Redox Behavior of Neptunium(V) in Tributyl Phosphate and <Emphasis Type="Italic">N</Emphasis>,<Emphasis Type="Italic">N</Emphasis>-Dihexyl Octanamide Extractants Dissolved in <Emphasis Type="Italic">n</Emphasis>-Dodecane

Spectrophotometric investigations have been carried out on the disproportionation of Np(V) to form Np(IV) and Np(VI) in 1.1 mol⋅L⁻¹ solutions of tributyl phosphate (TBP) and in N,N-dihexyl octanamide (DHOA) in n-dodecane medium. The Np(V) was found to coordinate with Np(IV) in 1.1 mol⋅L⁻¹ TBP solution in n-dodecane to form a mixed valence “cation–cation” complex by bonding through an axial oxo group on Np(V). By contrast, this interaction was less prominent in the case of 1.1 mol⋅L⁻¹ DHOA solutions. The effect of 1-octanol, added as phase modifier, on the disproportionation behavior of Np(V) was also investigated. An attempt was made to calculate the disproportionation/reduction rate constants for Np(V) under the conditions of these studies. Absorbance measurements on the Np stripped from organic phases revealed the occurrence of Np(V) in the aqueous phase.     

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.     

Separation of uranium and thorium using tris(2-ethylhexyl) phosphate as extractant

The distribution behavior of uranium and thorium has been investigated in a biphasic system of different aqueous nitric acid concentrations and a solution of tris(2-ethylhexyl) phosphate (TEHP) inn-dodecane at 25°C. The effect of different uranium and thorium concentrations in the aqueous phase on the extraction of these metal ions is evaluated. These results indicate that TEHP is a better choice than tri-n-butyl phosphate (TBP) for the separation of²³³U from the irradiated thorium matrix.     

Composition and structure of complexes formed in the reaction of uranyl di(2-ethylhexyl)phosphate and tributyl phosphate or di-N-octyl sulfoxide in benzene solutions

This paper reprots of³¹P NMR and IR studies of the interaction of tributyl phosphate (TBP) and di-n-octyl sulfoxide (DOSO) with polymer molecules of uranyl di-2-ethylhexyl phosphate (UO₂X₂)_p (I) in C₆H₆ sulutions. Detailed interpretations of the³¹P NMR spectra and the v_as(POO) IR bands and determination of the fraction of nonequivalent phosphorus atoms of X⁻ anions and uranium (VI) atoms as well as the concentration of U(VI)-bonded TBP in I have shown that only a single TBP or DOSO molecule is coordinated to the uranium atoms of polymer I at C_TBP=0.1–2 M or C_DOSO=0.1–0.5 M. In the case of 100% TBP, two TBP molecules are coordinated to some U(VI) atoms. Distribution of TBP (DOSO) molecules along the polymer chain agrees with the mean statistical value. The portion of terminal chalate POO-groups of X⁻ anions is determined. The dependence of the degree of (UO₂X₂)_p·nL (L=TBP, DOSO) polymerization on C_L is obtained. Saturation of solutions with water only slightly affects the terminal POO-groups and has no effects on the distribution of L along the polymer chain of I. Institute of Catalysis, Siberian Branch, Russian Academy of Sciences. Translated fromZhurnal Strukturnoi Khimii, Vol. 35, No. 6, pp. 66–73, November–December, 1994. Translated by K. Shaposhnikova     

A high efficient sorption of U(VI) from aqueous solution using amino-functionalized SBA-15

Uranium is one of the most hazardous heavy metal due to its long half-life radioactivity, high toxicity and mobility as aqueous uranyl ion (UO₂ ²⁺) under ordinary environmental conditions. Herein, amino functionalized SBA-15 (APSS) was developed as a rapid and efficient sorbent for removal of U(VI) from the environment. The APSS sample was synthesized by grafting method and was characterized by SEM, NMR, SAXS, and N₂ sorption/desorption isothermal experiments. The sorption of U(VI) by APSS was investigated under different conditions of pH, contact time, initial U(VI) concentration, ionic strength and solid–liquid ratio. The results show that the sorption of U(VI) by APSS is strongly dependent on pH but independent of ionic strength and solid–liquid ratios (m/V). The sorption is ultrafast with an equilibrium time of less than 30 min, and the sorption capacity is as large as 409 mg/g at pH 5.3 ± 0.1. Besides, the U(VI) sorption by APSS from extremely diluted solution and the desorption of U(VI) from APSS were also studied. It is found that 100 mg of APSS can almost completely remove the U(VI) ions from 4 L aqueous solution with the U(VI) concentration as low as 4.2 ppb and the sorbed U(VI) can be completely desorbed by 0.1 mol/L nitric acid. The results strongly reveal the high performance of the APSS material in the removal and preconcentration of U(VI) from the aqueous solution.     

Preconcentration of uranium(VI) and thorium(IV) from aqueous solutions using low-cost abundantly available sorbent

Olive cake as low-cost abundantly available sorbent has been characterized by N₂ at 77 K adsorption, porosity analysis, elemental analysis and IR spectra and has been used for preconcentrating of uranium(VI) and thorium(IV) ions prior to their determination spectrophotometrically. The optimum pH values for quantitative sorption of U(VI) and Th(IV) are 4–7 and 3–7, respectively. The enrichment factor for the preconcentration of U(VI) and Th(IV) were found to be 125 and 75 in the given order. The sorption capacity of olive cake is in the range of 2,260–15,000 μg g⁻¹ for Th(IV) and in the range of 1,090–17,000 μg g⁻¹ for U(VI) at pH 3–7. The sorbent exhibits good reusability and the uptake and stripping of the studied ions were fairly rapid. The elution of U(VI) and Th(IV) was performed with 0.3–1 M HCl/1–2 M HNO₃ and 0.3–0.8 M HCl/1 M HNO₃, respectively. The precision of the method was 1.8 RSD% for U(VI) and 2.5 RSD% for Th(IV) in a concentration of 1.00 μg mL⁻¹ for 10 replicate analysis. The influence of some electrolytes and cations as interferents was discussed. Separation of U(VI) and Th(IV) from other metal ions in synthetic solution was achieved.     

Mathematical modeling of solvent extraction of uranium from sulphate media employing 2-ethylhexyl phosphonic acid-mono-2-ethylhexyl ester (PC88A) and its mixture with trioctylphosphine oxide (TOPO) as extractants

The extraction of U(VI) from sulphate medium with 2-ethylhexyl phosphonic acid-mono-2-ethylhexyl ester (PC88A, H₂A₂ in dimeric form) in n-dodecane has been investigated under varying concentrations of sulphuric acid and uranium. Slope analysis of uranium (VI) distribution data as a function of PC88A concentration suggests the formation of monomeric species, viz. UO₂(HA₂)₂. This observation was further supported by the mathematical expression obtained during non-linear least square regression analysis of U(VI) distribution data correlating the percentage extraction (%E) and the acidity (H _i). A mathematical model correlating the experimental distribution ratio values of U(VI) (D _U) with initial acidity (H _i) and initial uranium concentrations (C _i) was developed: D_\textU = 12.98( ±0.90)/{ C_\texti ^{- 0.75( ±0.05)} ×[ H_\texti ]² } D_{\text{U}} = 12.98( \pm 0.90)/\left\{ {C_{\text{i}}^{ - 0.75( \pm 0.05)} \times \left[ {H_{\text{i}} } \right]^{2} } \right\} . This expression can be used to predict the concentration of uranium in organic as well as in aqueous phase at any C _i and H _i. The extraction data were used to calculate the conditional extraction constant (K _ex) values at different acidities (2–7 M H⁺), uranium (0.02–0.1 M) and PC88A (0.2–0.6 M) concentrations. These studies were also extended for the extraction of U(VI) using synergistic mixtures of PC88A and TOPO from sulphate medium.     

Determination of optimum process conditions for solvent extraction of thorium using Taguchi method

Solvent extraction of thorium was studied using Taguchi method. The effect of various parameters such as acid types (sulfuric, nitric, hydrochloric, sulfuric + nitric) and their concentrations from 0.001 to 4 M, initial thorium concentration (0.0001, 0.001, 0.01, 0.1 M) and solvent type (TBP, D₂EHPA, Cyanex921, Cyanex272) in the ranges of 0.001 to 1 M on thorium extraction efficiency were investigated. The maximum extraction of thorium was obtained while 0.001 M hydrochloric acid, 0.001 or 0.01 M thorium and Cyanex272 were used. Under these optimum conditions, the extraction percent and distribution coefficient of thorium were 98.7% and 73.8, respectively. Compared with the hydrochloric aqueous solution, the nitric acid system showed less variation in the extraction of thorium. The proposed process has been applied for the separation of Th(IV), U(VI), La(III), and Ce(III) from synthetic solution same as thorium ores (monazite).     

<Emphasis Type="Italic">p</Emphasis>-<Emphasis Type="Italic">tert</Emphasis>-Butylcalix[8]arene loaded silica gel for preconcentration of uranium(VI) via solid phase extraction

This paper reports silica gel loaded with p-tert-butylcalix[8]arene as a new solid phase extractor for determination of trace level of uranium. Effective extraction conditions were optimized in column methods prior to determination by spectrophotometry using arsenazo(III). The results showed that U(VI) ions can be sorbed at pH 6 in a mini-column and quantitative recovery of U(VI) (>95–98%) was achieved by stripping 0.4 mol L⁻¹ HCl. The sorption capacity of the functionalized sorbent is 0.072 mmol uranium(VI) g⁻¹ modified silica gel. The relative standard deviation and detection limit were 1.2% (n = 10) for 1 μg uranium(VI) mL⁻¹ solution and 0.038 μg L⁻¹, respectively. The method was employed to the preconcentration of U(VI) ions from spiked ground water samples.     

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.     

Studies on the radiolytic stability of N,N,N′,N′-tetra-2-ethylhexyl diglycolamide in n-dodecane solution containing different phase modifiers

The radiolytic stability of a branched diglycolamide extractant, namely N,N,N′,N′-tetra-2-ethylhexyl diglycolamide (T2EHDGA) dissolved in n-dodecane containing several phase modifiers, viz. N,N-dihexyloctanamide (DHOA), tri-n-butyl phosphate (TBP), 1-decanol and iso-decanol has been investigated. The distribution ratio of Am(III) decreased with increased radiation dose studied up to 1000 kGy. Nevertheless, all the composition of extractants showed satisfactory results up to 500 kGy, beyond which the extractants degraded drastically. The stripping behaviour of Am(III) with 0.2 M HNO₃ was found to be unaffected even with the ligand solution irradiated up to 1000 kGy. Extraction of fission product and structural elements was also investigated using the irradiated solvents and was found to be not significantly affected with increasing absorbed dose with the exception of Mo which showed sharp rise in the distribution coefficient values. Loading of Nd in the organic phase decreased with the irradiated solvent due to degradation of the carrier. The effect of the absorbed dose on physical parameters such as density, viscosity and interfacial tension of the solvents has also been investigated.     

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.     

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.     

Studies on extraction behaviour of molybdenum (VI) from acidic radioactive waste using 2(ethylhexyl) phosphonic acids,mono 2(ethylhexyl) ester (PC-88A)/<Emphasis Type="Italic">n</Emphasis>-dodecane

This paper describes the studies on the extraction of molybdenum (VI) from aqueous nitric acid medium by (2-ethylhexyl) phosphonic acid, mono (2-ethylhexyl) ester (PC-88A). The extraction affecting parameters such as concentration of HNO₃ in aqueous feed, effect of concentration of extractants, effect of diluents, and molybdenum concentration in the aqueous phase are investigated to optimize the extraction conditions for the quantitative separation of molybdenum from nitric acid medium. With increase of HNO₃ concentration in aqueous phase, percentage extraction was found to be decreased in all the cases. Percentage extraction of molybdenum increases with increase in PC-88A concentration till the 0.15 M of PC88A, and after that it becomes constant. Kerosene and n-dodecane was found to be most suitable diluents. Among the various strippants used 0.2 M (w/v) solution of Na₂CO₃ and 0.2 M (w/v) solution (NH₄)₂CO₃ are found to be the equally suitable for stripping of molybdenum from the loaded organic phase. The stripping of molybdenum from loaded organic layer by various reagents followed the order: (NH₄)₂CO₃ >Na₂CO₃ >0.1 M sodium salt of EDTA >2 M NaOH >8 M HNO₃. The optimized process conditions are employed to extract molybdenum (VI) from actual Davies–Gray waste as well as from diluted high level waste generated in the purex stream. More than 94% Mo(VI) was extracted from radioanalytical as well as from high level waste of purex process and quantitative recovery was achieved in both the cases when 0.2 M sodium carbonate was used as stripping agent.     

Smart thorium and uranium determination exploiting renewable solid-phase extraction applied to environmental samples in a wide concentration range

A smart fully automated system is proposed for determination of thorium and uranium in a wide concentration range, reaching environmental levels. The hyphenation of lab-on-valve (LOV) and multisyringe flow injection analysis (MSFIA), coupled to a long path length liquid waveguide capillary cell, allows the spectrophotometric determination of thorium and uranium in different types of environmental sample matrices achieving high selectivity and sensitivity levels. Online separation and preconcentration of thorium and uranium is carried out by means of Uranium and TEtraValents Actinides resin. The potential of the LOV–MSFIA makes possible the full automation of the system by the in-line regeneration of the column and its combination with a smart methodology is a step forward in automation. After elution, thorium(IV) and uranium(VI) are spectrophotometrically detected after reaction with arsenazo-III. We propose a rapid, inexpensive, and fully automated method to determine thorium(IV) and uranium(VI) in a wide concentration range (0–1,200 and 0–2,000 μg L^-1 Th and U, respectively). Limits of detection reached are 5.9 ηg L^-1 of uranium and 60 ηg L^-1 of thorium. Different water sample matrices (seawater, well water, freshwater, tap water, and mineral water), and a channel sediment reference material which contained thorium and uranium were satisfactorily analyzed with the proposed 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%.     

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%.     

Successive uranium and thorium adsorption from Egyptian monazite by solvent impregnated foam

The present work deals with uranium and thorium recovery from the Egyptian monazite sulfate leach liquor using the extraction chromatography technique (solvent impregnated material), where tributylamine (TBA) and di-n-octylamine (DOA) solvents were impregnated onto foam uranium and thorium separate recovery. The calculated theoretical capacities of the latter solvents were about 1.4 gU/g foam and 1.6 gTh/g foam, respectively. The attained uranium and thorium adsorption efficiencies (using ion-exchange columnar technique) were about 75 and 70% of its theoretical capacities, respectively. Using 1 M NaCl–0.1 M H₂SO₄ and 2 M H₂SO₄ as eluent solutions for uranium and thorium from the loaded solvents impregnated foam gave 95.8 and 98.7% elution efficiencies, respectively.     

Tuning the diglycolamides for modifier-free minor actinide partitioning

The unsymmetrical diglycolamides (DGAs) such as N,N-dihexyl-N′,N′-dioctyl-3-oxapentane-1,5-diamide (DHDODGA), N,N-didecyl-N′,N′-dioctyl-3-oxapentane-1,5-diamide (D²DODGA), N,N-didodecyl-N′,N′-dioctyl-3-oxapentane-1,5-diamide (D³DODGA), were synthesized, and characterized by IR, NMR, and mass spectroscopic techniques. The extraction behaviour of Am(III), Eu(III), and Sr(II) by the solutions of these unsymmetrical DGAs in n-dodecane was studied as a function of concentration of nitric acid and DGA. The distribution ratio of Am(III) and Eu(III) increased with increase in the concentration of nitric acid; whereas, the distribution ratio of Sr(II) reached a maximum at 4 M nitric acid followed by decrease at higher acidities. The extraction of Am(III) and Eu(III) in 0.1 M DGA/n-dodecane decreased in the order DHDODGA > D²DODGA > D³DODGA. However, the order changed upon lowering the concentration of DGA. The third-phase formation behaviour of nitric acid and neodymium(III) in 0.1 M DGA/n-dodecane was studied as a function of concentration of nitric acid. The limiting organic concentration of nitric acid and neodymium increased with increase in the chain length of alkyl group attached to amidic nitrogen. Near stoichiometric amount of neodymium(III) was loaded in 0.1 M D³DODGA/n-dodecane without the formation of third-phase from 3 to 4 M nitric acid medium. The study revealed that the unsymmetrical diglycolamides D²DODGA and D³DODGA are superior candidates for partitioning the minor actinides from high-level liquid waste.     

Polarographic waves of tri-n-butyl phosphate and polarographic determination of uranium in the presence of tri-n-butyl phosphate

The results of a study on the polarographic behaviour of TBP and its influence on the determination of uranyl ions is presented. The half-wave potential of the adsorption wave of TBP depends on the concentration of TBP, type of supporting elec trolyte and its concentration. In the presence of TBP the polarographic wave of U(VI) ion is changed. Below 7·10^?5 M TBP the polarographic wave of U(VI) is not affected, between 7·10^?5 and 2·10^?4 M TBP the shape, height and half-wave potential of U(VI) waves are changed and above 2·10^?4 M, up to saturated solution of TBP, the waves of U(VI) do, not change further. The bes supporting electrolytes for the determination of U(VI) are KNO₃ or NaClO₄ in concentrations of 0.1 to 0.5 M, pH 1–2 and TBP concentrations from 3·10^?4 to 1.2·10^?3 M.