Extraction of U(VI), Th(IV) and some fission products from nitric acid medium by sulfoxides and effect of γ-irradiation on the extraction

The extracting abilities for thorium, uranium and some fission products by five sulfoxides are given. The results show that di(2-ethylhexyl) sulfoxide (DEHSO) is not only completely miscible with kerosene, but also superior to tri-n-butyl phosphate in some properties. The extraction behavior of uranium, thorium and some fission products such as zirconium, niobium and ruthenium from aqueous nitric acid with DEHSO in kerosene has been studied over a wide range of conditions. DEHSO extracted uranium and fission products better than TBP under all conditions and is similar to TBP in extraction of thorium. A study of extraction mechanism indicates that U and Th are extracted as disolvates, whereas HNO₃ is extracted as monosolvate. Extraction of the two actinides decreases with increasing temperature, indicating the extraction to be exothermic. Preliminary studies show that -ray irradiated DEHSO extracts Zr and Nb to a smaller extent than irradiated TBP in the range of 10⁴–10⁷ rad.     

Extraction behavior of uranium,plutonium and some fission products with gamma-irradiated N,N′-dialkylamides

The extraction behavior of uranium(VI), plutonium(IV) and fission products like zirconium, ruthenium and europium from 3.5M nitric acid medium with gamma-irradiated dibutyl derivatives of hexanamide (DBHA), octanamide (DBOA) and decanamide (DBDA) in dodecane has been investigated as a function of absorbed dose up to 184 MRads. The results indicate that the K_d value for extraction of uranium(VI) decreases gradually, while K_d for extraction of plutonium(IV) decreases rapidly with dose up to 35 MRads, increasing thereafter with dose, indicating synergistic effects of radiolytic products at higher doses. Ruthenium and europium are not extracted in the entire dose range up to 184 MRads, while extraction of zirconium(IV) increases steadily up to 50 MRads and increases radiply thereafter, indicating synergistic effect of radiolytic products similar to that of plutonium(IV) beyond a dose of 50 MRads. The extractability of uranium(VI) and plutonium(IV) with 1M dibutyl decanamide (DBDA) in dodecane was studied for uranium loading up to 75 mg/ml and plutonium loading up to 3 mg/ml. The percent extraction was found to vary from 91 to 71 for uranium and 95 to 89 for plutonium, respectively. Quantitative stripping of uranium can be achieved with 0.01M nitric acid and plutonium with 0.5M nitric acid and 0.05M hydroxylamine soluton in two steps from an organic phase loaded with 53.2 mg/ml of uranium.     

Separation and Recovery of Uranium and Plutonium from Oxalate Supernatant

The separation of uranium and plutonium from oxalate supernatant, obtained after precipitating plutonium oxalate, containing ~10 g/l uranium and 30–100 mg/l plutonium in 3M HNO₃ and 0.10–0.18M oxalic acid solution has been carried out. In one extraction step with 30% TBP in dodecane: ~92% of uranium and ~7% of Pu is extracted. The raffinate containing the remaining U and Pu is extracted with 0.2M CMPO+1.2 M TBP in dodecane and near complete extraction of both the metal ions is achieved. The metal ions are back extracted from organic phases using suitable stripping agents. The recovery of both the metal ions separately is >99%. The uranium species extracted into the TBP phase from the HNO₃+oxalic acid medium was identified as UO₂(NO₃)₂·2TBP.     

Prompt gamma-ray analysis using cold and thermal guided neutron beams

The extraction behavior of uranium(VI), plutonium(IV) and some fission products like zirconium(IV), ruthenium(III) and europium(III) from 3.5M nitric acid with -irradiated organic phase pre-equilibratedn-dodecane solutions of dihexyl derivatives of hexanamide (DHHA), octanamide (DHOA) and decanamide (DHDA) has been investigated as a function of absorbed dose upto 184·10⁴ Gy. The results indicate that the extraction of uranium(VI) decreases gradually with dose upto 72·10⁴ Gy and becomes almost constant thereafter, while, the extraction of plutonium(IV) decreases upto a dose of 20·10⁴ Gy and then increases rapidly up to a dose of 82·10⁴ Gy indicating synergistic effects of radiolytic products formed at higher doses. Extraction of zirconium(IV) increases gradually upto a dose of 72·10⁴ Gy. Europium(III) does not get extracted with any of these amides in the entire dose range (0–184·10⁴ Gy) studied, however, ruthenium shows insignificant increase in extraction with dose. The decrease inD values noticed in the case of plutonium and zirconium after the dose of 72·10⁴ Gy which was attributed to the third phase formation and emulsification. Infrared studies confirm the final products of radiolysis as the respective amines and carboxylic acids. The degraded amide contents have been estimated by quantitative IR spectrophotometric technique. Extraction data obtained for uranium(VI) and plutonium(IV) with TBP/n-dodecane system have also been compared under similar experimental conditions.     

Solvent extraction of Tc(VII) by the mixture of TBP and 2-nitrophenyl octyl ether

The extraction of Tc(VII) by the mixture of tri-n-butyl phosphate (TBP) and 2-nitrophenyl octyl ether (NPOE) has been studied. 0.2M NPOE-TBP can extract Tc(VII) effectively from 1M HNO₃ and 1M NaOH solutions with distribution ratios of 57.1 and 12.3, respectively. The distribution ratio of Tc(VII) decreases with increasing (>0.5M) HNO₃ concentration but increases with the increase of NaOH concentration. A pH 9 NaOH solution has proven to be suitable for Tc(VII) stripping. A simple extraction-stripping cycle can remove Tc(VII) from a sodium hydroxide solution. A more sophisticated extraction process is proposed to remove Tc(VII) from nitric acid solution because the co-extracted HNO3 prevents the direct stripping of Tc(VII) by NaOH solution of pH 9.     

Studies on the extraction of americium(III) from aqueous nitrate media by sulfoxides

Solvent extraction behaviour of Am(III) from dilute nitric acid media with sulfoxides (R₂SO) in Solvesso-100 has been investigated over a wide range of conditions. Very poor extractability of Am necessitated the use of salting-out agents, viz., nitrates of Al, Mg, Ca, Li and NH ₄ ⁺ . Effects of certain variables such as acidity, extractant concentration, saltingout agent, temperature etc., on metal extraction by sulfoxides have been examined systematically. For a fixed sulfoxide concentration, extraction attains a maximum value up to around 0.2–0.4M HNO₃ and decreasing above 1M HNO₃. In contrast, increasing the concentration of sulfoxide (0.8M DISO, 1.3M DBuSO) gives almost quantitative Am extraction up to 1M HNO₃. For satisfactory extraction, di-n-octyl as well as di-n-hexyl sulfoxide are the most suitable extracting agents. Extractability of Am increases with increasing amounts of all the salting-out agents studied and their effect follows the sequence: Al³⁺>Mg²⁺>Ca²⁺>Li⁺>NH ₄ ⁺ ; this is also the relative dehydrating effect of the cations. The species extracted would appear to be Am(NO₃)₃.3R₂SO. Americium is easily stripped with 1–3M HNO₃ solutions from the loaded organic phase. Extraction decreases with increasing temperature, indicating the extraction to be exothermic. Extraction from partially non-aqueous solutions was also investigated.     

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

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

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 major,minor and trace elements in Iraqi vegetable samples by INAA

Our aim was to discover a method of separating zirconium(IV) and uranium(VI) from solutions. It is known that Zr(IV) and U(VI) are effectively extracted by tertiary amines from weak acidic sulfate solutions but the possibility of extraction decreases with increasing acidity. The transition from tertiary amine to primary amine Primene JMT enables the extraction of Zr also from more acidic solution. If both Zr and U are present in an aqueous solution, Zr is extracted preferentially and only the free part of the amine can convert uranium to an extract. The separation described below was carried out by preferentially stripping zirconium from the organic phase. The application of nitrate solution (2M HNO₃) to eliminate Zr from the solvent was tested. This method does not demand any special regeneration of the extraction agent and the amine nitrate, formed in the organic phase, can be used for further extraction of Zr without modification. Using this method of separation, a solution for producing pure ZrOCl₂·8H₂O was obtained.     

Synergistic extraction of uranium with mixtures of PC88A and neutral oxodonors

The extraction of uranium(VI) from nitric acid medium is investigated using 2-ethylhexyl phosphonic acid-mono-2-ethylhexyl ester (PC88A in dimeric form, H₂A₂) as extractant either alone or in combination with neutral extractants such as tri-n-butyl phosphate (TBP), trioctyl phosphine oxide (TOPO), and dioctyl sulfoxide (DOSO). The effects of different experimental parameters such as aqueous phase acidity (up to 10 M HNO₃), nature of diluent [xylene, carbon tetrachloride (CCl₄), n-dodecane and methyl iso-butyl ketone (MIBK)] and of temperature (303–333 K) on the extraction behavior of uranium were investigated. Synergistic extraction of uranium was observed between 0.5 and 6 M HNO₃. Use of MIBK as diluent was also studied. Temperature variation studies using PC88A as extractant showed exothermic nature of extraction process. Studies were carried out to optimize the conditions for the recovery of uranium from the raffinate generated during the purification of uranium from nitric acid medium. Inductively Couple Plasma Atomic Emission Spectroscopy (ICP-AES) and Energy Dispersive X-Ray Fluorescence (EDXRF) techniques were employed for analysis of uranium in equilibrated samples.     

Solvent extraction of zirconium from hydrochloric acid solutions by sulphoxides and their mixtures

The solvent extraction of zirconium from HCl solutions by dipentyl sulphoxide (DPSO), dioctyl sulphoxide (DOSO), tributyl phosphate (TBP), and their mixtures in various solvents has been studied. At a given H⁺ strength, the extraction coefficient η of the metal increases with an increase in Cl⁻ activity whereas it is almost independent of H⁺ at constant Cl⁻. Under otherwise identical conditions, η increases with an increase in the extractant concentration but is virtually independent of the metal ion concentration over a wide range. The species extracted are ZrCl₄·DPSO, ZrCl₄·DOSO, and ZrCl₄·2TBP. In the case of mixtures, the slope of the log η−log M extractant plot for one component decreases with an increase in the concentration of the second component, the lines crossing at a common point. Extraction is favoured by solvents of low dielectric constant. It is possible to separate zirconium from thorium and uranium by solvent extraction with sulphoxides.     

Alpha radiolytic degradation of <Emphasis Type="Italic">N,N</Emphasis>-dihexyl octanamide

The primary purpose of this study was to understand the alpha radiolytic degradation behavior of N,N-dihexyl octanamide (DHOA) vis a vis tributyl phosphate (TBP) solutions in n-dodecane under plutonium loading conditions. These studies were carried out as a function of dose on different Pu loaded samples (containing 0.002-10 g/L Pu) from 4 M HNO₃ medium. These Pu loaded solutions were evaluated for stripping behavior by contacting with 0.5 M NH₂OH at 0.5 M HNO₃ solutions. Organic phase analysis was carried out by gas chromatography (GC) and by visible spectrophotometry. These studies clearly indicated that Pu stripping becomes difficult with increased dose in the case of TBP system. On the other hand, no such problem was observed in DHOA system during stripping of plutonium, thereby indicating that DHOA is a promising candidate for the reprocessing of high burn up Pu rich spent fuels.     

Complexation of uranium(VI) with acetohydroxamic acid

The advanced separation extraction process based on tri-n-butyl phosphate organic phase called UREX is being developed to separate uranium from fission products and other actinides, and the acetohydroxamic acid (AHA) is employed to reduce and complex plutonium and neptunium in order to decrease their distribution to the TBP-organic phase. In this study, the extraction of uranium was performed from various aqueous matrices with different concentrations of HNO₃, LiNO₃, and AHA. Extraction of uranium increases with increasing both initial HNO₃ and total nitrate concentration. UV-VIS spectrophotometry confirmed that AHA is involved in the complex of uranium with TBP.     

Development of radiochemical analysis of uranium isotopes in highly contaminated soil samples

An accurate and reliable analytical technique of uranium isotopes in highly contaminated soil samples was developed and applied to the IAEA reference samples. The conventional TBP method of uranium isotopes is insufficient to completely purify uranium from actinides such as plutonium and americium isotopes in highly contaminated soil samples. For overcoming the demerits of the conventional TBP extraction method, sample materials were decomposed with HNO₃ and HF, and uranium isotopes were purified by TBP extraction and anion exchange columns and extraction chromatography. Among the purifying methods of uranium, with a TRU Spec resin column after TBP solvent extraction, uranium was completely separated from the radionuclides in a highly contaminated samples. With the modified TBP extraction method, it was found that the concentrations of uranium isotopes were consistent with the reference values reported by the IAEA.     

Recovery of uranium and thorium from zirconium oxychloride by solvent extraction

A solvent extraction process is proposed to recover uranium and thorium from the crystal waste solutions of zirconium oxychloride. The extraction of iron from hydrochloride medium with P350, the extraction of uranium from hydrochloride with N235, and the extraction of thorium from the mixture solutions of nitric acid and the hydrochloric acid with P350 was investigated. The optimum extraction conditions were evaluated with synthetic solutions by studying the parameters of extractant concentration and acidity. The optimum separation conditions for Fe (III) are recognized as 30% P350 and 4.5 to 6.0 M HCl. The optimum extraction conditions for U (VI) are recognized as 25% N235 and 4.5 to 6.0 M HCl. And the optimum extraction conditions for Th (VI) are recognized as 30% P350 and 2.5 to 3.5 M HNO₃ in the mixture solutions. The recovery of uranium and thorium from the crystal waste solutions of zirconium oxychloride was investigated also. The results indicate that the recoveries of uranium and thorium are 92 and 86%, respectively.     

Separation of uranium from different uranium oxide matrices employing supercritical carbon dioxide extraction

Uranium from different uranium oxide matrices was extracted with tri-n-butyl phosphate–nitric acid (TBP–HNO₃) adduct using supercritical carbon dioxide (SC CO₂). While 30 min dissolution time at 323 K was sufficient for U₃O₈ and UO₂ powder, UO₂ granule (at 333 K) and crushed green pellet (at 353 K) required 40 min. Crushed sintered pellet required 60 min at 353 K for complete dissolution. Influence of various experimental parameters such as temperature, pressure, volume of TBP–HNO₃ adduct, acidity of nitric acid used for preparing TBP–HNO₃ adduct and extraction time on uranium extraction efficiency was also investigated. For UO₂ powder, temperature of 323 K, pressure of 15.2 MPa, 1 mL TBP–HNO₃ adduct, 10 M nitric acid and 30 min extraction time was found to be optimum. ~70% uranium extraction efficiency was obtained on extraction with SC CO₂ alone which increased to 90% with the addition of 2.5% TBP in SC CO₂ stream. Extraction efficiency was found to vary linearly with TBP percentage and nearly complete uranium extraction (~99%) was observed with 20% TBP. Nearly complete extraction was also achieved with addition of 2.5% thenoyltrifluoroacetylacetone (TTA) in methanol. The optimized procedure was extended to remove uranium from simulated tissue paper waste matrix smeared with uranium oxide solids.     

Properties of PAN-TBP extraction chromatographic material

Three production routes of the preparation of a solid extractant based on tributylphosphate (TBP) embedded in the polyacrylonitrile matrix (PAN) have been studied. The method of direct PAN coagulation with TBP was found to be not viable due to the significant TBP solubility in the coagulation bath. The most suitable PAN-TBP solid extractant was prepared by the well-known impregnation method of ready-made neat PAN beads. The kinetics of uranium extraction from 3 mol L^?1 HNO₃, the effect of nitrate and nitric acids concentrations on the value of weight distribution coefficients D _g as well as the uranium “extraction isotherm” were determined for this material. Uranium extraction was rather fast, approximately 1 h was sufficient for the equilibrium achievement. Capacity for the uranium uptake, measured in batch experiments on PAN-TBP for 0.048 mol L^?1 of uranium in 3 mol L^?1 nitric acid, was found to be q = 0.363 mmol g^?1 (58 % of the theoretical capacity). It was concluded that PAN-TBP material behaves like TBP in liquid–liquid extraction. Extraction capacity determined in column experiments was lower (by about 23 %) than expected from the “extraction isotherm” due to the TBP leaching out of the column. The thus prepared material is therefore not very suitable for multicycle extraction and stripping and can be used once, particularly for the analytical purposes.     

Extraction behavior of uranium(VI), plutonium(IV), zirconium(IV), ruthenium(III) and europium(III) with γ-pre-irradiated solutions of N,N′-methylbutyl substituted amides in n-dodecane

The extraction of plutonium(IV), uranium(VI), zirconium(IV), europium(III) and ruthenium(III) with -pre-irradiated n-dodecane solutions of methylbutyl substituted hexanamide (MBHA), octanamide (MBOA) and decanamide (MBDA) from 3.5M HNO₃ has been studied as a function of absorbed dose up to 184×10⁴ Gray. The distribution ratios (K_d) of uranium(VI) decreased gradually up to a dose of 50×10⁴ Gray and became almost constant thereafter, while ruthenium(III) and europium(III) were not extracted in the entire dose range studied. The K_d values of Pu(IV) decreased gradually up to 10×10⁴ Gray, for MBOA, and 30×10⁴ Gray for MBHA and MBDA and then increased up to a dose of 72×10⁴ Gray, indicating the synergistic effect of radiolytic products at higher doses. The extraction of zirconium(IV) was found to increase gradually up to 72×10⁴ Gray. However, the steep fall in K_d values of plutonium(IV), zirconium(IV) beyond a dose of 72×10⁴ Gray was atrributed to third phase formation. The radiolytic degradation of amides was monitored by quantitative IR spectroscopy and was found to follow the order MBOA>MBDA>MBHA at 184×10⁴ Gray having the amines and carboxylic acids as the main radiolytic products.     

Liquid-liquid extraction of uranium(VI) by Cyanex 301/alamine 308 and their mixtures with TBP/DDSO

Quantitative extraction of uranium(VI) is observed from 0.2M HCl by 5% (v/v) Cyanex 301. The extraction decreases with increasing acid concentration. Mixtures of Cyanex 301 with tri-n-butyl phosphate (TBP), didecyl sulfoxide (DDSO) and Alamine 308 result in significant synergism in the extraction process, where a species of the type UO₂R₂. L is proposed to be extracted [RH=Cyanex 301 and L=TBP, DDSO or Alamine 308]. Significant extraction of uranium(VI) by 5% (v/v) Alamine 308 is observed at and above 2M HCl, which increases with further increase in acidity attaining a maximum at 6M, after which a slight decrease in extration is observed. Mixtures of Alamine 308 with TBP or DDSO result in a synergism, where a species of the type (R ₃ NH)₂ UO₂Cl₄. Lis extracted. [R ₃ N=Alamine 308, L=TBP or DDSO]. Mixtures of Alamine 308 and Cyanex 301 at 2M HCl result in a profound antagonism in the extraction of uranium(VI).     

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.