Progress in neutron activation analysis for uranium

A new type of extractant, sym-dibenzo-16-crown-5-oxyhydroxamic acid (HL) is introduced. The extractions of UO22+, Na+, K+, Sr2+, Ba2+ and Br- were studied with HL in chloroform. The results obtained show that UO22+ can be quantitatively extracted at pH values above 5, whereas the extractions of K+, Na+, Sr2+, Ba2+ and Br- are negligible in the pH range of 2 - 7. The dependence of the distribution ratio of U(VI) on both the concentration of the HL and pH are linear, and they have the same slope of 2. This suggests that U(VI) appears to form a 1:2 complex with ligand. Uranium(VI) can be selectively separated and concentrated from interfering elements such as Na, K, Sr and Br by solvent extraction with HL under specific conditions. The recovery of uranium is nearly 100% and the radionudear purity of uranium is greater than 99.99%. Therefore, neutron activation analysis has greatly improved the sensitivity and accuracy for the detection of trace uranium from seawater.     

Electrochemical separation of uranium and cerium in molten LiCl–KCl

The electrochemical separation of uranium from cerium in LiCl–KCl eutectic and the electrochemical behavior of Ce(III) were studied. According to the cyclic voltammogram of Ce(III) and the former result of U(III), electrodeposition potential was determined at ?1.65 V (vs Ag/AgCl). The uranium metal was successfully deposited and separated from cerium. The morphology of deposit and cross section of electrode were investigated by SEM, firstly uranium deposit alloys with stainless steel and forms a thin transition layer, and secondly the uranium metal layer grows from the transition layer. The separation factors of uranium/cerium on different recovery ratios were determined through a series of steps. It was found that the content of cerium in the deposit and separation factors declined with increasing the initial concentration of U³⁺ in molten salts; the separation factors remained stable at around 20 in different uranium recovery ratios.     

Anion-exchange separation and determination of thorium and uranium in Eskişehir-Beylikahir ore in Turkey

Eskişehir-Beylikahir ore contains rare earth elements (REE) and thorium as bastnaesite mixed with barite, calcite and fluorite and small amounts of uranium. The mineralogical composition and Th, U, REE contents of the ore samples taken from different sectors of the same region show important variations. For the spectrophotometric determination of Th and U in the ore, successive anion-exchange and solvent extraction procedures are carried out to separate U and Th from each other and from accompanying interfering elements. The accuracy and precision of the method is demonstrated by intercomparison of the thorium standard reference samples S-14, S-15 and S-16 of IAEA. In all cases very good agreement was obtained.     

Microwave-assisted dissolution of ceramic uranium dioxide in TBP–HNO3 complex

Microwave-assisted dissolution of ceramic uranium dioxide in tri-n-butyl phosphate (TBP)–HNO₃ complex was investigated. The research on dissolution of ceramic uranium dioxide in TBP–HNO₃ inclusion complex under microwave heating showed the efficiency of the use of this method. Nitric acid present in the inclusion complex participates both dissolution of UO₂, and oxidation of U(IV)–U(VI), the resulting UO₂(NO₃)₂ extracted with tri-n-butyl phosphate. Dissolution rate depends on both temperature of microwave dissolution process, and concentration of nitric acid present in the inclusion complex. The most intensive dissolution process is when the concentration of nitric acid ≥2 mol/L and the temperature of 120 °C. From the experimental data obtained by two kinetic models activation energies were calculated. At the average activation energy of UO₂ dissolution in TBP–HNO₃ complex equal 70 kJ/mol, and reaction order is close to one, i.e. the reaction takes place in an area close to kinetic.     

Selective extraction of uranium from uranium–beryllium ore by acid leaching

Journal of Radioanalytical and Nuclear Chemistry - Both uranium and beryllium are very important strategic metals and have been applied to many fields, such as nuclear industries, atomic energy,...     

Electrochemistry of uranium in LiF–BeF2 melt

A study of the electrochemistry of uranium in LiF–BeF₂ system important for molten salt reactor concept was conducted at W and Ni electrodes. Cyclic voltammetry and chronopotentiometry methods were used. Two-step reduction mechanism for U⁴⁺ ions involving one electron exchange in soluble/soluble U⁴⁺/U³⁺ system and three electron exchange in the second step were found on W electrode. Both processes were identified as reversible and diffusion-controlled. Based on voltammetric and chronopotentiometric measurements, the diffusion coefficient of U⁴⁺ ions at 813 K was calculated: D(U⁴⁺) = 1.26 × 10⁻⁶ cm² s⁻¹ and D(U⁴⁺) = 1.28 × 10⁻⁶ cm² s⁻¹, respectively. Formation of U–Ni alloys was observed on Ni electrode.     

Assay of uranium in crude diuranate cakes and MgF2 slag produced at the natural uranium conversion plants by γ-ray spectrometry

A transmission-corrected -ray counting method has been employed for the assay of uranium in crude Na₂U₂O₇ cakes produced at the Uranium Conversion Facilities. A 3×3 NaI(TI) detector was used in conjunction with a 400-channel analyzer. The observed count rate of the 1 MeV -ray emitted by the²³⁸U in the sample was corrected for sample self-attenuation, measured with a⁶⁵Zn (-energy 1115 keV) transmission source. A calibration factor determined by measuring a standard of known amount of radioactive material in the same form and geometry as the unknown sample was used to convert the transmission corrected count rate to the amount of uranium in the weighted sample. Another -spectrometric method is described for the assay of the U-content in the MgF₂ slag produced during the magnesiothermic reduction of UF₄ to U-metal ingots at the natural U-conversion plant.     

Solid-phase extraction–spectrophotometric determination of uranium(VI) in natural waters

A method for the extraction and determination of uranyl ion in natural waters using octadecyl bonded silica membrane disks modified with piroxicam and spectrophotometry with arsenazo(III) is proposed. The perconcentration step was studied with regard to experimental parameters such as amount of extractant, type and amount of eluent, pH, flow rates and tolerance limit of diverse ions on the recovery of uranyl ion. The limit of detection of the proposed method is 0.4 micro g L(-1) of uranyl. The method was applied to the recovery of uranyl from different water samples.     

Electrochemical behavior of uranium(III) in NaCl–KCl molten salt

The electro-redox behavior of uranium(III) on Mo electrode in NaCl–KCl molten salt in the temperature range 973–1073 K has been investigated using cyclic voltammetry electrochemical method and so on, such research will help to understand uranium behavior in pyro-reprocessing. The results showed that UCl₃ could be reduced into uranium metal in a quasi-reversible one-step process exchanging three electrons. The diffusion coefficients of U(III) ions were determined and the activation energy for diffusion was found to be 55.794 kJ mol⁻¹. The apparent standard potentials of U(III)/U(0) at several temperatures were calculated. The thermodynamic properties of UCl₃ have also been investigated.

     

Complexes triflates de l’uranium

Uranium triflate complexes. We review here the different preparations of uranium triflates that we have developed in the course of these last years in our laboratory. Protonation of 〚U〛–R and 〚U〛–NR₂ (R = alkyl) bonds with pyridinium triflate constitutes a general and efficient route towards triflate complexes. This method is very suitable for the preparation of organometallic compounds such as U(Cp)₃(OTf), U(Cp)₂(OTf)₂(py), U(Cp*)₂(OTf)₂, and U(Cot)(OTf)₂(py), which have been crystallographically characterised. The homoleptic species U(OTf)_n (n = 3, 4) are easily prepared by heating a mixture of uranium turnings or UH₃ in triflic acid. By adjusting the temperature to 120 or 180 °C, either U(OTf)₃ or U(OTf)₄ is isolated. Treatment of UO₃ with pure or aqueous solution of triflic acid leads to the non-solvated uranyl triflate UO₂(OTf)₂, which is more conveniently obtained by heating a suspension of UO₃ in triflic anhydride. This reactant is an excellent dehydrating agent and enables the preparation of UO₂(OTf)₂ and Ce(OTf)₄ from the hydrated starting materials.     

Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

This work studied a way to reclaim uranium from contaminated UO₂ oxide scraps as a sinterable UO₂ powder for UO₂ fuel pellet fabrication, which included a dissolution of the uranium oxide scraps in a carbonate solution with hydrogen peroxide and a UO₄ precipitation step. Dissolution characteristics of reduced and oxidized uranium oxides were evaluated in a carbonate solution with hydrogen peroxide, and the UO₄ precipitation were confirmed by acidification of uranyl peroxo–carbonate complex solution. An agglomerated UO₄ powder obtained by the dissolution and precipitation of uranium in the carbonate solution could not be pulverized into fine UO₂ powder by the OREOX process, because of submicron-sized individual UO₄ particles forming the agglomerated UO₄ precipitate. The UO₂ powder prepared from the UO₄ precipitate could meet the UO₂ powder specifications for UO₂ fuel pellet fabrication by a series of steps such as dehydration of UO₄ precipitate, reduction, and milling. The sinterability of the reclaimed UO₂ powder for fuel pellet fabrication was improved by adding virgin UO₂ powder in the reclaimed UO₂ powder. A process to reclaim the contaminated uranium scraps as UO₂ fuel powder using a carbonate solution was finally suggested.     

Characteristics of an oxa-diamide–HNO3 extractant in the supercritical fluid extraction of uranium

An extractant is required in the recovery process to drive the uranium to a stage that enables it to be extracted using the extraction solvent. This paper proposes the composition of a composite extractant, N,N,N′,N′-tetrabutyl-3-oxapentane-diamide–HNO₃ (TBODA–HNO₃) as an extractant, to successfully achieve the objective using supercritical carbon dioxide (sc-CO₂). The composite TBODA–HNO₃ extractant has a chemical composition of TBODA(HNO₃)_1.0(H₂O)_1.5. The U(IV) in the UO₂ containing solid phase is directly oxidized to U(VI) in the form of $ {\rm UO}_{2}^{2 + } $ in sc-CO₂, which contains a CO₂-soluble TBODA–HNO₃ extractant at 200 atm and 50 °C. The resulting $ {\rm UO}_{2}^{2 + } $ /TBODA complex can be consequently extracted using acetone-modified sc-CO₂. The chemical composition of the $ {\rm UO}_{2}^{2 + } $ /TBODA complex, which can be extracted by nonpolar sc-CO₂, is proposed in the form of an ion pair: [UO₂(TBODA)₂]²⁺–2( $ {\rm NO}_{3}^{ - } $ ).     

Development of a simple spectrophotometric method to estimate uranium concentration in LiCl–KCl matrix

“Off-the-shelf” clinical linear electron accelerators (LINAC) have been suggested as alternative to research accelerators once they are no longer suitable for the medical applications. We investigated feasibility of utilising a modified LINAC for instrumental photon activation analysis (IPAA) as a tool in environmental geochemistry studies. The IPAA results were compared with those obtained using a MT-25 research accelerator at Joint Institute for Nuclear Research in Dubna, Russia. To investigate soil pollution in Antalya, Turkey, 90 surface soil samples were analysed and significant enrichments of Ni, Cr and As were found.

     

Releases of radium from an abandoned uranium mine site: Žirovski Vrh uranium mine, Slovenia

Since the beginning of explorative uranium mining at the Žirovski Vrh uranium ore deposit area in 1968, a radioactivity monitoring programme has been carried out. The extent of the programme has varied according to the pre-operational, operational, and, finally, post-operational conditions. In this paper, our ten year results on the dissolved radium concentrations in surface waters, which have been contaminated and potentially affected by the uranium mining and milling activities, are reported. With the exception of waters drained from the hydrometallurgic waste site with radium content ranging from 2 to 9 kBqm⁻³, radium content is far below the drinking water limit of 1000 Bqm⁻³; in the Brebovŝčica stream, which collects all the waters affected by the mine, the present radium concentration does not exceed 10 Bqm⁻³.     

Colloidal transport of uranium in soil: Size fractionation and characterization by field-flow fractionation–multi-detection

The aim of this study was to characterize colloids associated with uranium by using an on-line fractionation/multi-detection technique based on asymmetrical flow field-flow fractionation (As-Fl-FFF) hyphenated with UV detector, multi angle laser light scattering (MALLS) and inductively coupling plasma-mass spectrometry (ICP-MS). Moreover, thanks to the As-Fl-FFF, the different colloidal fractions were collected and characterized by a total organic carbon analyzer (TOC). Thus it is possible to determine the nature (organic or inorganic colloids), molar mass, size (gyration and hydrodynamic radii) and quantitative uranium distribution over the whole colloidal phase. In the case of the site studied, two populations are highlighted. The first population corresponds to humic-like substances with a molar mass of (1500 ± 300) g mol⁻¹ and a hydrodynamic diameter of (2.0 ± 0.2) nm. The second one has been identified as a mix of carbonated nanoparticles or clays with organic particles (aggregates and/or coating of the inorganic particles) with a size range hydrodynamic diameter between 30 and 450 nm. Each population is implied in the colloidal transport of uranium: maximum 1% of the uranium content in soil leachate is transported by the colloids in the site studied, according to the depth in the soil. Indeed, humic substances are the main responsible of this transport in sub-surface conditions whereas nanoparticles drive the phenomenon in depth conditions.     

Determination of uranium and thorium in Eskişehir-Beylikahir ore samples by delayed neutron counting technique

The Eskiehir-Beylikahir district has the largest and richest thorium and rare earth elements deposits in Turkey. The uranium and thorium concentrations of samples taken from four different parts of this area have been determined by the delayed neutron counting technique. The results are compared with those of previous analyses by other techniques and found to be in good agreement.     

Application of gamma-ray spectrometry for the assay of uranium in crude UF4—An input material used for producing nuclear grade U-metal at the natural uranium conversion plants

A -spectrometric method has been developed for the assay of uranium in crude UF₄, which is used as a secondary source of input material for producing nuclear grade U-metal at natural uranium conversion plants. The method makes use of a NaI (Tl) detector coupled with a multichannel analyzer. The 1 MeV -ray of²³⁸U is used for calibration. A method for the fabrication of uniform -assay calibration standards is also suggested, based on the results of this investigation. The calibration standards were prepared by soaking the matrix in uranium solution and then drying the whole material. The amount of²³⁸U in the crude UF₄ sample was directly estimated by comparing the areas under the 1 MeV -ray peak of known calibration standards with the corresponding areas of the samples to be measured. 100 g each of the standard and the sample were counted. 5 crude UF₄ samples were analyzed by this method. The uranium contents in these samples were found to be in the range of 12.2–28.7 g. To compare the -ray spectrometry results with a completely independent method, chemical analysis by potentiometry of all the samples was also done. The -spectrometric results were found to agree within ±18% with the chemical analysis results.     

Quantitative extraction of uranium(VI) in aqueous solutions with n-alkylamine intercalates of γ-titanium phosphate

The intercalation compounds Ti(C₃H₇NH₂)(HPO₄)₂·H₂O (18.4 Å) (-TiP/n-Pr) and Ti(C₄H₉NH₂)(HPO₄)₂·H₂O (20.5 Å) (-TiP/n-Bu) have been prepared using -titanium phosphate, Ti(HPO₄)₂·2H₂O (11.6 Å), as precursor. The retention of UO ₂ ²⁺ , in aqueous solutions by -TiP is very low being only a superficial adsorption phenomenon. When the intercalated materials are used, the retention is quantitative until 95% of the cation exchange capacity in the -TiP/n-Pr case (c.e.c.=6.30 mequiv./g), and over 80% for the -TiP/n-Bu compound (c.e.c.=6.04 mequiv./g).