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Extraction of uranium from solid waste containing uranium and fluorine
Minerals Engineering 61 (2014) 32–39
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Extraction of uranium from solid waste containing uranium and fluorine Yusuke Ohashi ⇑, Shinji Murashita, Mitsuo Nomura Ningyo-toge Environmental Engineering Center, Japan Atomic Energy Agency, 1550 Kamisaibara, Kagamino-cho, Tomata-gun, Okayama 708-0698, Japan
a r t i c l e
i n f o
Article history: Received 21 October 2013 Accepted 5 March 2014 Available online 27 March 2014 Keywords: Uranium Fluorine Silicofluoride Solvent extraction
a b s t r a c t A significant amount of waste residue and adsorbent containing uranium generated by uranium refining and conversion R&D activities has been stored at the Ningyo-toge Environmental Engineering Center of the Japan Atomic Energy Agency. The treatment of these materials is one of the most important tasks in the decommissioning of the plant. These solid wastes also contain significant quantities of fluorine, and it is important to separate this out in order to recover pure uranium as a resource. As one method for separating fluorine from uranium, we have proposed the use of solvent extraction using tri-n-octylamine (TNOA), and silicofluoride (Na2SiF6) precipitation. The decontamination factor of fluorine for solutions, prepared by dissolving spent sodium fluoride (NaF) and uranium tetrafluoride (UF4) residue into sulfuric acid, became 5.2 and 6.6 respectively when the TNOA solvent extraction process was performed. The Na2SiF6 process using SiO2 is more effective in separating fluorine from a solution prepared by dissolving spent NaF and UF4 residue using sulfuric and hydrochloric acid, with the decontamination factor of fluorine being 14 and 28. Uranium remaining in solution after removal of fluorine was recovered as uranium peroxide. The fluorine content of uranium recovered from UF4 by SiO2 treatment was as low as 0.012 wt%, and that of uranium recovered from NaF was below 0.0075%. This is in accordance with the target conditions defined by the ASTM standard. Ó 2014 Elsevier Ltd. All rights reserved.
1. Introduction Uranium deposits were first discovered in the Ningyo-toge area in 1955. A test milling facility constructed in 1965 was used to establish the fundamental technology of the PNC process, producing uranium tetrafluoride directly from uranium ore. A Refining and Conversion Pilot Plant using this process was constructed in 1981 (Amamoto and Wakabayashi, 1993). A total of 385 t of UF6 had been produced in the plant by 1991, when its operation was discontinued. A considerable amount of data on the design, construction, and operation of a commercial plant was obtained from this facility. However, as no promising overseas uranium deposits were discovered, the process has not been used for practical commercial purposes. Small-scale reprocessed-uranium conversion testing was carried out at the Refining and Conversion Plant from 1982, in order to develop fundamental technologies for converting reprocessed UO3 into UF6. Thereafter, middle-scale testing and practical application testing was carried out. During demonstration testing, which continued until 2001, approximately 293 t of UF6 was produced. Consequently, both the practical application and economic ⇑ Corresponding author. E-mail address: [email protected] (Y. Ohashi). http://dx.doi.org/10.1016/j.mineng.2014.03.003 0892-6875/Ó 2014 Elsevier Ltd. All rights reserved.
prospects for reprocessed-uranium conversion technology were successfully demonstrated (Amamoto, 2000). On the other hand, as a result of the aforementioned tests a significant amount of scrap uranium, and adsorbent containing uranium, was generated. The treatment of scrap uranium and adsorbent is one of the most pressing issues in the decommissioning of the plant. This waste totals approximately 1500 tons, and is presently stored in drums. The uranium concentration for some varieties of this waste exceeds 50 wt%. Hence, it is necessary to recover as much uranium as possible from the waste and scrap uranium, in order to be able to dispose of it more easily and to store the uranium as a resource (Japan Atomic Energy Commission, 2000).
1.1. Uranium recovery method The production of yellow cake from ores is an essential step in the preparation of fuel for use in nuclear power plants. Sulfuric acid has been used for dissolving uranium ore, and this same method is expected to be applied to the recovery of uranium from scrap uranium (Litz and Coleman, 1979). Wet chemical decontamination processes using inorganic or organic acids have also been developed as effective decontamination methods for metal wastes, and sludge-like uranium bearing wastes
Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
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(Ikeda et al., 2002). On the basis of these previous studies, we have proposed using a combination of sulfuric acid and hydrochloric acid to obtain the highest dissolution of scrap uranium, as they not only have the requisite dissolving power but are also environmentally acceptable. In uranium recovery processes for ore, uranium is generally recovered from the eluate by precipitation with either ammonia or hydrogen peroxide (H2O2). The method for recovering uranium as an ammonium diuranate precipitate is regarded as a cheap method. However, metal hydroxide impurities can co-precipitate with the uranium, as the reaction is carried out in a neutral pH range. In addition, ammonia is regarded as an environmental pollutant. On the other hand, it has been reported that high purity uranium can be recovered from solution as uranium peroxide using H2O2, as this process is carried out under acidic conditions. The precipitation rate, specific gravity, and dewaterability of the slurry are important factors affecting precipitate production, with uranium peroxide offering benefits in all these areas. Hydrogen peroxide hardly generates secondary waste, because it decomposes to water and oxygen. Thus, we selected the uranium peroxide method for the recovery of uranium from solution. The formula for the reaction between uranium and H2O2 can be written as: (Yan, 1990; Gurevich and Susorva, 1972; Mattus and Lopez, 1983).
solidification, remained as a blockish residue waste in the reaction bath. NaF adsorbent was used for removal of hydrogen fluoride from the off-gas generated during the demonstration testing, which contains significant quantities of uranium and fluorine. Unless fluorine is removed from the solution in which the NaF and UF4 residues described above are dissolved, a significant amount of fluorine may become mixed with the uranium precipitate. There is a variety of techniques available to remove fluorine from acidic solutions, for example, it has been reported that fluorine can be removed as Na2SiF6 (Kumar et al., 2010). The technique to separate uranium in a sulfuric acid (H2SO4) and hydrochloric acid (HCl) solution from other metal elements, specifically solvent extraction with amine, has also been investigated (Sato, 1966; Metwally et al., 2005; Lee et al., 1984). Amine extractant can be disposed more easily in comparison with organophosphate extractant, because it decomposes to nitrogen and oxygen. Consequently, it is expected that pure uranium can be recovered from solution when fluorine is removed by using one of these methods. Fig. 2 shows the proposed process flowsheet. In this study, we investigated the dissolution behaviors of NaF and UF4 residue in HCl and H2SO4, as well as the precipitation and extraction behaviors of fluorine from aqueous HCl and H2SO4 solutions.
þ UO2þ 2 þ H2 O2 þ xH2 O ! UO4 xH2 O þ 2H
2. Experimental
Fig. 1 shows the points in the uranium conversion process where spent NaF adsorbents, sludge-like wastes, and residue wastes are generated. UF4 was produced from a solution, in which yellow cake was dissolved as uranous ions in a reaction bath. A portion of UF4, which could not be transported because of
2.1. Materials The concentration of major elements in the UF4 residue, and spent NaF, were investigated. Samples were dissolved in concentrated nitric acid, with the concentrations of cations and anions
Fig. 1. The points in the uranium conversion process where adsorbents and sludge-like wastes are generated.
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NaF
UF4 residue
HClor H 2 SO 4 H2O2
Dissolution
SiO2 Na2SO 4
Removalof fluorine
Removalof fluorine
H2O2
TNOA kerosene 2-ethylhexanol
Recovery of uranium
Uranium peroxide
Fig. 2. The proposed aqueous process to recover uranium from solid uranium bearing waste.
2.2. Dissolution experiment
Table 1 Element content in the solid uranium bearing wastes. Content rate (wt%)
NaF UF4
U
F
Na
26.1 69.4
25.1 22.1
32.7 –
in solution measured by atomic absorption spectrophotometry (AAS) and ion chromatography (IC). The results of these analyses are shown in Table 1. Fig. 3 shows the appearance of the spent NaF and UF4 residue, their compositions being investigated using an X-ray diffractometer. In the XRD pattern, the presence of Na3UO2F5 was confirmed in spent NaF, and the presence of UF4/2.5H2O was confirmed in the UF4 residue.
2.2.1. Dissolution of UF4 residue Dissolution experiments were carried out in air while stirring. Blockish UF4 residue was weighed (2 g), then dissolved in H2SO4 solution at a molar ratio of SO4/U = 2–3 (0.3–0.45 M, 40 mL), or HCl at a molar ratio of Cl/U = 6, (1 M, 40 mL) while changing the temperature from ambient to 80 °C. 30 wt% H2O2 was added to both solutions at the molar ratio of H2O2/U = 1:3 in order to oxidize uranium, with the solution then stirred for 1.5–24 h. After dissolution, the residual solid phase was filtered off, and the uranium concentration in the filtrate was measured by AAS. Residual solid phase was completely dissolved by H2SO4, and its uranium concentration was subsequently measured. 2.2.2. Dissolution of spent NaF Once again, dissolution experiments were carried out in air while stirring. Spent NaF was dissolved in water with a variable
Fig. 3. The appearance of the (a) spent NaF and (b) UF4 residues.
Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
solid–liquid ratio at ambient temperature. After dissolution, the residual solid phase was filtered off, and the uranium concentration of the filtrates, and residual solid phase, was measured using AAS and inductively coupled plasma atomic emission spectroscopy (ICP-AES).
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Na2SiF6. The following reaction formulae show the mechanism for the removal of fluorine
SiO2 nH2 O þ 4HF ! SiF4 þ ð2 þ nÞH2 O SiF4 þ 2HF ! H2 SiF6 H2 SiF6 þ 2Na ! Na2 SiF6
2.2.3. Time dependency of dissolution ratio Blockish UF4 was weighed (30 g), and dissolved in H2SO4 (1.3 M, 150 mL) with a variable molar ratio between H2O2 and uranium (H2O2/U = 1–2) at 80 °C. It was also dissolved in HCl (3.8 M, 150 mL) with H2O2 (a molar ratio of H2O2/U = 2). Milled UF4 was also weighed (50 g), and dissolved in HCl (1.25 M, 225 mL) with H2O2 (a molar ratio of H2O2/U = 5). Time dependencies of the dissolution ratio of uranium for each experiment were then investigated. Spent NaF (8 g) was dissolved in water (20 mL), and once again, the time dependency of the dissolution ratio of uranium was investigated. 2.3. Removal of fluorine using solvent extraction It is known that uranium in H2SO4 solution is extracted by TNOA, and the reaction formula can be written as:
2R3 N þ 2Hþ þ UO2 ðSO4 Þ2 ! ðR3 NHÞ2 UO2 ðSO4 Þ2 2 R3 : n Octyl group Extracted uranium is converted into a uranyl chloride complex salt by HCl, and back-extracted into water to recover pure uranium.
ðR3 NHÞ2 UO2 ðSO4 Þ2 þ 4HCl ! ðR3 NHÞ2 UO2 Cl4 þ 2H2 SO4
2.4.1. Removal of fluorine from a solution prepared by dissolving UF4 The sample solution was prepared by dissolving UF4 residue (30 g) in H2SO4 (1.3 M, 150 mL) with H2O2 (H2O2/U = 2). SiO2 was added into the solution at a variable molar ratio between SiO2 and F of 0.2–0.5 at 80 °C. Na2SO4 and NaCl were added to the solution respectively at a molar ratio of Na/SiO2 = 2 in order to precipitate H2SiF6 as Na2SiF6 at ambient temperature. Another sample solution was prepared by dissolving UF4 residue in HCl (1.3–2.0 M, 150 mL), to which NaOH was added at a molar ratio of Na/SiO2 = 2, in order to precipitate H2SiF6. Ba(OH)2 was also used to precipitate H2SiF6 as BaSiF6. It is expected that Ba(OH)2 is more effective for the removal of fluorine, since the solubility of BaSiF6 is lower than that of Na2SiF6. 2.4.2. Removal of fluorine from the solution prepared by dissolving NaF Fluorine tends to exist as NaF in water, hence, it is necessary to dissociate fluorine by the addition of a strong acid in order to react it with SiO2. A sample solution was prepared by dissolving NaF in water with H2SO4 (0.4–1.3 M, 20 mL). Another sample solution was also prepared by dissolving NaF into HCl (0.9–1.5 M, 20 mL). SiO2 was added into the solution at a molar ratio of SiO2/F = 0.25, with all experiments performed at ambient temperature.
ðR3 NHÞ2 UO2 Cl4 ! 2R3 NHCl þ UO2 Cl2 Sample solutions were prepared by dissolving UF4 residue (30 g) in sulfuric acid (SO4/U = 4, 2.5 M, 150 mL) with H2O2 (H2O2/U = 2). TNOA with a concentration of 0.2 mol L1, and 12 wt% of 2-ethylhexanol, were added to 50 mL of kerosene giving a controlled TNOA concentration of 0.2 M. 10 mL of the sample aqueous solution was then mixed with 50 mL of the extraction solvent for 2 min. The extraction operation was repeated four times, each time adding neat TNOA. The concentration of uranium and fluorine in the aqueous solution after treatment was measured by AAS, ICP-AES and IC. Furthermore, a sample solution was prepared by dissolving NaF in water with H2SO4 (molar ratio of SO4/U = 4, 1.7 M, 20 mL). 10 mL of this sample solution was then mixed with 50 mL of the extraction solvent for two minutes. Once again, the extraction operation was repeated four times with the addition of neat TNOA. In addition, the UV–visible absorption spectrum of the solvent after extraction was measured to examine the chemical forms of uranium species present in the solvent. Both extraction solvents containing uranium were washed with 10 mL of 0.1 M H2SO4. Following this, fluorine remaining in the solvent was converted to chloride ions using 20 mL of 8 M HCl. Uranium in the solvents was back extracted with 15 mL of water, and the solvent was washed with 10 wt% sodium carbonate. The removal performance of fluorine and the loss of uranium in wastewater generated from both processes were also investigated. 2.4. Removal of fluorine using SiO2 It has been reported that fluorine can be removed as Na2SiF6 in acidic solution. Firstly, tetrafluorosilane (SiF4) is generated from the fluorination of silicon dioxide (SiO2) with hydrogen fluoride (HF), with SiF4 further fluorinated to H2SiF6 with HF. Following this, H2SiF6 is then reacted with sodium and precipitated as
2.5. Uranium recovery experiment 2.5.1. Uranium recovery from solution without removal of fluorine Sample solutions were prepared by dissolving NaF (8 g) in water (20 mL), and dissolving UF4 residue (30 g) in HCl (1.3 M, 150 mL), respectively. 35% H2O2 was added to both solutions at a molar ratio of H2O2/U = 4. The pH of the solutions was controlled from 1.1 to 4.0. The sample solutions were stirred for 30 min, and the reaction temperature was varied from 40 to 70 °C. The generated precipitate was washed with 300 mL of water, and then dried at 105 °C for 24 h. Fluorine content, and the sodium content of the uranium precipitate, were investigated and compared using the assumed conditions defined by the ASTM standard. 2.5.2. Uranium recovery from solution after removal of fluorine by solvent extraction A sample solution was prepared by dissolving NaF (800 g) in H2SO4 (1.75 M, 2000 mL). Another sample solution was prepared by dissolving UF4 residue (200 g) in H2SO4 (3 M, 800 mL). The extraction solvent was prepared by mixing 0.6 M TNOA with 24% 2-ethyl hexanol and kerosene. 150 mL of the sample solution was mixed with 500 ml of the extraction solvent for 2 min. Extraction solvent containing uranium was washed with 100 mL of 0.1 M H2SO4, with any fluorine remaining in the solvent being converted to chloride ions by the addition of 200 mL of 8 M HCl. Uranium in the solvent was back extracted with 200 mL of water. H2O2 was added into the back extracted solution to change the molar ratio between uranium and H2O2 (H2O2/U = 2–4). The reaction temperature for precipitation of uranium was controlled from 55 to 80 °C, and the pH of the solution was controlled from 0.8 to 1.7. Precipitation time was also controlled from 1.0 h to 2.0 h. The fluorine and sodium contents of the uranium precipitate were investigated and compared.
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2.5.3. Uranium recovery from solution after removal of fluorine by Na2SiF6 A sample solution was prepared by dissolving UF4 residue (320 g) in HCl (2.4 M, 800 mL) with H2O2 (H2O2/U = 2.5). Another solution was prepared by dissolving NaF (300 g) into HCl (1.5 M, 900 mL). SiO2 was added to the solutions at a molar ratio of SiO2/ F = 0.25. Precipitation time was controlled at 1 h. H2O2 was added into the solution after removal of fluorine at a molar ratio of H2O2/ U = 2.5, in order to recover uranium. Reaction temperature was controlled at 80 °C, and the pH of the solution was controlled to between 1.5 and 1.6. Sodium and fluorine content rate of the recovered uranium were investigated.
Table 3 Dissolution ratio of uranium for spent NaF. Solid/liquid ratio
Dissolution ratio of uranium (%)
U (g/ L)
Temperature (°C)
Time (h)
1/20 2/20 3/20 4/20 5/20 10/20 15/20
92 98.9 100 95.4 95.8 76.6 56.7
12 25.8 39.7 49.8 62.5 100 111
R.T R.T R.T R.T R.T R.T R.T
2 2 2 2 2 2 2
3. Results and discussion 3.1. Dissolution experiment 3.1.1. Dissolution of the UF4 residue and spent NaF Table 2 shows the dissolution ratio of uranium for UF4 residue vs. reaction temperature, molar ratio (H2O2/U) and reaction time. Dissolution ratio of uranium (DR) was defined as follows:
½DRð%Þ ¼ 100
½Usol ½Usol þ ½Ures
where [U]sol = total amount of uranium in solution; [U]res = total amount of uranium in the insoluble residue. Approximately 37% of uranium was dissolved in 0.45 M H2SO4, and approximately 59% of uranium was dissolved in 1 M HCl, at ambient temperature after 24 h. This result indicates that it is difficult to dissolve blockish UF4 residue at ambient temperature. The dissolution ratio increased as the molar ratio of H2O2/U increased, and also increased with increasing temperature. UF4 residue was completely dissolved in 0.3 M sulfuric acid with H2O2 (H2O2/ U = 2) at 80 °C, however, a precipitate was generated within 24 h after dissolution was finished. This indicates that excess H2O2 reacts with uranium in the solution, resulting in the generation of UO4. On the other hand, approximately 90% of uranium was dissolved completely in 1 M HCl with H2O2 (H2O2/U = 2) at 60 °C, with no precipitate being generated. This result indicates that hydrogen peroxide use-efficiency is low in HCl, preventing the precipitation of uranium. Table 3 shows the dissolution ratio of uranium for spent NaF vs. the solid/liquid ratio. Uranium in spent NaF was completely dissolved in water at a solid/liquid ratio of more than 3/20 at ambient temperature. 3.1.2. Time dependency for the dissolution ratio of uranium Fig. 4 shows the dissolution ratio of uranium in UF4 residue vs. dissolution time. UF4 was dissolved more rapidly in H2SO4 compared with HCl. The dissolution ratio of uranium increased as the
Fig. 4. Uranium concentration in solution vs. dissolution time for UF4 residue.
added amount of H2O2 was increased. 97% of the uranium in blockish UF4 was dissolved in 1.3 M H2SO4 with H2O2 (H2O2/U = 1) at 80 °C after 7 h. Milled UF4 residue dissolved in HCl more rapidly than blockish UF4. Uranium was completely dissolved in 1.25 M HCl with H2O2 (H2O2/U = 5.2) at 80 °C after 5 h. Fig. 5 shows the dissolution ratio of uranium in spent NaF vs. dissolution time. Uranium in the spent NaF was completely dissolved in water after 1 h. 3.2. Removal of fluorine by solvent extraction Table 4 shows the conditions and results of extraction of uranium using TNOA solvent extraction. 40% of fluorine in the aqueous solution was extracted with uranium into TNOA. It is known that HF and uranyl fluoride (UO2F2) were extracted by TNOA, and the distribution ratio of fluorine is barely affected by uranium concentration (Beranova et al., 1975).
Table 2 Dissolution ratio of uranium for UF4 residue. Solution
H2O2/U molar ratio
U (g/L)
Dissolution ratio (%)
Temperature (°C)
Time (h)
H2SO4 0.45 M H2SO4 0.45 M H2SO4 0.3 M H2SO4 0.3 M H2SO4 0.3 M H2SO4 0.3 M H2SO4 0.45 M H2SO4 0.45 M HCl 1 M HCl 1 M HCl 1 M HCl 1 M
1 3 1 2 1 2 1 2 1 1 2 1
12.8 12.7 23 33 34 34.7 33.2 35 20.3 22 31.2 23
36.9 36.6 66.3 95.1 98 100 95.7 100 58.5 63.4 89.9 66.3
R.T R.T 60 60 80 80 80 80 R.T 60 60 80
24 24 1.5 3 1.5 3 1.5 3 24 1.5 3 1.5
Fig. 5. Uranium concentration in the solution vs. dissolution time for spent NaF.
Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
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Table 4 Conditions and results of extraction of uranium using TNOA.
Solvent
Aqueous phase (mL) Solvent (mL) Stirring time (min) Initial uranium concentration of aqueous phase (g/L) Initial fluorine concentration of aqueous phase (g/L) Uranium concentration of aqueous phase after solvent extraction (g/L) Fluorine concentration of aqueous phase after solvent extraction (g/L) Uranium concentration of back extracted solution (g/L) Fluorine concentration of back extracted solution (g/L) Loss of uranium after washing by 0.1 M H2SO4 (%) Loss of uranium after conversion by HCl (%) DF (ratio of F/U in the neat solution)/ (ratio of F/U in the treated solution)
NaF
UF4
TOA 2-Ethyl hexanol Kerosene 10 50 2 92
TOA 2-Ethyl hexanol Kerosene 10 50 2 92
53.6
33.4
43.6
58.3
32
22
37.8
29.3
4.2
1.6
13.5 6.9 5.2
11.3 3.4 6.6
Fig. 6 shows the absorption spectrum of TNOA with uranium after extraction under these experimental conditions. The shape of the spectrum was different from UO2F2 or UO2 F2 4 , and the shape of spectrum for the solvent was similar to that of UO2SO4. This result indicates that uranium is mainly extracted as UO2SO4 (Tian and Rao, 2009). Fluorine extracted with uranium was separated by back extraction, with the decontamination factor of fluorine (DF) defined as follows:
½DF ¼
½Fi =½Ui ½Fb =½Ub
where [F]i = fluorine concentration in the initial solution; [U]i = uranium concentration in the initial solution; [F]b = fluorine concentration in the back extracted solution; [U]b = uranium concentration in the back extracted solution. The decontamination factor of fluorine for the NaF solution, and the UF4 solution, was 5.2 and 6.6 respectively. On the other hand, approximately 13% of the uranium in the NaF aqueous solution was lost during the process of washing using 0.1 M H2SO4, and a further 6.9% of uranium was lost during the conversion process
Fig. 6. The absorption spectrum of TNOA with uranium after extraction.
Fig. 7. Isotherms of uranium extraction for the solution prepared by dissolving UF4 residue and spent NaF.
using 8 M HCl. Approximately 11.3% of the uranium in the UF4 aqueous solution was also lost during the same washing process, with 3.4% of uranium lost during the same conversion process. Fig. 7 shows isotherms of uranium extraction for the solution prepared by dissolving UF4 and spent NaF. It is seen that the loading capacity of uranium into TNOA solvent is 9.7 g l1, which is lower than stoichiometric capacity and seems to be caused by extraction of hydrofluoric acid into TNOA. 3.3. Removal of fluorine using SiO2
3.3.1. Removal of fluorine from a solution prepared by dissolving UF4 residue Table 5 shows the removal ratio of fluorine from a solution in which UF4 residue was dissolved. The result indicates that additive amounts of SiO2 have little effect on the removal of fluorine at a SiO2/F molar ratio greater than 0.2. The result also indicates that there is little difference in fluorine concentration after removal between two solutions, where NaCl and Na2SO4 were respectively used as a source of Na+ ions. More than 92.9% of fluorine was removed by both sodium sources. Correspondingly, the loss of uranium from solutions where NaCl and Na2SO4 were dissolved was 23.9%, and 16.8%, respectively. The results also indicate that uranium peroxide, which is generated by hydrogen peroxide remaining in the solution, was coprecipitated with Na2SiF6. For the solution prepared by dissolving UF4 into HCl, 92% of fluorine in solution was removed when NaOH was added as a source of Na. Loss of uranium was below 6% under all conditions. It appears that more H2O2 is consumed in HCl solution, as compared to a H2SO4 solution, when UF4 is dissolved. On the other hand, 93% of fluorine was removed when Ba(OH)2 was used for precipitation of fluorine. This result indicates that there is little difference between NaOH and Ba(OH)2 for the removal of fluorine. The decontamination factors of fluorine by Na2SiF6 and BaSiF6 treatment are up to 14, which is higher than with solvent extraction by TNOA. 3.3.2. Removal of fluorine from a solution prepared by dissolving NaF Table 6 shows the results for removal of fluorine and loss of uranium from NaF solutions. The results indicate that the removal ratio became high as pH decreased. 97% of fluorine was removed at a pH of 1.3 from 1.3 M H2SO4 solution, and 90% of fluorine
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Table 5 Removal ratio of fluorine from UF4 residue by SiO2 treatment. Solution type
Initial concentration (g/l) U
F
H2SO4 1.3 M H2SO4 1.3 M H2SO4 1.3 M H2SO4 1.3 M HCl 1.3 M HCl 1.3 M HCl 2 M HCl 2 M
80.6 79.2 83.0 80.0 47.9 64.9 97.5 120
28 28 28 26 15 20.5 30.4 37.4
Molar Na (Ba) resource ratio SiO2/F
Time Temperature pH Concentration after treatment (g/l) (h) (°C) U F
DF Loss of Removal uranium (%) ratio of fluorine (%)
0.25 0.25 0.2 0.5 0.2 0.2 0.2 0.2
2 2 2 2 2 2 2 2
18.0 23.0 30.0 35.5 6.0 3.0 1.6 5.4
NaCl/SiO2 = 2 NaCl/SiO2 = 2 Na2SO4/SiO2 = 1 NaCl/SiO2 = 2 Ba(OH)2/SiO2 = 1 Ba(OH)2/SiO2 = 1 NaOH/SiO2 = 2 NaOH/SiO2 = 2
R.T R.T R.T R.T R.T R.T R.T R.T
1 66.1 1 61.0 1.3 58.1 1.3 51.6 0.5 45.0 0.6 63.0 1 81.5 0.8 113.5
2.0 1.7 1.4 1.6 1.2 1.4 2.8 3.1
92.9 93.9 95.0 93.8 92.0 93.2 90.8 91.7
11.5 12.7 14.0 10.5 11.8 14.2 9.1 11.4
Table 6 Removal ratio of fluorine from spent NaF by SiO2 treatment. Solution type
Initial concentration (g/l) U
F
H2SO4 0.4 M H2SO4 0.9 M H2SO4 1.3 M HCl/U 0.9 M HCl/U 1.5 M
93.1 88.9 88.9 88.9 88.9
53.6 59 59 59 59
Molar ratio SiO2/F
Time (h)
Temperature (°C)
pH
0.25 0.25 0.25 0.25 0.25
2 2 2 2 2
R.T R.T R.T R.T R.T
3.5 2.3 1.3 3.5 1.3
was removed at a pH of 1.3 from 1.5 M HCl solution. The loss of uranium was below 11% over the entire pH range investigated. There is little difference between HCl and H2SO4 solutions. The decontamination factor of fluorine was up to 28, which is also higher than the result of solvent extraction by TNOA.
3.4. Uranium recovery experiment 3.4.1. Uranium recovery from solution without removal of fluorine Table 7 shows the results for a uranium precipitation experiment without removal of fluorine. The recovery ratio of uranium from the NaF solution is low in comparison to the UF4 solution, a fact that indicates that fluorine concentration affects uranium precipitation. Uranium can form complexes with fluorine, and this is considered to inhibit the formation of uranium peroxide by decreasing the concentration of free uranyl ions. These results indicate that removal of fluorine is important in order to improve the efficiency of recovery of uranium. The fluorine content of uranium recovered from NaF solution is higher than that recovered from UF4 solution, indicating that fluorine concentration in solution increases the fluorine content of uranium precipitates. For the UF4 solution, the uranium content was at most 64.2%, and the fluorine content reached a minimum of 1.7%. For the NaF solution, the maximum uranium content was 62.5%, with a minimum fluorine content of 5.8%, which does not meet the assumed conditions defined by the ASTM standard (Standard Specification for Uranium Ore Concentrate. ASTM C967-02).
Concentration after treatment (g/l) U
F
92.0 78.9 84.4 84.4 80.9
19 3.8 2.0 20.5 6.0
Loss of uranium (%)
Removal ratio of fluorine (%)
DF
1.2 11.2 5.1 5.1 9.0
64.6 93.6 96.6 65.3 89.8
2.8 13.8 28.0 2.7 8.9
3.4.2. Uranium precipitation after solvent extraction and SiO2 treatment Table 8 shows the precipitation conditions and impurities in uranium recovered from NaF and UF4 solutions after solvent extraction. The uranium recovery ratio for UF4 and NaF was up to 91.5%. The fluorine content of uranium recovered from UF4 by solvent extraction was as low as 0.020 wt%, and for uranium recovered from NaF was as low as 0.030 wt%. The Na content of uranium from NaF was as low as 0.0015 wt%, with a uranium content of more than 65%. For the uranium recovered from both NaF and UF4 the concentration of uranium, and the content of impurities, met the necessary requirements. Table 9 shows the precipitation conditions and impurities in uranium recovered from the NaF and UF4 solution after SiO2 treatment. The uranium recovery ratio for UF4 and NaF was up to 94%. The fluorine content of uranium recovered from UF4 by SiO2 treatment was as low as 0.012 wt%, and for uranium recovered from NaF it was below 0.0075 wt%. The lowest Na content of uranium from NaF was below 0.01 wt%. The uranium content rate of uranium recovered from solution after SiO2 treatment is at least 70%, which is higher than that of uranium recovered by solvent extraction. The fluorine content of uranium recovered by SiO2 treatment is roughly the same as that of uranium recovered by solvent extraction. Based on these results, uranium that meets the target conditions can be recovered from UF4 residue and spent NaF by solvent extraction and SiO2 treatment. However, it appears that the SiO2 treatment process is preferable to recover pure uranium from spent NaF and UF4, as the
Table 7 Uranium recovery ratio and impurity content of uranium recovered from solution without removal of fluorine.
UF4 UF4 NaF NaF
Precipitation time (h)
pH
2 2 1 2
4.2 1.1 4.2 3.8
Temperature (°C)
Molar ratio (H2O2/U)
70 60 70 40
8 4 4 4
Initial concentration (g/l) U
F
Na
94.0 94.0 95.0 86.0
28.0 28.0 52.0 47.0
1.1 1.1 39 35
Uranium recovery ratio (%) 95.0 91.0 28.1 25.0
Impurities content of recovered uranium (%) U
F
Na
55.0 64.2 45.8 62.5
6.2 1.7 18 5.8
4.8 1.3 13 3.8
39
Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39 Table 8 Uranium recovery ratio and impurity content of uranium recovered from solution after removal of fluorine by solvent extraction.
UF4 UF4 UF4 NaF NaF
Precipitation time (h)
pH
Temperature (°C)
Molar ratio (H2O2/U)
2 2 2 1 1
1.3 0.8 0.8 1.3 1.7
70 55 55 87 80
2 2 2 4 2
Initial concentration (g/l) U
F
Na
51.0 51.0 51.0 43.0 46.0
0.9 0.9 0.9 6.4 11.3
4.1 4.1 4.1 2.0 2.2
E03 E03 E03 E02 E02
Uranium recovery ratio (%) 91.5 86.0 81.5 80.9 82.5
Impurities content of recovered uranium (%) U
F
65.0 66.6 66.6 66.6 66.6
5.0 2.0 1.3 8.3 3.2
Na E03 E02 E03 E03 E02
5.0 8.3 3.3 1.5 1.5
E04 E04 E04 E03 E03
Table 9 Uranium recovery ratio and impurity content of uranium recovered from solution after removal of fluorine by SiO2 treatment.
UF4 UF4 NaF NaF
Precipitation time (h)
pH
Temperature (°C)
Molar ratio (H2O2/U)
1 1 1 1
1.6 1.5 1.6 1.6
80 80 80 80
2.5 2.5 2.5 2.5
Initial concentration (g/l) U
F
Na
27.9 40.9 18.2 41.5
1.3 0.7 0.4 0.8
10.4 12.6 6.5 7.0
amount of waste water generation is smaller, and the uranium content of the recovered uranium is higher than for the solvent extraction process. Based on these results, we would expect that pure uranium could be recovered from solid waste containing uranium and fluorine. 4. Conclusion Two aqueous processes for the separation of fluorine from uranium in solid waste were investigated. Based on the results of the present investigation, the following conclusions have been drawn. (1) UF4 residue can be dissolved in 1.25 M hydrochloric acid with H2O2 (H2O2/U = 5) and 0.3 M H2SO4 with H2O2 (H2O2/ U = 2). Uranium in spent NaF can be completely dissolved in water at a solid to liquid ratio of more than 1/20. (2) The decontamination factor of fluorine for solutions prepared by dissolving NaF and UF4 residue in H2SO4 became 5.2 and 6.6, respectively, when a TNOA solvent extraction process is performed. (3) Compared with solvent extraction, for which the decontamination factor of fluorine was 10–28, SiO2 treatment is more effective in separating fluorine from a solution prepared by dissolving NaF and UF4 residue using H2SO4 and HCl. (4) Uranium recovered from solution without the removal of fluorine contains 1.7–18% of fluorine. On the other hand, the fluorine content of uranium recovered from UF4 by SiO2 treatment was as low as 0.012 wt%, and that of uranium recovered from NaF was below 0.0075%. The Na content of uranium from NaF was below 0.0019 wt%. These values meet the conditions defined by the ASTM.
Uranium recovery ratio (%) 78.4 88.1 89.5 85.4
Impurities content of recovered uranium (%) U
F
71.8 70.0 70.8 70.8
1.2 2.0 6.0 7.5
Na E02 E02 E03 E03
1.9 1.7 9.0 1.9
E03 E03 E04 E03
References Amamoto, I., 2000. Development of uranium conversion technology. Saikuru-KikoGiho 9, 65–74 (in Japanese). Amamoto, I., Wakabayashi, S., 1993. Uranium refining and conversion at Ningyo Toge works. Shigen-to-Sozai 109 (12), 1170–1174 (in Japanese). Beranova, H., Kuca, L., 1975. Effect of nitric acid and uranium on the extraction of hydrofluoric acid by primary amines. Collect. Czech. Chem. Commun. 40, 3608– 3615. Gurevich, A.M., Susorva, N.A., 1972. Interaction of sulfate complexes of uranyl with hydrogen peroxide. Radiokhimiya l14 (6), 831–836. Ikeda, A., Aida, M., Fujii, Y., Kataoka, S., Annnen, S., Sato, J., 2002. Ion exchange separation for decontamination of centrifuge enrichment plant. J. Nucl. Sci. Technol. 39, 1099–1105. Japan Atomic Energy Commission. Basic concept for treatment and disposal of uranium waste, Japan. Atomic Energy Commission Report, 2000. (in Japanese). Kumar, M., Babu, M., Mankhand, T.R., Babu, Pandey, B.D., 2010. Precipitation of sodium silicofluoride (Na2SiF6) and cryolite (Na3AlF6) from HF/HCl leach liquors of alumino-silicates. Hydrometallurgy 104, 304–307. Lee, Te-Wei, Cheng, Wo-Long, Ting, Gann, 1984. Solvent extraction study on the separation of molybdenum-99 and uranium in sulfuric acid solution by tri-noctylamine in kerosene. Solvent Extr. Ion Exch. 2, 435–550. Litz J.E., Coleman R.B., 1979. A review of United States yellow cake precipitation Practice. Production of Yellow Cake and Uranium Fluorides. Proc. Advisory Group Mtg, Paris. pp. 101–115. Mattus, A.J., Lopez, M.F., 1983. Hydrogen Peroxide Precipitation of Uranium Eliminates Silica Problems Associated with a Magnesia Process. The metallurgical society of AIME, TMS paper selection, A83-28. Metwally, E., Salah, Sh.A., EI-Nagar, H.A., 2005. Extraction and separation of uranium (VI) and thorium (IV) using tri-n-dodecylamine impregnated resins. J. Nucl. Radiochem. Sci. 6 (2), 119–126. Sato, T., 1966. The extraction of uranium (VI) from hydrochloric acid solutions by tri-N-octylamine. J. Inorg. Nucl. Chem. 28, 1461–1467. Standard Specification for Uranium Ore Concentrate. ASTM C967-02. Tian, G., Rao, L., 2009. Effect of temperature on the complexation of uranium (VI) with fluoride in aqueous solutions. J. Inorg. Chem. 48 (14), 6748–6754. Yan, T.Y., 1990. Uranium precipitation from eluate using hydrogen peroxide. Miner. Metall. Process., 222–225.
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Minerals Engineering journal homepage: www.elsevier.com/locate/mineng
Extraction of uranium from solid waste containing uranium and fluorine Yusuke Ohashi ⇑, Shinji Murashita, Mitsuo Nomura Ningyo-toge Environmental Engineering Center, Japan Atomic Energy Agency, 1550 Kamisaibara, Kagamino-cho, Tomata-gun, Okayama 708-0698, Japan
a r t i c l e
i n f o
Article history: Received 21 October 2013 Accepted 5 March 2014 Available online 27 March 2014 Keywords: Uranium Fluorine Silicofluoride Solvent extraction
a b s t r a c t A significant amount of waste residue and adsorbent containing uranium generated by uranium refining and conversion R&D activities has been stored at the Ningyo-toge Environmental Engineering Center of the Japan Atomic Energy Agency. The treatment of these materials is one of the most important tasks in the decommissioning of the plant. These solid wastes also contain significant quantities of fluorine, and it is important to separate this out in order to recover pure uranium as a resource. As one method for separating fluorine from uranium, we have proposed the use of solvent extraction using tri-n-octylamine (TNOA), and silicofluoride (Na2SiF6) precipitation. The decontamination factor of fluorine for solutions, prepared by dissolving spent sodium fluoride (NaF) and uranium tetrafluoride (UF4) residue into sulfuric acid, became 5.2 and 6.6 respectively when the TNOA solvent extraction process was performed. The Na2SiF6 process using SiO2 is more effective in separating fluorine from a solution prepared by dissolving spent NaF and UF4 residue using sulfuric and hydrochloric acid, with the decontamination factor of fluorine being 14 and 28. Uranium remaining in solution after removal of fluorine was recovered as uranium peroxide. The fluorine content of uranium recovered from UF4 by SiO2 treatment was as low as 0.012 wt%, and that of uranium recovered from NaF was below 0.0075%. This is in accordance with the target conditions defined by the ASTM standard. Ó 2014 Elsevier Ltd. All rights reserved.
1. Introduction Uranium deposits were first discovered in the Ningyo-toge area in 1955. A test milling facility constructed in 1965 was used to establish the fundamental technology of the PNC process, producing uranium tetrafluoride directly from uranium ore. A Refining and Conversion Pilot Plant using this process was constructed in 1981 (Amamoto and Wakabayashi, 1993). A total of 385 t of UF6 had been produced in the plant by 1991, when its operation was discontinued. A considerable amount of data on the design, construction, and operation of a commercial plant was obtained from this facility. However, as no promising overseas uranium deposits were discovered, the process has not been used for practical commercial purposes. Small-scale reprocessed-uranium conversion testing was carried out at the Refining and Conversion Plant from 1982, in order to develop fundamental technologies for converting reprocessed UO3 into UF6. Thereafter, middle-scale testing and practical application testing was carried out. During demonstration testing, which continued until 2001, approximately 293 t of UF6 was produced. Consequently, both the practical application and economic ⇑ Corresponding author. E-mail address: [email protected] (Y. Ohashi). http://dx.doi.org/10.1016/j.mineng.2014.03.003 0892-6875/Ó 2014 Elsevier Ltd. All rights reserved.
prospects for reprocessed-uranium conversion technology were successfully demonstrated (Amamoto, 2000). On the other hand, as a result of the aforementioned tests a significant amount of scrap uranium, and adsorbent containing uranium, was generated. The treatment of scrap uranium and adsorbent is one of the most pressing issues in the decommissioning of the plant. This waste totals approximately 1500 tons, and is presently stored in drums. The uranium concentration for some varieties of this waste exceeds 50 wt%. Hence, it is necessary to recover as much uranium as possible from the waste and scrap uranium, in order to be able to dispose of it more easily and to store the uranium as a resource (Japan Atomic Energy Commission, 2000).
1.1. Uranium recovery method The production of yellow cake from ores is an essential step in the preparation of fuel for use in nuclear power plants. Sulfuric acid has been used for dissolving uranium ore, and this same method is expected to be applied to the recovery of uranium from scrap uranium (Litz and Coleman, 1979). Wet chemical decontamination processes using inorganic or organic acids have also been developed as effective decontamination methods for metal wastes, and sludge-like uranium bearing wastes
Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
33
(Ikeda et al., 2002). On the basis of these previous studies, we have proposed using a combination of sulfuric acid and hydrochloric acid to obtain the highest dissolution of scrap uranium, as they not only have the requisite dissolving power but are also environmentally acceptable. In uranium recovery processes for ore, uranium is generally recovered from the eluate by precipitation with either ammonia or hydrogen peroxide (H2O2). The method for recovering uranium as an ammonium diuranate precipitate is regarded as a cheap method. However, metal hydroxide impurities can co-precipitate with the uranium, as the reaction is carried out in a neutral pH range. In addition, ammonia is regarded as an environmental pollutant. On the other hand, it has been reported that high purity uranium can be recovered from solution as uranium peroxide using H2O2, as this process is carried out under acidic conditions. The precipitation rate, specific gravity, and dewaterability of the slurry are important factors affecting precipitate production, with uranium peroxide offering benefits in all these areas. Hydrogen peroxide hardly generates secondary waste, because it decomposes to water and oxygen. Thus, we selected the uranium peroxide method for the recovery of uranium from solution. The formula for the reaction between uranium and H2O2 can be written as: (Yan, 1990; Gurevich and Susorva, 1972; Mattus and Lopez, 1983).
solidification, remained as a blockish residue waste in the reaction bath. NaF adsorbent was used for removal of hydrogen fluoride from the off-gas generated during the demonstration testing, which contains significant quantities of uranium and fluorine. Unless fluorine is removed from the solution in which the NaF and UF4 residues described above are dissolved, a significant amount of fluorine may become mixed with the uranium precipitate. There is a variety of techniques available to remove fluorine from acidic solutions, for example, it has been reported that fluorine can be removed as Na2SiF6 (Kumar et al., 2010). The technique to separate uranium in a sulfuric acid (H2SO4) and hydrochloric acid (HCl) solution from other metal elements, specifically solvent extraction with amine, has also been investigated (Sato, 1966; Metwally et al., 2005; Lee et al., 1984). Amine extractant can be disposed more easily in comparison with organophosphate extractant, because it decomposes to nitrogen and oxygen. Consequently, it is expected that pure uranium can be recovered from solution when fluorine is removed by using one of these methods. Fig. 2 shows the proposed process flowsheet. In this study, we investigated the dissolution behaviors of NaF and UF4 residue in HCl and H2SO4, as well as the precipitation and extraction behaviors of fluorine from aqueous HCl and H2SO4 solutions.
þ UO2þ 2 þ H2 O2 þ xH2 O ! UO4 xH2 O þ 2H
2. Experimental
Fig. 1 shows the points in the uranium conversion process where spent NaF adsorbents, sludge-like wastes, and residue wastes are generated. UF4 was produced from a solution, in which yellow cake was dissolved as uranous ions in a reaction bath. A portion of UF4, which could not be transported because of
2.1. Materials The concentration of major elements in the UF4 residue, and spent NaF, were investigated. Samples were dissolved in concentrated nitric acid, with the concentrations of cations and anions
Fig. 1. The points in the uranium conversion process where adsorbents and sludge-like wastes are generated.
34
Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
NaF
UF4 residue
HClor H 2 SO 4 H2O2
Dissolution
SiO2 Na2SO 4
Removalof fluorine
Removalof fluorine
H2O2
TNOA kerosene 2-ethylhexanol
Recovery of uranium
Uranium peroxide
Fig. 2. The proposed aqueous process to recover uranium from solid uranium bearing waste.
2.2. Dissolution experiment
Table 1 Element content in the solid uranium bearing wastes. Content rate (wt%)
NaF UF4
U
F
Na
26.1 69.4
25.1 22.1
32.7 –
in solution measured by atomic absorption spectrophotometry (AAS) and ion chromatography (IC). The results of these analyses are shown in Table 1. Fig. 3 shows the appearance of the spent NaF and UF4 residue, their compositions being investigated using an X-ray diffractometer. In the XRD pattern, the presence of Na3UO2F5 was confirmed in spent NaF, and the presence of UF4/2.5H2O was confirmed in the UF4 residue.
2.2.1. Dissolution of UF4 residue Dissolution experiments were carried out in air while stirring. Blockish UF4 residue was weighed (2 g), then dissolved in H2SO4 solution at a molar ratio of SO4/U = 2–3 (0.3–0.45 M, 40 mL), or HCl at a molar ratio of Cl/U = 6, (1 M, 40 mL) while changing the temperature from ambient to 80 °C. 30 wt% H2O2 was added to both solutions at the molar ratio of H2O2/U = 1:3 in order to oxidize uranium, with the solution then stirred for 1.5–24 h. After dissolution, the residual solid phase was filtered off, and the uranium concentration in the filtrate was measured by AAS. Residual solid phase was completely dissolved by H2SO4, and its uranium concentration was subsequently measured. 2.2.2. Dissolution of spent NaF Once again, dissolution experiments were carried out in air while stirring. Spent NaF was dissolved in water with a variable
Fig. 3. The appearance of the (a) spent NaF and (b) UF4 residues.
Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
solid–liquid ratio at ambient temperature. After dissolution, the residual solid phase was filtered off, and the uranium concentration of the filtrates, and residual solid phase, was measured using AAS and inductively coupled plasma atomic emission spectroscopy (ICP-AES).
35
Na2SiF6. The following reaction formulae show the mechanism for the removal of fluorine
SiO2 nH2 O þ 4HF ! SiF4 þ ð2 þ nÞH2 O SiF4 þ 2HF ! H2 SiF6 H2 SiF6 þ 2Na ! Na2 SiF6
2.2.3. Time dependency of dissolution ratio Blockish UF4 was weighed (30 g), and dissolved in H2SO4 (1.3 M, 150 mL) with a variable molar ratio between H2O2 and uranium (H2O2/U = 1–2) at 80 °C. It was also dissolved in HCl (3.8 M, 150 mL) with H2O2 (a molar ratio of H2O2/U = 2). Milled UF4 was also weighed (50 g), and dissolved in HCl (1.25 M, 225 mL) with H2O2 (a molar ratio of H2O2/U = 5). Time dependencies of the dissolution ratio of uranium for each experiment were then investigated. Spent NaF (8 g) was dissolved in water (20 mL), and once again, the time dependency of the dissolution ratio of uranium was investigated. 2.3. Removal of fluorine using solvent extraction It is known that uranium in H2SO4 solution is extracted by TNOA, and the reaction formula can be written as:
2R3 N þ 2Hþ þ UO2 ðSO4 Þ2 ! ðR3 NHÞ2 UO2 ðSO4 Þ2 2 R3 : n Octyl group Extracted uranium is converted into a uranyl chloride complex salt by HCl, and back-extracted into water to recover pure uranium.
ðR3 NHÞ2 UO2 ðSO4 Þ2 þ 4HCl ! ðR3 NHÞ2 UO2 Cl4 þ 2H2 SO4
2.4.1. Removal of fluorine from a solution prepared by dissolving UF4 The sample solution was prepared by dissolving UF4 residue (30 g) in H2SO4 (1.3 M, 150 mL) with H2O2 (H2O2/U = 2). SiO2 was added into the solution at a variable molar ratio between SiO2 and F of 0.2–0.5 at 80 °C. Na2SO4 and NaCl were added to the solution respectively at a molar ratio of Na/SiO2 = 2 in order to precipitate H2SiF6 as Na2SiF6 at ambient temperature. Another sample solution was prepared by dissolving UF4 residue in HCl (1.3–2.0 M, 150 mL), to which NaOH was added at a molar ratio of Na/SiO2 = 2, in order to precipitate H2SiF6. Ba(OH)2 was also used to precipitate H2SiF6 as BaSiF6. It is expected that Ba(OH)2 is more effective for the removal of fluorine, since the solubility of BaSiF6 is lower than that of Na2SiF6. 2.4.2. Removal of fluorine from the solution prepared by dissolving NaF Fluorine tends to exist as NaF in water, hence, it is necessary to dissociate fluorine by the addition of a strong acid in order to react it with SiO2. A sample solution was prepared by dissolving NaF in water with H2SO4 (0.4–1.3 M, 20 mL). Another sample solution was also prepared by dissolving NaF into HCl (0.9–1.5 M, 20 mL). SiO2 was added into the solution at a molar ratio of SiO2/F = 0.25, with all experiments performed at ambient temperature.
ðR3 NHÞ2 UO2 Cl4 ! 2R3 NHCl þ UO2 Cl2 Sample solutions were prepared by dissolving UF4 residue (30 g) in sulfuric acid (SO4/U = 4, 2.5 M, 150 mL) with H2O2 (H2O2/U = 2). TNOA with a concentration of 0.2 mol L1, and 12 wt% of 2-ethylhexanol, were added to 50 mL of kerosene giving a controlled TNOA concentration of 0.2 M. 10 mL of the sample aqueous solution was then mixed with 50 mL of the extraction solvent for 2 min. The extraction operation was repeated four times, each time adding neat TNOA. The concentration of uranium and fluorine in the aqueous solution after treatment was measured by AAS, ICP-AES and IC. Furthermore, a sample solution was prepared by dissolving NaF in water with H2SO4 (molar ratio of SO4/U = 4, 1.7 M, 20 mL). 10 mL of this sample solution was then mixed with 50 mL of the extraction solvent for two minutes. Once again, the extraction operation was repeated four times with the addition of neat TNOA. In addition, the UV–visible absorption spectrum of the solvent after extraction was measured to examine the chemical forms of uranium species present in the solvent. Both extraction solvents containing uranium were washed with 10 mL of 0.1 M H2SO4. Following this, fluorine remaining in the solvent was converted to chloride ions using 20 mL of 8 M HCl. Uranium in the solvents was back extracted with 15 mL of water, and the solvent was washed with 10 wt% sodium carbonate. The removal performance of fluorine and the loss of uranium in wastewater generated from both processes were also investigated. 2.4. Removal of fluorine using SiO2 It has been reported that fluorine can be removed as Na2SiF6 in acidic solution. Firstly, tetrafluorosilane (SiF4) is generated from the fluorination of silicon dioxide (SiO2) with hydrogen fluoride (HF), with SiF4 further fluorinated to H2SiF6 with HF. Following this, H2SiF6 is then reacted with sodium and precipitated as
2.5. Uranium recovery experiment 2.5.1. Uranium recovery from solution without removal of fluorine Sample solutions were prepared by dissolving NaF (8 g) in water (20 mL), and dissolving UF4 residue (30 g) in HCl (1.3 M, 150 mL), respectively. 35% H2O2 was added to both solutions at a molar ratio of H2O2/U = 4. The pH of the solutions was controlled from 1.1 to 4.0. The sample solutions were stirred for 30 min, and the reaction temperature was varied from 40 to 70 °C. The generated precipitate was washed with 300 mL of water, and then dried at 105 °C for 24 h. Fluorine content, and the sodium content of the uranium precipitate, were investigated and compared using the assumed conditions defined by the ASTM standard. 2.5.2. Uranium recovery from solution after removal of fluorine by solvent extraction A sample solution was prepared by dissolving NaF (800 g) in H2SO4 (1.75 M, 2000 mL). Another sample solution was prepared by dissolving UF4 residue (200 g) in H2SO4 (3 M, 800 mL). The extraction solvent was prepared by mixing 0.6 M TNOA with 24% 2-ethyl hexanol and kerosene. 150 mL of the sample solution was mixed with 500 ml of the extraction solvent for 2 min. Extraction solvent containing uranium was washed with 100 mL of 0.1 M H2SO4, with any fluorine remaining in the solvent being converted to chloride ions by the addition of 200 mL of 8 M HCl. Uranium in the solvent was back extracted with 200 mL of water. H2O2 was added into the back extracted solution to change the molar ratio between uranium and H2O2 (H2O2/U = 2–4). The reaction temperature for precipitation of uranium was controlled from 55 to 80 °C, and the pH of the solution was controlled from 0.8 to 1.7. Precipitation time was also controlled from 1.0 h to 2.0 h. The fluorine and sodium contents of the uranium precipitate were investigated and compared.
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Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
2.5.3. Uranium recovery from solution after removal of fluorine by Na2SiF6 A sample solution was prepared by dissolving UF4 residue (320 g) in HCl (2.4 M, 800 mL) with H2O2 (H2O2/U = 2.5). Another solution was prepared by dissolving NaF (300 g) into HCl (1.5 M, 900 mL). SiO2 was added to the solutions at a molar ratio of SiO2/ F = 0.25. Precipitation time was controlled at 1 h. H2O2 was added into the solution after removal of fluorine at a molar ratio of H2O2/ U = 2.5, in order to recover uranium. Reaction temperature was controlled at 80 °C, and the pH of the solution was controlled to between 1.5 and 1.6. Sodium and fluorine content rate of the recovered uranium were investigated.
Table 3 Dissolution ratio of uranium for spent NaF. Solid/liquid ratio
Dissolution ratio of uranium (%)
U (g/ L)
Temperature (°C)
Time (h)
1/20 2/20 3/20 4/20 5/20 10/20 15/20
92 98.9 100 95.4 95.8 76.6 56.7
12 25.8 39.7 49.8 62.5 100 111
R.T R.T R.T R.T R.T R.T R.T
2 2 2 2 2 2 2
3. Results and discussion 3.1. Dissolution experiment 3.1.1. Dissolution of the UF4 residue and spent NaF Table 2 shows the dissolution ratio of uranium for UF4 residue vs. reaction temperature, molar ratio (H2O2/U) and reaction time. Dissolution ratio of uranium (DR) was defined as follows:
½DRð%Þ ¼ 100
½Usol ½Usol þ ½Ures
where [U]sol = total amount of uranium in solution; [U]res = total amount of uranium in the insoluble residue. Approximately 37% of uranium was dissolved in 0.45 M H2SO4, and approximately 59% of uranium was dissolved in 1 M HCl, at ambient temperature after 24 h. This result indicates that it is difficult to dissolve blockish UF4 residue at ambient temperature. The dissolution ratio increased as the molar ratio of H2O2/U increased, and also increased with increasing temperature. UF4 residue was completely dissolved in 0.3 M sulfuric acid with H2O2 (H2O2/ U = 2) at 80 °C, however, a precipitate was generated within 24 h after dissolution was finished. This indicates that excess H2O2 reacts with uranium in the solution, resulting in the generation of UO4. On the other hand, approximately 90% of uranium was dissolved completely in 1 M HCl with H2O2 (H2O2/U = 2) at 60 °C, with no precipitate being generated. This result indicates that hydrogen peroxide use-efficiency is low in HCl, preventing the precipitation of uranium. Table 3 shows the dissolution ratio of uranium for spent NaF vs. the solid/liquid ratio. Uranium in spent NaF was completely dissolved in water at a solid/liquid ratio of more than 3/20 at ambient temperature. 3.1.2. Time dependency for the dissolution ratio of uranium Fig. 4 shows the dissolution ratio of uranium in UF4 residue vs. dissolution time. UF4 was dissolved more rapidly in H2SO4 compared with HCl. The dissolution ratio of uranium increased as the
Fig. 4. Uranium concentration in solution vs. dissolution time for UF4 residue.
added amount of H2O2 was increased. 97% of the uranium in blockish UF4 was dissolved in 1.3 M H2SO4 with H2O2 (H2O2/U = 1) at 80 °C after 7 h. Milled UF4 residue dissolved in HCl more rapidly than blockish UF4. Uranium was completely dissolved in 1.25 M HCl with H2O2 (H2O2/U = 5.2) at 80 °C after 5 h. Fig. 5 shows the dissolution ratio of uranium in spent NaF vs. dissolution time. Uranium in the spent NaF was completely dissolved in water after 1 h. 3.2. Removal of fluorine by solvent extraction Table 4 shows the conditions and results of extraction of uranium using TNOA solvent extraction. 40% of fluorine in the aqueous solution was extracted with uranium into TNOA. It is known that HF and uranyl fluoride (UO2F2) were extracted by TNOA, and the distribution ratio of fluorine is barely affected by uranium concentration (Beranova et al., 1975).
Table 2 Dissolution ratio of uranium for UF4 residue. Solution
H2O2/U molar ratio
U (g/L)
Dissolution ratio (%)
Temperature (°C)
Time (h)
H2SO4 0.45 M H2SO4 0.45 M H2SO4 0.3 M H2SO4 0.3 M H2SO4 0.3 M H2SO4 0.3 M H2SO4 0.45 M H2SO4 0.45 M HCl 1 M HCl 1 M HCl 1 M HCl 1 M
1 3 1 2 1 2 1 2 1 1 2 1
12.8 12.7 23 33 34 34.7 33.2 35 20.3 22 31.2 23
36.9 36.6 66.3 95.1 98 100 95.7 100 58.5 63.4 89.9 66.3
R.T R.T 60 60 80 80 80 80 R.T 60 60 80
24 24 1.5 3 1.5 3 1.5 3 24 1.5 3 1.5
Fig. 5. Uranium concentration in the solution vs. dissolution time for spent NaF.
Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
37
Table 4 Conditions and results of extraction of uranium using TNOA.
Solvent
Aqueous phase (mL) Solvent (mL) Stirring time (min) Initial uranium concentration of aqueous phase (g/L) Initial fluorine concentration of aqueous phase (g/L) Uranium concentration of aqueous phase after solvent extraction (g/L) Fluorine concentration of aqueous phase after solvent extraction (g/L) Uranium concentration of back extracted solution (g/L) Fluorine concentration of back extracted solution (g/L) Loss of uranium after washing by 0.1 M H2SO4 (%) Loss of uranium after conversion by HCl (%) DF (ratio of F/U in the neat solution)/ (ratio of F/U in the treated solution)
NaF
UF4
TOA 2-Ethyl hexanol Kerosene 10 50 2 92
TOA 2-Ethyl hexanol Kerosene 10 50 2 92
53.6
33.4
43.6
58.3
32
22
37.8
29.3
4.2
1.6
13.5 6.9 5.2
11.3 3.4 6.6
Fig. 6 shows the absorption spectrum of TNOA with uranium after extraction under these experimental conditions. The shape of the spectrum was different from UO2F2 or UO2 F2 4 , and the shape of spectrum for the solvent was similar to that of UO2SO4. This result indicates that uranium is mainly extracted as UO2SO4 (Tian and Rao, 2009). Fluorine extracted with uranium was separated by back extraction, with the decontamination factor of fluorine (DF) defined as follows:
½DF ¼
½Fi =½Ui ½Fb =½Ub
where [F]i = fluorine concentration in the initial solution; [U]i = uranium concentration in the initial solution; [F]b = fluorine concentration in the back extracted solution; [U]b = uranium concentration in the back extracted solution. The decontamination factor of fluorine for the NaF solution, and the UF4 solution, was 5.2 and 6.6 respectively. On the other hand, approximately 13% of the uranium in the NaF aqueous solution was lost during the process of washing using 0.1 M H2SO4, and a further 6.9% of uranium was lost during the conversion process
Fig. 6. The absorption spectrum of TNOA with uranium after extraction.
Fig. 7. Isotherms of uranium extraction for the solution prepared by dissolving UF4 residue and spent NaF.
using 8 M HCl. Approximately 11.3% of the uranium in the UF4 aqueous solution was also lost during the same washing process, with 3.4% of uranium lost during the same conversion process. Fig. 7 shows isotherms of uranium extraction for the solution prepared by dissolving UF4 and spent NaF. It is seen that the loading capacity of uranium into TNOA solvent is 9.7 g l1, which is lower than stoichiometric capacity and seems to be caused by extraction of hydrofluoric acid into TNOA. 3.3. Removal of fluorine using SiO2
3.3.1. Removal of fluorine from a solution prepared by dissolving UF4 residue Table 5 shows the removal ratio of fluorine from a solution in which UF4 residue was dissolved. The result indicates that additive amounts of SiO2 have little effect on the removal of fluorine at a SiO2/F molar ratio greater than 0.2. The result also indicates that there is little difference in fluorine concentration after removal between two solutions, where NaCl and Na2SO4 were respectively used as a source of Na+ ions. More than 92.9% of fluorine was removed by both sodium sources. Correspondingly, the loss of uranium from solutions where NaCl and Na2SO4 were dissolved was 23.9%, and 16.8%, respectively. The results also indicate that uranium peroxide, which is generated by hydrogen peroxide remaining in the solution, was coprecipitated with Na2SiF6. For the solution prepared by dissolving UF4 into HCl, 92% of fluorine in solution was removed when NaOH was added as a source of Na. Loss of uranium was below 6% under all conditions. It appears that more H2O2 is consumed in HCl solution, as compared to a H2SO4 solution, when UF4 is dissolved. On the other hand, 93% of fluorine was removed when Ba(OH)2 was used for precipitation of fluorine. This result indicates that there is little difference between NaOH and Ba(OH)2 for the removal of fluorine. The decontamination factors of fluorine by Na2SiF6 and BaSiF6 treatment are up to 14, which is higher than with solvent extraction by TNOA. 3.3.2. Removal of fluorine from a solution prepared by dissolving NaF Table 6 shows the results for removal of fluorine and loss of uranium from NaF solutions. The results indicate that the removal ratio became high as pH decreased. 97% of fluorine was removed at a pH of 1.3 from 1.3 M H2SO4 solution, and 90% of fluorine
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Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39
Table 5 Removal ratio of fluorine from UF4 residue by SiO2 treatment. Solution type
Initial concentration (g/l) U
F
H2SO4 1.3 M H2SO4 1.3 M H2SO4 1.3 M H2SO4 1.3 M HCl 1.3 M HCl 1.3 M HCl 2 M HCl 2 M
80.6 79.2 83.0 80.0 47.9 64.9 97.5 120
28 28 28 26 15 20.5 30.4 37.4
Molar Na (Ba) resource ratio SiO2/F
Time Temperature pH Concentration after treatment (g/l) (h) (°C) U F
DF Loss of Removal uranium (%) ratio of fluorine (%)
0.25 0.25 0.2 0.5 0.2 0.2 0.2 0.2
2 2 2 2 2 2 2 2
18.0 23.0 30.0 35.5 6.0 3.0 1.6 5.4
NaCl/SiO2 = 2 NaCl/SiO2 = 2 Na2SO4/SiO2 = 1 NaCl/SiO2 = 2 Ba(OH)2/SiO2 = 1 Ba(OH)2/SiO2 = 1 NaOH/SiO2 = 2 NaOH/SiO2 = 2
R.T R.T R.T R.T R.T R.T R.T R.T
1 66.1 1 61.0 1.3 58.1 1.3 51.6 0.5 45.0 0.6 63.0 1 81.5 0.8 113.5
2.0 1.7 1.4 1.6 1.2 1.4 2.8 3.1
92.9 93.9 95.0 93.8 92.0 93.2 90.8 91.7
11.5 12.7 14.0 10.5 11.8 14.2 9.1 11.4
Table 6 Removal ratio of fluorine from spent NaF by SiO2 treatment. Solution type
Initial concentration (g/l) U
F
H2SO4 0.4 M H2SO4 0.9 M H2SO4 1.3 M HCl/U 0.9 M HCl/U 1.5 M
93.1 88.9 88.9 88.9 88.9
53.6 59 59 59 59
Molar ratio SiO2/F
Time (h)
Temperature (°C)
pH
0.25 0.25 0.25 0.25 0.25
2 2 2 2 2
R.T R.T R.T R.T R.T
3.5 2.3 1.3 3.5 1.3
was removed at a pH of 1.3 from 1.5 M HCl solution. The loss of uranium was below 11% over the entire pH range investigated. There is little difference between HCl and H2SO4 solutions. The decontamination factor of fluorine was up to 28, which is also higher than the result of solvent extraction by TNOA.
3.4. Uranium recovery experiment 3.4.1. Uranium recovery from solution without removal of fluorine Table 7 shows the results for a uranium precipitation experiment without removal of fluorine. The recovery ratio of uranium from the NaF solution is low in comparison to the UF4 solution, a fact that indicates that fluorine concentration affects uranium precipitation. Uranium can form complexes with fluorine, and this is considered to inhibit the formation of uranium peroxide by decreasing the concentration of free uranyl ions. These results indicate that removal of fluorine is important in order to improve the efficiency of recovery of uranium. The fluorine content of uranium recovered from NaF solution is higher than that recovered from UF4 solution, indicating that fluorine concentration in solution increases the fluorine content of uranium precipitates. For the UF4 solution, the uranium content was at most 64.2%, and the fluorine content reached a minimum of 1.7%. For the NaF solution, the maximum uranium content was 62.5%, with a minimum fluorine content of 5.8%, which does not meet the assumed conditions defined by the ASTM standard (Standard Specification for Uranium Ore Concentrate. ASTM C967-02).
Concentration after treatment (g/l) U
F
92.0 78.9 84.4 84.4 80.9
19 3.8 2.0 20.5 6.0
Loss of uranium (%)
Removal ratio of fluorine (%)
DF
1.2 11.2 5.1 5.1 9.0
64.6 93.6 96.6 65.3 89.8
2.8 13.8 28.0 2.7 8.9
3.4.2. Uranium precipitation after solvent extraction and SiO2 treatment Table 8 shows the precipitation conditions and impurities in uranium recovered from NaF and UF4 solutions after solvent extraction. The uranium recovery ratio for UF4 and NaF was up to 91.5%. The fluorine content of uranium recovered from UF4 by solvent extraction was as low as 0.020 wt%, and for uranium recovered from NaF was as low as 0.030 wt%. The Na content of uranium from NaF was as low as 0.0015 wt%, with a uranium content of more than 65%. For the uranium recovered from both NaF and UF4 the concentration of uranium, and the content of impurities, met the necessary requirements. Table 9 shows the precipitation conditions and impurities in uranium recovered from the NaF and UF4 solution after SiO2 treatment. The uranium recovery ratio for UF4 and NaF was up to 94%. The fluorine content of uranium recovered from UF4 by SiO2 treatment was as low as 0.012 wt%, and for uranium recovered from NaF it was below 0.0075 wt%. The lowest Na content of uranium from NaF was below 0.01 wt%. The uranium content rate of uranium recovered from solution after SiO2 treatment is at least 70%, which is higher than that of uranium recovered by solvent extraction. The fluorine content of uranium recovered by SiO2 treatment is roughly the same as that of uranium recovered by solvent extraction. Based on these results, uranium that meets the target conditions can be recovered from UF4 residue and spent NaF by solvent extraction and SiO2 treatment. However, it appears that the SiO2 treatment process is preferable to recover pure uranium from spent NaF and UF4, as the
Table 7 Uranium recovery ratio and impurity content of uranium recovered from solution without removal of fluorine.
UF4 UF4 NaF NaF
Precipitation time (h)
pH
2 2 1 2
4.2 1.1 4.2 3.8
Temperature (°C)
Molar ratio (H2O2/U)
70 60 70 40
8 4 4 4
Initial concentration (g/l) U
F
Na
94.0 94.0 95.0 86.0
28.0 28.0 52.0 47.0
1.1 1.1 39 35
Uranium recovery ratio (%) 95.0 91.0 28.1 25.0
Impurities content of recovered uranium (%) U
F
Na
55.0 64.2 45.8 62.5
6.2 1.7 18 5.8
4.8 1.3 13 3.8
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Y. Ohashi et al. / Minerals Engineering 61 (2014) 32–39 Table 8 Uranium recovery ratio and impurity content of uranium recovered from solution after removal of fluorine by solvent extraction.
UF4 UF4 UF4 NaF NaF
Precipitation time (h)
pH
Temperature (°C)
Molar ratio (H2O2/U)
2 2 2 1 1
1.3 0.8 0.8 1.3 1.7
70 55 55 87 80
2 2 2 4 2
Initial concentration (g/l) U
F
Na
51.0 51.0 51.0 43.0 46.0
0.9 0.9 0.9 6.4 11.3
4.1 4.1 4.1 2.0 2.2
E03 E03 E03 E02 E02
Uranium recovery ratio (%) 91.5 86.0 81.5 80.9 82.5
Impurities content of recovered uranium (%) U
F
65.0 66.6 66.6 66.6 66.6
5.0 2.0 1.3 8.3 3.2
Na E03 E02 E03 E03 E02
5.0 8.3 3.3 1.5 1.5
E04 E04 E04 E03 E03
Table 9 Uranium recovery ratio and impurity content of uranium recovered from solution after removal of fluorine by SiO2 treatment.
UF4 UF4 NaF NaF
Precipitation time (h)
pH
Temperature (°C)
Molar ratio (H2O2/U)
1 1 1 1
1.6 1.5 1.6 1.6
80 80 80 80
2.5 2.5 2.5 2.5
Initial concentration (g/l) U
F
Na
27.9 40.9 18.2 41.5
1.3 0.7 0.4 0.8
10.4 12.6 6.5 7.0
amount of waste water generation is smaller, and the uranium content of the recovered uranium is higher than for the solvent extraction process. Based on these results, we would expect that pure uranium could be recovered from solid waste containing uranium and fluorine. 4. Conclusion Two aqueous processes for the separation of fluorine from uranium in solid waste were investigated. Based on the results of the present investigation, the following conclusions have been drawn. (1) UF4 residue can be dissolved in 1.25 M hydrochloric acid with H2O2 (H2O2/U = 5) and 0.3 M H2SO4 with H2O2 (H2O2/ U = 2). Uranium in spent NaF can be completely dissolved in water at a solid to liquid ratio of more than 1/20. (2) The decontamination factor of fluorine for solutions prepared by dissolving NaF and UF4 residue in H2SO4 became 5.2 and 6.6, respectively, when a TNOA solvent extraction process is performed. (3) Compared with solvent extraction, for which the decontamination factor of fluorine was 10–28, SiO2 treatment is more effective in separating fluorine from a solution prepared by dissolving NaF and UF4 residue using H2SO4 and HCl. (4) Uranium recovered from solution without the removal of fluorine contains 1.7–18% of fluorine. On the other hand, the fluorine content of uranium recovered from UF4 by SiO2 treatment was as low as 0.012 wt%, and that of uranium recovered from NaF was below 0.0075%. The Na content of uranium from NaF was below 0.0019 wt%. These values meet the conditions defined by the ASTM.
Uranium recovery ratio (%) 78.4 88.1 89.5 85.4
Impurities content of recovered uranium (%) U
F
71.8 70.0 70.8 70.8
1.2 2.0 6.0 7.5
Na E02 E02 E03 E03
1.9 1.7 9.0 1.9
E03 E03 E04 E03
References Amamoto, I., 2000. Development of uranium conversion technology. Saikuru-KikoGiho 9, 65–74 (in Japanese). Amamoto, I., Wakabayashi, S., 1993. Uranium refining and conversion at Ningyo Toge works. Shigen-to-Sozai 109 (12), 1170–1174 (in Japanese). Beranova, H., Kuca, L., 1975. Effect of nitric acid and uranium on the extraction of hydrofluoric acid by primary amines. Collect. Czech. Chem. Commun. 40, 3608– 3615. Gurevich, A.M., Susorva, N.A., 1972. Interaction of sulfate complexes of uranyl with hydrogen peroxide. Radiokhimiya l14 (6), 831–836. Ikeda, A., Aida, M., Fujii, Y., Kataoka, S., Annnen, S., Sato, J., 2002. Ion exchange separation for decontamination of centrifuge enrichment plant. J. Nucl. Sci. Technol. 39, 1099–1105. Japan Atomic Energy Commission. Basic concept for treatment and disposal of uranium waste, Japan. Atomic Energy Commission Report, 2000. (in Japanese). Kumar, M., Babu, M., Mankhand, T.R., Babu, Pandey, B.D., 2010. Precipitation of sodium silicofluoride (Na2SiF6) and cryolite (Na3AlF6) from HF/HCl leach liquors of alumino-silicates. Hydrometallurgy 104, 304–307. Lee, Te-Wei, Cheng, Wo-Long, Ting, Gann, 1984. Solvent extraction study on the separation of molybdenum-99 and uranium in sulfuric acid solution by tri-noctylamine in kerosene. Solvent Extr. Ion Exch. 2, 435–550. Litz J.E., Coleman R.B., 1979. A review of United States yellow cake precipitation Practice. Production of Yellow Cake and Uranium Fluorides. Proc. Advisory Group Mtg, Paris. pp. 101–115. Mattus, A.J., Lopez, M.F., 1983. Hydrogen Peroxide Precipitation of Uranium Eliminates Silica Problems Associated with a Magnesia Process. The metallurgical society of AIME, TMS paper selection, A83-28. Metwally, E., Salah, Sh.A., EI-Nagar, H.A., 2005. Extraction and separation of uranium (VI) and thorium (IV) using tri-n-dodecylamine impregnated resins. J. Nucl. Radiochem. Sci. 6 (2), 119–126. Sato, T., 1966. The extraction of uranium (VI) from hydrochloric acid solutions by tri-N-octylamine. J. Inorg. Nucl. Chem. 28, 1461–1467. Standard Specification for Uranium Ore Concentrate. ASTM C967-02. Tian, G., Rao, L., 2009. Effect of temperature on the complexation of uranium (VI) with fluoride in aqueous solutions. J. Inorg. Chem. 48 (14), 6748–6754. Yan, T.Y., 1990. Uranium precipitation from eluate using hydrogen peroxide. Miner. Metall. Process., 222–225.