Small Amounts Of Uranium Research Articles

Effect of particle size analysis on the physicochemical characterization of Malaysian monazite

Rare earth elements (REEs) are the elements that have become extremely important to our world of technology owing to their unique physical and chemical properties, and widely used in advanced technology. Monazite is one of the mineral phosphate that contain a high amount of REEs including thorium and a small amount of uranium. In Malaysia, monazite is exist as by-product (‘amang’) from tin mining originated from Ipoh, Perak. The objective of this study is to determine the effects of the different particles size of raw and ground (milled) monazite on the physicochemical properties. The effectiveness of extraction and separation process of monazite was affected by the size of the particle. Based on the particle size distribution analysis, all the monazite samples were found to be present in the size range of 11-190 μm. The morphological and elemental composition of the monazite mineral have been characterized by field emission scanning electron microscope (FESEM) integrated with energy dispersive X-ray spectrometer (EDX). The FESEM micrographs of the monazite showed an irregular spherical structures with different sizes. The chemical composition and the mineral phases present in monazite were studied by wavelength-dispersive X-ray fluorescence spectrometer (WD-XRF) and X-ray diffraction (XRD). The XRF analysis confirmed the presence of the major REEs such as Ce, La and Nd elements in the mineral. XRD analysis also identified (Ce, La, Nd and Th)PO4 as the major phases along with quartz as minor phase. According to the FESEM-EDX and XRF analysis, smaller particle size of monazite gave higher content of REEs compare to raw monazite.

OxGamma: A MATLAB based application for the analysis of gamma-ray spectra

Gamma-ray spectrometry is a well-established technique which enables to perform qualitative and quantitative analyses of gamma-ray emitting radionuclides. In trapped charge dating, portable spectrometers can be used to acquire spectra in the field, allowing for the assessment of radionuclide concentrations and hence the determination of dose rates at the exact location where samples are collected for dating. Accurate assessment of concentrations and dose rates demands proper energy, width, and efficiency calibration of the instrument by comparison against one or several radioactive standards. The ‘Oxford blocks’ which consist of four concrete cubes doped with known concentrations of potassium [K-block], thorium [Th-block], and uranium [U-block], as well as a background [BG-block], were constructed for this purpose and have become widely adopted by researchers for calibrating their instruments. For inexperienced users, the calibration of detectors and the analysis of field spectra for determining the concentration of radionuclides can be a challenging and laborious task that often involves multiple steps (e.g., downloading and converting data files, peak identification and selection, energy calibration, true full-peak area calculation, efficiency calculation, etc.). The processing of calibration spectra from the Oxford blocks is further complicated by the fact that the K-block contains only ∼72% concrete and the Th-block contains a small amount of uranium. These factors must be accounted for in order to obtain accurate radionuclide concentrations and dose rates.Here, we present a practical and user-friendly MATLAB based application (OxGamma), available as a free download for both Windows OS and macOS, intended to facilitate the calibration process of portable gamma-ray detectors and to provide a quick and efficient means for determining the concentration of radionuclides (K, U and Th) and dose rates. The software has been tested and checked against results obtained independently and provides output using both the windows and threshold approaches. Crucially, it does not require users to possess in-depth experience in programming or gamma-spectrometry.

Magnetic assisted separation of uranium(VI) from aqueous phase using diethylenetriamine modified high capacity iron oxide adsorbent

Processing of uranium ore for the production of reactor-grade uranium results in the generation of large amount of aqueous waste containing small quantities of uranium, but higher than the guideline value (15 μg L−1) recommended by World Health Organization (WHO). The presence of uranium in aqueous waste poses several hazards and environmental issues due to the radioactive nature and migration behavior of uranium in the geosphere. In order to remove uranium (VI) from aqueous waste, the diethylenetriamine modified high capacity magnetic iron oxide adsorbent, abbreviated as Fe-DETA, was prepared and studied for the extraction of uranium from aqueous solutions. The Fe-DETA was characterized by X-ray diffraction, thermal analysis, infrared and Raman spectroscopy, and scanning electron microscopy. Significantly high amount (2 mmol g−1) of diethylenetriamine functional group (DETA) was anchored on magnetic iron oxide particles. The extraction of U(VI) from aqueous phase was studied as a function of pH of the aqueous phase, duration of the contact between Fe-DETA and aqueous phase, concentration of U(VI) in aqueous solution etc. The data on the rate of extraction of U(VI) in Fe-DETA was fitted into the first order and second order rate equations. The apparent uranium extraction capacity on Fe-DETA was determined to be 236 mg g−1, which correspond to the formation of 1:2 complex of U(VI) to diethylenetriamine ligand in Fe-DETA phase. The loaded U(VI) in Fe-DETA was quantitatively recovered using dilute Na2CO3 solution, and the recovered Fe-DETA was recycled for further extraction of U(VI), without any change in the apparent U(VI) extraction capacity.

Study on carbonation technology of alkaline uranium mother liquor

The characteristics of the main components of alkaline uranium mother liquor: A small amount of uranium, a certain amount of hydroxide, and a large amount of carbonate. Through the analysis of chemical reaction and diffusion about transfer between mother liquor and CO2, the influencing factors of mother liquor carbonation were determined: Gas-liquid transfer, reactant concentration, reaction temperature, etc. Tower counter-current was selected to investigate the influence of different types of fillers on carbonation effect. The stainless steel triangular spiral filler was optimized. The filling tower device for carbonation of uranium mother liquor was finalized, which greatly improved the gas-liquid distribution efficiency and carbon dioxide absorption rate. The effect of residual alkali concentration and reaction temperature of mother liquor on carbonation effect was investigated. The technology realizes the conversion of sodium hydroxide to sodium carbonate in the mother liquor. When sodium carbonate is converted to sodium bicarbonate, it can be converted to the saturated concentration of sodium bicarbonate, and CO2 is absorbed in both stages. The device can be used to control the degree of carbonation of the mother liquor according to the demand, and be returned as leaching agent or eluent, realized the full reuse of the remaining alkali in the mother liquor.

Phase distribution of uranium in matrices for immobilization of the rare earth–actinide fraction of high-level waste

The uranium distribution in candidate host materials for the rare earth–actinide fraction was studied. The compositions of the materials were set so as to obtain artificial phases with the structures of pyrochlore Ln2(Ti,Zr)2O7 (space group Fd3m, Z = 8) and of monoclinic (P21, Z = 4) and rhombic (Fddd, Z = 16) titanates with the formulas Ln2Ti2O7 and Ln4Ti9O24, respectively, where Ln is a mixture of lanthanides simulating the wastes. The uranium content was varied from 3.5 to 5.5 wt %. In some cases, an equiatomic amount of calcium was introduced into the charge to check the possibility of the replacement of two Ln3+ cations by the Ca2+ + U4+ pair. Samples were prepared by sintering of the oxide charge or by its melting and crystallization of the melt. Along with the prevalent target compounds, the products contained uranium and lanthanide oxides, titanate of Ln and U with brannerite structure, Ln titanosilicates, and other phases. The main uranium concentrators in the samples are pyrochlore, oxides, and brannerite; small amounts of uranium (1–2 wt %) are incorporated in titanosilicates and rhombic titanate. The addition of calcium does not influence the capacity of monoclinic and rhombic titanates for uranium. Based on these data and the results of previous studies, a conclusion is made that rhombic titanate Ln4Ti9O24 and brannerite (Ln,U)Ti2O6 deserve an additional study as candidate host materials for the rare earth–actinide fraction.

Micro-distribution of uranium in bone after contamination: new insight into its mechanism of accumulation into bone tissue.

After internal contamination, uranium rapidly distributes in the body; up to 20% of the initial dose is retained in the skeleton, where it remains for years. Several studies suggest that uranium has a deleterious effect on the bone cell system, but little is known regarding the mechanisms leading to accumulation of uranium in bone tissue. We have performed synchrotron radiation-based micro-X-ray fluorescence (SR μ-XRF) studies to assess the initial distribution of uranium within cortical and trabecular bones in contaminated rats' femurs at the micrometer scale. This sensitive technique with high spatial resolution is the only method available that can be successfully applied, given the small amount of uranium in bone tissue. Uranium was found preferentially located in calcifying zones in exposed rats and rapidly accumulates in the endosteal and periosteal area of femoral metaphyses, in calcifying cartilage and in recently formed bone tissue along trabecular bone. Furthermore, specific localized areas with high accumulation of uranium were observed in regions identified as micro-vessels and on bone trabeculae. These observations are of high importance in the study of the accumulation of uranium in bone tissue, as the generally proposed passive chemical sorption on the surface of the inorganic part (apatite) of bone tissue cannot account for these results. Our study opens original perspectives in the field of exogenous metal bio-mineralization.

Automatic potentiometric determination of uranium in the presence of organic solvents

Redox potentiometric titration of small amounts of uranium in mixed systems aqueous H3PO4 solutionorganic solvent using an automatic titrator was studied. The accuracy of the U determination strongly depends on the kind and concentration of the solvent. In determination of 0.1–0.2 mg of U in mixed phosphoric acid-organic solvent systems using ammonium vanadate as titrant, the determination accuracy decreases in the order acetone ≫ acetonitrile > nitromethane > ethanol. In the presence of acetone, 0.1 mg of U can be determined with high accuracy, which was impossible when using aqueous H3PO4 solutions. With all the organic components tested, the metrological characteristics are improved with an increase in the organic solvent content from 10 to 37 vol %.

Tensioned Metastable Fluid Detectors in Nuclear Security for Active Interrogation of Special Nuclear Materials―Part B

This paper (constituting Part B) addresses active interrogation for detecting Special Nuclear Materials (SN- Ms) and includes description of the transformational Tensioned Metastable Fluid Detector (TMFD) based method for optimal monitoring. One of the greatest difficulties in detection of SNMs by active interrogation is the task of distinguishing between the probing particles and the secondary particles that indicate the presence of SNMs. The TMFD’s selective insensitivity and γ photon blindness features are advantageous for alleviating this problem. The working principle of the TMFD is discussed along with its applications for security. The experimental work to date involving detection of small quantities of uranium with conventional detectors is discussed along with results of fission neutron detection. Statistically significant detection was achieved within 5 minutes of counting to ascertain and measure conclusive evidence for the presence of a 55g sample of uranium containing 235U. Results of simulations of three active detection techniques utilizing a TMFD system are presented. The process for using the TMFD to discriminate active source particles using timing and energy are described. These simulations indicate that it should be possible to utilize the TMFD system for optimal neutron-based interrogation of SNMs.

Evaluation of occupational exposure to naturally occurring radioactive materials in the Iranian ceramics industry

Zircon contains small amounts of uranium, thorium and radium in its crystalline structure. The ceramic industry is one of the major consumers of zirconium compounds that are used as an ingredient at ∼10-20 % by weight in glaze. In this study, seven different ceramic factories have been investigated regarding the presence of radioactive elements with focus on natural radioactivity. The overall objective of this investigation is to provide information regarding the radiation exposure to workers in the ceramic industry due to naturally occurring radioactive materials. This objective is met by collecting existing radiological data specific to glaze production and generating new data from sampling activities. The sampling effort involves the whole process of glaze production. External exposures are monitored using a portable gamma-ray spectrometer and environmental thermoluminescence dosimeters, by placing them for 6 months in some workplaces. Internal routes of exposure (mainly inhalation) are studied using air sampling, and gross alpha and beta counting. Measurement of radon gas and its progeny is performed by continuous radon gas monitors that use pulse ionisation chambers. Natural radioactivity due to the presence of ²³⁸U, ²³²Th and ⁴⁰K in zirconium compounds, glazes and other samples is measured by a gamma-ray spectrometry system with a high-purity germanium detector. The average concentrations of ²³⁸U and ²³²Th observed in the zirconium compounds are >3300 and >550 Bq kg⁻¹, respectively. The specific activities of other samples are much lower than in zirconium compounds. The annual effective dose from external radiation had a mean value of ∼0.13 mSv y⁻¹. Dust sampling revealed the greatest values in the process at the powdering site and hand weighing places. In these plants, the annual average effective dose from inhalation of long-lived airborne radionuclides was 0.226 mSv. ²²²Rn gas concentrations in the glaze production plant and storage warehouse were found to range from 10 to 213 Bq m⁻³. In this study, the estimated annual effective doses to exposed workers were <1 mSv y⁻¹.

An attempt for determining 235U/238U ratio for a trace amount of uranium: Search for an extinct radionuclide 247Cm in the early solar system

In order to search for isotopic anomalies of U caused by the disintegration of 247Cm in meteorite samples, an analytical scheme for determining 235U/238U ratio for a small amount of uranium (less than 1 ng U) was established. The isotopic ratio of U was determined by double-focusing inductively coupled plasma mass spectrometry (DF-ICP-MS). With a great improvement of chemical procedure for purifying the uranium fraction separated from meteorite samples, a variation in 235U/238U ratios was resolved at an order of ±4‰ for 100 ppt or 200 pg U in solution. For applying the procedure in searching an evidence of 247Cm as an extinct radionuclide in the early solar system, a selective chemical dissolution of constituent minerals of meteorites using EDTA and HCl was introduced and a chemical purification scheme of uranium for ICP-MS was established.

Is there a Nuclear Reactor at the Center of the Earth?

In this paper we discuss the Herndon hypothesis that a nuclear reactor is operating at the center of the Earth. Recent experimental evidence shows that some uranium can have partitioned into the core. There is no viable mechanism for the small amount of uranium that is dissolved in the molten metal to crystallize as a separate uranium phase (uranium metal or uranium sulfide) and migrate to the center of the core. There is no need for an extra heat source, as the total heat leaving the core can be easily provided by “classical” heat sources, which are also more than adequate to maintain the Earth’s magnetic field. It is unlikely that nuclear georeactors (fast breeder reactors) are operating at the Earth’s center.

Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with 235U is necessary, and the 235U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO2-UO2) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of 233U from the 232Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the 233U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial 235U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO2 fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.

Uraninite and fullerene in atmospheric particulates.

Particulates emitted from coal-burning power plants typically contain very small amounts of uranium (<10 ppm). Because of the extremely low concentrations, the form of the uranium has been unknown. Using a variety of advanced electron microscopy techniques, we have identified for the first time nanocrystals of uraninite, UO2+x, encapsulated in carbonaceous matter (< or = 50 nm) similar to fullerene. We have also identified, for the first time, closely associated fullerenes, C60. The "carbon-caged" nanocrystals of uraninite are protected from the immediate oxidation that would lead to increased mobility of uranium in the environment. Still, the presence of uranium in the very fine fraction of atmospheric particulates provides another pathway for radiation exposure.

Precipitation of Uranium and Plutonium from Alkaline Salt Solutions

Addition of sodium hydroxide to radioactive waste solutions produced a heterogeneous mixture of solids that exhibit different settling characteristics. Plutonium effectively coprecipitated with iron and uranium, and uranium with iron. Aluminum proved ineffective as a coprecipitating agent for either plutonium or uranium. Coprecipitation of uranium and plutonium occurs when the mole ratio of coprecipitating agent to actinide exceeds 1500. Addition of water to the alkaline slurries that simulate dilution during retrieval and pretreatment of high-level wastes results in dissolution of small amounts of uranium and plutonium in some slurries. The amount of uranium and plutonium dissolved did not saturate the solution in either actinide.

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

Eskisehir-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.

Uranium in phosphate rock and derivates

It is well known that phosphate rock contains a small amount of uranium. During the production of phosphate fertilizers, most uranium present is found in the phosphate fraction. Uranium was determined in the 70–200 ppm range in the phosphate rock and the final products by nuclear and classical methods.

A routine method for the determination of plutonium in urine with a low detection limit

A radiochemical method for the determination of plutonium in urine is described. The steps involved are a) co-precipitation of plutonium, b) wet ashing, c) hydrolysis, d) extraction from 2M HNO3 into capillary polypropylene columns coated with tri-n-octyl phosphineoxide 0.5M in toluene, and e) back-extraction of plutonium from the organic phase, f) electroplating onto stainless steel disks and α spectrometry, since plutonium is extracted together with small amounts of uranium naturally occurring in urine. High quality deposits for spectrometry are obtained because iron interference is eliminated before back-extraction. The radiochemical recovery of239Pu is 55.6±7.5% and the detection limit is 1.0 mBq per liter of urine.

Recovery of uranium, vanadium and fluorine from phosphate ores

A process for the recovery of uranium, vanadium and fluorine as byproducts of phosphoric acid obtained via hydrochloric acid (the cheapest locally produced acid) acidulation of phosphate ores was studied. At a redox potential of 180–200 mV, by introducing iron powder into the leaching circuit at 90°C, a relatively small amount of uranium and vanadium residue was obtained. The residue obtained was roasted with sodium chloride at 850°C to produce water-soluble sodium vanadate. Water leaching then removed the vanadium, leaving a uranium residue. The uranium was selectively leached by sodium carbonate at a redox potential of 480 mV. The product was then processed by alkaline decomposition to prepare the sodium diuranate concentrate necessary for uranium processing. Fluorine was recovered during the acid leaching process as sodium fluorosilicate by adding a stoichiometric amount of rock salt or sodium carbonate to the leaching circuit.

Photometric complex-formation titration of submicromole amounts of uranium

With DTPA present in excess, small amounts of uranium (100–200 μg) can be reduced quantitatively to the quadrivalent state in a Jones reductor (5% amalgamated Zn powder, 50 mesh). The excess of DTPA is back-titrated photometrically with thorium, with Semi-Xylenol Orange as the indicator. The accuracy is ± 0.2 μ g at the 200-μ g level. Metals forming complexes with DTPA at pH 2 interfere.

Head-end scheme for reprocessing of advanced nuclear fuels by direct dissolution in nitric acid

Advanced nuclear fuels-mixed carbides MC, mixed carbonitrides M(C, N), mixed nitrides MN (M = U 0.8Pu 0.2-show promise for use in future fast breeder reactors (FBR). One precondition for their acceptability as fuels is the mastering of the complete fuel cycle, especially the reprocessing step. In principle, the Purex solvent extraction process, developed for LWR fuel, is also applicable to the recovery of uranium and plutonium from mixed oxide FBR fuel. Additional preparative steps in the head-end of the process are needed for advanced fuels, if the reprocessing is based on the controlled oxidation of the fuel to MO 2 with O 2 or CO 2. Another way of preparing the feed to meet the requirements of the Purex process is the direct dissolution of the carbides or nitrides in nitric acid, as described in this paper. Specimens of 5–10 g of MC, M(C, N) or MN with burn-ups between 0 and 6.7 at % were dissolved under reflux in 2–12 M HNO 3 kept at constant temperatures between 30 and 85 °C for up to 24 hours. The evolved gases were continuously measured by infrared absorption (Miran-80) or samples were taken during and after the dissolution and analysed by mass spectrometry. In the case MC- and M(C, N)-dissolutions, the CO 2 formed was trapped in NaOH and the carbonate/bicarbonate titrated with HCl under computer control [1]. The resulting solutions were analysed for uranium and plutonium by titration (unirradiated MC), according to the methods of Eberle et al. [2] and Davies and Townsend [3], respectively. Additionally, spectrophotometric measurements in the range 400–850 nm were made. The remotely controlled spectrometer (Cary 17D and HP 9835A) was used to determine the oxidation states of U and Pu along with the total concentrations of these elements by multicomponent analysis, taking the carbon containing compounds in the solution into account [1]. The organic species, mainly mellitic acid and oxalic acid [4], were also identified by ion chromatography (Dionex D 10) [5]. The solutions were filtered through 1.6 μm fritted glass filters, the residues determined by weighing and characterized by x-ray diffraction analyses. After dissolution of the residue in H 2SO 4 and (NH 4) 2SO 4 the radioactive fission products were determined by γ-spectrometry, the uranium concentration was measured by isotope dilution mass spectrometry and the plutonium concentration by α-spectrometry. Gaseous Phase The off-gas from MC dissolution contains the gases N 2O, NO, NO 2 and CO 2; no hydrocarbons are formed when the nitric acid molarity is above 2 M. About 50% of the carbide carbon is released as CO 2, the remainder forms soluble organic compounds. The evolution of the predominant gas, NO, ceases at the end of the carbide dissolution, whereas the formation of NO 2 and CO 2 continues for many hours, indicating an oxidation of the carbonaceous by-products in the solution. Only small amounts of N 2O are found in the off-gas. Liquid Phase The MC-fuel dissolves in nitric acid producing a solution containing U(VI), Pu(IV) and carbon containing species [1]. Unirradiated or low burn-up fuels dissolve completely (>99.9%), whereas high burn-up fuels (>4 at%) give rise to minor amounts of residues, which can be removed by filtration. The oxalic acid complexes the plutonium and decreases its extractability with tirbutylphosphate (TBP) [6]. The carbon containing species can, however, be destroyed by oxidizing chemicals, such as MnO 2 or BiO 3 − [7], or by photochemical treatment [8]. The nuclear radiation absorbed by the sample solution also affects the organic compounds, so that their deleterious influence on the Pu-extraction disappears after storage. MN-fuels dissolve faster than MC under violent NO x evolution. The solution obtained contains U(VI) and Pu(IV). During M(C, N)-dissolution large amounts (20%) of Pu(VI) were occasionally observed; however Pu(VI) is unstable and transforms to Pu(IV) upon standing. Solid Phase The amounts of residues found by dissolution of high burn-up MC and MN are minimized in 8 M HNO 3 at 80 °C ( e.g. 0.5% and 0.3%, respectively). The MC residues contain only very small amounts of uranium and plutonium ( e.g. 0.04% and 0.3% of the initial fuel amounts, respectively), with a U/Pu ratio in the residue of 0.6, strongly deviating from the U/Pu ratio in the solution (about 3.9). Small amounts of PuO 2 were also found in the residue.