Development of a reconversion method for uranyl nitrate to oxide in the reconversion step of reprocessing of irradiated fuel

Ammonium uranyl carbonate (AUC) precipitation is developed for the conversion of uranyl nitrate to oxide in the uranium reconversion step of reprocessing of irradiated fuel by the addition of ammonium carbonate salt. Different precipitation conditions of AUC are studied. The solubility of AUC as a function of uranium concentration in the feed at different temperatures using ammonium carbonate salt as precipitant is studied. This study indicates that 95-99.8% of uranium is recovered as AUC by precipitating 5-125 g/l of uranium with loss of uranium (250-10 ppm) in the filtrate by adding ammonium carbonate salt. It is also observed that the solubility of AUC increased as the concentration of uranium decreased. Thermal decomposition is carried out by thermogravimetry/differential thermal analysis (TG/DTA) and evolved gas analysis-mass spectrometry (EGA-MS) to find out AUC decomposition and gases evolved during decomposition. Studies are also carried out to characterize AUC by using X-ray diffraction (XRD). The data show that AUC obtained by the above conditions is very much consistent with published information.     

Determination of protactinium-233 in irradiated thorium

A method has been devised for the absolute determination of ²³³pa in irradiated thorium meta1 by counting the β-emission of chemically separated and purified sources. The counter was calibrated by a source of ²³⁷Np of known activity in equilibrium with its 233Pa daughter, and checked by 4 π counting     

The dependence of build-up233U,232U,233Pa and fission products from ThO2 irradiated in HFETR on integral thermal neutron fluxes and neutron spectra

In this paper the dependence of build-up²³³U,²³²U,²³³Pa and fission products from ThO₂ irradiated in HFETR on integral thermal neutron fluxes and neutron spectra have been investigated. The yields of all above nuclides in ThO₂ increase with the increase of integral thermal neutron fluxes at different neutron spectra. The values of²³³U/²³²Th increase with the increases of _th and decreases with the increase of fast/thermal neutron ratios (_f/_th). The values of²³²U/²³³U increase with the increase of both _th and _f/_th ratio. The amount of fission products relative to original irradiated thorium decreases with the increase of _f/_th ratios. These results could be used to evaluate the behaviour of thorium-based nuclear fuel in reactor.     

Studies on the recovery of233U from phosphate containing aqueous waste using DBDECMP as extractant

A method for the recovery and purification of²³³U from phosphate containing analytical waste is developed. Extraction studies with DBDECMP (Di-butyl, N, N-diethylcarbamoyl-methyl phosphonate) in xylene were carried out to explore the feasibility of separation and purification of²³³U from such wastes. Based on the data obtained, optimum conditions for the recovery of²³³U are suggested.     

Development of selective separation method for thorium and rare earth elements from monazite liquor

The present work succeeded to develop new optional procedures to enhance the separation process of thorium and REEs. Selective precipitation of thorium with pyrophosphate was successfully attained for the upscale level in which, complete and efficient thorium separation (99%) was achieved with relatively low co-precipitation of REEs (average 15%) and Fe(III) (2.6%). On the other hand, promising and costless method has been developed to optimize the selective precipitation of REEs by adjusting the ratio of the free acids H₂SO₄ to H₃PO₄ at 5:1. It could be obviously demonstrated that about 65.3% of LREEs could be precipitated with a minor amount of thorium 11.9%. Finally, this proposed method could be successfully applied for production of Th and REEs with relatively high yield and purity in addition to low-cost–benefit.     

Preparation, characterization and application of thorium oxalate for sorption of plutonium and americium from aqueous solution

Synthetic inorganic exchangers exhibit good thermal and radiation stability. Thorium oxalate precipitate shows potential for co-precipitation of plutonium and americium from oxalate supernatant generated during plutonium oxalate precipitation. In the present study, efforts were made to prepare thorium oxalate precipitate to be used for column operation. Distribution ratios were determined to optimize conditions for sorption of plutonium and americium on thorium oxalate from nitric acid + oxalic acid solutions with composition similar to that of oxalate supernatant. Column experiments were also performed to evaluate the sorption capacity of thorium oxalate for plutonium and americium from the same medium. The result showed that, thorium oxalate prepared in 1.75M HNO₃ at 70 °C is suitable for column operations. These studies showed that plutonium and americium could be simultaneously removed from aqueous solutions with composition similar to plutonium oxalate waste using glass column packed with thorium oxalate and these nuclides could be recovered by eluting with 3M HNO₃.     

Use of microfiltration as first step in recovery of protein a from fermentation broth

The flux and transmission of protein A during microfiltration have been studied. We studied the performance of two commercial membranes: one made of nylon (Pall Ultipore Nylon₆₆, 0.2 μm) and one of polyether sulfone (Pall Omega, 0.16 μm). The Nylon₆₆ membrane had by far the best transmission of protein A although a previous study showed that bovine serum albumin (BSA), often used to characterize membranes, had much better transmission through the Omega membrane. The membrane manufacturer also states that the Omega membrane is the best membrane for this kind of application because it is a low-protein-binding membrane. The lower transmission of the Omega membrane for protein A was assumed to be owing to its smaller pores and higher charge density in combination with the larger Stokes radius for protein A. When the pH was lowered, the Nylon₆₆ membrane still had the higher transmission. It can thus be concluded that a membrane that is found suitable for the recovery process of one protein is not always the best choice for the recovery process for other proteins even though the membrane is low protein binding.     

Improvements in the recovery of palladium from reprocessing waste of spent nuclear fuel reported earlier

A collection-precipitation (CP) method and an extraction-collection-precipitation (ECP) method have been investigated for high speed and efficient recovery of palladium from reprocessing waste of spent nuclear fuel. For the CP method, a matrix feedstock solution and a small volume of inert solvent, whose specific gravity is greater than that of feedstock solution, as collecting and protective agent, were powerfully mixed with an appropriate amount of potassium iodide solution at 15°C for 1 minute, recovering PdI₂ precipitate after centrifuging. For the ECP method, an extraction complex K(benzo-15-crown-5)₂I in the same organic solvent as for the CP method was mixed with feedstock solution. For both methods, the percent recovery of palladium is more than 99% and 95% for irradiation doses of 1·10³ and 5·10³ Gy, respectively. Decontamination of the palladium product is good.     

An evaluation of the delayed neutron counting method for simultaneous analysis of uranium and thorium and for235U/238U isotopic ratio determination

This paper describes an evaluation of activation analysis by delayed neutron counting to determine uranium and thorium simultaneously in geological materials and to measure²³⁵U/²³⁸U isotopic ratios. A procedure to isolate the thorium before the irradiation was studied and adapted for use when the interference of uranium makes the nondestructive thorium analysis impossible.²³⁵U/²³⁸U ratios were determined in standards with²³⁵U abundances from about 0.5 to 93%, in milligram size samples. Discussion on precision, accuracy and total error of the method is presented.From a thesis submitted by M. J. A. ARMELIN to the University of São Paulo in partial fulfillment of the requirements for a Doctor of Science Degree in Nuclear Technology.     

Development of a microwave-assisted extraction method for the recovery of bioactive inositols from lettuce (Lactuca sativa) byproducts

A microwave-assisted extraction (MAE) method was developed for the extraction of bioactive inositols (D-chiro- and myo-inositols) from lettuce (Lactuca sativa) leaves as a strategy for the revalorization of these agrofood residues. Gas chromatography-mass spectrometry was selected for the simultaneous determination of inositols and sugars (glucose, fructose, and sucrose) in these samples. A Box–Behnken experimental design was used to maximize the extraction of inositols based on the results of single factor tests. Optimal conditions of the extraction process were as follows: liquid-to-solid ratio of 100:1 v/w, 40°C, 30 min extraction time, 20:80 ethanol:water (v/v), and one extraction cycle. When compared with conventional solid-liquid extraction (SLE), MAE was found to be more effective for the extraction of target bioactive carbohydrates (MAE 5.42 mg/g dry sample versus SLE 4.01 mg/g dry sample). Then, MAE methodology was applied to the extraction of inositols from L. sativa leaves of different varieties (var. longifolia, var. capitata and var. crispa). D-chiro- and myo-inositol contents varied between 0.57–7.15 and 0.83–3.48 mg/g dry sample, respectively. Interfering sugars were removed from the extracts using a biotechnological procedure based on the use of Saccharomyces cerevisiae for 24 h. The developed methodology was a good alternative to classical procedures to obtain extracts enriched in inositols from lettuce residues, which could be of interest for the agrofood industry.     

A method for the purification of bulk amount of thorium from its daughter products

A method has been developed to purify bulk amounts of natural thorium from its daughter products. The method is based on anion exchange separation of Th from Pb, Bi, Tl, Ra using Dowex 1 X 8 anion exchange resin conditioned to 2M HCl. The method has several advantages over other methods, namely the decontamination is quite high, Th separation is quantitative and it is suitable for the purification of bulk amounts of Th in a reasonably short time.     

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0–8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0–8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5–7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing.     

Purification of 233U from Thorium and Iron in the Reprocessing of Irradiated Thorium Oxide Rods

A two step precipitation using ammonium carbonate and oxalic acid as the precipitants for thorium and iron is developed for the purification of ²³³U. Ammonium carbonate is added to the feed to increase the pH of the solution. The effect of pH on the solubility of U, Th and Fe in an excess of ammonium carbonate is studied. This indicates that the solubility of Th and Fe is minimum at pH 7 and the recovery of uranium is maximum. The effect of the concentration of thorium and iron on the recovery of uranium at pH 7 is studied. This indicates that the ammonium carbonate precipitation tolerates 2 g/l of thorium and 10 g/l of iron keeping losses of uranium to a minimum. If the feed solution contains more than a tolerable concentration of thorium the precipitation is followed in two steps: (1) Bulk of the thorium is removed by oxalate precipitation, (2) the remaining thorium and iron in the supernatant are removed by ammonium carbonate precipitation. A flow sheet is proposed for the purification of ²³³U from thorium and iron present in a strip product concentrate obtained during the reprocessing of irradiated thorium rods.     

Preconcentration and recovery of uranium and thorium from Indian monazite sand by using a modified fly ash bed

Adsorption of uranium, as UO₂ ²⁺, and thorium, as Th⁴⁺, has been studied using a modified fly ash bed. Effects of pH and various ions like La³⁺, Fe³⁺, Ce⁴⁺, SiO₃ ^2- etc., have been examined. Synthetic mixtures of UO₂ ²⁺ and Th⁴⁺ in different concentrations were passed through the bed and eluted separately with various selective reagents viz. ammonium carbonate, sodium carbonate and acetic acid-sodium hydroxide buffer. Separations of these elements at ppm level are shown to be very effective. The separation of uranium and thorium in the presence of lanthanides in monazite sand has been studied successfully. In the analysis of monazite sand, the oxalate precipitation has been avoided. The method is simple and of very low cost. The modified fly ash bed can also be used to remove uranium from contaminated water.     

Radium accumulation in animal thyroid glands a possible method for uranium and thorium prospecting

A method of prospecting for uranium and thorium is proposed based on uptake of their radioactive daughters²²⁶Ra and²²⁸Ra by plants, the collection of plant material by herbivores, the concentration of the radioactive species by specific animal tissues, and the subsequent gamma-ray analysis of the tissues. This paper is based on work performed under United States Energy Research and Development Administration (formerly U.S.A.E.C.), Contract EY-76-C-06-1830.     

Determination of uranium isotopes in irradiated thorium dioxide by alpha spectrometry using WinALPHA for deconvolution of complex spectra

Isotopic composition of uranium obtained from irradiated thorium dioxide was determined using alpha spectrometry by employing WinALPHA for the deconvolution of the alpha spectra recorded using electrodeposited sources. The results obtained were found to agree within 1% with those determined by thermal ionization mass spectrometry. The deconvolution methodology is important since it is possible to account for the in-growth of ²²⁸Th, which interferes in the determination of ²³²U by alpha spectrometry. The present methodology has the potential to determine isotopic composition of uranium in the irradiated thorium based nuclear fuels, employing alpha spectrometry.     

Development of reliable analytical method for extraction and separation of thorium(IV) by Cyanex 272 in kerosene

A simple and selective spectrophotometric method has been developed for the extraction and separation of thorium(IV) from sodium salicylate media using Cyanex 272 in kerosene. Thorium(IV) was quantitatively extracted by 5 × 10⁻⁴ M Cyanex 272 in kerosene from 1 × 10⁻⁵M sodium salicylate medium. The extracted thorium(IV) was stripped out quantitatively from the organic phase with 4.0 M hydrochloric acid and determined spectrophotometrically with arsenazo(III) at 620 nm. The effect of concentrations of sodium salicylate, extractant, diluents, metal ion and strippants has been studied. Separation of thorium(IV) from other elements was achieved from binary as well as multicomponent mixtures such as uranium(VI), strontium(II), rubidium(I), cesium(I), potassium(I), Sodium(I), lithium(I), lead(II), barium(II), beryllium(II) etc. Using this method separation and determination of thorium(IV) in geological and real samples has been carried out. The method is simple, rapid and selective with good reproducibility (approximately ±2%).     

Development and validation of hybrid K-edge/K-XRF densitometer for assay of nuclear fuel from reprocessing plants

A non-destructive and noninvasive technique is preferable for characterization of the spent fuel solutions in the reprocessing plants, especially samples associated with high radiation dose. A hybrid K-edge/KXRF densitometer (HKED) technique has been designed, fabricated, and commissioned indigenously to accomplish the above objective for the spent fuel solutions from compact reprocessing facility for advanced fuels in lead cells (CORAL), Kalpakkam. This paper describes the customized design, development, calibration and validation of the HKED technique using a series of uranium solutions and corroborating the results with potentiometric method. Further, the pure and mixture of thorium and uranium as well as uranium and plutonium solutions were also assayed by HKED technique.