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.     

Electrochemical study on various types of anode materials in LiCl–KCl eutectic salt used in the electro-winning process

The electrochemical behaviors of the molybdenum, graphite, and glassy carbon anode have been observed during the electrolysis of uranium chloride in an UCl₃–LiCl–KCl molten salt. The cyclic voltammogram data of the molybdenum anode showed that Mo anode could be corroded by the formation of Mo compounds or its dissolution. However, the experimental data of the graphite and the glassy carbon anode showed that the chlorine evolution started from 1.2 V and the anode potential remained stable during uranium electrolysis. Therefore, the graphite and glassy carbon may be used as an inert anode in the electro-winning process of pyroprocessing.     

Carbon dots and polyurethane composite for photo-induced elimination of uranium under air atmosphere

Uranium is the main fuel of nuclear power and elimination uranium from nuclear wastewater is significant both in environmental protection and fuel recycle. Here we report for the first time the synthesis of carbon dots/polyurethane(CDs/PU) composite materials for the photoinduced elimination of uranium from water. Irradiated with visible light, CDs/PU could eliminate uranium efficiently with the generation of(UO₂)O₂·2H₂O as solid products in air. The further inve...     

<Superscript>99m</Superscript>Tc(CO)<Subscript>3</Subscript>-tosufloxacin dithiocarbamate complexation and radiobiological evaluation in male Wister rat model

The uranium ingot casting process is one of the steps which consolidate uranium deposits produced by electrorefiner in an ingot form in a pryprocessing technique. Since molten uranium metal reacts with a graphite crucible when the uranium is being dissolved, a graphite crucible cannot be used. Accordingly, a ceramic material must be selected which does not react with the dissolving uranium and this must be used as a coating material on the graphite crucible surface. As to this research, a reactivity experiments were performed between the coating layer and uranium by applying a thermal spray coating to the graphite material with alumina and YSZ ceramic material. As shown in the experimental result, the YSZ coating layer showed a stronger adhesive property on the side where there is no Ni–Al binding material. Moreover, no reaction was apparent between the YSZ coating layer and the uranium. Accordingly, the YSZ material and the process of thermal spray coating are considered to solve the reactive problem between uranium and a graphite crucible.     

Effects of current densities on the formation of LiCoO<Subscript>2</Subscript>/graphite lithium ion battery

The activation characteristics and the effects of current densities on the formation of a separate LiCoO₂ and graphite electrode were investigated and the behavior also was compared with that of the full LiCoO₂/graphite batteries using various electrochemical techniques. The results showed that the formation current densities obviously influenced the electrochemical impedance spectrum of Li/graphite, LiCoO₂/Li, and LiCoO₂/graphite cells. The electrolyte was reduced on the surface of graphite anode between 2.5 and 3.6 V to form a preliminary solid electrolyte interphase (SEI) film of anode during the formation of the LiCoO₂/graphite batteries. The electrolyte was oxidized from 3.95 V vs Li⁺/Li on the surface of LiCoO₂ to form a SEI film of cathode. A highly conducting SEI film could be formed gradually on the surface of graphite anode, whereas the SEI film of LiCoO₂ cathode had high resistance. The LiCoO₂ cathode could be activated completely at the first cycle, while the activation of the graphite anode needed several cycles. The columbic efficiency of the first cycle increased, but that of the second decreased with the increase in the formation current of LiCoO₂/graphite batteries. The formation current influenced the cycling performance of batteries, especially the high-temperature cycling performance. Therefore, the batteries should be activated with proper current densities to ensure an excellent formation of SEI film on the anode surface.     

Electrochemical-Mediated Regenerable FeII Active Sites for Efficient Uranium Extraction at Ultra-Low Cell Voltage

Nano-reduced iron (NRI) is a promising uranium adsorbent due to its strong reducibility and good selectivity, but it still faces the challenges of slow kinetics, limited and non-renewable active sites. In this work, we realized high efficiency uranium extraction under ultra-low cell voltage (−0.1 V) in seawater with 20 ppm UO₂(NO₃)₂ solution by coupling electrochemical mediated Fe^II/Fe^III redox and uranium extraction. The adsorption capacity and extraction efficiency of NRI after electrochemical uranium extraction (EUE) could reach 452 mg/g and 99.1 %, respectively. Combined with quasi-operando/operando characterization technologies, we clarified the mechanism of EUE and revealed that continuously regenerating Fe^II active sites by electroreduction could significantly enhance the property of EUE. This work here provides a new electrochemical mediated and low energy consumption uranium extraction strategy which also provides a reference for other metal resource recovery.     

Determination of trace impurities in electronic grade quartz: Comparison of inductively coupled plasma mass spectroscopy with other analytical techniques

An analytical procedure for the determination of uranium and thorium in the sub-ng/g range as well as of other trace elements in the ng/g to g/g range in high purity quartz samples is described. The results obtained by inductively coupled plasma mass spectroscopy (ICP-MS) are compared to those obtained by other analytical techniques (instrumental neutron activation analysis, INAA; flame atomic absorption spectrometry, AAS; Zeeman graphite furnace atomic absorption spectrometry, ZGFAAS; total reflection X-ray fluorescence analysis, TRFA; direct current arc optical emission spectrometry, DC-arc OES; and X-ray fluorescence analysis, XRFA). For the ICP-MS measurements, the decomposition of the samples is carried out with HF/HNO₃/H₂SO₄-mixtures. The results obtained by the different methods show reasonable agreement. For uranium and thorium, ICP-MS proves to be the most sensitive method: detection limits of about 50 pg/g can be achieved for both elements.Presented in part at the 1989 European Winter Conference on Plasma Spectrochemistry, Reutte, Austria     

Isolation of a Perfectly Linear Uranium(II) Metallocene

Reduction of the uranium(III) metallocene [(η⁵‐C₅ⁱPr₅)₂UI] ( 1 ) with potassium graphite produces the “second‐generation” uranocene [(η⁵‐C₅ⁱPr₅)₂U] ( 2 ), which contains uranium in the formal divalent oxidation state. The geometry of 2 is that of a perfectly linear bis(cyclopentadienyl) sandwich complex, with the ground‐state valence electron configuration of uranium(II) revealed by electronic spectroscopy and density functional theory to be 5f³ 6d¹. Appreciable covalent contributions to the metal‐ligand bonds were determined from a computational study of 2 , including participation from the uranium 5f and 6d orbitals. Whereas three unpaired electrons in 2 occupy orbitals with essentially pure 5f character, the fourth electron resides in an orbital defined by strong 7s‐6d mixing.     

Surface reactivity of uranium hexafluoride (UF6)

The present article reviews a selection of results obtained in the AREVA/CNRS/UCA joint research laboratory. It focuses on interfaces formed by uranium hexafluoride (UF₆) with chemical filter (purification), carbon (UF₆ storage), and metallic substrate (corrosion). As a matter of fact, along the nuclear fuel cycle, metallic surfaces of the fluorination reactors, cooling systems (for the liquefaction of UF₆), and storage containers are in contact with UF₆, either in the gas or in the liquid phase. For the removal of volatile impurities before the enrichment, surface of chemical filters with a high specific surface area must be enhanced for both selectivity and efficiency. To store depleted UF₆ (²³⁸U), graphite intercalation compounds are proposed and preliminary results are presented.     

Solvent extraction of uranium from lean grade acidic sulfate leach liquor with alamine 336 reagent

This paper describes the solvent extraction studies carried out on an acidic low assay uranium bearing leach liquor generated during sulfuric acid leaching of a refractory uranium ore using alamine 336?Cisodecenol?Ckerosene reagent combine. The leach liquor has a U₃O₈ content of about 270?mg/L, free acidity 2.4?N H₂SO₄ and total dissolved solids concentration of 260?g/L. Process parameteric variation studies indicated strong influence of free acidity of the leach liquor, alamine 336 concentration and aqueous to organic phase ratio on the extraction efficiency of uranium. An extraction efficiency of about 95% was achieved when the free acidity of leach liquor was 1?N H₂SO₄ or lower, using 2% (v/v) alamine 336 at ambient temperature with an aqueous to organic phase ratio of 1:1. The loading capacity under these conditions was 1.2?g/L of U₃O₈. About 98% of the uranium values could be stripped from the loaded organic using 1?N NaCl in 0.2?N H₂SO₄. The solvent extraction studies aided in developing a suitable process flowsheet for treating refractory uranium ores which need high acidity during leaching and relatively lower acidity for purification by solvent extraction.     

Separation studies of uranium and thorium using tetra(2-ethylhexyl) diglycolamide (TEHDGA) as an extractant

The extraction behavior of uranium, thorium and nitric acid has been investigated for the TEHDGA/isodecyl alcohol/n-dodecane solvent system. Conditional acid uptake constant (K _H) of TEHDGA/n-dodecane and the ratio of TEHDGA to nitric acid were obtained as 1.72 and 1:0.96, respectively. The extracted species of uranium and thorium in the organic phase were found to be UO₂(NO₃)₂·2TEHDGA and Th(NO₃)₄·2TEHDGA. A workable separation factor (D _Th/D _U) of the order of 300 was observed between thorium and uranium in the nitric acid range of 0.5M to 1.5M. Similar separation factor was also achieved at higher acidity when thorium was present in large concentration compared to uranium. These results indicate that TEHDGA solvent system could be a potential candidate for separation of thorium from uranium.     

低浓缩铀靶辐照后溶液中铀的化学种态及主要裂变元素的影响

利用化学种态分析软件CHEMSPEC计算了低浓缩铀靶辐照后溶液中铀(U)的化学种态分布及其主要裂变元素对U化学种态的影响。结果表明,在单组分体系中,pH值和铀酰浓度都会显著影响U的化学种态分布。随着铀酰浓度的增大,溶液中将会生成多核配合物;在较高的NO₃^-浓度下,U在溶液中主要以UO₂²⁺和UO₂NO₃⁺的形式存在。CO₂对不同浓度铀的种态分布影响结果表明,当铀酰浓度较低时,铀的化学种态多以碳酸铀酰的形式存在;当铀酰浓度较高时,铀的化学种态多以氢氧铀酰或柱铀矿沉淀的形式存在。计算发现,当裂片元素Tc、I、Mo的浓度小于0.01 mol·L^-1并分别以TcO₄^-、I^-、MoO₄^2-的种态存在时,这些裂片元素不改变铀的各化学种态的分布。     

低浓缩铀靶辐照后溶液中铀的化学种态及主要裂变元素的影响

利用化学种态分析软件CHEMSPEC计算了低浓缩铀靶辐照后溶液中铀(U)的化学种态分布及其主要裂变元素对U化学种态的影响。结果表明,在单组分体系中,pH值和铀酰浓度都会显著影响U的化学种态分布。随着铀酰浓度的增大,溶液中将会生成多核配合物;在较高的NO₃^-浓度下,U在溶液中主要以UO₂²⁺和UO₂NO³⁺的形式存在。CO₂对不同浓度铀的种态分布影响结果表明,当铀酰浓度较低时,铀的化学种态多以碳酸铀酰的形式存在;当铀酰浓度较高时,铀的化学种态多以氢氧铀酰或柱铀矿沉淀的形式存在。计算发现,当裂片元素Tc、I、Mo的浓度小于0.01mol·L^-1并分别以TcO₄^-、I^-、MoO₄^2-的种态存在时,这些裂片元素不改变铀的各化学种态的分布。     

Measurement of production date (age) of nanogram amount of uranium

JRC-Karlsruhe obtained a swipe sample from a highly enriched uranium seizure, which had taken place in 2011. Due to the very low amount of uranium (nanograms) a new method needed to be developed to determine the U production date (age). The particles on the swipe were collected on a pyrolytic graphite planchet using a vacuum impactor and they were subsequently leached with ccHNO₃. The “bulk” U isotopic composition (²³⁵U: 72.51?±?0.03 wt%) and the production date (December 1992?±?1 year) determined by MC-ICP-MS indicated that the material showed similarity with two other HEU cases seized earlier in Europe.

     

Comparison of direct kinetic phosphorescence analysis and recovery corrected kinetic phosphorescence analysis for the determination of natural uranium in human tissues

Summary In the analysis of biological samples with sub ng/g uranium concentrations, pre-concentration has been shown to improve the detection limit for the determination of uranium. Recovery corrected kinetic phosphorescence analysis (KPA) combines pre-concentration and separation of uranium by anion-exchange from human tissues dissolved in 6M HCl, with the radiochemical yield determined by alpha-spectrometry, using ²³²U as a tracer. Total uranium is determined by KPA after correction for chemical recovery. Twenty-one randomly selected dissolved tissue samples from the United States Transuranium and Uranium Registries (USTUR) Case 0242 were chosen for comparative analyses. The set of samples included dissolved bone and soft tissues. Uranium concentrations for seven of the samples had not been previously reported. Direct KPA could not be used to determine uranium concentrations of five unreported tissues. Three of these tissues had uranium concentrations at or below the KPA L_Q value of 0.028 ng/ml and two tissues had known matrix interferences. All seven of the unreported tissues were successfully analyzed by recovery corrected KPA; concentrations ranged from 9 to 1380 ng per tissue, including those that could not be analyzed by direct KPA due to matrix problems. Recovery corrected KPA gives results similar to direct KPA where matrix interferences and low detection limits are not encountered. A comparison of the direct method of KPA versus recovery corrected KPA shows marked improvement for the determination of uranium in samples that heretofore either uranium was not detected or the sample had to be drastically diluted to minimize matrix effects in order to measure uranium.     

Preparing graphene oxide–copper composite material from spent lithium ion batteries and catalytic performance analysis

The wide use of lithium ion batteries (LIBs) has created much waste, which has become a global issue. It is vital to recycle waste LIBs considering their environmental risks and resource characteristics. Anode graphite from spent LIBs still possess a complete layer structure and contain some oxygen-containing groups between layers, which can be reused to prepare high value-added products. Given the intrinsic defect structure of anode graphite, copper foils in LIB anode electrodes, and excellent properties of graphene, graphene oxide–copper composite material was prepared in this work. Anode graphite was firstly purified to remove organic impurities by calcination and remove lithium. Purified graphite was used to prepare graphene oxide–copper composite material after oxidation to graphite oxide, ultrasonic exfoliation to graphene oxide (GO), and Cu²⁺ adsorption. Compared with natural graphite, preparing graphite oxide using anode graphite consumed 40% less concentrated H₂SO₄ and 28.6% less KMnO₄. Cu²⁺ was well adsorbed by 1.0 mg L^?1 stable GO suspension at pH 5.3 for 120 min. Graphene oxide–copper composite material could be successfully obtained after 6 h absorption, 3 h bonding between GO and Cu²⁺ with 3/100 of GO/CuSO₄ mass ratio. Compared to CuO, graphene oxide–copper composite material had better catalytic photodegradation performance on methylene blue, and the electric field further improved the photodegradation efficiency of the composite material.     

Separation of Uranium Mixing with its Decayed Nuclides and its Activated Nuclides

Extractions of uranium from the acidic media, HCl and MNO₃ at different concentrations, by the tertiary amines, trilaurylamine, triisooctyllamine and trioctylamine were systematically investigated. In experiments, ²³²U was used as the tracer and aliquots of both phases after extractions were taken, ashed at 500°C, placed onto stainless steel planchets, and measured by a surface-barrier Si(Li) detector connected to an alpha spectrometer. One unique condition found useful for complete separation of uranium was that of the radioactive uranium present in 8 M HCl being extracted with 10% trilaurylamine in xylene. In that case, greater than 95% of the uranium could be extracted into the organic solution, whereas all of the radionuclides of a series of its decayed products, ²²⁸h, ²²⁴Ra, ²²⁰Rn, ²¹⁶Po and ²¹²Bi remained completely in the aqueous solution. The same procedure could also be used for the separation of uranium from a mixture with its activated nuclides. Plutonium-239 and Am were used as the tracers representing the activated nuclides of uranium. It was found that ²⁴¹Am was absolutely not extracted, but that ²³⁹Pu could be extracted with ≥95% efficiency from the medium of 8 M HCl into 10% trilaurylamine in xylene. However, ²³⁹Pu could be easily stripped using the solution mixture of 8 M HCl and 0.05 M NH₄I and excluded from the organic solution.     

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.     

Electrosorption of uranium ions on activated carbon fibers

A study on the electrosorption of uranium (U(VI)) ions onto a porous activated carbon fiber was performed to treat lagoon sludge containing 100 mg/L uranium and high concentration of chemical salts composed 3.8% NaNO₃, 19.8% NH₄NO₃, 1.9% Ca(NO₃)₂. The applied negative potential increased the adsorption kinetics and capacity in comparison to the open-circuit potential adsorption for uranium ions. When applying potential at −0.9 V (vs. Ag/AgCl) and pH 4, above 99% of the uranium is selectively removed from the 100 mg/L influent by electrosorption, and the cumulative amount of uranium for 50 h is about 600 mg_uranium/g_ACF. The high selectivity of elctrosorption process for uranium was probably caused by the difference of charge density of cations. More than 99% of adsorbed uranium ions was desorbed at a potential of +1.2 V and pH 3. The electrosorption of uranium onto the porous activated carbon fiber electrode is due to an ion exchange type reaction between the uranium ions and surface acid groups on carbon surface. Cyclic electrosorption test consisting of adsorption and desorption step shows that the activated carbon fiber electrode is easily regenerated in situ, indicating it is a reversible process.     

Diffusive gradients in thin films technique for uranium measurements in river water

The diffusive gradients in thin films technique (DGT) was used for uranium measurements in water. DGT devices with Dowex resin binding phase (Dow DGT) were tested in synthetic river water, which gave 84% response to total uranium concentration. The devices were also deployed in natural river water and compared to devices with other types of binding phases, Chelex 100 resin beads imbedded in polyacrylamide hydrogel (Chelex DGT) and DE 81 anion exchange membrane (DE DGT), deployed in the same location at the same time. The measurement by Dow DGT was the lowest among the different types of the DGT devices, 45% of total uranium, while measurement by DE DGT was the highest, 98% of total uranium. The results achieved by the three types of DGT devices were explained by three DGT working mechanisms, equilibrium between complexes of resin/uranyl carbonates and complexes of resin/competitive ligands in water, effective reduction of uranyl carbonate concentration by the binding phase and dissociation of UO₂(CO₃)₂²⁻ and UO₂(CO₃)₃⁴⁻ within the diffusive layer in a DGT device. It is hoped that by deploying the DGT devices with different binding phases in natural waters, additional information on uranium speciation could be obtained.