Recovery of uranium from phosphate by carbonate solutions

Uranium concentrations were analyzed in the Syrian phosphate deposits. Mean concentrations were found between 50 and 110 ppm. As a consequence, an average phosphate dressing of 22 kg/ha phosphate would charge the soil with 5–20 g/ha uranium when added as a mineral fertilizer. Fine grinding phosphate produced at the Syrian mines was used for uranium recovery by carbonate leaching. The formation of the soluble uranyl tricarbonate anion UO₂(CO₃)₃ ⁴⁻ permits using alkali and sodium bicarbonate salts for the nearly selective dissolution of uranium from phosphate. Separation of iron, aluminum, titanium, etc., from uranium during leaching was carried out. Formation of some small amounts of molybdates, vanadates, phosphates, aluminates, and some complex metals was investigated. This process could be used before the manufacture of Tri-Super Phosphate (TSP) fertilizer, and the final products would contain less uranium quantities.     

Recovery of plutonium from aqueous waste solutions using porous zirconia spherical particles

Based on the Linssi database and UniSampo/Shaman software, an automated analysis platform has been setup for the analysis of large amounts of gamma-spectra from the primary coolant monitoring systems of a CANDU reactor. Thus, a database inventory of gaseous and volatile fission products in the primary coolant of a CANDU reactor has been established. This database is comprised of 15,000 spectra of radioisotope analysis records. Records from the database inventory were retrieved by a specifically designed data-mining module and subjected to further analysis. Results from the analysis were subsequently used to identify the reactor coolant half-life of ¹³⁵Xe and ¹³³Xe, as well as the correlations of ¹³⁵Xe and ⁸⁸Kr activities.     

Recovery of plutonium from phosphate containing aqueous analytical waste solutions using macroporous anion exchange resin

Distribution ratios of Pu(IV) between 7.5M HNO₃+0.75M H₃PO₄+0.3M H₂SO₄ media and a macroporous anion-exchange resin Amberlyst A-26 (MP) increased from 40 to 250 when 1M aluminium nitrate was added to the aqueous medium. When 1M ferric nitrate was used in place of aluminium nitrate the distribution ratio further increased to 850. The 10% Pu(IV) breakthrough capacities with a 5 ml bed resin column, using synthetic feed solutions containing 1M aluminium nitrate, were 1.4 g l^–1, 3.2 g l^–1 at flow rates of 30 ml per hour and 10 ml per hour, respectively. The corresponding 10% Pu(IV) breakthrough capacities in the presence of 1M ferric nitrate were 8.5 g l^–1 and 12.8 g l^–1. More than 97% of plutonium could be recovered from actual analytical phosphate waste solutions.     

Recovery of plutonium and cerium from alumina crucible

Journal of Radioanalytical and Nuclear Chemistry - Plutonium recovery is inevitable from plutonium bearing alumina crucibles generated over the years as part of carbon analysis during chemical...     

Precipitation of plutonium oxalate from homogeneous solutions

A method for the precipitation of plutonium(IV) oxalate from homogeneous solutions using diethyl oxalate is reported. The precipitate obtained is crystalline and easily filterable with yields in the range of 92–98% for precipitations involving a few mg to g quantities of plutonium. Decontamination factors for common impurities such as U(VI), Am(III) and Fe(III) were determined. TGA and chemical analysis of the compound indicate its composition as Pu(C₂O₄)₂·6H₂O. Data are obtained on the solubility of the oxalate in nitric acid and in mixtures of nitric acid and oxalic acid of varying concentrations. Green PuO₂ obtained by calcination of the oxalate has specifications within the recommended values for trace foreign substances such as chlorine, fluorine, carbon and nitrogen.     

Recovery of plutonium and americium from laboratory acidic waste solutions using tri-n-octylamine and octylphenyl-N-N- diisobutylcarbamoylmethylphosphine oxide

Plutonium from acidic waste solutions has been recovered quantitatively using tri-n-octylamine (TnOA) in xylene and americium using a mixture of octylphenyl-N-N- diisobutylcarbamoylmethylphosphine oxide (CMPO) and TBP in dodecane by extraction and extraction chromatographic methods. The Pu ( IV ) TnOA species extracted into the organic phase from higher nitric acid concentrations has been confirmed as (R(3)NH)(2)Pu(NO(3))(6) (where R(3)N = TnOA by employing slope analysis as well as spectrophotometric studies.     

Studies on sorption of plutonium from carbonate medium on polyacrylhydroxamic acid resin

Polyacrylhydroxamic acid resin synthesized by functionalization of polyacrylamide with hydroxylamine has been investigated for the sorption of plutonium(IV) from carbonate medium, aiming at its application for the removal of plutonium from alkali wash effluent generated during purification of TBP in PUREX process. Batch experiments have been carried out to determine distribution coefficient of plutonium(IV) between this exchanger and various compositions of carbonate medium. Effect of the concentration of sodium carbonate, sodium bicarbonate and pH of the solution on the distribution coefficient have been studied to optimize the conditions for the uptake of Pu(IV) by this exchanger. Column experiments were carried out to determine the practical capacity of the exchanger for plutonium. Elution studies were also carried out to recover the loaded plutonium from the ion exchange column The exchanger displayed good exchange capacity for Pu(IV) from feed solution simulating the conditions of carbonate wash effluent generated in PUREX process. The exchanger also exhibited fast elution of Pu, suggesting the feasibility of using it for the recovery of Pu from carbonate based wash effluent.     

Removal of plutonium from carbonate medium using titania microspheres prepared by sol-gel route

Spherical beads of titania were prepared by a sol-gel route. During the preparation a cationic surfactant was added in the feed broth to modify the surface characteristics of the microspheres. The absorption behavior of Pu(IV) from carbonate medium on these microspheres was studied by batch experiments. Loading and elution behavior of Pu(IV) on a bed of titania microspheres was studied and the practical capacity was determined. This revised version was published online in July 2006 with corrections to the Cover Date.     

Studies on sorption of plutonium on inorganic ion exchangers from sodium carbonate medium

Sorption of Pu(IV) from sodium carbonate medium has been investigated by using three inorganic ion exchangers, viz. alumina, silica gel and hydrous titanium oxide (HTO). Distribution ratios (D) of Pu(IV) for its sorption on these ion exchangers have been determined. The values are 700, 10³ and 10⁴ for alumina, silica gel and hydrous titanium oxide, respectively, from 0.1M sodium carbonate medium. The high distribution ratios indicate their suitability for the removal of Pu(IV) from sodium carbonate waste streams. Pu(IV) breakthrough capacities have been determined with 5 ml bed at a flow rate of 30 ml per hour. The 10% Pu(IV) breakthough capacities for alumina and silica gel are 3 g l^–1 and 14 g l^–1, respectively. The capacity of HTO is 60 g of Pu(IV) per liter of exchanger at 4% Pu(IV) breakthrough.     

Complexing of tetravalent plutonium in aqueous solutions

Complex formation by tetravalent plutonium at I = 2 and [H⁺] = 2 M, has been studied by solvent extraction methods using vanadium(V) to eliminate the possibility of a change in the oxidation state of Pu(IV) during the experiment or complexing of Pu(IV) by the holding oxidant itself. Sulphate complexing of Pu(IV) has been studied using three extractants, viz. thenoyltrifluoroacetone, dinoylnaphthalene sulphonic acid and tri-n-butylphosphate, and nitrate, chloride and fluoride complexing of Pu(IV) has been studied using the thenoyltrifluoroacetone extraction method.     

Recovery of plutonium from graphite crucibles generated in inert gas/vacuum fusion technique

A simple and rapid method has been developed for the separation and purification of plutonium from solid analytical waste (containing uranium, plutonium, nickel and graphite) generated during the analysis of the nuclear fuels for their oxygen and nitrogen contents by inert gas fusion technique. The method is based on crushing the graphite crucibles, electromagnetic separation of plutonium-nickel alloy, dissolution in nitric acid and ion exchange purification of plutonium. Recovery of plutonium is better than 98%. The method may be extendable for the recovery of any other valuable materials from such analytical waste.     

Hydrogen generation in nitric acid solutions of plutonium

Summary Literature data on the radiolytic generation of hydrogen in nitric acid solutions of plutonium are used to construct a model that predicts G(H₂) as a function of the nitric acid and plutonium concentrations. The model indicates that G(H₂) decreases with increasing concentration of nitric acid, in agreement with most experimental observations. The effect of the plutonium concentration on G(H₂) is secondary to the effect of the acid concentration. An equation for interpolating Gvalues for total gas is included.     

Dielectric properties of propylene carbonate and propylene carbonate solutions

Permittivity data at frequencies from 0.9 to 12 GHz for propylene carbonate and for the solutions of NaI, NaClO₄, Bu₄NI, Bu₄NClO₄, ZnBr₂, and Ca(ClO₄)₂ in propylene carbonate at 25°C are reported and discussed. The contaminating influence of water on the dielectric spectra is shown. Measurements were executed by the method of travelling waves with equipment known to produce data of high precision. Evaluation of the data is performed on the basis of models presupposing one or more relaxation regions. The dielectric spectra of all salts with the exception of ZnBr₂ yield relaxation time distributions with a single critical relaxation time or can be analyzed by assuming two critical relaxation times for the solvent. ZnBr₂ solutions show a supplementary relaxation region at low frequencies which is attributed to the solute. The variation of permittivities at zero frequency with the salt concentration is discussed in the framework of kinetic depolarization. Solvation numbers are estimated.     

Recovery of plutonium from oxalate supernatant using 1-phenyl-3-methyl-4-benzoyl-5-pyrazolone

Summary Extraction of Pu(IV) from oxalate supernatant was carried out employing 1-phenyl-3-methyl-4-benzoyl-5-pyrazolone (PMBP) in xylene as extractant. The conditions for quantitative extraction were determined by the variation of ligand, oxalic acid and nitric acid concentration. Quantitative stripping was achieved using a mixture of 0.4M oxalic acid and 0.4M ammonium oxalate. Extraction of Pu(IV) from synthetic oxalate supernatant solution containing 3M nitric acid and 0.2M oxalic acid was investigated under various loading conditions employing 1-phenyl-3-methyl-4-benzoyl-5-pyrazolone in xylene as extractant. Under uranium loading conditions the Pu extraction decreased significantly while with increased Pu loading whereas the D_Pu value was influenced marginally. The effect of a redox reagent on Pu extraction was also investigated.     

Recovery of137Cs from acidic fission products solutions

A procedure for the isolation of¹³⁷Cs from acidic fission products solutions, based on the use of silica gel and zirconium phosphate ion exchangers, is presented. The¹³⁷Cs recovered by the ion-exchange process is converted to powder by coprecipitation of cesium with ammonium molybdophosphate.¹³⁷Cs pellets have been prepared by compression of¹³⁷Cs ammonium molybdophosphate powder using a hydraulic press. An important aspect of this procedure is that it does not require neutralization of the Purex waste.