Application of hollow fiber supported liquid membrane for the separation of americium from the analytical waste

Americium from analytical solid waste containing U and metallic impurities was separated using hollow fiber supported liquid membrane (HFSLM) technique impregnated with DHOA–TODGA from nitric acid medium. An aliquot of 5 g of the solid waste containing Am (19.95 mg) as minor actinide and of U (2,588 mg), Fe (1,360 mg), Ca (1,810 mg) and Na (3,130 mg) as major impurities was processed. The feed solution obtained after the dissolution of the residue in ~4 M HNO₃ was passed through HFSLM module. In the first stage using 1 M DHOA–dodecane U was recovered while Am and other impurities were left in the raffinate. In the second stage, 0.5 M DHOA + 0.1 M TODGA/dodecane was used for the separation of Am from other impurities. Though, majority of the elements were separated in this cycle, Ca was co extracted along with the americium. CMPO extraction chromatographic technique was used for further separation of americium from Ca. Significant decontamination factors were achieved in this three step separation process with respect to U, Fe, Na and Ca with ~77 % recovery of americium.     

Extraction studies of uranium into a third-phase of thorium nitrate employing tributyl phosphate and <Emphasis Type="Italic">N,N</Emphasis>-dihexyl octanamide as extractants in different diluents

Extraction behavior of 1 × 10⁻²–0.1 M U(VI) from aqueous phases containing 0.86 M Th(IV) at 4 M HNO₃ in 1.1 M tributyl phosphate (TBP) and 1.1 M N,N-dihexyl octanamide (DHOA) solutions in different diluents viz. n-dodecane, 10% 1-octanol + n-dodecane, and decahydronaphthalene (decalin) was studied. Third-phase formation was observed in both the extractants using n-dodecane as diluent. There was a gradual decrease in Th(IV) concentration in the third-phase (heavy organic phase, HOP) with increased aqueous U(VI) concentration [0.71 M (no U(VI))–0.61 M (0.1 M U(VI)) for TBP; 0.27 M (no U(VI))–0.22 M (0.1 M U(VI)) for DHOA]. The HOP volume in case of DHOA was ~2.2 times of that of TBP. Uranium concentration in HOP increased with its initial concentration in the aqueous phase [from 1.8 × 10⁻² M (0.01 M U(VI))–0.162 M (0.1 M U(VI)) for TBP; from 1.4 × 10⁻² M (0.01 M U(VI))–0.14 M (0.1 M U(VI)) for DHOA] suggesting that Th(IV) was being replaced by U(VI). An empirical correlation was developed for predicting the concentrations of uranium and thorium in HOP for both the extractants. No third-phase appeared during the extraction of uranium and thorium from the aqueous phases employing 10% 1-octanol + n-dodecane, or decalin as diluents, and therefore, were better choices as diluent for alleviating the third-phase formation during the reprocessing of spent thorium based fuels, and for the recovery of thorium from high-level waste solutions.     

色谱分离ICP-AES法测定高纯度八氧化三铀中的13种微量杂质

采用磷酸三丁酯（TBP）萃淋树脂色层分离铀，用电感耦合等离子体-原子发射光谱法测定分离后的离纯度铀氧化物的杂质元素Al、Ca、Cr、Cu、Fu、Mn、Mo、Ni、P、Ti、V、Zn、Zr，除Al、Fe、Mo外的其余10种元素的测定结果符合标准物质定值的要求。     

Extraction of uranium(VI) from nitric acid medium by 1.1M tri-n-butylphosphate in ionic liquid diluent

Summary A systematic study on the extraction of U(VI) from nitric acid medium by tri-n-butylphosphate (TBP) dissolved in a non-traditional diluent namely 1-butyl-3-methylimidazolium hexafluorophosphate (bmimPF₆) ionic liquid (IL) is reported. The results are compared with those obtained using TBP/n-dodecane (DD). The distribution ratio for the extraction of U(VI) from nitric acid by 1.1M TBP/bmimPF₆ increases with increasing nitric acid concentration. The U(VI) distribution ratios are comparable in the nitric acid concentration range of 0.01M to 4M, to the ratios measured using 1.1M TBP/DD. In contrast to the extraction behavior of TBP/DD, the D values continued to increase with the increase in the concentration of nitric acid above 4.0M. The stoichiometry of uranyl solvate extracted by 1.1M TBP/IL is similar to that of TBP/DD system, wherein two molecules of TBP are associated with one molecule of uranyl nitrate in the organic phase. Ionic liquid alone also extracts uranium from nitric acid, albeit to a small extent. The exothermic enthalpy accompanying the extraction of U(VI) in TBP/bmimPF₆ decreases with increasing nitric acid and with TBP concentrations.     

Precipitation and purification of uranium from rock phosphate

This study was carried-out to leach uranium from rock phosphate using sulphuric acid in the presence of potassium chlorate as an oxidant and to investigate the relative purity of different forms of yellow cakes produced with ammonia, magnesia and sodium hydroxide as precipitants, as well as purification of the products with TBP and matching its impurity levels with specifications of the commercial products. Alpha-particle spectrometry was used for determination of activity concentration of uranium isotopes in rock phosphate, resulting phosphoric acid, and in different forms of the yellow cake. Likewise, atomic absorption spectroscopy was used for determination of impurities. On the average, the equivalent mass concentration of uranium was 119.38 ± 79.66 ppm (rock phosphate) and 57.85 ± 20.46 ppm (phosphoric acid) with corresponding low percent of dissolution (48 %) which is considered low. The isotopic ratio (²³⁴U:²³⁸U) in all stages of hydrometallurgical process was not much different from unity indicating lack of fractionation. Upon comparing the levels of impurities in different form of crude yellow cakes, it was found that the lowest levels were measured in hydrated trioxide (UO₃·xH₂O). This implies that saturated magnesia is least aggressive relative to other precipitants and gives relatively pure crude cake. Therefore, it was used as an index to judge the relative purity of other forms of yellow cakes by taking the respective elemental ratios. The levels of impurities (Fe, Zn, Mn, Cu, Ni, Cd and Pb) in the purified yellow cake were found comparable with those specified for commercial products.     

Determination of trace impurities in uranium, thorium and plutonium matrices by solvent extraction and inductively coupled plasma atomic emission spectrometry

Studies on the determination of trace metallic impurities in nuclear materials such as uranium, thorium and plutonium are described. The bulk of the matrix is separated by batch extraction from their nitric acid solutions using 2-ethylhexyl hydrogen 2-ethylhexyl phosphonate (KSM-17, equivalent to PC88-A). The final aqueous phase containing the metallic impurities is fed to a high-temperature source inductively coupled plasma and the analysis is carried out employing a computer-controlled multichannel direct-reading spectrometer. The studies also included the recovery of impurities at various acidities and spectral interferences of the above matrices over the analyte elements. Based on the above studies, methods were standardized for the determination of 19 elements, viz. Al, B, Be, Ca, Cd, Cr, Cu, Fe, Mg, Mn, Ni, Pb, Si, Zn, Ce, Dy, Eu, Gd and Sm, in U/Th/Pu solutions. The relative standard deviation for various elements is in the range 1-5%.     

Selective ion exchange separation of uranium from concomitant impurities in uranium materials and subsequent determination of the impurities by ICP-OES

An ion exchange method has been developed for the separation of uranium from trace level metallic impurities prior to their determination by inductively coupled plasma optical emission spectrometry (ICP-OES) in uranium materials. Selective separation of uranium from trace level metallic impurities consisting Cr, Co, Cu, Fe, Mn, Cd, Gd, Dy, Ni, and Ca was achieved on anion exchange resin Dowex 1 × 8 in sulphate medium. The resin (100–200 mesh, in chloride form) was packed in a small Teflon column (7.8 cm × 0.8 cm I.D.) and brought into sulphate form by passing 0.2 N ammonium sulphate solution. Optimum experimental conditions including pH and concentration of sulphate in the liquid phase were investigated for the effective uptake of uranium by the column. Uranium was selectively retained on the column as anionic complex with sulphate, while impurities were passed through the column. Post column solution was collected and analyzed by ICP-OES for the determination of metallic impurities. Up to 2,500 μg/mL of uranium was retained with >99% efficiency after passing 25 mL sample through the column at pH 3. Percentage recoveries obtained for most of the metallic impurities were >95% with relative standard deviations <5%. The method established was applied for the determination of gadolinium in urania–gadolinia (UO₂–Gd₂O₃) ceramic nuclear fuel and excellent results were achieved. Solvent extraction method using tributylphosphate (TBP) as extractant was also applied for the separation of uranium in urania–gadolinia nuclear fuel samples prior to the determination of gadolinium by ICP-OES. The results obtained with the present method were found very comparable with those of the solvent extraction method.     

Purification of americium from assorted analytical waste in hydrochloric acid medium

During the simultaneous extraction of plutonium and uranium using anion exchange chromatographic technique from analytical waste in hydrochloric acid medium, ²⁴¹Am which is invariably present in the plutonium bearing fuel samples remains in the effluent. A two step separation scheme was developed for the recovery and purification of Am from the assorted waste to facilitate the disposal of large volume of aqueous waste and the purified Am solution was utilized for spectroscopic investigation. The separation scheme involved solvent extraction using 0.1 M TODGA + 0.5 M DHOA for separation of americium from Fe, Pb, Ni and Na followed by extraction chromatographic technique using CMPO on inert support as stationary phase for separation of Ca from Am. A systematic study on the extraction behavior of Am from hydrochloric acid medium revealed that out of four extraction systems well known for actinide partitioning namely 0.1 M TODGA + 0.5 M DHOA, 1 M DMDBTDMA, 0.2 M CMPO + 1.2 M TBP and 30% TRPO, only 0.1 M TODGA + 0.5 M DHOA extracts americium from 7.5 M HCl feed acidity. A comparative study involving CMPO solvent extraction and column chromatographic technique revealed that elution of Am from column is satisfactory as compared to inefficient stripping of Am from organic phase in solvent extraction technique using 0.1 M HNO₃. The purity of the final solution was checked for 17 elements of interest and was found to be 98% pure, while the overall recovery of this two step separation scheme was found to be 95%.     

Direct determination of U(VI) in TBP-kerosene extracts using first derivative ultraviolet spectroscopy

Summary The extractive first order derivative spectrophotometry is a selective method for the separation and determination of U(VI) using tri-n-butyl phosphate (TBP), which combines the roles of solvent and complexing agent. The complex is formed by extracting U(VI) from an aqueous 6M sodium nitrate solution at initial pH 3.0 into a 25% solution of TBP in kerosene. This extraction also separates U(VI) from many diverse ions that interfere. After extraction, the determination of uranium shows good accuracy and precision with relative standard deviation of 1.5% (n = 5) at 20 ppm using zero-order spectrum at l_max = 250 nm. Calibration curve was also found to obey Beer's law in the range of 10-100 ppm with 3.33 ppm detection limit. However, these accuracy and precision have been improved to give relative standard deviation of 0.7% (n = 5) at 20 ppm with a lower detection limit of 2.24 ppm using the first-derivative spectrum at l = 263 nm comparing to the normal one.     

Determination of analytes at trace level in uranium matrix by ICP-AES without chemical/physical separation

Determination of trace metallic constituents in nuclear materials e.g. U, Pu, Am, Zr etc. by Atomic Emission Spectroscopy requires the separation of the major matrix without the loss of analytes at trace level. For DC Arc carrier distillation technique, carrier is used to separate the matrix physically according to the volatility of the analytes while appropriate extractant in suitable diluent is used for chemical separation in inductively coupled plasma atomic emission spectroscopy (ICP-AES). In the present study an attempt was made to develop a methodology for the determination of B, Cd, Mg, Zn, Al, Sr and Sc at trace level (up to 0.1 μg/mL) in uranium matrix without any chemical or physical separation. It involves identification of suitable analytical lines of uranium for its ICP-AES determination; study the spectral interference of uranium to choose interference free analytical lines, optimization of instrumental and experimental parameters etc. The method was validated using synthetic samples.     

Comparative studies on co-extraction of uranium(VI) and different mineral acid from aqueous feed solutions using TBP,TOPO and TOA

Equilibrium and kinetics of co-extraction of hexavalent uranium and mineral acids from aqueous solutions into a hydrocarbon phase (paraffin) using tri n-butyl phosphate (TBP), tri-n-octyl phosphine oxide (TOPO) and tri-n-octyl amine (TOA) has been studied. Relative rates of extraction of uranium(VI) and mineral acid by different complexing ligands were measured simultaneously using bulk-liquid membrane system. Acid extraction by complexing ligands was found to be significant. Wherever there was a possibility of the formation of the third phase, isodecanol was used as an organic phase modifier. Study revealed that isodecanol promotes acid extraction and substantially reduces distribution coefficient of U(VI) into the hydrocarbon phase. The rate of acid extraction by different ligand was in the order of TOPO > TOA > TBP–isodecanol > TBP, whereas the rate of extraction of uranium(VI) was in the order TOPO > TOA > TBP > TBP–isodecanol. A kinetic model was developed to predict concentration of acid and U(VI) in the feed, organic and the strip phase during extraction. The mass transfer coefficients for acid and metal were determined by fitting the model to the observed concentration–time data.     

磷酸三丁酯萃淋树脂色层分离电感耦合等离子体质谱法测定富稀土样中微量铀和钍

研究了磷酸三丁酯萃淋树脂色层分离,电感耦合等离子体质谱法测定富稀土样中微量铀、钍的方法。样品经消解后,以磷酸三丁酯萃淋树脂为固定相、8 mol/L硝酸为流动相过柱分离,样品中的大部分稀土元素随流动相流出,而铀和钍则被固定相吸附,用去离子水洗脱后,再用电感耦合等离子体质谱仪测定。铀、钍的检出限分别为0.06,0.16μg/L,测定结果的相对标准偏差均小于10%(n=5),加标回收率为98%~105%。对稀土矿石标准物质进行测定,测定值与推荐值相符。该法操作简便,测定结果可靠,适于富稀土样中微量铀、钍的测定。     

Separation and inductively coupled plasma optical emission spectrometric (ICP-OES) determination of trace impurities in nuclear grade uranium oxide

A simple and rapid inductively coupled plasma optical emission spectrometric method for the determination of trace level impurities like REEs, Y, Cd, Co, V, Mg, B, Ca, Cr, Mn, Ni, Cu, Zn and Al in uranium oxide samples is described. The method involves solvent extraction separation of uranium from 6 M HNO₃ acid medium using di (2-ethyl hexyl) phosphoric acid in toluene, which selectively separates uranium leaving behind the trace impurities in the aqueous media, for quantification by ICP-OES. The method has been applied to few synthetic samples and five certified reference U₃O₈ standards. The results are compared with other methods such as TBP-TOPO-CCl₄ and 1,2 diaminocyclohexane N,N,N′,N′-tetra acetic acid (CyDTA)–ammonium hydroxide (NH₄OH) separation techniques. Different experimental parameters like contact time, acidity, aqueous to organic ratio etc., are optimized for better and accurate results. The method is simple, rapid, accurate and precise for all the studied elements, showing a relative standard deviation of 1.5–12.0% at trace levels studied (5.5–12% at 0.2 μg/mL and 1.5–6.0% at 0.5 μg/mL), on the synthetic samples prepared from high purity oxides.     

Purification and recovery of plutonium from uranium oxide obtained in the fast reactor fuel reprocessing plant using hydroxyurea as reductant

Spent fuel discharged from Fast Breeder Test Reactor (FBTR) in Kalpakkam is being reprocessed by modified plutonium uranium reduction extraction (PUREX) process using 30% TBP (tributylphosphate) as extractant in the presence of heavy normal paraffin (HNP) as diluent. Partitioning of uranium (U) and plutonium (Pu) is carried out using oxalate precipitation method. Uranium oxide product obtained by this method contains appreciable amount of plutonium which has to be recovered. Recovery of plutonium from this uranium oxide product is carried out by reducing Pu to inextractable Pu(III) using hydroxyurea (HU) and then uranium is extracted into 30% TBP. A small amount of Pu which is extracted in the organic phase is stripped back to aqueous phase by scrubbing with scrubbing agent containing 0.1 M HU in 4 M nitric acid. Similarly U and Pu are co-extracted into 30% TBP and then Pu is removed by scrubbing with 0.1 M HU in 4 M nitric acid. Further decontamination from Pu is obtained in the stripping stages. By this method Pu contamination in the uranium oxide is brought from 7300 ppm to 0.4–3 ppm (wt/wt). This uranium product obtained can be handled on table top.     

Determination of Ultra-Trace Amouris of Uranium by Icp-Aes Technique

Abstract

An Atomic Emission Spectrometric method based on the use of Inductively coupled argon Plasma source has been developed for the determination of ultra trace concentrations of uranium in aqueous solutions. Using the optimised experimental conditions for the ICP source, uranium can be determined at 0.05 μg/ml concentration in 0.3M HNO₃ solutions with either of the two analytical lines viz. 3859.6 Å or 4090.1 Å. The precision of determinations at the lowest detection limit is better than 2% R.S.D. The uranium estimation has been corrected for the interference due to the presence of eighteen metallic elements using an inter-element correction procedure. A number of spiked samples and NBL reference samples with concomitant impurities have been analysed using the standardised procedure and good agreement has been observed with their certified values.     

A novel rapid detection approach for the analysis of radionuclides in environmental samples using graphite MALDI mass spectrometry

In nuclear technology, tri-n-butyl phosphate (TBP) diluted with a hydrocarbon diluent such as n-dodecane or NPH is the most frequently used solvent in liquid–liquid extraction for fuel reprocessing. This extraction, known as the plutonium uranium refining by extraction, is still considered as the most dominant process for the extraction of uranium and plutonium from irradiated fuels. The solubility of pure TBP in water is about 0.4 g/L at 25 °C. This is enough to create trouble during evaporation of raffinate and product solutions. Solubility data for undiluted TBP and TBP (diluted in inert hydrocarbon diluent) in various concentrations of nitric acid is not adequate in the literature. The solubility data generated in the present study provide complete information on the solubility of TBP in various nitric acid concentrations (0–15.7 M) at room temperature. The effect of heavy metal ion concentration such as uranium and various fission products on the solubility of TBP in nitric acid is also presented. The results obtained from gas chromatographic technique were compared with spectrophotometric technique by converting the organic phosphate into inorganic phosphate. The generated data is of direct relevance to reprocessing applications.     

Comparative extraction efficiencies of tri-<Emphasis Type="Italic">n</Emphasis>-butyl phosphate and <Emphasis Type="Italic">N,N</Emphasis>-dihexyloctanamide for uranium recovery using supercritical CO<Subscript>2</Subscript>

Extraction of uranium from tissue paper, synthetic soil, and from its oxides (UO₂, UO₃ and U₃O₈) was carried out using supercritical carbon dioxide modified with methanol solutions of extractants such as tri-n-butyl phosphate (TBP) or N,N-dihexyl octanamide (DHOA). The effects of temperature, pressure, extractant/nitric acid (nitrate) concentration, and of hydrogen peroxide on uranium extraction were investigated. The dissolution and extraction of uranium in supercritical CO₂ modified with TBP, from oxide samples followed the order: UO₃ ≫ UO₂ > U₃O₈. Addition of hydrogen peroxide in the modifier solution enhanced the dissolution/extraction of uranium in dynamic mode. DHOA appeared better than TBP for recovery of uranium from different oxide samples. Similar enhancement in uranium extraction was observed in static mode experiments in the presence of hydrogen peroxide. Uranium estimation in the extracted fraction was carried out by spectrophotometry employing 2-(5-bromo-2-pyridylazo)-5-diethylaminophenol (Br-PADAP) as the chromophore.     

Preconcentration and determination of ultra-trace amounts of U(VI) and Th(IV) using titan yellow-impregnated Amberlite XAD-7 resin

A simple and sensitive method for the determination of ultra trace amounts of U(VI) and Th(IV) ions by spectrophotometric method after solid-phase extraction on a new extractant-impregnated resin (EIR) has been reported. The new EIR was synthesised by impregnating a weakly polar polymeric adsorbent, Amberlite XAD-7, with titan yellow (TY) as extractant. The analytical method is based on the simultaneous adsorption of analyte ions in a mini-column packed with TY/XAD-7 and performing sequential elution with 0.5% (w/v) Na₂CO₃ for uranium and 2.0 M HCl for thorium. The influences of the analytical parameters including pH, salting out agent and sample volume were investigated. The interference effects of foreign ions on the retention of the analyte ions were also explored. The limits of detection for U(VI) and Th(IV) were as low as 50 and 25 ng L^?1, respectively. Relative standard deviations (n = 7) for U(VI) and Th(IV) were 3.1% and 2.9%, respectively. The method was successfully applied to the determination of ultra trace amounts of U(VI) and Th(IV) in different real matrices including industrial wastewater samples and environmental waters. The proposed method was validated using three certified reference materials and the results were in good agreement with the certified values.     

Extractive separation of Th(IV), U(VI), Ti(IV), La(III) and Fe(III) from Zarigan ore

In this study, the effects of various extraction parameters such as extractant types (Cyanex302, Cyanex272, TBP), acid type (nitric, sulfuric, hydrochloric) and their concentrations were studied on the thorium separation efficiency from uranium(VI), titanium(IV), lanthanum(III), iron(III) using Taguchi^??s method. Results showed that, all these variables had significant effects on the selective thorium separation. The optimum separations of thorium from uranium, titanium and iron were achieved by Cyanex302. The aqueous solutions of 0.01 and 1 M nitric acid were found as the best aqueous conditions for separating of thorium from titanium (or iron) and uranium, respectively. The combination of 0.01 M nitric acid and Cyanex272 were found that to be the optimum conditions for the selective separation of thorium from lanthanum. The results also showed that TBP could selectively extract all studied elements into organic phase leaving thorium behind in the aqueous phase. Detailed experiments showed that 0.5 M HNO₃ is the optimum acid concentration for separating of thorium from other elements with acidic extractants such as Cyanex272 and Cyanex302. The two-stage process containing TBP-Cyanex302 was proposed for separation thorium and uranium from Zarigan ore leachate.     

Characterization of Portuguese geomaterials,the clay component of <Emphasis Type="Italic">rañas</Emphasis>, as potential liners for low and intermediate radioactive disposal sites

This paper describes the development of a separation method for americium from the effluents emanating from anion exchange column, used for the recovery of plutonium from analytical waste solutions. The waste contained uranium, sodium, calcium and iron as the major impurities as estimated by ICP-AES method. ~99% pure americium was obtained by three separation steps using solvent extraction and extraction chromatography techniques. In the first step, uranium was quantitatively separated by giving five contacts of equal volumes of 30% TBP in n-dodecane. Fe and Na were separated in the next step using 0.1 M TODGA + 0.5 M DHOA as the extractant. In the last step, Am was separated from the co-extracted Ca (about 76%) using CMPO loaded extraction chromatographic column. The overall recovery was >80% with decontamination factor (D.F.) from the impurities being >3000 while the purity of the product was 99%.