Mixture Of Uranium Research Articles

Improving the reliability of internal dosimetry for uranium workers: Application of the ICRP/OIR uranium model to long-term occupational intakes

In the Occupational Intakes of Radionuclides (OIR) report series, the International Commission on Radiological Protection (ICRP) published updated biokinetic models and dosimetric data associated with the internal exposure of workers, which are consistent with the recommendations of ICRP Publication 103. The present study focused on the application of the new OIR uranium model published in ICRP Publication 137, which is relevant for the interpretation of measurements of activity in the body and in excreta, and for calculating the committed effective dose. The main objective of this study was to assess the impacts of the new OIR biokinetic models and dose coefficients compared with the old ones based on ICRP Publications 60/78/119. CIEMAT and the University of Salamanca in Spain have been working on re-interpreting in vitro bioassay data obtained from workers with long-term exposure to inhaling uranium oxides during the fabrication of nuclear fuel elements using low enriched uranium, which is now classified as OIR type M/S material. The effects on the retention/excretion models and dose coefficients were studied for 234U, 235U, 238U, and the uranium mixture. Committed effective doses were re-assessed using BIOKMOD code by applying the new OIR excretion model and dose coefficients considering acute and/or chronic inhalation intake scenarios. A reduction by roughly a factor of four was obtained compared with former doses based on ICRP Publications 60/78/119, which is an important effect to take into consideration.

Heterogeneous Structure, Mechanisms of Counterion Exchange, and the Spacer Salt Effect in Complex Molten Salt Mixtures Including LaCl3.

Complex molten chloride salt mixtures of uranium, magnesium, and sodium are top candidates for promising nuclear energy technologies to produce electricity based on molten salt reactors. From a local structural perspective, LaCl3 is similar to UCl3 and hence a good proxy to study these complex salt mixtures. As fission products, lanthanide salts and their mixtures are also very important in their own right. This article describes from an experimental and theory perspective how very different the structural roles of MgCl2 and NaCl are in mixtures with LaCl3. We find that, whereas MgCl2 becomes an integral part of multivalent ionic networks, NaCl separates them. In a recent article (J. Am. Chem. Soc. 2022, 144, 21751-21762) we have called the disruptive behavior of NaCl "the spacer salt effect". Because of the heterogeneous nature of these salt mixtures, there are multiple structural motifs in the melt, each with its particular free energetics. Our work identifies and quantifies these; it also elucidates the mechanisms through which Cl- ions exchange between Mg2+-rich and La3+-rich environments.

Deformation–Induced Mechanical Synthesis of U and Fe

The phase composition of metallic α-U and Fe after mechanical synthesis in conditions of severe (mega) plastic deformation at room temperature using rotational Bridgman anvils was studied using Mössbauer spectroscopy, scanning and transmission electron microscopy. It was shown that mechanical synthesis results in U6Fe and UFe2 intermetallic formation with a precursor represented by UFe2(D) and UFe3(D) defective phases and a defective dispersed mechanical mixture of iron and uranium. Low-level annealing at 300 °C results in the ordering of the defective phases and transition of a dispersed mechanical mixture of iron and uranium into U6Fe and UFe2 intermetallics. The diffusion mechanism of intermetallic formation in conditions of cold deformation of iron and uranium mixture was established, and the high deformation and thermal phase stability of intermetallics U6Fe and UFe2 was shown.

Evaluation of neutronic performance for the VVER-1000 reactor core with regenerated uranium-plutonium fuel

The possibility has been considered for the VVER-1000 reactor fuel loading to be formed based on regenerated fuel with the use of the spent fuel accumulated in reactors of the same type. A study was undertaken to investigate the change in the isotopic composition of the plutonium discharged from a thermal reactor in the course of its multiple reprocessing and recycle in a thermal neutron reactor. To obtain an equilibrium isotopic composition of the reactor-grade plutonium, 3D neutronic calculations were performed for the stationary fuel cycles of a VVER-1000 serial reactor with conventional oxide fuel and oxide fuel based on regenerated uranium and based on an undivided mixture of uranium and plutonium oxides from SNF. The neutronic performance of reactor cores was compared for the above mentioned fuel types in the course of the fuel company, including the following: in-core radial power density shaping, values of reactivity coefficients for various thermal parameters, reactivity control system efficiency, etc.

Exposure of Lemna minor (Common Duckweed) to Mixtures of Uranium and Perfluorooctanoic Acid (PFOA).

A variety of processes, both natural and anthropogenic, can have a negative impact on surface waters, which in turn can be detrimental to human and environmental health. Few studies have considered the ecotoxicological impacts of concurrently occurring contaminants, and that is particularly true for mixtures that include contaminants of emerging concern (CEC). Motivated by this knowledge gap, the present study considers the potential ecotoxicity of environmentally relevant contaminants in the representative aquatic plant Lemna minor (common duckweed), a model organism. More specifically, biological effects associated with exposure of L. minor to a ubiquitous radionuclide (uranium [U]) and a fluorinated organic compound (perfluorooctanoic acid [PFOA], considered a CEC), alone and in combination, were monitored under controlled laboratory conditions. Lemna minor was grown for 5 days in small, aerated containers. Each treatment consisted of four replicates with seven plants each. Treatments were 0, 0.3, and 3 ppb PFOA; 0, 0.5, and 5 ppb U; and combinations of these. Plants were observed daily for frond number and signs of chlorosis and necrosis. Other biological endpoints examined at the conclusion of the experiment were chlorophyll content and antioxidant capacity. In single-exposure experiments, a slight stimulatory effect was observed on frond number at 0.3 ppb PFOA, whereas both concentrations of U had a detrimental effect on frond number. In the dual-exposure experiment, the combinations with 5 ppb U also had a detrimental effect on frond number. Results for chlorophyll content and antioxidant capacity were less meaningful, suggesting that environmentally relevant concentrations of PFOA and U have only subtle effects on L. minor growth and health status. Environ Toxicol Chem 2023;42:2412-2421. © 2023 The Authors. Environmental Toxicology and Chemistry published by Wiley Periodicals LLC on behalf of SETAC.

Burn-Up Analysis of Plutonium and Minor Actinide Recycling on Gas Cooled Fast Reactor (GCFR) Using SRAC COREBN

Research on the analysis of recycling, plutonium, and minor actinides in the Gas Cooled Fast Reactor (GFCR) using SRAC COREBN has been done. This research uses a fuel mixture of uranium, plutonium, and minor actinides. The analysis was conducted with computational simulation using the COREBN code, an additional code to SRAC. The purpose of this research is to find out the influence of the addition of plutonium and actinide on the composition of nuclear fuel at the end of reactor operation and the limitation of the value of the multiplication factor (keff at the end of the reactor burn up period). Found in this research is the final value, the multiplication factor (keff) of 1.19964, and the conversion ratio value of 0.766813 in the burn-up period of 1515 days as well as the maximum power density value of 125.85 watts/cm³, the relative power density value at the radius y of 1,230263 and radius x equal to 1.19737 the atomic density value experienced a change in the number of nuclides in the types of nuclides U235, U238, Pu239, Pu241, Np237, and Am243 at the end of the burn-up period.

Diffuse-interface modelling of multicomponent diffusion and phase separation in the U-O-Zr ternary system

The ternary mixture of uranium, oxygen and zirconium is investigated as a minimal model for corium, the mixture that forms after the meltdown of a nuclear reactor. Like corium, U-O-Zr exhibits a liquid–liquid phase separation between a metal-rich and an oxide-rich phase at high temperatures. A CALPHAD database built on an associate model is used to set up a ternary Cahn–Hilliard model, which can describe two-phase patterns in U-O-Zr. The interface structure and properties, which depend on the choice on the gradient energy coefficients in the free-energy functional, are studied in detail. It is found that interface adsorption is generally present due to the diffuse character of the interface, but that its magnitude is small, such that the model remains a robust and useful tool for future simulations of corium pool stratification dynamics.

New measurements of cumulative photofission yields of 239Pu, 235U and 238U with a 17.5 MeV Bremsstrahlung photon beam and progress toward actinide differentiation

In the frame of a long-term research program on the characterization of large radioactive waste packages by photofission, the Nuclear Measurement Laboratory of CEA IRESNE has measured cumulative yields of 239Pu, 235U and 238U photofission products by using a Bremsstrahlung photon beam produced by a 17.5 MeV linear electron accelerator. A characterization of the energy of the Bremsstrahlung photon beam has been carried out by photon activation analysis with different samples of gold, nickel, uranium, zinc and zirconium. The contribution of neutron fission in the different samples has also been estimated by MCNP simulations in order to assess as precisely as possible the photofission yields. Finally, 26 cumulative photofission product yields are reported for 239Pu, 28 for 238U and 26 for 235U, with half-lives ranging from 14 min to more than 3 days, some of them being not recorded so far in the literature. Among these reported photofission product yields, 18 have been measured for all 3 actinides, which can thus be used for their discrimination. A differentiation criterion based on delayed gamma-ray ratios has been established to determine the most efficient photofission product couples to estimate the enrichment of a 235U/238U mixture or the fissile fraction (235U+239Pu)/actinide mass in a mixture of uranium and plutonium.

402 Quantifying Heavy Metals in Interstitial Fluid for Remote Monitoring of Chronic Exposures

OBJECTIVES/GOALS: Our hypothesis is that microneedle array (MA) extraction of interstitial fluid (ISF) will enable minimally invasive quantitation of heavy metal (HM) exposure. We aim to establish analytical parameters for ICP-MS analysis of HMs, quantify baseline HM content in ISF vs other fluids, and characterize a mixed HM exposure model. METHODS/STUDY POPULATION: Ten healthy human volunteers were recruited into the study, approved by the UNM Human Research and Resource Committee Each subject had blood and urine collected. ISF was also collected using 3D-printed MAs inserted into the forearm. Additionally, twelve Sprague Dawley rats were unexposed (n=6) or exposed (n=6) to ad libitum water containing a mixture of uranium (U), cadmium (Cd), vanadium (V), and arsenic (As), each at 5X the maximum contaminant level (MCL) for drinking water under a protocol approved by the UNM animal care and use program. Human and animal fluids were analyzed, using ICP-MS, to quantify the levels of U, Cd, V, and As. RESULTS/ANTICIPATED RESULTS: Recent advances in ISF extraction and analysis suggest a minimally invasive method that can be adapted to monitor HM exposure and biological loads longitudinally in both localized and dispersed communities. ISF can be collected with MAs and is a rich source of disease and exposure biomarkers. However, determining biological loads remains challenging due to the need to collect blood or urine from a dispersed rural population over time. Our preliminary results suggest similar HM concentrations in ISF, compared with blood in small unexposed human and animal populations. All four metals can be successfully quantified in tandem using ICP-MS. DISCUSSION/SIGNIFICANCE: Chronic exposure to heavy metals (HM) is associated with detrimental health effects. Exposure to multiple HMs is suspected to have additive or synergistic harmful effects. We envision a wearable microneedle patch that could be mailed to individuals or distributed through community centers, worn for a few hours, and returned to a central laboratory.

Dissolution of a Mixture of Uranium and Plutonium Compounds

Dissolution of a Mixture of Uranium and Plutonium Compounds

ИЗГОТОВЛЕНИЕ И ИССЛЕДОВАНИЕ ТАБЛЕТОК СМЕШАННЫХ НИТРИДОВ УРАНА, ПЛУТОНИЯ, АМЕРИЦИЯ И НЕПТУНИЯ

The method of carbothermic synthesis produced 260 g tablets of mixed nitrides of uranium, plutonium, americium and neptunium with a mass fraction of americium 0.61%. For the production of tablets were used uranium oxide manufactured by the water method, as well as plutonium dioxide containing impurities of uranium and americium oxides with a mass of 0.9 g americium, obtained by volumetric crystallization method in a NaCl — 2CsCl melt. Neptunium dioxide and americium oxide were added to the mixture of uranium and plutonium oxides before carbothermic synthesis. The resulting tablets had a density of 11.6-11.9 g / cm3. By methods of gamma-spectrometry, scanning electron microscopy and x-ray microanalysis is showed that the heat treatment of the mixture of the source chemicals - oxides of uranium, plutonium, americium, neptunium with carbon black at a temperature 1600-1500 °C for 72 h (24 h in nitrogen and 48 h in nitrogen-hydrogen atmospheres), and also the subsequent sintering synthesized by the press powders for 48 h at a temperature of 1800°C in nitrogen-hydrogen atmosphere doesn't leads in average to the significant losses of americium.

Kinetics of Reaction of Oxygen with Uranium(IV) Chloride in Alkali Chloride Melts

Alkali chloride melts have numerous potential applications in nuclear fuel cycle including pyrochemical reprocessing of spent nuclear fuels, uranium electrowinning and electrorefining. Oxygen is a common technological impurity that can affect uranium speciation and behavior in fused salts. The present work was devoted to studying the reactions of oxygen with solutions of uranium tetrachloride in molten alkali chlorides. The experiments were performed in LiCl–KCl, NaCl–KCl–CsCl and NaCl–CsCl eutectic based melts at 450–750 oC. Pure oxygen and argon–oxygen mixtures (containing ca. 1 and 10 % O2) were used. Amount of oxygen passed through the melt varied from less than one to over 100 moles per mole of uranium present. Effect of moisture (0.4–2.5 % H2O) presence in oxygen or Ar–O2 mixtures was also investigated. The course of the reaction was followed by in situ electronic absorption spectroscopy measurements with the spectra recorded at the certain time intervals.Depending on temperature, cationic melt composition and oxygen-to-uranium molar ration the reaction resulted in oxidation of uranium(IV) to soluble uranyl chloride and/or precipitation of uranium dioxide. Analysis of the spectra provided the information on kinetics of U(IV) concentration change. Increasing temperature, O2 : U(IV) molar ratio or decreasing mean radius of alkali cations of the solvent melt resulted in faster decrease of U(IV) concentration in the melt.Under certain conditions U(IV) can be oxidized to UO2Cl4 2– without precipitation of UO2. Therefore sparging the melt with oxygen can be used as a way of separating uranium from certain fission products, for example rare earth elements. Rare earth chlorides react with oxygen yielding oxychlorides or oxides insoluble in alkali chloride melts. Interaction of oxygen with melts containing a mixture of uranium and rare earth chlorides was therefore also investigated and an example of the spectra recorded in LiCl–KCl–UCl4–NdCl3 melt is shown in Fig.Fig. Spectra recorded in the course of reaction of O2 with LiCl–KCl–UCl4–NdCl3 melt at 550 oC. Arrows show the direction spectra changed. Figure 1

Kinetics of Reaction of Oxygen with Uranium(IV) Chloride in Alkali Chloride Melts

Reaction of oxygen with solutions of uranium tetrachloride in molten alkali chlorides was studied. The experiments were performed in LiCl–KCl, NaCl–KCl–CsCl, and NaCl–CsCl eutectic based melts at 450–750oC. Pure oxygen and argon–oxygen mixtures (containing ca. 1 and 10% O2) were used. Amount of oxygen passed through the melt varied from less than one to over 100 moles per mole of uranium present. In addition the reaction of oxygen with the melts containing a mixture of uranium(IV) and neodymium chlorides was investigated. The course of the reactions was followed by in situ electronic absorption spectroscopy measurements with the spectra recorded at the certain time intervals. Depending on temperature, cationic melt composition, and oxygen-to-uranium molar ratio, the reaction resulted in oxidation of uranium(IV) to soluble uranyl chloride and/or precipitation of uranium oxide. Analysis of the spectra provided the information on kinetics of U(IV) concentration change and the reaction rates were determined. Increasing temperature, O2 : U(IV) molar ratio or decreasing mean radius of alkali cations of the solvent melt resulted in faster decrease of U(IV) concentration in the melt.

Development of a Technology for Dissolving a Homogeneous Mixture of Uranium and Plutonium Oxides Obtained by the Plasma-Chemical Method

A plasma technology for the synthesis of mixed uranium and plutonium oxides UO2–PuO2 by denitration of mixed solutions of these metals has been developed at the Siberian Chemical Combine. The production process of U and Pu oxides includes the preparation of mixed initial solutions, their denitration in a plasma reactor with an inductive high-frequency discharge, and isolation of powders from the vapor–gas flow in vortex precipitators or filters with metal–fabric meshes. The particles of the obtained oxide powders possess the features of nanocrystalline and nanoscale materials containing crystalline and amorphous phases in the structure. The equipment for the implementation of this technology has been developed (Zenith plasma plant), on which mixed U and Pu oxides have been produced, and over 60 kg of oxides UO2–PuO2 with the plutonium content of 25 wt % has been synthesized, from which industrial fuel rods for the BN-600 reactor have been fabricated. After the operation of the fuel rods in the BN-600 reactor and corresponding exposure, the cores of the irradiated fuel rods should be dissolved for subsequent radiochemical treatment. The results of the studies on the search for the conditions of dissolution and effective solvents for mixed U and Pu oxides obtained by the plasma technology of decomposition of mixed solutions of uranium and plutonium are presented. It has been found that no complete dissolution of the synthesized oxides occurs in solutions of HNO3 and HNO3 with the addition of the reducing agents for tetravalent plutonium—ascorbic acid and hydrazine. Complete dissolution of the mixed oxides most effectively occurs in solutions of 10 M HNO3 with the addition of 0.25–1.0 g/L HF.

Selective Separation of Neptunium from an Acidic Feed Containing a Mixture of Actinides Using Dialkylamides

ABSTRACTExtraction of Np(IV), Pu(IV), U(VI), and Np(VI) was studied using N,N-di-2-ethylhexyl derivatives iso-butyramide (D2EHIBA) and n-propanamide (D2EHPrA). D2EHIBA showed higher selectivity for U(VI) over Np(IV) and Pu(IV). A flowsheet for the separation of Np(IV) from a mixture of uranium, neptunium, and plutonium in 3 M HNO3 was developed exploiting the selectivity of D2EHIBA for U(VI) over the tetravalent actinides followed by the selective stripping of Np(IV) from the di-n-hexyl-octanamide (DHOA) medium using acetohydroxamic acid (AHA). The results indicated that >90% of the neptunium present in the mixture could be recovered at the final step with a decontamination factor >103.

Effect of Electrode Materials on electro kinetic remediation of Uranium Contaminated Soil

In this study, three different electrode materials (graphite, stainless steel, and titanium rods) were used to explore the electro kinetic (EK) remediation of soil contaminated with uranium (U). All EK experiments were carried out in a soil cell reactor with a moisture content of 25.1% and a total uranium mixture of 90.6 mg/kg at a constant voltage gradient (1V/cm) for 5 days. The experimental results show that the EK repair process with graphite rod as the electrode material has the highest current peak, and the overall current level is higher than the other two electrode materials. Compared with stainless steel and titanium rods, the removal efficiency of using graphite as the electrode material U in contaminated soil reached 34.11%. However, the removal rates using stainless steel and titanium rods were lower, 15.79% and 18.77%, respectively. In the EK experiment, the main removal is from the exchangeable part of U to the carbon-bound U composition, while the remaining forms of uranium are difficult to remove under this condition, so in order to be more efficient, the uranium in the soil is removed with maximum reduction. It is necessary to use it in conjunction with other enhancement technologies.

The effects of U-232 on enrichment and material attractiveness over time

The effect of 232U on enrichment and material attractiveness provides a significant firewall against proliferation. Mixing 232U with natural uranium creates major obstacles to the weaponization process. The decay heat of 232U is significant and increases over time by a factor greater than 5 through the decay of 232U daughter products. The dose rate also becomes significant as 208Tl builds up and emits an energetic gamma ray of 2.615 MeV. These factors, along with the bare critical mass and spontaneous fission neutron rate are used to determine the material’s attractiveness through a Figure of Merit analysis. Considering 232U is lighter than 235U, attempting to enrich a mixture of natural uranium and 232U to weapons grade will also increase the concentration of 232U and thus increase the material’s unattractiveness. This study simulates natural uranium with a 232U concentration of 0.03 at.% being enriched to weapons grade and determines the attractiveness of the resulting uranium vector. The results show that the material becomes unattractive within a year of initial separation. In addition, the alpha decay of 232U in gaseous UF6 will result in the decomposition of UF6 molecules. The rate of this decomposition in the initial natural uranium mixture, the mixture enriched to 3 at.% 235U, and this mixture enriched to 90 at.% 235U is calculated. These rates are notable and may pose a significant firewall to obtaining weapons grade uranium.

Neutronics analysis of JSI TRIGA Mark II reactor benchmark experiments with SuperMC3.3

Abstract Jozef Stefan Institute (JSI), TRIGA Mark II reactor employs the homogeneous mixture of uranium and zirconium hydride fuel type. Since its upgrade, a series of fresh fuel steady state experimental benchmarks have been conducted. The benchmark results have provided data for testing computational neutronics codes which are important for reactor design and safety analysis. In this work, we investigated the JSI TRIGA Mark II reactor neutronics characteristics: the effective multiplication factor and two safety parameters, namely the control rod worth and the fuel temperature reactivity coefficient using SuperMC. The modeling and real-time cross section generation methods of SuperMC were evaluated in the investigation. The calculation analysis indicated the following: the effective multiplication factor was influenced by the different cross section data libraries; the control rod worth evaluation was better with Monte Carlo codes; the experimental fuel temperature reactivity coefficient was smaller than calculated results due to change in water temperature. All the results were in good agreement with the experimental values. Hence, SuperMC could be used for the designing and benchmarking of other TRIGA Mark II reactors.

INTERCOMPARISON ON INTERNAL DOSE ASSESSMENT FOR NUCLEAR POWER PLANTS IN KOREA.

The intercomparison test is a quality assurance activity performed for internal dose assessment. In Korea, the intercomparison test on internal dose assessment was carried out for nuclear facilities in May 2018. The test involved four nuclear facilities in Korea, and seven exposure scenarios were applied. These scenarios cover the intake of 131I, a uranium mixture, 60Co and tritium under various conditions. This paper only reviews the participant results of three scenarios pertinent to the operation of nuclear power plants and adopts the statistical evaluation method, used in international intercomparison tests, to determine the significance values of the results. Although no outliers were established in the test, improvements in the internal dose assessment procedure were derived. These included the selection of intake time, selection of lung absorption type according to the chemical form and consideration of the contribution of previous intake.

Conceptual potential of a pyroelectrochemical technology for the thorium engagement in the fast neutron fuel cycle

The use of thorium in combination with plutonium in nuclear power generation offers a solution to the problem of reducing the accumulation of long-lived transplutonium nuclides. Along with this, the existing uranium fuel cycle (UFC) has such disadvantage as the vulnerability to unauthorized use of nuclear materials. The thorium fuel cycle (TFC) is devoid of these drawbacks. The engagement of thorium in nuclear power is possible provided the availability of an appropriate technology for reprocessing irradiated thorium. A fuel cycle based on thorium oxide may not differ in principle from the already developed pyrochemical fuel cycle involving uranium and plutonium oxides. Thorium oxide is most commonly obtained in compact state by electrolysis of molten salts from thorium-containing electrolytes. The most thorough studies of physical and chemical and electrochemical behavior of thorium in molten haloids of alkali and alkaline-earth metals were conducted in the 1960ies and the 1970ies. Since extensive experimental material has been accumulated by now for justification of the use of pyroelectrochemical and chemical processes for regeneration of fuel in molten salts, then it has also been proposed that technologies for fuel reprocessing in molten chlorides of alkali metals should be applied resulting in a crystalline product that can be used for the fuel element fabrication. Unlike uranium and plutonium, thorium behavior in molten salt environments is less complex. In molten salts, thorium exists predominantly in the form of Th4+, and the mixture of uranium and thorium dioxides with ThO2 content reaching up to 50 % can be obtained by electrolysis of molten salts. Therefore, the existing amount of knowledge about the chemistry of thorium allows regarding the use of pyrochemical processes in production of thorium oxide as highly promising, and the available data on the physical and chemical properties of thorium and its compounds in high-temperature molten salts makes it possible to state that the pyroelectrochemical technology can be potentially used in production and reprocessing of thorium fuel.