations, would be to repeat the extraction in a second membrane contactor with fresh TBP (Kumar et al., 2005). In this study two MBSX contacting steps were used with fresh TBP for each contact applied to a concentrated U(VI) solution (98,000 ± 6000 mg L−1). After 30 min each, using an A/O ratio of 1:1, 87.7 ± 2.3% and 96.8 ± 1.5% were recovered in the 1st and 2nd contacting step respectively. It is clear that a two-step extraction using fresh TBP in both yielded a satisfactory recovery of concentrated U(VI) from a simulated 99Mo production residue feed, confirming results obtained in literature (Gupta et al., 2007). 3.2. The influence of AHA on U(VI) extraction It has been shown with actual dissolved 99Mo production residue (Goede, 2012) that Pu extraction decreased with increasing U and AHA concentration. Since it was shown that no Pu extraction was
3. Results and discussion 3.1. Extraction For optimal extraction 30% (v/v) TBP in kerosene (Dixit et al., 2012; Roy et al., 2010) was used to extract U(VI) from a simulated 99Mo production residue feed containing 3 M HNO3 (Pathak et al., 2013; Roy et al., 2010) and 7500 mg L−1 U(VI). The two phases were contacted for 30 min (in triplicate) in the MBSX set-up. The results in Table 1 confirmed that after 10 min 96% U(VI) was extracted. The fast extraction can be ascribed to the membrane contactor's large contact surface area (1.4 m2) resulting in the high U(VI) mass transfer rates attained (Dixit et al., 2012). When contacting various U(VI) concentrations (Fig. 1) complete extraction from 100 mg L1 U(VI) was attained in less than 5 min. However, at higher U(VI) concentrations TBP in the membrane pores became
Fig. 1. U(VI) extraction from simulated 99Mo production residue at 25 °C, A/O = 1:1 as a function of the U(VI) concentration and time.
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Table 2 The influence of 0.1 M AHA and 98,000 ± 6000 mg L−1 U(VI) in simulated 99Mo production residue feed using two MBSX contacting steps with fresh TBP each contact at 25 °C and A/O = 1:1 for 30 min. [AHA]
% Extraction
(M) 0 0.1
Contact step 1 85.4 ± 4.3 87.7 ± 4.4
Contact step 2 95.5 ± 4.8 96.4 ± 4.8
observed at a U:Pu ratio of 1154:1 with 0.1 M AHA from an irradiated 99 Mo production residue (Stassen and Suthiram, 2015), MBSX runs were conducted with 0.1 M AHA using a feed of 98,000 ± 6000 mg L−1U(VI) in the simulated 99Mo production residue and again two MBSX contacting steps with fresh TBP for each contact (Table 2). It is clear that 0.1 M AHA did not influence U(VI) extraction, as confirmed by previous researchers (Schroeder et al., 2001; Tkac et al., 2008). AHA is able to form a complex with U(VI) in both the aqueous (UO2(NO3)(AHA)) and organic phases (UO2(NO3)(AHA)·2TBP) by forming a hydrogen-bond with the polar TBP. However, only a small fraction of AHA is extracted into the TBP phase, decreasing with an increasing HNO3 concentration in the aqueous feed solution and therefore the amount of AHA extracted into the organic phase will be negligible (Pathak et al., 2013; Tkac et al., 2008). Literature also showed that the stability constants for U(VI)-AHA is lower than for the Pu(IV)-AHA complex (Chatterjee, 1978; Dutt and Seshadri, 1968) and therefore U(VI) will not preferentially form complexes with AHA in the presence of Pu(IV) (Taylor et al., 1998). 3.3. Fission product extraction and the influence of AHA on fission product extraction After the 99Mo production residue was dissolved in the (NH4)2CO3/ H2O2 solution and the U(VI) purified with ion exchange on alumina, 60 Co, 137Cs, 106Ru, 125Sb, 90Sr and 239Pu still remained in solution at concentrations below 1 mg L− 1. Hence, in order to determine whether final purification with TBP in MBSX could be achieved, fission product surrogates of 60Co, 137Cs, 106Ru, 125Sb and 90Sr (1 mg L−1) were added to the simulated 99Mo production residue feed solution in the presence and absence of 0.1 M AHA. The results are presented in Table 3. Fission product extraction yields obtained in Table 3 were below the levels of experimental uncertainty and therefore it can be assumed that none of the fission products were extracted with 30% TBP/kerosene even at concentrations above those obtained from the alumina column's eluate. These results were confirmed from literature showing that 30% TBP/kerosene is a highly selective U(VI) extractant (Kumar et al., 2011; Kumbasar, 2010; Roy et al., 2010). When comparing the extractions in the absence and presence of AHA, it is clear that AHA did not influence U(VI) or the fission product extractions again confirming other studies (Dixit et al., 2012; Hlushak et al., 2011; Pathak et al., 2013). 3.4. Stripping To determine the stripping of a loaded organic phase containing 7500 mg L−1 U(VI), a 0.5 M (NH4)2CO3 solution was used. According to the results in Figs. 2, 99% of U(VI) in the loaded organic phase could
Table 3 Extraction from a feed solution containing a multi-component mixture of 60Co, 137Cs, 106 Ru, 125Sb and 90Sr surrogates (1 mg L−1) at 25 °C for 20 min with A/O = 1:1 in the presence and absence of 0.1 M AHA in the feed. %
Co
Cs
Ru
Sb
Sr
Extraction 6.9 ± 2.4 3.3 ± 1.4 3.8 ± 0.8 4.4 ± 1.2 2.0 ± 2.4 Extraction (0.1 M AHA) 5.4 ± 1.2 2.8 ± 1.2 3.1 ± 0.9 4.2 ± 2.1 1.8 ± 1.8 Experimental uncertainty 9.6 4.9 4.8 6.7 5.4
Fig. 2. The recovery of U(VI) with 0.5 M (NH4)2CO3 from a loaded organic phase containing 7500 mg L−1 U(VI) using MBSX at 25 °C, A/O = 1:1 for 210 min.
be recovered in a single contacting step after 210 min. While similar efficiencies were attained using Na2CO3 and NaHCO3 stripping solutions (Gabelman and Hwang, 1999; Pabby and Sastre, 2013; Yang et al., 1996), the disadvantages of carbonate solutions (see Introduction) were avoided. The 210 min required to reach equilibrium can be ascribed to the high D value of U(VI) in (NH4)2CO3, implying that the majority of the mass transfer resistance is situated in the organic phase. Since the hydrophobic polypropylene membrane contactor is wetted by the organic phase, the membrane mass transfer resistance is added to the total mass transfer resistance further increasing the time required to reach equilibrium (Gabelman and Hwang, 1999). The added membrane mass transfer resistance could be avoided by using a hydrophilic membrane contactor where the aqueous phase will wet the membrane resulting in a negligible membrane mass transfer resistance (Gupta et al., 2007; Kumar et al., 2005; Pabby and Sastre, 2013).
4. Conclusions Complete U(VI) extraction in a single MBSX contacting step (or membrane contactor) could be achieved at 100 mg L−1 U(VI). However, in order to completely extract U(VI) from higher U(VI) concentrations such as 98,000 mg L− 1, two MBSX contacting steps (or membrane contactors) were required. Results showed that 0.1 M AHA did not influence U(VI) extraction. Co, Cs, Ru, Sb and Sr surrogates were not extracted with 30% TBP/kerosene even at fission product concentrations above those present in the alumina column's eluate. Again AHA did not influence fission product extraction. Almost complete (99%) U(VI) recovery from the loaded organic phase (7500 mg L−1U(VI)) was achieved with 0.5 M (NH4)2CO3 in a single contacting step after 210 min. The study has shown that MBSX can successfully be used in a process to recover U(VI) from a simulated 99Mo production residue in the presence of AHA using TBP for extraction and (NH4)2CO3 for stripping.
Acknowledgements This study was financially supported by the National Research Foundation of South Africa (NRF PDP 90592) and Necsa (South African Nuclear Energy Corporation SOC Ltd). The authors wish to express their sincere gratitude to Mike Britton, Josias Mokgawa, Lize Stassen, Amelia Goede and the rest of the Nuclear Waste Research group at Necsa for their kind consultation and help.
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Contents lists available at ScienceDirect
Hydrometallurgy journal homepage: www.elsevier.com/locate/hydromet
Technical note
Uranium recovery from simulated molybdenum-99 production residue using non-dispersive membrane based solvent extraction Maretha Fourie a,b,⁎, Willem Carl Meyer Hunter Meyer a,c, Derik Jacobus van der Westhuizen b, Henning Manfred Krieg b a b c
South African Nuclear Energy Corporation SOC Ltd (Necsa), Elias Motsoaledi Street, R104 Pelindaba, P.O. Box 582, Pretoria 0001, South Africa Chemical Resource Beneficiation, North-West University, Potchefstroom campus, Private Bag X6001, Potchefstroom 2531, South Africa Centre for Applied Radiation Science and Technology, North-West University, Mafikeng campus, Private Bag X2046, Mmabatho 2735, South Africa
a r t i c l e
i n f o
Article history: Received 15 March 2016 Received in revised form 27 June 2016 Accepted 4 July 2016 Available online 5 July 2016 Keywords: Uranium Tributyl phosphate Fission products Acetohydroxamic acid Membrane based solvent extraction Alkaline 99Mo production residue
a b s t r a c t Commercial alkaline dissolution molybdenum-99 (99Mo) producers are currently not recovering enriched uranium (U) from 99Mo production waste which has a unique actinide and fission product composition. During the development of a process to recover U from 99Mo production residue, the suitability of non-dispersive membrane based solvent extraction (MBSX) for this process was investigated. The residue was dissolved in nitric acid (HNO3) and extracted using tributyl phosphate (TBP). Acetohydroxamic acid (AHA) was added to the feed to prevent co-extraction of plutonium (Pu). Complete U extraction with MBSX was achieved at low U(VI) concentrations in the feed while more than one MBSX contacting step was required for high U(VI) concentrations. Ammonium carbonate ((NH4)2CO3) was investigated as an alternative stripping solution to traditionally used dilute HNO3. Almost complete U(VI) stripping (99%) from 7500 mg L−1 U(VI) in the organic phase was obtained in a single contacting step with 0.5 M (NH4)2CO3. Practically none of the characteristic fission products present in the 99Mo production residue were extracted with AHA and also neither influencing U(VI) nor fission product extraction. The study has shown that MBSX was suitable for selective extraction (with AHA added) and stripping (0.5 M (NH4)2CO3) of U(VI) from a simulated 99Mo production residue. © 2016 Elsevier B.V. All rights reserved.
1. Introduction U recovery from 99Mo production waste will decrease the residue volumes that have to be disposed of while the recovered enriched U can be used during isotope production, thereby reducing production costs. During alkaline dissolution of the target plates used in the 99Mo process, a residue precipitates that contain most of the enriched U. The U in this residue can be dissolved in a (NH4)2CO3/H2O2 solution, which can then be further purified using firstly ion exchange followed by solvent extraction (Stassen and Suthiram, 2015). Ammonia (NH3) and carbon dioxide (CO2) are released during steam stripping of the U containing (NH4)2CO3 solution which then forms uranium trioxide (UO3·2H2O) (Kweto et al., 2012), that can be dissolved in HNO3 similar to what is being done during spent fuel dissolution. While various different technologies such as dispersive solvent extraction, ion exchange and precipitation have been investigated and
⁎ Corresponding author at: South African Nuclear Energy Corporation SOC Ltd (Necsa), Elias Motsoaledi Street, R104 Pelindaba, P.O. Box 582, Pretoria 0001, South Africa. E-mail addresses: [email protected] (M. Fourie), [email protected] (W.C.M.H. Meyer), [email protected] (D.J. van der Westhuizen), [email protected] (H.M. Krieg).
http://dx.doi.org/10.1016/j.hydromet.2016.07.001 0304-386X/© 2016 Elsevier B.V. All rights reserved.
used for the recovery of U from liquid waste (Gupta et al., 2005), they pose specific disadvantages when used in the nuclear industry (Dixit et al., 2012; Gabelman and Hwang, 1999). A promising contacting method entails the use of a micro-porous hollow fiber membrane contactor for non-dispersive solvent extraction. Advantages of using membrane contactors include a constant and large contact surface for mass transfer which is independent on flow rates, being able to significantly vary phase ratios, leading to reduced secondary waste volumes (Rathore et al., 2001), high energy efficiencies, while the process is environmentally benign and easy to operate (Pabby and Sastre, 2013). In addition the processing and maintenance costs are low (Roy et al., 2010), the modular design of the membrane contactor ensures easy scale up possibilities (Dixit et al., 2012) and their small footprint makes them ideal to use in hot cells and glove boxes. In general, the nuclear industry uses the PUREX process with TBP (Tkac et al., 2008) to recover U from spent fuel. Additionally AHA can be added to the aqueous feed solution to reduce Pu(IV) to Pu(III) while also forming an unextractable hydrophilic Pu complex (Paulenova et al., 2009; Taylor et al., 1998). Traditionally (PUREX process) dilute HNO3 is used as the stripping solution (Gupta et al., 2007) however, more than one stripping contact is required for complete U(VI) recovery (Tkac et al., 2008) due to the
M. Fourie et al. / Hydrometallurgy 164 (2016) 330–333
unfavorable D of U(VI) in TBP and dilute HNO3 (Gupta et al., 2005). It has been shown that U(VI) forms stable uranyl-tricarbonate (UO2(CO3)−4 3 ) complexes when stripped from TBP using carbonate solutions at high pH values, which occurs naturally when using (NH4)2CO3 due to an ammonia-ammonium buffer at pH = 9 (Kweto et al., 2012). In addition, (NH4)2CO3 is a milder alkali than sodium carbonate (Na2CO3) or sodium bicarbonate (NaHCO3) leading to a reduced chance of U(VI) precipitation (Dixit et al., 2012). (NH4)2CO3 also decomposes above 60 °C to form NH3, CO2 and H2O (Kweto et al., 2012) which will reduce secondary waste volumes created during U(VI) recovery. While the extraction of U(VI) using TBP is well documented, it has, to our knowledge, never been investigated for the recovery of U(VI) from the residue formed during the alkaline dissolution of U/Al target plates during 99Mo production with its unique composition, specifically using MBSX technologies. It was accordingly the aim of this study to develop a MBSX based process to recover U(VI) from a simulated 99Mo production residue in the presence of AHA using TBP for the extraction and (NH4)2CO3 for the stripping. 2. Materials and methods All of the chemicals used were of A.R. grade and used without further purification. Uranyl nitrate (UO2(NO3)2) stock solutions were prepared by dissolving natural uranium(VI) oxide in 3 M HNO3 (70% sourced from Sigma Aldrich) and milli-Q water at 60 °C, stirred at 250 rpm for 60 min. The stock solution contained U(VI) (83,050 ± 0.5 mg L−1), K (12.5 ± 2.8 mg L−1), Na (8.0 ± 1.7 mg L−1), Al (20.2 ± 1.3 mg L−1), Mo (3.4 ± 0.3 mg L−1) and W (1.0 ± 0.2 mg L−1). TBP (97%) and kerosene (reagent grade) were sourced from Sigma-Aldrich. ICP (inductively coupled plasma) standard solutions (1000 mg L−1) for Co, Cs, Ru, Sb and Sr from Ultraspec were used as fission product surrogates. Dilute HNO3 and 0.5 M (NH4)2CO3 (sourced from Sigma Aldrich) stripping solutions were prepared using milli-Q water. AHA (sourced from Sigma Aldrich) was dissolved in the aqueous feed containing U(VI), HNO3 and milli-Q water. Isopropyl alcohol (C3H8O) from Merck and milli-Q water was used to clean the polypropylene hollow fiber membrane contactor. The MBSX set-up and procedure (Gupta et al., 2005) contained Cole Parmer gear pumps, pressure gauges from Wika, Aalborg PTFE flowmeters, a polypropylene Liqui-Cel™ Extra-Flow 2.5 × 8″ membrane contactor, polyethylene tubing and perfluoroalkoxy alkane (PFA) fittings. U(VI) concentrations were analysed spectrophotometrically with the carbonate (Rodden, 1950) and Bromo-Padap (Johnson and Florence, 1971) methods on a Perkin-Elmer Lambda 1050 High Performance UV/Vis/NIR spectrophotometer, using the “Wavelength Quant” method. In addition to ICP, U(VI) concentrations were also analysed with X-Ray fluorescence (XRF) on a Panalytical Axios instrument with the measurement done in the liquid samples under a helium medium. Analysis assurance was done using direct spikes of Rh in the samples.
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Table 1 U(VI) extraction from simulated 99Mo production residue containing 7500 mg L−1 U(VI) at 25 °C and A/O = 1:1 after 30 min. Time (min)
% Extraction
5 10 15 20 30
91.7 ± 7.0 95.8 ± 5.1 95.2 ± 5.6 95.0 ± 5.7 94.6 ± 5.5
saturated with the UO2(NO3)2·2TBP complex slowing the diffusion into the bulk organic phase resulting in an decreased mass transfer rate and extraction efficiency. This tendency was also observed in other studies for a variety of metals (Kumar et al., 2005; Kumar et al., 2011). While the saturation of TBP could be overcome by increasing the TBP concentration, it would i) lead to an increase in the organic phase's viscosity resulting in a lower diffusion coefficient for the UO2(NO3)2·2TBP complex (Davies and Gray, 1964; Roy et al., 2010), ii) result in increased HNO3-TBP complexation thereby reducing the HNO3 concentration in the feed, leading again to lower U(VI) extraction efficiencies (Gupta et al., 2007; Roy et al., 2010) and iii)increase the probability of extracting unwanted fission products such as Zr and Nb (Gupta et al., 2005). This confirms that complete U(VI) extraction from a simulated 99Mo production residue with a single MBSX contact (or membrane contactor) is only possible at lower U(VI) concentrations. Another approach to achieve complete U(VI) extraction from simulated 99Mo production residue with high U(VI) concentrations, would be to repeat the extraction in a second membrane contactor with fresh TBP (Kumar et al., 2005). In this study two MBSX contacting steps were used with fresh TBP for each contact applied to a concentrated U(VI) solution (98,000 ± 6000 mg L−1). After 30 min each, using an A/O ratio of 1:1, 87.7 ± 2.3% and 96.8 ± 1.5% were recovered in the 1st and 2nd contacting step respectively. It is clear that a two-step extraction using fresh TBP in both yielded a satisfactory recovery of concentrated U(VI) from a simulated 99Mo production residue feed, confirming results obtained in literature (Gupta et al., 2007). 3.2. The influence of AHA on U(VI) extraction It has been shown with actual dissolved 99Mo production residue (Goede, 2012) that Pu extraction decreased with increasing U and AHA concentration. Since it was shown that no Pu extraction was
3. Results and discussion 3.1. Extraction For optimal extraction 30% (v/v) TBP in kerosene (Dixit et al., 2012; Roy et al., 2010) was used to extract U(VI) from a simulated 99Mo production residue feed containing 3 M HNO3 (Pathak et al., 2013; Roy et al., 2010) and 7500 mg L−1 U(VI). The two phases were contacted for 30 min (in triplicate) in the MBSX set-up. The results in Table 1 confirmed that after 10 min 96% U(VI) was extracted. The fast extraction can be ascribed to the membrane contactor's large contact surface area (1.4 m2) resulting in the high U(VI) mass transfer rates attained (Dixit et al., 2012). When contacting various U(VI) concentrations (Fig. 1) complete extraction from 100 mg L1 U(VI) was attained in less than 5 min. However, at higher U(VI) concentrations TBP in the membrane pores became
Fig. 1. U(VI) extraction from simulated 99Mo production residue at 25 °C, A/O = 1:1 as a function of the U(VI) concentration and time.
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Table 2 The influence of 0.1 M AHA and 98,000 ± 6000 mg L−1 U(VI) in simulated 99Mo production residue feed using two MBSX contacting steps with fresh TBP each contact at 25 °C and A/O = 1:1 for 30 min. [AHA]
% Extraction
(M) 0 0.1
Contact step 1 85.4 ± 4.3 87.7 ± 4.4
Contact step 2 95.5 ± 4.8 96.4 ± 4.8
observed at a U:Pu ratio of 1154:1 with 0.1 M AHA from an irradiated 99 Mo production residue (Stassen and Suthiram, 2015), MBSX runs were conducted with 0.1 M AHA using a feed of 98,000 ± 6000 mg L−1U(VI) in the simulated 99Mo production residue and again two MBSX contacting steps with fresh TBP for each contact (Table 2). It is clear that 0.1 M AHA did not influence U(VI) extraction, as confirmed by previous researchers (Schroeder et al., 2001; Tkac et al., 2008). AHA is able to form a complex with U(VI) in both the aqueous (UO2(NO3)(AHA)) and organic phases (UO2(NO3)(AHA)·2TBP) by forming a hydrogen-bond with the polar TBP. However, only a small fraction of AHA is extracted into the TBP phase, decreasing with an increasing HNO3 concentration in the aqueous feed solution and therefore the amount of AHA extracted into the organic phase will be negligible (Pathak et al., 2013; Tkac et al., 2008). Literature also showed that the stability constants for U(VI)-AHA is lower than for the Pu(IV)-AHA complex (Chatterjee, 1978; Dutt and Seshadri, 1968) and therefore U(VI) will not preferentially form complexes with AHA in the presence of Pu(IV) (Taylor et al., 1998). 3.3. Fission product extraction and the influence of AHA on fission product extraction After the 99Mo production residue was dissolved in the (NH4)2CO3/ H2O2 solution and the U(VI) purified with ion exchange on alumina, 60 Co, 137Cs, 106Ru, 125Sb, 90Sr and 239Pu still remained in solution at concentrations below 1 mg L− 1. Hence, in order to determine whether final purification with TBP in MBSX could be achieved, fission product surrogates of 60Co, 137Cs, 106Ru, 125Sb and 90Sr (1 mg L−1) were added to the simulated 99Mo production residue feed solution in the presence and absence of 0.1 M AHA. The results are presented in Table 3. Fission product extraction yields obtained in Table 3 were below the levels of experimental uncertainty and therefore it can be assumed that none of the fission products were extracted with 30% TBP/kerosene even at concentrations above those obtained from the alumina column's eluate. These results were confirmed from literature showing that 30% TBP/kerosene is a highly selective U(VI) extractant (Kumar et al., 2011; Kumbasar, 2010; Roy et al., 2010). When comparing the extractions in the absence and presence of AHA, it is clear that AHA did not influence U(VI) or the fission product extractions again confirming other studies (Dixit et al., 2012; Hlushak et al., 2011; Pathak et al., 2013). 3.4. Stripping To determine the stripping of a loaded organic phase containing 7500 mg L−1 U(VI), a 0.5 M (NH4)2CO3 solution was used. According to the results in Figs. 2, 99% of U(VI) in the loaded organic phase could
Table 3 Extraction from a feed solution containing a multi-component mixture of 60Co, 137Cs, 106 Ru, 125Sb and 90Sr surrogates (1 mg L−1) at 25 °C for 20 min with A/O = 1:1 in the presence and absence of 0.1 M AHA in the feed. %
Co
Cs
Ru
Sb
Sr
Extraction 6.9 ± 2.4 3.3 ± 1.4 3.8 ± 0.8 4.4 ± 1.2 2.0 ± 2.4 Extraction (0.1 M AHA) 5.4 ± 1.2 2.8 ± 1.2 3.1 ± 0.9 4.2 ± 2.1 1.8 ± 1.8 Experimental uncertainty 9.6 4.9 4.8 6.7 5.4
Fig. 2. The recovery of U(VI) with 0.5 M (NH4)2CO3 from a loaded organic phase containing 7500 mg L−1 U(VI) using MBSX at 25 °C, A/O = 1:1 for 210 min.
be recovered in a single contacting step after 210 min. While similar efficiencies were attained using Na2CO3 and NaHCO3 stripping solutions (Gabelman and Hwang, 1999; Pabby and Sastre, 2013; Yang et al., 1996), the disadvantages of carbonate solutions (see Introduction) were avoided. The 210 min required to reach equilibrium can be ascribed to the high D value of U(VI) in (NH4)2CO3, implying that the majority of the mass transfer resistance is situated in the organic phase. Since the hydrophobic polypropylene membrane contactor is wetted by the organic phase, the membrane mass transfer resistance is added to the total mass transfer resistance further increasing the time required to reach equilibrium (Gabelman and Hwang, 1999). The added membrane mass transfer resistance could be avoided by using a hydrophilic membrane contactor where the aqueous phase will wet the membrane resulting in a negligible membrane mass transfer resistance (Gupta et al., 2007; Kumar et al., 2005; Pabby and Sastre, 2013).
4. Conclusions Complete U(VI) extraction in a single MBSX contacting step (or membrane contactor) could be achieved at 100 mg L−1 U(VI). However, in order to completely extract U(VI) from higher U(VI) concentrations such as 98,000 mg L− 1, two MBSX contacting steps (or membrane contactors) were required. Results showed that 0.1 M AHA did not influence U(VI) extraction. Co, Cs, Ru, Sb and Sr surrogates were not extracted with 30% TBP/kerosene even at fission product concentrations above those present in the alumina column's eluate. Again AHA did not influence fission product extraction. Almost complete (99%) U(VI) recovery from the loaded organic phase (7500 mg L−1U(VI)) was achieved with 0.5 M (NH4)2CO3 in a single contacting step after 210 min. The study has shown that MBSX can successfully be used in a process to recover U(VI) from a simulated 99Mo production residue in the presence of AHA using TBP for extraction and (NH4)2CO3 for stripping.
Acknowledgements This study was financially supported by the National Research Foundation of South Africa (NRF PDP 90592) and Necsa (South African Nuclear Energy Corporation SOC Ltd). The authors wish to express their sincere gratitude to Mike Britton, Josias Mokgawa, Lize Stassen, Amelia Goede and the rest of the Nuclear Waste Research group at Necsa for their kind consultation and help.
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