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Sample records for uranium oxide powders

  1. Metallization of uranium oxide powders by lithium reduction

    International Nuclear Information System (INIS)

    Kim, I. S.; Seo, J. S.; Oh, S. C.; Hong, S. S.; Lee, W. K.

    2002-01-01

    Laboratory scale experiments on the reduction of uranium oxide powders into metal by lithium were performed in order to determine the equipment setup and optimum operation conditions. The method of filtration using the porous magnesia filter was introduced to recover uranium metal powders produced. Based on the laboratory scale experimental results, mock-up scale (20 kg U/batch) metallizer was designed and made. The applicability to the metallization process was estimated with respect to the thermal stability of the porous magnesia filter in the high temperature molten salt, the filtration of the fine uranium metal powders, and the operability of the equipment

  2. Production and analysis of ultradispersed uranium oxide powders

    Science.gov (United States)

    Zajogin, A. P.; Komyak, A. I.; Umreiko, D. S.; Umreiko, S. D.

    2010-05-01

    Spectroscopic studies are made of the laser plasma formed near the surface of a porous body containing nanoquantities of uranium compounds which is irradiated by two successive laser pulses. The feasibility of using laser chemical methods for obtaining nanoclusters of uranium oxide particles in the volume of a porous body and the simultaneous possibility of determining the uranium content with good sensitivity are demonstrated. The thermochemical and spectral characteristics of the analogs of their compounds with chlorine are determined and studied. The possibility of producing uranium dioxides under ordinary conditions and their analysis in the reaction products is demonstrated.

  3. Development of Pneumatic Transport System (PTS) for safe handling of uranium oxide powder in UMP/UED

    International Nuclear Information System (INIS)

    Manna, S.; Satpati, S.K.; Roy, S.B.

    2009-01-01

    Tonnage quantity radioactive uranium oxide powder of particle size sub micron to 100 micron is handled in Uranium Metal Plant (UMP), UED/BARC for production of nuclear grade uranium metal, required for fuelling research reactors - Dhruva and Cirus. A Pneumatic Transfer System (PTS) using vacuum has been introduced and is being used for handling radioactive powder to improve radiation protection

  4. Synthesis of Uranium nitride powders using metal uranium powders

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Kim, Dong Joo; Oh, Jang Soo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik

    2012-01-01

    Uranium nitride (UN) is a potential fuel material for advanced nuclear reactors because of their high fuel density, high thermal conductivity, high melting temperature, and considerable breeding capability in LWRs. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. The carbothermic reduction has an advantage in the production of fine powders. However it has many drawbacks such as an inevitable engagement of impurities, process burden, and difficulties in reusing of expensive N 15 gas. Manufacturing concerns issued in the carbothermic reduction process can be solved by changing the starting materials from oxide powder to metals. However, in nitriding process of metal, it is difficult to obtain fine nitride powders because metal uranium is usually fabricated in the form of bulk ingots. In this study, a simple reaction method was tested to fabricate uranium nitride powders directly from uranium metal powders. We fabricated uranium metal spherical powder and flake using a centrifugal atomization method. The nitride powders were obtained by thermal treating those metal particles under nitrogen containing gas. We investigated the phase and morphology evolutions of powders during the nitriding process. A phase analysis of nitride powders was also a part of the present work

  5. Development of a pneumatic transport system for bulk transfer of metal grade uranium oxide powder

    International Nuclear Information System (INIS)

    Manna, S.; Satpati, S.K.; Roy, S.B.

    2010-01-01

    Uranium oxide powder is a commonly handled ceramic powder in nuclear industries. Design of the powder transfer system is an important aspect because of some of its typical characteristics. Pneumatic transport system has been widely used in transferring powder from one place to another. A pneumatic transport system using vacuum has been presented in the paper. This is used for bulk transfer of UO 3 powder. The system consists of a cyclone separator and filter cloth at the top of the cyclone separator. The pneumatic transfer system provides high efficiency with sustainable performance and it is a compact, robust, handy and moveable unit. No degradation of the powder quality has been observed during transfer. The system provides highly efficient, easy and safe transfer of radioactive powder, better working environment for the operator. (author)

  6. Pyrophoric behaviour of uranium hydride and uranium powders

    Science.gov (United States)

    Le Guyadec, F.; Génin, X.; Bayle, J. P.; Dugne, O.; Duhart-Barone, A.; Ablitzer, C.

    2010-01-01

    Thermal stability and spontaneous ignition conditions of uranium hydride and uranium metal fine powders have been studied and observed in an original and dedicated experimental device placed inside a glove box under flowing pure argon. Pure uranium hydride powder with low amount of oxide (Oxidation mechanisms are proposed.

  7. The use dynamic avalanching and fractal analysis to characterise uranium oxide powders

    International Nuclear Information System (INIS)

    Hobbs, J.W.; Rhodes, D.

    2000-01-01

    Direct thermal denitration is an attractive method of co-converting mixed-metal nitrate solutions of plutonium and uranium into oxide because of its apparent simplicity. Such benefits are often marred by the relatively poor powder quality and handling characteristics, which can be overcome by modifications to the process chemistry. To ensure that powder synthesis routes under assessment require the minimal further processing it is necessary to be able to characterise the powder fully in term of the key fundamental properties. This paper will demonstrate the use of a dynamic avalanching technique, fractal analysis and morphology to assess processing behaviour. The use of dynamic avalanching to uniquely characterise the chaotic flow properties of urania powders has proved successful and results have shown that this technique is capable of detecting small differences in processing behaviour due changes in morphologies and particle size distribution. This technique has promise for being able to provide nearly instantaneous feedback to the powder generation process being monitored (e.g. calcination, milling, mixing). The use of fractals to describe powders is an interesting characterisation tool when combined with morphological shape factors and the flow index. (authors)

  8. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Clark, C.R.; Meyer, M.K.

    1998-01-01

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  9. Comparison of physical chemical properties of powders and respirable aerosols of industrial mixed uranium and plutonium oxide fuels

    International Nuclear Information System (INIS)

    Eidson, A.F.

    1982-01-01

    Studies were performed to characterize physical and chemical properties which may be important in determining the metabolism of accidentally released, inhaled aerosols of industrial mixed uranium and plutonium oxide fuels and to compare the properties of bulk powders and the respirable fraction they include. X-ray diffraction measurements showed that analysis of mixed-oxide powders from four process steps served to characterize their respirable fractions. IR spectroscopy was useful as a method to detect organic binders that were not observed by X-ray diffraction methods. Both X-ray diffraction and IR spectroscopy methods can be used in combination to identify the sources of a complex aerosol that might be released from more than one fabrication step. Isotopic distributions in powders and aerosols showed that information important for radiation dose to tissue calculations or Pu lung burden estimates can be obtained by analysis of powders. (U.K.)

  10. Nuclear forensic analysis of uranium oxide powders interdicted in Victoria, Australia

    Energy Technology Data Exchange (ETDEWEB)

    Kristo, Michael Joseph [Lawrence Livermore National Laboratory, Livermore, CA (United States); Keegan, Elizabeth; Colella, Michael [Australian Nuclear Science and Technology Organisation, Kirrawee, NSW (Australia); and others

    2015-07-01

    Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (∝ 1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K{sub 2}(UO{sub 2}){sub 3}O{sub 4} . 4H{sub 2}O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (∝ 380 μg/g total elemental impurities). The chemical form of the uranium was primarily UO{sub 3} . 2H{sub 2}O, with minor phases of U{sub 3}O{sub 8} and UO{sub 2}. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of {sup 236}U and {sup 232}U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.

  11. Manufacture of uranium dioxide powder

    International Nuclear Information System (INIS)

    Becker, M.

    1976-01-01

    Uranium dioxide powder is prepared by the AUC (ammonium uranyl carbonate) method. Supplementing the known process steps, the AUC, after separation from the mother liquor, is washed with an ammonium hydrogen carbonate or an NH 4 OH solution and is subsequently post-treated with a liquid which reduces the surface tension of the residual water in an AUC. Such a liquid is, for instance, alcohol

  12. New titrimetric method for oxygen to metal ratio in uranium oxide powders

    International Nuclear Information System (INIS)

    Ray, Vinod Kumar; Brahmananda Reddy, G.; Balaji Rao, Y.; Subba Rao, Y.

    2015-01-01

    O/U ratio is of high importance to both U 3 O 8 and UO 2 powders for different reasons. In UO 2 powder it is a guiding parameter for sintering process where as for U 3 O 8 , it indicates efficiency of ammonium di-uranate (ADU) to U 3 O 8 conversion process. In the present method for O/U determination, UO 2 and U 3 O 8 powders are dissolved in 4.5 M sulphuric acid and little HF by heating on hot plate. Subsequently, optimized quantity of phosphoric acid is added on cooling, for getting sharp end point. The resultant solution is titrated with standard potassium dichromate using barium diphenylamine sulphonate (BDS) as an indicator. The expanded uncertainties calculated for UO 2 and U 3 O 8 powders are ±0.004 and ±0.006 O/U ratio units respectively at 95 % confidence level. (author)

  13. Determination of lithium and potassium in uranium oxide powders and pellets by Flame Atomic Emission Spectrometric method

    International Nuclear Information System (INIS)

    Jat, J.R.; Balaji Rao, Y.; Prasada Rao, G.; Prahlad, B.

    2012-01-01

    The present paper describes a method developed at Control Laboratory, NFC which includes prior separation of lithium and potassium from uranium matrix before their measurements. Solvent extraction, using Tri-n-Butyl Phosphate (TBP) in CCI 4 followed by Tri-n-Octyl Phosphine Oxide (TOPO) in CCI 4 , is employed for prior separation of Li and K. The resultant aqueous solution was analyzed by Flame-Atomic Emission Spectrometric (AES) method. Solvent extraction conditions are optimized for measurement of Li and K in the same aliquot. Experimental conditions such as instrument calibration, flame condition, fuel flow, sample flow rate through nebulizer, burner height etc. are also optimized. Under the optimal condition the detection limits achieved for lithium is 0.02 ppm and 0.2 ppm for potassium. A RSD of ± 3 % for Li at 0.05 ppm and ± 4% for K at 1 ppm level has been achieved in this method. The results of lithium in the sample are compared with the values obtained by Inductively Coupled Plasma-Optical Emission Spectrometry (ICP-OES). Similarly, values of potassium are compared with Flame-Atomic Absorption Spectrometry (Flame-AAS) technique. The comparisons are in good agreement. The above method is simple, sensitive, reproducible and can be used for measurement of lithium and potassium in UO 2 powder and pellets on regular basis

  14. Reference materials for nondestructive assay of special nuclear material. Volume 1. Uranium oxide plus graphite powder

    International Nuclear Information System (INIS)

    Sprinkle, J.K.; Likes, R.N.; Parker, J.L.; Smith, H.A.

    1983-10-01

    This manual describes the fabrication of reference materials for use in gamma-ray-based nondestructive assay of low-density uranium-bearing samples. The sample containers are 2-l bottles. The reference materials consist of small amounts of UO 2 spread throughout a graphite matrix. The 235 U content ranges from 0 to 100 g. The manual also describes the far-field assay procedure used with low-resolution detectors

  15. Laser-induced breakdown spectroscopy measurements of uranium and thorium powders and uranium ore

    Energy Technology Data Exchange (ETDEWEB)

    Judge, Elizabeth J. [Chemistry Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Barefield, James E., E-mail: jbarefield@lanl.gov [Chemistry Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Berg, John M. [Manufacturing Engineering and Technology Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Clegg, Samuel M.; Havrilla, George J.; Montoya, Velma M.; Le, Loan A.; Lopez, Leon N. [Chemistry Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-05-01

    Laser-induced breakdown spectroscopy (LIBS) was used to analyze depleted uranium and thorium oxide powders and uranium ore as a potential rapid in situ analysis technique in nuclear production facilities, environmental sampling, and in-field forensic applications. Material such as pressed pellets and metals, has been extensively studied using LIBS due to the high density of the material and more stable laser-induced plasma formation. Powders, on the other hand, are difficult to analyze using LIBS since ejection and removal of the powder occur in the laser interaction region. The capability of analyzing powders is important in allowing for rapid analysis of suspicious materials, environmental samples, or trace contamination on surfaces since it most closely represents field samples (soil, small particles, debris etc.). The rapid, in situ analysis of samples, including nuclear materials, also reduces costs in sample collection, transportation, sample preparation, and analysis time. Here we demonstrate the detection of actinides in oxide powders and within a uranium ore sample as both pressed pellets and powders on carbon adhesive discs for spectral comparison. The acquired LIBS spectra for both forms of the samples differ in overall intensity but yield a similar distribution of atomic emission spectral lines. - Highlights: • LIBS analysis of mixed actinide samples: depleted uranium oxide and thorium oxide • LIBS analysis of actinide samples in powder form on carbon adhesive discs • Detection of uranium in a complex matrix (uranium ore) as a precursor to analyzing uranium in environmental samples.

  16. Infrared absorption study of ammonium uranates and uranium oxide powders formed during their thermal decomposition

    International Nuclear Information System (INIS)

    Rofail, N.H.; ELfekey, S.A.

    1992-01-01

    Ammonium uranates (AU) were precipitated from a nuclear-pure uranyl nitrate solution using different precipitating agents. IR spectra of the obtained uranates and oxides formed during their thermal decomposition have been studied. The results indicated that the precipitating agent, mode of stirring, washing and calcining temperature are important factors for a specific oxide formation.4 FIG., 3 TAB

  17. Melting of Uranium Metal Powders with Residual Salts

    International Nuclear Information System (INIS)

    Jin-Mok Hur; Dae-Seung Kang; Chung-Seok Seo

    2007-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) of the Korea Atomic Energy Research Institute focuses on the conditioning of Pressurized Water Reactor spent oxide nuclear fuel. After the oxide reduction step of the ACP, the resultant metal powders containing ∼ 30 wt% residual LiCl-Li 2 O should be melted for a consolidation of the fine metal powders. In this study, we investigated the melting behaviors of uranium metal powders considering the effects of a LiCl-Li 2 O residual salt. (authors)

  18. Study of the low temperature oxidation of uranium powders and its application to the sintering of uranium oxide powders; Etude de l'oxydation des poudres dtranium a basse temperature et son application au frittage de poudres d'uranium oxyde

    Energy Technology Data Exchange (ETDEWEB)

    Conte-Albert, M. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-06-01

    The uranium oxygen reaction has been studied with a view to obtaining U-UO{sub 2} samples containing about 20 per cent by weight of UO{sub 2} starting from spherical grain uranium powder (36 {mu} < {phi} < 50 {mu}). The techniques used are micrography, thermogravimetry, sintering under pressure, radio-crystallography. At 170 deg. C in air or argon + oxygen mixtures, the uranium oxide formed is always UO{sub 2} and it is uniformly distributed around the initial uranium spheres. These mixed powders can easily be sintered under pressure in the {gamma}-phase. The density of the samples obtained is 85 to 90 per cent of the theoretical density. The influence of UO{sub 2} on the properties of uranium has been shown by the use of dilatometry and thermal cycling in the {alpha} phase. The temperatures at which the phase changes {alpha} {r_reversible} {beta} and {beta} {r_reversible} {gamma} occur are lowered, the remnant expansion is decreased. High density samples resist well to thermal cycling; the characteristic defects of uranium: high distortion, wrinkled surface, have almost disappeared. Heat treatments in a secondary vacuum at 1050 deg. C cause crystallization of UO{sub 2} in a geometrical form and the appearance of a phase of the F.C.C. crystalline type having the composition U{sub W}C{sub X}O{sub Y}N{sub Z}. This phase causes a new decrease in the {alpha} {r_reversible} {beta}, {beta} {r_reversible} {gamma} transformation temperatures for the uranium. After ten dilatometric cycles the remanent expansion of the sample is about 0.5 per cent. The resistance to thermal cycling of a low density sample which has been heat-treated is similar to that of a high density sample which has not undergone a heat treatment. (author) [French] La reaction uranium-oxygene a ete etudiee pour permettre l'obtention d'echantillons U-UO{sub 2} a 20 pour cent en poids environ d'UO{sub 2}, a partir de billes d'uranium pulverulent (36 {mu} < {phi} < 50 {mu}). Les

  19. Optimization of dissolution process parameters for uranium ore concentrate powders

    Energy Technology Data Exchange (ETDEWEB)

    Misra, M.; Reddy, D.M.; Reddy, A.L.V.; Tiwari, S.K.; Venkataswamy, J.; Setty, D.S.; Sheela, S.; Saibaba, N. [Nuclear Fuel Complex, Hyderabad (India)

    2013-07-01

    Nuclear fuel complex processes Uranium Ore Concentrate (UOC) for producing uranium dioxide powder required for the fabrication of fuel assemblies for Pressurized Heavy Water Reactor (PHWR)s in India. UOC is dissolved in nitric acid and further purified by solvent extraction process for producing nuclear grade UO{sub 2} powder. Dissolution of UOC in nitric acid involves complex nitric oxide based reactions, since it is in the form of Uranium octa oxide (U{sub 3}O{sub 8}) or Uranium Dioxide (UO{sub 2}). The process kinetics of UOC dissolution is largely influenced by parameters like concentration and flow rate of nitric acid, temperature and air flow rate and found to have effect on recovery of nitric oxide as nitric acid. The plant scale dissolution of 2 MT batch in a single reactor is studied and observed excellent recovery of oxides of nitrogen (NO{sub x}) as nitric acid. The dissolution process is automated by PLC based Supervisory Control and Data Acquisition (SCADA) system for accurate control of process parameters and successfully dissolved around 200 Metric Tons of UOC. The paper covers complex chemistry involved in UOC dissolution process and also SCADA system. The solid and liquid reactions were studied along with multiple stoichiometry of nitrous oxide generated. (author)

  20. Uranium Dioxide Powder Flow ability Improvement Using Sol-Gel

    International Nuclear Information System (INIS)

    Juanda, D.; Sambodo Daru, G.

    1998-01-01

    The improvement of flow ability characteristics of uranium dioxide powder has been done using sol-gel process. To anticipate a pellet mass production with uniform pellet dimension, the uranium dioxide powder must be have a spherical form. Uranium dioxide spherical powder has been diluted in acid transformed into sol colloidal solution. To obtain uranium dioxide spherical form, the uranium sol-colloidal solution has been dropped in a hot paraffin ( at the temperature of 90 0 C) to form gelatinous colloid and then dried at 800 0 C, and sintered at the temperature of 1700 0 C. The flow ability of spherical uranium dioxide powder has been examined by using Flowmeter Hall (ASTM. B. 213-46T). The measurement result reveals that the spherical uranium dioxide powder has a flow ability twice than that of unprocessed uranium dioxide powder

  1. Uranium oxide recovering method

    International Nuclear Information System (INIS)

    Ota, Kazuaki; Takazawa, Hiroshi; Teramae, Naoki; Onoue, Takeshi.

    1997-01-01

    Nitrates containing uranium nitrate are charged in a molten salt electrolytic vessel, and a heat treatment is applied to prepare molten salts. An anode and a cathode each made of a graphite rod are disposed in the molten salts. AC voltage is applied between the anode and the cathode to conduct electrolysis of the molten salts. Uranium oxides are deposited as a recovered product of uranium, on the surface of the anode. The nitrates containing uranium nitrate are preferably a mixture of one or more nitrates selected from sodium nitrate, potassium nitrate, calcium nitrate and magnesium nitrate with uranium nitrate. The nitrates may be liquid wastes of nitrates. The temperature for the electrolysis of the molten salts is preferably from 150 to 300degC. The voltage for the electrolysis of the molten salts is preferably an AC voltage of from 2 to 6V, more preferably from 4 to 6V. (I.N.)

  2. Welding of a powder metallurgy uranium alloy

    International Nuclear Information System (INIS)

    Holbert, R.K.; Doughty, M.W.; Alexander-Morrison, G.M.

    1989-01-01

    The interest at the Oak Ridge Y-12 Plant in powder metallurgy (P/M) uranium parts is due to the potential cost savings in the fabrication of the material, to achieving a more homogeneous product, and to the reduction of uranium scrap. The joining of P/M uranium-6 wt-% niobium (U-6Nb) alloys by the electron beam (EB) welding process results in weld porosity. Varying the EB welding parameters did not eliminate the porosity. Reducing the oxygen and nitrogen content in this P/M uranium material did minimize the weld porosity, but this step made the techniques of producing the material more difficult. Therefore, joining wrought and P/M U-6Nb rods with the inertia welding technique is considered. Since no gases will be evolved with the solid-state welding process and the weld area will be compacted, porosity should not be a problem in the inertia welding of uranium alloys. The welds that are evaluated are wrought-to-wrought, wrought-to-P/M, and P/M-to-P/M U-6Nb samples

  3. Complex defects in the oxidation of uranium

    International Nuclear Information System (INIS)

    MacCrone, R.K.; Sankaran, S.; Shatynski, S.R.; Colmenares, C.A.

    1986-01-01

    We are reporting EPR results obtained with uranium powder samples fully oxidized in dry air, water vapor, and air/water vapor mixtures. The results reported previously are confirmed and additional paramagnetic centers, associated with chemisorbed species, have been identified. The temperature dependence of the g-value for these centers from room temperature to 10K is also reported

  4. Method for converting uranium oxides to uranium metal

    International Nuclear Information System (INIS)

    Duerksen, W.K.

    1988-01-01

    A method for converting uranium oxide to uranium metal is described comprising the steps of heating uranium oxide in the presence of a reducing agent to a temperature sufficient to reduce the uranium oxide to uranium metal and form a heterogeneous mixture of a uranium metal product and oxide by-products, heating the mixture in a hydrogen atmosphere at a temperature sufficient to convert uranium metal in the mixture to uranium hydride, cooling the resulting uranium hydride-containing mixture to a temperature sufficient to produce a ferromagnetic transition in the uranium hydride, magnetically separating the cooled uranium hydride from the mixture, and thereafter heating the separated uranium hydride in an inert atmosphere to a temperature sufficient to convert the uranium hydride to uranium metal

  5. Uranium bed oxidation vacuum process system

    International Nuclear Information System (INIS)

    McLeland, H.L.

    1977-01-01

    Deuterium and tritium gases are occluded in uranium powder for release into neutron generator tubes. The uranium powder is contained in stainless steel bottles, termed ''beds.'' If these beds become damaged, the gases must be removed and the uranium oxidized in order not to be flammable before shipment to ERDA disposal grounds. This paper describes the system and methods designed for the controlled degassing and oxidation process. The system utilizes sputter-ion, cryo-sorption and bellows pumps for removing the gases from the heated source bed. Removing the tritium gas is complicated by the shielding effect of helium-3, a byproduct of tritium decay. This effect is minimized by incremental pressure changes, or ''batch'' processing. To prevent runaway exothermic reaction, oxidation of the uranium bed is also done incrementally, or by ''batch'' processing, rather than by continuous flow. The paper discusses in detail the helium-3 shielding effect, leak checks that must be made during processing, bed oxidation, degree of gas depletion, purity of gases sorbed from beds, radioactivity of beds, bed disposal and system renovation

  6. A study on the oxidation behavior of uranium

    International Nuclear Information System (INIS)

    Kim, Tae Kook; Kang, Kweon Ho; Kim, Kil Jeang; Kang, Il Sik; Jung, Kyung Whan

    1998-03-01

    When storing depleted uranium wastes, careful handling is required due to their very high oxidation rates. To determine the oxidation mechanism and oxidation rate of depleted uranium wastes, the most important factors to be considered in their treatment, an experiment was carried out by varying the heating rates of the Air-Controlled Oxidizer. The experiment, showed that depleted uranium wastes are pulverized after complete oxidation because of the density difference and then converted to UO 2 , U 3 O 7 , U 3 O 8 . The grain size of pulverized powder decrease with increased temperature. (author). 30 refs., 5 tabs., 28 figs

  7. Electrolytic recovery of uranium oxides

    International Nuclear Information System (INIS)

    Gurr, W.R.

    1979-01-01

    A method is described for extracting uranium oxide from a solution of one or more uranium compounds, e.g. leach liquors, comprising subjecting the solution to electrolysis utilizing a high current density, e.g. 500 to 4000 amp/m 2 , whereby uranium oxide is formed at the cathode and is recovered. The method is particularly suited to a continuous process using a rotating cathode cell. (author)

  8. Method for converting uranium oxides to uranium metal

    Science.gov (United States)

    Duerksen, Walter K.

    1988-01-01

    A process is described for converting scrap and waste uranium oxide to uranium metal. The uranium oxide is sequentially reduced with a suitable reducing agent to a mixture of uranium metal and oxide products. The uranium metal is then converted to uranium hydride and the uranium hydride-containing mixture is then cooled to a temperature less than -100.degree. C. in an inert liquid which renders the uranium hydride ferromagnetic. The uranium hydride is then magnetically separated from the cooled mixture. The separated uranium hydride is readily converted to uranium metal by heating in an inert atmosphere. This process is environmentally acceptable and eliminates the use of hydrogen fluoride as well as the explosive conditions encountered in the previously employed bomb-reduction processes utilized for converting uranium oxides to uranium metal.

  9. Uranium plutonium oxide fuels

    International Nuclear Information System (INIS)

    Cox, C.M.; Leggett, R.D.; Weber, E.T.

    1981-01-01

    Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

  10. Obtention of powdered UO{sub 2} oxide by water vapor action on calcium-reduced uranium; Obtention d'oxyde UO{sub 2} pulverulent par action de la vapeur d'eau sur l'uranium calciothermique

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, Claude; Barnoud, Louis [Commissariat a l' energie atomique et aux energies alternatives - CEA, Centre d' etudes nucleaires de Grenoble (France)

    1960-07-01

    The oxidation method used allows to control and to easily vary the reaction kinetics in order to obtain oxidized uranium samples with predetermined UO{sub 2} contents. The solid reagent, made of spherical uranium grains of constant size, is an interesting material for the application of heterogeneous kinetic principles. An experimental continuation of this study is expected. Reprint of a paper published in Comptes rendus des seances de l'Academie des Sciences, t. 250, p. 1495-1497, sitting of 22 February 1960 [French] La methode d'oxydation utilisee permet de controler, et de faire varier facilement la vitesse de reaction de facon a obtenir des echantillons d'uranium oxydes en UO{sub 2} a des teneurs predeterminees. En outre le reactif solide, compose de grains d'uranium spheriques de meme taille constitue un materiau interessant pour l'application des principes de la cinetique heterogene. Un prolongement experimental de cette etude est envisage. Reproduction d'un article publie dans les Comptes rendus des seances de l'Academie des Sciences, t. 250, p. 1495-1497, seance du 22 fevrier 1960.

  11. Set up of Uranium-Molybdenum powder production (HMD process)

    International Nuclear Information System (INIS)

    Lopez, Marisol; Pasqualini, Enrique E.; Gonzalez, Alfredo G.

    2003-01-01

    Powder metallurgy offers different alternatives for the production of Uranium-Molybdenum (UMo) alloy powder in sizes smaller than 150 microns. This powder is intended to be used as a dispersion fuel in an aluminum matrix for research, testing and radioisotopes production reactors (MTR). A particular process of massive hydriding the UMo alloy in gamma phase has been developed. This work describes the final adjustments of process variables to obtain UMo powder by hydriding-milling-de hydriding (HMD) and its capability for industrial scaling up. (author)

  12. Standard test methods for analysis of sintered gadolinium oxide-uranium dioxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 These test methods cover procedures for the analysis of sintered gadolinium oxide-uranium dioxide pellets to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Section Carbon (Total) by Direct CombustionThermal Conductivity Method C1408 Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method Chlorine and Fluorine by Pyrohydrolysis Ion-Selective Electrode Method C1502 Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide Gadolinia Content by Energy-Dispersive X-Ray Spectrometry C1456 Test Method for Determination of Uranium or Gadolinium, or Both, in Gadolinium Oxide-Uranium Oxide Pellets or by X-Ray Fluorescence (XRF) Hydrogen by Inert Gas Fusion C1457 Test Method for Determination of Total Hydrogen Content of Uranium Oxide Powders and Pellets by Carrier Gas Extraction Isotopic Uranium Composition by Multiple-Filament Surface-Ioni...

  13. Method for preparing a sinterable uranium dioxide powder

    International Nuclear Information System (INIS)

    Thornton, T.A.; Holaday, V.D. Jr.

    1985-01-01

    This invention provides an improved method for preparing a sinterable uranium dioxide powder for the preparation of nuclear fuel, using microwave radiation in a microwave induction furnace. The starting compound may be uranyl nitrate hexahydrate, ammonium diuranate or ammonium uranyl carbonate. The starting compound is heated in a microwave induction furnace for a period of time sufficient for compound decomposition. The decomposed compound is heated in a microwave induction furnace in a reducing atmosphere for a period of time sufficient to reduce the decomposed compound to uranium dioxide powder

  14. Oxygen isotope fractionation in uranium oxides

    International Nuclear Information System (INIS)

    Zheng Yongfei

    1995-01-01

    Thermodynamic oxygen isotope factors for uranium oxides have been calculated by means of the modified increment method. The sequence of 18 O-enrichment in the uranium oxides with respect to the common rock-forming minerals is predicted as follows: spinel 3 < illite. Two sets of self-consistent fractionation factors between the uranium oxides and water and between the uranium oxides and the other minerals have been obtained for 0∼1200 degree C. The theoretical results are applicable to the isotopic geothermometry of uranium ores when pairing with other gangue minerals in hydrothermal uranium deposits

  15. Sintering of uranium oxide of high specific surface area

    International Nuclear Information System (INIS)

    Bel, Alain; Francois, Bernard; Delmas, Roger; Caillat, Roger

    1959-01-01

    The extent to which a uranium oxide powder deriving from ammonium uranate or uranium peroxide lends itself to the sintering process depends largely on its specific surface area. When this is greater than 5 m 2 / g there is an optimum temperature for sintering in hydrogen. This temperature becomes less as the specific area of the powder is greater. Reprint of a paper published in Comptes rendus des seances de l'Academie des Sciences, t. 249, p. 1045-1047, sitting of 21 September 1959 [fr

  16. Isotope exchange between gaseous hydrogen and uranium hydride powder

    International Nuclear Information System (INIS)

    Shugard, Andrew D.; Buffleben, George M.; Johnson, Terry A.; Robinson, David B.

    2014-01-01

    Highlights: • Isotope exchange between hydrogen gas and uranium hydride powder can be rapid and reversible. • Gas–solid exchange rate is controlled by transport within ∼0.7 μm hydride particles. • Gas chromatographic separation of hydrogen isotopes using uranium hydride is feasible. - Abstract: Isotope exchange between gaseous hydrogen and solid uranium hydride has been studied by flowing hydrogen (deuterium) gas through packed powder beds of uranium deuteride (hydride). We used a residual gas analyzer system to perform real-time analysis of the effluent gas composition. We also developed an exchange and transport model and, by fitting it to the experimental data, extracted kinetic parameters for the isotope exchange reaction. Our results suggest that, from approximately 70 to 700 kPa and 25 to 400 °C, the gas-to-solid exchange rate is controlled by hydrogen and deuterium transport within the ∼0.7 μm diameter uranium hydride particles. We use our kinetic parameters to show that gas chromatographic separation of hydrogen and deuterium using uranium hydride could be feasible

  17. Characterization of uranium silicide powder using XRD

    International Nuclear Information System (INIS)

    Garcia, Rafael H.L.; Saliba-Silva, Adonis M.; Carvalho, Elita F.U.; Lima, Nelson B.; Ichikawa, Rodrigo U.; Martinez, Luiz G.

    2013-01-01

    Uranium silicide (U 3 Si 2 ) is an intermetallic used as nuclear fuel in most modern MTR - Materials Test Reactor. Dispersed in aluminum, this fuel allows high uranium densities, up to 4.8 gU/cm 3 . At IPEN, the fabrication of fuel elements based on U 3 Si 2 for the IEA-R1 reactor is carried out in the Nuclear Fuel Center (CCN), by vacuum induction melting of uranium and silicon, followed by grinding. Before employed in a nuclear reactor, U 3 Si 2 must be submitted to a strict quality control, which includes granulometry, density, X-ray radiography for dispersion homogeneity, chemical and crystallographic characterization. Concerning phase composition for a qualified fuel, the fraction of U 3 Si 2 should be higher than 80wt.%. Aiming at the development of a routine methodology for quantification of phases via analysis of XRD data using the Rietved method, six samples from two production baths of CCN were submitted to X-ray diffraction. The data were analyzed using software GSAS and line profile analysis methods. The results suggest that fusion product have preferred orientation and grinding step is important for a better refinement. (author)

  18. Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O' Brien

    2013-02-01

    Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.

  19. Process for fabricating mixed-oxide powders

    International Nuclear Information System (INIS)

    Elmaleh, D.; Giraudel, A.

    1975-01-01

    A physical-chemical process for fabricating homogeneous powders suitable for sintering is described. It can be applied to the synthesis of all mixed oxides having mutually compatible and water soluble salts. As a specific example, the fabrication of lead titanate-zirconate powders used to make hot pressed ceramics is described. These ceramics show improved piezoelectric properties [fr

  20. Influence of uranium hydride oxidation on uranium metal behaviour

    International Nuclear Information System (INIS)

    Patel, N.; Hambley, D.; Clarke, S.A.; Simpson, K.

    2013-01-01

    This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

  1. Influence of uranium hydride oxidation on uranium metal behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Patel, N.; Hambley, D. [National Nuclear Laboratory (United Kingdom); Clarke, S.A. [Sellafield Ltd (United Kingdom); Simpson, K.

    2013-07-01

    This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

  2. Metallurgical examination of powder metallurgy uranium alloy welds

    International Nuclear Information System (INIS)

    Morrison, A.G.M.; Dobbins, A.G.; Holbert, R.K.; Doughty, M.W.

    1986-01-01

    Inertia welding provided a successful technique for joining full density, powder metallurgy uranium-6 wt pct niobium alloy. Initial joining attempts concentrated on the electron beam method, but this method failed to produce a sound weld. The electron beam welds and the inertia welds were evaluated by radiography and metallography. Electron beam welds were attempted on powder metallurgy plates which contained various levels of oxygen and nitrogen. All welds were porous. Sixteen inertia welds were made and all welds were radiographically sound. The tensile properties of the joints were found to be equivalent to the p/m base metal properties

  3. Analysis of Hazards Associated with a Process Involving Uranium Metal and Uranium Hydride Powders

    Energy Technology Data Exchange (ETDEWEB)

    Bullock, J.S.

    2000-05-01

    An analysis of the reaction chemistry and operational factors associated with processing uranium and uranium hydride powders is presented, focusing on a specific operation in the Development Division which was subjected to the Job Hazard Analysis (JHA) process. Primary emphasis is on the thermodynamic factors leading to pyrophoricity in common atmospheres. The discussion covers feed powders, cold-pressed and hot-pressed materials, and stray material resulting from the operations. The sensitivity of the various forms of material to pyrophoricity in common atmospheres is discussed. Operational recommendations for performing the work described are given.

  4. Milling uranium silicide powder for dispersion nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, E.; Silva, D.G.; Souza, J.A.B.; Durazzo, M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Riella, H.G. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)

    2009-07-01

    Full text: Uranium silicide (U3Si2) is presently considered the best fuel qualified so far in terms of uranium loading and performance. Stability of the U3Si2 fuel with uranium density of 4.8 g/cm3 was confirmed by burnup stability tests performed during the Reduced Enrichment for Research and Test Reactors (RERTR) program. This fuel was chosen to compose the first core of the new Brazilian Multipurpose Research Reactor (RMB), planned to be constructed in the next years. This new reactor will consume bigger quantities of U3Si2 powder, when compared with the small consumption of the IEA-R1 research reactor of IPEN-CNEN/SP, the unique MTR type research reactor operating in the country. At the present time, the milling operation of U3Si2 ingots is made manually. In order to increase the powder production capacity, the manual milling must be replaced by an automated procedure. This paper describes a new milling machine and procedure developed to produce U3Si2 powder with higher efficiency. (author)

  5. Predictor of regulation of uranium dioxide powder pressing process

    International Nuclear Information System (INIS)

    Motta, Eduardo Souza; Araujo, Victor Hugo Leal de; Bernardelli, Sergio Henrique

    2007-01-01

    One of the most important steps of the uranium dioxide pellets fabrication used in the nuclear fuel elements is the green pellets pressing. The target density of the pellets after the sintering process determines the density of the green pellet. To meet the same sintered target density the green density may vary according to the powder characteristics. These variations implies in changing the regulation of the press for different powder's patches. The regulation done empirically imply in productivity loss and necessity of reprocessing the pellets pressed during the press regulation and also depends on the operator experience. At this work, was developed an artificial neural network feed forward back propagation to predict the press regulation, depending on the powder characteristics and the green pellet's target density. The results obtained at INB - Industrias Nucleares do Brasil S. A. during the fabrication of the fifth recharge of Angra II nuclear power plant are presented. (author)

  6. Operating experience in processing of differently sourced deeply depleted uranium oxide and production of deeply depleted uranium metal ingots

    International Nuclear Information System (INIS)

    Manna, S.; Ladola, Y.S.; Sharma, S.; Chowdhury, S.; Satpati, S.K.; Roy, S.B.

    2009-01-01

    Uranium Metal Plant (UMP) of BARC had first time experience on production of three Depleted Uranium Metal (DUM) ingots of 76kg, 152kg and 163kg during March 1991. These ingots were produced by processing depleted uranyl nitrate solution produced at Plutonium Plant (PP), Trombay. In recent past Uranium Metal Plant (UMP), Uranium Extraction Division (UED), has been assigned to produce tonnage quantity of Deeply DUM (DDUM) from its oxide obtained from PP, PREFRE and RMP, BARC. This is required for shielding the high radioactive source of BHABHATRON Tele-cobalt machine, which is used for cancer therapy. The experience obtained in processing of various DDU oxides is being utilized for design of large scale DDU-metal plant under XIth plan project. The physico- chemical characteristics like morphology, density, flowability, reactivity, particle size distribution, which are having direct effect on reactivity of the powders of the DDU oxide powder, were studied and the shop-floor operational experience in processing of different oxide powder were obtained and recorded. During campaign trials utmost care was taken to standardized all operating conditions using the same equipment which are in use for natural uranium materials processing including safety aspects both with respect to radiological safety and industrial safety. Necessary attention and close monitoring were specially arranged and maintained for the safety aspects during the trial period. In-house developed pneumatic transport system was used for powder transfer and suitable dust arresting system was used for reduction of powder carry over

  7. Production of sized particles of uranium oxides and uranium oxyfluorides

    International Nuclear Information System (INIS)

    Knudsen, I.E.; Randall, C.C.

    1976-01-01

    A process is claimed for converting uranium hexafluoride (UF 6 ) to uranium dioxide (UO 2 ) of a relatively large particle size in a fluidized bed reactor by mixing uranium hexafluoride with a mixture of steam and hydrogen and by preliminary reacting in an ejector gaseous uranium hexafluoride with steam and hydrogen to form a mixture of uranium and oxide and uranium oxyfluoride seed particles of varying sizes, separating the larger particles from the smaller particles in a cyclone separator, recycling the smaller seed particles through the ejector to increase their size, and introducing the larger seed particles from the cyclone separator into a fluidized bed reactor where the seed particles serve as nuclei on which coarser particles of uranium dioxide are formed. 9 claims, 2 drawing figures

  8. Oxidation of nano-sized aluminum powders

    International Nuclear Information System (INIS)

    Vorozhtsov, A.B.; Lerner, M.; Rodkevich, N.; Nie, H.; Abraham, A.; Schoenitz, M.; Dreizin, E.L.

    2016-01-01

    Highlights: • Weight gain measured in TG oxidation experiments was split between particles of different sizes. • Reaction kinetics obtained by isoconversion explicitly accounting for the effect of size distribution. • Activation energy is obtained as a function of oxide thickness for growth of amorphous alumina. • Oxidation mechanism for nanopowders remains the same as for coarser aluminum powders. - Abstract: Oxidation of aluminum nanopowders obtained by electro-exploded wires is studied. Particle size distributions are obtained from transmission electron microscopy (TEM) images. Thermo-gravimetric (TG) experiments are complemented by TEM and XRD studies of partially oxidized particles. Qualitatively, oxidation follows the mechanism developed for coarser aluminum powder and resulting in formation of hollow oxide shells. Sintering of particles is also observed. The TG results are processed to account explicitly for the particle size distribution and spherical shapes, so that oxidation of particles of different sizes is characterized. The apparent activation energy is obtained as a function of the reaction progress using model-free isoconversion processing of experimental data. A complete phenomenological oxidation model is then proposed assuming a spherically symmetric geometry. The oxidation kinetics of aluminum powder is shown to be unaffected by particle sizes reduced down to tens of nm. The apparent activation energy describing growth of amorphous alumina is increasing at the very early stages of oxidation. The higher activation energy is likely associated with an increasing homogeneity in the growing amorphous oxide layer, initially containing multiple defects and imperfections. The trends describing changes in both activation energy and pre-exponent of the growing amorphous oxide are useful for predicting ignition delays of aluminum particles. The kinetic trends describing activation energies and pre-exponents in a broader range of the oxide

  9. Nuclear energy - Uranium dioxide powder and sintered pellets - Determination of oxygen/uranium atomic ratio by the amperometric method. 2. ed.

    International Nuclear Information System (INIS)

    2007-01-01

    This International Standard specifies an analytical method for the determination of the oxygen/uranium atomic ratio in uranium dioxide powder and sintered pellets. The method is applicable to reactor grade samples of hyper-stoichiometric uranium dioxide powder and pellets. The presence of reducing agents or residual organic additives invalidates the procedure. The test sample is dissolved in orthophosphoric acid, which does not oxidize the uranium(IV) from UO 2 molecules. Thus, the uranium(VI) that is present in the dissolved solution is from UO 3 and/or U 3 O 8 molecules only, and is proportional to the excess oxygen in these molecules. The uranium(VI) content of the solution is determined by titration with a previously standardized solution of ammonium iron(II) sulfate hexahydrate in orthophosphoric acid. The end-point of the titration is determined amperometrically using a pair of polarized platinum electrodes. The oxygen/uranium ratio is calculated from the uranium(VI) content. A portion, weighing about 1 g, of the test sample is dissolved in orthophosphoric acid. The dissolution is performed in an atmosphere of nitrogen or carbon dioxide when sintered material is being analysed. When highly sintered material is being analysed, the dissolution is performed at a higher temperature in purified phosphoric acid from which the water has been partly removed. The cooled solution is titrated with an orthophosphoric acid solution of ammonium iron(II) sulfate, which has previously been standardized against potassium dichromate. The end-point of the titration is detected by the sudden increase of current between a pair of polarized platinum electrodes on the addition of an excess of ammonium iron(II) sulfate solution. The paper provides information about scope, principle, reactions, reagents, apparatus, preparation of test sample, procedure (uranium dioxide powder, sintered pellets of uranium dioxide, highly sintered pellets of uranium dioxide and determination

  10. Electrochemical preparation of new uranium oxide phases

    International Nuclear Information System (INIS)

    Smolenskij, V.V.; Lyalyushkin, N.V.; Bove, A.L.; Komarov, V.K.; Kapshukov, I.I.

    1992-01-01

    Behaviour of uranium ions in oxidation states 3+ and 4+ in molten chlorides of alkali metals in the temperature range of 700-900 degC in the atmosphere of an inert gas was studied by the method of cyclic voltametry. It is shown that as a result of introduction of crystal uranium dioxide into the salt melt formation of uranium oxide ions of the composition UO + and UO 2+ occurs, the ions participating in electrode reactions and bringing about formation of the following uranium oxides on the cathode: UO and, presumably, U 3 O 4 . Oxides UO and U 3 O 4 are thermodynamically unstable at low temperatures and decompose into uranium oxide of the composition UO 2-x , where x varies from 0 to 0.05, and metal uranium

  11. Electrical impedance studies of uranium oxide

    International Nuclear Information System (INIS)

    Hampton, R.N.

    1986-11-01

    The thesis presents data on the electrical properties of uranium oxide at temperatures from 1700K to 4.2K, and pressures between 25 K bar and 70 K bar. The impedance data were analysed using the technique of complex plane representation to establish the conductivity and dielectric constant of uranium dioxide. The thermophysical data were compared with previously reported experimental and theoretical work on uranium dioxide and other fluorite structured oxides. (U.K.)

  12. Process for producing uranium oxide rich compositions from uranium hexafluoride

    International Nuclear Information System (INIS)

    DeHollander, W.R.; Fenimore, C.P.

    1978-01-01

    Conversion of gaseous uranium hexafluoride to a uranium dioxide rich composition in the presence of an active flame in a reactor defining a reaction zone is achieved by separately introducing a first gaseous reactant comprising a mixture of uranium hexafluoride and a reducing carrier gas, and a second gaseous reactant comprising an oxygen-containing gas. The reactants are separated by a shielding gas as they are introduced to the reaction zone. The shielding gas temporarily separates the gaseous reactants and temporarily prevents substantial mixing and reacting of the gaseous reactants. The flame occurring in the reaction zone is maintained away from contact with the inlet introducing the mixture to the reaction zone. After suitable treatment, the uranium dioxide rich composition is capable of being fabricated into bodies of desired configuration for loading into nuclear fuel rods. Alternatively, an oxygen-containing gas as a third gaseous reactant is introduced when the uranium hexafluoride conversion to the uranium dioxide rich composition is substantially complete. This results in oxidizing the uranium dioxide rich composition to a higher oxide of uranium with conversion of any residual reducing gas to its oxidized form

  13. Dissolution of uranium oxide TBP-HNO3 complex

    International Nuclear Information System (INIS)

    Mizuno, Mineo; Kosaka, Yuji; Mori, Yukihide; Shimada, Takashi

    2002-12-01

    As a head end process for the pulverization of the spent fuel, the mechanical method (the shredder method) and the pyro-chemical method (oxidisation heat-treatment) have been examined. UO 2 is a main ingredient of Uranium oxide powder by the mechanical method, and U 3 O 8 is that by the pyro-chemical method. Moreover, the particle size of the pulverized powder depend on the conditions of the pulverizing process. As it was considered that the difference of dissolution rates of samples was caused by the difference of sample chemical forms and dissolution temperature, parametric surveys on chemical form and particle size of powder and dissolution temperature were carried out, and the following results were obtained. 1) The remarkable difference of dissolution rate between U 3 O 8 powder (average particle size 3.7 μm) and UO 2 powder (average particle size 2.4 μm) which have comparatively similar particle size was not observed. 2) It was confirmed that the dissolution rate became lower according to the particle size increase (average particle size 2.4 μm-1 mm). And it was considered that dissolution rate had strong dependency on particle size, according to the results that the powder with 1 mm particle size did not dissolute completely after 5 hours test. 3) The temperature dependency of the dissolution rate was confirmed by dissolution test with UO 2 powder (average particle size 2.4 μm-1 mm). The higher dissolution rate was obtained in the higher dissolution temperature, and 11 kcal/mol was obtained as activation energy of dissolution. 4) In the dissolution test of UO 2 powder, the nitric acid concentration started to change earlier than that of U 3 O 8 powder and concentration change range became larger compared with that in the dissolution test of U 3 O 8 powder. It was considered that those differences were caused by difference in mole ratio of Uranium and nitric acid which are consumed in the dissolution reaction (3:7 for U 3 O 8 , 3:8 for UO 2 ). 5) In case

  14. Green strength of zirconium sponge and uranium dioxide powder compacts

    International Nuclear Information System (INIS)

    Balakrishna, Palanki; Murty, B. Narasimha; Sahoo, P.K.; Gopalakrishna, T.

    2008-01-01

    Zirconium metal sponge is compacted into rectangular or cylindrical shapes using hydraulic presses. These shapes are stacked and electron beam welded to form a long electrode suitable for vacuum arc melting and casting into solid ingots. The compact electrodes should be sufficiently strong to prevent breakage in handling as well as during vacuum arc melting. Usually, the welds are strong and the electrode strength is limited by the green strength of the compacts, which constitute the electrode. Green strength is also required in uranium dioxide (UO 2 ) powder compacts, to withstand stresses during de-tensioning after compaction as well as during ejection from the die and for subsequent handling by man and machine. The strengths of zirconium sponge and UO 2 powder compacts have been determined by bending and crushing respectively, and Weibul moduli evaluated. The green density of coarse sponge compact was found to be larger than that from finer sponge. The green density of compacts from lightly attrited UO 2 powder was higher than that from unattrited category, accompanied by an improvement in UO 2 green crushing strength. The factors governing green strength have been examined in the light of published literature and experimental evidence. The methodology and results provide a basis for quality control in metal sponge and ceramic powder compaction in the manufacture of nuclear fuel

  15. Studies on the sintering behaviour of uranium dioxide powder compacts

    International Nuclear Information System (INIS)

    Das, P.; Chowdhury, R.

    1988-01-01

    Uranium dioxide fuel pellets are normally made from their precursor ammonium diuranate, followed by calcination, subsequent reduction to sinterable grade powders and a post operation treatment of pressing and sintering. The low temperature calcined powders, usually exhibiting non-crystalline behaviour (under X-ray diffraction studies) progressively transforms into a crystalline variety on subsequent heat treatment at higher temperature. It is observed however that powders calcined between 800 to 900 0 C exhibit enhanced densification behaviour when sintered at higher temperatures. The isothermal shrinkage versus time plot of the sintered compacts are well described by a hyperbolic relationship which takes care of the observed shrinkage (λ) as caused due to a cumulative effect from the initial sintering of the powder compacts at zero time (α) and that caused due to the structural transformation from a non-crystalline modification with increased thermal treatment (β). The derived equation is a modification of the sintering mechanism of the viscous flow type proposed by Frenkel, involving sintering of an amorphous phase, the viscosity of the latter is presumed to increase with increasing thermal treatment to assume the final modified form as λ=t/(α+βt), where t = time, λ = shrinkage and α and β are the unknown parameters. (orig.)

  16. Oxidation of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Orman, S.

    1976-01-01

    The corrosion behaviour of uranium in oxygen, water and water + oxygen mixtures is compared and contrasted. A considerable amount of work, much of it conflicting, has been published on the U + H 2 O and U + H 2 O + O 2 systems. An attempt has been made to summarise this data and to explain the reasons for the lack of agreement between the experimental results. The evidence for the mechanism involving OH - ion diffusion as the reacting entity in both the U + H 2 O and U + O 2 + H 2 O reactions is advanced. The more limited corrosion data on some lean uranium alloys and on some higher addition alloys referred to as stainless materials is summarised together with some previously unreported results obtained with these materials at AWRE. The data indicates that in the absence of oxygen the lean alloys behave in a similar manner to uranium and evolve hydrogen in approximately theoretical quantities. But the stainless alloys absorb most of the product hydrogen and assessments of reactivity based on hydrogen evolution would be very inaccurate. The direction that future corrosion work on these materials should take is recommended

  17. Reaction of uranium oxides with chlorine and carbon or carbon monoxide to prepare uranium chlorides

    Energy Technology Data Exchange (ETDEWEB)

    Haas, P.A.; Lee, D.D.; Mailen, J.C.

    1991-11-01

    The preferred preparation concept of uranium metal for feed to an AVLIS uranium enrichment process requires preparation of uranium tetrachloride (UCI{sub 4}) by reacting uranium oxides (UO{sub 2}/UO{sub 3}) and chlorine (Cl{sub 2}) in a molten chloride salt medium. UO{sub 2} is a very stable metal oxide; thus, the chemical conversion requires both a chlorinating agent and a reducing agent that gives an oxide product which is much more stable than the corresponding chloride. Experimental studies in a quartz reactor of 4-cm ID have demonstrated the practically of some chemical flow sheets. Experimentation has illustrated a sequence of results concerning the chemical flow sheets. Tests with a graphite block at 850{degrees}C demonstrated rapid reactions of Cl{sub 2} and evolution of carbon dioxide (CO{sub 2}) as a product. Use of carbon monoxide (CO) as the reducing agent also gave rapid reactions of Cl{sub 2} and formation of CO{sub 2} at lower temperatures, but the reduction reactions were slower than the chlorinations. Carbon powder in the molten salt melt gave higher rates of reduction and better steady state utilization of Cl{sub 2}. Addition of UO{sub 2} feed while chlorination was in progress greatly improved the operation by avoiding the plugging effects from high UO{sub 2} concentrations and the poor Cl{sub 2} utilizations from low UO{sub 2} concentrations. An UO{sub 3} feed gave undesirable effects while a feed of UO{sub 2}-C spheres was excellent. The UO{sub 2}-C spheres also gave good rates of reaction as a fixed bed without any molten chloride salt. Results with a larger reactor and a bottom condenser for volatilized uranium show collection of condensed uranium chlorides as a loose powder and chlorine utilizations of 95--98% at high feed rates. 14 refs., 7 figs., 14 tabs.

  18. PROCESSES OF CHLORINATION OF URANIUM OXIDES

    Science.gov (United States)

    Rosenfeld, S.

    1958-09-16

    An improvement is described in the process fur making UCl/sub 4/ from uranium oxide and carbon tetrachloride. In that process, oxides of uranium are contacted with carbon tetrachloride vapor at an elevated temperature. It has been fuund that the reaction product and yield are improved if the uranlum oxide charge is disposed in flat trays in the reaction zone, to a depth of not more than 1/2 centimeter.

  19. Uranium metal and uranium dioxide powder and pellets - Determination of nitrogen content - Method using ammonia-sensing electrode. 1. ed.

    International Nuclear Information System (INIS)

    1994-01-01

    This International Standard specifies an analytical method for determining the nitrogen content in uranium metal and uranium dioxide powder and pellets. It is applicable to the determination of nitrogen, present as nitride, in uranium metal and uranium dioxide powder and pellets. The concentration range within which the method can be used is between 9 μg and 600 μg of nitrogen per gram. Interference can occur from metals which form complex ammines, but these are not normally present in significant amounts

  20. Some characteristics of uranium oxides in China

    International Nuclear Information System (INIS)

    Xu, Guoqing; Wang, Aizhen; Gu, Qifang; Zhang, Jingyi; Zhang, Zhaoming; Huang, Yuzhu

    1981-01-01

    According to the analytical data of seventy-seven samples from several tens of uranium ore deposits and occurrences in China, chemical properties, cell dimensions and reflectance of uranium oxides are studied. Chemical properties of uranium oxides from different types of uranium ore deposits and the influence of various mineralization ages and hosts on the compositions of uranium oxides are presented. The influence of these factor such as mineralization temperatures, the compositions of hosts and geochemical background on the compositions of uranium oxides are evident. Lead in proterozoic uranium oxides is relatively enriched by the decay of radio-active elements. Cell dimensions have positive correlation with mineralization ages, formation temperatures and concentration of rare earths and Pb, and negative correlation with the oxidation coefficient. The cell size is an exponential function of the content in CaO. It is suggested that among the factors of influence the most important is the mineralization temperature. The size of ionic radius of elements substituted U 4 + and autooxidation of U 4 + during the process of the decay of radioactive elements are of secondary importance. The reflectance is independent of the content of CaO and SiO 2 . The reflectance is positively correlative with the cell size and negatively correlative with oxidation coefficient. The relation between the reflectance and the content of PbO is logarithmic

  1. Uranium oxidation: characterization of oxides formed by reaction with water

    International Nuclear Information System (INIS)

    Fuller, E.L. Jr.; Smyrl, N.R.; Condon, J.B.; Eager, M.H.

    1983-01-01

    Three different uranium oxide samples have been characterized with respect to the different preparation techniques. Results show that the water reaction with uranium metal occurs cyclically forming laminar layers of oxide which spall off due to the strain at the oxide/metal interface. Single laminae are released if liquid water is present due to the prizing penetration at the reaction zone. The rate of reaction of water with uranium is directly proportional to the amount of adsorbed water on the oxide product. Rapid transport is effected through the open hydrous oxide product. Dehydration of the hydrous oxide irreversibly forms a more inert oxide which cannot be rehydrated to the degree that prevails in the original hydrous product of uranium oxidation with water. 27 figures

  2. Oxidation and crystal field effects in uranium

    Energy Technology Data Exchange (ETDEWEB)

    Tobin, J. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Booth, C. H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Shuh, D. K. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); van der Laan, G. [Diamond Light Source, Didcot (United Kingdom); Sokaras, D. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States); Weng, T. -C. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States); Yu, S. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bagus, P. S. [Univ. of North Texas, Denton, TX (United States); Tyliszczak, T. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Nordlund, D. [Stanford Synchrotron Radiation Lightsource, Stanford, CA (United States)

    2015-07-06

    An extensive investigation of oxidation in uranium has been pursued. This includes the utilization of soft x-ray absorption spectroscopy, hard x-ray absorption near-edge structure, resonant (hard) x-ray emission spectroscopy, cluster calculations, and a branching ratio analysis founded on atomic theory. The samples utilized were uranium dioxide (UO2), uranium trioxide (UO3), and uranium tetrafluoride (UF4). As a result, a discussion of the role of non-spherical perturbations, i.e., crystal or ligand field effects, will be presented.

  3. Standard specification for uranium oxides with a 235U content of less than 5 % for dissolution prior to conversion to nuclear-grade uranium dioxide

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This specification covers uranium oxides, including processed byproducts or scrap material (powder, pellets, or pieces), that are intended for dissolution into uranyl nitrate solution meeting the requirements of Specification C788 prior to conversion into nuclear grade UO2 powder with a 235U content of less than 5 %. This specification defines the impurity and uranium isotope limits for such urania powders that are to be dissolved prior to processing to nuclear grade UO2 as defined in Specification C753. 1.2 This specification provides the nuclear industry with a general standard for such uranium oxide powders. It recognizes the diversity of conversion processes and the processes to which such powders are subsequently to be subjected (for instance, by solvent extraction). It is therefore anticipated that it may be necessary to include supplementary specification limits by agreement between the buyer and seller. 1.3 The scope of this specification does not comprehensively cover all provisions for prevent...

  4. Standard specification for blended uranium oxides with 235U content of less than 5 % for direct hydrogen reduction to nuclear grade uranium dioxide

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This specification covers blended uranium trioxide (UO3), U3O8, or mixtures of the two, powders that are intended for conversion into a sinterable uranium dioxide (UO2) powder by means of a direct reduction process. The UO2 powder product of the reduction process must meet the requirements of Specification C 753 and be suitable for subsequent UO2 pellet fabrication by pressing and sintering methods. This specification applies to uranium oxides with a 235U enrichment less than 5 %. 1.2 This specification includes chemical, physical, and test method requirements for uranium oxide powders as they relate to the suitability of the powder for storage, transportation, and direct reduction to UO2 powder. This specification is applicable to uranium oxide powders for such use from any source. 1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of th...

  5. Failure mechanisms for compacted uranium oxide fuel cores

    International Nuclear Information System (INIS)

    Berghaus, D.G.; Peacock, H.B.

    1980-01-01

    Tension, compression, and shear tests were performed on test specimens of aluminum-clad, compacted powder fuel cores to determine failure mechanisms of the core material. The core, which consists of 70% uranium oxide in an aluminum matrix, frequently fails during post-extrusion drawing. Tests were conducted to various strain levels up to failure of the core. Sections were made of tested specimens to microscopically study initiation of failure. Two failure modes wee observed. Tensile failure mode is initiated by prior tensile failure of uranium oxide particles with the separation path strongly influenced by the arrangement of particles. Delamination mode consists of the separation of laminae formed during extrusion of tubes. Separation proceeds from fine cracks formed parallel to the laminae. Tensile failure mode was experienced in tension and shear tests. Delamination mode was produced in compression tests

  6. Study on uranium metallization yield of spent Pressurized Water Reactor fuels and oxidation behavior of fission products in uranium metals

    International Nuclear Information System (INIS)

    Choi, Ke Chon; Lee, Chang Heon; Kim, Won Ho

    2003-01-01

    Metallization yield of uranium oxide to uranium metal from lithium reduction process of spent Pressurized Water Reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metallization yield was measured. Metallization yield of the solid part was 90.7∼95.9 wt%, and the powder being 77.8∼71.5 wt% individually. Oxidation behaviour of the quarternary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At 600∼700 .deg. C, weight increments of allow of No, Ru, Rh and Pd was 0.40∼0.55 wt%. Phase change on the surface of the allow was started at 750 .deg. C. In particular, Mo was rapidly oxidized and then the alloy lost 0.76∼25.22 wt% in weight

  7. Standard test methods for chemical and mass spectrometric analysis of nuclear-grade gadolinium oxide (Gd2O3) powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 These test methods cover procedures for the chemical and mass spectrometric analysis of nuclear-grade gadolinium oxide powders to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Carbon by Direct CombustionThermal Conductivity C1408 Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method Total Chlorine and Fluorine by Pyrohydrolysis Ion Selective Electrode C1502 Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide Loss of Weight on Ignition 7-13 Sulfur by CombustionIodometric Titration Impurity Elements by a Spark-Source Mass Spectrographic C761 Test Methods for Chemical, Mass Spectrometric, Spectrochemical,Nuclear, and Radiochemical Analysis of Uranium Hexafluoride C1287 Test Method for Determination of Impurities In Uranium Dioxide By Inductively Coupled Plasma Mass Spectrometry Gadolinium Content in Gadolinium Oxid...

  8. Formation conditions of uranium minerals in oxidation zone of uranium deposits

    International Nuclear Information System (INIS)

    Li Youzhu

    2005-01-01

    The paper concerns about the summary and classification of hydrothermal uranium deposit with oxidation zone. Based on the summary of observation results of forty uranium deposits located in CIS and Bulgaria which are of different sizes and industrial-genetic types, analysis on available published information concerning oxidation and uranium mineral enrichment in supergenic zone, oxidation zone classification of hydrothermal uranium had been put forward according to the general system of the exogenetic uranium concentration. (authors)

  9. The creation of a uranium oxide industry, from the laboratory stage to a pilot plant (1961)

    International Nuclear Information System (INIS)

    Caillat, R.; Delange, M.; Sauteron, J.

    1961-01-01

    The qualities of uranium oxide, in particular its good in-pile characteristics and its resistance to corrosion by the usual heat-exchange fluids, have led to this material being chose at the present time as a nuclear fuel in many power reactors, either planned or under construction. A great effort has been made these last few years in France in studying processes for transforming powdered uranium oxide into a dense material with satisfactory behaviour in a neutron flux. The laboratories at Saclay have studied the physico-chemical features of the phenomena accompanying the calcination of uranium peroxide or ammonium uranate to give uranium trioxide, and the subsequent reduction of the latter to dioxide as well as the sintering of the powders obtained. This work has made it possible on one hand to prepare powder of known specific surface area, and on the other to show the overriding influence of this factor, all other things being equal, on the behaviour of powders during sintering in a hydrogen atmosphere. The work has led to defining two methods for sintering stoichiometric uranium oxide of high density. The technological study of the preparation of the powder and its industrial production are carried out at the plant of Le Bouchet which produces at the moment powders of known characteristics suitable for sintering in hydrogen at 1650 deg. C without prior grinding. The industrial sintering is carried out by the Compagnie industrielle des Combustibles Atomiques Frittes who has set up a pilot plant having a capacity of 25 metric tons/year, for the Commissariat l'Energie Atomique and has been operating this plant since May 1958. This plant is presented by a film entitled 'uranium oxide'. (author) [fr

  10. Improvement of removal characteristics for uranium by immobilization of diphosil powder into alginate bead

    International Nuclear Information System (INIS)

    Kim, K. J.; Kang, I. S.; Lee, Y. H.; Son, J. S.; Hong, K. P.

    2003-01-01

    Chemical wastes containing small amounts of uranium can not be disposed of them as industrial wastes. Especially for the removal of uranium, In this study, the method of immobilizing Diphosil powder within alginate beads is adopted to make a bead from powdered resin. Sodium alginate bead itself showed a capability to uptake uranium to above 60%, but the value was decreased to below 30% after equilibrium. The rate of uranium adsorption increased with increasing content of Diphosil in sodium alginate bead. Diphosil resin itself showed very fast uptake of uranium from early stages, and then the rates were leveled off. Diphosil bead showed a improved capability to uptake uranium considering Diphosil content in the bead, and a considerable potential for further applications of a continuous process by using a bead form of Diphosil

  11. Contribution to characteristics of uranium oxides

    International Nuclear Information System (INIS)

    Fritsche, R.; Dahlkamp, F.J.

    2001-01-01

    Uranium oxides from pegmatitic, metamorphic and metasomatic uranium occurrences were investigated with the objective to check for differences in their physico-chemical properties and, whether such properties are sufficiently distinct to be applied as an exploration tool. Research methods included microscopy, electron microprobe and X-ray diffractometry amended by determinations of reflectance, Vickers hardness, unit-cell dimension and oxidation grade. Tentative research results are as follows: (a) U-oxides (uraninites) of pegmatites always contain significant amounts of Th (1,5-10 wt.% ThO2). (b) U-oxides from metasomatic environments have high, but variable contents of Fe, Ca, Ti, Si and Th (around 10 wt.%), Th being low. (c) U-oxides crystallised during metamorphism contain minor impurities of the above listed elements (total of oxides < 2 wt.%). (d) Redistributed U-oxides have elevated amounts of these elements. (e) Unit-cell dimensions of U-oxides tend to reflect a complex function of formation temperature, oxidation grade and the influence of incorporated elements caused by their radius and electro-negativity. (f) A global negative correlation of unit-cell dimension and oxidation grade of uranium oxides is indicated but based on widely varying ratios of the two parameters. (g) Colloform U-oxide (pitchblende) is characterised by elevated Ca-contents (1-5 wt.% CaO) and an almost complete lack of Th (< 1 wt.% ThO2). (h) Idiomorphic U-oxide (uraninite) is commonly low in Ca (< 1.5 wt.% CaO) but contains relatively high Th values. (i) The reflectance of U-oxides generally correlates positively with Vickers hardness and unit-cell dimension, but the incorporation of other elements in the lattice of U-oxides may cause strong interference. (author)

  12. Reaction of hydrogen peroxide with uranium zirconium oxide solid solution - Zirconium hinders oxidative uranium dissolution

    Science.gov (United States)

    Kumagai, Yuta; Takano, Masahide; Watanabe, Masayuki

    2017-12-01

    We studied oxidative dissolution of uranium and zirconium oxide [(U,Zr)O2] in aqueous H2O2 solution to estimate (U,Zr)O2 stability to interfacial reactions with H2O2. Studies on the interfacial reactions are essential for anticipating how a (U,Zr)O2-based molten fuel may chemically degrade after a severe accident. The fuel's high radioactivity induces water radiolysis and continuous H2O2 generation. Subsequent reaction of the fuel with H2O2 may oxidize the fuel surface and facilitate U dissolution. We conducted our experiments with (U,Zr)O2 powder (comprising Zr:U mole ratios of 25:75, 40:60, and 50:50) and quantitated the H2O2 reaction via dissolved U and H2O2 concentrations. Although (U,Zr)O2 reacted more quickly than UO2, the dissolution yield relative to H2O2 consumption was far less for (U,Zr)O2 compared to that of UO2. The reaction kinetics indicates that most of the H2O2 catalytically decomposed to O2 at the surface of (U,Zr)O2. We confirmed the H2O2 catalytic decomposition via O2 production (quantitative stoichiometric agreement). In addition, post-reaction Raman scattering spectra of the undissolved (U,Zr)O2 showed no additional peaks (indicating a lack of secondary phase formation). The (U,Zr)O2 matrix is much more stable than UO2 against H2O2-induced oxidative dissolution. Our findings will improve understanding on the molten fuels and provide an insight into decommissioning activities after a severe accident.

  13. DEVELOPMENT OF HIGH-DENSITY U/AL DISPERSION PLATES FOR MO-99 PRODUCTION USING ATOMIZED URANIUM POWDER

    OpenAIRE

    RYU, HO JIN; KIM, CHANG KYU; SIM, MOONSOO; PARK, JONG MAN; LEE, JONG HYUN

    2013-01-01

    Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99) production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 g-U/cm3 were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compou...

  14. Studies on O/M ratio determination in uranium oxide, plutonium oxide and uranium-plutonium mixed oxide

    International Nuclear Information System (INIS)

    Sampath, S.; Chawla, K.L.

    1975-01-01

    Thermogravimetric studies were carried out in unsintered and sintered samples of uranium oxide, plutonium oxide and uranium-plutonium mixed oxide under different atmospheric conditions (air, argon and moist argon/hydrogen). Moisture loss was found to occur below 200 0 C for uranium dioxide samples, upto 700 0 C for sintered plutonium dioxide and negligible for sintered samples. The O/M ratios for non-stoichiometric uranium dioxide (sintered and unsintered), plutonium dioxide and mixed uranium and plutonium oxides (sintered) could be obtained with a precision of +- 0.002. Two reference states UOsub(2.000) and UOsub(2.656) were obtained for uranium dioxide and the reference state MOsub(2.000) was used for other cases. For unsintered plutonium dioxide samples, accurate O/M ratios could not be obtained of overlap of moisture loss with oxygen loss/gain. (author)

  15. Plutonium oxides and uranium and plutonium mixed oxides. Carbon determination

    International Nuclear Information System (INIS)

    Anon.

    Determination of carbon in plutonium oxides and uranium plutonium mixed oxides, suitable for a carbon content between 20 to 3000 ppm. The sample is roasted in oxygen at 1200 0 C, the carbon dioxide produced by combustion is neutralized by barium hydroxide generated automatically by coulometry [fr

  16. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  17. Preparation and utilization of metal oxide fine powder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joon Soo; Jang, Hee Dong; Lim, Young Woong; Kim, Sung Don; Lee, Hi Sun; Lee, Hoo In; Kim, Chul Joo; Shim, Gun Joo; Jang, Dae Kyu [Korea Inst. of Geology Mining and Materials, Taejon (Korea, Republic of)

    1995-12-01

    Metal oxide fine powders finds many applications in industry as new materials. It is very much necessary for the development of such powders to improve the domestic industry. The purpose of present research is to develop a process for the preparation and utilization of metal oxide fine powder. This project is consisted of two main subjects. (1) Production of ultrafine metal oxide powder: Ultrafine metal oxide powder is defined as a metal oxide powder of less than 100 nanometer in particle size. Experiments for the control of particle size and distributions in the various reaction system and compared with results of (2 nd year research). Various reaction systems were adopted for the development of feasible process. Ultrafine particles could be prepared even higher concentration of TiCl{sub 4} and lower gas flowrate compared to TiCl{sub 4}-O{sub 2} system in the TiCl{sub 4}-Air-H{sub 2}O system. Ultrafine Al{sub 2}O{sub 3} powders also prepared with the change of concentration and gas flowrate. Experiments on the treatment of surface characteristics of ultrafine TiO{sub 2} powders were investigated using esterification and surface treating agents. A mathematical model that can predict the particle size and distribution was also developed. (2) Preparation of cerium oxide for high-grade polishing powder: Used cerium polishing powder was recycled for preparation of high grade cerium oxide polishing powder. Also, cerium hydroxide which was generated as by-product in processing of monazite ore was used as another material. These two materials were leached respectively by using acid, and the precipitate was gained in each leached solution by adjusting pH of the solution, and by selective crystallization. These precipitates were calcined to make high grade cerium oxide polishing powder. The effect of several experimental variables were investigated, and the optimum conditions were obtained through the experiments. (author). 81 refs., 49 figs., 27 tabs.

  18. Characterization of depleted uranium oxides fabricated using different processing methods

    International Nuclear Information System (INIS)

    Hastings, E.P.; Lewis, C.; FitzPatrick, J.; Rademacher, D.; Tandon, L.

    2008-01-01

    Identifying both physical and chemical characteristics of Special Nuclear Material (SNM) production processes is the corner stone of nuclear forensics. Typically, processing markers are based on measuring an interdicted sample's bulk chemical properties, such as the elemental or isotopic composition, or focusing on the chemical and physical morphology of only a few particles. Therefore, it is imperative that known SNM processes be fully characterized from bulk to trace level for each particle size range. This report outlines a series of particle size measurements and fractionation techniques that can be applied to a bulk SNM powders, categorizing both chemical and physical properties in discrete particle size fractions. This will be demonstrated by characterizing the process signatures of a series of different depleted uranium oxides prepared at increasing firing temperatures (350-1100 deg C). Results will demonstrate how each oxides' material density, particle size distribution, and morphology varies. (author)

  19. Synthesis of vanadium oxide powders by evaporative decomposition of solutions

    International Nuclear Information System (INIS)

    Lawton, S.A.; Theby, E.A.

    1995-01-01

    Powders of the vanadium oxides V 2 O 4 , V 6 O 13 , and V 2 O 5 were produced by thermal decomposition of aqueous solutions of vanadyl sulfate hydrate in atmospheres of N 2 , H 2 mixed with N 2 , or air. The composition of the oxide powder was determined by the reactor temperature and gas composition. Residual sulfur concentrations in powders produced by decomposition at 740 C were less than 1 at.%, and these powders consisted of hollow, roughly spherical aggregates of particles less than 1 microm in diameter

  20. Determination of uranium in uranium metal, uranium oxides, and uranyl nitrate solutions by potentiometric titration

    International Nuclear Information System (INIS)

    Tucker, H.L.; McElhaney, R.J.

    1983-01-01

    A simple, fast method for the determination of uranium in uranium metal, uranium oxides, and uranyl nitrate solutions has been adapted from the Davies-Gray volumetric method to meet the needs of Y-12. One-gram duplicate aliquots of uranium metal or uranium oxide are dissolved in 1:1 HNO 3 and concentrated H 2 SO 4 to sulfur trioxide fumes, and then diluted to 100-mL volume. Duplicate aliquots are then weighed for analysis. For uranyl nitrate samples, duplicate aliquots containing between 50 and 150 mg of U are weighed and analyzed directly. The weighed aliquot is transferred to a Berzelius beaker; 1.5 M sulfamic acid is added, followed in order by concentrated phosphoric acid, 1 M ferrous sulfate, and (after a 30-second interval) the oxidizing reagent. After a timed 3-minute waiting period, 100 mL of the 0.1% vanadyl sulfate-sulfuric acid mixture is added. The sample is then titrated past its endpoint with standard potassium dichromate, and the endpoint is determined by second derivative techniques on a mV/weight basis

  1. Uranium Oxide Aerosol Transport in Porous Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

    2012-01-23

    The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

  2. Contribution to the study of the sintering of uranium oxide; Contribution a l'etude du frittage de l'oxyde d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Bel, A; Carteret, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The sintering ofnium oxide has been considered and the following factors have been particularly taken in consideration: - the particle size and the particles in shape of the initial powder, - the specific area of the initial powder, - the chemical composition of the oxide, - and the medium in which the sintering was carried out. A method of sintering uranium oxide on semi-industrial scale is presented. (author)Fren. [French] On xamine l'influence de differents facteurs sur le frittage de l'oxyde d'uranium. Sont particulierement prises en consideration: - la taille et la forme des grains de la poudre initiale, - la surface specifique de la poudre initiale, - la composition chimique de l'oxyde, - ainsi que la nature de l'atmosphere durant le frittage. D'autre part, une technique de frittage de l'oxyde d'uranium a l'echelle semi-industrielle est presentee. (auteur)

  3. Aqueous dissolution rates of uranium oxides

    International Nuclear Information System (INIS)

    Steward, S.A.; Mones, E.T.

    1994-10-01

    An understanding of the long-term dissolution of waste forms in groundwater is required for the safe disposal of high level nuclear waste in an underground repository. The main routes by which radionuclides could be released from a geological repository are the dissolution and transport processes in groundwater flow. Because uranium dioxide is the primary constituent of spent nuclear fuel, the dissolution of its matrix in spent fuel is considered the rate-limiting step for release of radioactive fission products. The purpose of our work has been to measure the intrinsic dissolution rates of uranium oxides under a variety of well-controlled conditions that are relevant to a repository and allow for modeling. The intermediate oxide phase U 3 O 8 , triuranium octaoxide, is quite stable and known to be present in oxidized spent fuel. The trioxide, UO 3 , has been shown to exist in drip tests on spent fuel. Here we compare the results of essentially identical dissolution experiments performed on depleted U 3 O 8 and dehyrated schoepite or uranium trioxide monohydrate (UO 3 ·H 2 O). These are compared with earlier work on spent fuel and UO 2 under similar conditions

  4. Hydrothermal synthesis of fine oxide powders

    Indian Academy of Sciences (India)

    Unknown

    Inorganic powders are among the most important factors in many fields of materials such as ceramics, catalysts, medicines, food, etc. There are many papers and books related to powders preparation by many authors (Veale. 1972; Kato and Yamaguchi 1983; Vincenzini 1983;. Brinker et al 1984; Johnson Jr. 1987; Messing ...

  5. Compacted and Sintered Microstructure Depending on Uranium Powder Size in Zr-U Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Chang Gun; Jun, Hyun-Joon; Ju, Jung Hwan; Lee, Ho Jin; Lee, Chong-Tak; Kim, Hyung Lae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-03-15

    In case of the uranium (U) and zirconium (Zr) powders which have been utilized for the production of a metallic fuel in the various nuclear applications, the homogenous distribution of U powders in the Zr-U pellet has influenced significantly on the nuclear fuel performance. The inhomogeneity in a powder process was changed by various intricate factors, e.g. powder size, shape, distribution and so on. Particularly, the U inhomogeneity in the Zr-U pellets occurs by segregation derived from the great gaps of densities between Zr and U during compaction of the mixed powders. In this study, the relationship between powder size and homogeneity was investigated by using the different-sized U powders. The microstructure in Zr-U pellets reveals more homogeneity when the weight ration of Zr and U powders are close to 1. In addition, homogeneous pellets which were produced by fine U powders have higher density because the homogeneity affects the alloying reaction during sintering and the densification behavior of pore induced by powder size.

  6. Spray drying of beryllium oxide powder

    International Nuclear Information System (INIS)

    Sepulveda, J.L.; Kahler, D.A.

    1991-01-01

    Forming of beryllia ceramics through dry pressing requires the agglomeration of the powder through spray drying. To produce high quality fired ceramics it is necessary to disperse/grind the primary powder prior to binder addition. Size reduction of the powder is accomplished using an aqueous system in Vibro-Energy mills (VEM) charged with beryllia media to minimize contamination. Two VEM mills of different size were used to characterize the grinding operation. Details of the grinding kinetics are described within the context of the Macroscopic Population Balance Model approach. Spray drying of the ceramic slurry was accomplished with both a centrifugal atomizer and a two fluid nozzle atomizer. Two different spray dryers were used. Important operating parameters affecting the size distribution of the spray dried powder are discussed

  7. Positron lifetime studies on thorium oxide powders

    International Nuclear Information System (INIS)

    Upadhyaya, D.D.; Muraleedharan, R.V.; Sharma, B.D.

    1982-01-01

    Positron lifetime spectra have been studied for ThO 2 powders, calcined at different temperatures and having different particle sizes. Three lifetime components could be resolved, the longest component being of low intensity. An observed strong dependence on the particle size of the annihilation process and the variation of positronium diffusion constant is explained on the basis of defect density variations in these powders. (author)

  8. Standard test method for the determination of uranium by ignition and the oxygen to uranium (O/U) atomic ratio of nuclear grade uranium dioxide powders and pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 This test method covers the determination of uranium and the oxygen to uranium atomic ratio in nuclear grade uranium dioxide powder and pellets. 1.2 This test method does not include provisions for preventing criticality accidents or requirements for health and safety. Observance of this test method does not relieve the user of the obligation to be aware of and conform to all international, national, or federal, state and local regulations pertaining to possessing, shipping, processing, or using source or special nuclear material. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. 1.4 This test method also is applicable to UO3 and U3O8 powder.

  9. Determination of trace elements in ceramic uranium dioxide pellets powders CRMs by ICP-AES

    International Nuclear Information System (INIS)

    Liu Husheng; Li Jun

    1997-01-01

    The 237-quaternary ammonium extraction resin chromatography is used to the separation of 6 trace elements in ceramic uranium dioxide pellets powders, which are used as certified reference materials (CRMs). The sample is dissolved in 6.5 mol/L HNO 3 and uranium is separated by chromatographic column. the 6 trace elements Al, Ba, Co, Ta, Ti and V contained in the elutriant are determined by using ICP directly reading spectrometer. For a 300 mg sample, the lowest determinable concentration of impurities in ceramic UO 2 pellets powders CRMs is (0.016-0.250) x 10 -6 . The relative standard deviation is less than 7.5%. The proposed method provides excellent and accurate analytical data for the ceramic UO 2 pellets powders samples (CRMs)

  10. Inhalation toxicology of industrial plutonium and uranium oxide aerosols I. Physical chemical characterization

    International Nuclear Information System (INIS)

    Eidson, A.F.; Mewhinney, J.A.

    1978-01-01

    In the fabrication of mixed plutonium and uranium oxide fuel, large quantities of dry powders are processed, causing dusty conditions in glove box enclosures. Inadvertent loss of glove box integrity or failure of air filter systems can lead to human inhalation exposure. Powdered samples and aerosol samples of these materials obtained during two fuel fabrication process steps have been obtained. A regimen of physical chemical tests of properties of these materials has been employed to identify physical chemical properties which may influence their biological behavior and dosimetry. Materials to be discussed are 750 deg. C heat-treated, mixed uranium and plutonium oxides obtained from the ball milling operation and 1750 deg. C heat-treated, mixed uranium and plutonium oxides obtained from the centerless grinding of fuel pellets. Results of x-ray diffraction studies have shown that the powder generated by the centerless grinding of fuel pellets is best described as a solid solution of UO x and PuO x consistent with its temperature history. In vitro dissolution studies of both mixed oxide materials indicate a generally similar dissolution rate for both materials. In one solvent, the material with the higher temperature history dissolves more rapidly. The x-ray diffraction and in vitro dissolution results as well as preliminary results of x-ray photoelectron spectroscopic analyses will be compared and the implications for the associated biological studies will be discussed. (author)

  11. Electrochemical oxidation of nitrite on nanodiamond powder electrode

    Energy Technology Data Exchange (ETDEWEB)

    Chen, L.H.; Zang, J.B.; Wang, Y.H.; Bian, L.Y. [State Key Laboratory of Metastable Materials Science and Technology, Yanshan University, Qinhuangdao 066004 (China); College of Materials Science and Engineering, Yanshan University, Qinhuangdao 066004 (China)

    2008-03-10

    Nanodiamond (ND) powder electrodes were fabricated and the electrochemical properties were investigated in the solution containing nitrite in this article. This electrode exhibits substantial catalytic ability toward the oxidation of nitrite anions. The electrochemical oxidation mechanism of nitrite on the ND powder electrode is discussed. The oxidation of NaNO{sub 2} is a two-electron transfer process. The electrode reaction rate constant k is estimated to be 2.013 x 10{sup -4} cm/s and (1 - {alpha})n{sub {alpha}} is 0.1643. The peak current increases linearly with the rising of the concentration of NaNO{sub 2}. (author)

  12. Superficial evolution and compacting aptitude of uranium dioxide powders

    International Nuclear Information System (INIS)

    Danroc, J.

    1982-04-01

    Long term storage of UO 2 powder improves slightly shaping and solidity of compacted powder. The aim of this work is the study of material evolution and the increase of this evolution rate for application to industrial fabrication. Aging in wet air at different temperatures is examined. Evolution of texture and superficial composition is followed. Below 80 0 C UO 3 , 2H 2 O is formed at crystal surface and thermal decomposition gives different hydrates. Kinetics of the transformation is studied. Oxidohydratation in liquid phase is rapid with hydrogen peroxide. The aged or treated material is compacted and mechanical behaviour is examined. Improvement is explained by inter-layer water molecule of the superficial hydrate giving lubricant and pseudo-plastic properties [fr

  13. The compatibility of stainless steels with particles and powders of uranium carbide and low-sulphur UCS fuels

    International Nuclear Information System (INIS)

    Venter, S.

    1978-05-01

    Slightly hyperstoichiometric (U,Pu)C is a potential nuclear fuel for fast breeder reactors. The excess carbon above the stoichiometric amount results in a higher carbon activity in the fuel, and carbon is transferred to the stainless steel cladding, resulting in embrittlement of the cladding. It is with this problem of carbon transfer from the fuel to the cladding that this thesis is concerned. For practical reasons, UC and not (U,Pu)C was used as the fuel. The theory of decarburisation of carbide fuel and the carburisation of stainless steel, the facilities constructed for the project at the Atomic Energy Board, and the experimental techniques used, including preparation of the fuels, are discussed. The effect of a number of variables of uranium carbide fuel on its compatibility behaviour with stainless steels was investigated, as well as the effect om microstructure and type of stainless steel (304, 304 L and 316) on the rate of carburisation. These studies can be briefly summarised under the following headings: powder-particle size; surface oxidation of uranium carbide; preparation temperature of uranium carbide; low sulfur UCS fuels; uranium sulfide and the microstructure and type of steel. The author concludes that: the effect of surface oxidation and particle size must be taken into account when evaluating out-of-pile tests; the possible effects of surface oxidation must be taken into account when considering vibro-compacted carbide fuels; there is no advantage in replacing a fraction of the carbon atoms by sulphur atoms in slightly hyperstoichiometric carbide fuels, and the type and thermo-mechanical treatment of the stainless steel used as cladding material in a fuel pin is not important as far as the rate of carburisation by the fuel is concerned

  14. Discussion on the interlayer oxidation and uranium metallogenesis in Qianjiadian uranium deposit, Songliao Basin

    International Nuclear Information System (INIS)

    Pang Yaqing; Chen Xiaolin; Fang Xiheng; Sun Ye

    2010-01-01

    Through systematic drill core observation, section contrast and analysis,it is proved that the ore-controlling interlayer oxidation zone of Qianjiadian uranium deposit is mainly composed by the red oxidized sandstone and locally distributed yellow and off-white sandstones. The red sandstone contains charcoal fragments, pyrite, ilmenite, siderite, which have been oxidized intensively, and it can be deduced that their original color was gray and became red due to the oxidization. The distribution of the oxidation zone is mainly controlled by the sedimentary facies,which also controll uranium metallization. The uranium orebodies mainly developed in the thinning or pinch parts of the red oxidation zone in section. On the plans, the uranium mineralization distributes near the front of the red interlayer oxidation zone. (authors)

  15. Simulation of uranium oxides reduction kinetics by hydrogen. Reactivities of germination and growth

    International Nuclear Information System (INIS)

    Brun, C.

    1997-01-01

    The aim of this work is to simulate the reduction by hydrogen of the tri-uranium octo-oxide U 3 O 8 (obtained by uranium trioxide calcination) into uranium dioxide. The kinetics curves have been obtained by thermal gravimetric analysis, the hydrogen and steam pressures being defined. The geometrical modeling which has allowed to explain the trend of the kinetics curves and of the velocity curves is an anisotropic germination-growth modeling. The powder is supposed to be formed of spherical grains with the same radius. The germs of the new UO 2 phase appear at the surface of the U 3 O 8 grains with a specific germination frequency. The growth reactivity is anisotropic and is very large in the tangential direction to the grains surface. Then, the uranium dioxide growths inside the grain and the limiting step is the grain surface. The variations of the growth reactivity and of the germination specific frequency in terms of the gases partial pressures and of the temperature have been explained by two different mechanisms. The limiting step of the growth mechanism is the desorption of water in the uranium dioxide surface. Concerning the germination mechanism the limiting step is a water desorption too but in the tri-uranium octo-oxide surface. The same geometrical modeling and the same germination and growth mechanisms have been applied to the reduction of a tri-uranium octo-oxide obtained by calcination of hydrated uranium trioxide. The values of the germination specific frequency of this solid are nevertheless weaker than those of the solid obtained by direct calcination of the uranium trioxide. (O.M.)

  16. Preparation of cerium oxide for lens polishing powder

    International Nuclear Information System (INIS)

    Injarean, Uthaiwan; Rodthongkom, Chouvana; Pichestapong, Pipat; Changkrurng, Kalaya

    2003-10-01

    Cerium is an element of rare earth group which is called lanthanide series. It is found in the ores like monazite and xenotime which are the tailings of tin mines in the south of Thailand. Cerium is used mostly as lens polishing powder besides the applications in other industries. In this study, cerium extracted from monazite ore breakdown by alkaline process was used for the preparation of lens polishing powder. Cerium hydroxide cake from the process was dissolved by hydrochloric acid and precipitated with oxalic acid. The oxalate precipitate then was calcined to oxide powder and its particle size was measured. Precipitation conditions being studied are concentration of feed cerium chloride solution, concentration of oxalic acid used for the precipitation, concentration of sulfuric acid used as precipitation control reagent and the precipitation temperature. It was found that the appropriate precipitation conditions yielded the fine oxide powder with particle size about 12μm. The oxide powder can be ground to the size of 1-3 μm which is suitable for making lens polishing powder

  17. Method and apparatus for the production of metal oxide powder

    Science.gov (United States)

    Harris, Michael T.; Scott, Timothy C.; Byers, Charles H.

    1992-01-01

    The present invention provides a method for preparing metal oxide powder. A first solution, which is substantially organic, is prepared. A second solution, which is an aqueous solution substantially immiscible in the first solution, is prepared and delivered as drops to the first solution. The drops of the second solution are atomized by a pulsed electric field forming micro-drops of the second solution. Reagents in the first solution diffuse into and react with reactants in the micro-drops of the second solution forming metal hydroxide or oxalate particles. The metal hydroxide or metal oxalate particles are then recovered and dried to produce the metal oxide powder. An apparatus for preparing a metal oxide powder is also disclosed.

  18. Method and apparatus for the production of metal oxide powder

    Science.gov (United States)

    Harris, M.T.; Scott, T.C.; Byers, C.H.

    1992-06-16

    The present invention provides a method for preparing metal oxide powder. A first solution, which is substantially organic, is prepared. A second solution, which is an aqueous solution substantially immiscible in the first solution, is prepared and delivered as drops to the first solution. The drops of the second solution are atomized by a pulsed electric field forming micro-drops of the second solution. Reagents in the first solution diffuse into and react with reactants in the micro-drops of the second solution forming metal hydroxide or oxalate particles. The metal hydroxide or metal oxalate particles are then recovered and dried to produce the metal oxide powder. An apparatus for preparing a metal oxide powder is also disclosed. 2 figs.

  19. Some characterisation profiles of uranium oxides pertaining to-ceramic fuel preparation

    International Nuclear Information System (INIS)

    El-Fekey, S.A.; Bishay, A.F.; Farah, M.Y.; Hammad, F.H.

    1978-01-01

    The effect of several variables during various stages of preparation of UO 2 powder such as precipitation, calcination and reduction on some physico-chemical properties of intermediate oxides and UO 2 powders were investigated. The methods used for characterisation were powder density, specific surface area by permeability, X-ray analysis and O/U ratio by T.G.A. The powder density of uranates was found to increase with the decrease of uranium content in uranyl nitrate solution, the increase of pH and duration of precipitation. The surface area of uranates was found to increase with the increase of pH, temperature of precipitation and the decrease of duration of precipitation. The effect of calcination for 18 hours showed that as the calcination temperature was increased for 100-800 0 C, the powder density increses and the surface area decreases. The variaton in powder density of uranates were eliminated when calcination was carried out at 700 0 C for 18 hours. The progress of the formation of intermediate oxide was followed by X-rays analysis. Two steps reduction methods, two hours at 400 0 C followed by two hours at 700 0 C, eliminated the effect of preparatory history on the 0/U ratio and produced powders having more or less the same 0/U ratio

  20. Development of metallic uranium recovery technology from uranium oxide by Li reduction and electrorefining

    International Nuclear Information System (INIS)

    Tokiwai, Moriyasu; Kawabe, Akihiro; Yuda, Ryouichi; Usami, Tsuyoshi; Fujita, Reiko; Nakamura, Hitoshi; Yahata, Hidetsugu

    2002-01-01

    The purpose of the study is to develop technology for pre-treatment of oxide fuel reprocessing through pyroprocess. In the pre-treatment process, it is necessary to reduce actinide oxide to metallic form. This paper outlines some experimental results of uranium oxide reduction and recovery of refined metallic uranium in electrorefining. Both uranium oxide granules and pellets were used for the experiments. Uranium oxide granules was completely reduced by lithium in several hours at 650degC. Reduced uranium pellets by about 70% provided a simulation of partial reduction for the process flow design. Almost all adherent residues of Li and Li 2 O were successfully washed out with fresh LiCl salt. During electrorefining, metallic uranium deposited on the iron cathode as expected. The recovery efficiencies of metallic uranium from reduced uranium oxide granules and from pellets were about 90% and 50%, respectively. The mass balance data provided the technical bases of Li reduction and refining process flow for design. (author)

  1. URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

    Science.gov (United States)

    Handwerk, J.H.; Noland, R.A.; Walker, D.E.

    1957-09-10

    In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

  2. Comparison of oxidants in alkaline leaching of uranium ore

    International Nuclear Information System (INIS)

    Sreenivas, T.; Rajan, K.C.; Srinivas, K.; Anand Rao, K.; Manmadha Rao, M.; Venkatakrishnan, R.R.; Padmanabhan, N.P.H.

    2007-01-01

    The uranium minerals occurring in various ore deposits consists of predominantly uranous ion (U +4 ), necessitating use of an oxidant and other lixiviants for efficient dissolution during leaching. Unlike acid leaching route, where uranium minerals dissolution could be achieved efficiently with cheaper lixiviants, processing of ores by alkaline leaching route involve expensive lixiviants and drastic leaching conditions. Alkaline leaching of uranium ores becomes economical only upon using cheaper and efficient oxidants and conservation of other reagents by their recycle. The present paper gives efficacy of various oxidants - KMnO 4 , NaOCl, Cu - NH 3 , air and oxygen, in the leaching of uranium from a low-grade dolostone hosted uranium ore of India. A comparison based on technical merits and cost of the oxidant chemicals is discussed. (author)

  3. Powder processing of high Tc oxide superconductors and their properties

    International Nuclear Information System (INIS)

    Vajpei, A.C.; Upadhyaya, G.S.

    1992-01-01

    Powder processing of ceramics is an established technology and in the area of high T c superconductors, its importance is felt even more significantly. The present monograph is an attempt in this direction to explore the perspectives and practice of powder processing routes towards control and optimization of the microstructure and pertinent properties of high T c oxide superconductors. The monograph consists of 6 chapters. After a very brief introduction (Chapter 1), Chapter 2 describes various classes of high T c oxide superconductors and their phase equilibria. Chapter 3 highlights the preparation of oxide superconductor powders through various routes and details their subtle distinctions. Chapter 4 briefly covers characterisation of the oxide superconductors, laying emphasis on the process-analysis and microstructure. Chapter 5 describes in detail various fabrication techniques for bulk superconductors through the powder routes. The last Chapter (Chapter 6) describing properties of bulk oxide superconductors, discusses the role of subtituents, compositional variations and processing methods on such properties. References are given at the end of each chapter. (orig.)

  4. Ultrafiltration evaluation with depleted uranium oxide

    International Nuclear Information System (INIS)

    Weisbrod, K.R.; Schake, A.R.; Morgan, A.N.; Purdy, G.M.; Martinez, H.E.; Nelson, T.O.

    1998-03-01

    Scientists at the Los Alamos National Laboratory Plutonium Facility are using electrodissolution in neutral to alkaline solutions to decontaminate oralloy parts that have surface plutonium contamination. Ultrafiltration of the electrolyte stream removes precipitate so that the electrolyte stream to the decontamination fixture is precipitate free. This report describes small-scale laboratory ultrafiltration experiments that the authors performed to determine conditions necessary for full-scale operation of an ultrafiltration module. Performance was similar to what they observed in the ferric hydroxide system. At 12 psi transmembrane pressure, a shear rate of 12,000 sec -1 was sufficient to sustain membrane permeability. Ultrafiltration of uranium(VI) oxide appears to occur as easily as ultrafiltration of ferric hydroxide. Considering the success reported in this study, the authors plan to add ultrafiltration to the next decontamination system for oralloy parts

  5. Effects of pressure on thermal transport in plutonium oxide powder

    International Nuclear Information System (INIS)

    Bielenberg, Patricia; Prenger, F. Coyne; Veirs, Douglas Kirk; Jones, Jerry

    2004-01-01

    Radial temperature profiles in plutonium oxide (PuO 2 ) powder were measured in a cylindrical vessel over a pressure range of 0.055 to 334.4 kPa with two different fill gases, helium and argon. The fine PuO 2 powder provides a very uniform self-heating medium amenable to relatively simple mathematical descriptions. At low pressures ( 2 powder has small particle sizes (on the order of 1 to 10 μm), random particle shapes, and high porosity so a more general model was required for this system. The model correctly predicts the temperature profiles of the powder over the wide pressure range for both argon and helium as fill gases. The effective thermal conductivity of the powder bed exhibits a pressure dependence at higher pressures because the pore sizes in the interparticle contact area are relatively small (less than 1 μm) and the Knudsen number remains above the continuum limit at these conditions for both fill gases. Also, the effective thermal conductivity with argon as a fill gas is higher than expected at higher pressures because the solid pathways account for over 80% of the effective powder conductivity. The results obtained from this model help to bring insight to the thermal conductivity of very fine ceramic powders with different fill gases.

  6. Oxidation-extraction of uranium from wet-process phosphoric acid

    International Nuclear Information System (INIS)

    Lawes, B.C.

    1985-01-01

    The invention involves an improvement to the reductive stripping process for recovering uranium values from wet-process phosphoric acid solution, where uranium in the solution is oxidized to uranium (VI) oxidation state and then extracted from the solution by contact with a water immiscible organic solvent, by adding sufficient oxidant, hydrogen peroxide, to obtain greater than 90 percent conversion of the uranium to the uranium (VI) oxidation state to the phosphoric acid solution and simultaneously extracting the uranium (VI)

  7. Sintering uranium oxide using a preheating step

    International Nuclear Information System (INIS)

    Jensen, N.J.; Nivas, Y.; Packard, D.R.

    1977-01-01

    Compacted pellets of uranium oxide or uranium oxide with one or more additives are heated in a kiln in a process having a preheating step, a sintering step, a reduction step, and a cooling step in a controlled atmosphere. The process is practiced to give a range of temperature and atmosphere conditions for obtaining optimum fluoride removal from the compacted pellets along with optimum sintering in a single process. The preheating step of this process is conducted in a temperature range of about 600 0 to about 900 0 C and the pellets are held for at least twenty min, and preferably about 60 min, in an atmosphere having a composition in the range of about 10 to about 75 vol % hydrogen with the balance being carbon dioxide. The sintering step is conducted at a temperature in the range of about 900 0 C to 1500 0 C in the presence of an atmosphere having a composition in the range of about 0.5 to about 90 vol % hydrogen with the balance being carbon dioxide. The reduction step reduces the oxygen to metal ratio of the pellets to a range of about 1.98 to 2.10:1 and this is accomplished by gradually cooling the pellets for about 30 to about 120 min from the temperature of the sintering step to about 1100 0 C in an atmosphere of about 10 to 90 vol % hydrogen with the balance being carbon dioxide. Thereafter the pellets are cooled to about 100 0 C under a protective atmosphere, and in one preferred practice the same atmosphere used in the reduction step is used in the cooling step. The preheating, sintering and reduction steps may also be conducted with their respective atmospheres having an initial additional component of water vapor and the water vapor can comprise up to about 20 vol %

  8. DEVELOPMENT OF HIGH-DENSITY U/AL DISPERSION PLATES FOR MO-99 PRODUCTION USING ATOMIZED URANIUM POWDER

    Directory of Open Access Journals (Sweden)

    HO JIN RYU

    2013-12-01

    Full Text Available Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99 production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 g-U/cm3 were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compounds by a chemical reaction of the uranium particles and aluminum matrix. Thus, these target plates can be treated with the same alkaline dissolution process that is used for conventional UAlx dispersion targets, while increasing the uranium density in the target plates

  9. Treatment of uranium turning with the controllable oxidizing process

    International Nuclear Information System (INIS)

    Shen Bingyi; Zhang Yonggang; Zhen Huikuan

    1989-02-01

    The concept, procedure and safety measures of the controllable oxidizing for uranium turning is described. The feasibility study on technological process has been made. The process provided several advantages such as: simplicity of operation, no pollution environment, safety, high efficiency and low energy consumption. The process can yield nuclear pure uranium dioxide under making no use of a great number of chemical reagent. It may supply raw material for fluoration and provide a simply method of treatment for safe store of uranium turning

  10. Effect of CO on surface oxidation of uranium metal

    International Nuclear Information System (INIS)

    Wang, X.; Fu, Y.; Xie, R.

    1997-01-01

    The surface reactions of uranium metal with carbon monoxide at 25 and 200 deg C have been studied by X-ray photoelectron spectroscopy (XPS);respectively. Adsorption of carbon monoxide on the surface layer of uranium metal leads to partial reduction of surface oxide and results in U4f photoelectron peak shifting to the lower binding energy. The content of oxygen in the surface oxide is decreased and O1s/O4f ratio decreases with increasing the exposure of carbon monoxide. The investigation indicates the surface layer of uranium metal has resistance to further oxidation in the atmosphere of carbon monoxide. (author)

  11. Oxidizing attack process of uranium ore by a carbonated liquor

    International Nuclear Information System (INIS)

    Maurel, Pierre; Nicolas, Francois.

    1981-01-01

    A continuous process for digesting a uraniferous ore by oxidation with a recycling aqueous liquor containing alkaline carbonates and bicarbonates in solution as well as uranium in a concentration close to its solubility limit at digestion temperature, and of recuperation of the precipitated uranium within the solid phase remaining after digestion. The digestion is carried out by spraying oxygen into the hot reactional medium in order not only to permit oxidation of the uranium and its solubilization but also to ensure that the sulphides of impurities and organic substances present in the ore are oxidized [fr

  12. Uranium induces oxidative stress in lung epithelial cells

    International Nuclear Information System (INIS)

    Periyakaruppan, Adaikkappan; Kumar, Felix; Sarkar, Shubhashish; Sharma, Chidananda S.; Ramesh, Govindarajan T.

    2007-01-01

    Uranium compounds are widely used in the nuclear fuel cycle, antitank weapons, tank armor, and also as a pigment to color ceramics and glass. Effective management of waste uranium compounds is necessary to prevent exposure to avoid adverse health effects on the population. Health risks associated with uranium exposure includes kidney disease and respiratory disorders. In addition, several published results have shown uranium or depleted uranium causes DNA damage, mutagenicity, cancer and neurological defects. In the current study, uranium toxicity was evaluated in rat lung epithelial cells. The study shows uranium induces significant oxidative stress in rat lung epithelial cells followed by concomitant decrease in the antioxidant potential of the cells. Treatment with uranium to rat lung epithelial cells also decreased cell proliferation after 72 h in culture. The decrease in cell proliferation was attributed to loss of total glutathione and superoxide dismutase in the presence of uranium. Thus the results indicate the ineffectiveness of antioxidant system's response to the oxidative stress induced by uranium in the cells. (orig.)

  13. Plasma metallization of aluminium oxide powder

    International Nuclear Information System (INIS)

    Smirnov, A.I.; Petrunichev, V.A.

    1981-01-01

    The sintering ability of cermets of metallized granulas of aluminium and matrix materials, such as chromium, nickel and nichrome is studied. Deformation tests of samples of cermets with molybdenum coated granules show satisfactory results at normal and high temperatures without fracture of metall-oxide interfaces [ru

  14. Chlorination of uranium oxides with CCl4 using a mechanochemical method

    Science.gov (United States)

    Kitawaki, Shinichi; Nagai, Takayuki; Sato, Nobuaki

    2013-08-01

    A chlorination method for uranium oxides at low temperature was investigated by using a mechanochemical method. In particular, the possibility of the chlorination of uranium oxides, such as UO2 and U3O8, via mechanochemical reaction with CCl4 was studied using a planetary ball mill. Mechanochemical experiments were conducted to evaluate the effect of milling time, CCl4/uranium oxide molar ratio, and revolution speed on the reaction. The synthesized products were then subjected to X-ray diffraction analysis, and it was found that the chlorination of U3O8 with CCl4 to UOCl2, UCl4, and U2O2Cl5 proceeded. However, the chlorination reaction could not be observed when using UO2 powder as the raw material. The chlorination reaction could not be observed when using UO2 powder as the raw material. The chlorination of U3O8 with CCl4 to form UOCl2, UCl4, and U2O2Cl5 via mechanochemical reaction occurs at room temperature. The ratio of chlorination increases with milling time when the appropriate amount of CCl4 is employed. However, the use of excess liquid CCl4 decreases the mechanochemical effect.

  15. Chlorination of uranium oxides with CCl4 using a mechanochemical method

    International Nuclear Information System (INIS)

    Kitawaki, Shinichi; Nagai, Takayuki; Sato, Nobuaki

    2013-01-01

    Highlights: • UCl 4 or UOCl 2 could be synthesized from U 3 O 8 with CCl 4 by using a planetary ball mill. • The chlorination could not be observed when using UO 2 powder as the starting material. • Extension of milling time was effective for chlorinating U 3 O 8 with the appropriate amount of CCl 4 . -- Abstract: A chlorination method for uranium oxides at low temperature was investigated by using a mechanochemical method. In particular, the possibility of the chlorination of uranium oxides, such as UO 2 and U 3 O 8 , via mechanochemical reaction with CCl 4 was studied using a planetary ball mill. Mechanochemical experiments were conducted to evaluate the effect of milling time, CCl 4 /uranium oxide molar ratio, and revolution speed on the reaction. The synthesized products were then subjected to X-ray diffraction analysis, and it was found that the chlorination of U 3 O 8 with CCl 4 to UOCl 2 , UCl 4 , and U 2 O 2 Cl 5 proceeded. However, the chlorination reaction could not be observed when using UO 2 powder as the raw material

  16. Chlorination of uranium oxides with CCl{sub 4} using a mechanochemical method

    Energy Technology Data Exchange (ETDEWEB)

    Kitawaki, Shinichi, E-mail: kitawaki.shinichi@jaea.go.jp [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai, Naka, Ibaraki 319-1194 (Japan); Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Miyagi 980-8577 (Japan); Nagai, Takayuki [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai, Naka, Ibaraki 319-1194 (Japan); Sato, Nobuaki [Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Miyagi 980-8577 (Japan)

    2013-08-15

    Highlights: • UCl{sub 4} or UOCl{sub 2} could be synthesized from U{sub 3}O{sub 8} with CCl{sub 4} by using a planetary ball mill. • The chlorination could not be observed when using UO{sub 2} powder as the starting material. • Extension of milling time was effective for chlorinating U{sub 3}O{sub 8} with the appropriate amount of CCl{sub 4}. -- Abstract: A chlorination method for uranium oxides at low temperature was investigated by using a mechanochemical method. In particular, the possibility of the chlorination of uranium oxides, such as UO{sub 2} and U{sub 3}O{sub 8}, via mechanochemical reaction with CCl{sub 4} was studied using a planetary ball mill. Mechanochemical experiments were conducted to evaluate the effect of milling time, CCl{sub 4}/uranium oxide molar ratio, and revolution speed on the reaction. The synthesized products were then subjected to X-ray diffraction analysis, and it was found that the chlorination of U{sub 3}O{sub 8} with CCl{sub 4} to UOCl{sub 2}, UCl{sub 4}, and U{sub 2}O{sub 2}Cl{sub 5} proceeded. However, the chlorination reaction could not be observed when using UO{sub 2} powder as the raw material.

  17. Oxide Dispersion Strengthened Iron Aluminide by CVD Coated Powders

    Energy Technology Data Exchange (ETDEWEB)

    Asit Biswas Andrew J. Sherman

    2006-09-25

    This I &I Category2 program developed chemical vapor deposition (CVD) of iron, aluminum and aluminum oxide coated iron powders and the availability of high temperature oxidation, corrosion and erosion resistant coating for future power generation equipment and can be used for retrofitting existing fossil-fired power plant equipment. This coating will provide enhanced life and performance of Coal-Fired Boilers components such as fire side corrosion on the outer diameter (OD) of the water wall and superheater tubing as well as on the inner diameter (ID) and OD of larger diameter headers. The program also developed a manufacturing route for readily available thermal spray powders for iron aluminide coating and fabrication of net shape component by powder metallurgy route using this CVD coated powders. This coating can also be applid on jet engine compressor blade and housing, industrial heat treating furnace fixtures, magnetic electronic parts, heating element, piping and tubing for fossil energy application and automotive application, chemical processing equipment , heat exchanger, and structural member of aircraft. The program also resulted in developing a new fabrication route of thermal spray coating and oxide dispersion strengthened (ODS) iron aluminide composites enabling more precise control over material microstructures.

  18. Elaboration and characterisation of yttrium oxide and hafnium oxide powders by the sol-gel process

    International Nuclear Information System (INIS)

    Hours, T.

    1988-01-01

    The two classical sol-gel processes, colloidal and polymeric are studied for the preparation of yttrium oxide and hafnium oxide high performance powders. In the colloidal process, controlled and reproducible conditions for the preparation of yttrium oxide and hafnium oxide sols from salts or alkoxides are developed and the hydrothermal synthesis monodisperse hafnium oxide colloids is studied. The polymeric process is studied with hafnium ethyl-hexylate, hydrolysis kinetics for controlled preparation of sols and gels is investigated. Each step of preparation is detailed and powders obtained are characterized [fr

  19. Design of a uranium-dioxide powder spheroidization system by plasma processing

    Science.gov (United States)

    Cavender, Daniel

    The plasma spheroidization system (PSS) is the first process in the development of a tungsten-uranium dioxide (W-UO2) ceramic-metallic (cermet) fuel for nuclear thermal rocket (NTR) propulsion. For the purposes of fissile fuel retention, UO2 spheroids ranging in size from 50 - 100 micrometers (μm) in diameter will be encapsulated in a tungsten shell. The PSS produces spherical particles by melting angular stock particles in an argon-hydrogen plasma jet where they become spherical due to surface tension. Surrogate CeO 2 powder was used in place of UO2 for system and process parameter development. Stock and spheroidized powders were micrographed using optical and scanning electron microscopy and evaluated by statistical methods to characterize and compare the spherocity of pre and post process powders. Particle spherocity was determined by irregularity parameter. Processed powders showed a statistically significant improvement in spherocity, with greater that 60% of the examined particles having an irregularity parameter of equal to or lower than 1.2, compared to stock powder.

  20. Uranium oxidation: Characterization of oxides formed by reaction with water by infrared and sorption analyses

    Science.gov (United States)

    Fuller, E. L.; Smyrl, N. R.; Condon, J. B.; Eager, M. H.

    1984-04-01

    Three different uranium oxide samples have been characterized with respect to the different preparation techniques. The results show that the water reaction with uranium metal occurs cyclically forming laminar layers of oxide which spall off due to the strain at the oxide/metal interface. Single laminae are released if liquid water is present due to the prizing penetration at the reaction zone. The rate of reaction of water with uranium is directly proportional to the amount of adsorbed water on the oxide product. Rapid transport is effected through the open hydrous oxide product. Dehydration of the hydrous oxide irreversibly forms a more inert oxide which cannot be rehydrated to the degree that prevails in the original hydrous product of uranium oxidation with water. Inert gas sorption analyses and diffuse reflectance infrared studies combined with electron microscopy prove valuable in defining the chemistry and morphology of the oxidic products and hydrated intermediates.

  1. Precision determination of uranium in uranium oxide by constant-current coulometry

    International Nuclear Information System (INIS)

    Xu Laili; Wang Chunhuan.

    1990-01-01

    A method of constant-current coulometric titration for determination of uranium in uranium oxide is described. This method involves preliminary reduction of U (VI) in H 2 SO 4 -H 3 PO 4 medium by Cr (II) as a reductant, followed by air oxidation of excess of Cr (II), addition of solid K 2 Cr 2 O 7 in quantity slightly more than that of the required for quantitative oxidation of U (IV) and final titration of excess of K 2 Cr 2 O 7 with the electrogenerated Fe (II). The endpoint is determined amperometrically. The effect of various factors on the sample treatment and reduction-oxidation processes has been examined. The precision of the method as indicated by the standard deviation of an individual observation is less than 0.01% for l gram uranium oxide

  2. Formation of corrosion-resistant oxide film on uranium

    International Nuclear Information System (INIS)

    Petit, G.S.

    1976-01-01

    A vacuum heat-treatment method was developed for coating metallic uranium with an adherent protective film of uranium oxide. The film is prepared by vacuum heat-treating the metallic uranium at 625 0 C for 1 h while controlling the amount of oxygen being metered into the furnace. Uranium coupons with the protective film were exposed for several hundred hours in a corrosion test bath at 95 0 C and 100 percent RH without corroding. Film thicknesses ranging from 5 to 25 μm (0.0002 to 0.001 in.) were prepared and corrosion tested; the film thickness can be controlled to less than +-2.5 μm (+-0.0001 in.). The oxide film is hard, nonwetting, and very adherent. The resulting surface finish of the metal is equivalent to that of the original finish. The advantages of the oxide films over other protective coatings are given. 12 fig

  3. Determination of uranium content and its impurities in the AUC and UO2 powders

    International Nuclear Information System (INIS)

    Boybul; Arif Nugroho

    2012-01-01

    The analysis of uranium (U) content and its impurities in the ammonium uranyl carbonate (AUC) and uranium dioxide (UO 2 ) produced from research reactor fuel element production installation, PT. BATAN Teknologi have been carried out. Uranium content in the powders was analyzed by potentiometric titration methods and impurity contents was analyzed by atomic absorption spectrophotometer (AAS) and by inductively coupled plasma-atomic emission spectroscopy (ICP-AES). The purpose of this study was to determine of impurity elements in the AUC and UO 2 powder resulting from the production process if it meets the required specifications. It is reported that U content in the AUC is 48.62 wt% and that in the UO 2 is 88.08 wt%. The precision and accuracy analysis of the U content is 0,235% and 0,151%. In case of impurities in the AUC powders, it is reported that the analytical results of Zn, Ni, Cd, Co, Mn, Mg, Fe, Cu and Cr at 10.15 ppm, 1.12 ppm, not detection, not detection, not detection, 0.30 ppm, 216.07 ppm, not detection, and 31.36 ppm, respectively, while that UO 2 are 11.31 ppm, 72.14 ppm, not detection, not detection, 6.25 ppm, 8.65 ppm, 298.24 ppm, 12.75 ppm and 32, 23 ppm. The U and impurity contents in both the AUC and UO 2 fulfill the specification of nuclear fuel for RSG-GAS research reactor. (author)

  4. Study of reactions between uranium-plutonium mixed oxide and uranium nitride and between uranium oxide and uranium nitride; Etude des reactions entre l`oxyde mixte d`uranium-plutonium et le nitrure d`uranium et entre l`oxyde d`uranium et le nitrure d`uranium

    Energy Technology Data Exchange (ETDEWEB)

    Lecraz, C

    1993-06-11

    A new type of combustible elements which is a mixture of uranium nitride and uranium-plutonium oxide could be used for Quick Neutrons Reactors. Three different studies have been made on the one hand on the reactions between uranium nitride (UN) and uranium-plutonium mixed oxide (U,Pu)O{sub 2}, on the other hand on these between UN and uranium oxide UO{sub 2}. They show a sizeable reaction between nitride and oxide for the studied temperatures range (1573 K to 1973 K). This reaction forms a oxynitride compound, MO{sub x} N{sub y} with M=U or M=(U,Pu), whose crystalline structure is similar to oxide`s. Solubility of nitride in both oxides is studied, as the reaction kinetics. (TEC). 32 refs., 48 figs., 22 tabs.

  5. Spectrophotometric study of bio-sorption of uranium on glass grade spodumene shell powder

    International Nuclear Information System (INIS)

    Parakudyil, A.S.; Pillai, A.K.; Reddy, A.V.R.; Singal, R.K.; Sharma, P.K.; Michael, K.M.

    2012-01-01

    Separation of uranium found in iron ore leachates was done by extraction chromatography using glass grade spodumene shellpowder (GSS) in nitric acid medium and analyzed spectrophotometrically. The influences of metal ion concentration, pH and adsorption capacity of biomass were investigated. Biosorption is a potential method of separation of heavy and trace metals from waste water and effluents from various sources. The adsorption capacities of biomass were investigated by batch experiments and column experiments. In the present study, glass grade spodumene shell powder (GSS) in acidic medium is being used as a biosorbent

  6. ( Citrus maxima ) peel powder on lipid oxidation of

    African Journals Online (AJOL)

    The presence of phenolic compounds in Shaddock peel powder (SPP) and its effect on lipid oxidation in cooked and raw semimembranosus (SM) muscle in goats were examined. The SPP was applied to cooked and raw minced SM at the rate of 0, 0.5, 1.0 and 1.5% of the weight of the meat and compared with 0.01% ...

  7. Standard specification for nuclear-grade zirconium oxide powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification defines the physical and chemical requirements for zirconium oxide powder intended for fabrication into shapes, either entirely or partially of zirconia, for use in a nuclear reactor core. 1.2 The material described herein shall be particulate in nature. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  8. Standard specification for nuclear-grade beryllium oxide powder

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This specification defines the physical and chemical requirements of nuclear-grade beryllium oxide (BeO) powder to be used in fabricating nuclear components. This specification does not include requirements for health and safety. It recognizes the material as a Class B poison and suggests that producers and users become thoroughly familiar with and comply to applicable federal, state and local regulations and handling guidelines. Special tests and procedures are given

  9. PREPARATION OF METAL OXIDE POWDERS FROM METAL LOADED VERSATIC ACID

    OpenAIRE

    KAKIHATA, Takayuki; USAMI, Kensuke; YAMAMOTO, Hideki; SHIBATA, Junji

    1998-01-01

    A production process for metal oxide powders was developed using a solvent extraction method. Versatic Acid 10 and D2EHPA solutions containing copper, zinc and nickel were used for a precipitation-stripping process, where oxalic acid was added to the solution as a precipitation reagent.Copper, zinc and nickel oxalates were easily formed in an aqueous phase, and 99.9% of precipitation was obtained for each metal during this process. These metal oxalates were easily converted to metal oxides by...

  10. Kinetics of the reduction of uranium oxide catalysts

    International Nuclear Information System (INIS)

    Heynen, H.W.G.; Camp-van Berkel, M.M.; Bann, H.S. van der

    1977-01-01

    The reduction of uranium oxide and uranium oxide on alumina catalysts by ethylbenzene and by hydrogen has been studied in a thermobalance. Ethylbenzene mole fractions between 0.0026 and 0.052 and hydrogen mole fractions between 0.1 and 0.6 were applied at temperatures of 425--530 0 C. During the reduction the uranium oxides are converted into UO 2 . The rate of reduction of pure uranium oxide appears to be constant in the composition region UO/sub 2.6/-UO/sub 2.25/. The extent of this region is independent of the concentration of the reducing agents and of the reaction temperature. The constant rate is explained in terms of a constant oxygen pressure which is in equilibrium with the two solid phases, U 3 O/sub 8-x/ and U 4 O 9 . The reduction rate is first order in hydrogen and zero order in ethylbenzene with activation energies of 120 and 190 kJ mol -1 , respectively. Oxygen diffusion through the lattice is probably not rate limiting. The reduction behavior of uranium oxide on alumina is different from that of pure uranium oxide; the rate of reduction continuously decreases with increasing degree of reduction. An explanation for this behavior has been given by visualizing this catalyst as a set of isolated uranium oxide crystallites with a relative wide variation of diameters, in an alumina matrix. At the beginning of the reduction, carbon dioxide and water are the only reaction products. Thereafter, benzene is found as well and, finally, at U/O ratios below 2.25, styrene also appears in the reactor outlet

  11. A simple economic and quantitative method for the removal of uranium from Gogi mine water using powdered red brick

    International Nuclear Information System (INIS)

    Nathan, Usha; Cyriac, Bincy; Hegde, G.N.; Premadas, A.; Rai, A.K.

    2011-01-01

    A simple and economical method for the removal of uranium from Gogi mine water using the powdered red brick as a good adsorbent is discussed. Preliminary studies for the removal of uranium using brick showed encouraging results. Further studies were carried to find the amount and size of brick for the quantitative removal of uranium. The results of these studies showed that 50 g of brick with 10 mesh size was enough to remove uranium quantitatively from 100 ml of mine water containing 1800 μg/L of uranium. However the column studies indicated considerable decrease (∼ 5 g for 100 ml of mine water) in the amount of brick required to remove uranium from 100 ml of mine water

  12. Heat-induced redistribution of surface oxide in uranium

    International Nuclear Information System (INIS)

    Swissa, E.; Shamir, N.; Bloch, J.; Mintz, M.H.; Israel Atomic Energy Commission, Beersheba. Nuclear Research Center-Negev)

    1990-01-01

    The redistribution of oxygen and uranium metal at the vicinity of the metal-oxide interface of native and grown oxides due to vacuum thermal annealing was studied for uranium and uranium-chromium alloy using Auger depth profiling and metallographic techniques. It was found that uranium metal is segregating out through the uranium oxide layer for annealing temperatures above 450deg C. At the same time the oxide is redistributed in the metal below the oxide-metal interface in a diffusion like process. By applying a diffusion equation of a finite source, the diffusion coefficients for the process were obtained from the oxygen depth profiles measured for different annealing times. An Arrhenius like behavior was found for the diffusion coefficient between 400 and 800deg C. The activation energy obtained was E a =15.4±1.9 kcal/mole and the pre-exponential factor, D 0 =1.1x10 -8 cm 2 /s. An internal oxidation mechanism is proposed to explain the results. (orig.)

  13. Heat-induced redistribution of surface oxide in uranium

    Science.gov (United States)

    Swissa, Eli; Shamir, Noah; Mintz, Moshe H.; Bloch, Joseph

    1990-09-01

    The redistribution of oxygen and uranium metal at the vicinity of the metal-oxide interface of native and grown oxides due to vacuum thermal annealing was studied for uranium and uranium-chromium alloy using Auger depth profiling and metallographic techniques. It was found that uranium metal is segregating out through the uranium oxide layer for annealing temperatures above 450°C. At the same time the oxide is redistributed in the metal below the oxide-metal interface in a diffusion like process. By applying a diffusion equation of a finite source, the diffusion coefficients for the process were obtained from the oxygen depth profiles measured for different annealing times. An Arrhenius like behavior was found for the diffusion coefficient between 400 and 800°C. The activation energy obtained was Ea = 15.4 ± 1.9 kcal/mole and the pre-exponential factor, D0 = 1.1 × 10 -8cm2/ s. An internal oxidation mechanism is proposed to explain the results.

  14. Exploration and discovery of the Pine Ridge uranium deposits, Powder River Basin, Wyoming, USA

    International Nuclear Information System (INIS)

    Doelger, M.

    2014-01-01

    The Pine Ridge uranium deposits are named for a newly identified area between the Pumpkin Buttes and Southern Powder River Basin (PRB) mining districts. This regional prospect, covering nine contiguous townships, is northwest of the Cameco Smith Ranch mine and west of the Uranium One Allemand-Ross project in Converse County, Wyoming. Surface mapping and 350+ measured sections of well exposed outcrops have identified 250 target sandstones and contributed to a model of the complex braided stream channel architecture within the Eocene Watsatch and Paleocene Fort Union Formations. The uranium-bearing sandstones occur in 3- D bundles of vertically aggrading river systems flowing into the PRB from distant uranium source areas of the Granite Mountains to the west and the northern Laramie Range to the south. Large volumes of mudstone overbank and swamp facies separate the individual river systems laterally, resulting in greater vertical reservoir continuity from sandstones stacking. At least five major paleo river systems have been identified and named. High organic content, within the host formations, and rising veils of hydrocarbon gases from underlying oil and gas deposits have resulted in classic roll front uranium deposits in individual sandstones and intervals. Mineralization in stacked sandstone bundles several hundred feet thick show a crescent-shaped distribution within the shallow mineralized interval “attic”, the “cellar” at the base of the alteration cell, and the furthest basin-ward “front door”. World-class uranium resource potential has been identified along 208 miles of redox boundary string length mapped from the 1522 control points consisting of outcrop data, pre-existing uranium drilling, oil and gas wells, and proprietary drilling in 2012 and 2013 by Stakeholder. All data is managed in ARC VIEW GIS with 3-D capability, which will be demonstrated. Very few restrictions apply to the project area. Uranium holes are permitted solely by the

  15. Standard specification for sintered gadolinium oxide-uranium dioxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification is for finished sintered gadolinium oxide-uranium dioxide pellets for use in light-water reactors. It applies to gadolinium oxide-uranium dioxide pellets containing uranium of any 235U concentration and any concentration of gadolinium oxide. 1.2 This specification recognizes the presence of reprocessed uranium in the fuel cycle and consequently defines isotopic limits for gadolinium oxide-uranium dioxide pellets made from commercial grade UO2. Such commercial grade UO2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unirradiated uranium. UO2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design. 1.3 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aw...

  16. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean M

    2011-04-29

    outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis

  17. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    International Nuclear Information System (INIS)

    McDeavitt, Sean M.

    2011-01-01

    beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A - MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled 'Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled 'Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors' Appendix B - External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, 'Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, 'Uranium Powder Production Using a Hydride-Dehydride Process,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C - Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys' presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow

  18. Extraction separation studies of uranium(VI) by amine oxides

    International Nuclear Information System (INIS)

    Ejaz, M.

    1975-01-01

    The extraction of uranium(VI) by two amine oxides, 4-(5-nonyl)pyridine oxide and trioctylamine oxide has been studied. The extraction behavior of these two N-oxides is compared. The dependence of extraction on the type of amine oxide and acid, nature of organic diluent, and amine oxide concentration has been investigated. The influence of the concentration of the metal and salting-out agents is described. The possible mechanism of extraction is discussed in the light of the results of extraction isotherms, loading radiodata, and log-log plots of amine oxide concentration vs distribution ratio. The separation factors for a number of metal ions are reported, and the separation of uranium from some fission elements has also been achieved

  19. Simulation of uranium and plutonium oxides compounds obtained in plasma

    Science.gov (United States)

    Novoselov, Ivan Yu.; Karengin, Alexander G.; Babaev, Renat G.

    2018-03-01

    The aim of this paper is to carry out thermodynamic simulation of mixed plutonium and uranium oxides compounds obtained after plasma treatment of plutonium and uranium nitrates and to determine optimal water-salt-organic mixture composition as well as conditions for their plasma treatment (temperature, air mass fraction). Authors conclude that it needs to complete the treatment of nitric solutions in form of water-salt-organic mixtures to guarantee energy saving obtainment of oxide compounds for mixed-oxide fuel and explain the choice of chemical composition of water-salt-organic mixture. It has been confirmed that temperature of 1200 °C is optimal to practice the process. Authors have demonstrated that condensed products after plasma treatment of water-salt-organic mixture contains targeted products (uranium and plutonium oxides) and gaseous products are environmental friendly. In conclusion basic operational modes for practicing the process are showed.

  20. Studies on the fluorination of tri uranium octa oxide to Uranium tetrafluoride

    Energy Technology Data Exchange (ETDEWEB)

    Rofail, N H; Elfekey, S A [Nuclear chemistry department, hot laboratories centre, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    Uranium tetrafluoride suitable for both uranium metal and hexafluoride preparations, was prepared by fluorination of U{sub 3} O{sub 8} with C F{sub 2} Cl{sub 2}. It was found that the oct oxide must have certain physical and chemical specifications to satisfy the specifications needed for subsequent operations. X-ray diffraction analysis, infra red investigations and chemical analysis confirm that the obtained uranium tetrafluoride contains more than 97% of U F{sub 4} with tap density equals to 3.5 g/cc. 3 FIGS., 2 TABS.

  1. The ignitability potential of uranium {open_quotes}roaster oxide{close_quotes}

    Energy Technology Data Exchange (ETDEWEB)

    Stakebake, J.L.

    1994-11-01

    The oxidation of uranium to form Uranium `roaster oxide` was investigated with respect to concerns of unreacted metal remaining in the roaster oxide matrix. It was found that ignition of unreacted uranium chips in the roaster oxide as synthesized is unlikely under normal storage conditions.

  2. Studies on yttrium oxide coatings for corrosion protection against molten uranium

    International Nuclear Information System (INIS)

    Chakravarthy, Y.; Bhandari, Subhankar; Pragatheeswaran; Thiyagarajan, T.K.; Ananthapadmanabhan, P.V.; Das, A.K.; Kumar, Jay; Kutty, T.R.G.

    2012-01-01

    Yttrium oxide is resistant to corrosion by molten uranium and its alloys. Yttrium oxide is recommended as a protective oxide layer on graphite and metal components used for melting and processing uranium and its alloys. This paper presents studies on the efficacy of plasma sprayed yttrium oxide coatings for barrier applications against molten uranium

  3. Pentavalent Uranium Chemistry - Synthetic Pursuit Of A Rare Oxidation State

    Energy Technology Data Exchange (ETDEWEB)

    Graves, Christopher R [Los Alamos National Laboratory; Kiplinger, Jaqueline L [Los Alamos National Laboratory

    2009-01-01

    This feature article presents a comprehensive overview of pentavalent uranium systems in non-aqueous solution with a focus on the various synthetic avenues employed to access this unusual and very important oxidation state. Selected characterization data and theoretical aspects are also included. The purpose is to provide a perspective on this rapidly evolving field and identify new possibilities for future developments in pentavalent uranium chemistry.

  4. Fundamental study on recovery uranium oxide from HEPA filters

    International Nuclear Information System (INIS)

    Izumida, T.; Noguchi, Y.

    1993-01-01

    Large numbers of spent HEPA filters are produced at uranium fuel fabrication facilities. Uranium oxide particles have been collected on these filters. Then, a spent HEPA filter treatment system was developed from the viewpoint of recovering the UO 2 and minimizing the volume. The system consists of a mechanical separation process and a chemical dissolution process. This paper describes the results of fundamental experiments on recovering UO 2 from HEPA filters

  5. Method and device to produce pourable, directly pressable uranium dioxide powder. Verfahren und Vorrichtung zur Herstellung von rieselfaehigem, direkt verpressbarem Urandioxid-Pulver

    Energy Technology Data Exchange (ETDEWEB)

    Boerner, P.; Isensee, H.J.; Vietzke, H.

    1978-08-17

    The uranium dioxide powder is produced from uranium peroxide which is obtained by continuous precipitation of uranyl nitrate solutions. By varying the precipitation conditions, one can exactly adjust the desired properties of the UO/sub 2/ powder, there is no 'post sintering'. The individual process steps are shown in detail.

  6. A study on the reduction of uranium oxide to uranium metal in LiCl molten salt

    International Nuclear Information System (INIS)

    Seo, J. S.; Hur, J. M.; Lee, W. K.; Hong, S. S.; Kang, D. S.; Park, S. W.

    2002-01-01

    Research for the analysis on a metallization process of uranium oxide in LiCl-Li molten salt was carried out. Effect of a concentration of Li 2 O on the metallization process was also studied. The new concept, electrochemical reduction of uranium oxide in LiCl-Li 2 O molten salt was proposed. The concept is based on the integrated process of metallization of UO 2 with simultaneous electrochemical reduction of Li 2 O which is recycled in a closed system. In a LiCl-Li molten salt system, U 3 O 8 whose conversion ratio to U turns out to be 97.1%, showed a better metallization characteristic than UO 2 . It is verified that electrochemically reduced Li is well deposited on the UO 2 powder cathode through a porous magnesia filter in LiCl-Li 2 O molten salt. In that process Li 2 O was from by the reduction process of UO 2 to U. This electrochemical reduction process showed good results to covert UO 2 to U

  7. The Resonance Absorption of Uranium Metal and Oxide

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E; Lundgren, G

    1962-06-15

    The resonance integrals for uranium metal and uranium oxide have been determined for a 1/E flux. The following results were obtained Metal RI 2.95 + 25.8{radical}(S/M); Oxide RI = 4.15 + 26.6{radical}(S/M). The oxide value agrees with the expression found earlier at this laboratory. But the result for the metal is 4. 5 % larger than the earlier one. In addition, the resonance absorption in a R1 fuel rod has been compared with that for a cadmium-covered rod placed in an approximate cell boundary flux. The former came out 3 % larger than the latter. A comparison of the fuel rod absorption with that for a 1/E flux yields a corresponding figure of 7 %. The neutron flux was monitored below the lowest resonance in uranium.

  8. Standard test methods for chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade uranium dioxide powders and pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1999-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade uranium dioxide powders and pellets to determine compliance with specifications. 1.2 This test method covers the determination of uranium and the oxygen to uranium atomic ratio in nuclear-grade uranium dioxide powder and pellets. 1.4 This test method covers the determination of chlorine and fluorine in nuclear-grade uranium dioxide. With a 1 to 10-g sample, concentrations of 5 to 200 g/g of chlorine and 1 to 200 μg/g of fluorine are determined without interference. 1.5 This test method covers the determination of moisture in uranium dioxide samples. Detection limits are as low as 10 μg. 1.6 This test method covers the determination of nitride nitrogen in uranium dioxide in the range from 10 to 250 μg. 1.7 This test method covers the spectrographic analysis of nuclear-grade UO2 for the 26 elements in the ranges indicated in Table 2. 1.8 For simultaneous determination of trace ele...

  9. Study on microstructure change of Uranium nitride coated U-7wt%Mo powder by heat treatment

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Woo Hyoung; Park, Jae Soon; Lee, Hae In; Kim, Woo Jeong; Yang, Jae Ho; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Uranium-molybdenum alloy particle dispersion fuel in an aluminum matrix with a high uranium density has been developed for a high performance research reactor in the RERTR program. In order to retard the fuel-matrix interaction in U-Mo/Al dispersion fuel in which the U-Mo fuel particles were dispersed in Al matrix, nitride layer coated U-Mo fuel particle has been designed and techniques to fabricate nitride-layer coated U-7wt%Mo particles have been developed in our lab. In this study, uranium nitride coated U-Mo particle has heat treatment for several times and degree. And we suggested for interaction layer remedy in U-Mo dispersion fuel. We investigate effect of heat treatment interaction layer evolution on uranium nitride coated U-Mo powder. The EDS and XRD analysis to investigate the phase evolution in uranium nitride coated layer is also a part of the present work

  10. Analysis of impurities in uranium oxide by ICP-MS

    International Nuclear Information System (INIS)

    Paul, M.; Hager, J.

    1989-01-01

    The method offers detection limits of less than 0.1 μg/g in solid samples and a wide linear dynamic working range and allows a fast sample throughput. The poster summarizes recent authors' experience with the analysis of high purity uranium oxide. The samples are analyzed using both conventional solution analysis and solid sampling method, using a new powerful laser sampling system attached to the ICP mass spectrometer. The total analysis time required for one uranium oxide sample is about 4 minutes including sample handling, data acquisition and processing. The results achieved from semiquantitative and quantitative analysis applying both techniques are in good agreement with the reference values. (author)

  11. Uranium Metal to Oxide Conversion by Air Oxidation –Process Development

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, A

    2001-12-31

    Published technical information for the process of metal-to-oxide conversion of uranium components has been reviewed and summarized for the purpose of supporting critical decisions for new processes and facilities for the Y-12 National Security Complex. The science of uranium oxidation under low, intermediate, and high temperature conditions is reviewed. A process and system concept is outlined and process parameters identified for uranium oxide production rates. Recommendations for additional investigations to support a conceptual design of a new facility are outlined.

  12. Transformations of highly enriched uranium into metal or oxide

    International Nuclear Information System (INIS)

    Nollet, P.; Sarrat, P.

    1964-01-01

    The enriched uranium workshops in Cadarache have a double purpose on the one hand to convert uranium hexafluoride into metal or oxide, and on the other hand to recover the uranium contained in scrap materials produced in the different metallurgical transformations. The principles that have been adopted for the design and safety of these workshops are reported. The nuclear safety is based on the geometrical limitations of the processing vessels. To establish the processes and the technology of these workshops, many studies have been made since 1960, some of which have led to original achievements. The uranium hexafluoride of high isotopic enrichment is converted either by injection of the gas into ammonia or by an original process of direct hydrogen reduction to uranium tetrafluoride. The uranium contained m uranium-zirconium metal scrap can be recovered by combustion with hydrogen chloride followed treatment of the uranium chloride by fluorine in order to obtain the uranium in the hexafluoride state. Recovery of the uranium contained m various scrap materials is obtained by a conventional refining process combustion of metallic scrap, nitric acid dissolution of the oxide, solvent purification by tributyl phosphate, ammonium diuranate precipitation, calcining, reduction and hydro fluorination into uranium tetrafluoride, bomb reduction by calcium and slag treatment. Two separate workshops operate along these lines one takes care of the uranium with an isotopic enrichment of up to 3 p. 100, the other handles the high enrichments. The handling of each step of this process, bearing in mind the necessity for nuclear safety, has raised some special technological problems and has led to the conception of new apparatus, in particular the roasting furnace for metal turnings, the nitric acid dissolution unit, the continuous precipitator and ever safe filter and dryer for ammonium diuranate, the reduction and hydro fluorination furnace and the slag recovery apparatus These are

  13. Sensitivity analysis of uranium solubility under strongly oxidizing conditions

    International Nuclear Information System (INIS)

    Liu, L.; Neretnieks, I.

    1999-01-01

    To evaluate the effect of geochemical conditions in the repository on the solubility of uranium under strongly oxidizing conditions, a mathematical model has been developed to determine the solubility, by utilizing a set of nonlinear algebraic equations to describe the chemical equilibria in the groundwater environment. The model takes into account the predominant precipitation-dissolution reactions, hydrolysis reactions and complexation reactions that may occur under strongly oxidizing conditions. The model also includes the solubility-limiting solids induced by the presence of carbonate, phosphate, silicate, calcium, and sodium in the groundwater. The thermodynamic equilibrium constants used in the solubility calculations are essentially taken from the NEA Thermochemical Data Base of Uranium, with some modification and some uranium minerals added, such as soddyite, rutherfordite, uranophane, uranyl orthophosphate, and becquerelite. By applying this model, the sensitivities of uranium solubility to variations in the concentrations of various groundwater component species are systematically investigated. The results show that the total analytical concentrations of carbonate, phosphate, silicate, and calcium in deep groundwater play the most important role in determining the solubility of uranium under strongly oxidizing conditions

  14. Oxidation experiment of metal uranium waste for the treatment of depleted uranium waste

    International Nuclear Information System (INIS)

    Kang, K. H.; Kwac, K. I.; Kim, K. J.

    2001-01-01

    A study was conducted on the oxidation behavior of U-Ti chips(Depleted Uranium, DU chips) using an XRD and a thermogravimetric analyzer in the temperature range from 250 to 500 .deg. C in air. At the temperature lower than 400 .deg. C, DU chips were converted to UO 2 , U 3 O 7 and U 3 O 8 whereas at the temperature higher than 400 .deg. C, DU chips were completely converted to U 3 O 8 , the most stable form of uranium oxide. The activation energy for the oxidation of U-Ti chips is found, 44.9 kJ/mol and the oxidation rate in terms of weight gain (%) can be expressed as ; dW/dt=8.4 x 10 2 e(-44.9 kJ/mol /RT) wt %/min (250≤T(deg. C)≤500) where W=weight gain (%), t=time and T=temperature

  15. Voltametric determination of O:U relation in uranium oxide

    International Nuclear Information System (INIS)

    Carvalho, F.M.S. de; Abrao, A.

    1988-07-01

    Uranium oxide samples are dissolved in hot concentrated H 3 PO 4 - H 2 SO 4 mixture and the solution diluted with 1M H 2 SO 4 . One aliquot of such solution (A) is used to record the first voltamogram which gives the U(VI) content. To a second aliquot HNO 3 and H 2 O 2 is added to oxidise uranium to the hexavalent state (B) and the second voltamogram is recorded from 0.0 to 0.4 V X SCE. The O:U ratio in the original sample is calculated by the expression: O/U = 2.000 + [U (VI) soln.A/% U(VI) soln. B]. The method provides an accurate means for determining O to U ratios in high-purity uranium dioxide, fuel pellets and a variety of oxides prepared for developmental work on ceramic fuel materials. (author) [pt

  16. A method of eliminating the surface defect in low-temperature oxidation powder added UO2 pellet

    International Nuclear Information System (INIS)

    Yoo, H. S.; Lee, S. J.; Kim, J. I.; Jeon, K. R.; Kim, J. W.

    2002-01-01

    A study on methods to eliminate surface defect shown in low-temperature oxidation powder added UO 2 pellet has been performed. Powders oxidized at 350 .deg. C for 4 hrs were prepared and mixed with UO 2 powder after crushing them. After being sintered, surfaces of the pellet were inspected both visually and optically. A large number of defects were observed on the surface of the specimens in which low-temperature oxidation powders were directly mixed or master mixed with UO 2 powder while both specimens produced from mixed powders including milled oxidation powders and powders that were milled totally after mixing had clean surfaces. However, optical examination showed considerably large defected pores in the milled oxidation powder added pellet and it was confirmed that the inner defects can be eliminated completely only when milling the entire mixture on UO 2 and low-temperature oxidation powder, but not by crushing only oxidation powder

  17. Graphene oxide powders with different oxidation degree, prepared by synthesis variations of the Hummers method

    International Nuclear Information System (INIS)

    Guerrero-Contreras, Jesus; Caballero-Briones, F.

    2015-01-01

    Graphene oxide (GO) powders with different oxidation degree estimated through the relative intensity of the infrared absorption bands related to oxygen containing groups were prepared through variations of the Hummers method. The GO powders were analyzed by Transmission Electron Microscopy, Energy dispersive spectroscopy, X-ray Photoelectron Spectroscopy, Fourier Transform Infrared Spectroscopy, Raman spectroscopy, X-ray Diffraction, UV–VIS spectroscopy and Electrical Resistance measurements. Several square micron GO sheets with low wrinkling were obtained. Oxygen to carbon ratio is around 0.2 in all the samples although a strong variance in the relative intensity of the oxygen related infrared bands is evident. Thus, the oxidation degree was estimated from the FTIR measurements using the quotient between the C–O related bands area to the total area under the spectra. FTIR shows presence of hydroxyl (–OH), epoxy (C–O–C), carboxyl (–COOH) and carbonyl (C=O) moieties and evidence of intermolecular interactions between adjacent groups. These interactions influence the exfoliation degree, the absorbance of the GO suspensions, as well as the electrical resistance, while the crystalline domain sizes, estimated from XRD and Raman do not show a noticeable behavior related with the composition and molecular structure. The results indicate that the electrical resistance is influenced mainly by the surface chemistry of the GO powders and not only by the O/C ratio. The control of the surface chemistry of GO powders would allow their use as additives in organic bulk heterojunction solar cells with enhanced photoconversion efficiency. - Highlights: • Powders of graphene oxide with different oxidation degree were prepared through variations of the Hummers method. • Raman spectroscopy and XRD demonstrated similar crystallite domain size in the samples. • Electrical resistance, exfoliation degree and optical absorption depend on the molecular structure.

  18. Graphene oxide powders with different oxidation degree, prepared by synthesis variations of the Hummers method

    Energy Technology Data Exchange (ETDEWEB)

    Guerrero-Contreras, Jesus; Caballero-Briones, F., E-mail: fcaballero@ipn.mx

    2015-03-01

    Graphene oxide (GO) powders with different oxidation degree estimated through the relative intensity of the infrared absorption bands related to oxygen containing groups were prepared through variations of the Hummers method. The GO powders were analyzed by Transmission Electron Microscopy, Energy dispersive spectroscopy, X-ray Photoelectron Spectroscopy, Fourier Transform Infrared Spectroscopy, Raman spectroscopy, X-ray Diffraction, UV–VIS spectroscopy and Electrical Resistance measurements. Several square micron GO sheets with low wrinkling were obtained. Oxygen to carbon ratio is around 0.2 in all the samples although a strong variance in the relative intensity of the oxygen related infrared bands is evident. Thus, the oxidation degree was estimated from the FTIR measurements using the quotient between the C–O related bands area to the total area under the spectra. FTIR shows presence of hydroxyl (–OH), epoxy (C–O–C), carboxyl (–COOH) and carbonyl (C=O) moieties and evidence of intermolecular interactions between adjacent groups. These interactions influence the exfoliation degree, the absorbance of the GO suspensions, as well as the electrical resistance, while the crystalline domain sizes, estimated from XRD and Raman do not show a noticeable behavior related with the composition and molecular structure. The results indicate that the electrical resistance is influenced mainly by the surface chemistry of the GO powders and not only by the O/C ratio. The control of the surface chemistry of GO powders would allow their use as additives in organic bulk heterojunction solar cells with enhanced photoconversion efficiency. - Highlights: • Powders of graphene oxide with different oxidation degree were prepared through variations of the Hummers method. • Raman spectroscopy and XRD demonstrated similar crystallite domain size in the samples. • Electrical resistance, exfoliation degree and optical absorption depend on the molecular structure.

  19. Practical considerations of pyrite oxidation control in uranium tailings

    International Nuclear Information System (INIS)

    1984-05-01

    The problems posed by the oxidation of pyrite in uranium tailings include the generation of sulfuric acid and acid sulfate metal salts. These have substantial negative impacts on watercourse biota by themselves, and the lowered pH levels tend to mobilize heavy metals present in the tailings the rate of oxidation of pyrite at lower pH levels is catalyzed by sulfur and iron oxidizing bacteria present in soils. No single clear solution to the problems came from this study. Exclusion of air is a most important preventative of bacterial catalysis of oxidation. Bactericides, chemically breaking the chain of integrated oxidation reactions, maintaining anaerobic conditions, or maintaining a neutral or alkaline pH all reduce the oxidation rate. Removal of pyrite by flotation will reduce but not eliminate the impact of pyrite oxidation. Controlled oxidation of the remaining sulfide in the flotation tails would provide an innocuous tailing so far as acidity generation is concerned

  20. Study on the solid state chemistry of ternary uranium oxides

    International Nuclear Information System (INIS)

    Yamashita, Toshiyuki

    1988-03-01

    With the increase of burnup of uranium oxide fuels, various kinds of fission products are formed, and the oxygen atoms combined with the consumed heavy atoms are freed. The solid state chemical and/or thermodynamic properties of these elements at high temperatures are complex, and have not been well clarified. In the present report, an approach was taken that the chemical interactions between UO 2 and these fission products can be regarded as causing overlapped effects of composing ternary uranium oxides, and formation reactions and phase behavior were studied for several ternary uranium oxides with typical fission product elements such as alkaline earth metals and rare earth elements. Precise determination methods for the composition of ternary uranium oxides were developed. The estimated accuracies for x and y values in M y U 1-y O 2+x were ± 0.006 and ± 0.004, respectively. The thermodynamic properties and the lattice parameters of the phases in the Ca-U-O and Pr-U-O systems were discussed in relation to the composition determined by the methods. Crystal structure analyses of cadmium monouranates were made with X-ray diffraction method. (author) 197 refs

  1. Resuspension of uranium-plutonium oxide particles from burning Plexiglas

    International Nuclear Information System (INIS)

    Pickering, S.

    1987-01-01

    Nuclear fuel materials such as Uranium-Plutonium oxide must be handled remotely in gloveboxes because of their radiotoxicity. These gloveboxes are frequently constructed largely of combustible Plexiglas sheet. To estimate the potential airborne spread of radioactive contamination in the event of a glovebox fire, the resuspension of particles from burning Plexiglas was investigated. (author)

  2. Controlled synthesis of thorium and uranium oxide nano-crystals

    International Nuclear Information System (INIS)

    Hudry, Damien; Apostolidis, Christos; Walter, Olaf; Gouder, Thomas; Courtois, Eglantine; Kubel, Christian; Meyer, Daniel

    2013-01-01

    Very little is known about the size and shape effects on the properties of actinide compounds. As a consequence, the controlled synthesis of well-defined actinide-based nano-crystals constitutes a fundamental step before studying their corresponding properties. In this paper, we report on the non-aqueous surfactant-assisted synthesis of thorium and uranium oxide nano-crystals. The final characteristics of thorium and uranium oxide nano-crystals can be easily tuned by controlling a few experimental parameters such as the nature of the actinide precursor and the composition of the organic system (e.g., the chemical nature of the surfactants and their relative concentrations). Additionally, the influence of these parameters on the outcome of the synthesis is highly dependent on the nature of the actinide element (thorium versus uranium). By using optimised experimental conditions, monodisperse isotropic uranium oxide nano-crystals with different sizes (4.5 and 10.7 nm) as well as branched nano-crystals (overall size ca. 5 nm), nano-dots (ca. 4 nm) and nano-rods (with ultra-small diameters of 1 nm) of thorium oxide were synthesised. (authors)

  3. On the stability of sub-stoichiometric uranium oxides

    Science.gov (United States)

    Winer, K.; Colmenares, C. A.; Smith, R. L.; Wooten, F.

    1986-12-01

    The oxidation of clean, high-purity polycrystalline uranium metal surfaces for low exposures to dry oxygen was studied with AES and XPS in an attempt to substantiate claims for the formation of a stable UO surface phase at ambient temperatures. We found no evidence for such a surface phase and found instead that grossly sub-stoichiometric surface oxides were formed after sequential oxygen saturation and heating.

  4. Characterization and uranium bioleaching performance of mixed iron- and sulfur-oxidizers versus iron-oxidizers

    International Nuclear Information System (INIS)

    Qian Li; Jing Sun; Dexin Ding; Qingliang Wang; Wenge Shi; Eming Hu; Xiaoyu Jiang; University of South China, Hengyang; Xingxing Wang

    2017-01-01

    In order to develop and apply mixed iron- and sulfur-oxidizers in uranium bioleaching, the characteristics of a mixed iron- and sulfur-oxidizing consortium (Consortium ISO) were comparatively investigated versus an iron-oxidizing consortium (Consortium IO). The results showed, the Consortium ISO exerted stronger oxidative ability and acid-producing ability than Consortium IO did. The synergy of sulfur-oxidizers and iron-oxidizers could change the structure and properties of the passivation substance, and work positively for eliminating the accumulation of passivation substance. In the bioleaching process, the uranium bioleaching experiments showed the recovery percentage of uranium reached 99.5% with Consortium ISO, 6.3% more than that of Consortium IO. (author)

  5. Preparation and characterization of uranium alkoxides through oxidation of uranium metal

    International Nuclear Information System (INIS)

    Gordon, P.L.; Sauer, N.N.; Burns, C.J.; Watkin, J.G.; Van Der Sluys, W.G.

    1993-01-01

    Currently the authors are investigating the preparation of halide-containing uranium alkoxides by simultaneous halogen and alcohol oxidation of uranium metal. They recently reported the formation of U 2 I 4 (O-i-Pr) 4 (HO-i-Pr) 2 which upon addition of excess isopropanol forms UI 2 (O-i-Pr) 2 (HO-i-Pr) 2 . They report further characterization and reactivity for this monomeric species. Attempts to prepare similar complexes are being made using chlorine gas in the presence of other alcohols. They describe this ongoing research

  6. Method to manufacture a nuclear fuel from uranium-plutonium monocarbide or uranium-plutonium mononitride

    International Nuclear Information System (INIS)

    Krauth, A.; Mueller, N.

    1977-01-01

    Pure uranium carbide or nitride is converted with plutonium oxide and carbon (all in powder form) to uranium-plutonium monocarbide or mononitride by cold pressing and sintering at about 1600 0 C. Pure uranium carbide or uranium nitride powder is firstly prepared without extensive safety measures. The pure uranium carbide or nitride powder can also be inactivated by using chemical substances (e.g. stearic acid) and be handled in air. The sinterable uranium carbide or nitride powder (or also granulate) is then introduced into the plutonium line and mixed with a nonstoichiometrically adjusted, prereacted mixture of plutonium oxide and carbon, pressed to pellets and reaction sintered. The surface of the uranium-plutonium carbide (higher metal content) can be nitrated towards the end of the sinter process in a stream of nitrogen. The protective layer stabilizes the carbide against the water and oxygen content in air. (IHOE) [de

  7. Conversion and Blending Facility highly enriched uranium to low enriched uranium as oxide. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

  8. Method for preparation of uranium hydride

    International Nuclear Information System (INIS)

    Gorski, M.S.; Goncalves, Miriam; Mirage, A.; Lima, W. de.

    1985-01-01

    A method for preparation of Uranium Hydride starting from Hidrogen and Uranium is described. In the temperature range of 250 0 up to 350 0 C, and pressures above 10torr, Hydrogen reacts smoothly with Uranium turnings forming a fine black or dark gray powder (UH 3 ). Samples containing a significant amount of oxides show a delay before the reaction begging. (Author) [pt

  9. Preparation of uranium-based oxide catalysts; Preparation de catalyseurs oxydes a base d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Bressat, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1965-07-01

    We have studied the thermal decomposition of uranyl and uranium IV oxalates as a mean of producing uranium dioxide. We have isolated the main intermediate phases of the decompositions and have indexed the lines of their X-ray diffraction patterns. The oxides produced by the decomposition are ill-defined and unstable: they strongly absorb atmospheric oxygen with modification of the composition and, in certain cases, of the structure (pyrophoric oxide). With a view to obtaining stable oxides, we have prepared mixed uranium-thorium oxalates. In order to prepare an oxalate having a homogeneous composition, it is necessary to adopt a well-defined preparation method: the addition of solutions of thorium and uranium IV nitrates to a continually saturated oxalic acid solution. The mixed oxide obtained from the thermal decomposition of an oxalate U{sub x}Th{sub 1-x}(C{sub 2}O{sub 4}){sub 2}, 2 H{sub 2}O at 500 C for 24 hours in a current of oxygen leads to a cubic structure which is well-defined both in the bulk and superficially when x is less than 0.35. Above this atomic concentration of uranium, some uranium moves out of the lattice in the form of UO{sub 3} or U{sub 3}O{sub 8} according to the temperature. The mixed oxide is not stoichiometric,(U{sub x}Th{sub 1-x}O{sub 2+y}) and the average degree of oxidation of the uranium varies with the temperature and partial oxygen pressure. The oxides thus formed have a high surface area. By dissolving the mixed oxalates in a concentrated solution of ammonium oxalate, it is possible to deposit the catalyst on a support, but the differences in the solubilities of the thorium and uranium IV oxalates in the ammonium oxalate make it impossible to prepare double salts formed either of thorium and uranium and of ammonium. (author) [French] Nous avons etudie la decomposition thermique des oxalates d'uranyle et d'uranium IV en vue d'aboutir au dioxide d'uranium. Nous avons pu isoler les principales phases intermediaires des decompositions

  10. Specification for nuclear-grade beryllium oxide powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification defines the physical and chemical requirements of nuclear-grade beryllium oxide (BeO) powder to be used in fabricating nuclear components. 1.2 This specification does not include requirements for health and safety. , , It recognizes the material as a Class B poison and suggests that producers and users become thoroughly familiar with and comply to applicable federal, state, and local regulations and handling guidelines. 1.3 Special tests and procedures are given in Annex A1 and Annex A2. 1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  11. Characterization of manganese-gallium mixed oxide powders

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez Escribano, V.; Fernandez Lopez, E.; Sanchez Huidobro, P. [Universidad de Salamanca, Dept. de Quimica Inorganica (Spain); Panizza, M.; Resini, C.; Busca, G. [UNiversita di Genova, Dipt. di Ingegneria Chimica e di Processo, Genova (Italy); Resini, C. [Istituto Nazionale di Fisica della Materia, INFM (Spain); Gallardo- Amores, J.M. [Universidad Complutense, Dept. de Quimica Inorganica, Lab. Complutense de Altas Presiones, Madrid (Spain)

    2003-12-01

    Mn-Ga mixed oxides have been prepared by coprecipitation of the corresponding oxo-hydroxides as powders and have been characterized in relation to their structural and optical properties. The materials have been characterized by XRD, TG-DTA, skeletal IR and UV-visible-NIR spectroscopies. Large solubility of Mn in the diaspore type {alpha}-GaOOH oxo-hydroxide has been found. The spinel related structures of hausmannite Mn{sub 3}O{sub 4} and of {beta}-gallia present large reciprocal solubilities at least in a metastable form. At high temperature also bixbyite-type {alpha}-Mn{sub O3} solid solutions containing up to 20% at. Ga have been observed. (authors)

  12. Characterization of manganese?gallium mixed oxide powders

    Science.gov (United States)

    Sánchez Escribano, Vicente; Fernández López, Enrique; Sánchez Huidobro, Paula; Panizza, Marta; Resini, Carlo; Gallardo-Amores, José M.; Busca, Guido

    2003-11-01

    MnGa mixed oxides have been prepared by coprecipitation of the corresponding oxo-hydroxides as powders and have been characterized in relation to their structural and optical properties. The materials have been characterized by XRD, TG-DTA, skeletal IR and UV-visible-NIR spectroscopies. Large solubility of Mn in the diaspore type α-GaOOH oxo-hydroxide has been found. The spinel related structures of hausmannite Mn 3O 4 and of β-gallia present large reciprocal solubilities at least in a metastable form. At high temperature also bixbyite-type α-Mn 2O 3 solid solutions containing up to 20% at. Ga have been observed.

  13. Oxidation kinetics of reaction products formed in uranium metal corrosion

    International Nuclear Information System (INIS)

    Totemeier, T. C.

    1998-01-01

    The oxidation behavior of uranium metal ZPPR fuel corrosion products in environments of Ar-4%O 2 and Ar-20%O 2 were studied using thermo-gravimetric analysis (TGA). These tests were performed to extend earlier work in this area specifically, to assess plate-to-plate variations in corrosion product properties and the effect of oxygen concentration on oxidation behavior. The corrosion products from two relatively severely corroded plates were similar, while the products from a relatively intact plate were not reactive. Oxygen concentration strongly affected the burning rate of reactive products, but had little effect on low-temperature oxidation rates

  14. Oxidation kinetics of reaction products formed in uranium metal corrosion.

    Energy Technology Data Exchange (ETDEWEB)

    Totemeier, T. C.

    1998-04-22

    The oxidation behavior of uranium metal ZPPR fuel corrosion products in environments of Ar-4%O{sub 2} and Ar-20%O{sub 2} were studied using thermo-gravimetric analysis (TGA). These tests were performed to extend earlier work in this area specifically, to assess plate-to-plate variations in corrosion product properties and the effect of oxygen concentration on oxidation behavior. The corrosion products from two relatively severely corroded plates were similar, while the products from a relatively intact plate were not reactive. Oxygen concentration strongly affected the burning rate of reactive products, but had little effect on low-temperature oxidation rates.

  15. Oxidation of uranium monocarbide in dry or moist oxygen

    International Nuclear Information System (INIS)

    Herrmann, B.; Herrmann, F.J.

    1968-01-01

    The kinetics of oxidation of uranium monocarbide either in dry or moist air or in oxygen-argon mixtures, has been studied thermogravimetrically, between 500 and 800 C in a circulating atmosphere. In all cases the oxidation leads to the formation of U 3 O 8 . Between 500 and 700 C, the activation energy is about 21 +3 kcal/mole. It seems to decrease between 700 and 800 C, but the reaction follows always a linear rate law. In moist air, the oxidation proceeds more swiftly, due to an increase in the reactional interface. An evaluation of the over-temperature has been made at 800 C. (author) [fr

  16. Comparison of open cycles of uranium and mixed oxides of thorium-uranium using advanced reactors

    International Nuclear Information System (INIS)

    Gonçalves, Letícia C.; Maiorino, José R.

    2017-01-01

    A comparative study of the mass balance and production costs of uranium oxide fuels was carried out for an AP1000 reactor and thorium-uranium mixed oxide in a reactor proposal using thorium called AP-Th1000. Assuming the input mass values for a fuel load the average enrichment for both reactors as well as their feed mass was determined. With these parameters, the costs were calculated in each fuel preparation process, assuming the prices provided by the World Nuclear Association. The total fuel costs for the two reactors were quantitatively compared with 18-month open cycle. Considering enrichment of 20% for the open cycle of mixed U-Th oxide fuel, the total uranium consumption of this option was 50% higher and the cost due to the enrichment was 70% higher. The results show that the use of U-Th mixed oxide fuels can be advantageous considering sustainability issues. In this case other parameters and conditions should be investigated, especially those related to fuel recycling, spent fuel storage and reduction of the amount of transuranic radioactive waste

  17. Water Sorption and Gamma Radiolysis Studies for Uranium Oxides

    Energy Technology Data Exchange (ETDEWEB)

    Icenhour, A.S.

    2002-02-27

    During the development of a standard for the safe, long-term storage of {sup 233}U-containing materials, several areas were identified that needed additional experimental studies. These studies were related to the perceived potential for the radiolytic generation of large pressures or explosive concentrations of gases in storage containers. This report documents the results of studies on the sorption of water by various uranium oxides and on the gamma radiolysis of uranium oxides containing various amounts of sorbed moisture. In all of the experiments, {sup 238}U was used as a surrogate for the {sup 233}U. For the water sorption experiments, uranium oxide samples were prepared and exposed to known levels of humidity to establish the water uptake rate. Subsequently, the amount of water removed was studied by heating samples in a oven at fixed temperatures and by thermogravimetric analysis (TGA)/differential thermal analysis (DTA). It was demonstrated that heating at 650 C adequately removes all moisture from the samples. Uranium-238 oxides were irradiated in a {sup 60}Co source and in the high-gamma-radiation fields provided by spent nuclear fuel elements of the High Flux Isotope Reactor. For hydrated samples of UO{sub 3}, H{sub 2} was the primary gas produced; but the total gas pressure increase reached steady value of about 10 psi. This production appears to be a function of the dose and the amount of water present. Oxygen in the hydrated UO{sub 3} sample atmosphere was typically depleted, and no significant pressure rise was observed. Heat treatment of the UO{sub 3} {center_dot} xH{sub 2}O at 650 C would result in conversion to U{sub 3}O{sub 8} and eliminate the H{sub 2} production. For all of the U{sub 3}O{sub 8} samples loaded in air and irradiated with gamma radiation, a pressure decrease was seen and little, if any, H{sub 2} was produced--even for samples with up to 9 wt % moisture content. Hence, these results demonstrated that the efforts to remove trace

  18. Literature information applicable to the reaction of uranium oxides with chlorine to prepare uranium tetrachloride

    Energy Technology Data Exchange (ETDEWEB)

    Haas, P.A.

    1992-02-01

    The reaction of uranium oxides and chlorine to prepare anhydrous uranium tetrachloride (UCl{sub 4}) are important to more economical preparation of uranium metal. The most practical reactions require carbon or carbon monoxide (CO) to give CO or carbon dioxide (CO{sub 2}) as waste gases. The chemistry of U-O-Cl compounds is very complex with valances of 3, 4, 5, and 6 and with stable oxychlorides. Literature was reviewed to collect thermochemical data, phase equilibrium information, and results of experimental studies. Calculations using thermodynamic data can identify the probable reactions, but the results are uncertain. All the U-O-Cl compounds have large free energies of formation and the calculations give uncertain small differences of large numbers. The phase diagram for UCl{sub 4}-UO{sub 2} shows a reaction to form uranium oxychloride (UOCl{sub 2}) that has a good solubility in molten UCl{sub 4}. This appears more favorable to good rates of reaction than reaction of solids and gases. There is limited information on U-O-Cl salt properties. Information on the preparation of titanium, zirconium, silicon, and thorium tetrachlorides (TiCl{sub 4}, ZrCl{sub 4}, SiCl{sub 4}, ThCl{sub 4}) by reaction of oxides with chlorine (Cl{sub 2}) and carbon has application to the preparation of UCl{sub 4}.

  19. Literature information applicable to the reaction of uranium oxides with chlorine to prepare uranium tetrachloride

    International Nuclear Information System (INIS)

    Haas, P.A.

    1992-02-01

    The reaction of uranium oxides and chlorine to prepare anhydrous uranium tetrachloride (UCl 4 ) are important to more economical preparation of uranium metal. The most practical reactions require carbon or carbon monoxide (CO) to give CO or carbon dioxide (CO 2 ) as waste gases. The chemistry of U-O-Cl compounds is very complex with valances of 3, 4, 5, and 6 and with stable oxychlorides. Literature was reviewed to collect thermochemical data, phase equilibrium information, and results of experimental studies. Calculations using thermodynamic data can identify the probable reactions, but the results are uncertain. All the U-O-Cl compounds have large free energies of formation and the calculations give uncertain small differences of large numbers. The phase diagram for UCl 4 -UO 2 shows a reaction to form uranium oxychloride (UOCl 2 ) that has a good solubility in molten UCl 4 . This appears more favorable to good rates of reaction than reaction of solids and gases. There is limited information on U-O-Cl salt properties. Information on the preparation of titanium, zirconium, silicon, and thorium tetrachlorides (TiCl 4 , ZrCl 4 , SiCl 4 , ThCl 4 ) by reaction of oxides with chlorine (Cl 2 ) and carbon has application to the preparation of UCl 4

  20. Dissolution of uranium oxide materials in simulated lung fluid

    International Nuclear Information System (INIS)

    Scripsick, R.C.; Soderholm, S.C.

    1985-01-01

    Depleted uranium (DU) oxide aerosols prepared in the laboratory and collected in the field were tested to characterize their dissolution in simulated lung fluid and to determine how dissolution is affected by aerosol preparation. DU, a by-product of the uranium fuel cycle, has been selected by the US military for use in several types of munitions. During development, manufacture, testing, and use of these munitions, opportunities exist for inhalation exposure to various (usually oxide) aerosol forms of DU. The hazard potential associated with such exposures is closely related to the chemical form, the size of the DU aerosol material, and its dissolution properties. Five DU sample materials produced by exposing uranium alloy penetrators to certain controlled oxidation atmospheres were studied (oxidation temperatures ranged from 500 to 900 0 C). In addition, two DU sample materials collected in the field were provided by the US Air Force. All sample materials were generated as aerosols and the respirable fraction was separated and collected. Data suggest that under some conditions a rapidly dissolving U 3 O 8 fraction may be formed concurrent with the production of UO 2

  1. Experimental investigation and thermodynamic simulation of the uranium oxide-zirconium oxide-iron oxide system in air

    Czech Academy of Sciences Publication Activity Database

    Petrov, Y. B.; Udalov, Y. P.; Šubrt, Jan; Bakardjieva, Snejana; Sázavský, P.; Kiselová, M.; Selucký, P.; Bezdička, Petr; Joumeau, C.; Piluso, P.

    2011-01-01

    Roč. 37, č. 2 (2011), s. 212-229 ISSN 1087-6596 Institutional research plan: CEZ:AV0Z40320502 Keywords : uranium oxide * zirconium oxide * iron oxide * fusibility curve * oxygen partial pressure * crystallization * phase composition Subject RIV: CA - Inorganic Chemistry Impact factor: 0.492, year: 2011

  2. Ionization and Coulomb explosion of small uranium oxide clusters

    International Nuclear Information System (INIS)

    Ross, Matt W; Castleman, A W Jr

    2012-01-01

    Femtosecond pulses are used to study the strong-field ionization and subsequent Coulomb explosion of small uranium oxide clusters. The resulting high atomic charge states are explored as a function of laser intensity and compared to ionization rates calculated using semi-classical tunneling theory with sequential ionization potential values. The gap in laser intensity between saturation intensity values for the 7s, 6d, and 5f orbitals are identified and quantified. Extreme charge states of oxygen up to O 4+ are observed indicating multiple ionization enhancement processes occurring within the clusters. The peak splittings of the atomic charge states are explored and compared to previous results on transition metal oxide species. Participation of the 5f orbitals in bonding is clearly identified based on the saturation intensity dependence of oxygen to uranium metal.

  3. Characteristics of uranium oxide cathode for neutron streak camera

    International Nuclear Information System (INIS)

    Niki, H.; Itoga, K.; Yamanaka, M.; Yamanaka, T.; Yamanaka, C.

    1986-01-01

    In laser fusion research, time-resolved neutron measurements require 20ps resolution in order to obtain the time history of the D-T burn. Uranium oxide was expected to be a sensitive material as a cathode of a neutron streak camera because of its large fission cross section. The authors report their measurements of some characteristics of the uranium oxide cathode connected to a conventional streak tube. 14 MeV neutron signal were observed as the bright spots on a TV monitor using a focus mode opration. Detection efficiency was ∼ 1 x 10 -6 for 1 μm thick cathode. Each signal consisted of more than several tens of components, which were corresponding to the secondary electrons dragged out from the cathode by a fission fragment. Time resolution is thought to be limited mainly by the transit time spread of the secondary electrons. 14ps resolution was obtained by a streak mode operation for a single fission event

  4. Uranium tetracyclopentadienyl interaction with carbon oxide and dioxide

    International Nuclear Information System (INIS)

    Leonov, M.R.; Solov'eva, G.V.; Kozina, I.Z.; Bolotova, G.T.

    1983-01-01

    Using the methods of gas-liquid chromatography, IR and UV spectroscopy and element analysis, the reactions of tetracyclogentadienyluranium with carbon oxide and dioxide have been studied. It is shown that complete uranium cyclopentadienyl π-complex-tetracyclopentadienyluranium - in pentane under normal conditions for 100 hr reacts with carbon oxide and dioxide with the formation of polymeric complex ([(etasup(5)-Csub(5)Hsub(5))x(-CO-)U(etasup(5)-Csub(5)Hsub(4))(-CO-)]sub(2)]sub(n), in which two uranium atoms are bonded with two bridge fragments (eta 5 -C 5 H 4 -CO-), and dimeric complex [(eta 5 -C 5 H 5 ) 2 UH 2 xCO 2 ] 2 respectively

  5. Thermal diffusivity and conductivity of thorium- uranium mixed oxides

    Science.gov (United States)

    Saoudi, M.; Staicu, D.; Mouris, J.; Bergeron, A.; Hamilton, H.; Naji, M.; Freis, D.; Cologna, M.

    2018-03-01

    Thorium-uranium oxide pellets with high densities were prepared at the Canadian Nuclear Laboratories (CNL) by co-milling, pressing, and sintering at 2023 K, with UO2 mass contents of 0, 1.5, 3, 8, 13, 30, 60 and 100%. At the Joint Research Centre, Karlsruhe (JRC-Karlsruhe), thorium-uranium oxide pellets were prepared using the spark plasma sintering (SPS) technique with 79 and 93 wt. % UO2. The thermal diffusivity of (Th1-xUx)O2 (0 ≤ x ≤ 1) was measured at CNL and at JRC-Karlsruhe using the laser flash technique. ThO2 and (Th,U)O2 with 1.5, 3, 8 and 13 wt. % UO2 were found to be semi-transparent to the infrared wavelength of the laser and were coated with graphite for the thermal diffusivity measurements. This semi-transparency decreased with the addition of UO2 and was lost at about 30 wt. % of UO2 in ThO2. The thermal conductivity was deduced using the measured density and literature data for the specific heat capacity. The thermal conductivity for ThO2 is significantly higher than for UO2. The thermal conductivity of (Th,U)O2 decreases rapidly with increasing UO2 content, and for UO2 contents of 60% and higher, the conductivity of the thorium-uranium oxide fuel is close to UO2. As the mass difference between the Th and U atoms is small, the thermal conductivity decrease is attributed to the phonon scattering enhanced by lattice strain due to the introduction of uranium in ThO2 lattice. The new results were compared to the data available in the literature and were evaluated using the classical phonon transport model for oxide systems.

  6. An investigation on fuel meats extruded with atomized U-10wt% Mo powder for uranium high-density dispersion fuel

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Don-Bae; Sohn, Dong-Seong

    1997-01-01

    The RERTR program has been making an effort to develop dispersion fuels with uranium densities of 8 to 9 g U/cm3 for research and test reactors. Using atomized U-10wt%Mo powder, fuel meats have been fabricated successfully up to 55 volume % of fuel powder. The uranium density of an extruded meat with a 55 volume % of fuel powder was obtained to be 7.7 g/cm3. A relatively high porosity of 7.3% was formed due to cracking of particles, presumably induced by the impingement among agglomerated particles. Tensile test results indicated that the strength of fuel meats with 55% volume fraction decreased some and a little of ductility was maintained. Examination on the fracture surface revealed that some U-10%Mo particles appeared to be broken by the tensile force in brittle rupture mode. The increase of broken particles in high fuel fraction is considered to be induced mainly by the impingement among agglomerated particles. Uranium loading density is assumed to be improved through the development of the better homogeneous dispersion technology. (author)

  7. Electrical conductivity of uranium-antimony oxide catalysts

    International Nuclear Information System (INIS)

    Golunski, S.E.; Nevell, T.G.; Hucknall, D.J.

    1985-01-01

    The relative ionic and electronic contributions to the electrical conductivity of a uranium-antimony oxide catalyst and of USbO 5 have been determined from measurements of a.c. and d.c. conductance. Under inert atmospheres (390 to 775 K) conduction in the catalyst (predominantly USb 3 O 10 together with small proportions of Sb 2 O 4 and USbO 5 ) is associated with both electronic and effectively charged atomic point defects. Only electronic conduction occurs in USbO 5 . Under oxygen (10 to 70 kPa, 493 to 682 K) both materials are n-type semiconductors at higher temperatures, but at lower temperatures semiconducting behaviour varies with the pressure of oxygen. Heating USbO 5 in oxygen induces an ionic contribution to conductivity. Ionic conduction in the catalyst is eliminated by heating in hydrogen or propene at 470 K but is restored by heating in oxygen. It is suggested that both charged oxygen vacancies and interstitial oxide ions are involved in interactions of gaseous components with uranium-antimony oxides. With alkenes, interstitial oxide ions give rise to the products of selective partial oxidation. (author)

  8. Electrical conductivity of uranium-antimony oxide catalysts

    Energy Technology Data Exchange (ETDEWEB)

    Golunski, S.E.; Nevell, T.G. (Portsmouth Polytechnic (UK)); Hucknall, D.J. (Southampton Univ. (UK). Dept. of Chemistry)

    1985-05-01

    The relative ionic and electronic contributions to the electrical conductivity of a uranium-antimony oxide catalyst and of USbO/sub 5/ have been determined from measurements of a.c. and d.c. conductance. Under inert atmospheres (390 to 775 K) conduction in the catalyst (predominantly USb/sub 3/O/sub 10/ together with small proportions of Sb/sub 2/O/sub 4/ and USbO/sub 5/) is associated with both electronic and effectively charged atomic point defects. Only electronic conduction occurs in USbO/sub 5/. Under oxygen (10 to 70 kPa, 493 to 682 K) both materials are n-type semiconductors at higher temperatures, but at lower temperatures semiconducting behaviour varies with the pressure of oxygen. Heating USbO/sub 5/ in oxygen induces an ionic contribution to conductivity. Ionic conduction in the catalyst is eliminated by heating in hydrogen or propene at 470 K but is restored by heating in oxygen. It is suggested that both charged oxygen vacancies and interstitial oxide ions are involved in interactions of gaseous components with uranium-antimony oxides. With alkenes, interstitial oxide ions give rise to the products of selective partial oxidation.

  9. Measurement of fission track of uranium particle by solid state nuclear track detector

    International Nuclear Information System (INIS)

    Son, S. C.; Pyo, H. W.; Ji, K. Y.; Kim, W. H.

    2002-01-01

    In this study, we discussed results of the measurement of fission tracks for the uranium containing particles by solid state nuclear track detector. Uranium containing silica and uranium oxide particles were prepared by uranium sorption onto silica powder in weak acidic medium and laser ablation on uranium pellet, respectively. Fission tracks for the uranium containing silica and uranium oxide particles were detected on Lexan plastic detector. It was found that the fission track size and shapes depend on the particle size uranium content in particles. Correlation of uranium particle diameter with fission track radius was also discussed

  10. Alpha Radiolysis of Sorbed Water on Uranium Oxides and Uranium Oxyfluorides

    Energy Technology Data Exchange (ETDEWEB)

    Icenhour, A.S.

    2003-09-10

    The radiolysis of sorbed water and other impurities contained in actinide oxides has been the focus of a number of studies related to the establishment of criteria for the safe storage and transport of these materials. Gamma radiolysis studies have previously been performed on uranium oxides and oxyfluorides (UO{sub 3}, U{sub 3}O{sub 8}, and UO{sub 2}F{sub 2}) to evaluate the long-term storage characteristics of {sup 233}U. This report describes a similar study for alpha radiolysis. Uranium oxides and oxyfluorides (with {sup 238}U as the surrogate for {sup 233}U) were subjected to relatively high alpha radiation doses (235 to 634 MGy) by doping with {sup 244}Cm. The typical irradiation time for these samples was about 1.5 years, which would be equivalent to more than 50 years irradiation by a {sup 233}U sample. Both dry and wet (up to 10 wt % water) samples were examined in an effort to identify the gas pressure and composition changes that occurred as a result of radiolysis. This study shows that several competing reactions occur during radiolysis, with the net effect that only very low pressures of hydrogen, nitrogen, and carbon dioxide are generated from the water, nitrate, and carbon impurities, respectively, associated with the oxides. In the absence of nitrate impurities, no pressures greater than 1000 torr are generated. Usually, however, the oxygen in the air atmosphere over the oxides is consumed with the corresponding oxidation of the uranium oxide. In the presence of up to 10 wt % water, the oxides first show a small pressure rise followed by a net decrease due to the oxygen consumption and the attainment of a steady-state pressure where the rate of generation of gaseous components is balanced by their recombination and/or consumption in the oxide phase. These results clearly demonstrate that alpha radiolysis of either wet or dry {sup 233}U oxides will not produce deleterious pressures or gaseous components that could compromise the long-term storage of

  11. Uranium accumulation in valentinite within the oxidation zone of an antimony occurrence

    Energy Technology Data Exchange (ETDEWEB)

    Sergeev, I.P.; Kurilo, M.V.

    1985-01-01

    As a result of mineralogic-radiogeochemical study of real composition of oxidation zone of antimony occurrence represented by quartz - antimonite vien in silicificated sandstones the previousy undescribed phenomenon of uranium concentration in valentinite Sb/sub 2/O/sub 3/ one of antimonite oxidation products has been found. According to the data of fission radiography the enrichment of valentinite with uranium is clearly seen, particularly of its concentrically zonal aggregates. The valentinite is the basic uranium mineral-concentrator in the oxidation zone, whereas in the initial mineral - antimonite as well as in the product of its more complete oxidation - stibiconite - uranium is lacking. Probably the crystallochemical properties of anhydrous antimony oxide (valentinite) by analogy with those for iron (goethite hematite) and titanium (leucoxene) are the most favourable for uranium adsorption as compared with high-water antimony oxides (stibiconite) or iron (limonite), which do not usually sorb uranium.

  12. Uranium accumulation in valentinite within the oxidation zone of an antimony occurrence

    International Nuclear Information System (INIS)

    Sergeev, I.P.; Kurilo, M.V.

    1985-01-01

    As a result of mineralogic-radiogeochemical study of real composition of oxidation zone of antimony occurrence represented by quartz - antimonite vien in silicificated sandstones the previousy undescribed phenomenon of uranium concentration in valentinite Sb 2 O 3 one of antimonite oxidation products has been found. According to the data of fission radiography the enrichment of valentinite with uranium is clearly seen, particularly of its concentrically zonal aggregates. The valentinite is the basic uranium mineral-concentrator in the oxidation zone, whereas in the initial mineral - antimonite as well as in the product of its more complete oxidation - stibiconite - uranium is lacking. Probably the crystallochemical properties of anhydrous antimony oxide (valentinite) by analogy with those for iron (goethite hematite) and titanium (leucoxene) are the most favourable for uranium adsorption as compared with high-water antimony oxides (stibiconite) or iron (limonite), which do not usually sorb uranium

  13. Storage stability of cauliflower soup powder: The effect of lipid oxidation and protein degradation reactions.

    Science.gov (United States)

    Raitio, Riikka; Orlien, Vibeke; Skibsted, Leif H

    2011-09-15

    Soups based on cauliflower soup powders, prepared by dry mixing of ingredients and rapeseed oil, showed a decrease in quality, as evaluated by a sensory panel, during the storage of the soup powder in the dark for up to 12weeks under mildly accelerated conditions of 40°C and 75% relative humidity. Antioxidant, shown to be effective in protecting the rapeseed bulk oil, used for the powder preparation, had no effect on storage stability of the soup powder. The freshly prepared soup powder had a relatively high concentration of free radicals, as measured by electron spin resonance spectroscopy, which decreased during storage, and most remarkably during the first two weeks of storage, with only marginal increase in lipid hydroperoxides as primary lipid oxidation products, and without any increase in secondary lipid oxidation products. Analyses of volatiles by SPME-GC-MS revealed a significant increase in concentrations of 2-methyl- and 3-methyl butanals, related to Maillard reactions, together with an increase in 2-acetylpyrrole concentration. The soup powders became more brown during storage, as indicated by a decreasing Hunter L-value, in accord with non-enzymatic browning reactions. A significant increase in the concentrations of dimethyl disulfide in soup powder headspace indicated free radical-initiated protein oxidation. Protein degradation, including Maillard reactions and protein oxidation, is concluded to be more important than lipid oxidation in determining the shelf-life of dry cauliflower soup powder. Copyright © 2011 Elsevier Ltd. All rights reserved.

  14. Examination of Zinc Oxide Nanoparticles as a Fluorescent Fingerprint Detection Powder

    International Nuclear Information System (INIS)

    Tun Tun Lin

    2010-12-01

    Detection of latent fingerprint was performed using zinc oxide nanoparticles which were produced by simple and efficient method in aqueous media from zinc nitrate. Synthesized ZnO nanoparticles were characterized by XRD, SEM and AFM for ZnO purification and particle size examination. In this paper an effort has been made to compare the results of using ZnO nanoparticles and conventional fingerprint powders such as ZnO bulk powder, CaO, TiO2, printer toner powder and graphite. Fingerprints on different materials were also examined by the use of ZnO and Graphite powder, which is currently used in the Central Intelligence Department of Myanmar Police Force.From this research, it was observed that zinc oxide nanoparticles powder produced a much clearer picture of the fingerprints, compared to conventional powders and it has very good quality at sticking to the fingerprint residue but not to the background surface.

  15. Surface Characterization and Electrochemical Oxidation of Metal Doped Uranium Dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeongmook; Kim, Jandee; Youn, Young-Sang; Kim, Jong-Goo; Ha, Yeong-Keong; Kim, Jong-Yun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Trivalent element in UO{sub 2} matrix makes the oxygen vacancy from loss of oxygen for charge compensation. Tetravalent element alters lattice parameter of UO{sub 2} due to diameter difference between the tetravalent element and replaced U. These structural changes have significant effect on not only relevant fuel performance but also the kinetics of fuel oxidation. Park and Olander explained the stabilization of Ln (III)-doped UO{sub 2} against oxidation based on oxygen potential calculations. In this work, we have been investigated the effect of Gd{sup 3+} and Th{sup 4+} doping on the UO{sub 2} structure with Raman spectroscopy and X-ray diffraction to characterize the surface structure of nuclear fuel material. For Gd doped UO{sub 2}, its electrochemical oxidation behaviors are also investigated. The Gd and Th doped uranium dioxide solid solution pellets with various doping level were investigated by XRD, Raman spectroscopy, SEM, electrochemical experiments to investigate surface structure and electro chemical oxidation behaviors. The lattice parameter evaluated from XRD spectra indicated the formation of solid solutions. Raman spectra showed the existence of the oxygen vacancy. SEM images showed the grain structure on the surface of Gd doped uranium dioxide depending on doping level and oxygen-to-metal ratio.

  16. The creation of a uranium oxide industry, from the laboratory stage to a pilot plant (1961); Creation d'une industrie de l'oxyde d'uranium du laboratoire a l'usine pilote (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Caillat, R.; Delange, M.; Sauteron, J. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Hauser, R. [Compagnie Industrielle des Combustibles atomiques frittes (France)

    1961-07-01

    The qualities of uranium oxide, in particular its good in-pile characteristics and its resistance to corrosion by the usual heat-exchange fluids, have led to this material being chose at the present time as a nuclear fuel in many power reactors, either planned or under construction. A great effort has been made these last few years in France in studying processes for transforming powdered uranium oxide into a dense material with satisfactory behaviour in a neutron flux. The laboratories at Saclay have studied the physico-chemical features of the phenomena accompanying the calcination of uranium peroxide or ammonium uranate to give uranium trioxide, and the subsequent reduction of the latter to dioxide as well as the sintering of the powders obtained. This work has made it possible on one hand to prepare powder of known specific surface area, and on the other to show the overriding influence of this factor, all other things being equal, on the behaviour of powders during sintering in a hydrogen atmosphere. The work has led to defining two methods for sintering stoichiometric uranium oxide of high density. The technological study of the preparation of the powder and its industrial production are carried out at the plant of Le Bouchet which produces at the moment powders of known characteristics suitable for sintering in hydrogen at 1650 deg. C without prior grinding. The industrial sintering is carried out by the Compagnie industrielle des Combustibles Atomiques Frittes who has set up a pilot plant having a capacity of 25 metric tons/year, for the Commissariat l'Energie Atomique and has been operating this plant since May 1958. This plant is presented by a film entitled 'uranium oxide'. (author) [French] Les qualites de l'oxyde d'uranium, en particulier son bon comportement en pile et sa resistance a la corrosion par les fluides caloporteurs habituels, font choisir aujourd'hui ce materiau comme combustible de nombreux reacteurs de

  17. The creation of a uranium oxide industry, from the laboratory stage to a pilot plant (1961); Creation d'une industrie de l'oxyde d'uranium du laboratoire a l'usine pilote (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Caillat, R; Delange, M; Sauteron, J [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Hauser, R [Compagnie Industrielle des Combustibles atomiques frittes (France)

    1961-07-01

    The qualities of uranium oxide, in particular its good in-pile characteristics and its resistance to corrosion by the usual heat-exchange fluids, have led to this material being chose at the present time as a nuclear fuel in many power reactors, either planned or under construction. A great effort has been made these last few years in France in studying processes for transforming powdered uranium oxide into a dense material with satisfactory behaviour in a neutron flux. The laboratories at Saclay have studied the physico-chemical features of the phenomena accompanying the calcination of uranium peroxide or ammonium uranate to give uranium trioxide, and the subsequent reduction of the latter to dioxide as well as the sintering of the powders obtained. This work has made it possible on one hand to prepare powder of known specific surface area, and on the other to show the overriding influence of this factor, all other things being equal, on the behaviour of powders during sintering in a hydrogen atmosphere. The work has led to defining two methods for sintering stoichiometric uranium oxide of high density. The technological study of the preparation of the powder and its industrial production are carried out at the plant of Le Bouchet which produces at the moment powders of known characteristics suitable for sintering in hydrogen at 1650 deg. C without prior grinding. The industrial sintering is carried out by the Compagnie industrielle des Combustibles Atomiques Frittes who has set up a pilot plant having a capacity of 25 metric tons/year, for the Commissariat l'Energie Atomique and has been operating this plant since May 1958. This plant is presented by a film entitled 'uranium oxide'. (author) [French] Les qualites de l'oxyde d'uranium, en particulier son bon comportement en pile et sa resistance a la corrosion par les fluides caloporteurs habituels, font choisir aujourd'hui ce materiau comme combustible de nombreux reacteurs de puissance en construction ou en

  18. Effect of ethylene oxide sterilization on the osteoinductivity of demineralized allograft bone powder

    International Nuclear Information System (INIS)

    Hamid Reza Aghayan; Babak Arjmand; Mehdi Golestani; Farokh Tirgari

    2008-01-01

    Full text: Ethylene oxide has been widely used for secondary sterilization of bone allograft to reduce the risk of infection and associated complications. In this study we investigate the effects of ethylene oxide gas sterilization on the osteoinductivity of demineralized bone Powder. Eighteen rats received two separate implants consisting of 30 mg aseptically prepared and 30 mg ethylene oxide-sterilized demineralized bone powder. The demineralized bone powder from each group was placed into two separate muscle pouch created in the paravertebral muscles of each rat. After 4 weeks each implantation site was removed with 0.5 cm normal tissue around the implant. Histological examination was done to determine the presence or absence of osteoinduction. All except one of eighteen aseptically prepared demineralized bone powder sites histologically contained new bone elements (94.4%) and fourteen (77.7%) of ethylene oxide sterilized demineralized bone powder sites showed evidence of new bone elements (p>0.05). There is no significant difference in osteoblast formation in two groups. We concluded that ethylene oxide sterilization in 42 degree C did not significantly reduce the osteoinductivity of demineralized bone powder. So ethylene oxide can be considered as a suitable but not perfect method for secondary sterilization of demineralized bone powder. (Author)

  19. Simulation of uranium oxides reduction kinetics by hydrogen. Reactivities of germination and growth; Modelisation de la cinetique de reduction d`oxydes d`uranium par l`hydrogene. Reactivites de germination et de croissance

    Energy Technology Data Exchange (ETDEWEB)

    Brun, C

    1997-12-04

    The aim of this work is to simulate the reduction by hydrogen of the tri-uranium octo-oxide U{sub 3}O{sub 8} (obtained by uranium trioxide calcination) into uranium dioxide. The kinetics curves have been obtained by thermal gravimetric analysis, the hydrogen and steam pressures being defined. The geometrical modeling which has allowed to explain the trend of the kinetics curves and of the velocity curves is an anisotropic germination-growth modeling. The powder is supposed to be formed of spherical grains with the same radius. The germs of the new UO{sub 2} phase appear at the surface of the U{sub 3}O{sub 8} grains with a specific germination frequency. The growth reactivity is anisotropic and is very large in the tangential direction to the grains surface. Then, the uranium dioxide growths inside the grain and the limiting step is the grain surface. The variations of the growth reactivity and of the germination specific frequency in terms of the gases partial pressures and of the temperature have been explained by two different mechanisms. The limiting step of the growth mechanism is the desorption of water in the uranium dioxide surface. Concerning the germination mechanism the limiting step is a water desorption too but in the tri-uranium octo-oxide surface. The same geometrical modeling and the same germination and growth mechanisms have been applied to the reduction of a tri-uranium octo-oxide obtained by calcination of hydrated uranium trioxide. The values of the germination specific frequency of this solid are nevertheless weaker than those of the solid obtained by direct calcination of the uranium trioxide. (O.M.) 45 refs.

  20. Role of oxidizing agent in the chemistry of in-situ uranium leaching

    International Nuclear Information System (INIS)

    Carlson, R.H.; Norris, R.D.; Schellinger, R.

    1982-01-01

    Synthetic two-component mixtures (uraninite and iron sulfide) as well as native uranium ores obtained from Texas and Wyoming have been examined. Physical/chemical ore properties are correlated with observed laboratory leach response. Data show a large inherent selectivity of oxidant for uranium in the early stages of a leach period. Uranium head grade was found to increase in a nearly linear fashion with hydrogen peroxide concentration in the leach solution. As uranium in the ore is depleted, uranium response decreases and the oxidant serves mainly to leach iron sulfide gangue material. 6 refs

  1. Powder of a copper oxide superconductor precursor, fabrication process and use for the preparation of superconducting oxide

    International Nuclear Information System (INIS)

    Dehaudt, P.

    1990-01-01

    The precursor powder comprises at least a copper compound (hydroxide, oxide and hydroxynitrates), at least a rare earth and/or yttrium compound (nitrates, hydroxides and hydroxynitrates) or bismuth oxide and at least an alkaline earth nitrate. It can be prepared by atomization drying of a suspension a copper precipitate or coprecipitate and other elements of the superconducting oxide in solution [fr

  2. The crystal structures and powder diffraction patterns of the uranium tellurides

    Energy Technology Data Exchange (ETDEWEB)

    Snyder, R.L. (State Univ. of New York, Alfred, NY (USA). Inst. of Ceramic Superconductivity); Nichols, M.C.; Boehme, D.R. (Sandia National Labs., Livermore, CA (USA))

    1990-10-03

    A critical review of all of the reported structures and powder diffraction patterns in the uranium telluride system has been undertaken. Structures that are correct: Cubic -- UTe: no experimental pattern exists. Retain calculated 15--865. Cubic --U{sub 3}Te{sub 4}: retain the poor quality 12--610 but adopt the pattern calculated here. Cubic U{sub 2}Te{sub 3}: no experimental pattern exists. Adopt pattern calculated here. Orthorhombic UTe{sub 2}: Adopt the new pattern of Boehme et al. Monoclinic {alpha}UTe{sub 3} Adopt the new pattern of Boehme et al. Monoclinic {alpha}UTe{sub 3} Adopt the new pattern of Boehme et al. Orthorhombic {beta}UTe{sub 3}: Adopt pattern calculated here. Orthorhombic UTe{sub 5}: Adopt the new pattern of Boehme et al. Structures in need of refinement: Orthorhombic U{sub 2}Te{sub 3}:Adopt pattern calculated here over 34--807. Hexagonal U{sub 7}Te{sub 12}: Adopt pattern calculated here but retain 24--1368. Orthorhombic UTe{sub 1.78}: Adopt pattern calculated here and retain our modified 21--1404 reported for U{sub 4}Te{sub 7}. Orthorhombic UTe{sub 2.5}: Adopt pattern calculated here. Orthorhombic UTe{sub 3.4}: Accept recent pattern of Boehme et al. Phases for which no structures or reliable patterns exist: Orthorhombic U{sub 3}Te{sub 4}: no published pattern. Tetragonal U{sub 3}Te{sub 5}: three patterns 21--1407, 34--766 and 34--896 exit but all are of very poor quality. Phases which probably do not exist: Tetragonal UTe{sub 1.78}, Tetragonal UTe{sub 2}, Cubic UTe{sub 2} U{sub 3}Te{sub 7}(21--1402), U{sub 3}Te{sub 8}(21--1406).

  3. Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

    Science.gov (United States)

    Bhattacharya, Sumit

    High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the

  4. Behavior of Colorado Plateau uranium minerals during oxidation

    Science.gov (United States)

    Garrels, Robert Minard; Christ, C.L.

    1956-01-01

    Uranium occurs as U(VI) and U(IV) in minerals of the Colorado Plateau ores. The number of species containing U(VI) is large, but only two U(IV) minerals are known from the Plateau: uraninite, and oxide, and coffinite, a hydroxy-silicate. These oxidize to yield U(VI) before reacting significantly with other mineral constituents. Crystal-structure analysis has shown that U(VI) invariable occurs as uranyl ion, UO2+2. Uranyl ion may form complex carbonate or sulfate ions with resulting soluble compounds, but only in the absence of quinquevalent vanadium, arsenic, or phosphorous. In the presence of these elements in the +5 valence state, the uranyl ion is fixed in insoluble layer compounds formed by union of uranyl ion with orthovanadate, orthophosphate, or orthoarsenate. Under favorable conditions UO2+2 may react to form the relatively insoluble rutherfordine, UO2CO3, or hydrated uranyl hydroxides. These are rarely found on the Colorado Plateau as opposed to their excellent development in other uraniferous areas, a condition which is apparently related to the semiarid climate and low water table of the Plateau. Uranium may also be fixed as uranyl silicate, but little is known about minerals of this kind. In the present study emphasis has been placed on a detailing of the chemical and crystal structural changes which occur in the oxidation paragenetic sequence.

  5. Small cell experiments for electrolytic reduction of uranium oxides to uranium metal using fluoride salts

    International Nuclear Information System (INIS)

    Haas, P.A.; Adcock, P.W.; Coroneos, A.C.; Hendrix, D.E.

    1994-01-01

    Electrolytic reduction of uranium oxide was proposed for the preparation of uranium metal feed for the atomic vapor laser isotope separation (AVLIS) process. A laboratory cell of 25-cm ID was operated to obtain additional information in areas important to design and operation of a pilot plant cell. Reproducible test results and useful operating and control procedures were demonstrated. About 20 kg of uranium metal of acceptable purity were prepared. A good supply of dissolved UO 2 feed at the anode is the most important controlling requirement for efficient cell operation. A large fraction of the cell current is nonproductive in that it does not produce a metal product nor consume carbon anodes. All useful test conditions gave some reduction of UF 4 to produce CF 4 in addition to the reduction of UO 2 , but the fraction of metal from the reduction of UF 4 can be decreased by increasing the concentration of dissolved UO 2 . Operation of large continuous cells would probably be limited to current efficiencies of less than 60 pct, and more than 20 pct of the metal would result from the reduction of UF 4

  6. Sol-gel preparation of uranium oxide spheres

    International Nuclear Information System (INIS)

    Dolezal, J.; Urbanek, V.

    1978-01-01

    Information is presented on problems of preparing nuclear fuel by the sol-gel method. Basic data on different process types are given. A more detailed description of the method of preparation of spherical particles of uranium oxide gel developed and used at the Nuclear Research Institute at Rez is given. Advantages and disadvantages of sol-gel materials are discussed in comparison with fuel materials prepared by classical precipitation methods. The feasibility of the sol-gel methods for preparing other materials is shortly mentioned and their application outlined. (author)

  7. Uranium reduction by carbon oxide during ore formation

    International Nuclear Information System (INIS)

    Matyash, I.V.; Gavrusevich, I.B.; Pasal'skaya, L.F.; Shcherba, D.I.

    1981-01-01

    Using the method of gas chromatography the gas content in Pre-Cambrian granitoils of various types and in natrometasomatites associted with them is studied. It is established that granites associated with ore-bearing albitites have sharply elevated amounts of CO as compared with granites, which do not include mineralization. Simultaneously in ore samples the absence or sharply low amounts of CO as compared with ore-free samples is observed, that is reverse dependence of CO and ore components. Carbon oxide is the reducing agent of uranium mineralization and alongside with other reducing agents can be a geochemical barrier in the process of ore formation [ru

  8. NMR study of nanophase Al/Al-oxide powder and consolidated composites

    International Nuclear Information System (INIS)

    Suits, B.H.; Apte, P.; Wilken, D.E.; Siegel, R.W.

    1994-10-01

    27 Al Nuclear Magnetic Resonance (NMR) measurements from aluminum powders and consolidated nanophase aluminum made from those powders are presented. The signals from the metal and surface oxidation are easily separated and are compared before and after consolidation. The results presented indicate that the oxide coating becomes the interface region within the nanophase composite material and that during consolidation the metal has undergone a deformation equivalent to that seen for bulk material under a compressive strain of between 4% and 8%

  9. Uranium oxide catalysts: environmental applications for treatment of chlorinated organic waste from nuclear industry.

    Science.gov (United States)

    Lazareva, Svetlana; Ismagilov, Zinfer; Kuznetsov, Vadim; Shikina, Nadezhda; Kerzhentsev, Mikhail

    2018-02-05

    Huge amounts of nuclear waste, including depleted uranium, significantly contribute to the adverse environmental situation throughout the world. An approach to the effective use of uranium oxides in catalysts for the deep oxidation of chlorine-containing hydrocarbons is suggested. Investigation of the catalytic activity of the synthesized supported uranium oxide catalysts doped with Cr, Mn and Co transition metals in the chlorobenzene oxidation showed that these catalysts are comparable with conventional commercial ones. Physicochemical properties of the catalysts were studied by X-ray diffraction, temperature-programmed reduction with hydrogen (H 2 -TPR), and Fourier transform infrared spectroscopy. The higher activity of Mn- and Co-containing uranium oxide catalysts in the H 2 -TPR and oxidation of chlorobenzene in comparison with non-uranium catalysts may be related to the formation of a new disperse phase represented by uranates. The study of chlorobenzene adsorption revealed that the surface oxygen is involved in the catalytic process.

  10. Role of thermal analysis in uranium oxide fuel fabrication process

    International Nuclear Information System (INIS)

    Balaji Rao, Y.; Yadav, R.B.

    2006-01-01

    The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO 2+x and U 3 O 8 powders as an alternate technique. Further, the temperature difference (ΔT max ) between the two exothermic peaks obtained in UO 2+x powder oxidation is related to sintered density of UO 2 pellets. (author)

  11. Method of converting uranium fluoride to intermediate product for uranium oxide manufacture with recycling or reusing valuable materials

    International Nuclear Information System (INIS)

    Baran, V.; Moltasova, J.

    1982-01-01

    Uranium fluoride is acted upon by water with nitrate containing a cation capable of binding fluoride ions. The uranium is extracted, for instance, with tributyl phosphate with the generated organic phase containing the prevalent proportion of uranium and representing the required intermediate product and the aqueous phase from which is isolated the fluorine component which may be used within the fuel cycle. The nitrate component of the aqueous phase is recycled following treatment. It is also possible to act on uranium fluoride directly with an aqueous solution. Here the cations of nitrate form with the fluorides soluble nondissociated complexes and reduce the concentration of free fluoride ions. The nitrate +s mostly used in an amount corresponding to its solubility in the system prior to the introduction of UF 6 . The uranium from the solution with the reduced concentration of free fluoride ions is extracted into the reaction system under such conditions as to make the prevalent majority of fluorides and an amount of uranium smaller than 5x10 -2 mol/l remain in the aqueous phase and that such an amount of fluorides should remain in the organic phase which is smaller than corresponds to the fluorine/uranium molar ratio in the organic phase. Uranium contained in the organic phase is processed into uranium oxide, with advantage into UO 2 . From the isolated compounds of fluorine and the cation of the nitrate gaseous HF is released which is used either inside or outside of the fuel cycle. (J.P.)

  12. Thermogravimetric study of reduction of oxides present in oxidized nickel-base alloy powders

    Science.gov (United States)

    Herbell, T. P.

    1976-01-01

    Carbon, hydrogen, and hydrogen plus carbon reduction of three oxidized nickel-base alloy powders (a solid solution strengthened alloy both with and without the gamma prime formers aluminum and titanium and the solid solution strengthened alloy NiCrAlY) were evaluated by thermogravimetry. Hydrogen and hydrogen plus carbon were completely effective in reducing an alloy containing chromium, columbium, tantalum, molybdenum, and tungsten. However, with aluminum and titanium present the reduction was limited to a weight loss of about 81 percent. Carbon alone was not effective in reducing any of the alloys, and none of the reducing conditions were effective for use with NiCrAlY.

  13. Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour

    OpenAIRE

    T. L. Martin; C. Coe; P. A. J. Bagot; P. Morrall; G. D. W Smith; T. Scott; M. P. Moody

    2016-01-01

    Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (∼5 nm) interfacia...

  14. Characterisation of oxidised aluminium powder: Validation of a new anodic oxidation bench

    Energy Technology Data Exchange (ETDEWEB)

    Gascoin, Nicolas, E-mail: Nicolas.Gascoin@univ-orleans.fr [PRISME Institute, Orleans University, 63 avenue de Lattre de Tassigny, 18020 Bourges (France); Gillard, Philippe; Baudry, Guillaume [PRISME Institute, Orleans University, 63 avenue de Lattre de Tassigny, 18020 Bourges (France)

    2009-11-15

    Aluminium powder is of major interest in many applications but it presents a risk due to its high explosibility, particularly when dispersed in air. The safety is directly linked to the particles oxidation because the Minimum Ignition Energy (MIE), which is required to initiate an Al dust explosion, increases with the oxide layer thickness. This study provides a controlled method to furnish reproducible homogeneous set of powder for such safety studies. Thanks to a new experimental bench, the influence on the oxidation rate of seven treatment parameters is investigated (current density, time of treatment, acid concentration, mass of powder, particles size, stirring, neutralisation by ammonia solution). The oxide content is plotted versus the current density, the time and the acid concentration to provide reference curves for further elaboration of oxidised powder. The particles size of sieved powder is measured before and after treatment by different methods (optical and Scanning Electron Microscopes, laser measurement). A high refinement of the powder in terms of size distribution is achieved thanks to the employed sieving. The present bench and the elaborated procedure are of great interest to provide well-calibrated oxidised powder directly available for safety studies. The time must be adjusted, depending on the wanted oxide content - from 2 to 18 wt.% - and the other treatment parameters must be kept constant: acid concentration (5 wt.%), current density (1 A dm{sup -2}), treated powder (20 g). In these conditions, the ratio of the oxide layer thickness on the particles diameter is found to be constant for a given oxide content whatever the particles size.

  15. Characterisation of oxidised aluminium powder: Validation of a new anodic oxidation bench

    International Nuclear Information System (INIS)

    Gascoin, Nicolas; Gillard, Philippe; Baudry, Guillaume

    2009-01-01

    Aluminium powder is of major interest in many applications but it presents a risk due to its high explosibility, particularly when dispersed in air. The safety is directly linked to the particles oxidation because the Minimum Ignition Energy (MIE), which is required to initiate an Al dust explosion, increases with the oxide layer thickness. This study provides a controlled method to furnish reproducible homogeneous set of powder for such safety studies. Thanks to a new experimental bench, the influence on the oxidation rate of seven treatment parameters is investigated (current density, time of treatment, acid concentration, mass of powder, particles size, stirring, neutralisation by ammonia solution). The oxide content is plotted versus the current density, the time and the acid concentration to provide reference curves for further elaboration of oxidised powder. The particles size of sieved powder is measured before and after treatment by different methods (optical and Scanning Electron Microscopes, laser measurement). A high refinement of the powder in terms of size distribution is achieved thanks to the employed sieving. The present bench and the elaborated procedure are of great interest to provide well-calibrated oxidised powder directly available for safety studies. The time must be adjusted, depending on the wanted oxide content - from 2 to 18 wt.% - and the other treatment parameters must be kept constant: acid concentration (5 wt.%), current density (1 A dm -2 ), treated powder (20 g). In these conditions, the ratio of the oxide layer thickness on the particles diameter is found to be constant for a given oxide content whatever the particles size.

  16. On bacteria oxidizing enlargement scale test for uranium in-situ leaching at. 381 mine

    International Nuclear Information System (INIS)

    Hu Kaiguang; Wang Qingliang; Liu Yingjiu; Shi Wenge; Hu Shihe; Hu Yincai; Fang Qiu

    1999-01-01

    The results of enlarged scale test of bacteria as oxidizer for uranium in-situ leaching at No 381 mine showed that redox potential of the oxidized absorbed tailing water by bacteria is more than 510 mV, without any effects on after treatments by using bacteria as oxidizer and reduce oxidizer costs 70% compared with H 2 O 2 as oxidizer

  17. Grain growth kinetics in uranium-plutonium mixed oxides

    International Nuclear Information System (INIS)

    Sari, C.

    1986-01-01

    Grain growth rates were investigated in uranium-plutonium mixed oxide specimens with oxygen-to-metal ratios 1.97 and 2.0. The specimens in the form of cylindrical pellets were heated in a temperature gradient similar to that existing in a fast reactor. The results are in agreement with the cubic rate law. The mean grain size D(μm) after annealing for time t (min) is represented by D 3 -D 0 3 =1.11x10 12 . exp(-445870/RT).t and D 3 -D 0 3 =2.55x10 9 .exp(-319240/RT).t for specimens with overall oxygen-to-metal ratios 1.97 and 2.0, respectively (activation energies expressed in J/mol). An example for the influence of the oxygen-to-metal ratio on the grain growth in mixed oxide fuel during operation in a fast reactor is also given. (orig.)

  18. Study of uranium oxidation states in geological material.

    Science.gov (United States)

    Pidchenko, I; Salminen-Paatero, S; Rothe, J; Suksi, J

    2013-10-01

    A wet chemical method to determine uranium (U) oxidation states in geological material has been developed and tested. The problem faced in oxidation state determinations with wet chemical methods is that U redox state may change when extracted from the sample material, thereby leading to erroneous results. In order to quantify and monitor U redox behavior during the acidic extraction in the procedure, an analysis of added isotopic redox tracers, (236)U(VI) and (232)U(IV), and of variations in natural uranium isotope ratio ((234)U/(238)U) of indigenous U(IV) and U(VI) fractions was performed. Two sample materials with varying redox activity, U bearing rock and U-rich clayey lignite sediment, were used for the tests. The Fe(II)/Fe(III) redox-pair of the mineral phases was postulated as a potentially disturbing redox agent. The impact of Fe(III) on U was studied by reducing Fe(III) with ascorbic acid, which was added to the extraction solution. We observed that ascorbic acid protected most of the U from oxidation. The measured (234)U/(238)U ratio in U(IV) and U(VI) fractions in the sediment samples provided a unique tool to quantify U oxidation caused by Fe(III). Annealing (sample heating) to temperatures above 500 °C was supposed to heal ionizing radiation induced defects in the material that can disturb U redox state during extraction. Good agreement between two independent methods was obtained for DL-1a material: an average 38% of U(IV) determined by redox tracer corrected wet chemistry and 45% for XANES. Copyright © 2013 Elsevier Ltd. All rights reserved.

  19. Guideline tests on the corrosion of sintered uranium oxide by lead and sodium about 450 deg. C

    International Nuclear Information System (INIS)

    Portnoff, A.; Pointud, R.

    1958-05-01

    Within the frame of the investigation of behaviour of a fuel element (such as UO 2 ) under irradiation, the authors report the study of the physical-chemical action of the coolant at different temperatures on the body to be irradiated. Thus, sintered uranium oxide has been submitted to corrosion by lead and by sodium during 250 hours at temperatures between 400 and 500 C. The physical characteristics of the UO 2 powder and of different sintered UO 2 pellets produced from this powder under different sintering processes are indicated, as well as the results of a spectrographic analysis of the sintered UO 2 . Corrosion devices, treatments and obtained results are reported for corrosion by lead and by sodium. In the latter case, extraction processes are discussed (using butyl alcohol, or melting in vaseline oil)

  20. The FTIR study of uranium oxides by the method of light pipe reflection spectroscopy

    International Nuclear Information System (INIS)

    Bao Zhu Yu; Hansen, W.N.

    1988-01-01

    Light pipe infrared reflection spectra of UO 2 , UO 3 , U 3 O 8 have been studied by using an FTIR spectrometer. The uranium oxide powders were ground to ensure fine particle size and distributed on the inner surface of a straight glass pipe with gold coating. The infrared beam from the inter-ferometer was focused into one end of the pipe at 45 0 incidence and then the transmitted beam was refocused by a pair of Cassegrainian type mirrors. The resultant spectra show the infrared characteristics of the ...-U-O-U-O-..., uranyl ion UO 2 2+ bond vibration and the active lattice vibrations predicted by group theory calculations. In comparison to the transmission spectra measured by authors or reported in literature, this 45 0 incident light pipe method as well as the previous light pipe method offer advantages of sensitivity, ease of acquisition and interpretation, and require a very small sample. It confirms the power of the light pipe method for studying powders and its special utility for the infrared studies of hazardous materials. (Author)

  1. The pressure bonding ability of uranium dioxide powders in relation to the evolution of their surface properties

    International Nuclear Information System (INIS)

    Danroc, J.

    1982-09-01

    The long term storage of sinterable uranium dioxide powders generally improves their pressure bonding ability and the strength of the resulting green pellets. Evidence of the gradual evolution of the surface texture and composition of these powders during storage at room temperature and pressure has been provided by infrared spectroscopy, X-ray diffraction and thermogravimetric and microcalorimetric methods. These techniques demonstrated the existence of a thin adherent surface layer of UO 3 2H 2 0. Such a natural evolutionary process can be reproduced and substantially amplified by subjecting the powder to thermal treatments at temperatures up to 90 0 C in a moist air environment. It was shown that powder treated in this manner could be more readily compacted into strong green pellets than could raw material. The tensile strength, commonly regarded as a quality test for such pellets and measured by the brazilian method, was found to be at least twice that of normal pellets. The high density and geometric integrity of these sintered products ensures the extrapolation of these preparation techniques to the mass production of nuclear reactor fuel pellets [fr

  2. Production of uranium metal via electrolytic reduction of uranium oxide in molten LiCl and salt distillation

    International Nuclear Information System (INIS)

    Eun-Young Choi; Chan Yeon Won; Dae-Seung Kang; Sung-Wook Kim; Ju-Sun Cha; Sung-Jai Lee; Wooshin Park; Hun Suk Im; Jin-Mok Hur

    2015-01-01

    Recovery of metallic uranium has been achieved by electrolytic reduction of uranium oxide in a molten LiCl-Li 2 O electrolyte at 650 deg C, followed by the removal of the residual salt by vacuum distillation at 850 deg C. Four types of stainless steel mesh baskets, with various mesh sizes (325, 1,400 and 2,300 meshes) and either three or five ply layers, were used both as cathodes and to contain the reduced product in the distillation stage. The recovered uranium had a metal fraction greater than 98.8 % and contained no residual salt. (author)

  3. Determination of the oxygen-metal-ratio of uranium-americium mixed oxides

    International Nuclear Information System (INIS)

    Bartscher, W.

    1982-01-01

    During the dissolution of uranium-americium mixed oxides in phosphoric acid under nitrogen tetravalent uranium is oxidized by tetravalent americium. The obtained hexavalent uranium is determined by constant potential coulometry. The coulombs measured are equivalent to the oxygen in excess of the minimum composition of UO 2 x AmO 1 . 5 . The total uranium content of the sample is determined in a subsequent coulometric titration. The oxygen-metal ratio of the sample can be calculated for a given uranium-americium ratio. An excess of uranium dioxide is necessary in order to suppress the oxidation of water by tetravalent americium. The standard deviation of the method is 0.0017 O/M units. (orig.) [de

  4. Dissolution of thorium/uranium mixed oxide in nitric acid-hydrofluoric acid solution

    International Nuclear Information System (INIS)

    Filgueiras, S.A.C.

    1984-01-01

    The dissolution process of thorium oxide and mixed uranium-thorium oxide is studied, as a step of the head-end of the fuel reprocessing. An extensive bibliography was analysed, concerning the main aspects of the system, specially the most important process variables. Proposed mechanisms and models for the thorium oxide dissolution are presented. The laboratory tests were performed in two phases: at first, powdered thoria was used as the material to be dissolved. The objective was to know how changes in he concentrations of the dissolvent solution components HNO 3 , HF and Al(NO 3 ) 3 affect the dissolution rate. The tests were planned according to the fractional factorial method. Thes results showed that it is advantageous to work with powdered material, since the reaction occurs rapidly. And, if the Thorex solution (HNO 3 13M, HF 0.05M and Al(NO 3 ) 3 0.10M) is a suitable dissolvent, it was verified that it is possible to reduce the concentration of either nitric or fluoridric acid, without reducing the reaction rate to an undesirable value. It was also observed significant interaction between the components of the dissolvent solution. In the second phase of the tests, (Th, 5%U)O 2 sintered pellets were used. The main goals were to know the pellets dissolution behaviour and to compare the results for different pellets among themselves. It was observed that the metallurgical history of the material strongly influences its dissolution, specially the density and the microstructure. It was also studied how the (Th,U)O 2 mass/Thorex solution volume ratio affects the time needed to obtain an 1 M Th/liter solution. The activation energy for the reaction was obtained. (Author) [pt

  5. Oxidized Metal Powders for Mechanical Shock and Crush Safety Enhancers; TOPICAL

    International Nuclear Information System (INIS)

    GARINO, TERRY J.

    2002-01-01

    The use of oxidized metal powders in mechanical shock or crush safety enhancers in nuclear weapons has been investigated. The functioning of these devices is based on the remarkable electrical behavior of compacts of certain oxidized metal powders when subjected to compressive stress. For example, the low voltage resistivity of a compact of oxidized tantalum powder was found to decrease by over six orders of magnitude during compaction between 1 MPa, where the thin, insulating oxide coatings on the particles are intact, to 10 MPa, where the oxide coatings have broken down along a chain of particles spanning the electrodes. In this work, the behavior of tantalum and aluminum powders was investigated. The low voltage resistivity during compaction of powders oxidized under various conditions was measured and compared. In addition, the resistivity at higher voltages and the dielectric breakdown strength during compaction were also measured. A key finding was that significant changes in the electrical properties persist after the removal of the stress so that a mechanical shock enhancer is feasible. This was verified by preliminary shock experiments. Finally, conceptual designs for both types of enhancers are presented

  6. Formation of oxides particles in ferritic steel by using gas-atomized powder

    International Nuclear Information System (INIS)

    Liu Yong; Fang Jinghua; Liu Donghua; Lu Zhi; Liu Feng; Chen Shiqi; Liu, C.T.

    2010-01-01

    Oxides dispersion strengthened (ODS) ferritic steel was prepared by using gas-atomized pre-alloyed powder, without the conventional mechanical alloying process. By adjusting the volume content of O 2 in the gas atmosphere Ar, the O level in the ferritic powder can be well controlled. The O dissolves uniformly in the ferritic powder, and a very thin layer of oxides forms on the powder surface. After hot deformation, the primary particle boundaries, which retain after sintering, can be disintegrated and near fully dense materials can be obtained. The oxide layer on the powder surface has a significant effect on the microstructural evolution. It may prevent the diffusion in between the primary particles during sintering, and may dissolve and/or induce the nucleation of new oxides in the ferritic matrix during recrystallization. Two kinds of oxide particles are found in the ferritic steel: large (∼100 nm) Ti-rich and fine (10-20 nm) Y-Ti-rich oxides. The hardness of the ferritic steel increases with increasing annealing temperatures, however, decreases at 1400 deg. C, due to the coarsening of precipitates and the recrystallization microstructure.

  7. A gravimetric method for the determination of oxygen in uranium oxides and ternary uranium oxides by addition of alkaline earth compounds

    International Nuclear Information System (INIS)

    Fujino, Takeo; Tagawa, Hiroaki; Adachi, Takeo; Hashitani, Hiroshi

    1978-01-01

    A simple gravimetric determination of oxygen in uranium oxides and ternary uranium oxides is described. In alkaline earth uranates which are formed by heating in air at 800-1100 0 C, uranium is in the hexavalent state over certain continuous ranges of alkaline earth-to-uranium ratios. Thus, if an alkaline earth uranate or a compound containing an alkaline earth element, e.g. MgO, is mixed with the oxide sample and heated in air under suitable conditions, oxygen can be determined from the weight change before and after the reaction. The standard deviation of the O:U ratio for a UOsub(2+x) test sample is +-0.0008-0.001, if a correction is applied for atmospheric moisture absorbed during mixing. (Auth.)

  8. Reoxidation of uranium in electrolytically reduced simulated oxide fuel during residual salt distillation

    International Nuclear Information System (INIS)

    Eun-Young Choi; Jin-Mok Hur; Min Ku Jeon; University of Science and Technology, Yuseong-gu, Daejeon

    2017-01-01

    We report that residual salt removal by high-temperature distillation causes partial reoxidation of uranium metal to uranium oxide in electrolytically reduced simulated oxide fuel. Specifically, the content of uranium metal in the above product decreases with increasing distillation temperatures, which can be attributed to reoxidation by Li 2 O contained in residual salt (LiCl). Additionally, we estimate the fractions of Li 2 O reacted with uranium metal under these conditions, showing that they decrease with decreasing temperature, and calculate some thermodynamic parameters of the above reoxidation. (author)

  9. Preparation and Characterization of Uranium Oxides in Support of the K Basin Sludge Treatment Project

    Energy Technology Data Exchange (ETDEWEB)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2008-07-08

    Uraninite (UO2) and metaschoepite (UO3·2H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21°C and 50°C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.004±0.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21°C than the particles prepared at 50°C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.

  10. Preparation and Characterization of Uranium Oxides in Support of the K Basin Sludge Treatment Project

    International Nuclear Information System (INIS)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2008-01-01

    Uraninite (UO2) and metaschoepite (UO3-2H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21 C and 50 C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.004 ± 0.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21 C than the particles prepared at 50 C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.

  11. Thermogravimetric study of the reaction of uranium oxides with fluorine

    International Nuclear Information System (INIS)

    Komura, Motohiro; Sato, Nobuaki; Kirishima, Akira; Tochiyama, Osamu

    2008-01-01

    Thermogravimetric study of the reaction of uranium oxides with fluorine was conducted by TG-DTA method using anti-corrosion type differential thermobalance. When UO 2 was heated from R.T. to 500 deg. C in 5% F 2 /He atmosphere, the weight increase appeared at ca. 250 deg. C with an exothermic peak. Then the weight decreased slightly with a small exothermic peak followed by the complete volatilization with a large exothermic peak at ca. 350 deg. C. At a flow rate of 15, 30, 60 ml min -1 , there seemed to be no significant change for the fluorination of UO 2 . With the different heating rates of 1, 2, 5 and 10 deg. C min -1 , the fluorination peak shifted to higher temperature with increasing heating rates. For the comparison with thermogravimetric results, phase analysis by XRD method was conducted for the products obtained at different temperatures. At 260 deg. C, the product was UO 2 with a small amount of the intermediate compound, UO 2 F. The amount of this compound increased with increasing temperature up to 320 deg. C. Then another phase of UO 2 F 2 appeared at 340 deg. C but it was immediately fluorinated to the volatile fluoride. When U 3 O 8 was used as a starting material, it was found that the steep weight decrease peak appeared at ca. 350 deg. C and the uranium volatilized completely. This result suggests that fluorination of U 3 O 8 occurs at this temperature forming UF 6 . Uranium trioxide showed the similar fluorination behavior to that of U 3 O 8

  12. Tailored Core Shell Cathode Powders for Solid Oxide Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Swartz, Scott [NexTech Materials, Ltd.,Lewis Center, OH (United States)

    2015-03-23

    In this Phase I SBIR project, a “core-shell” composite cathode approach was evaluated for improving SOFC performance and reducing degradation of lanthanum strontium cobalt ferrite (LSCF) cathode materials, following previous successful demonstrations of infiltration approaches for achieving the same goals. The intent was to establish core-shell cathode powders that enabled high performance to be obtained with “drop-in” process capability for SOFC manufacturing (i.e., rather than adding an infiltration step to the SOFC manufacturing process). Milling, precipitation and hetero-coagulation methods were evaluated for making core-shell composite cathode powders comprised of coarse LSCF “core” particles and nanoscale “shell” particles of lanthanum strontium manganite (LSM) or praseodymium strontium manganite (PSM). Precipitation and hetero-coagulation methods were successful for obtaining the targeted core-shell morphology, although perfect coverage of the LSCF core particles by the LSM and PSM particles was not obtained. Electrochemical characterization of core-shell cathode powders and conventional (baseline) cathode powders was performed via electrochemical impedance spectroscopy (EIS) half-cell measurements and single-cell SOFC testing. Reliable EIS testing methods were established, which enabled comparative area-specific resistance measurements to be obtained. A single-cell SOFC testing approach also was established that enabled cathode resistance to be separated from overall cell resistance, and for cathode degradation to be separated from overall cell degradation. The results of these EIS and SOFC tests conclusively determined that the core-shell cathode powders resulted in significant lowering of performance, compared to the baseline cathodes. Based on the results of this project, it was concluded that the core-shell cathode approach did not warrant further investigation.

  13. The Dissolution of Uranium Oxides in HB-Line Phase 1 Dissolvers

    International Nuclear Information System (INIS)

    Gray, J.H.

    2003-01-01

    A series of characterization and dissolution studies has been performed to define flowsheet conditions for the dissolution of uranium oxide materials in dissolvers. The samples selected for analysis were uranium oxide materials. The selection of these uranium oxide materials for characterization and dissolution studies was based on high enriched uranium content and trace levels of plutonium. Test results from the characterization study identified ferric oxide (Fe2O3) and iron/chromium/nickel (Fe/Cr/Ni) particles as impurities along with the tri-uranium oxide (U3O8) and uranium trioxide (UO3). The weight percent uranium in this material was found to vary depending on the impurity content. The trace impurity plutonium appears to be associated with the Fe/Cr/Ni particles. A small amount of absorbed moisture and waters of hydration is present. Most of the uranium oxides easily dissolved in low-molar nitric acid solutions without fluoride within one to two hours at solution temperature s between 60-80 degrees C. A small amount of residue remained following this dissolution step. To assure complete dissolution of uranium from these oxide materials, an additional dissolution step at 90 degrees C to boiling for at least one to two hours has been suggested. Only trace amounts of iron associated with Fe2O3 and Fe/Cr/Ni particles will dissolve during the dissolution steps. Neither hydrogen nor heat will be generated during the dissolution of these uranium oxide materials in nitric acid solutions. Some brown nitrogen dioxide (NO2) fumes will be generated during the dissolution of U3O8

  14. Oxidative lixiviation of pitchblende and precipitation of uranium with hydrogen

    International Nuclear Information System (INIS)

    Mouret, P.

    1958-01-01

    Earlier work on the preparation of uranium by F.A. Forward and his colleagues has shown the possibilities presented by oxidative lixiviation of ores in a carbonate medium, and the catalytic reduction of uranyl carbonate solutions by hydrogen. The carbonate attack is of considerable interest because of the selectivity of the uranium dissolution, which means it can be applied particularly to the treatment of low grade ores with a reduced consumption of cheap reagents. The subsequent reduction with hydrogen is of the same nature, and not only enables relatively dilute uranyl carbonate solutions to be treated, but also avoids any significant alteration of the attacking solution which can therefore be used again in the lixiviation stage. The experimental work, undertaken at the request of the Commissariat a I'Energie Atomique, was aimed at determining the quantitative characteristics of each of the two stages in order to ascertain their possibilities for industrial application to the principal low grade ores found in France. (author) [fr

  15. Solid state processing of massive uranium mononitride, using uranium and uranium higher nitride powders as starting materials (1962); Preparation a l'etat solide de mononitrure d'uranium massif a partir de poudres d'uranium et de nitrures superieurs d'uranium (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Molinari, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1960-12-15

    The mechanism and the optimum conditions for preparing uranium mononitride have been studied. The results have been used for hot pressing (250 kg/cm{sup 2}, 1000 deg. C, under vacuum) a mixture of powders of uranium and uranium higher nitrides. The products obtained have been identified by X-ray measurements and may be - at will and depending upon the stoichiometry - either UN, or a cermet a U{sub {alpha}}-UN. As revealed by the curved shape of grain boundaries, the sinters obtained here do not easily evolve towards physico-chemical equilibrium when submitted to heat treatment. This behaviour is quite different from the one observed with uranium monocarbide prepared by a similar method. This fact may be ascribed to the insolubility in the matrix UN of particles of UO{sub 2} being present as impurities. The density, hardness and thermal conductivity of these products are higher than those measured on uranium nitride or cermets U-UN obtained by other methods. (author) [French] Apres une etude prealable du mecanisme et des conditions optimales de nitruration de l'uranium, on a montre qu'il est possible de preparer par frittage sous charge (250 kg/cm{sup 2}, 1000 deg. C sous vide) d'un melange de poudres d'uranium et de nitrures superieurs d'uranium, un produit qui a ete identifie par diffraction de rayons X. On peut ainsi obtenir a volonte, soit le monocarbure UN, soit un cermet U{sub {alpha}}-UN dans le cas de compositions sous-stoechiometriques. Au contraire du monocarbure d'uranium prepare dans des conditions analogues, les produits obtenus ici, soumis a un traitement thermique, n'evoluent pas facilement vers un etat d'equilibre physico-chimique caracterise par l'existence de joints de grains rectilignes. On attribue ce phenomene a l'insolubilite de l'impurete UO{sub 2} dans UN. La densite, la durete, la conductibilite thermique de ces produits se revelent superieures a celles des nitrures d'uranium ou des cermets U-UN obtenus par les autres methodes. (auteur)

  16. Fire hazards analysis for the uranium oxide (UO3) facility

    International Nuclear Information System (INIS)

    Wyatt, D.M.

    1994-01-01

    The Fire Hazards Analysis (FHA) documents the deactivation end-point status of the UO 3 complex fire hazards, fire protection and life safety systems. This FHA has been prepared for the Uranium Oxide Facility by Westinghouse Hanford Company in accordance with the criteria established in DOE 5480.7A, Fire Protection and RLID 5480.7, Fire Protection. The purpose of the Fire Hazards Analysis is to comprehensively and quantitatively assess the risk from a fire within individual fire areas in a Department of Energy facility so as to ascertain whether the objectives stated in DOE Order 5480.7, paragraph 4 are met. Particular attention has been paid to RLID 5480.7, Section 8.3, which specifies the criteria for deactivating fire protection in decommission and demolition facilities

  17. Dosage of boron traces in graphite, uranium and beryllium oxide

    International Nuclear Information System (INIS)

    Coursier, J.; Hure, J.; Platzer, R.

    1955-01-01

    The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [fr

  18. Caramel, uranium oxide fuel plates for water cooled reactors

    International Nuclear Information System (INIS)

    Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

    1979-01-01

    The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

  19. U3O8 obtained from metallic uranium

    International Nuclear Information System (INIS)

    Lopez, Marisol; Gonzalez, Alfredo; Pasqualini, Enrique E.

    2003-01-01

    Enriched uranium oxide, U 3 O 8 , used as nuclear powder in MTR's, can be obtained by direct oxidation of metallic uranium at 800 C degrees. Maximum density, 8.2 gr/cm 3 , is achieved after grinding and a high temperature treatment at 1400 C degrees. All the process is highly controllable and performed in dry environments. (author)

  20. Oxidation kinetics of hydride-bearing uranium metal corrosion products

    Science.gov (United States)

    Totemeier, Terry C.; Pahl, Robert G.; Frank, Steven M.

    The oxidation behavior of hydride-bearing uranium metal corrosion products from Zero Power Physics Reactor (ZPPR) fuel plates was studied using thermo-gravimetric analysis (TGA) in environments of Ar-4%O 2, Ar-9%O 2, and Ar-20%O 2. Ignition of corrosion product samples from two moderately corroded plates was observed between 125°C and 150°C in all environments. The rate of oxidation above the ignition temperature was found to be dependent only on the net flow rate of oxygen in the reacting gas. Due to the higher net oxygen flow rate, burning rates increased with increasing oxygen concentration. Oxidation rates below the ignition temperature were much slower and decreased with increasing test time. The hydride contents of the TGA samples from the two moderately corroded plates, determined from the total weight gain achieved during burning, were 47-61 wt% and 29-39 wt%. Samples from a lightly corroded plate were not reactive; X-ray diffraction (XRD) confirmed that they contained little hydride.

  1. Oxidation kinetics of hydride-bearing uranium metal corrosion products

    International Nuclear Information System (INIS)

    Totemeier, T.C.; Pahl, R.G.; Frank, S.M.

    1998-01-01

    The oxidation behavior of hydride-bearing uranium metal corrosion products from zero power physics reactor (ZPPR) fuel plates was studied using thermo-gravimetric analysis (TGA) in environments of Ar-4%O 2 , Ar-9%O 2 , and Ar-20%O 2 . Ignition of corrosion product samples from two moderately corroded plates was observed between 125 C and 150 C in all environments. The rate of oxidation above the ignition temperature was found to be dependent only on the net flow rate of oxygen in the reacting gas. Due to the higher net oxygen flow rate, burning rates increased with increasing oxygen concentration. Oxidation rates below the ignition temperature were much slower and decreased with increasing test time. The hydride contents of the TGA samples from the two moderately corroded plates, determined from the total weight gain achieved during burning, were 47-61 wt% and 29-39 wt%. Samples from a lightly corroded plate were not reactive; X-ray diffraction (XRD) confirmed that they contained little hydride. (orig.)

  2. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-09-20

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

  3. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project

  4. Microstructural evolution of ferritic steel powder during mechanical alloying with iron oxide

    Energy Technology Data Exchange (ETDEWEB)

    Wen, Yuren; Liu, Yong; Liu, Donghua; Tang, Bei [Central South Univ., State Key Lab. of Powder Metallurgy, Changsha (China); Liu, C.T. [The Hong Kong Polytechnic Univ., Dept. of Mechanical Engineering, Hong Kong (China)

    2011-02-15

    Mechanical alloying of mixed powders is of great importance for preparing oxide dispersion strengthened ferritic steels. In this study, the microstructural evolution of ferritic steel powder mixed with TiH{sub x}, YH{sub 2} and Fe{sub 2}O{sub 3} in the process of mechanical alloying is systematically investigated by using X-ray diffraction analysis, scanning electron microscopy, transmission electron microscopy and microhardness tests. It is found that titanium, yttrium hydrides and iron oxide are completely dissolved during milling, and homogeneous element distribution can be achieved after milling for 12 h. The disintegration of the composite powder particles occurs at 24 h and reaches the balance of welding and fracturing after 36 h. The oxygen content increases sharply with the disintegration of powder particles due to the absorption of oxygen at the solid/gas interface from the milling atmosphere, which is the main source of extra oxygen in the milled powder. Grain refinement down to nanometer level occurs due to the severe plastic deformation of particles; however, the grain size does not change much with further disintegration of particles. The hardness increases with milling time and then becomes stable during further milling. The study indicates that the addition of iron oxide and hydrides may be more beneficial for the dispersion and homogenization of chemical compositions in the powder mixture, thus shortening the mechanical alloying process. (orig.)

  5. Study of uranium oxide milling in order to obtain nanostructured UC{sub x} target

    Energy Technology Data Exchange (ETDEWEB)

    Guillot, Julien, E-mail: guillotjulien@ipno.in2p3.fr [Institut de Physique Nucléaire CNRS/IN2P3 UMR 8608 – Université Paris Sud, F-91406 ORSAY Cedex (France); Tusseau-Nenez, Sandrine; Roussière, Brigitte; Barré-Boscher, Nicole [Institut de Physique Nucléaire CNRS/IN2P3 UMR 8608 – Université Paris Sud, F-91406 ORSAY Cedex (France); Brisset, François [ICMMO UMR 8182 CNRS – Université Paris Sud, F-91405 Orsay Cedex (France); Mhamed, Maher Cheikh; Lau, Christophe [Institut de Physique Nucléaire CNRS/IN2P3 UMR 8608 – Université Paris Sud, F-91406 ORSAY Cedex (France); Nowak, Sophie [ITODYS, Université Paris Diderot, Sorbonne Paris Cité, CNRS UMR-7086, F-75013 Paris (France); IPSL CNRS UMR 7583 Universités Paris Est Créteil et Paris Diderot, F-94010 Créteil Cedex (France)

    2016-05-01

    A R&D program is developed at the ALTO facility to provide new beams of exotic neutron-rich nuclei, as intense as possible. In the framework of European projects, it has been shown that the use of refractory targets with nanometric structure allows us to obtain beams of nuclei unreachable until now. The first parameter to be controlled in the processing to obtain targets with a homogeneous nanostructure is the grinding of uranium dioxide, down to 100 nm grain size. In this study, dry and wet grinding routes are studied and the powders are analyzed in terms of phase stabilization, specific surface area and grain morphology. It appears that the grinding, as well dry as wet, leads to the decrease of the particle size. The oxidation of UO{sub 2} is observed whatever the grinding. However, the dry grinding is the most efficient and leads to the oxidation of UO{sub 2} into U{sub 4}O{sub 9} and U{sub 3}O{sub 7} whose quantities increase with the grinding time while crystallite sizes decrease.

  6. Modeling conversion of ammonium diuranate (ADU) into uranium dioxide (UO{sub 2}) powder

    Energy Technology Data Exchange (ETDEWEB)

    Hung, Nguyen Trong; Thuan, Le Ba [Institute for Technology of Radioactive and Rare Elements (ITRRE), 48 Lang Ha, Dong Da, Ha Noi (Viet Nam); Khoai, Do Van [Institute for Technology of Radioactive and Rare Elements (ITRRE), 48 Lang Ha, Dong Da, Ha Noi (Viet Nam); Current Postdoctoral Fellow at Tokai Reprocessing Technology Development Center, Japan Atomic Energy Agency (JAEA), 4-33 Tokaimura, Nakagun, Ibaraki, 319-1194 (Japan); Lee, Jin-Young, E-mail: jylee@kigam.re.kr [Convergence Research Center for Development of Mineral Resources (DMR), Korea Institute of Geoscience and Mineral Resources (KIGAM), Daejeon, 34132 (Korea, Republic of); Jyothi, Rajesh Kumar, E-mail: rkumarphd@kigam.re.kr [Convergence Research Center for Development of Mineral Resources (DMR), Korea Institute of Geoscience and Mineral Resources (KIGAM), Daejeon, 34132 (Korea, Republic of)

    2016-10-15

    In the paper, Brandon mathematical model that describes the relationship between the essential fabrication parameters [reduction temperature (T{sub R}), calcination temperature (T{sub C}), calcination time (t{sub C}) and reduction time (t{sub R})] and specific surface area of ammonium diuranate (ADU)-derived UO{sub 2} powder products was established. The proposed models can be used to predict and control the specific surface area of UO{sub 2} powders prepared through ADU route. Suitable temperatures for conversion of ADU and ammonium uranyl carbonate (AUC) was examined with the proposed model through assessment of the sinterability of UO{sub 2} powders.

  7. Investigation of the oxidative processes in intermetallic Sm Co5 powder during heat treatment

    International Nuclear Information System (INIS)

    Talijan, Nadezda M.; Milutinovic-Nikolic, Aleksandra; Stajic-Trosic, Jasna T.; Jovanovic, Zarko D.

    1996-01-01

    Understanding of the thermal stability of intermetallic Sm Co 5 powder is essential for designing the working atmosphere in all phases of the technological procedure in the production of sintered Sm Co 5 magnets to obtain maximal magnetic properties. The thermal stability of the Sm Co 5 powder with defined chemical composition and particle size was investigated in the interval from 20 to 900 deg C. Commercial Sm Co 5 powder was used in this experiment. The powder was milled in anhydrous toluene in an agate mortar to fine powder of quality used in the production of sintered magnets. All the experiments were carried out with powder of an average particle size of 7.23μm, established by SEM. THe thermal stability of the Sm Co 5 powder in static air atmosphere was investigated by thermogravimetric analysis (TGA) using a DuPont Thermal Analyzer. Investigation of the behaviour of Sm Co 5 powder during heating was carried out using new samples of Sm Co 5 powder for each of the investigated temperature cycles. It was found by TGA that up to 200 deg C, the oxidation of Sm Co 5 was negligible. X-ray diffraction of the thermogravimetric experimental residue of the Sm Co 5 powder, heated at 240 deg C, yielded only the presence of the Sm Co 5 phase. By X-ray diffraction different crystal forms were identified depending on the maximal heating temperature. The following phases were identified: Sm 2 O 3 , Co, Co O, Co 3 O 4 and Sm Co O 3 . According to TG and X-ray results, for each of the investigated temperatures, the corresponding chemical reactions were established. The experimental data from both the thermal and X-ray investigations confirm that the phases of pressing and aligning the Sm Co 5 powder, in the process of producing sintered Sm Co 5 magnets, may be performed without a protective atmosphere. (author)

  8. Determination of Uranium In UO2 And U3O8 Powder Using UV-VIS Spectrophotometry

    International Nuclear Information System (INIS)

    Natalia Adventini; Diah Dwiana Lestiani; Muhayatun; Endah Damastuti

    2009-01-01

    Lab. TAR PTNBR BATAN - Bandung has been accredited by National Accreditation Committee on May 2 nd , 2006 as a test laboratory with number LP-311-ID, has to maintain its laboratory performance by participating in a proficiency test. In this activity, the determination of uranium in 2 samples of UO 2 with A1 and A2 codes and other 2 samples of U 3 O 8 with B1 and B2 codes using UV-Vis spectrophotometry was carried out. Colouring method was used by reacting thiocyanate ion with the uranyl ion in acidic solution to develop a stable yellow colour of uranyl thiocyanate complex solution and measured at wavelength of 380 nm. The result gave that concentration of uranium in A1, A2, B1 and B2 samples were 77.95; 75.29; 64.58 and 63.69% respectively. The Z-score value for A samples was - 1.99, meanwhile for B samples the Z score value of between laboratory was −1.29 with intra laboratory was -1,09. It meant that Z-score values for both samples were in good category. From this result, it showed that UV-Vis spectrophotometry is one of the several methods that can be used to determine uranium in UO 2 and U 3 O 8 powder. The Lab. TAR’s proficiency test for determination of uranium in UO 2 and U 3 O 8 gave a good result and it was hoped to support BATAN's program in the nuclear fuel field. (author)

  9. Study on the surface oxidation resistance of uranium metal in the atmosphere of carbon monoxide

    International Nuclear Information System (INIS)

    Wang Xiaolin; Fu Yibei; Xie Renshou

    1999-01-01

    The surface reactions of different layers on uranium metal with carbon monoxide at 25, 80 and 200 degree C are studied by X-ray photoelectron spectroscopy (XPS). The experimental results show that the carbon monoxide is adsorbed on the surface oxide layer of uranium and interacted each other. The content of oxygen in the surface oxide and O/U ratio are decreased with increasing the exposure of carbon monoxide to the surface layer. The effect of reduction on the metal surface is more obviously with a higher temperature and increasing of layer thickness. The investigation indicates the uranium metal has resistance to further oxidation in the atmosphere of carbon monoxide

  10. Influence of carbon monoxide to the surface layer of uranium metal and its oxides

    International Nuclear Information System (INIS)

    Wang Xiaoling; Fu Yibei; Xie Renshou; Huang Ruiliang

    1996-09-01

    The surface structures of uranium metal and triuranium octaoxide (U 3 O 8 ) and the influence of carbon monoxide to the surface layers have been studied by X-ray photoelectron spectroscopy (XPS). After exposure to carbon monoxide, contents of oxygen in the surface oxides of uranium metal and U 3 O 8 are decreased and O/U ratios decrease 7.2%, 8.0% respectively. The investigation indicated the surface layers of uranium metal and its oxides were forbidden to further oxidation in the atmosphere of carbon monoxide. (11 refs., 9 figs., 2 tabs.)

  11. Uranium

    International Nuclear Information System (INIS)

    Cuney, M.; Pagel, M.; Leroy, J.

    1992-01-01

    First, this book presents the physico-chemical properties of Uranium and the consequences which can be deduced from the study of numerous geological process. The authors describe natural distribution of Uranium at different scales and on different supports, and main Uranium minerals. A great place in the book is assigned to description and classification of uranium deposits. The book gives also notions on prospection and exploitation of uranium deposits. Historical aspects of Uranium economical development (Uranium resources, production, supply and demand, operating costs) are given in the last chapter. 7 refs., 17 figs

  12. Determination of size and shape distributions of metal and ceramic powders

    International Nuclear Information System (INIS)

    Jovanovic, DI.

    1961-01-01

    For testing the size and shape distributions of metal and ceramic uranium oxide powders the following method for analysing the grain size of powders were developed and implemented: microscopic analysis and sedimentation method. A gravimetry absorption device was constructed for determining the specific surfaces of powders

  13. The spectrographic analysis of plutonium oxide or mixed plutonium oxide/uranium oxide fuel pellets by the dried residue technique

    International Nuclear Information System (INIS)

    Jarbo, G.J.; Faught, P.; Hildebrandt, B.

    1980-05-01

    An emission spectrographic method for the quantitative determination of metallic impurities in plutonium oxide and mixed plutonium oxide/uranium oxide is described. The fuel is dissolved in nitric acid and the plutonium and/or uranium extracted with tributyl phosphate. A small aliquot of the aqueous residue is dried on a 'mini' pyrolitic graphite plate and excited by high voltage AC spark in an oxygen atmosphere. Spectra are recorded in a region which has been specially selected to record simultaneously lines of boron and cadmium in the 2nd order and all the other elements of interest in the 1st order. Indium is used as an internal standard. The excitation of very small quantities of the uraniumm/plutonium free residue by high voltage spark, together with three separate levels of containment reduce the hazards to personnel and the environment to a minimum with limited effect on sensitivity and accuracy of the results. (auth)

  14. The role of the native oxide shell on the microwave sintering of copper metal powder compacts

    International Nuclear Information System (INIS)

    Mahmoud, Morsi M.; Link, Guido; Thumm, Manfred

    2015-01-01

    Highlights: • Thin oxide native layer had a critical role on microwave sintering of copper. • Explain why microwaves interact with copper powder differently than its bulk. • Abnormal expansion in copper is due to the plastic deformation and crack formation. • In-situ setup gives important insight about the microwave sintering of metals. • Microwave sintering is a promising candidate technology in powder metallurgy. - Abstract: Successful microwave sintering of several metal powders had been reported by many researchers with remarkable improvements in the materials properties and/or in the overall process. However, the concept behind microwave heating of metal powders has not been fully understood till now, as it is well known that bulk metals reflect microwaves. The progress of microwave sintering of copper metal powder compacts was investigated via combining both in-situ electrical resistivity and dilatometry measurements that give important information about microstructural changes with respect to the inter-particle electrical contacts during sintering. The sintering behavior of copper metal powders was depending on the type of the gas used, particle size, the initial green density, the soaking sintering time and the thin oxide layer on the particles surfaces. The thin copper oxide native layer (ceramics) that thermodynamically formed on the particles surfaces under normal handling and ambient environmental conditions had a very critical and important role in the microwave absorption and interaction, the sintering behavior and the microstructural changes. This finding could help to have a fundamental understanding of why MW’s interact with copper metal powder in a different way than its bulk at room temperature, i.e. why a given metal powder could be heated using microwaves while its bulk reflects it

  15. Effect of fabrication conditions on the properties of indium tin oxide powders

    International Nuclear Information System (INIS)

    Xie Wei

    2008-01-01

    This paper reports that indium tin oxide (ITO) crystalline powders are prepared by coprecipitation method. Fabrication conditions mainly as sintering temperature and Sn doping content are correlated with the phase, microstructure, infrared emissivity in and powder resistivity of indium tin oxides by means of x-ray diffraction, Fourier transform infrared, and transmission electron microscope. The optimum sintering temperature of 1350°C and Sn doping content 6∼8wt% are determined. The application of ITO in the military camouflage field is proposed. (condensed matter: electronic structure, electrical, magnetic, and optical properties)

  16. Oxidation state analyses of uranium with emphasis on chemical speciation in geological media

    International Nuclear Information System (INIS)

    Ervanne, H.

    2004-01-01

    This thesis focuses on chemical methods suitable for the determination of uranium redox species in geological materials. Nd-coprecipitation method was studied for the determination of uranium oxidation states in ground waters. This method is ideally suited for the separation of uranium oxidation states in the field, which means that problems associated with the instability of U(IV) during transport are avoided. An alternative method, such as ion exchange, is recommended for the analysis of high saline and calcium- and iron-rich ground waters. U(IV)/Utot was 2.8-7.2% in ground waters in oxidizing conditions and 60-93% in anoxic conditions. From thermodynamic model calculations applied to results from anoxic ground waters it was concluded that uranium can act as redox buffer in granitic ground waters. An ion exchange method was developed for the analysis of uranium oxidation states in different solid materials of geological origin. These included uranium minerals, uraniumbearing minerals, fracture coatings and bulk rock. U(IV)/Utot was 50-70% in uraninites, 5.8-8.7% in secondary uranium minerals, 15-49% in different fracture coatings and 64- 77% in samples from deep bedrock. In the uranium-bearing minerals, U(IV)/Utot was 33-43% (allanites), 5.9% (fergusonite) and 93% (monazite). Although the ion exchange method gave reliable results, there is a risk for the conversion of uranium oxidation states during analysis of heterogeneous samples due to the redox reactions that take place in the presence of some iron compounds. This risk was investigated in a study of several common iron-bearing minerals. The risk for conversion of uranium oxidation states can be screened by sample selection and minimized with use of a redox buffer compound such as polyacrylic acid (PAA). In studies of several carboxylic acids, PAA was found to be the most suitable for extending the application of the method. The stability of uranium oxidation states during analysis and the selectivity

  17. Metallographic preparation of sintered oxides, carbides and nitrides of uranium and plutonium

    International Nuclear Information System (INIS)

    Martin, A.; Arles, L.

    1967-12-01

    We describe the methods of polishing, attack and coloring used at the section of plutonium base ceramics studies. These methods have stood the test of experience on the uranium and plutonium carbides, nitrides and carbonitrides as well on the mixed uranium and plutonium oxides. These methods have been particularly adapted to fit to the low dense and sintered samples [fr

  18. Procedure for the conversion of a metal oxide powder to a fine grained ceramic material

    International Nuclear Information System (INIS)

    Ferrell, L.J.

    1978-01-01

    A procedure for sintering metal oxides is described which gives a product with significantly smaller grain size and better grain size distribution than previous processes. The procedure is presented as applied to aluminium oxide, but it is also stated to be applicable to uranium dioxide. A pellet density of within 1/2 percent of the theoretical maximum can be obtained. No grinding or surface treatment of the pellets is necessary. (JIW)

  19. Study on the surface oxidation of uranium in different gaseous atmospheres

    International Nuclear Information System (INIS)

    Wang Xiaoling; Fu Yibei; Xie Renshou

    1996-03-01

    The studying for the surface oxidation of uranium and oxide by X-ray photoelectron spectroscopy (XPS), Auger electron spectroscopy (AES), secondary ion mass spectroscopy (SIMS), and the surface oxidation of uranium in different gaseous atmospheres such as O 2 , H 2 , CO, CO 2 , H 2 O(v) and air were reviewed. The surface oxidation of uranium is greatly influenced by a number of parameters including atmospheric temperature, pressure, diffusion of adsorbed gas atoms through the oxide layer, surface and interface chemical component, and defect structure and electron nature of the oxide layer. The initial oxidation mechanism and kinetics have been discussed. Suggestions for future work have also been presented. (32 refs., 7 figs., 5 tabs.)

  20. Method of manufacturing mixed stock powders for nuclear fuel elements

    International Nuclear Information System (INIS)

    Hirayama, Satoshi.

    1980-01-01

    Purpose: To alleviate the limit of the present reactor operating conditions by uniformly mixing an additive to the main content as an uranium dioxide or mixture of the uranium dioxide with plutonium dioxide. Method: A mixed stock powder is obtained by adding an additive of at least two of aluminium oxide, beryllium oxide, calcium oxide, magnesium oxide, silicon oxide, sodium oxide, potassium oxide, phosphorus oxide, titanium oxide and iron oxide to suspension having ammonia water as dispersion medium to start the deposition of precipitation at a step of precipitating ammonium diuranate or plutionium hydroxide of a main content of uranium dioxide or mixture of uranium dioxide and plutonium dioxide and deposited precipitate is calcinated and reduced. (Yoshihara, H.)

  1. Advanced oxide powders processing based on cascade plasma

    International Nuclear Information System (INIS)

    Solonenko, O P; Smirnov, A V

    2014-01-01

    Analysis of the potential advantages offered to thermal spraying and powder processing by the implementation of plasma torches with inter-electrode insert (IEI) or, in other words, cascade plasma torches (CPTs) is presented. The paper provides evidence that the modular designed single cathode CPT helps eliminate the following major disadvantages of conventional plasma torches: plasma parameters drifting, 1-5 kHz pulsing of plasma flow, as well as excessive erosion of electrodes. More stable plasma results in higher quality, homogeneity and reproducibility of plasma sprayed coatings and powders treated. In addition, CPT offers an extremely wide operating window, which allows better control of plasma parameters, particle dwell time and, consequently, particle temperature and velocity within a wide range by generating high enthalpy quasi-laminar plasmas, medium enthalpy transient plasmas, as well as relatively low enthalpy turbulent plasmas. Stable operation, flexibility with plasma gases as well as wide operating window of CPT should help significantly improve the existing plasma spraying processes and coatings, and also help develop new advanced technologies

  2. Organic matter in uranium concentration during ancient bed oxidation of carboniferons sediments

    International Nuclear Information System (INIS)

    Kruglova, V.G.; Uspenskij, V.A.; Dement'ev, P.K.; Kochenov, A.V.

    1984-01-01

    Changes in the organic matter accompanying the process of epigenetic ore formation are studied using the example of a deposit localized in carboniferous molasse strata of the Cretaceous period. Peculiarities of the organic matter as the main mineralization agent are studied by a complex of physical and themical methods. A distinct relationship between the uranium concentration and the degree of organic matter oxigenation is a most characteristic feature of the ore localization, however, there is no direct correlation between the contents of uranium and organic matter in ores. Uranium minerallzation was accumulated during infiltration of acid uraniferous.waters into grey stratum in the process of the bed oxidation zone formation oxidizing. Brown coal matter possessing a maximum adsorbability, as compared to other sedimentary rocks, apprared to be the uranium precipitator. The adsorption was accompanie by the formation of proper uranium minerals (coffinite, pitchblende) due to uranium reduction by oxidizing organic matter. Thus, the oxidative epigenesis was an are-forming process with the uranium concentration on organic matter proportionally to oxidation of the latter

  3. Development of Oxide Dispersion Strengthened (ODS) Ferritic Steel Through Powder Forging

    Science.gov (United States)

    Kumar, Deepak; Prakash, Ujjwal; Dabhade, Vikram V.; Laha, K.; Sakthivel, T.

    2017-04-01

    Oxide dispersion strengthened (ODS) ferritic steels are candidates for cladding tubes in fast breeder nuclear reactors. In this study, an 18%Cr ODS ferritic steel was prepared through powder forging route. Elemental powders with a nominal composition of Fe-18Cr-2 W-0.2Ti (composition in wt.%) with 0 and 0.35% yttria were prepared by mechanical alloying in a Simoloyer attritor under argon atmosphere. The alloyed powders were heated in a mild steel can to 1473 K under flowing hydrogen atmosphere. The can was then hot forged. Steps of sealing, degassing and evacuation are eliminated by using powder forging. Heating ODS powder in hydrogen atmosphere ensures good bonding between alloy powders. A dense ODS alloy with an attractive combination of strength and ductility was obtained after re-forging. On testing at 973 K, a loss in ductility was observed in yttria-containing alloy. The strength and ductility increased with increase in strain rate at 973 K. Reasons for this are discussed. The ODS alloy exhibited a recrystallized microstructure which is difficult to achieve by extrusion. No prior particle boundaries were observed after forging. The forged compacts exhibited isotropic mechanical properties. It is suggested that powder forging may offer several advantages over the traditional extrusion/HIP routes for fabrication of ODS alloys.

  4. Criticality of moderated and undermoderated low-enriched uranium oxide systems

    International Nuclear Information System (INIS)

    Goebel, G.R.

    1980-06-01

    Uranium oxide was enriched to 4.46 wt % 235 U compacted to a density of 4.68 g/cm 3 . The uranium oxide was packed into cubical aluminum cans and water added to the oxide until an H/U atomic ratio of 0.77 was achieved. A 5 x 5 x 5 array of uranium oxide cans for the experiments were used when no plastic moderator material was placed between cans. High enriched uranium drivers were used to achieve criticality. Criticality was achieved for smaller arrays without a driver when 24.5 mm plastic moderator material was placed between the cans. Twelve critical experiments are reported, six in each reflector

  5. Chlorination of antimony and its volatilization treatment of waste antimony-uranium composite oxide catalyst

    International Nuclear Information System (INIS)

    Sawada, K.; Enokida, Y.

    2011-01-01

    For the waste antimony-uranium composite oxide catalyst, the chlorination of antimony and its volatilization treatment were proposed, and evaluated using hydrogen chloride gas at 873-1173 K. During the treatment, the weight loss of the composite oxide sample, which resulted from the volatilization of antimony, was confirmed. An X-ray diffraction analysis showed that uranium oxide, U 3 O 8 , was formed during the reaction. After the treatment at 1173 K for 1 h, almost all the uranium contained in the waste catalyst was dissolved by a 3 M nitric acid solution at 353 K within 10 min, although that of the non-treated catalyst was less than 0.1%. It was found that the chlorination and volatilization treatment was effective to separate antimony from the composite oxide catalyst and change uranium into its removable form. (orig.)

  6. Physical characterization of steel and stainless steel metal powders

    International Nuclear Information System (INIS)

    Lavilla, A.O.; Lucchesi, C.G.; Sandin, O.O.

    1991-01-01

    A methodology has been developed for the physical characterization of steel powders (obtained by atomization) for later sintering and for the construction of porous sheets and filtrating tubes, capable of operating at temperatures between 600 deg C and 800 deg C in corrosive atmospheres. This methodology was based on the equipment and methods used for the physical characterization of uranium oxide powders. (Author) [es

  7. Influence of pH on the adsorption of uranium ions by oxidized activated carbon and chitosan

    International Nuclear Information System (INIS)

    Park, G.I.; Park, H.S.; Woo, S.I.

    1999-01-01

    The adsorption characteristics of uranyl ions on surface-oxidized carbon were compared with those of powdered chitosan over a wide pH range. In particular, an extensive analysis was made on solution pH variation during the adsorption process or after adsorption equilibrium. Uranium adsorption on the two adsorbents was revealed to be strongly dependent on the initial pH of the solution. A quantitative comparison of the adsorption capacities of the two adsorbents was made, based on the isotherm data obtained at initial pH 3, 4, and 5. In order to analyze the adsorption kinetics incorporated with pH effects, batch experiments at various initial pH values were carried out, and solution pH profiles with the adsorption time were also evaluated. The breakthrough behavior in a column packed with oxidized carbon was also characterized with respect to the variation of effluent pH. Based on these experimental results, the practical applicability of oxidized carbon for uranium removal from acidic radioactive liquid waste was suggested

  8. Nuclear energy - Determination of chlorine and fluorine in uranium dioxide powder and sintered pellets

    International Nuclear Information System (INIS)

    2008-01-01

    This International Standard describes a method for determining the chlorine and fluorine concentrations in uranium dioxide and in sintered fuel pellets by pyrohydrolysis of samples, followed either by liquid ion-exchange chromatography or by selective electrode measurement of chlorine and fluorine ions. Many ion-exchange chromatography systems and ion-selective electrode measurement systems are available

  9. Fabrication of powder from ductile uranium alloys for use as nuclear dispersion

    International Nuclear Information System (INIS)

    Durazzo, M.; Leal Neto, R.M.; Rocha, C.J.; Urano de Carvalho, E.; Riella, H.G.

    2014-01-01

    This work forms part of the studies presently ongoing at IPEN investigating the feasibility of powdering ductile U-10wt%Mo alloy by hydriding-milling-de-hydriding of the gamma phase (HMD). Hydriding was conducted at room temperature in a Sievert apparatus following heat treatment activation. Hydrided pieces were fragile enough to be hand milled to the desired particle size range. Hydrogen was removed by heating the samples under high vacuum. X-ray diffraction analysis of the hydrided material showed an amorphous-like pattern that is completely reversed following de-hydriding. The hydrogen content of the hydrided samples corresponds to a trihydride, i.e. (U,Mo)H 3 . SEM analysis of HMD powder particles revealed equi-axial powder particles together with some plate-like particles. A hypothesis for the amorphous hydride phase formation is suggested. (authors)

  10. An Investigation on the Persistence of Uranium Hydride during Storage of Simulant Nuclear Waste Packages

    OpenAIRE

    Stitt , C. A.; Harker , N. J.; Hallam , K. R.; Paraskevoulakos , C.; Banos , A.; Rennie , S.; Jowsey , J.; Scott , T. B.

    2015-01-01

    International audience; Synchrotron X-rays have been used to study the oxidation of uranium and uranium hydride when encapsulated in grout and stored in de-ionised water for 10 months. Periodic synchrotron X-ray tomography and X-ray powder diffraction have allowed measurement and identification of the arising corrosion products and the rates of corrosion. The oxidation rates of the uranium metal and uranium hydride were slower than empirically derived rates previously reported for each reacta...

  11. Characterizing uranium oxide reference particles for isotopic abundances and uranium mass by single particle isotope dilution mass spectrometry

    International Nuclear Information System (INIS)

    Kraiem, M.; Richter, S.; Erdmann, N.; Kühn, H.; Hedberg, M.; Aregbe, Y.

    2012-01-01

    Highlights: ► A method to quantify the U mass in single micron particles by ID-TIMS was developed. ► Well-characterized monodisperse U-oxide particles produced by an aerosol generator were used. ► A linear correlation between the mass of U and the volume of particle(s) was found. ► The method developed is suitable for determining the amount of U in a particulate reference material. - Abstract: Uranium and plutonium particulate test materials are becoming increasingly important as the reliability of measurement results has to be demonstrated to regulatory bodies responsible for maintaining effective nuclear safeguards. In order to address this issue, the Institute for Reference Materials and Measurements (IRMM) in collaboration with the Institute for Transuranium Elements (ITU) has initiated a study to investigate the feasibility of preparing and characterizing a uranium particle reference material for nuclear safeguards, which is finally certified for isotopic abundances and for the uranium mass per particle. Such control particles are specifically required to evaluate responses of instruments based on mass spectrometric detection (e.g. SIMS, TIMS, LA-ICPMS) and to help ensuring the reliability and comparability of measurement results worldwide. In this paper, a methodology is described which allows quantifying the uranium mass in single micron particles by isotope dilution thermal ionization mass spectrometry (ID-TIMS). This methodology is characterized by substantial improvements recently achieved at IRMM in terms of sensitivity and measurement accuracy in the field of uranium particle analysis by TIMS. The use of monodisperse uranium oxide particles prepared using an aerosol generation technique developed at ITU, which is capable of producing particles of well-characterized size and isotopic composition was exploited. The evidence of a straightforward correlation between the particle volume and the mass of uranium was demonstrated in this study

  12. Evidence for Single Metal Two Electron Oxidative Addition and Reductive Elimination at Uranium

    OpenAIRE

    Gardner, Benedict M; Kefalidis, Christos E; Lu, Erli; Patel, Dipti; Mcinnes, Eric; Tuna, Floriana; Wooles, Ashley; Maron, Laurent; Liddle, Stephen

    2017-01-01

    Reversible single-metal two-electron oxidative addition and reductive elimination are common fundamental reactions for transition metals that underpin major catalytic transformations. However, these reactions have never been observed together in the f-block because these metals exhibit irreversible one- or multi-electron oxidation or reduction reactions. Here, we report that azobenzene oxidises sterically and electronically unsaturated uranium(III) complexes to afford a uranium(V)-imido compl...

  13. An unusual temperature dependence in the oxidation of oxycarbide layers on uranium

    Science.gov (United States)

    Ellis, Walton P.

    1981-09-01

    An anomalous temperature dependence has been observed for the oxidation kinetics of outermost oxycarbide layers on polycrystalline uranium metal. Normally, oxidation or corrosion reactions are expected to proceed more rapidly as the temperature is elevated. Thus, it came as a surprise when we observed that the removal of the outermost atomic layers of carbon from uranium oxycarbide by O 2 reproducibly proceeds at a much faster rate at 25°C than at 280°C.

  14. Simple process to fabricate nitride alloy powders

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Kim, Dong-Joo; Kim, Keon Sik; Rhee, Young Woo; Oh, Jang-Soo; Kim, Jong Hun; Koo, Yang Hyun

    2013-01-01

    Uranium mono-nitride (UN) is considered as a fuel material [1] for accident-tolerant fuel to compensate for the loss of fissile fuel material caused by adopting a thickened cladding such as SiC composites. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. Among them, a direct nitriding process of metal is more attractive because it has advantages in the mass production of high-purity powders and the reusing of expensive 15 N 2 gas. However, since metal uranium is usually fabricated in the form of bulk ingots, it has a drawback in the fabrication of fine powders. The Korea Atomic Energy Research Institute (KAERI) has a centrifugal atomisation technique to fabricate uranium and uranium alloy powders. In this study, a simple reaction method was tested to fabricate nitride fuel powders directly from uranium metal alloy powders. Spherical powder and flake of uranium metal alloys were fabricated using a centrifugal atomisation method. The nitride powders were obtained by thermal treating the metal particles under nitrogen containing gas. The phase and morphology evolutions of powders were investigated during the nitriding process. A phase analysis of nitride powders was also part of the present work. KAERI has developed the centrifugal rotating disk atomisation process to fabricate spherical uranium metal alloy powders which are used as advanced fuel materials for research reactors. The rotating disk atomisation system involves the tasks of melting, atomising, and collecting. A nozzle in the bottom of melting crucible introduces melt at the center of a spinning disk. The centrifugal force carries the melt to the edge of the disk and throws the melt off the edge. Size and shape of droplets can be controlled by changing the nozzle size, the disk diameter and disk speed independently or simultaneously. By adjusting the processing parameters of the centrifugal atomiser, a spherical and flake shape

  15. Synthesis and characterization of nanocomposite powders of calcium phosphate/titanium oxide for biomedical applications

    Energy Technology Data Exchange (ETDEWEB)

    Delima, S.A.; Camargo, N.H.A.; Souza, J.C.P.; Gemelli, E., E-mail: sarahamindelima@hotmail.com, E-mail: dem2nhac@joinville.udesc.br, E-mail: souzajulio@joinville.udesc.br, E-mail: gemelli@joinville.udesc.br [Universidade do Estado de Santa Catarina (UDESC), Joinville, SC (Brazil). Centro de Ciencias Tecnologicas

    2009-07-01

    The nanostructured bioceramics of calcium phosphate are current themes of research and they are becoming important as bone matrix in regeneration of tissues in orthopedic and dental applications. Nanocomposite powders of calcium phosphate, reinforced with nanometric particles of titanium oxide, silica oxide and alumina oxid ealpha, are being widely studied because they offer new microstructures, nanostructures and interconnected microporosity with high superficial area of micropores that contribute to osteointegration and osteoinduction processes. This study is about the synthesis of nanocomposites powders of calcium phosphate reinforced with 1%, 2%, 3% and 5% in volume of titanium oxide and its characterization through the techniques of X-Ray Diffraction (XRD), Scanning Electron Microscopy (SEM), Differential Thermal Analysis (DTA), Thermogravimetry (TG) and Dilatometry. (author)

  16. Synthesis and characterization of nanocomposite powders of calcium phosphate/titanium oxide for biomedical applications

    International Nuclear Information System (INIS)

    Delima, S.A.; Camargo, N.H.A.; Souza, J.C.P.; Gemelli, E.

    2009-01-01

    The nanostructured bioceramics of calcium phosphate are current themes of research and they are becoming important as bone matrix in regeneration of tissues in orthopedic and dental applications. Nanocomposite powders of calcium phosphate, reinforced with nanometric particles of titanium oxide, silica oxide and alumina oxid ealpha, are being widely studied because they offer new microstructures, nanostructures and interconnected microporosity with high superficial area of micropores that contribute to osteointegration and osteoinduction processes. This study is about the synthesis of nanocomposites powders of calcium phosphate reinforced with 1%, 2%, 3% and 5% in volume of titanium oxide and its characterization through the techniques of X-Ray Diffraction (XRD), Scanning Electron Microscopy (SEM), Differential Thermal Analysis (DTA), Thermogravimetry (TG) and Dilatometry. (author)

  17. Interstitial pressure dependence of the thermal conductivity of some rare earth oxide powders

    International Nuclear Information System (INIS)

    Pradeep, P.

    1997-01-01

    Thermal transport properties of powdered materials depend upon interstitial gas pressure. The present study reports the experimental results for the effective thermal conductivity of three rare earth oxide powders viz. yttrium oxide, samarium oxide, and gadolinium oxide, at various interstitial pressures by using transient plane source (TPS) method. A theoretical model is also proposed for the interpretation of the variation of the effective thermal conductivity with interstitial gas pressure. Its validity is found to be good in low pressure range of 45 mm Hg to normal pressure when compared with the experimental results. Also an attempt has been made to calculate the variation of thermal conductivity with interstitial pressure in the high pressure range up to 2 kbar using the proposed model. (author)

  18. First-principles study on oxidation effects in uranium oxides and high-pressure high-temperature behavior of point defects in uranium dioxide

    Science.gov (United States)

    Geng, Hua Y.; Song, Hong X.; Jin, K.; Xiang, S. K.; Wu, Q.

    2011-11-01

    Formation Gibbs free energy of point defects and oxygen clusters in uranium dioxide at high-pressure high-temperature conditions are calculated from first principles, using the LSDA+U approach for the electronic structure and the Debye model for the lattice vibrations. The phonon contribution on Frenkel pairs is found to be notable, whereas it is negligible for the Schottky defect. Hydrostatic compression changes the formation energies drastically, making defect concentrations depend more sensitively on pressure. Calculations show that, if no oxygen clusters are considered, uranium vacancy becomes predominant in overstoichiometric UO2 with the aid of the contribution from lattice vibrations, while compression favors oxygen defects and suppresses uranium vacancy greatly. At ambient pressure, however, the experimental observation of predominant oxygen defects in this regime can be reproduced only in a form of cuboctahedral clusters, underlining the importance of defect clustering in UO2+x. Making use of the point defect model, an equation of state for nonstoichiometric oxides is established, which is then applied to describe the shock Hugoniot of UO2+x. Furthermore, the oxidization and compression behavior of uranium monoxide, triuranium octoxide, uranium trioxide, and a series of defective UO2 at 0 K are investigated. The evolution of mechanical properties and electronic structures with an increase of the oxidation degree are analyzed, revealing the transition of the ground state of uranium oxides from metallic to Mott insulator and then to charge-transfer insulator due to the interplay of strongly correlated effects of 5f orbitals and the shift of electrons from uranium to oxygen atoms.

  19. Purifications of calcium carbonate and molybdenum oxide powders for neutrinoless double beta decay experiment, AMoRE

    International Nuclear Information System (INIS)

    Park, HyangKyu

    2015-01-01

    The AMoRE (Advanced Mo based Rare process Experiment) collaboration is going to use calcium molybdate crystals to search for neutrinoless double beta decay of 100 Mo isotope. In order to make the crystal, we use calcium carbonate and molybdenum oxide powders as raw materials. Therefore it is highly necessary to reduce potential sources for radioactive backgrounds such as U and Th in the powders. In this talk, we will present our studies for purification of calcium carbonate and molybdenum oxide powders

  20. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The article briefly discusses the Australian government policy and the attitude of political party factions towards the mining and exporting of the uranium resources in Australia. Australia has a third of the Western World's low-cost uranium resources

  1. Uranium

    International Nuclear Information System (INIS)

    Poty, B.; Cuney, M.; Bruneton, P.; Virlogeux, D.; Capus, G.

    2010-01-01

    With the worldwide revival of nuclear energy comes the question of uranium reserves. For more than 20 years, nuclear energy has been neglected and uranium prospecting has been practically abandoned. Therefore, present day production covers only 70% of needs and stocks are decreasing. Production is to double by 2030 which represents a huge industrial challenge. The FBR-type reactors technology, which allows to consume the whole uranium content of the fuel, is developing in several countries and will ensure the long-term development of nuclear fission. However, the implementation of these reactors (the generation 4) will be progressive during the second half of the 21. century. For this reason an active search for uranium ores will be necessary during the whole 21. century to ensure the fueling of light water reactors which are huge uranium consumers. This dossier covers all the aspects of natural uranium production: mineralogy, geochemistry, types of deposits, world distribution of deposits with a particular attention given to French deposits, the exploitation of which is abandoned today. Finally, exploitation, ore processing and the economical aspects are presented. Contents: 1 - the uranium element and its minerals: from uranium discovery to its industrial utilization, the main uranium minerals (minerals with tetravalent uranium, minerals with hexavalent uranium); 2 - uranium in the Earth's crust and its geochemical properties: distribution (in sedimentary rocks, in magmatic rocks, in metamorphic rocks, in soils and vegetation), geochemistry (uranium solubility and valence in magmas, uranium speciation in aqueous solution, solubility of the main uranium minerals in aqueous solution, uranium mobilization and precipitation); 3 - geology of the main types of uranium deposits: economical criteria for a deposit, structural diversity of deposits, classification, world distribution of deposits, distribution of deposits with time, superficial deposits, uranium

  2. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  3. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  4. The oxidation behaviour of uranium in air at 348-765 K

    International Nuclear Information System (INIS)

    Bennett, M.J.; Price, J.B.

    1981-01-01

    The oxidation behaviour of adjusted uranium has been examined in air, at atmospheric pressure, at 348-765 K. Particular emphasis has been directed to the role of swelling resulting from prior irradiation of the metal to a burn-up of 5600-9100 MWd/t and an addition of 2.5 x 10 4 vpm water vapour to the air. Pre-irradiation of uranium enhanced its attack by air at 348-523 K, the enhancement increasing progressively with percentage swelling. This effect resulted primarily from the break-up of the uranium surface during oxidation with the generation of a greater reaction surface area and was independent of the temperature of oxidation in dry air and also above 423 K in wet air. At lower temperatures, however, the water vapour addition increased the effective reaction rate, possibly by the transistory involvement of uranium hybride. The influence of the water vapour increased with swelling of the irradiated uranium and was greater than that exerted on the oxidation of unirradiated uranium at comparable temperatures. With increasing temperature above 623 K, swelling had a progressively decreasing influence upon the attack of irradiated uranium in both environments. (orig.)

  5. Method for continuous synthesis of metal oxide powders

    Science.gov (United States)

    Berry, David A.; Haynes, Daniel J.; Shekhawat, Dushyant; Smith, Mark W.

    2015-09-08

    A method for the rapid and continuous production of crystalline mixed-metal oxides from a precursor solution comprised of a polymerizing agent, chelated metal ions, and a solvent. The method discharges solution droplets of less than 500 .mu.m diameter using an atomizing or spray-type process into a reactor having multiple temperature zones. Rapid evaporation occurs in a first zone, followed by mixed-metal organic foam formation in a second zone, followed by amorphous and partially crystalline oxide precursor formation in a third zone, followed by formation of the substantially crystalline mixed-metal oxide in a fourth zone. The method operates in a continuous rather than batch manner and the use of small droplets as the starting material for the temperature-based process allows relatively high temperature processing. In a particular embodiment, the first zone operates at 100-300.degree. C., the second zone operates at 300-700.degree. C., and the third operates at 700-1000.degree. C., and fourth zone operates at at least 700.degree. C. The resulting crystalline mixed-metal oxides display a high degree of crystallinity and sphericity with typical diameters on the order of 50 .mu.m or less.

  6. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

    A discussion is given of uranium as an energy source in The Australian economy. Figures and predictions are presented on the world supply-demand position and also figures are given on the added value that can be achieved by the processing of uranium. Conclusions are drawn about Australia's future policy with regard to uranium (R.L.)

  7. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  8. Study on the influence of carbon monoxide to the surface oxide layer of uranium metal

    International Nuclear Information System (INIS)

    Wang Xiaolin; Duan Rongliang; Fu Yibei; Xie Renshou; Zuo Changming; Zhao Chunpei; Chen Hong

    1997-01-01

    The influence of carbon monoxide to the surface oxide layer of uranium metal has been studied by X-ray photoelectron spectroscopy (XPS) and gas chromatography (GC). Carbon monoxide adsorption on the oxide layer resulted in U4f peak shifting to the lower binding energy. The content of oxygen in the oxide is decreased and the atomic ratio (O/U) is decreased by 7.2%. The amount of carbon dioxide in the atmosphere after the surface reaction is increased by 11.0%. The investigation indicates that the surface layer can prevent the further oxidation uranium metal in the atmosphere of carbon monoxide

  9. Method for analysing radium in powder samples and its application to uranium prospecting

    International Nuclear Information System (INIS)

    Gong Xinxi; Hu Minzhi.

    1987-01-01

    The decayed daughter of Rn released from the power sample (soil) in a sealed bottle were collected on a piece of copper and the radium in the sample can be measured by counting α-particles with an Alphameter for uranium prospection, thus it is called the radium method. This method has many advantages, such as high sensitivity (the lowest limit of detection for radium sample per gram is 2.7 x 10 -15 g), high efficiency, low cost and easy to use. On the basis of measuring more than 700 samples taken along 20 sections in 8 deposits, the results show that the radium method is better than γ-measurement and equal to 210 Po method for the capability to descover anomalies. The author also summarizes the anomaly intensities of radium method, 210 Po method and γ-measurement respectively at the surface with deep blind ores, with or without surficial mineralization, and the figures of their profiles and the variation of Ra/ 210 Po ratios. According to the above-mentioned distinguishing features, the uranium mineralization located in deep and/or shallow parts can be distinguishd. The combined application of radium, 210 Po and γ-measurement methods may be regarded as one of the important methods used for anomaly assessment. Based on the experiments of the radium measurements with 771 stream sediments samples in an area of 100 km 2 , it is demonstrated that the radium mehtod can be used in the stages of uranium reconnaissance and prospecting

  10. The evaluation, design and implementation of an automated storage and retrieval system for uranium trioxide powder (UO3) at Sellafield

    International Nuclear Information System (INIS)

    Fitt, C.R.; Mather, K.

    1993-01-01

    The paper initially sets out the methods used to evaluate the requirements for an automated system to store and retrieve drums of radioactive Uranium Trioxide (UO3) power arising from the Thermal Oxide Reprocessing Plant (THORP) at Sellafield Cumbria. This is followed by a description of the configuration of storage vaults used and of the development of a Self Guided Vehicle (SGV) to operate remotely within these vaults. The system evolved is based on a combination of well proven mechanical equipment and control techniques and the implementation of the design together with testing and control procedures are described. (author)

  11. Preparation of uranium-based oxide catalysts; Preparation de catalyseurs oxydes a base d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Bressat, R. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1965-07-01

    We have studied the thermal decomposition of uranyl and uranium IV oxalates as a mean of producing uranium dioxide. We have isolated the main intermediate phases of the decompositions and have indexed the lines of their X-ray diffraction patterns. The oxides produced by the decomposition are ill-defined and unstable: they strongly absorb atmospheric oxygen with modification of the composition and, in certain cases, of the structure (pyrophoric oxide). With a view to obtaining stable oxides, we have prepared mixed uranium-thorium oxalates. In order to prepare an oxalate having a homogeneous composition, it is necessary to adopt a well-defined preparation method: the addition of solutions of thorium and uranium IV nitrates to a continually saturated oxalic acid solution. The mixed oxide obtained from the thermal decomposition of an oxalate U{sub x}Th{sub 1-x}(C{sub 2}O{sub 4}){sub 2}, 2 H{sub 2}O at 500 C for 24 hours in a current of oxygen leads to a cubic structure which is well-defined both in the bulk and superficially when x is less than 0.35. Above this atomic concentration of uranium, some uranium moves out of the lattice in the form of UO{sub 3} or U{sub 3}O{sub 8} according to the temperature. The mixed oxide is not stoichiometric,(U{sub x}Th{sub 1-x}O{sub 2+y}) and the average degree of oxidation of the uranium varies with the temperature and partial oxygen pressure. The oxides thus formed have a high surface area. By dissolving the mixed oxalates in a concentrated solution of ammonium oxalate, it is possible to deposit the catalyst on a support, but the differences in the solubilities of the thorium and uranium IV oxalates in the ammonium oxalate make it impossible to prepare double salts formed either of thorium and uranium and of ammonium. (author) [French] Nous avons etudie la decomposition thermique des oxalates d'uranyle et d'uranium IV en vue d'aboutir au dioxide d'uranium. Nous avons pu isoler les principales phases

  12. Melting temperature of uranium - plutonium mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

  13. Melting temperature of uranium - plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Hirosawa, Takashi

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

  14. An autoclave system for uranium oxide dissolution experiments

    International Nuclear Information System (INIS)

    Nykyri, Mikko

    1985-05-01

    According to the decision in principle of the Council of State of Finland the nuclear energy producers must provide preparedness for carrying out the final disposal of spent nuclear fuel in Finland. By the present principal concept the spent fuel will be disposed deep into the granitic bedrock. A parameter needed by risk analysis models is the dissolution rate of the uranium oxide matrix in the fuel pellets. In order to approach conditions prevailing deep in the groundwater, and autoclave system for dissolution experiments was developed at the Technical Research Centre of Finland. The low oxygen content and high pressure at elevated temperatures are simulated in the system. 20 MPa and 100 deg C are the upper operation limits of pressure and temperature. Water can be changed in the experiment autoclave without remarkable pressure and temperature variations. This has been arranged by using three pressure vessels: a supply vessel, a dissolution vessel and a depletion vessel. The extreme vessels serve pressure balancing purposes during water exchange. The water is deoxygenated during a preparation phase in the supply vessel by flushing it with nitrogen gas. Polytetrafluoroethylene is the principal material in contact with the water. A redox electrode couple was developed for potential measurements inside the dissolution vessel. The reference electrode is of Ag/AgCl-type with saturated KC1 electrolyte. A platinum wire operates as a measuring electrode

  15. Oxygen potential of uranium--plutonium oxide as determined by controlled-atmosphere thermogravimetry

    International Nuclear Information System (INIS)

    Swanson, G.C.

    1975-10-01

    The oxygen-to-metal atom ratio, or O/M, of solid solution uranium-plutonium oxide reactor fuel is a measure of the concentration of crystal defects in the oxide which affect many fuel properties, particularly, fuel oxygen potential. Fabrication of a high-temperature oxygen electrode, employing an electro-active tip of oxygen-deficient solid-state electrolyte, intended to confirm gaseous oxygen potentials is described. Uranium oxide and plutonium oxide O/M reference materials were prepared by in situ oxidation of high purity metals in the thermobalance. A solid solution uranium-plutonium oxide O/M reference material was prepared by alloying the uranium and plutonium metals in a yttrium oxide crucible at 1200 0 C and oxidizing with moist He at 250 0 C. The individual and solid solution oxides were isothermally equilibrated with controlled oxygen potentials between 800 and 1300 0 C and the equilibrated O/M ratios calculated with corrections for impurities and buoyancy effects. Use of a reference oxygen potential of -100 kcal/mol to produce an O/M of 2.000 is confirmed by these results. However, because of the lengthy equilibration times required for all oxides, use of the O/M reference materials rather than a reference oxygen potential is recommended for O/M analysis methods calibrations. (auth)

  16. Study of the oxidation state of arsenic and uranium in individual particles from uranium mine tailings, Hungary

    International Nuclear Information System (INIS)

    Alsecz, A.; Osan, J.; Palfalvi, J.; Torok, Sz.; Sajo, I.; Mathe, Z.; Simon, R.; Falkenberg, G.

    2007-01-01

    Uranium ore mining and milling have been terminated in the Mecsek Mountains (southwest Hungary) in 1997. Mine tailings ponds are located between two important water bases, which are resources of the drinking water of the city of Pecs and the neighbouring villages. The average U concentration of the tailings material is 71.73 μg/g, but it is inhomogeneous. Some microscopic particles contain orders of magnitude more U than the rest of the tailings material. Other potentially toxic elements are As and Pb of which chemical state is important to estimate mobility, because in mobile form they can risk the water basis and the public health. Individual U-rich particles were selected with solid state nuclear track detector (SSNTD) and after localisation the particles were investigated by synchrotron radiation based microanalytical techniques. The distribution of elements over the particles was studied by micro beam X-ray fluorescence (μ-XRF) and the oxidation state of uranium and arsenic was determined by micro X-ray absorption near edge structure (μ-XANES) spectroscopy. Some of the measured U-rich particles were chosen for studying the heterogeneity with μ-XRF tomography. Arsenic was present mainly in As(V) and uranium in U(VI) form in the original uranium ore particles, but in the mine tailings samples uranium was present mainly in the less mobile U(IV) form. Correlation was found between the oxidation state of As and U in the same analyzed particles. These results suggest that dissolution of uranium is not expected in short term period. (authors)

  17. Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour.

    Science.gov (United States)

    Martin, T L; Coe, C; Bagot, P A J; Morrall, P; Smith, G D W; Scott, T; Moody, M P

    2016-07-12

    Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (~5 nm) interfacial layer of uranium hydride was observed at the oxide-metal interface. Exposure to deuterated water vapour produced an equivalent deuteride signal at the metal-oxide interface, confirming the hydride as originating via the water vapour oxidation mechanism. Hydroxide ions were detected uniformly throughout the oxide, yet showed reduced prominence at the metal interface. These results support a proposed mechanism for the oxidation of uranium in water vapour environments where the transport of hydroxyl species and the formation of hydride are key to understanding the observed behaviour.

  18. Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour

    Science.gov (United States)

    Martin, T. L.; Coe, C.; Bagot, P. A. J.; Morrall, P.; Smith, G. D. W.; Scott, T.; Moody, M. P.

    2016-07-01

    Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (~5 nm) interfacial layer of uranium hydride was observed at the oxide-metal interface. Exposure to deuterated water vapour produced an equivalent deuteride signal at the metal-oxide interface, confirming the hydride as originating via the water vapour oxidation mechanism. Hydroxide ions were detected uniformly throughout the oxide, yet showed reduced prominence at the metal interface. These results support a proposed mechanism for the oxidation of uranium in water vapour environments where the transport of hydroxyl species and the formation of hydride are key to understanding the observed behaviour.

  19. High resolution Transmission Electron Microscopy characterization of a milled oxide dispersion strengthened steel powder

    Energy Technology Data Exchange (ETDEWEB)

    Loyer-Prost, M., E-mail: marie.loyer-prost@cea.fr [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Merot, J.-S. [Laboratoire d’Etudes des Microstructures – UMR 104, CNRS/ONERA, BP72-29, Avenue de la Division Leclerc, 92 322, Châtillon (France); Ribis, J. [DEN-Service de Recherches de Métallurgie Appliquée, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Le Bouar, Y. [Laboratoire d’Etudes des Microstructures – UMR 104, CNRS/ONERA, BP72-29, Avenue de la Division Leclerc, 92 322, Châtillon (France); Chaffron, L. [DEN-Service de Recherches de Métallurgie Appliquée, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Legendre, F. [DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2016-10-15

    Oxide Dispersion Strengthened (ODS) steels are promising materials for generation IV fuel claddings as their dense nano-oxide dispersion provides good creep and irradiation resistance. Even if they have been studied for years, the formation mechanism of these nano-oxides is still unclear. Here we report for the first time a High Resolution Transmission Electron Microscopy and Energy Filtered Transmission Electron Microscopy characterization of an ODS milled powder. It provides clear evidence of the presence of small crystalline nanoclusters (NCs) enriched in titanium directly after milling. Small NCs (<5 nm) have a crystalline structure and seem partly coherent with the matrix. They have an interplanar spacing close to the (011) {sub bcc} iron structure. They coexist with larger crystalline spherical precipitates of 15–20 nm in size. Their crystalline structure may be metastable as they are not consistent with any Y-Ti-O or Ti-O structure. Such detailed observations in the as-milled grain powder confirm a mechanism of Y, Ti, O dissolution in the ferritic matrix followed by a NC precipitation during the mechanical alloying process of ODS materials. - Highlights: • We observed an ODS ball-milled powder by high resolution transmission microscopy. • The ODS ball-milled powder exhibits a lamellar microstructure. • Small crystalline nanoclusters were detected in the milled ODS powder. • The nanoclusters in the ODS milled powder are enriched in titanium. • Larger NCs of 15–20 nm in size are, at least, partly coherent with the matrix.

  20. Distribution characteristics of interlayer oxidation zone and its relationship with sedimentary facies and uranium mineralization in QJD uranium deposit, Songliao basin, NE China

    International Nuclear Information System (INIS)

    Chen Xiaolin; Xiang Weidong; Li Tiangang; Fang Xiheng; Xia Yuliang; Pang Yaqing; Zheng Jiwei; Zhang Mingyu; Zhang Zegui; Tian Shifeng

    2006-01-01

    QJD uranium deposit is a sandstone-type uranium deposit with the special shape of the interlayer oxidation zone. After studying the palaeoclimate condition and sedimentation of host sandstones, contrasting drilling cross sections, this paper suggests that primary colors of host sandstones are mostly in gray, red and yellow colors of sandstones are the result of oxidation alteration. According to the positions of drill holes with red and yellow alteration sandstones, the plane distribution of oxidation roll fronts of the Upper and Lower Member of Yaojia Formation are delineated. By contrasting the plane distribution of oxidation fronts and sedimentary facies, it can be found that oxidation fronts are obviously controlled by interchannel fine-grained sediments. The movement of uranium-bearing oxidizing groundwater can be obstructed by interchannel sediments and pelitic interbeds of host sandstones. As a result, redox interface will be developed and uranium can be concentrated in neighbouring sandstones. (authors)

  1. Assessing the Renal Toxicity of Capstone Depleted Uranium Oxides and Other Uranium Compounds

    International Nuclear Information System (INIS)

    Roszell, Laurie E.; Hahn, Fletcher; Lee, Robyn B.; Parkhurst, MaryAnn

    2009-01-01

    The primary target for uranium toxicity is the kidney. The most frequently used guideline for uranium kidney burdens is the International Commission on Radiation Protection (ICRP) value of 3 (micro)g U/g kidney, a value that is based largely upon chronic studies in animals. In the present effort, we have developed a risk model equation to assess potential outcomes of acute uranium exposure. Twenty-seven previously published case studies in which workers were acutely exposed to soluble compounds of uranium (as a result of workplace accidents) were analyzed. Kidney burdens of uranium for these individuals were determined based on uranium in the urine, and correlated with health effects observed over a period of up to 38 years. Based upon the severity of health effects, each individual was assigned a score (- to +++) and then placed into an Effect Group. A discriminant analysis was used to build a model equation to predict the Effect Group based on the amount of uranium in the kidneys. The model equation was able to predict the Effect Group with 85% accuracy. The risk model was used to predict the Effect Group for Soldiers exposed to DU as a result of friendly fire incidents during the 1991 Gulf War. This model equation can also be used to predict the Effect Group of new cases in which acute exposures to uranium have occurred

  2. Method of making metal oxide ceramic powders by using a combustible amino acid compound

    Science.gov (United States)

    Pederson, Larry R.; Chick, Lawrence A.; Exarhos, Gregory J.

    1992-01-01

    This invention is directed to the formation of homogeneous, aqueous precursor mixtures of at least one substantially soluble metal salt and a substantially soluble, combustible co-reactant compound, typically an amino acid. This produces, upon evaporation, a substantially homogeneous intermediate material having a total solids level which would support combustion. The homogeneous intermediate material essentially comprises highly dispersed or solvated metal constituents and the co-reactant compound. The intermediate material is quite flammable. A metal oxide powder results on ignition of the intermediate product which combusts same to produce the product powder.

  3. The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

    International Nuclear Information System (INIS)

    Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W.

    1998-01-01

    While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study, the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation

  4. The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

    Energy Technology Data Exchange (ETDEWEB)

    Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W. [Los Alamos National Lab., NM (United States). Materials Science and Technology Div.

    1998-12-31

    While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation.

  5. The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

    Energy Technology Data Exchange (ETDEWEB)

    Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W.

    1998-12-31

    While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study, the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation.

  6. Beryllium Project: developing in CDTN of uranium dioxide fuel pellets with addition of beryllium oxide to increase the thermal conductivity

    International Nuclear Information System (INIS)

    Ferreira, Ricardo Alberto Neto; Camarano, Denise das Merces; Miranda, Odair; Grossi, Pablo Andrade; Andrade, Antonio Santos; Queiroz, Carolinne Mol; Gonzaga, Mariana de Carvalho Leal

    2013-01-01

    Although the nuclear fuel currently based on pellets of uranium dioxide be very safe and stable, the biggest problem is that this material is not a good conductor of heat. This results in an elevated temperature gradient between the center and its lateral surface, which leads to a premature degradation of the fuel, which restricts the performance of the reactor, being necessary to change the fuel before its full utilization. An increase of only 5 to 10 percent in its thermal conductivity, would be a significant increase. An increase of 50 percent would be a great improvement. A project entitled 'Beryllium Project' was developed in CDTN - Centro de Desenvolvimento da Tecnologia Nuclear, which aimed to develop fuel pellets made from a mixture of uranium dioxide microspheres and beryllium oxide powder to obtain a better heat conductor phase, filling the voids between the microspheres to increase the thermal conductivity of the pellet. Increases in the thermal conductivity in the range of 8.6% to 125%, depending on the level of addition employed in the range of 1% to 14% by weight of beryllium oxide, were obtained. This type of fuel promises to be safer than current fuels, improving the performance of the reactor, in addition to last longer, resulting in great savings. (author)

  7. Influence of attrition scrubbing, ultrasonic treatment, and oxidant additions on uranium removal from contaminated soils

    International Nuclear Information System (INIS)

    Timpson, M.E.; Elless, M.P.; Francis, C.W.

    1994-01-01

    As part of the Uranium in Soils Integrated Demonstration Project being conducted by the US Department of Energy, bench-scale investigations of selective leaching of uranium from soils at the Fernald Environmental Management Project site in Ohio were conducted at Oak Ridge National Laboratory. Two soils (storage pad soil and incinerator soil), representing the major contaminant sources at the site, were extracted using carbonate- and citric acid-based lixiviants. Physical and chemical processes were used in combination with the two extractants to increase the rate of uranium release from these soils. Attrition scrubbing and ultrasonic dispersion were the two physical processes utilized. Potassium permanganate was used as an oxidizing agent to transform tetravalent uranium to the hexavalent state. Hexavalent uranium is easily complexed in solution by the carbonate radical. Attrition scrubbing increased the rate of uranium release from both soils when compared with rotary shaking. At equivalent extraction times and solids loadings, however, attrition scrubbing proved effective only on the incinerator soil. Ultrasonic treatments on the incinerator soil removed 71% of the uranium contamination in a single extraction. Multiple extractions of the same sample removed up to 90% of the uranium. Additions of potassium permanganate to the carbonate extractant resulted in significant changes in the extractability of uranium from the incinerator soil but had no effect on the storage pad soil

  8. Oxidation of naturally reduced uranium in aquifer sediments by dissolved oxygen and its potential significance to uranium plume persistence

    Science.gov (United States)

    Davis, J. A.; Smith, R. L.; Bohlke, J. K.; Jemison, N.; Xiang, H.; Repert, D. A.; Yuan, X.; Williams, K. H.

    2015-12-01

    The occurrence of naturally reduced zones is common in alluvial aquifers in the western U.S.A. due to the burial of woody debris in flood plains. Such reduced zones are usually heterogeneously dispersed in these aquifers and characterized by high concentrations of organic carbon, reduced mineral phases, and reduced forms of metals, including uranium(IV). The persistence of high concentrations of dissolved uranium(VI) at uranium-contaminated aquifers on the Colorado Plateau has been attributed to slow oxidation of insoluble uranium(IV) mineral phases found in association with these reducing zones, although there is little understanding of the relative importance of various potential oxidants. Four field experiments were conducted within an alluvial aquifer adjacent to the Colorado River near Rifle, CO, wherein groundwater associated with the naturally reduced zones was pumped into a gas-impermeable tank, mixed with a conservative tracer (Br-), bubbled with a gas phase composed of 97% O2 and 3% CO2, and then returned to the subsurface in the same well from which it was withdrawn. Within minutes of re-injection of the oxygenated groundwater, dissolved uranium(VI) concentrations increased from less than 1 μM to greater than 2.5 μM, demonstrating that oxygen can be an important oxidant for uranium in such field systems if supplied to the naturally reduced zones. Dissolved Fe(II) concentrations decreased to the detection limit, but increases in sulfate could not be detected due to high background concentrations. Changes in nitrogen species concentrations were variable. The results contrast with other laboratory and field results in which oxygen was introduced to systems containing high concentrations of mackinawite (FeS), rather than the more crystalline iron sulfides found in aged, naturally reduced zones. The flux of oxygen to the naturally reduced zones in the alluvial aquifers occurs mainly through interactions between groundwater and gas phases at the water table

  9. Possible domestication of uranium oxides using biological assistance reduction

    Directory of Open Access Journals (Sweden)

    Slah Hidouri

    2017-01-01

    Full Text Available Uranium has been defined in material research engineering field as one of the most energetic radioactive elements in the entire Mendeleev periodic table. The manipulation of uranium needs higher theories and sophisticated apparatus even in nuclear energy extraction or in many other chemical applications. Above the nuclear exploitation level, the chemical conventional approaches used, require a higher temperature and pressure to control the destination of ionic form. However, it has been discovered later that at biological scale, the manipulation of this actinide is possible under friendly conditions. The review summarizes the relevant properties of uranium element and a brief characterization of nanoparticles, based on some structural techniques. These techniques reveal the common link between chemical approaches and biological assistance in nanoparticles. Also, those biological entities have been able to get it after reduction. Uranium is known for its ability to destroy ductile materials. So, if biological cell can really reduce uranium, then how does it work?

  10. Preparation of nano-iron oxide red pigment powders by use of cyanided tailings

    International Nuclear Information System (INIS)

    Li Dengxin; Gao Guolong; Meng Fanling; Ji Chong

    2008-01-01

    On one hand, cyanided tailings are one kind of pollutants. On the other hand, they contain a lot of valuable elements. So utilization of them can bring social and environmental benefits. In this paper, cyanided tailings were used to prepare nano-iron oxide red pigment powders by an ammonia process with urea as precipitant. At first, cyanided tailings were oxidized by nitric acid. Then, the oxidizing mixture was separated into solid and liquid parts. The liquid mixture was reduced by scrap iron and the impurity of it was removed by use of NH 3 .H 2 O. Then, the seed crystal of γ-FeOOH was obtained, when the pure liquid reacted with ammonia liquid at the selected experimental conditions. At last, nano-iron oxide red pigment powders were prepared. The structure, morphology and size distribution of seed crystal and iron oxide red were characterized systematically by means of X-ray diffraction (XRD), transmission electron microscope (TEM) and laser particle size analyzer (LPSA). The results revealed that typical iron oxide nanoparticles were α-Fe 2 O 3 with particle size of 50-70 nm. Furthermore, the factors that affected the hue and quality of the seed crystal and iron oxide red pigment were also discussed

  11. Annihilation characteristics of positrons in oxide powders in relation to catalytic activities

    International Nuclear Information System (INIS)

    Ito, K.; Ohtsu, Y.; Tanigawa, S.; Enomura, A.; Tsuda, N.

    1982-01-01

    The annihilation chaaracteristics in magnesium oxide powders were studied by the measurements of Doppler broadening of annihilation radiations. MgO powders are well known as a solid base and are utilized as a catalyst for the reactions which start by extracting protons from molecules such as decomposition of alcohol. The isochronal annealing behavior of annihilation characteristics in the process Mg(OH) 2 → MgO was found to correspond just to the change in the number of basic points, specific surface area and catalytic activities in some reactions. From the results of the thermal equilibrium measurements of MgO powders after dehydration, the temperature dependence of S parameter can be considered as the thermal activation process of the escape of positrons from trapped states at surface to form positroniums. The derived value of this activation energy was 0.187 eV. (Auth.)

  12. Preparation of Phosphonic Acid Functionalized Graphene Oxide-modified Aluminum Powder with Enhanced Anticorrosive Properties

    Science.gov (United States)

    He, Lihua; Zhao, Yan; Xing, Liying; Liu, Pinggui; Wang, Zhiyong; Zhang, Youwei; Liu, Xiaofang

    2017-07-01

    To improve the anticorrosive performance of aluminum powder, a common functional filler in polymer coatings, we report a novel method to prepare graphene oxide modified aluminum powder (GO-Al) using 3-aminoproplyphosphoic acid as "link" agent. The GO nanosheets were firstly functionalized with 3-aminoproplyphosphoic acid (APSA) by the reaction of amine groups of APSA and the epoxy groups of GO. Subsequently, a layer of GO nanosheets uniformly and tightly covered the surface of flaky aluminum particle though the strong linking strength between -PO(OH)2 functional groups of the modified GO and aluminum. The hydrogen evolution experiment suggests that the GO attached on the aluminum powder could effectively improve the anticorrosive performance of the pigments.

  13. Preparation of high density (8 to 9) uranium oxide UO2

    International Nuclear Information System (INIS)

    Eichner, C.; Ertaud, A.; Ortel, Y.; Stohr, J.; Vautrey, L.

    1948-10-01

    This report describes the process elaborated for the preparation of high density UO 2 . The thermal decomposition of uranium peroxide leads to UO 3 which is reduced by an hydrogen flow to obtain UO 2 . A UO 2 powder of good quality is obtained for temperatures below 650 deg. C. The powder is pulverized to obtain an homogeneous grain size and compressed inside a die to make pellets. Pellets are sintered up to 1600 deg. C in a reducing atmosphere and following a temperature rise law of 150 deg. C/hour. The equipment used (furnaces, gases purifier, control equipment, power supplies, thermoregulation systems) is described at the end. (J.S.)

  14. Thermochemistry of uranium(VI), arsenic, and alkali metal triple oxides

    International Nuclear Information System (INIS)

    Karyakin, N.V.; Chernorukov, G.N.

    1994-01-01

    The standard enthalpies of reactions of stoichiometric mixtures of potassium dyhydrogen orthoarsenate, uranium(VI) oxide, alkali metal nitrates, and of mixtures of triple oxides with the general formula M I AsUO 6 (M I =Li, Na, K, Rb, and Cs) and potassium nitrate with aqueous solution of hydrofluoric acid were determined an an adiabatic calorimeter at 298.15 K. The standard enthalpies of formation of uranium(VI), arsenic, and alkali metal triple oxides at 298.15 K were calculated form the data obtained. 8 refs., 1 tab

  15. Determination of uranium in plutonium--238 metal and oxide by differential pulse polarography

    International Nuclear Information System (INIS)

    Fawcett, N.C.

    1976-01-01

    A differential pulse polarographic method was developed for the determination of total uranium in 238 Pu metal and oxides. A supporting electrolyte of 0.5 M ascorbic acid in 0.15 N H 2 SO 4 was found satisfactory for the determination of 500 ppM or more of uranium in 10 mg or less of plutonium. A relative standard deviation of 0.27 to 4.3 percent was obtained in the analysis of samples ranging in uranium content from 0.65 to 2.79 percent. The limit of detection was 0.18 μg ml -1 . Peak current was a linear function of uranium concentration up to at least 100 μg ml -1 . Amounts of neptunium equal to the uranium content were tolerated. The possible interference of a number of other cations and anions were investigated

  16. Electrochemical behavior of uranium oxide in a LiCl-Li2O molten salt with the integrated cathode assembly

    International Nuclear Information System (INIS)

    Park, Sung Bin; Park, Byung Heung; Kang, Dae Seoung; Kwon, Seon Gil; Seo, Chung Seok; Park, Seong Won

    2005-01-01

    Electrochemical reduction of uranium oxide to uranium metal was studied in a LiCl-Li 2 O molten salt system. By means of a cyclic voltammetry and a chronopotentiometry, the electrolytic reduction of uranium oxide has been studied to establish the reduction mechanisms and the effects of the thickness of the uranium oxide on the overpotential of the cathode and anode were investigated. From the voltamograms, the reduction potentials of the uranium oxide and Li 2 O were obtained and the two mechanisms of the electrolytic reduction were considered with regards to the applied cathode potential. In the chronopotentiograms, the exchange current, the transfer coefficient and the maximum allowable current based on the Tafel behavior were obtained according to the thickness of the uranium oxide which is loaded into the porous MgO membrane. (author)

  17. Measurement of the oxidation-extraction of uranium from wet-process phosphoric acid

    International Nuclear Information System (INIS)

    Lawes, B.C.

    1985-01-01

    The present invention relates to processes for the recovery of uranium from wet-process phosphoric acid and more particularly to the oxidation-extraction steps in the DEPA-TOPO process for such recovery. A more efficient use of oxidant is obtained by monitoring the redox potential during the extraction step

  18. Biomembrane oxidizing tank used in the process of bacterial heap leaching of uranium ore

    International Nuclear Information System (INIS)

    Meng Yunsheng; Fan Baotuan; Liu Jian; Zheng Ying; Liu Chao

    2004-01-01

    The construction characteristic of biomembrane oxidizing tank and specialty of packing material used in the process of bacterial heap leaching of uranium ore are introduced in this paper. Method for designing biomembrane oxidizing tank, layout principle of aeration system and measurements on running management are summarized

  19. Effects of cocoa powder and dark chocolate on LDL oxidative susceptibility and prostaglandin concentrations in humans.

    Science.gov (United States)

    Wan, Y; Vinson, J A; Etherton, T D; Proch, J; Lazarus, S A; Kris-Etherton, P M

    2001-11-01

    Flavonoids are polyphenolic compounds of plant origin with antioxidant effects. Flavonoids inhibit LDL oxidation and reduce thrombotic tendency in vitro. Little is known about how cocoa powder and dark chocolate, rich sources of polyphenols, affect these cardiovascular disease risk factors. We evaluated the effects of a diet high in cocoa powder and dark chocolate (CP-DC diet) on LDL oxidative susceptibility, serum total antioxidant capacity, and urinary prostaglandin concentrations. We conducted a randomized, 2-period, crossover study in 23 healthy subjects fed 2 diets: an average American diet (AAD) controlled for fiber, caffeine, and theobromine and an AAD supplemented with 22 g cocoa powder and 16 g dark chocolate (CP-DC diet), providing approximately 466 mg procyanidins/d. LDL oxidation lag time was approximately 8% greater (P = 0.01) after the CP-DC diet than after the AAD. Serum total antioxidant capacity measured by oxygen radical absorbance capacity was approximately 4% greater (P = 0.04) after the CP-DC diet than after the AAD and was positively correlated with LDL oxidation lag time (r = 0.32, P = 0.03). HDL cholesterol was 4% greater after the CP-DC diet (P = 0.02) than after the AAD; however, LDL-HDL ratios were not significantly different. Twenty-four-hour urinary excretion of thromboxane B(2) and 6-keto-prostaglandin F(1)(alpha) and the ratio of the 2 compounds were not significantly different between the 2 diets. Cocoa powder and dark chocolate may favorably affect cardiovascular disease risk status by modestly reducing LDL oxidation susceptibility, increasing serum total antioxidant capacity and HDL-cholesterol concentrations, and not adversely affecting prostaglandins.

  20. Characterization of dense lead lanthanum titanate ceramics prepared from powders synthesized by the oxidant peroxo method

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Alexandre H. [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Departamento de Quimica, UFSCar-Universidade Federal de Sao Carlos, Rod.Washington Luis km 235, CP 676 Sao Carlos, SP 13565-905 (Brazil); Souza, Flavio L., E-mail: flavio.souza@ufabc.edu.br [Centro de Ciencias Naturais e Humanas, Universidade Federal do ABC, Rua Santa Adelia 166, Bangu, Santo Andre, SP 09210-170 (Brazil); Chiquito, Adenilson J., E-mail: chiquito@df.ufscar.br [Departamento de Fisica, UFSCar-Federal University of Sao Carlos, Rod.Washington Luis km 235, CP 676 Sao Carlos, SP 13565-905 (Brazil); Longo, Elson, E-mail: elson@iq.unesp.br [Instituto de Quimica de Araraquara, UNESP-Universidade Estadual Paulista, Rua Francisco Degni, CP 355 Araraquara, SP 14801-907 (Brazil); Leite, Edson R., E-mail: derl@power.ufscar.br [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Departamento de Quimica, UFSCar-Universidade Federal de Sao Carlos, Rod.Washington Luis km 235, CP 676 Sao Carlos, SP 13565-905 (Brazil); Camargo, Emerson R., E-mail: camargo@ufscar.br [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Departamento de Quimica, UFSCar-Universidade Federal de Sao Carlos, Rod.Washington Luis km 235, CP 676 Sao Carlos, SP 13565-905 (Brazil)

    2010-12-01

    Nanosized powders of lead lanthanum titanate (Pb{sub 1-x}La{sub x}TiO{sub 3}) were synthesized by means of the oxidant-peroxo method (OPM). Lanthanum was added from 5 to 30% in mol through the dissolution of lanthanum oxide in nitric acid, followed by the addition of lead nitrate to prepare a solution of lead and lanthanum nitrates, which was dripped into an aqueous solution of titanium peroxo complexes, forming a reactive amorphous precipitate that could be crystallized by heat treatment. Crystallized powders were characterized by FT-Raman spectroscopy and X-ray powder diffraction, showing that tetragonal perovskite structure is obtained for samples up to 25% of lanthanum and cubic perovskite for samples with 30% of lanthanum. Powders containing 25 and 30% in mol of lanthanum were calcined at 700 deg. C for 2 h, and in order to determine the relative dielectric permittivity and the phase transition behaviour from ferroelectric-to-paraelectric, ceramic pellets were prepared and sintered at 1100 or 1150 deg. C for 2 h and subjected to electrical characterization. It was possible to observe that sample containing 25% in mol of La presented a normal behaviour for the phase transition, whereas the sample containing 30% in mol of La presented a diffuse phase transition and relaxor behaviour.

  1. Characterization of dense lead lanthanum titanate ceramics prepared from powders synthesized by the oxidant peroxo method

    International Nuclear Information System (INIS)

    Pinto, Alexandre H.; Souza, Flavio L.; Chiquito, Adenilson J.; Longo, Elson; Leite, Edson R.; Camargo, Emerson R.

    2010-01-01

    Nanosized powders of lead lanthanum titanate (Pb 1-x La x TiO 3 ) were synthesized by means of the oxidant-peroxo method (OPM). Lanthanum was added from 5 to 30% in mol through the dissolution of lanthanum oxide in nitric acid, followed by the addition of lead nitrate to prepare a solution of lead and lanthanum nitrates, which was dripped into an aqueous solution of titanium peroxo complexes, forming a reactive amorphous precipitate that could be crystallized by heat treatment. Crystallized powders were characterized by FT-Raman spectroscopy and X-ray powder diffraction, showing that tetragonal perovskite structure is obtained for samples up to 25% of lanthanum and cubic perovskite for samples with 30% of lanthanum. Powders containing 25 and 30% in mol of lanthanum were calcined at 700 deg. C for 2 h, and in order to determine the relative dielectric permittivity and the phase transition behaviour from ferroelectric-to-paraelectric, ceramic pellets were prepared and sintered at 1100 or 1150 deg. C for 2 h and subjected to electrical characterization. It was possible to observe that sample containing 25% in mol of La presented a normal behaviour for the phase transition, whereas the sample containing 30% in mol of La presented a diffuse phase transition and relaxor behaviour.

  2. Depositing laser-generated nanoparticles on powders for additive manufacturing of oxide dispersed strengthened alloy parts via laser metal deposition

    Science.gov (United States)

    Streubel, René; Wilms, Markus B.; Doñate-Buendía, Carlos; Weisheit, Andreas; Barcikowski, Stephan; Henrich Schleifenbaum, Johannes; Gökce, Bilal

    2018-04-01

    We present a novel route for the adsorption of pulsed laser-dispersed nanoparticles onto metal powders in aqueous solution without using any binders or surfactants. By electrostatic interaction, we deposit Y2O3 nanoparticles onto iron-chromium based powders and obtain a high dispersion of nano-sized particles on the metallic powders. Within the additively manufactured component, we show that the particle spacing of the oxide inclusion can be adjusted by the initial mass fraction of the adsorbed Y2O3 particles on the micropowder. Thus, our procedure constitutes a robust route for additive manufacturing of oxide dispersion-strengthened alloys via oxide nanoparticles supported on steel micropowders.

  3. Ore controlling oxidized zonation epigenetic uranium-coal deposits and regularities in lignite transformations

    International Nuclear Information System (INIS)

    Uspenskij, V.A.; Kulakova, Ya.M.

    1982-01-01

    Complex of analytical methods was used to study epigenetic transformations in uranium-coal ore manifestation. To clarify the principle scheme of oxidized zonation in coals the materials, related to three similar objects were used. When comparing obtained epigenetic column with columns of similar ore objects the principle scheme of oxidized epigenetic zonation for ancient infiltration uranium-coal deposits was specified; general regularities of eignite transformations and characteristics of profile distribution of uranium and accessory metal zonations were revealed. Infiltration processes, proceeded in coal measureses, formed the steady epigenetic oxidized zonation: O - zone of barren unoxidized coals, 1 - zone of ore-bearing unoxidized coals, 2 - zone of weakly ore-bearing oxidized coals, 3 - zone of oxidized terrigenous rocks with zonules of development of yellow and red iron hydroxides. Capacities of some zones and zonules reflect the intensity and duration of ore-forming processes. Distribution of U and accessory elements obeys completely epigenetic zonation. It is assumed, that ancient infiltration uranium-coal deposits formed due to weakly uranium-bearing oxygen-containing waters

  4. A biamperometric method for the determination of O/U ratio in uranium oxide

    International Nuclear Information System (INIS)

    Xavier, Mary; Nair, P.R.; Aggarwal, S.K.

    2007-01-01

    The methodology based on the dissolution of the uranium dioxide samples in H 2 SO 4 + HF mixture and the indirect determination of U(VI) by biamperometric redox titration is a simple method for determining ratio in hyperstoichiometric UO 2 powders.Analytical methods for % measurements in hyperstoichiometric. The present paper describes a simple method based on the determination of U(IV) and total U by biamperometric titration

  5. The collection of uranium from sea water with hydrous metal oxide, 3

    International Nuclear Information System (INIS)

    Yamashita, Hisao; Nakajima, Fumito; Ozawa, Yoshihiro; Murata, Toshifumi.

    1980-01-01

    The effect of diverse ions present in sea water on the uranium adsorption is elucidated in the present paper. The uranium-adsorption experiments were conducted using sea water and a solution containing 0.72 mol dm -3 NaCl and 2.3 x 10 -3 mol dm -3 NaHCO 3 . The uranium uptake was about ten times larger from the NaCl-NaHCO 3 solution than from sea water. The ions which depressed the uranium uptake were the calcium, magnesium, and fluoride present in sea water. Among these ions, calcium had the largest effect on the uranium uptake. The analysis of calcium and carbonate in the adsorbent after the adsorption experiment has revealed that the molar ratio between calcium and carbonate was about one. It was considered that calcium carbonate was deposited on the adsorbent during the uranium adsorption. The specific surface area and the pore volume decreased after the deposition of calcium carbonate. It was supposed that the decrease in the uranium uptake was caused by the coverage of the surface of hydrous titanium(IV) oxide with calcium carbonate. Magnesium ions depressed the uranium uptake in the same manner as calcium ions. The effect of the magnesium ions, however, was relatively small compared with that of the calcium ions. (author)

  6. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

    Events in the Canadian uranium industry during 1980 are reviewed. Mine and mill expansions and exploration activity are described, as well as changes in governmental policy. Although demand for uranium is weak at the moment, the industry feels optimistic about the future. (LL)

  7. About the nitriding of powder uranium by nitrogen - Extract from the proceedings of the sessions of the Academy of Sciences, t. 253, p. 1100-1102, session of the 28 August 1961

    International Nuclear Information System (INIS)

    Moreau, Claude; Philippot, Joseph

    1961-01-01

    The authors report the study of powder uranium nitriding by nitrogen performed by thermogravimetry between 300 and 700 C. Results highlight the complexity of a pulverulent gas-solid reaction, and notably the influence of granulometry. Uranium powder is prepared by calciothermy, and presents nearly spherical grains with a diameter between 2 and 20 microns. Preliminary tests, performed on grains with heterogeneous size, showed that the reaction started between 300 and 350 C. Isotherm curves are discussed [fr

  8. The effect of hydrogen peroxide on uranium oxide films on 316L stainless steel

    Science.gov (United States)

    Wilbraham, Richard J.; Boxall, Colin; Goddard, David T.; Taylor, Robin J.; Woodbury, Simon E.

    2015-09-01

    For the first time the effect of hydrogen peroxide on the dissolution of electrodeposited uranium oxide films on 316L stainless steel planchets (acting as simulant uranium-contaminated metal surfaces) has been studied. Analysis of the H2O2-mediated film dissolution processes via open circuit potentiometry, alpha counting and SEM/EDX imaging has shown that in near-neutral solutions of pH 6.1 and at [H2O2] ⩽ 100 μmol dm-3 the electrodeposited uranium oxide layer is freely dissolving, the associated rate of film dissolution being significantly increased over leaching of similar films in pH 6.1 peroxide-free water. At H2O2 concentrations between 1 mmol dm-3 and 0.1 mol dm-3, formation of an insoluble studtite product layer occurs at the surface of the uranium oxide film. In analogy to corrosion processes on common metal substrates such as steel, the studtite layer effectively passivates the underlying uranium oxide layer against subsequent dissolution. Finally, at [H2O2] > 0.1 mol dm-3 the uranium oxide film, again in analogy to common corrosion processes, behaves as if in a transpassive state and begins to dissolve. This transition from passive to transpassive behaviour in the effect of peroxide concentration on UO2 films has not hitherto been observed or explored, either in terms of corrosion processes or otherwise. Through consideration of thermodynamic solubility product and complex formation constant data, we attribute the transition to the formation of soluble uranyl-peroxide complexes under mildly alkaline, high [H2O2] conditions - a conclusion that has implications for the design of both acid minimal, metal ion oxidant-free decontamination strategies with low secondary waste arisings, and single step processes for spent nuclear fuel dissolution such as the Carbonate-based Oxidative Leaching (COL) process.

  9. The crystallographic structure of the air-grown oxide on depleted uranium metal

    International Nuclear Information System (INIS)

    Jones, Christopher P.; Petherbridge, James R.; Davis, Sean A.; Jones, Jonathon A.; Scott, Thomas B.

    2016-01-01

    Highlights: • Oxidation of depleted uranium coupons under ambient conditions and 150 °C. • Oxide characterised using SEM, TEM and electron backscatter diffraction analysis, • Layer comprises of UO 2 crystallites 12 nm in diameter. • Preferred [110] growth direction normal to the surface of the metal. • Oxide growth direction is independent of the underlying crystal orientation. - Abstract: Oxide formation on depleted uranium metal was investigated using a combination of electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM) characterisation. Diffraction analysis of the oxide revealed an FCC crystalline formation of UO 2 crystallites whilst TEM data indicated an average grain size of 12 nm with a standard deviation of 3.8 nm. EBSD analysis revealed a preferential texture of [110] normal to the surface of the metal. This data implied that lattice matching between the oxide and the underlying metal did not occur, therefore, the observed preferential growth direction is independent of the underlying crystal orientation.

  10. Effects of pH on uranium uptake and oxidative stress responses induced in Arabidopsis thaliana

    OpenAIRE

    Saenen, Eline; Horemans, Nele; Vanhoudt, Nathalie; Vandenhove, Hildegarde; Biermans, Geert; Van Hees, May; Wannijn, Jean; Vangronsveld, Jaco; Cuypers, Ann

    2013-01-01

    Uranium (U) causes oxidative stress in Arabidopsis thaliana plants grown at pH 5.5. However, U speciation and its toxicity strongly depend on environmental parameters, for example pH. It is unknown how different U species determine U uptake and translocation within plants and how they might affect the oxidative defense mechanisms of these plants. The present study analyzed U uptake and oxidative stress-related responses in A. thaliana (Columbia ecotype) under contrasted U chemical speciation ...

  11. Exploration on trickle leaching of uranium ore by refreshed liquor of bacterial oxidation

    International Nuclear Information System (INIS)

    Chen Shian; Huang Xiangfu; Fan Baotuan

    1995-01-01

    The paper describes the adaptation of the domesticated thiobacillus ferroxidans to the trickle leaching conditions of uranium ore. When the bacterial leaching liquor through multiple cycles of oxidation and regeneration was used to return to the trickle leaching, the following results were obtained: the extraction rate was more than 95%, the acid consumption was saved by 30%, and the consumed 2.0% pyrolusite (MnO 2 40%) was eliminated. The following problems are discussed: the basic principle, process and some factors influencing the process of the trickle leaching of uranium ore using regenerated liquor of bacterial oxidation, counter-current trickle leaching mode, oxidation and regeneration techniques of bacterial leaching liquor and other technological problems on the process of uranium extraction by thiobacillus ferroxidans

  12. XPS characterization of the anodic oxide film formed on uranium metal in sodium hydroxide solution

    International Nuclear Information System (INIS)

    Fu Xiaoguo; Wang Xiaolin; Guo Huanjun; Wang Qingfu; Zhao Zhengping; Zhong Yongqiang

    2002-01-01

    X-ray photoelectron spectroscopy (XPS) is used to examine the anodic oxide film formed on uranium metal in 0.8 mol/L NaOH solution. The U4f 7/2 fitting spectra suggests that the anodic oxide film is composed of uranium trioxide and a small amount of UO 2+x . Under UHV condition, the U4f peak shifts to the lower binding energy, while a gradual increase in the intensity of U5f peak and the broad of U4f peak are also observed. All of these changes are due to reduction of uranium trioxide in the anodic oxide film. XPS quantitative analysis confirms the occurrence of reduction reaction

  13. Effect of process parameters on surface oxides on chromium-alloyed steel powder during sintering

    International Nuclear Information System (INIS)

    Chasoglou, D.; Hryha, E.; Nyborg, L.

    2013-01-01

    The use of chromium in the PM steel industry today puts high demands on the choice and control of the atmosphere during the sintering process due to its high affinity to oxygen. Particular attention is required in order to control the surface chemistry of the powder which in turn is the key factor for the successful sintering and production of PM parts. Different atmosphere compositions, heating rates and green densities were employed while performing sintering trials on water atomized steel powder pre-alloyed with 3 wt.% Cr in order to evaluate the effect on surface chemical reactions. Fracture surfaces of sintered samples were examined using high resolution scanning electron microscopy combined with X-ray microanalysis. The investigation was complemented with thermogravimetric (TG) studies. Reaction products in particulate form containing strong-oxide forming elements such as Cr, Si and Mn were formed during sintering for all conditions. Processing in vacuum results in intensive inter-particle neck development during the heating stage and consequently in the excessive enclosure of surface oxide which is reflected in less good final mechanical properties. Enhanced oxide reduction was observed in samples processed in hydrogen-containing atmospheres independent of the actual content in the range of 3–10 vol.%. An optimum heating rate was required for balancing reduction/oxidation processes. A simple model for the enclosure and growth of oxide inclusions during the sinter-neck development is proposed. The obtained results show that significant reduction of the oxygen content can be achieved by adjusting the atmosphere purity/composition. - Highlights: ► A local atmosphere microclimate is very important for sintering of PM steels. ► High risk of surface oxide enclosure between 800 and 1000 °C. ► Coalescence and agglomeration of enclosed oxides take place during sintering. ► The effect of different process parameters on the oxide reduction is examined. ► A

  14. White paper on possible inclusion of mixed plutonium-uranium oxides in DOE-STD-3013-96

    International Nuclear Information System (INIS)

    Haschke, J.M.; Venetz, T.; Szempruch, R.; McClard, J.W.

    1997-11-01

    This report assesses stabilization issues concerning mixed plutonium-uranium oxides containing 50 mass % Pu. Possible consequences of uranium substitution on thermal stabilization, specific surface areas, moisture readsorption behavior, loss-on-ignition analysis, and criticality safety of the oxide are examined and discussed

  15. Polarography applied to the determination of uranium oxide composition; Application de la polarographie a la determination de la composition d'oxydes d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Nens, C; Canton, C; Molina, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires, Departement de Chimie, Services d' Etudes Chimiques et d' Analyse, Service d' Etudes Analytiques, Section de Chimie Analytique du Plutonium et d' Electroanalyse

    1967-03-01

    An analytical method based on conventional polarography has been developed, for the determination of the O/U ratio in uranium oxides. The dissolution of the samples is effected by means of molten ammonium bifluoride. After a transfer to aqueous solution, polarography is used to determine the oxide composition by measurement of both the hexavalent and the total uranium. (author) [French] Une methode d'analyse utilisamt la polarographie conventionnelle a ete mise au point pour la determination du rapport O/U dans les oxydes d'uranium. La mise en solution des echgantillons est realisee en milieu de bifluorure d'ammonium fondu. Apres passage en salution aqueuse, la polarographie permet d'atteindre la composition de l'oxyde par determination de l'uranium hexavalent et de l'uranium total. (auteur)

  16. The relations between hydrodynamic characteristics and interbedding oxidation zone type uranium mineralization

    International Nuclear Information System (INIS)

    Bai Jingping

    2001-01-01

    Infiltrating type hydrodynamic way controls the formation of interbedding oxidation zone type uranium deposit. The author analyzes hydrodynamic condition of Songliao basin and concludes that during evolution and development of Songliao basin, Water-bearing petrofabric of Mingshui Formation and above inherit completely infiltrating hydrodynamic way as they were deposited and that Sifangtai Formation inherit the way to some extent, that below Sifangtai Formation water bearing petrofabric were completely reformed in northern part of Songliao watershed. The contact line between infiltrating and out filtrating type hydrodynamic way, e.g. underground water dividing lines formed in different geological period, restricts development of interbedding oxidation zone in this period and controls uranium mineralization

  17. Reaction of Antimony-Uranium Composite Oxide in the Chlorination Treatment of Waste Catalyst - 13521

    International Nuclear Information System (INIS)

    Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi

    2013-01-01

    The effect of oxygen gas concentration on the chlorination treatment of antimony-uranium composite oxide catalyst waste was investigated by adding different concentrations of oxygen at 0-6 vol% to its chlorination agent of 0.6 or 6 vol% hydrogen chloride gas at 1173 K. The addition of oxygen tended to prevent the chlorination of antimony in the oxide. When 6 vol% hydrogen chloride gas was used, the addition of oxygen up to 0.1 vol% could convert the uranium contained in the catalyst to U 3 O 8 without any significant decrease in the reaction rate compared to that of the treatment without oxygen. (authors)

  18. Synthesis and characterization of lead zirconate titanate powders obtained by the oxidant peroxo method

    Energy Technology Data Exchange (ETDEWEB)

    Camargo, Emerson R. [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Department of Chemistry, UFSCar-Federal University of Sao Carlos, Rod. Washington Luis km 235, CP 676, Sao Carlos SP, 13565-905 (Brazil)], E-mail: camargo@ufscar.br; Leite, Edson R. [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Department of Chemistry, UFSCar-Federal University of Sao Carlos, Rod. Washington Luis km 235, CP 676, Sao Carlos SP, 13565-905 (Brazil)], E-mail: derl@power.ufscar.br; Longo, Elson [Department of Biochemistry, Chemistry Institute of Araraquara, UNESP, Sao Paulo State University Rua Francisco Degni, CP 355 Araraquara SP, 14801-907 Brazil (Brazil)], E-mail: elson@iq.unesp.br

    2009-02-05

    Lead zirconate titanate (PbZr{sub 1-x}Ti{sub x}O{sub 3}) was synthesized by the 'oxidant peroxo method (OPM)' with 'x' between 0.25 and 0.50. Titanium metal was dissolved into a hydrogen peroxide/ammonia aqueous solution, followed by the addition of lead and zirconium nitrate solution. The amorphous precipitated precursor obtained was crystallized by heat treatment between 400 and 1000 deg. C. Images of transmission microscopy showed spherical particles with average diameter between 20 and 60 nm, and the presence of necks between particles treated at 700 deg. C. All of the unpressed powders were characterized by X-ray diffractometry and FT-Raman spectroscopy. Powder samples with 'x' up to 0.35 showed rhombohedral structure when treated at temperatures higher than 500 deg. C, and tetragonal structure when 'x' was higher than 0.40. Analysis of XRD and Raman spectroscopy of the precursor powders showed amorphous-like structures, however powders treated at 400 deg. C showed a structure identified as an intermediate pyrochlore phase, independently of the Zr and Ti mole ratio.

  19. Synthesis and characterization of lead zirconate titanate powders obtained by the oxidant peroxo method

    International Nuclear Information System (INIS)

    Camargo, Emerson R.; Leite, Edson R.; Longo, Elson

    2009-01-01

    Lead zirconate titanate (PbZr 1-x Ti x O 3 ) was synthesized by the 'oxidant peroxo method (OPM)' with 'x' between 0.25 and 0.50. Titanium metal was dissolved into a hydrogen peroxide/ammonia aqueous solution, followed by the addition of lead and zirconium nitrate solution. The amorphous precipitated precursor obtained was crystallized by heat treatment between 400 and 1000 deg. C. Images of transmission microscopy showed spherical particles with average diameter between 20 and 60 nm, and the presence of necks between particles treated at 700 deg. C. All of the unpressed powders were characterized by X-ray diffractometry and FT-Raman spectroscopy. Powder samples with 'x' up to 0.35 showed rhombohedral structure when treated at temperatures higher than 500 deg. C, and tetragonal structure when 'x' was higher than 0.40. Analysis of XRD and Raman spectroscopy of the precursor powders showed amorphous-like structures, however powders treated at 400 deg. C showed a structure identified as an intermediate pyrochlore phase, independently of the Zr and Ti mole ratio

  20. Influence of Diatomite and Mineral Powder on Thermal Oxidative Ageing Properties of Asphalt

    Directory of Open Access Journals (Sweden)

    Yongchun Cheng

    2015-01-01

    Full Text Available Ageing of asphalt affects the performances of asphalt pavement significantly. Therefore, effects of diatomite and mineral powder on ageing properties of asphalt were investigated systematically in order to improve the antiageing property of mixture. Thin film oven test (TFOT was used to conduct the short term ageing in laboratory. Softening points, penetrations, force ductility, low temperature creep properties, and viscosities of asphalt mastics were tested before and after TFOT, respectively. Results indicated that percent retained penetration (PRP increased with the increasing of fillers. Increment of softening point (ΔT, ductility retention rate (DRR, deformation energy ageing index (JAI, and viscosity ageing index (VAI of asphalt mastics nonlinearly decreased with the increasing of fillers. Ageing of asphalt was reduced by diatomite and mineral powder. And the antiageing effect of diatomite was better than that of mineral powder as a result of its porous structure. It is suggested that the mineral powder could be reasonably replaced by diatomite in order to reduce thermal oxidative ageing of asphalt mixture. The optimal content of diatomite 12.8% is also suggested for engineering.

  1. Influence of packaging atmosphere on the formation of cholesterol oxides in [gamma]-irradiated egg powder

    Energy Technology Data Exchange (ETDEWEB)

    Lebovics, V.K.; Gaal, O. (National Inst. of Food Hygiene and Nutrition, Budapest (Hungary)); Farkas, J.; Somogyi, L. (University of Horticulture and Food Industry, Budapest (Hungary))

    1993-09-01

    In the present work the influence of aerobic and anoxic conditions have been comparatively investigated to study the chemical changes of cholesterol in [gamma]-irradiated egg powder. The irradiation treatment was carried out with powdered egg packed under air and also under vacuum in polyethylene (PE) bags and in laminated, oxygen impermeable three-layer (polyester-aluminium-polyethylene) foil to dosage levels 2 and 4 kGy at room temperature. The cholesterol oxidation is demonstrated by thin-layer chromatograms. In the egg powder wrapped in Pe bags the predominant cholesterol derivatives -7-hydroxycholesterol isomers ([alpha] and [beta]) - accumulated in significant amounts (increasing with dose) while vacuum-packaging in laminated foil considerably suppressed the quantity of these products and prevented the formation of cholesterol 5[alpha], 6[alpha]-epoxide as well as 7-ketocholesterol. Little or no change was observed in non-irradiated (control) vacuum-packed egg powder stored at approximately 22[sup o]C for up to 5 months. Peroxide values showed changes parallel to the formation of COPs. (author).

  2. Ammonium carbonate and/or bicarbonate plus alkaline chlorate oxidant for recovery of uranium values

    International Nuclear Information System (INIS)

    Stapp, P.R.

    1983-01-01

    In accordance with the present invention, uranium values are extracted from materials containing uranium in valence states lower than its hexavalent state by contacting the materials containing uranium with an aqueous alkaline leach solution containing an alkaline chlorate in an amount sufficient to oxidize at least a portion of the uranium in valence states lower than its hexavalent state to its hexavalent state. In a further embodiment of the present invention, the alkaline leach solution is an aqueous solution of a carbonate selected from the group consisting of ammonium carbonate, ammonium bicarbonate and mixtures thereof. In yet another embodiment of the present invention, at least one catalytic compound of a metal selected from the group consisting of copper, cobalt, iron, nickel, chromium and mixtures thereof adapted to assure the presence of the ionic species Cu ++ , Co ++ , Fe +++ , Ni ++ , Cr +++ and mixtures thereof, respectively, during the contacting of the material containing uranium with the alkaline leach solution and in an amount sufficient to catalyze the oxidation of at least a portion of the uranium in its lower valence states to its hexavalent state, is present

  3. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

    Evidence of expanding markets, improved prices and the short supply of uranium became abundantly clear in 1975, providing the much needed impetus for widespread activity in all phases of uranium operations. Exploration activity that had been at low levels in recent years in Canada was evident in most provinces as well as the Northwest Territories. All producers were in the process of expanding their uranium-producing facilities. Canada's Atomic Energy Control Board (AECB) by year-end had authorized the export of over 73,000 tons of U/sub 3/0/sub 8/ all since September 1974, when the federal government announced its new uranium export guidelines. World production, which had been in the order of 25,000 tons of U/sub 3/0/sub 8/ annually, was expected to reach about 28,000 tons in 1975, principally from increased output in the United States.

  4. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  5. The collection of uranium from sea water with hydrous metal oxide, 2

    International Nuclear Information System (INIS)

    Yamashita, Hisao; Nakajima, Fumito; Ozawa, Yoshihiro; Murata, Toshifumi.

    1980-01-01

    The mechanism of uranium adsorption from sea water by hydrous titanium(IV) oxide was investigated. The uranium adsorption experiments were conducted using a solution containing NaCl, NaHCO 3 , and uranium. Thermochemical calculation showed that the tris(carbonato)dioxouranate(VI) ion (UO 2 (CO 3 ) 3 ) 4- found in sea water existed in the solution at pH 8 and 25 0 C. The uranium uptake varied with the pH of the solution, exhibiting a minimum value at pH 8. The enthalpy change, delta H, and the activation energy, E, of the uranium adsorption were found to be 23.6 kJ mol -1 and 52.7 kJ mol -1 respectively. The analysis of carbonate in the adsorbent showed that the carbonate ion in(UO 2 (CO 3 ) 3 ) 4- was released into the solution during the uranium adsorption. On the basis of the present experimental results, the mechanism of uranium adsorption was discussed. (author)

  6. Redox reaction between graphene oxide and In powder to prepare In2O3/reduced graphene oxide hybrids for supercapacitors

    Science.gov (United States)

    Xu, Xiaoyang; Wu, Tao; Xia, Fengling; Li, Yi; Zhang, Congcong; Zhang, Lei; Chen, Mingxi; Li, Xichuan; Zhang, Li; Liu, Yu; Gao, Jianping

    2014-11-01

    A facile and quick route for the chemical reduction of graphene oxide (GO) using In powder as a reductant has been established. The reduction of GO by In powder is traced by UV-visible absorption spectroscopy, and the obtained reduced graphene oxide (rGO) is analyzed. The In3+ ions produced during the reaction between the GO and the In powder are chemically transformed to In2O3 and then form In2O3/rGO hybrids. The In2O3/rGO hybrids are used as electrode materials and their electrochemical performance are studied using cyclic voltammetry and galvanostatic charge/discharge. The In2O3/rGO hybrids demonstrate excellent electrochemical performance and their highest specific capacitance is 178.8 F g-1 which is much higher than that of either In2O3 or rGO. In addition, the In2O3/rGO hybrids are also very stable.

  7. Determination of size and shape distributions of metal and ceramic powders; Odredjivanje raspodele velicina, specificne povrsine i oblika metalnih i keramickih prahova

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, DI [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    For testing the size and shape distributions of metal and ceramic uranium oxide powders the following method for analysing the grain size of powders were developed and implemented: microscopic analysis and sedimentation method. A gravimetry absorption device was constructed for determining the specific surfaces of powders.

  8. Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development

    International Nuclear Information System (INIS)

    Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

    2007-01-01

    The uranium dioxide is the most used substance as nuclear reactor fuel for presenting many advantages such as: high stability even when it is in contact with water in high temperatures, high fusion point, and high capacity to retain fission products. The conventional fuel is made with ceramic sintered pellets of uranium dioxide stacked inside fuel rods, and presents disadvantages because its low thermal conductivity causes large and dangerous temperature gradients. Besides, the thermal conductivity decreases further as the fuel burns, what limits a pellet operational lifetime. This research developed a new kind of fuel pellets fabricated with uranium dioxide kernels and beryllium oxide filling the empty spaces between them. This fuel has a great advantage because of its higher thermal conductivity in relation to the conventional fuel. Pellets of this kind were produced, and had their thermophysical properties measured by the flash laser method, to compare with the thermal conductivity of the conventional uranium dioxide nuclear fuel. (author) (author)

  9. Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

    International Nuclear Information System (INIS)

    Hanus, D.; Kleykamp, H.

    1982-01-01

    Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

  10. Study the oxidation kinetics of uranium using XRD and Rietveld method

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Yanzhi; Guan Weijun; Wang Qinguo; Wang Xiaolin; Lai Xinchun; Shuai Maobing, E-mail: yanzhizh@163.com [China Academy of Engineering Physics, PO Box 919-71, Mianyang, Sichuan, 621900 (China)

    2010-03-15

    The surface oxidation of uranium metal has been studied by X-ray diffraction (XRD) and Rietveld method in the range of 50{approx}300deg. C in air. The oxidation processes are analyzed by XRD to determine the extent of surface oxidation and the oxide structure. The dynamics expression for the formation of UO{sub 2} was derived. At the beginning, the dynamic expression was nonlinear, but switched to linear subsequently for uranium in air and humid oxygen. That is, the growth kinetics of UO{sub 2} can be divided into two stages: nonlinear portion and linear portion. Using the kinetic data of linear portion, the activation energy of reaction between uranium and air was calculated about 46.0 kJ/mol. However the content of oxide as a function of time was linear in humid helium ambience. Contrast the dynamics results, it prove that the absence of oxygen would accelerate the corrosion rate of uranium in the humid gas. We can find that the XRD and Rietveld method are a useful convenient method to estimate the kinetics and thermodynamics of solid-gas reaction.

  11. Study the oxidation kinetics of uranium using XRD and Rietveld method

    Science.gov (United States)

    Zhang, Yanzhi; Guan, Weijun; Wang, Qinguo; Wang, Xiaolin; Lai, Xinchun; Shuai, Maobing

    2010-03-01

    The surface oxidation of uranium metal has been studied by X-ray diffraction (XRD) and Rietveld method in the range of 50~300°C in air. The oxidation processes are analyzed by XRD to determine the extent of surface oxidation and the oxide structure. The dynamics expression for the formation of UO2 was derived. At the beginning, the dynamic expression was nonlinear, but switched to linear subsequently for uranium in air and humid oxygen. That is, the growth kinetics of UO2 can be divided into two stages: nonlinear portion and linear portion. Using the kinetic data of linear portion, the activation energy of reaction between uranium and air was calculated about 46.0 kJ/mol. However the content of oxide as a function of time was linear in humid helium ambience. Contrast the dynamics results, it prove that the absence of oxygen would accelerate the corrosion rate of uranium in the humid gas. We can find that the XRD and Rietveld method are a useful convenient method to estimate the kinetics and thermodynamics of solid-gas reaction.

  12. Synthesis and Characterization of Oxide Feedstock Powders for the Fuel Cycle R and D Program

    International Nuclear Information System (INIS)

    Voit, Stewart L.; Vedder, Raymond James; Johnson, Jared A.

    2010-01-01

    Nuclear fuel feedstock properties, such as physical, chemical, and isotopic characteristics, have a significant impact on the fuel fabrication process and, by extension, the in-reactor fuel performance. This has been demonstrated through studies with UO 2 spanning greater than 50 years. The Fuel Cycle R and D Program with The Department of Energy Office of Nuclear Energy has initiated an effort to develop a better understanding of the relationships between oxide feedstock, fresh fuel properties, and in-reactor fuel performance for advanced mixed oxide compositions. Powder conditioning studies to enable the use of less than ideal powders for ceramic fuel pellet processing are ongoing at Los Alamos National Laboratory (LANL) and an understanding of methods to increase the green density and homogeneity of pressed pellets has been gained for certain powders. Furthermore, Oak Ridge National Laboratory (ORNL) is developing methods for the co-conversion of mixed oxides along with techniques to analyze the degree of mixing. Experience with the fabrication of fuel pellets using co-synthesized multi-constituent materials is limited. In instances where atomically mixed solid solutions of two or more species are needed, traditional ceramic processing methods have been employed. Solution-based processes may be considered viable synthesis options, including co-precipitation (AUPuC), direct precipitation, direct-conversion (Modified Direct Denitration or MDD) and internal/external gelation (sol-gel). Each of these techniques has various advantages and disadvantages. The Fiscal Year 2010 feedstock development work at ORNL focused on the synthesis and characterization of one batch of UO x and one batch of U 80 Ce 20 O x . Oxide material synthesized at ORNL is being shipped to LANL for fuel fabrication process development studies. The feedstock preparation was performed using the MDD process which utilizes a rotary kiln to continuously thermally denitrate double salts of ammonium

  13. Critical experiments on low-enriched uranium oxide system with H/U=1.25

    International Nuclear Information System (INIS)

    Oh, I.; Rothe, R.E.; Tuck, G.

    1982-01-01

    Fifteen (15) critical experiments were performed on a horizontal split table machine using 4.48%-enriched sup(235)U uranium oxide(U 3 O 8 ). The oxide was compacted to a density of 4.68g/cm 3 and placed in 152 mm cubical aluminum cans. Water was added to achive an H/U of 1.25. Various arrays of oxide cans were distributed on each half of the split table, and the separation between halves reduced until criticality occurred. The critical table separation varied from 3.59 mm to 18.40 mm. Twelve (12) experiments required the addition of a high-enriched(-93 %sup(235)U) metal or solution driver to achieve criticality. These experiments were performed in a plastic, concrete, or thin steel reflector. Three additional experiments in the plastic reflector contained either 9.3-mm- or 24.3-mm-thick plastic moderator material between the oxide cans and did not require a driver to achieve criticality. Critical uranium driver masses ranged from 9.999 kg to 14.000 kg (solution driver), and from 25.378 kg to 29.278 kg (metal driver) for 5X5X5 arrays of uranium oxide cans. Always, one or four of these 125 cans had to be removed to make room for the drivers. Therefore, the uranium oxide masses used were 1823.8 kg and 1863.5 kg. For the moderated experiments, the uranium oxide mass ranged between 574.4 kg and 1210.0 kg. (Author)

  14. Gas atomized precursor alloy powder for oxide dispersion strengthened ferritic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Rieken, Joel [Iowa State Univ., Ames, IA (United States)

    2011-12-13

    Gas atomization reaction synthesis (GARS) was employed as a simplified method for producing precursor powders for oxide dispersion strengthened (ODS) ferritic stainless steels (e.g., Fe-Cr-Y-(Ti,Hf)-O), departing from the conventional mechanical alloying (MA) process. During GARS processing a reactive atomization gas (i.e., Ar-O2) was used to oxidize the powder surfaces during primary break-up and rapid solidification of the molten alloy. This resulted in envelopment of the powders by an ultra-thin (t < 150 nm) metastable Cr-enriched oxide layer that was used as a vehicle for solid-state transport of O into the consolidated microstructure. In an attempt to better understand the kinetics of this GARS reaction, theoretical cooling curves for the atomized droplets were calculated and used to establish an oxidation model for this process. Subsequent elevated temperature heat treatments, which were derived from Rhines pack measurements using an internal oxidation model, were used to promote thermodynamically driven O exchange reactions between trapped films of the initial Cr-enriched surface oxide and internal Y-enriched intermetallic precipitates. This novel microstructural evolution process resulted in the successful formation of nano-metric Y-enriched dispersoids, as confirmed using high energy X-ray diffraction and transmission electron microscopy (TEM), equivalent to conventional ODS alloys from MA powders. The thermal stability of these Y-enriched dispersoids was evaluated using high temperature (1200°C) annealing treatments ranging from 2.5 to 1,000 hrs of exposure. In a further departure from current ODS practice, replacing Ti with additions of Hf appeared to improve the Y-enriched dispersoid thermal stability by means of crystal structure modification. Additionally, the spatial distribution of the dispersoids was found to depend strongly on the original rapidly solidified microstructure. To exploit this, ODS microstructures were engineered from

  15. Study of interaction of uranium, plutonium and rare earth fluorides with some metal oxides in fluoric salt melts

    International Nuclear Information System (INIS)

    Gorbunov, V.F.; Novoselov, G.P.; Ulanov, S.A.

    1976-01-01

    Interaction of plutonium, uranium, and rare-earth elements (REE) fluorides with aluminium and calcium oxides in melts of eutectic mixture LiF-NaF has been studied at 800 deg C by X-ray diffraction method. It has been shown that tetravalent uranium and plutonium are coprecipitated by oxides as a solid solution UO 2 -PuO 2 . Trivalent plutonium in fluorides melts in not precipitated in the presence of tetravalent uranium which can be used for their separation. REE are precipitated from a salt melt by calcium oxide and are not precipitated by aluminium oxide. Thus, aluminium oxide in a selective precipitator for uranium and plutonium in presence of REE. Addition of aluminium fluoride retains trivalent plutonium and REE in a salt melt in presence of Ca and Al oxides. The mechanism of interacting plutonium and REE trifluorides with metal oxides in fluoride melts has been considered

  16. Spectrochemical method of uranium determination in sea water

    International Nuclear Information System (INIS)

    Koval'chuk, L.I.; Koryukova, V.P.; Andrianov, A.M.

    1979-01-01

    A spectrochemical method of uranium determination in sea water is reported. The method involves the use of hydrated titanium oxide as a concentrator and a substrate for the analysis. The uranium-containing concentrate mixed with carbon powder (1:1) is burned in the alternating current ark (i=15 A) and the spectra are recorded by a diffraction spectrometer. The analytical line of uranium is 2865.14 A. The variation coefficient is 12%

  17. Metallogenic characteristics, model and exploration prospect for the paleo-interlayer-oxidation type sandstone-hosted uranium deposits in China

    International Nuclear Information System (INIS)

    Huang Jingbai; Li Shengxiang

    2007-01-01

    In this paper, the paleo-interlayer-oxidation type sandstone-hosted uranium deposits occurred in the Meso-Cenozoic continental basins in China are divided into 3 subtype, they are stratum over lapping buried subtype, structure-uplifting destroy subtype and faulted-folding conserved subtype. The metallogenic characteristics, metallogenic model and exploration prospect for these 3 subtypes uranium deposits are discussed. It is proposed that the paleo-interlayer-oxidation type sandstone-hosted uranium deposits, besides the recent interlayer oxidation type sandstone-hosted uranium deposits, are of great prospecting potential in the Meso-Cenozoic continental basins in China. Therefore, the metallogenic theory of these types uranium deposits should be conscientiously summarized and replenished continuously so as to propel forward the exploration of the sandstone-hosted uranium deposits in China. (authors)

  18. The study on microb and organic metallogenetic process of the interlayer oxidized zone uranium deposit. A case study of the Shihongtan uranium deposit in Turpan-Hami basin

    International Nuclear Information System (INIS)

    Qiao Haiming; Shang Gaofeng

    2010-01-01

    Microbial and organic process internationally leads the field in the study of metallogenetic process presently. Focusing on Shi Hongtan uranium deposit, a typical interlayer oxidized zone sandstone-type deposit, this paper analyzes the geochemical characteristics of microb and organic matter in the deposit, and explores the interaction of microb and organic matter. It considers that the anaerobic bacterium actively takes part in the formation of the interlayer oxidized zone, as well as the mobilization and migration of uranium. In the redox (oxidation-reduction) transition zone, sulphate-reducing bacteria reduced sulphate to stink damp, lowing Eh and acidifying pH in the groundwater, which leads to reducing and absorbing of uranium, by using light hydrocarbon which is the product of the biochemical process of organism and the soluble organic matter as the source of carbon. The interaction of microb and organic matter controls the metallogenetic process of uranium in the deposit. (authors)

  19. Electrodeposition of uranium metal by reduction of uranium oxides in molten Lif-KF=NaF-CaF 2-UF4

    International Nuclear Information System (INIS)

    Pao, D.S.; Burris, L.; Steunenberg, R.K.; Tomczuk, Z.

    1990-01-01

    Although electrolytic reduction of uranium oxides was shown to be feasible in the early 1960's it is recognized that considerable improvement in the electrolytic reduction technology must be achieved for practical applications. This exploratory work on electrolytic reduction of uranium oxide was undertaken to investigate potential improvements in the technology. The approach taken was to deposit solid uranium metal directly on a solid cathode at temperatures below the melting point of uranium (1132 degrees C). The lower temperature electrolytic reduction process has several advantages over the existing chemical reduction processes. It lessens materials problems and special heating and insulating requirements associated with high-temperature operations. It removes most impurities. It does not produce the large quantities of byproduct oxides wastes typical of chemical reduction processes

  20. A study on the formation of uranium carbide in an induction furnace

    International Nuclear Information System (INIS)

    Song, In Young; Lee, Yoon Sang; Kim, Eung Soo; Lee, Don Bae; Kim, Chang Kyu

    2005-01-01

    Uranium is a typical carbide-forming element. Three carbides, UC, U 2 C 3 and UC 2 , are formed in the uranium-carbon system. The most important of these as fuel is uranium monocarbide UC. It is well known that Uranium carbides can be obtained by three basic methods: 1) by reaction of uranium metal with carbon; 2) by reaction of uranium metal powder with gaseous hydrocarbons; 3) by reaction of uranium oxides with carbon. The use of uranium monocarbide, or materials based on it, has great prospects as fuel for nuclear reactors. It is quite possible that uranium dicarbide UC 2 may also acquire great importance as a fuel, particularly in dispersion fuel elements with graphite matrix. In the present study, uranium carbides are obtained by direct reaction of uranium metal with graphite in a high frequency induction furnace

  1. Study on thermo-oxide layers of uranium-niobium alloy

    International Nuclear Information System (INIS)

    Luo Lizhu; Yang Jiangrong; Zhou Ping

    2010-01-01

    Surface oxides structure of uranium-niobium alloys which were annealed under different temperatures (room temperature, 100, 200, 300 degree C, respectively)in air were studied by X-ray photoelectron spectroscopy (XPS) analysis and depth profile. Thickness of thermo-oxide layers enhance with the increasing oxide temperature, and obvious changes to oxides structure are observed. Under different delt temperatures, Nb 2 O 5 are detected on the initial surface of U-Nb alloys, and a layer of NbO mixed with some NbO x (0 2 O 5 and Nb metal. Dealing samples in air from room temperature to 200 degree C, non-stoichiometric UO 2+x (UO 2 + interstitial oxygen, P-type semiconductor) are found on initial surface of U-Nb alloys, which has 0.7 eV shift to lower binding energy of U 4f 7/2 characteristics comparing to that of UO 2 . Under room temperature, UO 2 are commonly detected in the oxides layer, while under temperature of 100 and 200 degree C, some P-type UO 2+x are found in the oxide layers,which has a satellite at binding energy of 396.6 eV. When annealing at 300 degree C, higher valence oxides, such as U 3 O 8 or UO x (2 5/2 and U 4f 7/2 peaks are 392.2 and 381.8 eV, respectively. UO 2 mixed uranium metal are the main compositions in the oxide layers. From the results, influence of temperature to oxidation of uranium is more visible than to niobium in uranium-niobium alloys. (authors)

  2. Reducing retrogradation and lipid oxidation of normal and glutinous rice flours by adding mango peel powder.

    Science.gov (United States)

    Siriamornpun, Sirithon; Tangkhawanit, Ekkarat; Kaewseejan, Niwat

    2016-06-15

    Green and ripe mango peel powders (MPP) were added to normal rice flour (NRF) and glutinous rice flour (GRF) at three levels (400, 800 and 1200 ppm) and their effects on physicochemical properties and lipid oxidation inhibition were investigated. Overall, MPP increased the breakdown viscosity and reduced the final viscosity in rice flours when compared to the control. Decreasing in retrogradation was observed in both NRF and GRF with MPP added of all levels. MPP addition also significantly inhibited the lipid oxidation of all flours during storage (30 days). Retrogradation values were strongly negatively correlated with total phenolic and flavonoid contents, but not with fiber content. The hydrogen bonds and hydrophilic interactions between phenolic compounds with amylopectin molecule may be involved the decrease of starch retrogradation, especially GRF. We suggest that the addition of MPP not only reduced the retrogradation but also inhibited the lipid oxidation of rice flour. Copyright © 2016 Elsevier Ltd. All rights reserved.

  3. A comparative study on determination of composition of uranium thorium mixed oxides by tube and radioisotope excited EDXRF

    International Nuclear Information System (INIS)

    Dhara, Sangita; Sanjay Kumar, S.; Misra, N.L.; Aggarwal, S.K.; Singh, Ajit Kumar; Lodha, G.S.

    2009-01-01

    Energy Dispersive X-ray Fluorescence (EDXRF) methods for determination of uranium and thorium in their mixed oxide matrices using tube and radioisotope excitation sources have been developed. The methodology involves preparation of mixed oxide calibration/sample mixtures of uranium and thorium oxides, mixing of fixed amount of internal standard Yttrium in form of Yttrium oxide, pelletizing these mixtures after thorough mixing and recording their EDXRF spectra using Rh target as well as 109 Cd radioisotope source. The samples were analysed for uranium and thorium on the basis of calibration plots

  4. Physicochemical characterization of Capstone depleted uranium aerosols III: morphologic and chemical oxide analyses.

    Science.gov (United States)

    Krupka, Kenneth M; Parkhurst, Mary Ann; Gold, Kenneth; Arey, Bruce W; Jenson, Evan D; Guilmette, Raymond A

    2009-03-01

    The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using x-ray diffraction (XRD), and particle morphologies were examined using scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS). The oxidation state of a DU aerosol is important as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles were spherical, occasionally with dendritic or lobed surface structures. Others appear to have fractures that perhaps resulted from abrasion and comminution, or shear bands that developed from plastic deformation of the DU material. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small bits of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of Health Physics to interpret the

  5. Physicochemical Characterization of Capstone Depleted Uranium Aerosols III: Morphologic and Chemical Oxide Analyses

    International Nuclear Information System (INIS)

    Krupka, Kenneth M.; Parkhurst, MaryAnn; Gold, Kenneth; Arey, Bruce W.; Jenson, Evan D.; Guilmette, Raymond A.

    2009-01-01

    The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using X-ray diffraction (XRD) and particle morphologies using scanning electron microscopy/energy dispersive spectrometry (SEM/EDS). The oxidation state of a DU aerosol is important as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles appear to have been fractured (perhaps as a result of abrasion and comminution); others were spherical, occasionally with dendritic or lobed surface structures. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small chunks of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of The Journal of Health Physics to interpret the results of lung solubility studies and in selecting input parameters for

  6. Study on Exothermic Oxidation of Acrylonitrile-butadiene-styrene (ABS Resin Powder with Application to ABS Processing Safety

    Directory of Open Access Journals (Sweden)

    Jenq-Renn Chen

    2010-08-01

    Full Text Available Oxidative degradation of commercial grade ABS (Acrylonitrile-butadiene-styrene resin powders was studied by thermal analysis. The instabilities of ABS containing different polybutadiene (PB contents with respect to temperature were studied by Differential Scanning Calorimeter (DSC. Thermograms of isothermal test and dynamic scanning were performed. Three exothermic peaks were observed and related to auto-oxidation, degradation and oxidative decomposition, respectively. Onset temperature of the auto-oxidation was determined to be around 193 °C. However, threshold temperature of oxidation was found to be as low as 140 °C by DSC isothermal testing. Another scan of the powder after degeneration in air showed an onset temperature of 127 °C. Reactive hazards of ABS powders were verified to be the exothermic oxidation of unsaturated PB domains, not the SAN (poly(styrene-acrylonitrile matrix. Heat of oxidation was first determined to be 2,800 ± 40 J per gram of ABS or 4,720 ± 20 J per gram of PB. Thermal hazards of processing ABS powder are assessed by adiabatic temperature rise at process conditions. IR spectroscopy associated with heat of oxidation verified the oxidative mechanism, and these evidences excluded the heat source from the degradation of SAN. A specially prepared powder of ABS without adding anti-oxidant was analyzed by DSC for comparing the exothermic behaviors. Exothermic onset temperatures were determined to be 120 °C and 80 °C by dynamic scanning and isothermal test, respectively. The assessment successfully explained fires and explosions in an ABS powder dryer and an ABS extruder.

  7. Thermochemistry of the complex oxides of uranium, vanadium, and alkali metals

    International Nuclear Information System (INIS)

    Karyakin, N.V.; Chernorukov, N.G.; Suleimanov, E.V.; Kharyushina, E.A.

    1992-01-01

    The standard enthalpies of the formation at T 298.15 K of complex oxides of uranium(VI), vanadium(V) and alkali metals with the general formula M 1 VUO 6 where M 1 = Na, K, Rb, and Cs, were calculated from the results of calorimetric experiments and from published data. 8 refs., 1 tab

  8. A system for the synthesis of uranium hexafluoride by high pressure fluorination of uranium oxides

    International Nuclear Information System (INIS)

    Elizalde T, J.; Saniger B, J.M.; Nava S, R.

    1986-01-01

    An equipment for the synthesis of uranium hexafluoride by a direct fluorination method is reported. The equipment is composed by a gaseous fluorine supply, a gas burette, a reactor tube inside a protective shield, a soda-lime chemical trap and a vacuum system. The fluorination is accomplished at a pressure of about 70 kg/cm 2 (1000 lb in 2 ), using gaseous fluorine. (Author). 5 refs, 4 figs, 2 tabs

  9. The effect of mixed oxidants and powdered activated carbon on the removal of natural organic matter.

    Science.gov (United States)

    Alvarez-Uriarte, Jon I; Iriarte-Velasco, Unai; Chimeno-Alanís, Noemí; González-Velasco, Juan R

    2010-09-15

    Present paper studies the influence of electrochemically generated mixed oxidants on the physicochemical properties of natural organic matter, and especially from the disinfection by-products formation point of view. The study was carried out in a full scale water treatment plant. Results indicate that mixed oxidants favor humic to non-humic conversion of natural organic matter. Primary treatment preferentially removes the more hydrophobic fraction. This converted the non-humic fraction in an important source of disinfection by-products with a 20% contribution to the final trihalomethane formation potential (THMFP(F)) of the finished water. Enhanced coagulation at 40 mg l(-1) of polyaluminium chloride with a moderate mixing intensity (80 rpm) and pH of 6.0 units doubled the removal efficiency of THMFP(F) achieved at full scale plant. However, gel permeation chromatography data revealed that low molecular weight fractions were still hardly removed. Addition of small amounts of powdered activated carbon, 50 mg l(-1), allowed reduction of coagulant dose by 50% whereas removal of THMFP(F) was maintained or even increased. In systems where mixed oxidants are used addition of powdered activated carbon allows complementary benefits by a further reduction in the THMFP(F) compared to the conventional only coagulation-flocculation-settling process. Copyright 2010 Elsevier B.V. All rights reserved.

  10. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

    Developments in the Australian uranium industry during 1980 are reviewed. Mine production increased markedly to 1841 t U 3 O 8 because of output from the new concentrator at Nabarlek and 1131 t of U 3 O 8 were exported at a nominal value of $37.19/lb. Several new contracts were signed for the sale of yellowcake from Ranger and Nabarlek Mines. Other developments include the decision by the joint venturers in the Olympic Dam Project to sink an exploration shaft and the release of an environmental impact statement for the Honeymoon deposit. Uranium exploration expenditure increased in 1980 and additions were made to Australia's demonstrated economic uranium resources. A world review is included

  11. Uranium

    International Nuclear Information System (INIS)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-01-01

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U 3 O 8 ; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables

  12. Thorium effect on the oxidation of uranium: Photoelectron spectroscopy (XPS/UPS) and cyclic voltammetry (CV) investigation on (U{sub 1−x}Th{sub x})O{sub 2} (x = 0 to 1) thin films

    Energy Technology Data Exchange (ETDEWEB)

    Cakir, P., E-mail: pelincakir@outlook.com [European Commission, Joint Research Centre, P.O. Box 2340, D-76125, Karlsruhe (Germany); Department of Radiation Science and Technology, Delft University of Technology, Mekelweg 15, 2629, JB Delft (Netherlands); Eloirdi, R.; Huber, F. [European Commission, Joint Research Centre, P.O. Box 2340, D-76125, Karlsruhe (Germany); Konings, R.J.M. [European Commission, Joint Research Centre, P.O. Box 2340, D-76125, Karlsruhe (Germany); Department of Radiation Science and Technology, Delft University of Technology, Mekelweg 15, 2629, JB Delft (Netherlands); Gouder, T. [European Commission, Joint Research Centre, P.O. Box 2340, D-76125, Karlsruhe (Germany)

    2017-01-30

    Highlights: • XRD and XPS data of U{sub x}Th{sub 1-x}O{sub 2} films are in agreement with data obtained on bulk. • Oxygen affinity of thorium is much stronger than uranium. • Oxidation of uranium decreases as a function of thorium in the matrix. • XPS made pre and post CV shows thorium enrichment indicating a protective layer. • Higher initial uranium content is directly proportional to higher oxidation states. - Abstract: Thin films of U{sub 1−x}Th{sub x}O{sub 2} (x = 0 to 1) have been deposited via reactive DC sputter technique and characterized by X-ray/Ultra-violet Photoelectron Spectroscopy (XPS/UPS), X-ray Powder Diffractometer (XRD) and Cyclic Voltammetry (CV) in order to understand the effect of Thorium on the oxidation mechanism. During the deposition, the competition between uranium and thorium for oxidation showed that thorium has a much higher affinity for oxygen. Deposition conditions, time and temperature were also the subject of this study, to look at the homogeneity and the stability of the films. While core level and valence band spectra were not altered by the time of deposition, temperature was affecting the oxidation state of uranium and the valence band due to the mobility increase of oxygen through the film. X-ray diffraction patterns, core level spectra obtained for U{sub 1−x}Th{sub x}O{sub 2} versus the composition showed that lattice parameters follow the Vegard's law and together with the binding energies of U-4f and Th-4f are in good agreement with literature data obtained on bulk compounds. To study the effect of thorium on the oxidation of U{sub 1−x}Th{sub x}O{sub 2} films, we used CV experiments at neutral pH of a NaCl solution in contact with air. The results indicated that thorium has an effect on the uranium oxidation as demonstrated by the decrease of the current of the oxidation peak of uranium. XPS measurements made before and after the CV, showed a relative enrichment of thorium at the extent of uranium at

  13. Plasma decontamination of uranium oxide from stainless steel surfaces

    International Nuclear Information System (INIS)

    Veilleux, J.M.; El-Genk, M.S.; Chamberlin, E.P.

    1997-01-01

    The U.S. Department of Energy (DOE) is expected to have 845000 m 3 of transuranic (TRU) waste by the year 2000 that has accumulated during the development and assembly of the nation's nuclear stockpile. The TRU disposal costs alone range up to $28000/m 3 , which could be reduced to $1800/m 3 or less by treating and converting the material to low-level waste. Plasma-based processes have been shown to remove plutonium and uranium surface contaminants from metallic components and could be used to treat TRU with significant cost avoidance, estimated at over $1.0 billion. Martz and Hess conducted the initial work of plutonium etching in low-power radio-frequency (rf) plasma with etch rates ranging from 0.007 to 0.025 kg/m 2 ·h. Veilleux et al. reported that plasma decontamination of uranium from the interior of aluminum objects results in etch rates an order of magnitude greater. The current work reports on removal rates of uranium from stainless steel surfaces and includes estimates of the etch rates and characteristic times for removal

  14. Induction of oxidative stress related responses in Arabidopsis thaliana after uranium exposure

    International Nuclear Information System (INIS)

    Vanhoudt, N.

    2009-01-01

    Uranium contamination in the environment has resulted from releases linked with nuclear fuel cycle activities and from industries extracting and processing materials containing naturally occurring radionuclides (for example phosphate industry). Uranium toxicity effects are predominantly studied on man and animal species, but little information is available for plants. If phytomanagement of uranium contaminated soils is considered, biological effects on the vegetation have to be investigated. Information on the contamination impact can also be used for risk assessment and derivation of clean-up standards. Plants can experience oxidative stress when they are exposed to environmental stress situations (for example exposure to heavy metals). Reactive oxygen species (ROS) are produced in both stressed and unstressed cells potentially leading to cellular damage. Consequently, plants have developed an antioxidative defence system comprising ROS-scavenging enzymes (e.g. SOD (superoxide dismutase), CAT (catalase) and metabolites (e.g. ascorbate, glutathione). Previous results showed that uranium exposure can cause an imbalance between the oxidative and antioxidative capacities of the plant cells. The present study aimed to analyse biological effects induced in Arabidopsis thaliana after bioaccumulation of uranium and to define possible dose-effect relationships. Subtle effects on the antioxidative defence system (enzymes, metabolites viewed as early responses for individual disturbances (growth, nutrient profile) were analysed

  15. Uranium metal oxidation, grinding, and encapsulation in BorobondR: TRU waste management - 59279

    International Nuclear Information System (INIS)

    Cook, Kevin S.; Addington, Larry A.; Utley, Beth

    2012-01-01

    Hydrogen generation mitigation for K Basin sludge was examined by encapsulation of uranium metal in BoroBond R , pre-oxidation of uranium metal with Fenton's reagent and grinding of Densalloy SD170, an irradiated uranium metal surrogate. Encapsulation in BoroBond R resulted in pressure increase rates at 60 deg. C ranging from 0.116 torr/h to 0.186 torr/h compared to 0.240 torr/h for a uranium metal in water standard. Samples cast with higher water content led to increased rates. A Fenton's reagent system consisting of a simple reagent mix of FeSO 4 .7H 2 O, H 2 O 2 and HCl effectively oxidized 1/4'' cubes of uranium metal in under four days at room temperature. Increased peroxide addition rate, increased FeSO 4 .7H 2 O concentration and low pH all increase the corrosion rate. Densalloy SD170 with an average particle size of 581 μm with 7.63 % of particles less than 90 μm was milled so that over 90 % of the Densalloy mass measured less than 90 μm in 6 hours of milling. Acceptable wear rates were seen on wear components that were from standard materials (Nitronic SS and 440SS). (authors)

  16. Comparative study of the oxidation of various qualities of uranium in carbon dioxide at high temperatures

    International Nuclear Information System (INIS)

    Desrues, R.; Paidassi, J.

    1965-01-01

    Uranium samples of six different qualities were subjected, in the temperature range 400 - 1000 C, to the action of carbon dioxide carefully purified to eliminate oxygen and water vapour; the resulting oxidation was followed micro-graphically and also (but only in the range 400 - 700 C) gravimetrically using an Ugine-Eyraud microbalance. A comparison of the results leads to the following 3 observations. First, the oxidation of the six uraniums studied obeys a linear law, (followed at 700 C by an accelerating law). The rates of reaction differ by a maximum of 100 per cent, the higher purity grades being oxidized more slowly except at 700 C when the reverse is true. Secondly, simultaneously with the growth, of an approximately uniform film of uranium dioxide on the metal, there occurs a localized attack in the form of blisters in the immediate neighbourhood of the monocarbide inclusions in the uranium. The relative importance of this attack is greater for lower oxidation temperatures and for a larger size, number and inequality of distribution of the inclusions, that is to say for higher carbon concentrations in the uranium (which have values from 7 to 1000 ppm in our tests). Thirdly, for oxidation temperatures above 600 C blistering is much less pronounced, but at 700 C the beginning of a general deformation of the sample occurs, which, above 750 C, becomes much greater; this leads to an acceleration of the reaction rate with respect to the linear law. In view of the over-heating, the sample must already be in the γ-phase which is particularly easily deformed; furthermore this expansion phenomenon is more pronounced when the sample is more plastic and therefore purer. (authors) [fr

  17. Fabrication of dendritic silver-coated copper powders by galvanic displacement reaction and their thermal stability against oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yu-Seon [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Farad Materials Co., Ltd., Suwon 16419 (Korea, Republic of); An, Chang Yong; Kannan, Padmanathan Karthick; Seo, Nary [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Zhuo, Kai [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Farad Materials Co., Ltd., Suwon 16419 (Korea, Republic of); Yoo, Tae Kyong [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Chung, Chan-Hwa, E-mail: chchung@skku.edu [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Farad Materials Co., Ltd., Suwon 16419 (Korea, Republic of)

    2016-12-15

    Highlights: • The dendritic silver-coated copper powders with high specific surface area have been prepared using a simple wet chemical reduction process at room temperature. • It is found that the Cu starts to be oxidized into Cu{sub 2}O followed by CuO at elevated temperatures. • The more amount of Ag-coating provides the less oxidation, which confirms that the Ag-shell prevents the Cu-core from oxidation. • The resistivity of dendritic 33.27 wt.% Ag-coated Cu powders was measured to 25.67 μΩ cm after the annealing at 150 °C for 30 min. - Abstract: Two steps of wet chemical processes have been developed for the preparation of core-shell nanostructures of copper and silver, which is a facile and low cost method for the production of large quantity of dendritic powders. First step involves a galvanic displacement reaction with hydrogen evolution which is the motive force of spontaneous electrochemical reaction. To achieve the core-shell structure, silver has been coated on the dendritic copper using the galvanic displacement reaction. The dendritic silver-coated copper powders exhibit high surface-area, excellent conductivity, and good oxidation resistance. It has been found that silver-coated copper powders maintain the electrical conductivity even after annealing at 150 °C for several to tens of minutes, thus it is a promising material and an alternative to pure silver powders in printed electronics application.

  18. Fabrication of dendritic silver-coated copper powders by galvanic displacement reaction and their thermal stability against oxidation

    International Nuclear Information System (INIS)

    Park, Yu-Seon; An, Chang Yong; Kannan, Padmanathan Karthick; Seo, Nary; Zhuo, Kai; Yoo, Tae Kyong; Chung, Chan-Hwa

    2016-01-01

    Highlights: • The dendritic silver-coated copper powders with high specific surface area have been prepared using a simple wet chemical reduction process at room temperature. • It is found that the Cu starts to be oxidized into Cu_2O followed by CuO at elevated temperatures. • The more amount of Ag-coating provides the less oxidation, which confirms that the Ag-shell prevents the Cu-core from oxidation. • The resistivity of dendritic 33.27 wt.% Ag-coated Cu powders was measured to 25.67 μΩ cm after the annealing at 150 °C for 30 min. - Abstract: Two steps of wet chemical processes have been developed for the preparation of core-shell nanostructures of copper and silver, which is a facile and low cost method for the production of large quantity of dendritic powders. First step involves a galvanic displacement reaction with hydrogen evolution which is the motive force of spontaneous electrochemical reaction. To achieve the core-shell structure, silver has been coated on the dendritic copper using the galvanic displacement reaction. The dendritic silver-coated copper powders exhibit high surface-area, excellent conductivity, and good oxidation resistance. It has been found that silver-coated copper powders maintain the electrical conductivity even after annealing at 150 °C for several to tens of minutes, thus it is a promising material and an alternative to pure silver powders in printed electronics application.

  19. Calibration Tools for Measurement of Highly Enriched Uranium in Oxide and Mixed Uranium-Plutonium Oxide with a Passive-Active Neutron Drum Shuffler

    International Nuclear Information System (INIS)

    Mount, M; O'Connell, W; Cochran, C; Rinard, P

    2003-01-01

    Lawrence Livermore National Laboratory (LLNL) has completed an extensive effort to calibrate the LLNL passive-active neutron drum (PAN) shuffler (Canberra Model JCC-92) for accountability measurement of highly enriched uranium (HEU) oxide and HEU in mixed uranium-plutonium (U-Pu) oxide. Earlier papers described the PAN shuffler calibration over a range of item properties by standards measurements and an extensive series of detailed simulation calculations. With a single normalization factor, the simulations agree with the HEU oxide standards measurements to within ±1.2% at one standard deviation. Measurement errors on mixed U-Pu oxide samples are in the ±2% to ±10% range, or ±20 g for the smaller items. The purpose of this paper is to facilitate transfer of the LLNL procedure and calibration algorithms to external users who possess an identical, or equivalent, PAN shuffler. Steps include (1) measurement of HEU standards or working reference materials (WRMs); (2) MCNP simulation calculations for the standards or WRMs and a range of possible masses in the same containers; (3) a normalization of the calibration algorithms using the standard or WRM measurements to account for differences in the 252 Cf source strength, the delayed-neutron nuclear data, effects of the irradiation protocol, and detector efficiency; and (4) a verification of the simulation series trends against like LLNL results. Tools include EXCEL/Visual Basic programs which pre- and post-process the simulations, control the normalization, and embody the calibration algorithms

  20. Elemental and Isotopic Analysis of Uranium Oxide an NIST Glass Standards by FEMTOSECOND-LA-ICP-MIC-MS

    International Nuclear Information System (INIS)

    Ebert, Chris; Zamzow, Daniel S.; McBay, Eddie H.; Bostick, Debra A.; Bajic, Stanley J.; Baldwin, David P.; Houk, R.S.

    2009-01-01

    The objective of this work was to test and demonstrate the analytical figures of merit of a femtosecond-laser ablation (fs-LA) system coupled with an inductively coupled plasma-multi-ion collector-mass spectrometer (ICP-MIC-MS). The mobile fs-LA sampling system was designed and assembled at Ames Laboratory and shipped to Oak Ridge National Laboratory (ORNL), where it was integrated with an ICP-MIC-MS. The test period of the integrated systems was February 2-6, 2009. Spatially-resolved analysis of particulate samples is accomplished by 100-shot laser ablation using a fs-pulsewidth laser and monitoring selected isotopes in the resulting ICP-MS transient signal. The capability of performing high sensitivity, spatially resolved, isotopic analyses with high accuracy and precision and with virtually no sample preparation makes fs-LA-ICP-MIC-MS valuable for the measurement of actinide isotopes at low concentrations in very small samples for nonproliferation purposes. Femtosecond-LA has been shown to generate particles from the sample that are more representative of the bulk composition, thereby minimizing weaknesses encountered in previous work using nanosecond-LA (ns-LA). The improvement of fs- over ns-LA sampling arises from the different mechanisms for transfer of energy into the sample in these two laser pulse-length regimes. The shorter duration fs-LA pulses induce less heating and cause less damage to the sample than the longer ns pulses. This results in better stoichiometric sampling (i.e., a closer correlation between the composition of the ablated particles and that of the original solid sample), which improves accuracy for both intra- and inter-elemental analysis. The primary samples analyzed in this work are (a) solid uranium oxide powdered samples having different 235 U to 238 U concentration ratios, and (b) glass reference materials (NIST 610, 612, 614, and 616). Solid uranium oxide samples containing 235 U in depleted, natural, and enriched abundances were

  1. Confirmation test of powder mixing process in J-MOX

    International Nuclear Information System (INIS)

    Ota, Hiroshi; Osaka, Shuichi; Kurita, Ichiro

    2009-01-01

    Japan Nuclear Fuel Ltd. (hereafter, JNFL) MOX Fuel Fabrication Plant (hereafter, J-MOX) is what fabricates MOX fuel for domestic light water power plants. Development of design concept of J-MOX was started mid 90's and the frame of J-MOX process was clarified around 2000 including adoption of MIMAS process as apart of J-MOX powder process. JNFL requires to take an answer to any technical question that has not been clarified ever before by world's MOX and/or Uranium fabricators before it commissions equipment procurement. J-MOX is to be constructed adjacent to the Rokkasho Reprocessing Plant (RRP) and to utilize MH-MOX powder recovered at RRP. The combination of the MIMAS process and the MH-MOX powder is what has never tried in the world. Therefore JNFL started a series of confirmation tests of which the most important is the powder test to confirm the applicability of MH-MOX powder to the MIMAS process. The MH-MOX powder, consisting of 50% plutonium oxide and 50% uranium oxide, originates JAEA development utilizing microwave heating (MH) technology. The powder test started with laboratory scale small equipment utilizing both uranium and the MOX powder in 2000, left a solution to tough problem such as powder adhesion onto equipment, and then was followed by a large scale equipment test again with uranium and the MOX powder. For the MOX test, actual size equipment within glovebox was manufactured and installed in JAEA plutonium fuel center in 2005, and based on results taken so far an understanding that the MIMAS equipment, with the MH-MOX powder, can present almost same quality MOX pellet as what is introduced as fabricated in Europe was developed. The test was finished at the end of Japanese fiscal year (JFY) 2007, and it was confirmed that the MOX pellets fabricated in this test were almost satisfied with the targeted specifications set for domestic LWR MOX fuels. (author)

  2. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Recent decisions by the Australian Government will ensure a significant expansion of the uranium industry. Development at Roxby Downs may proceed and Ranger may fulfil two new contracts but the decision specifies that apart from Roxby Downs, no new mines should be approved. The ACTU maintains an anti-uranium policy but reaction to the decision from the trade union movement has been muted. The Australian Science and Technology Council (ASTEC) has been asked by the Government to conduct an inquiry into a number of issues relating to Australia's role in the nuclear fuel cycle. The inquiry will examine in particular Australia's nuclear safeguards arrangements and the adequacy of existing waste management technology. In two additional decisions the Government has dissociated itself from a study into the feasibility of establishing an enrichment operation and has abolished the Uranium Advisory Council. Although Australian reserves account for 20% of the total in the Western World, Australia accounts for a relatively minor proportion of the world's uranium production

  3. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The French Government has decided to freeze a substantial part of its nuclear power programme. Work has been halted on 18 reactors. This power programme is discussed, as well as the effect it has on the supply of uranium by South Africa

  4. Electrochemical reduction behavior of U3O8 powder in a LiCl molten salt

    International Nuclear Information System (INIS)

    Jeong, Sang Mun; Shin, Ho-Sup; Hong, Sun-Seok; Hur, Jin-Mok; Do, Jae Bum; Lee, Han Soo

    2010-01-01

    The reduction path of the U 3 O 8 powder vol-oxidized at 1200 deg. C has been determined by a series of electrochemical experiments in a 1 wt.% Li 2 O/LiCl molten salt. Various reaction intermediates are observed by during electrolysis of U 3 O 8 . The formation of the metallic uranium is caused from two different reduction paths, a direct reduction of uranium oxide and an electro-lithiothermic reduction. As the uranium oxide is converted to the metallic uranium, the lithium metal is more actively formed in the cathode basket. The reducibility of the rare earth oxides with the U 3 O 8 powder has been tested by constant voltage electrolysis. The results suggest the advanced vol-oxidation could lead to the enhancement in the reducibility of the rare earth fission products.

  5. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 2. of ISO 7097 describes procedures for determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with cerium(IV) and ISO 7097-1 uses a titration with potassium dichromate

  6. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 1. of ISO 7097 describes procedures for the determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with potassium dichromate and ISO 7097-2 uses a titration with cerium(IV)

  7. Formation of peripheral porosity regions around urania in zirconia-urania mixed oxide powder compact sintering

    International Nuclear Information System (INIS)

    Das, P.; Choudhury, R.

    1992-01-01

    Sintering studies of zirconia-urania mixed oxide powder compacts (in stages of 5% urania up to a maximum of 20% addition) were carried out at temperatures between 1000-1400deg C for various soaking periods. The formation of a peripheral porosity region around comparatively coarser urania particle was a characteristic feature in this mixed oxide sintered compact. At even a higher sintering temperature (1800deg C), where extensive solid solution formation takes place, this porosity region demarcates the solutionized particles from the host zirconia apparently acting as a discontinuity in the system. Relative shrinkage difference between the dissimilar particles probably contributes to the porosity regions around the minor second phase at a lower temperature while at higher temperature generation of 'Kirkendall porosity' may be responsible for such an effect. (orig.)

  8. Oxidation of 1-butene over uranium oxide (UO3)-antimony oxide (Sb2O3) catalysts

    NARCIS (Netherlands)

    Simons, T.; Houtman, P.N.; Schuit, G.C.A.

    1971-01-01

    The oxidative dehydrogenation of butene to butadiene over U-Sb catalysts was investigated. The presence of two compds., (UO2)Sb3O7 and Sb3U3O14, reported by Grasselli and Callahan (1969), was confirmed with (UO2)Sb3O7 being the actual catalyst. The reaction is first order in butene and zero order in

  9. Process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide

    International Nuclear Information System (INIS)

    Heremanns, R.H.; Vandersteene, J.J.

    1983-01-01

    The invention concerns a process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide in the form of PuO 2 . Mixed fuels consisting of uranium oxide and plutonium oxide are being used more and more. The plants which prepare these mixed fuels have around 5% of the total mass of fuels as fabrication residue, either as waste or scrap. In view of the high cost of plutonium, it has been attempted to recover this plutonium from the fabrication residues by a process having a purchase price lower than the price of plutonium. The problem is essentially to separate the plutonium, the uranium and the impurities. The residues are fluorinated, the UF 6 and PuF 6 obtained are separated by selective absorption of the PuF 6 on NaF at a temperature of at least 400 0 C, the complex obtained by this absorption is dissolved in nitric acid solution, the plutonium is precipitated in the form of plutonium oxalate by adding oxalic acid, and the precipitated plutonium oxalate is calcined

  10. High-temperature, Knudsen cell-mass spectroscopic studies on lanthanum oxide/uranium dioxide solid solutions

    International Nuclear Information System (INIS)

    Sunder, S.; McEachern, R.; LeBlanc, J.C.

    2001-01-01

    Knudsen cell-mass spectroscopic experiments were carried out with lanthanum oxide/uranium oxide solid solutions (1%, 2% and 5% (metal at.% basis)) to assess the volatilization characteristics of rare earths present in irradiated nuclear fuel. The oxidation state of each sample used was conditioned to the 'uranium dioxide stage' by heating in the Knudsen cell under an atmosphere of 10% CO 2 in CO. The mass spectra were analyzed to obtain the vapour pressures of the lanthanum and uranium species. It was found that the vapour pressure of lanthanum oxide follows Henry's law, i.e., its value is directly proportional to its concentration in the solid phase. Also, the vapour pressure of lanthanum oxide over the solid solution, after correction for its concentration in the solid phase, is similar to that of uranium dioxide. (authors)

  11. Effect of cooling rate on achieving thermodynamic equilibrium in uranium-plutonium mixed oxides

    Science.gov (United States)

    Vauchy, Romain; Belin, Renaud C.; Robisson, Anne-Charlotte; Hodaj, Fiqiri

    2016-02-01

    In situ X-ray diffraction was used to study the structural changes occurring in uranium-plutonium mixed oxides U1-yPuyO2-x with y = 0.15; 0.28 and 0.45 during cooling from 1773 K to room-temperature under He + 5% H2 atmosphere. We compare the fastest and slowest cooling rates allowed by our apparatus i.e. 2 K s-1 and 0.005 K s-1, respectively. The promptly cooled samples evidenced a phase separation whereas samples cooled slowly did not due to their complete oxidation in contact with the atmosphere during cooling. Besides the composition of the annealing gas mixture, the cooling rate plays a major role on the control of the Oxygen/Metal ratio (O/M) and then on the crystallographic properties of the U1-yPuyO2-x uranium-plutonium mixed oxides.

  12. Formation conditions and prospecting criteria for sandstone uranium deposit of interlayer oxidation type

    International Nuclear Information System (INIS)

    Huang Shijie

    1994-01-01

    This paper comprehensively analyses the geotectonic setting and favourable conditions, such as structure of the basin, sedimentary facies and paleogeography, geomorphology and climate, hydrodynamics and hydrogeochemistry, the development of interlayered oxidation etc, necessary for the formation of sandstone uranium deposit of interlayered oxidation type. The following prospecting criteria is proposed, namely: abundant uranium source, arid climate, stable big basin, flat-lying sandstone bed, big alluvial fan, little change in sedimentary facies, intercalation of sandstone and mudstone beds, shallow burying of sandstone bed, well-aquiferous sandstone bed, high permeability of sandstone bed, development of interlayered oxidation, and high content of reductant in sandstone. In addition, the 6 in 1 hydrogenic genetic model is proposed

  13. Uranium oxide nanocrystals by microwave-assisted thermal decomposition. Electronic and structural properties

    International Nuclear Information System (INIS)

    Leduc, Jennifer; Mathur, Sanjay; Pacold, Joseph I.; Shuh, David K.; Dong, Chung-Li

    2018-01-01

    Uranium oxides have attracted much attention not only in the context of nuclear energy generation but also for their application as pristine catalysts or as supports for other (transition metal) oxides and (precious) metals. Their propensity to adopt high coordination numbers and manifest multiple oxidation states (from +II to +VI) makes them attractive candidates for catalyzed transformation reactions. Herein, we report a new synthesis route to phase-pure, crystalline UO 2 nanoparticles via microwave-assisted decomposition of a molecular uranium(IV) precursor. The electronic structure and optical absorption properties of these nanocrystals were investigated using spectroscopic methods to evaluate their suitability for photo(electro)catalytic applications. (copyright 2018 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  14. Uranium oxide nanocrystals by microwave-assisted thermal decomposition. Electronic and structural properties

    Energy Technology Data Exchange (ETDEWEB)

    Leduc, Jennifer; Mathur, Sanjay [Institute of Inorganic Chemistry, University of Cologne (Germany); Pacold, Joseph I.; Shuh, David K. [Chemical Sciences Division, The Glenn T. Seaborg Center, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Dong, Chung-Li [Department of Physics, Tamkang University, Tamsui, Taiwan (China)

    2018-01-17

    Uranium oxides have attracted much attention not only in the context of nuclear energy generation but also for their application as pristine catalysts or as supports for other (transition metal) oxides and (precious) metals. Their propensity to adopt high coordination numbers and manifest multiple oxidation states (from +II to +VI) makes them attractive candidates for catalyzed transformation reactions. Herein, we report a new synthesis route to phase-pure, crystalline UO{sub 2} nanoparticles via microwave-assisted decomposition of a molecular uranium(IV) precursor. The electronic structure and optical absorption properties of these nanocrystals were investigated using spectroscopic methods to evaluate their suitability for photo(electro)catalytic applications. (copyright 2018 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  15. First principle active neutron coincidence counting measurements of uranium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Goddard, Braden, E-mail: goddard.braden@gmail.com [Nuclear Security Science and Policy Institute, Texas A and M University, College Station, Texas 77843 (United States); Charlton, William [Nuclear Security Science and Policy Institute, Texas A and M University, College Station, Texas 77843 (United States); Peerani, Paolo [European Commission, EC-JRC-ITU, Ispra (Italy)

    2014-03-01

    Uranium is present in most nuclear fuel cycle facilities ranging from uranium mines, enrichment plants, fuel fabrication facilities, nuclear reactors, and reprocessing plants. The isotopic, chemical, and geometric composition of uranium can vary significantly between these facilities, depending on the application and type of facility. Examples of this variation are: enrichments varying from depleted (∼0.2 wt% {sup 235}U) to high enriched (>20 wt% {sup 235}U); compositions consisting of U{sub 3}O{sub 8}, UO{sub 2}, UF{sub 6}, metallic, and ceramic forms; geometries ranging from plates, cans, and rods; and masses which can range from a 500 kg fuel assembly down to a few grams fuel pellet. Since {sup 235}U is a fissile material, it is routinely safeguarded in these facilities. Current techniques for quantifying the {sup 235}U mass in a sample include neutron coincidence counting. One of the main disadvantages of this technique is that it requires a known standard of representative geometry and composition for calibration, which opens up a pathway for potential erroneous declarations by the State and reduces the effectiveness of safeguards. In order to address this weakness, the authors have developed a neutron coincidence counting technique which uses the first principle point-model developed by Boehnel instead of the “known standard” method. This technique was primarily tested through simulations of 1000 g U{sub 3}O{sub 8} samples using the Monte Carlo N-Particle eXtended (MCNPX) code. The results of these simulations showed good agreement between the simulated and exact {sup 235}U sample masses.

  16. Influence of Turmeric Rhizome Powder diets on decreasing oxidative stress caused by heat stress inbroiler model

    Directory of Open Access Journals (Sweden)

    Seyyed Javad Hosseini-Vashan

    2012-08-01

    Full Text Available Background and Aim: Production of reactive oxygen species (ROS increases during oxidative stress conditions, which stimulates diabetes, inflammatory reactions, rheumatism and anemia. Some antioxidant properties of turmeric rhizome powder (TRP were revealed by previous researchers. The present study was conducted to evaluate the influence of TRP on decreasing effects of oxidative stress resulted from heat stress in broiler chickens.   Materials and Methods: In this experimental study, two-hundred-sixty-four 1-day-old broilers were divided into 3 dietary treatments. The dietary treatments involved 0(control, 0.4 and 0.8% turmeric rhizome powder (cases. In order to create oxidative stress, the ambient temperature was daily raised from 21 to 33oc for 5 hours (11a.m-4p.m throughout the 28th-42nd days. Blood lipids, Glutathione peroxidase (GPx ,superoxide dismutase (SOD, and Tiobarbituric acid reaction score (TBARS were determined at the end of the experiment.   Results: The results revealed that total cholesterol and triglyceride were not affected. The 0.4 TRP diet decreased blood LDL (46.7±3.01 compared to basal group (52.0±2.17. HDL increased in broilers fed 0.8% TRP (74.0 ± 3.87 compared to chickens with basal diet (63.7± 2.98. Enzyme activity of GPx improved in broilers fed TRP diets (225.9± 11.52 as compared to chickens with basal diet(183.1± 8.52 however, the TRP diet did not affect enzyme activity of SOD (P > 0.05. The TBARS index decreased in broilers fed TRP (0.76 ± 0.0052 in basal vs.0.49 ± 0.0032 in 0.8% TRP.   Conclusion: The major bioactive component of TRP is Curcumin that can improve the antioxidant properties under oxidative stress and high ambient temperature.

  17. Fabrication of TiC-TiO{sub 2} composite powders by thermal plasma oxidation of titanium carbide powder; Tanka chitan funmatsu no plasma sanka hanno ni yori seiseishita TiC-TiO{sub 2} fukugo funmatsu

    Energy Technology Data Exchange (ETDEWEB)

    Ishigaki, T.; Li, Y.; Haneda, H. [National institute for Research Inorganic materials, Tsukuba (Japan); Kataoka, E. [Showa Cabot Supermetals K.K., Tokyo (Japan)

    2000-09-15

    TiC-TiO{sub 2} composite powders were prepared by in-flight oxidation of titanium carbide powder in RF induction thermal plasmas. Original titanium carbide powder of 20 - 38 {mu}m in particle size was axially injected into the center of argon-oxygen plasma. The powders were partially spheroidized and evaporated through the plasma treatment. X-ray diffraction of plasma-treated powders showed the formation of titanium dioxides, both rutile and anatase phases. The phase content of the plasma-prepared powders strongly depended on the plasma conditions, such as the plasma generating pressure and the oxygen flow rate in plasma generating gas. Especially, the increase of oxygen flow rate in plasma gas gave rise to the increase of heat transfer from plasma to powder particles, exothermic heat of oxidation reaction and cooling rate of plasma, giving the increase of spheroidization ratio, formation ratio of titanium dioxides, and content of anatase phases. (author)

  18. Kinetic Study of Zn2+ and Cd2+ Ions Sorption by Ceric Oxide Powder

    International Nuclear Information System (INIS)

    Hassan, H.S.; Abd El-Rahman, K.M.; El Sayed, A.A.

    2008-01-01

    Ceric Oxide powder was chemically synthesized and characterized using infrared spectra and x-ray diffraction. The sorptive removal of Zinc and Cadmium ions from aqueous waste solution using synthetic ceric oxide powder was investigated using batch technique. Experiments were carried out as a function of ph, particle size, solute concentration and temperature. The uptake of zinc was found to be greater than that of cadmium. A comparison of kinetic models applied to the sorption process of each ion was evaluated for the pseudo first order, the pseudo second order, and homogeneous particle diffusion kinetic models, respectively. The results showed that both the pseudo second order and the homogeneous particle diffusion model (HPDM) were found to best correlate the experimental rate data. The numerical values of the rate constants and particle diffusion coefficients were determined from the graphical representation of the proposed models. Activation energy (Ε a ) and entropy (δ S * ) of activation for each sorption process were also calculated from the linearized form of Arrhenius equation

  19. Nanosized lead lanthanum titanate (PLT) ceramic powders synthesized by the oxidant peroxo method

    Energy Technology Data Exchange (ETDEWEB)

    Camargo, Emerson R. [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Department of Chemistry, UFSCar-Federal University of Sao Carlos, Rod.Washingtin Luis km 235, CP 676, Sao Carlos SP 13565-9905 (Brazil)], E-mail: camargo@ufscar.br; Barrado, Cristiano M. [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Department of Chemistry, UFSCar-Federal University of Sao Carlos, Rod.Washingtin Luis km 235, CP 676, Sao Carlos SP 13565-9905 (Brazil); Ribeiro, Caue [EMBRAPA Instrumentacao Agropecuaria, Rua XV de Novembro 1452, Sao Carlos SP 13560-970 (Brazil)], E-mail: caue@cnpdia.embrapa.br; Longo, Elson [Department of Biochemistry, Chemistry Institute of Araraquara, UNESP-Sao Paulo State University, Rua Francisco Degni, CP 355, Araraquara SP 14801-907 (Brazil)], E-mail: elson@iq.unesp.br; Leite, Edson R. [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Department of Chemistry, UFSCar-Federal University of Sao Carlos, Rod.Washingtin Luis km 235, CP 676, Sao Carlos SP 13565-9905 (Brazil)], E-mail: derl@power.ufscar.br

    2009-05-05

    For the first time it is reported the synthesis of lead titanate modified with rare earth by the oxidant-peroxo method (OPM). Lanthanum was added up to 20% in mol through the dissolution of lanthanum oxide in nitric acid, followed by the addition of a solution of lead and lanthanum nitrate into an aqueous solution of titanium peroxo complexes. The amorphous precipitate formed was heat-treated at different temperatures in the range from 400 to 900 deg. C for crystallization. Powders were characterized by Raman spectroscopy and X-ray diffraction. Tetragonal perovskite structure was observed for the samples up to 15% of lanthanum substitution and cubic perovskite for sample with 20% of lanthanum. Crystallographic domains calculated by Scherrer equation showing a probable suppression of the crystallite growth in function of lanthanum content. It was observed shifting to lower frequencies of Raman modes in the range between 100 and 400 cm{sup -1} and the vanishing of the A1(2TO) and E(1LO) modes could be attributed to transition phase from tetragonal to cubic. Electronic microscopy image revealed that the powders annealed at height temperature are spherical with sharp size distribution.

  20. Nanosized lead lanthanum titanate (PLT) ceramic powders synthesized by the oxidant peroxo method

    International Nuclear Information System (INIS)

    Camargo, Emerson R.; Barrado, Cristiano M.; Ribeiro, Caue; Longo, Elson; Leite, Edson R.

    2009-01-01

    For the first time it is reported the synthesis of lead titanate modified with rare earth by the oxidant-peroxo method (OPM). Lanthanum was added up to 20% in mol through the dissolution of lanthanum oxide in nitric acid, followed by the addition of a solution of lead and lanthanum nitrate into an aqueous solution of titanium peroxo complexes. The amorphous precipitate formed was heat-treated at different temperatures in the range from 400 to 900 deg. C for crystallization. Powders were characterized by Raman spectroscopy and X-ray diffraction. Tetragonal perovskite structure was observed for the samples up to 15% of lanthanum substitution and cubic perovskite for sample with 20% of lanthanum. Crystallographic domains calculated by Scherrer equation showing a probable suppression of the crystallite growth in function of lanthanum content. It was observed shifting to lower frequencies of Raman modes in the range between 100 and 400 cm -1 and the vanishing of the A1(2TO) and E(1LO) modes could be attributed to transition phase from tetragonal to cubic. Electronic microscopy image revealed that the powders annealed at height temperature are spherical with sharp size distribution.

  1. Kinetic study of Cs+ and Eu3+ ions sorption by zirconium oxide powder

    International Nuclear Information System (INIS)

    Hanafi, H.A.; Hassan, H.S.; Hamed, M.M.

    2009-01-01

    Full text: Zirconium oxide powder was chemically synthesized by sol-gel method and characterized using infrared spectra and x-ray diffraction. The sorptive removal of cesium and europium ions from aqueous waste solution using synthetic zirconium oxide powder was investigated using batch technique. Experiments were carried out as a function of pH, time and temperature. The uptake of europium was found to be greater than that of cesium. A comparison of kinetic models applied to the sorption process of each ion was evaluated for the pseudo first order, the pseudo second order, and homogeneous particle diffusion kinetic models, respectively. The results showed that both the pseudo second order and the homogeneous particle diffusion models (HPDM) were found to best correlate the experimental rate data. The numerical values of the rate constants and particle diffusion coefficients were determined from the graphical representation of the proposed models. Activation energy (Ea) and entropy (Δ S*) of activation for each sorption process were also calculated from the linearized form of Arrhenius equation. (author)

  2. Analysis of nuclear grade uranium oxides by atomic absorption spectrometry with electrothermal atomization

    International Nuclear Information System (INIS)

    Batistoni, D.A.; Erlijman, L.H.; Pazos, A.L.

    1986-01-01

    The application of atomic absorption spectrometry for the determination of five trace impurities in nuclear grade uranium oxides is described. The elements were separated from the uranium matrix by extraction chromatography and determined in 5.5 M nitric acid by electrothermal atomization using pyrolytic graphite coated tubes. Two elements, cadmium and chromium, with different volatility characteristics were employed to investigate the operating conditions. Drying and ashing conditions were studied for both elements. Ramp and constant potential (step) heating modes have also been studied and compared. Good reproducibility and a longer life of graphite tubes were obtained with ramp atomization. Detection limits (in micrograms per gram of uranium) were: Cd 0.01; Cr 0.1; Cu 0.4; Mn 0.04 and Ni 0.2. (author) [es

  3. Does uranium exposure induce oxidative stress and genotoxicity in the teleostean Danio rerio? first experimental results

    International Nuclear Information System (INIS)

    Barillet, S.; Devaux, A.; Simon, O.; Buet, A.; Pradines, C.

    2004-01-01

    Within the Envirhom research program, key advances have been obtained in uranium bioaccumulation and underlying mechanisms understanding in various biological models at the individual level. However, considering different scales of biological effects (from early to delayed ones, from low to high level of organization) is crucial to provide ecologically relevant indicators. Organisms counteract stress induced by pollutant exposure through a wide range of physiological responses being both dose and time dependent. Effects at higher hierarchical levels are always preceded by early changes in biological processes, from subtle biochemical disturbances to impaired physiological functions, increased susceptibility to other stresses, reduced life-span Within this global context, preliminary experiments were carried out on adult zebra fish (Danio rerio), to assess early changes after short-term uranium exposure. Among the subsequent primary subcellular damages oxidative stress and genotoxicity (characterizing both chemo-toxicity and radiotoxicity) are relevant endpoints, thus requiring the knowledge of dose-effects relationships as a first operational approach to provide useful tool in predicting possible effects of U exposure. Zebra fish has been selected due to its small size (facilitating its maintenance) and its extended use in eco-toxicological studies. Moreover, its short life-cycle will allow to carry out chronic exposure experiments (along the whole life-cycle). Four uranium concentrations (0, 20, 100 and 500μg.L -1 ) and five sampling times (0, 0.5, 1, 5 and 10 days) were selected. Catalase, glutathione peroxidase (GPx) and superoxide dismutase (SOD) activities were measured as oxidative stress bio-markers. DNA damage level was assessed in zebra fish erythrocytes using the comet assay. Uranium bioaccumulation was concurrently studied to understand observed bio-marker responses. Further experiments, dedicated to the assessment of the impact of chronic uranium

  4. Ianthinite: A rare hydrous uranium oxide mineral from Akkavaram ...

    Indian Academy of Sciences (India)

    In the area of ianthinite occurrence (figure 1), the basement ... Akkavaram area, from where ianthinite is reported. (figure 1), falls in ... feldspar grains from separated yellowish material .... oxidizing environment and its relevance to the disposal.

  5. Dissolution of synthetic uranium dibutyl phosphate deposits in oxidizing and reducing chemical formulations

    International Nuclear Information System (INIS)

    Rufus, A.L.; Sathyaseelan, V.S.; Narasimhan, S.V.; Velmurugan, S.

    2013-01-01

    Graphical abstract: SEM of the U-DBP coated stainless steel coupon before and after exposure to chemical formulation containing acid permanganate at 80 °C. -- Highlights: •Combination of oxidation and reduction processes efficiently dissolves U-DBP deposits. •NP and NAC formulations are compatible with SS-304. •Dissolved uranium and added chemicals are effectively removed via ion exchangers. -- Abstract: Permanganate and nitrilotriacetic acid (NTA) based dilute chemical formulations were evaluated for the dissolution of uranium dibutyl phosphate (U-DBP), a compound that deposits over the surfaces of nuclear reprocessing plants and waste storage tanks. A combination of an acidic, oxidizing treatment (nitric acid with permanganate) followed by reducing treatment (NTA based formulation) efficiently dissolved the U-DBP deposits. The dissolution isotherm of U-DBP in its as precipitated form followed a logarithmic fit. The same chemical treatment was also effective in dissolving U-DBP coated on the surface of 304-stainless steel, while resulting in minimal corrosion of the stainless steel substrate material. Investigation of uranium recovery from the resulting decontamination solutions by ion exchange with a bed of mixed anion and cation resins showed quantitative removal of uranium

  6. Dissolution of synthetic uranium dibutyl phosphate deposits in oxidizing and reducing chemical formulations

    Energy Technology Data Exchange (ETDEWEB)

    Rufus, A.L.; Sathyaseelan, V.S.; Narasimhan, S.V.; Velmurugan, S., E-mail: svelu@igcar.gov.in

    2013-06-15

    Graphical abstract: SEM of the U-DBP coated stainless steel coupon before and after exposure to chemical formulation containing acid permanganate at 80 °C. -- Highlights: •Combination of oxidation and reduction processes efficiently dissolves U-DBP deposits. •NP and NAC formulations are compatible with SS-304. •Dissolved uranium and added chemicals are effectively removed via ion exchangers. -- Abstract: Permanganate and nitrilotriacetic acid (NTA) based dilute chemical formulations were evaluated for the dissolution of uranium dibutyl phosphate (U-DBP), a compound that deposits over the surfaces of nuclear reprocessing plants and waste storage tanks. A combination of an acidic, oxidizing treatment (nitric acid with permanganate) followed by reducing treatment (NTA based formulation) efficiently dissolved the U-DBP deposits. The dissolution isotherm of U-DBP in its as precipitated form followed a logarithmic fit. The same chemical treatment was also effective in dissolving U-DBP coated on the surface of 304-stainless steel, while resulting in minimal corrosion of the stainless steel substrate material. Investigation of uranium recovery from the resulting decontamination solutions by ion exchange with a bed of mixed anion and cation resins showed quantitative removal of uranium.

  7. Sodium oxide and uranium oxide aerosol experiments: NSPP Tests 106-108 and Tests 204-207, data record report

    Energy Technology Data Exchange (ETDEWEB)

    Adams, R.E.; Kress, T.S.; Tobias, M.L.

    1981-03-01

    This data record report describes three sodium oxide aerosol tests and four uranium oxide aerosol tests conducted in the Nuclear Safety Pilot Plant project at Oak Ridge National Laboratory. The goal of this project is to establish the validity (or level of conservatism) of the aerosol behavioral code, HAARM-3, and follow-on codes under development at the Battelle Columbus Laboratories for the US Nuclear Regulatory Commission. Descriptions of the seven tests with tables and graphs summarizing the results are included. 92 figs.

  8. Remote sensing technology prospecting methods of interlayer oxidation zone type sandstone uranium deposit in Yili basin

    International Nuclear Information System (INIS)

    Huang Xianfang; Huang Shutao; Pan Wei; Feng Jie; Liu Dechang; Zhang Jingbo; Xuan Yanxiu; Rui Benshan

    1998-12-01

    Taking Yili Basin as an example, remote sensing technology and method of interlayer oxidation zone type sandstone uranium deposit have systematically been summarized. Firstly, principle, methods and procedures of the second development of scientific experimental satellite photograph have been elaborated in detail. Three dimensional stereo simulation, display, and multi-parameters extraction have been recommended. Secondarily, the research is focused on prospective section image features in different type images and their geological implications and on establishing recognition keys of promising areas. Finally, based on above research results, three graded predictions, i.e. regional prospect, promising sections and favourable location in the deposit have been made step by step and reconnaissance and prospecting range are gradually reduced. The practice has indicated that breakthrough progress has been made in application to prospect prognosis of interlayer oxidation zone type sandstone uranium deposit and good verified results have been obtained

  9. Comparison of two lung clearance models based on the dissolution rates of oxidized depleted uranium

    International Nuclear Information System (INIS)

    Crist, K.C.

    1984-10-01

    An in-vitro dissolution study was conducted on two respirable oxidized depleted uranium samples. The dissolution rates generated from this study were then utilized in the International Commission on Radiological Protection Task Group lung clearance model and a lung clearance model proposed by Cuddihy. Predictions from both models based on the dissolution rates of the amount of oxidized depleted uranium that would be cleared to blood from the pulmonary region following an inhalation exposure were compared. It was found that the predictions made by both models differed considerably. The difference between the predictions was attributed to the differences in the way each model perceives the clearance from the pulmonary region. 33 references, 11 figures, 9 tables

  10. Study about uranium oxides at high temperature by X-ray diffraction

    International Nuclear Information System (INIS)

    Costa, M.I.

    1978-01-01

    In this work a technique to study the lattice parameters in the crystalline substances at hight temperature by X-rays diffraction is developed. The results obtained agree very well with the experimental data found in the literature. The crystalline structure of uranium oxide at different temperature is studied in detail by this technique. At the range of the temperature investigated, i.e., 20 0 C to 640 0 C, the following forms for uranium oxide: U 3 O 8 in its hexagonal modification, cubic UO 2 , cubic U 4 O 9 and tetragonal U 3 O 7 is observed. The appearance of two hexagonal units observed in this work is identified by Milne. A good reproducibillity is observed for measurements at the same temperature [pt

  11. Study on metallogenetic prospect of interlayer oxidation zone sandstone type uranium deposit in Shanganning basin

    International Nuclear Information System (INIS)

    Wang Jinping

    1998-01-01

    As Compared with orogenic zone basin, which the interlayer oxidation zone sandstone type uranium deposits are found, the Shanganning basin a continental platform type basin is distinct either in the geodynamic background and the post-basin hydrogeological evolution or in the appearance of the metallogenetic dynamics-orogenesis. The prediction criteria summarized for interlayer oxidation zone type U-deposits in Middle Asia therefore can not be completely applied in such a basin. Based on analysis of the typical regional geological setting, the hydrogeology of the Meso-Cenozoic cover is studied in detail. Three hydrogeological cycles have been divided, and prospects of uranium deposits have been clarified and the most promising target have been proposed

  12. Comparison of two lung clearance models based on the dissolution rates of oxidized depleted uranium

    Energy Technology Data Exchange (ETDEWEB)

    Crist, K.C.

    1984-10-01

    An in-vitro dissolution study was conducted on two respirable oxidized depleted uranium samples. The dissolution rates generated from this study were then utilized in the International Commission on Radiological Protection Task Group lung clearance model and a lung clearance model proposed by Cuddihy. Predictions from both models based on the dissolution rates of the amount of oxidized depleted uranium that would be cleared to blood from the pulmonary region following an inhalation exposure were compared. It was found that the predictions made by both models differed considerably. The difference between the predictions was attributed to the differences in the way each model perceives the clearance from the pulmonary region. 33 references, 11 figures, 9 tables.

  13. Uranium and main oxides in soil in the Northeast part of Parana basil

    International Nuclear Information System (INIS)

    Fagundes, I.C.; Bonotto, D.M.; Jimenez-Rueda, J.R.

    2015-01-01

    Uranium is a litophile element that migrates to crust together with other light silicates. This work evaluated the relationships among the concentrations of uranium, major oxides (SiO_2, Al_2O_3, Na_2O, K_2O, CaO, MgO, Fe_2O_3, MnO, TiO_2, P_2O_5) and organic matter in different horizons of a soil profile located over siltstone from Tatui Formation, Piracicaba River sub-basin, Sao Paulo State, Brazil. Different techniques were utilized for data acquisition, for instance, alpha spectrometry, X-ray fluorescence spectrometry and smectrophotometry. The major U concentrations were found in horizons enriched in Fe_2O_3, indicating its tending to be retained in iron oxides. (author)

  14. Reaction of Antimony-Uranium Composite Oxide in the Chlorination Treatment of Waste Catalyst - 13521

    Energy Technology Data Exchange (ETDEWEB)

    Sawada, Kayo [EcoTopia Science Institute (Japan); Hirabayashi, Daisuke; Enokida, Youichi [Department of Materials, Physics and Energy Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan)

    2013-07-01

    The effect of oxygen gas concentration on the chlorination treatment of antimony-uranium composite oxide catalyst waste was investigated by adding different concentrations of oxygen at 0-6 vol% to its chlorination agent of 0.6 or 6 vol% hydrogen chloride gas at 1173 K. The addition of oxygen tended to prevent the chlorination of antimony in the oxide. When 6 vol% hydrogen chloride gas was used, the addition of oxygen up to 0.1 vol% could convert the uranium contained in the catalyst to U{sub 3}O{sub 8} without any significant decrease in the reaction rate compared to that of the treatment without oxygen. (authors)

  15. Method of decontamination for uranium oxide particles floating in liquid waste

    International Nuclear Information System (INIS)

    Terakado, Tsutomu; Ebara, Tsuneo; Sato, Kuniaki.

    1981-01-01

    Purpose: To rapidly treat liquid waste containing uranium oxide particles floating in it and to enable substantially complete decontamination. Method: An iron salt such as ferrous sulfate or the like is added to liquid waste with floating uranium oxide particles, an alkaline solution such as caustic soda or the like is then added to the liquid waste while feeding compressed air at 0.1 to 0.02 l/sec. per ton of liquid waste, and the pH of the liquid waste is made to from 6.5 to 7.5. Thereafter, the feed of compressed air is stopped, the liquid waste is allowed to stand, and is then filtered. (Aizawa, K.)

  16. The controlling role of positive structures over the metallogenesis and emplacement of inter layer oxidation sandstone type uranium deposits

    International Nuclear Information System (INIS)

    Gu Kangheng; Chen Zuyi

    2010-01-01

    The positive structures in this paper mean the geological structures related to the occurrence of U-metallogenic zones or U-deposit such as anticlines, uplifts and uplifted fault-blocks. Occurrence features of interlayer oxidation sandstone type deposit at the southern margin of Yili basin and southwestern margin of Turpan-Hami basin, the northeastern margin of Jiudong basin illustrate that the sandstone-hosted uranium deposits, the U-mineralized sections and the uranium occurrences are always selectively emplaced on/in positive structures. The reasons for this lie in the formation mechanism of sandstone-hosted U-deposits. The positive structures raised the elevation of ore-hosting sandstone horizon and make it close to ground surface or exposed at the ground surface, which result in the infiltration of uranium and oxygen bearing groundwater from recharge area into host sandstone horizon, and the interlayer oxidation of host sandstone, as well as the dissolution and the migration of uranium in host sandstone, and the reduction mineralization at the oxidation-reduction interface. Sufficient attention should be paid to the controlling role of positive structures over the metallogenesis and emplacement of sandstone-hosted uranium deposits. They could act as an important criterion for recognizing and prognosticating potential uranium mineralized areas in uranium metallogenic zones or uranium-productive sedimentary basins. (authors)

  17. Thermodynamic analysis and experimental study on the chlorination of uranium oxide by gas-solid reaction

    International Nuclear Information System (INIS)

    Shin, Y.J.; Kim, I.S.; Shin, H.S.; Ro, S.G.; Park, H.S.

    1998-01-01

    In order to determine the operating condition of an uranium chlorination process with U 3 O 8 -C-Cl 2 system, the experimental conditions have been evaluated preliminarily by the thermochemical analysis and experimentally confirmed in this study. The dry-type chlorination of U 3 O 8 occurs as irreversible and exothermic reaction and produces many kinds of chloride compounds such as UCl 3 , UCl 4 , UCl 5 , and UCl 6 in the air and humidity controlled argon environment. Taking account of Gibbs free energy and vapor pressure for various chloride compounds, the proper temperature range of chlorination appears to be 863 to 953 K in aspects of increasing reaction rate and the yield of nonvolatile product. In the course of the experimental confirmation the powder of U 3 O 8 is perfectly converted into uranium chlorides within 4 hours above 863 K, and then the maximum fraction of uranium chloride remaining in the reactor is about 30% of total conversion mass. (author)

  18. Influences of structures on the interlayer oxidation zone sandstone-type uranium deposits on the southern margin of Yili basin

    International Nuclear Information System (INIS)

    Wang Mou; Li Shengfu

    2006-01-01

    Based on geology and the theory of hydromorphic origin uranium deposit, structural conditions of uranium formation on the southern margin of Yili Basin are analyzed from two aspects of structural movements and deformation. It is suggested that the subsidiary structures caused by the neotectonic movement are the major factor that control and reform the interlayer oxidation zone sandstone-type uranium deposit, and the differences lie in the tectonics at the eastern and western section on the southern margin of Yili Basin. At the western section, because Mesozoic and Cenozoic strata are tilted by the subsidiary structures, some strata on the margin of the basin outcrop at the surface and suffer from the weathering and erosion, which is favorable for the formation of large size uranium deposits. But at the eastern section, the fault and fold are predominant, outcropping at the surface, cause the redistribution of the uranium, which is favorable for the formation of small size uranium deposits. (authors)

  19. Effect of Tara (Caesalpinia spinosa Pod Powder on the Oxidation and Colour Stability of Pork Meat Batter During Chilled Storage

    Directory of Open Access Journals (Sweden)

    Monika Skowyra

    2015-01-01

    Full Text Available The eff ect of dried pods of Caesalpinia spinosa, known as tara, on pH, cooking loss, lipid oxidation, colour stability and texture of model meat systems stored at 4 °C for 21 days was investigated. Tara pod powder showing a potential antioxidant activity was added at 0.02, 0.04 and 0.08 % (by mass directly to the pork batt er and compared with a synthetic antioxidant, butylated hydroxyanisole (BHA and control (no added antioxidants. The addition of tara pod powder at 0.02 % was as effective as BHA (0.02 % in retarding lipid oxidation in pork products during storage. Results showed that redness increased after the addition of tara pod powder. Specifically, 0.02 % of tara pod powder was effective in keeping the red colour of meat batter stored under illumination at 4 °C for 48 h. Hardness of pork products was the lowest in samples manufactured with tara pod powder compared with control. Results highlight the potential of using tara pod powder as natural functional ingredient in the development of pork products with enhanced quality and shelf life.

  20. Studies involving direct heating of uranium and plutonium oxides by microwaves

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G K; Malav, R K; Karande, A P; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

    1997-08-01

    Nuclear fuel fabrication and recovery of nuclear materials from scraps generated during fabrication involve heating steps like dewaxing, sintering, roasting of scraps, calcination, etc. The dielectric properties of uranium and plutonium oxides place them in the category of materials which are susceptible to absorption of microwaves. The studies were carried out to explore the microwave heating technique for these steps required in nuclear fuel fabrication and scrap recovery laboratories. (author). 1 ref.

  1. Investigation of the kinetics of the reactions of oxidation, nitration, and hydrogenation of uranium

    International Nuclear Information System (INIS)

    Adda, Y.

    1955-06-01

    Various physico-chemical methods have been used to investigate the kinetics of the oxidation hydridation and nitridation of uranium. The experimental results show that the kinetics of these reactions are influenced by many factors also the Pilling and Bedworth rule is valid only under very limited conditions. The disagreement between this rule and the experimental results could be explained by the existence of numerous mechanical faults in the compounds obtained by the dry corrosion of the metal. (author) [fr

  2. Depleted uranium oxides as spent-nuclear-fuel waste-package invert and backfill materials

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Haire, M.J.

    1997-01-01

    A new technology has been proposed in which depleted uranium, in the form of oxides or silicates, is placed around the outside of the spent nuclear fuel waste packages in the geological repository. This concept may (1) reduce the potential for repository nuclear criticality events and (2) reduce long-term release of radionuclides from the repository. As a new concept, there are significant uncertainties

  3. Thermal expansion studies on uranium-neodymium mixed oxide solid solutions

    International Nuclear Information System (INIS)

    Panneerselvam, G.; Venkata Krishnan, R.; Antony, M.P.; Nagarajan, K.

    2012-01-01

    Uranium-Neodymium mixed oxides solid solutions (U 1-y Nd y ) O 2 (y=0.2-0.95) were prepared by combustion synthesis using citric acid as fuel. Structural characterization and computation of lattice parameter was carried out from room temperature X-ray diffraction measurements. Single-phase fluorite structure was observed up to y=0.80. For solid solutions with y>0.80 additional Nd 2 O 3 lines were visible

  4. A study on oxidation treatment of uranium metal chip under controlling atmosphere for safe storage

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Ji, Chul Goo; Bae, Sang Oh; Woo, Yoon Myeoung; Kim, Jong Goo; Ha, Yeong Keong

    2011-01-01

    The U metal chips generated in developing nuclear fuel and a gamma radioisotope shield have been stored under immersion of water in KAERI. When the water of the storing vessels vaporizes or drains due to unexpected leaking, the U metal chips are able to open to air. A new oxidation treatment process was raised for a long time safe storage with concepts of drying under vacuum, evaporating the containing water and organic material with elevating temperature, and oxidizing the uranium metal chips at an appropriate high temperature under conditions of controlling the feeding rate of oxygen gas. In order to optimize the oxidation process the uranium metal chips were completely dried at higher temperature than 300 .deg. C and tested for oxidation at various temperatures, which are 300 .deg. C, 400 .deg. C, and 500 .deg. C. When the oxidation temperature was 400 .deg. C, the oxidized sample for 7 hours showed a temperature rise of 60 .deg. C in the self-ignition test. But the oxidized sample for 14 hours revealed a slight temperature rise of 7 .deg. C representing a stable behavior in the self-ignition test. When the temperature was 500 .deg. C, the shorter oxidation for 7 hours appeared to be enough because the self-ignition test represented no temperature rise. By using several chemical analyses such as carbon content determination, X-ray deflection (XRD), Infrared spectra (IR) and Thermal gravimetric analysis (TGA) on the oxidation treated samples, the results of self-ignition test of new oxidation treatment process for U metal chip were interpreted and supported

  5. Preliminary study of uranium in Pennsylvanian and lower Permian strata in the Powder River Basin, Wyoming and Montana, and the Northern Great Plains

    International Nuclear Information System (INIS)

    Dunagan, J.F. Jr.; Kadish, K.A.

    1977-11-01

    Persistent and widespread radiometric anomalies occur in Pennsylvanian and Lower Permian strata in the subsurface of the northern Great Plains and the Powder River Basin. The primary host lithology of these anomalies is shale interbedded with sandstone, dolomite, and dolomitic sandstone. Samples from the project area indicate that uranium is responsible for some anomalies. In some samples there seems to be a correlation between high uranium content and high organic-carbon content, which possibly indicates that carbonaceous material acted as a trapping mechanism in some strata. The Pennsylvanian and Permian rocks studied are predominantly marine carbonates and clastics, but there are rocks of fluvial origin in the basal Pennsylvanian of Montana, North Dakota, and South Dakota and in the Pennsylvanian and Permian deposits on the east flank of the Laramie Mountains. Fine-grained clastic rocks that flank the Chadron arch in western Nebraska are possibly of continental origin. The trend of the Chadron arch approximately parallels the trend of radiometric anomalies in the subsurface Permian-Pennsylvanian section. Possible source areas for uranium in the sediments studied were pre-Pennsylvanian strata of the Canadian Shield and Precambrian igneous rocks of the Ancestral Rocky Mountains

  6. Atom economy and green elimination of nitric oxide using ZrN powders.

    Science.gov (United States)

    Chen, Ning; Wang, Jigang; Yin, Wenyan; Li, Zhen; Li, Peishen; Guo, Ming; Wang, Qiang; Li, Chunlei; Wang, Changzheng; Chen, Shaowei

    2018-05-01

    Nitric oxide (NO) may cause serious environmental problems, such as acid rain, haze weather, global warming and even death. Herein, a new low-cost, highly efficient and green method for the elimination of NO using zirconium nitride (ZrN) is reported for the first time, which does not produce any waste or any by-product. Relevant experimental parameters, such as reaction temperature and gas concentration, were investigated to explore the reaction mechanism. Interestingly, NO can be easily decomposed into nitrogen (N 2 ) by ZrN powders at 600°C with ZrN simultaneously transformed into zirconium dioxide (ZrO 2 ) gradually. The time for the complete conversion of NO into N 2 was approximately 14 h over 0.5 g of ZrN at a NO concentration of 500 ppm. This green elimination process of NO demonstrated good atom economy and practical significance in mitigating environmental problems.

  7. Mass synthesis of yttrium oxide nano-powders using radio frequency (RF) plasma

    International Nuclear Information System (INIS)

    Ghorui, S.; Sahasrabudhe, S.N.; Chakravarthy, Y.; Nagaraj, A.; Das, A.K.; Dhamale, G.

    2014-01-01

    Mass synthesis of nano-phase Yttrium Oxide (Y 2 O 3 ) from commercially available coarse grain powder is reported. Nano-sized high purity Y 2 O 3 is an important and critical constituent of ceramics like YAG (Yttrium aluminum garnet: Y 3 Al 5 O 12 ) for laser applications. The system is characterized in terms of its thermal and electrical behavior. Boltzmann plot technique is used to measure axial variation of temperature of the generated plasma. The synthesized particles are characterized in terms of XRD, SEM, TEM and BET analyses for qualification of the developed system. Major features observed are efficient conversion into nanometer-sized highly spherical particles, narrow size distribution, highly crystallite nature and highly pure phases. The particle distribution (from TEM) peaks within 20-30 nm. Average particle sizes determined from different methods like XRD, TEM and BET are very close to each other and point toward particle sizes within 20 to 30 nm

  8. Sintering uranium oxide in the reaction product of hydrogen-carbon dioxide mixtures

    International Nuclear Information System (INIS)

    De Hollander, W.R.; Nivas, Y.

    1975-01-01

    Compacted pellets of uranium oxide alone or containing one or more additives such as plutonium dioxide, gadolinium oxide, titanium dioxide, silica, and alumina are heated to 900 to 1599 0 C in the presence of a mixture of hydrogen and carbon dioxide, either alone or with an inert carrier gas and held at the desired temperature in this atmosphere to sinter the pellets. The sintered pellets are then cooled in an atmosphere having an oxygen partial pressure of 10 -4 to 10 -18 atm of oxygen such as dry hydrogen, wet hydrogen, dry carbon monoxide, wet carbon monoxide, inert gases such as nitrogen, argon, helium, and neon and mixtures of ayny of the foregoing including a mixture of hydrogen and carbon dioxide. The ratio of hydrogen to carbon dioxide in the gas mixture fed to the furnace is controlled to give a ratio of oxygen to uranium atoms in the sintered particles within the range of 1.98:1 to about 2.10:1. The water vapor present in the reaction products in the furnace atmosphere acts as a hydrolysis agent to aid removal of fluoride should such impurity be present in the uranium oxide. (U.S.)

  9. Evidence for single metal two electron oxidative addition and reductive elimination at uranium.

    Science.gov (United States)

    Gardner, Benedict M; Kefalidis, Christos E; Lu, Erli; Patel, Dipti; McInnes, Eric J L; Tuna, Floriana; Wooles, Ashley J; Maron, Laurent; Liddle, Stephen T

    2017-12-01

    Reversible single-metal two-electron oxidative addition and reductive elimination are common fundamental reactions for transition metals that underpin major catalytic transformations. However, these reactions have never been observed together in the f-block because these metals exhibit irreversible one- or multi-electron oxidation or reduction reactions. Here we report that azobenzene oxidises sterically and electronically unsaturated uranium(III) complexes to afford a uranium(V)-imido complex in a reaction that satisfies all criteria of a single-metal two-electron oxidative addition. Thermolysis of this complex promotes extrusion of azobenzene, where H-/D-isotopic labelling finds no isotopomer cross-over and the non-reactivity of a nitrene-trap suggests that nitrenes are not generated and thus a reductive elimination has occurred. Though not optimally balanced in this case, this work presents evidence that classical d-block redox chemistry can be performed reversibly by f-block metals, and that uranium can thus mimic elementary transition metal reactivity, which may lead to the discovery of new f-block catalysis.

  10. Supplementation of Citrus maxima Peel Powder Prevented Oxidative Stress, Fibrosis, and Hepatic Damage in Carbon Tetrachloride (CCl4) Treated Rats.

    Science.gov (United States)

    Chowdhury, Mohammed Riaz Hasan; Sagor, Md Abu Taher; Tabassum, Nabila; Potol, Md Abdullah; Hossain, Hemayet; Alam, Md Ashraful

    2015-01-01

    Citrus maxima peel is rich in natural phenolic compounds and has a long use in the traditional medicine. HPLC-DAD analysis on Citrus maxima peel powder exhibited the presence of various phenolic compounds such as caffeic acid and (-)-epicatechin. To determine the plausible hepatoprotective activity of Citrus maxima peel powder, we used carbon tetrachloride (CCl4) treated rat model. Liver damage in rats was confirmed by measuring the AST, ALT, and ALP enzyme activities. In addition, lipid peroxidation products (MDA), nitric oxide, advanced protein oxidation products level (APOP), and catalase activities were also analyzed along with the histological profiling for the inflammatory cell infiltration, collagen, and iron deposition in liver. Dietary supplementation of Citrus maxima peel powder exhibited significant reduction of serum AST, ALT, and ALP activities in carbon tetrachloride treated rats. Moreover, Citrus maxima peel powder also showed a significant reduction of the oxidative stress markers (MDA, NO, and APOP level) and restored the catalase activity in CCl4 treated rats. Histological examination of the liver section revealed reduced inflammatory cells infiltration, collagen, and iron deposition in CCl4 treated rats. The results from this study demonstrated that Citrus maxima peel powder produced significant hepatoprotective action in CCl4 administered rats.

  11. Supplementation of Citrus maxima Peel Powder Prevented Oxidative Stress, Fibrosis, and Hepatic Damage in Carbon Tetrachloride (CCl4 Treated Rats

    Directory of Open Access Journals (Sweden)

    Mohammed Riaz Hasan Chowdhury

    2015-01-01

    Full Text Available Citrus maxima peel is rich in natural phenolic compounds and has a long use in the traditional medicine. HPLC-DAD analysis on Citrus maxima peel powder exhibited the presence of various phenolic compounds such as caffeic acid and (−-epicatechin. To determine the plausible hepatoprotective activity of Citrus maxima peel powder, we used carbon tetrachloride (CCl4 treated rat model. Liver damage in rats was confirmed by measuring the AST, ALT, and ALP enzyme activities. In addition, lipid peroxidation products (MDA, nitric oxide, advanced protein oxidation products level (APOP, and catalase activities were also analyzed along with the histological profiling for the inflammatory cell infiltration, collagen, and iron deposition in liver. Dietary supplementation of Citrus maxima peel powder exhibited significant reduction of serum AST, ALT, and ALP activities in carbon tetrachloride treated rats. Moreover, Citrus maxima peel powder also showed a significant reduction of the oxidative stress markers (MDA, NO, and APOP level and restored the catalase activity in CCl4 treated rats. Histological examination of the liver section revealed reduced inflammatory cells infiltration, collagen, and iron deposition in CCl4 treated rats. The results from this study demonstrated that Citrus maxima peel powder produced significant hepatoprotective action in CCl4 administered rats.

  12. Insertion compounds of transition-metal and uranium oxides

    International Nuclear Information System (INIS)

    Chippindale, A.M.; Dickens, P.G.; Powell, A.V.

    1991-01-01

    Several transition-metal and actinide oxides, in which the metal occurs in a high oxidation state, have open covalent structures and are capable of incorporating alkali and other electropositive metals under mild conditions to form insertion compounds A x MO n . These are solids which have several features in common: Over a range of compositions, A x MO n exists as one or more stable or metastable phases in which the structure of the parent oxide MO n is largely retained and the insertion element A is accommodated interstitially. Insertion is accompanied by a redox process A=A i . + e - M in which M is reduced and the electronic properties of the parent oxide change to those typical of a mixed-valence compound. The insertion process xA + MO n = A x MO n can be reversed, at least to some extent, by chemical or electrochemical reaction, with retention of structure (topotactic reaction). This review concentrates on methods of synthesis, characterisation, crystal structure and thermochemistry of these insertion compounds. It updates and extends previous work. (author)

  13. The temperature coefficient of the resonance integral for uranium metal and oxide

    Energy Technology Data Exchange (ETDEWEB)

    Blomberg, P; Hellstrand, E; Homer, S

    1960-06-15

    The temperature coefficient of the resonance integral in uranium metal and oxide has been measured over a wide temperature range for rods with three different diameters. The results for metal agree with most earlier results from activation measurements but differ as much as a factor of two from results obtained with reactivity methods. For oxide only one measurement has been reported recently. Our value is considerably lower than the result of that measurement. The experiments will continue in order to find the reason for the large discrepancy mentioned above.

  14. The temperature coefficient of the resonance integral for uranium metal and oxide

    International Nuclear Information System (INIS)

    Blomberg, P.; Hellstrand, E.; Homer, S.

    1960-06-01

    The temperature coefficient of the resonance integral in uranium metal and oxide has been measured over a wide temperature range for rods with three different diameters. The results for metal agree with most earlier results from activation measurements but differ as much as a factor of two from results obtained with reactivity methods. For oxide only one measurement has been reported recently. Our value is considerably lower than the result of that measurement. The experiments will continue in order to find the reason for the large discrepancy mentioned above

  15. Anaerobic U(IV) Bio-oxidation and the Resultant Remobilization of Uranium in Contaminated Sediments

    International Nuclear Information System (INIS)

    Coates, John D.

    2005-01-01

    A proposed strategy for the remediation of uranium (U) contaminated sites is based on immobilizing U by reducing the oxidized soluble U, U(VI), to form a reduced insoluble end product, U(IV). Due to the use of nitric acid in the processing of nuclear fuels, nitrate is often a co-contaminant found in many of the environments contaminated with uranium. Recent studies indicate that nitrate inhibits U(VI) reduction in sediment slurries. However, the mechanism responsible for the apparent inhibition of U(VI) reduction is unknown, i.e. preferential utilization of nitrate as an electron acceptor, direct biological oxidation of U(IV) coupled to nitrate reduction, and/or abiotic oxidation by intermediates of nitrate reduction. Recent studies indicates that direct biological oxidation of U(IV) coupled to nitrate reduction may exist in situ, however, to date no organisms have been identified that can grow by this metabolism. In an effort to evaluate the potential for nitrate-dependent bio-oxidation of U(IV) in anaerobic sedimentary environments, we have initiated the enumeration of nitrate-dependent U(IV) oxidizing bacteria. Sediments, soils, and groundwater from uranium (U) contaminated sites, including subsurface sediments from the NABIR Field Research Center (FRC), as well as uncontaminated sites, including subsurface sediments from the NABIR FRC and Longhorn Army Ammunition Plant, Texas, lake sediments, and agricultural field soil, sites served as the inoculum source. Enumeration of the nitrate-dependent U(IV) oxidizing microbial population in sedimentary environments by most probable number technique have revealed sedimentary microbial populations ranging from 9.3 x 101 - 2.4 x 103 cells (g sediment)-1 in both contaminated and uncontaminated sites. Interestingly uncontaminated subsurface sediments (NABIR FRC Background core FB618 and Longhorn Texas Core BH2-18) both harbored the most numerous nitrate-dependent U(IV) oxidizing population 2.4 x 103 cells (g sediment)-1

  16. A Study on the Electrolytic Reduction Mechanism of Uranium Oxide in a LiCl-Li2O Molten Salt

    International Nuclear Information System (INIS)

    Oh, Seung Chul; Hur, Jin Mok; Seo, Chung Seok; Park, Seong Won

    2003-01-01

    This study proposed a new electrolytic reduction technology that is based on the integration of simultaneous uranium oxide metallization and Li 2 O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxides to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, investigation of reaction mechanism, the characteristics of closed recycle of Li 2 O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt increase of metallization yield, and simplification of process.

  17. A study on the electrolytic reduction of uranium oxide in a LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Su, J. S.; Hu, J. M.; Hong, S. S.; Jang, D. S.; Park, S. W.

    2003-01-01

    New electrolytic reduction technology was proposed that is based on the integration of metallization of uranium oxide and Li 2 O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxides to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, investigation of reaction mechanism, the characteristics of closed recycle of Li 2 O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt, increase of metallization yield, and simplification of process

  18. Fabrication of nuclear fuel by powder injection moulding: Study of the binders systems and the de-binding of feedstock containing actinide powder

    International Nuclear Information System (INIS)

    Bricout, J.

    2012-01-01

    Powder Injection Moulding (PIM) is identified as an innovative process for the nuclear fuel fabrication. Technological breakthrough compared to the current process of powder metallurgy, the impact of actinide powder's specificities on the different steps of PIM is performed. Alumina powders simulating actinide powder have been implemented with a reference binders system. Thermal and rheological studies show the injectability and the de-binding of feedstocks with adequate solid loading (≥50 %vol), thanks to the de-agglomeration during the mixing step, which allow to obtain net shape fuel pellet. Specific surface area of powders, acting as a key role in behaviour's feedstocks, has been integrated in analysis models of viscosity prediction according to the shear rate. Also conducted studies on uranium oxide powder show that the selected binders systems, which have a compatible rheological behaviour with PIM process, impact the de-agglomeration of powder and final microstructure of the fuel pellet, consistent with the results obtained on alumina powders. Independent behaviour of binders and uranium oxide powder, showing no adverse chemical reaction against the PIM process, show a residual mass of carbon of about 150 ppm after sintering. Binders system using polystyrene, resistant to radiolysis phenomena and loadable more than 50 %(vol) of actinide powder, shows the promising potential of PIM process for the fuel fabrication. (author) [fr

  19. Direct electrochemical reduction of solid uranium oxide in molten fluoride salts

    Science.gov (United States)

    Gibilaro, Mathieu; Cassayre, Laurent; Lemoine, Olivier; Massot, Laurent; Dugne, Olivier; Malmbeck, Rikard; Chamelot, Pierre

    2011-07-01

    The direct electrochemical reduction of UO 2 solid pellets was carried out in LiF-CaF 2 (+2 mass.% Li 2O) at 850 °C. An inert gold anode was used instead of the usual reactive sacrificial carbon anode. In this case, oxidation of oxide ions present in the melt yields O 2 gas evolution on the anode. Electrochemical characterisations of UO 2 pellets were performed by linear sweep voltammetry at 10 mV/s and reduction waves associated to oxide direct reduction were observed at a potential 150 mV more positive in comparison to the solvent reduction. Subsequent, galvanostatic electrolyses runs were carried out and products were characterised by SEM-EDX, EPMA/WDS, XRD and microhardness measurements. In one of the runs, uranium oxide was partially reduced and three phases were observed: nonreduced UO 2 in the centre, pure metallic uranium on the external layer and an intermediate phase representing the initial stage of reduction taking place at the grain boundaries. In another run, the UO 2 sample was fully reduced. Due to oxygen removal, the U matrix had a typical coral-like structure which is characteristic of the pattern observed after the electroreduction of solid oxides.

  20. The effect of hydrogen peroxide on uranium oxide films on 316L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Wilbraham, Richard J., E-mail: r.wilbraham@lancaster.ac.uk [The Lloyd’s Register Foundation Centre for Nuclear Engineering, Engineering Department, Lancaster University, Bailrigg, Lancashire LA1 4YR (United Kingdom); Boxall, Colin, E-mail: c.boxall@lancaster.ac.uk [The Lloyd’s Register Foundation Centre for Nuclear Engineering, Engineering Department, Lancaster University, Bailrigg, Lancashire LA1 4YR (United Kingdom); Goddard, David T., E-mail: dave.t.goddard@nnl.co.uk [National Nuclear Laboratory, Preston Laboratory, Springfields, Preston, Lancashire PR4 0XJ (United Kingdom); Taylor, Robin J., E-mail: robin.j.taylor@nnl.co.uk [National Nuclear Laboratory, Central Laboratory, Seascale, Cumbria CA20 1PG (United Kingdom); Woodbury, Simon E., E-mail: simon.woodbury@nnl.co.uk [National Nuclear Laboratory, Central Laboratory, Seascale, Cumbria CA20 1PG (United Kingdom)

    2015-09-15

    Highlights: • The first report of the presence of both UO{sub 2} and polymeric UO{sub 2}{sup 2+} in the same electrodeposited U oxide sample. • The action of H{sub 2}O{sub 2} on electrodeposited U oxides is described using corrosion based concepts. • Electrodeposited U oxide freely dissolves at hydrogen peroxide concentrations <100 μmol dm{sup −3}. • At [H{sub 2}O{sub 2}] > 0.1 mmol dm{sup −3} dissolution is inhibited by formation of a studtite passivation layer. • At [H{sub 2}O{sub 2}] ⩾ 1 mol dm{sup −3} studtite formation competes with uranyl–peroxide complex formation. - Abstract: For the first time the effect of hydrogen peroxide on the dissolution of electrodeposited uranium oxide films on 316L stainless steel planchets (acting as simulant uranium-contaminated metal surfaces) has been studied. Analysis of the H{sub 2}O{sub 2}-mediated film dissolution processes via open circuit potentiometry, alpha counting and SEM/EDX imaging has shown that in near-neutral solutions of pH 6.1 and at [H{sub 2}O{sub 2}] ⩽ 100 μmol dm{sup −3} the electrodeposited uranium oxide layer is freely dissolving, the associated rate of film dissolution being significantly increased over leaching of similar films in pH 6.1 peroxide-free water. At H{sub 2}O{sub 2} concentrations between 1 mmol dm{sup −3} and 0.1 mol dm{sup −3}, formation of an insoluble studtite product layer occurs at the surface of the uranium oxide film. In analogy to corrosion processes on common metal substrates such as steel, the studtite layer effectively passivates the underlying uranium oxide layer against subsequent dissolution. Finally, at [H{sub 2}O{sub 2}] > 0.1 mol dm{sup −3} the uranium oxide film, again in analogy to common corrosion processes, behaves as if in a transpassive state and begins to dissolve. This transition from passive to transpassive behaviour in the effect of peroxide concentration on UO{sub 2} films has not hitherto been observed or explored, either in terms

  1. Uranium

    International Nuclear Information System (INIS)

    Battey, G.C.; McKay, A.D.

    1988-01-01

    Production for 1986 was 4899 t U 3 O 8 (4154 t U), 30% greater than in 1985, mainly because of a 39% increase in production at Ranger. Exports for 1986 were 4166 t U 3 O 8 at an average f.o.b. unit value of $40.57/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1985-86 fiscal year was $50.2 million. Plans were announced to increase the nominal capacity of the processing plant at Ranger from 3000 t/year U 3 O 8 to 4500 t and later to 6000 t/year. Construction and initial mine development at Olympic Dam began in March. Production is planned for mid 1988 at an annual rate of 2000 t U 3 O 8 , 30 000 t Cu, and 90 000 oz (2800 kg) Au. The first long-term sales agreement was concluded in September 1986. At the Manyingee deposit, testing of the alkaline solution mining method was completed, and the treatment plant was dismantled. Spot market prices (in US$/lb U 3 O 8 ) quoted by Nuexco were generally stable. From January-October the exchange value fluctuated from US$17.00-US$17.25; for November and December it was US$16.75. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U at December 1986 were estimated as 462 000 t U, 3000 t U less than in 1985. This represents 30% of the total low-cost RAR in the WOCA (World Outside the Centrally Planned Economy Areas) countries. Australia also has 257 000 t U in the low-cost Estimated Additional Resources Category I, 29% of the WOCA countries' total resources in this category

  2. Influence of uranyl speciation and iron oxides on uranium biogeochemical redox reactions

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, B.D.; Amos, R.T.; Nico, P.S.; Fendorf, S.

    2010-03-15

    Uranium is a pollutant of concern to both human and ecosystem health. Uranium's redox state often dictates its partitioning between the aqueous- and solid-phases, and thus controls its dissolved concentration and, coupled with groundwater flow, its migration within the environment. In anaerobic environments, the more oxidized and mobile form of uranium (UO{sub 2}{sup 2+} and associated species) may be reduced, directly or indirectly, by microorganisms to U(IV) with subsequent precipitation of UO{sub 2}. However, various factors within soils and sediments may limit biological reduction of U(VI), inclusive of alterations in U(VI) speciation and competitive electron acceptors. Here we elucidate the impact of U(VI) speciation on the extent and rate of reduction with specific emphasis on speciation changes induced by dissolved Ca, and we examine the impact of Fe(III) (hydr)oxides (ferrihydrite, goethite and hematite) varying in free energies of formation on U reduction. The amount of uranium removed from solution during 100 h of incubation with S. putrefaciens was 77% with no Ca or ferrihydrite present but only 24% (with ferrihydrite) and 14% (no ferrihydrite) were removed for systems with 0.8 mM Ca. Imparting an important criterion on uranium reduction, goethite and hematite decrease the dissolved concentration of calcium through adsorption and thus tend to diminish the effect of calcium on uranium reduction. Dissimilatory reduction of Fe(III) and U(VI) can proceed through different enzyme pathways, even within a single organism, thus providing a potential second means by which Fe(III) bearing minerals may impact U(VI) reduction. We quantify rate coefficients for simultaneous dissimilatory reduction of Fe(III) and U(VI) in systems varying in Ca concentration (0 to 0.8 mM), and using a mathematical construct implemented with the reactive transport code MIN3P, we reveal the predominant influence of uranyl speciation, specifically the formation of uranyl

  3. Anti-oxidant effect of extracts of kinnow rind, pomegranate rind and seed powders in cooked goat meat patties.

    Science.gov (United States)

    Devatkal, Suresh K; Narsaiah, K; Borah, A

    2010-05-01

    To overcome the disadvantages of using synthetic anti-oxidants in meat products, an investigation was carried out to evaluate the anti-oxidant effect of extracts of fruit by-products viz., kinnow rind powder (KRP), pomegranate rind powder (PRP) and pomegranate seed powder (PSP) in goat meat patties. Total phenolics content, DPPH radical scavenging activity and effect of these extracts on instrumental color, sensory attributes and TBARS values during storage (4+/-1 degrees C) of goat meat patties were evaluated. Results showed that these extracts are rich sources of phenolic compounds having free radical scavenging activity. Hunter Lab L value significantly (Pgoat meat patties was observed in PRP, PSP and KRP as compared to control patties. Average TBARS values (mg/kg meat) during refrigerated storage (4+/-1 degrees C) were significantly lower in PRP, followed by PSP and KRP as compared to control. The overall anti-oxidant effect was in the order of PRP>PSP>KRP. It was concluded that extracts of above fruits by-product powders have potential to be used as natural anti-oxidants in meat products. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

  4. Thorium/uranium mixed oxide nano-crystals: Synthesis, structural characterization and magnetic properties

    International Nuclear Information System (INIS)

    Hudry, Damien; Griveau, Jean-Christophe; Apostolidis, Christos; Colineau, Eric; Rasmussen, Gert; Walter, Olaf; Wang, Di; Venkata Sai Kiran Chakravadhaluna; Courtois, Eglantine; Kubel, Christian

    2014-01-01

    One of the primary aims of the actinide community within nano-science is to develop a good understanding similar to what is currently the case for stable elements. As a consequence, efficient, reliable and versatile synthesis techniques dedicated to the formation of new actinide-based nano-objects (e.g., nano-crystals) are necessary. Hence, a 'library' dedicated to the preparation of various actinide based nano-scale building blocks is currently being developed. Nano-scale building blocks with tunable sizes, shapes and compositions are of prime importance. So far, the non-aqueous synthesis method in highly coordinating organic media is the only approach which has demonstrated the capability to provide size and shape control of actinide-based nano-crystals (both for thorium and uranium, and recently extended to neptunium and plutonium). In this paper, we demonstrate that the non-aqueous approach is also well adapted to control the chemical composition of the nano-crystals obtained when mixing two different actinides. Indeed, the controlled hot co-injection of thorium acetylacetonate and uranyl acetate (together with additional capping agents) into benzyl ether can be used to synthesize thorium/uranium mixed oxide nano-crystals covering the full compositional spectrum. Additionally, we found that both size and shape are modified as a function of the thorium/uranium ratio. Finally, the magnetic properties of the different thorium/uranium mixed oxide nano-crystals were investigated. Contrary to several reports, we did not observe any ferromagnetic behavior. As a consequence, ferromagnetism cannot be described as a universal feature of nano-crystals of non-magnetic oxides as recently claimed in the literature. (authors)

  5. Reduction of uranium and plutonium oxides by aluminum. Application to the recycling of plutonium

    International Nuclear Information System (INIS)

    Gallay, J.

    1968-01-01

    A process for treating plutonium oxide calcined at high temperatures (1000 to 2000 deg. C) with a view to recovering the metal consists in the reduction of this oxide dissolved in a mixture of aluminium, sodium and calcium fluorides by aluminium at about 1180 deg. C. The first part of the report presents the results of reduction tests carried out on the uranium oxides UO 2 and U 3 O 8 ; these are in agreement with the thermodynamic calculations of the exchange reaction at equilibrium. The second part describes the application of this method to plutonium oxides. The Pu-Al alloy obtained (60 per cent Pu) is then recycled in an aqueous medium. (author) [fr

  6. Description and classification of uranium oxide hydrate sheet topologies

    International Nuclear Information System (INIS)

    Miller, M.L.; Burns, P.C.; Ewing, R.C.; Finch, R.J.

    1996-01-01

    The uranyl oxide hydrates (UOH) are important corrosion products of uraninite and UO 2 in spent nuclear fuel under oxidizing conditions. However, the systematics of the crystal chemistry, thermodynamic parameters, and solubilities of this mineral group are poorly understood. With the exception of the synthetic UO 2 (OH) 2 polymorphs, all UOH crystal structures are based on sheets of edge-sharing 5 and 4-coordinated uranyl dipyramids. This structural similarity suggests that it is possible to develop a model by which to estimate the thermodynamic behavior of UOHs from data on structural endmember phases. Toward this end, a method of quantitatively describing all known UOH sheets has been developed. Only four structural unit chains are required to construct the uranyl oxide hydrate sheets (as well as the structurally similar U 3 O 8 sheets). The H-chain is restricted to α-UO 2 (OH) 2 and is made up of hexagonally coordinated uranyl ions sharing opposing edges. The arrowhead chain composed of pentagonal dipyramids sharing edges and alternating with trigonal vacancies is present in all other UOH sheets. These arrowhead chains are directed and can occur in both an Up-arrow and Down-arrow sense within a single sheet. The P-chain consists of edge-sharing pentagonal dipyramids forming a zigzag chain. The P-chain is flanked on both sides by arrowhead chains of the same sense. The remaining structural unit is a discontinuous chain of rhombic dipyramids. This R-chain is produced when nested adjacent Up-arrow and Down-arrow arrowhead chains are translated by a diagonal shift. This chain occurs in sheets which contain only 4-coordinate uranyl ion and those containing both 4- and 5-coordinate uranyl ions

  7. Effects of Lithium Dopant on Size and Morphology of Magnesium Oxide Nano powders

    International Nuclear Information System (INIS)

    Mohd Sufri Mastuli; Siti Nur Hazlinda Hasbu; Noraziahwati Ibrahim; Mohd Azizi Nawawi; Mohd Sufri Mastuli

    2014-01-01

    Lithium doped of magnesium oxide powders have been synthesized using the sol-gel method with magnesium acetate tetrahydrate, oxalic acid dihydrate and lithium acetate dihydrate used as the starting materials. The dried sol-gel products were calcined at 950 degree Celsius for 36 h to form the Li doped-MgO samples. The calcined samples were characterized using X-Ray diffraction (XRD) and field emission scanning electron microscopy (FESEM). The present work is investigated the effect of lithium ion on the band gap energy of studied samples. The band gap energies were obtained from a Tauc plot that drawn based on absorption edge of each sample that measured using a UV-Vis spectrophotometer. It is found that the doped and undoped MgO samples showed a slightly different in their band gap energies. The lithium ion that present in the MgO as a dopant affects the crystallite size and morphology of the final products. Our study shows that the lithium dopant can modified optical properties of the metal oxide which to be beneficial in some industrial applications. (author)

  8. Subsurface Conditions Controlling Uranium Incorporation in Iron Oxides: A Redox Stable Sink

    International Nuclear Information System (INIS)

    Fendorf, Scott

    2016-01-01

    Toxic metals and radionuclides throughout the U.S. Department of Energy Complex pose a serious threat to ecosystems and to human health. Of particular concern is the redox-sensitive radionuclide uranium, which is classified as a priority pollutant in soils and groundwaters at most DOE sites owing to its large inventory, its health risks, and its mobility with respect to primary waste sources. The goal of this research was to contribute to the long-term mission of the Subsurface Biogeochemistry Program by determining reactions of uranium with iron (hydr)oxides that lead to long-term stabilization of this pervasive contaminant. The research objectives of this project were thus to (1) identify the (bio)geochemical conditions, including those of the solid-phase, promoting uranium incorporation in Fe (hydr)oxides, (2) determine the magnitude of uranium incorporation under a variety of relevant subsurface conditions in order to quantify the importance of this pathway when in competition with reduction or adsorption; (3) identify the mechanism(s) of U(VI/V) incorporation in Fe (hydr)oxides; and (4) determine the stability of these phases under different biogeochemical (inclusive of redox) conditions. Our research demonstrates that redox transformations are capable of achieving U incorporation into goethite at ambient temperatures, and that this transformation occurs within days at U and Fe(II) concentrations that are common in subsurface geochemical environments with natural ferrihydrites - inclusive of those with natural impurities. Increasing Fe(II) or U concentration, or initial pH, made U(VI) reduction to U(IV) a more competitive sequestration pathway in this system, presumably by increasing the relative rate of U reduction. Uranium concentrations commonly found in contaminated subsurface environments are often on the order of 1-10 μM, and groundwater Fe(II) concentrations can reach exceed 1 mM in reduced zones of the subsurface. The redox-driven U(V) incorporation

  9. Subsurface Conditions Controlling Uranium Incorporation in Iron Oxides: A Redox Stable Sink

    Energy Technology Data Exchange (ETDEWEB)

    Fendorf, Scott [Stanford Univ., CA (United States)

    2016-04-05

    Toxic metals and radionuclides throughout the U.S. Department of Energy Complex pose a serious threat to ecosystems and to human health. Of particular concern is the redox-sensitive radionuclide uranium, which is classified as a priority pollutant in soils and groundwaters at most DOE sites owing to its large inventory, its health risks, and its mobility with respect to primary waste sources. The goal of this research was to contribute to the long-term mission of the Subsurface Biogeochemistry Program by determining reactions of uranium with iron (hydr)oxides that lead to long-term stabilization of this pervasive contaminant. The research objectives of this project were thus to (1) identify the (bio)geochemical conditions, including those of the solid-phase, promoting uranium incorporation in Fe (hydr)oxides, (2) determine the magnitude of uranium incorporation under a variety of relevant subsurface conditions in order to quantify the importance of this pathway when in competition with reduction or adsorption; (3) identify the mechanism(s) of U(VI/V) incorporation in Fe (hydr)oxides; and (4) determine the stability of these phases under different biogeochemical (inclusive of redox) conditions. Our research demonstrates that redox transformations are capable of achieving U incorporation into goethite at ambient temperatures, and that this transformation occurs within days at U and Fe(II) concentrations that are common in subsurface geochemical environments with natural ferrihydrites—inclusive of those with natural impurities. Increasing Fe(II) or U concentration, or initial pH, made U(VI) reduction to U(IV) a more competitive sequestration pathway in this system, presumably by increasing the relative rate of U reduction. Uranium concentrations commonly found in contaminated subsurface environments are often on the order of 1-10 μM, and groundwater Fe(II) concentrations can reach exceed 1 mM in reduced zones of the subsurface. The redox-driven U(V) incorporation

  10. Oxidative Uranium Release from Anoxic Sediments under Diffusion-Limited Conditions.

    Science.gov (United States)

    Bone, Sharon E; Cahill, Melanie R; Jones, Morris E; Fendorf, Scott; Davis, James; Williams, Kenneth H; Bargar, John R

    2017-10-03

    Uranium (U) contamination occurs as a result of mining and ore processing; often in alluvial aquifers that contain organic-rich, reduced sediments that accumulate tetravalent U, U(IV). Uranium(IV) is sparingly soluble, but may be mobilized upon exposure to nitrate (NO 3 - ) and oxygen (O 2 ), which become elevated in groundwater due to seasonal fluctuations in the water table. The extent to which oxidative U mobilization can occur depends upon the transport properties of the sediments, the rate of U(IV) oxidation, and the availability of inorganic reductants and organic electron donors that consume oxidants. We investigated the processes governing U release upon exposure of reduced sediments to artificial groundwater containing O 2 or NO 3 - under diffusion-limited conditions. Little U was mobilized during the 85-day reaction, despite rapid diffusion of groundwater within the sediments and the presence of nonuraninite U(IV) species. The production of ferrous iron and sulfide in conjunction with rapid oxidant consumption suggested that the sediments harbored large concentrations of bioavailable organic carbon that fueled anaerobic microbial respiration and stabilized U(IV). Our results suggest that seasonal influxes of O 2 and NO 3 - may cause only localized mobilization of U without leading to export of U from the reducing sediments when ample organic carbon is present.

  11. Plutonium-uranium mixed oxide characterization by coupling micro-X-ray diffraction and absorption investigations

    Science.gov (United States)

    Degueldre, C.; Martin, M.; Kuri, G.; Grolimund, D.; Borca, C.

    2011-09-01

    Plutonium-uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The potential differences of metal redox state and microstructural developments of the matrix before and after irradiation are commonly analysed by electron probe microanalysis. In this work the structure and next-neighbor atomic environments of Pu and U oxide features within unirradiated homogeneous MOX and irradiated (60 MW d kg -1) MOX samples was analysed by micro-X-ray fluorescence (μ-XRF), micro-X-ray diffraction (μ-XRD) and micro-X-ray absorption fine structure (μ-XAFS) spectroscopy. The grain properties, chemical bonding, valences and stoichiometry of Pu and U are determined from the experimental data gained for the unirradiated as well as for irradiated fuel material examined in the center of the fuel as well as in its peripheral zone (rim). The formation of sub-grains is observed as well as their development from the center to the rim (polygonization). In the irradiated sample Pu remains tetravalent (>95%) and no (oxidation in the rim zone. Any slight potential plutonium oxidation is buffered by the uranium dioxide matrix while locally fuel cladding interaction could also affect the redox of the fuel.

  12. A review of the oxidation of uranium dioxide at temperatures below 400oC

    International Nuclear Information System (INIS)

    McEachern, R.J.; Taylor, P.

    1997-01-01

    A critical review of the extensive literature on the air oxidation Of U0 2 at temperatures below 400 o C is presented. The key parameters that affect the rate Of U0 2 oxidation are examined systematically, and their importance to the reaction rate is evaluated. The formation of U 30 7/U 4 0 9 on unirradiated U0 2 powders follows the discrete-layer mechanism and displays diffusion-controlled kinetics. In contrast, U 3 0 8 formation on unirradiated U0 2 displays sigmoidal 'nucleation-and-growth' kinetics. Low-temperature oxidation of used fuel tends to proceed by rapid grain-boundary oxidation followed by simultaneous intragranular oxidation throughout the sample. The activation energy for the formation Of U 3 0 7 /U 4 0 9 is 96 kJ mol -1 for U0 2 powders, 99 kJ mol -1 for sintered pellets and 106 kJ mol -1 for used fuel. The activation energy for the formation Of U 3 0 8 is temperature dependent. The best estimate of the activation energy below ∼325 o C is 154 kJ mol -1 , but all the kinetic data incorporate substantial approximations so that further study is required to properly predict the behaviour of used fuel under low-temperature ( o C) dry-air storage conditions, based on high-temperature (200 to 350 o C) laboratory data. (author). 204 refs., 5 tabs., 4 figs

  13. Late-occurring pulmonary pathologies following inhalation of mixed oxide (uranium + plutonium oxide) aerosol in the rat.

    Science.gov (United States)

    Griffiths, N M; Van der Meeren, A; Fritsch, P; Abram, M-C; Bernaudin, J-F; Poncy, J L

    2010-09-01

    Accidental exposure by inhalation to alpha-emitting particles from mixed oxide (MOX: uranium and plutonium oxide) fuels is a potential long-term health risk to workers in nuclear fuel fabrication plants. For MOX fuels, the risk of lung cancer development may be different from that assigned to individual components (plutonium, uranium) given different physico-chemical characteristics. The objective of this study was to investigate late effects in rat lungs following inhalation of MOX aerosols of similar particle size containing 2.5 or 7.1% plutonium. Conscious rats were exposed to MOX aerosols and kept for their entire lifespan. Different initial lung burdens (ILBs) were obtained using different amounts of MOX. Lung total alpha activity was determined by external counting and at autopsy for total lung dose calculation. Fixed lung tissue was used for anatomopathological, autoradiographical, and immunohistochemical analyses. Inhalation of MOX at ILBs ranging from 1-20 kBq resulted in lung pathologies (90% of rats) including fibrosis (70%) and malignant lung tumors (45%). High ILBs (4-20 kBq) resulted in reduced survival time (N = 102; p inhalation result in similar risk for development of lung tumors as compared with industrial plutonium oxide.

  14. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

  15. Irradiation behavior of uranium oxide-aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

  16. Study of the discontinuous dissolution of uranium oxides

    International Nuclear Information System (INIS)

    Bueno, L.A.O.; Floh, B.; Araujo, J.A. de.

    1978-01-01

    Dissolution studies of UO 2 and U 3 O 8 particles and pellets in nitric acid were carried out to find the best flowsheet conditions for treatment of irradiated materials. All experiments were accomplished with unirradiated oxides at room-and boiling point temperature of the nitric acid solutions, the acid molarity ranging from 1 up to 12M in stoichiometric-and (100% up to 300%) excess conditions. The UO 2 (10g) and U 3 O 8 (10g) are easily dissolved (10 and 50s), respectively, at boiling point of 6M nitric acid solution. At the same conditions compacted pellets are dissolved in 29 min (U 3 O 8 =5g) and in 330 min (UO 2 =20g) [pt

  17. Investigation of the effects of phase transformations in micro and nano aluminum powders on kinetics of oxidation using thermogravimetric analysis.

    Science.gov (United States)

    Saceleanu, Florin; Atashin, Sanam; Wen, John Z

    2017-07-26

    Aluminum micro and nanoparticles are key ingredients in the synthesis of nano energetic materials. Hence it is important to characterize the kinetics and the rate controlling process of their oxidation. The literature shows that the mass diffusion and phase transformation within the aluminum oxide shell are important. However, the description of physical processes regarding simultaneous oxidation and phase transformation is lacking. In this paper, the controlled thermogravimetric (TGA) oxidation of 40-60 nm and 1 µm Al powders is investigated at constant heating rates and under isothermal conditions, respectively, upon varying the partial pressure of oxygen. It is found that the core-shell model of homogenous oxidation is applicable to explain the TGA results when the shell does not undergo phase transformation, which predicts the apparent activation energy in good agreement with the literature data. On the other hand, the simultaneous oxidation and phase transformation is able to be addressed using the JMAK model which reveals key parameters of the rate controlling processes. Mass diffusion is indeed rate determining during the oxidation of Al micro and nanopowders while the kinetics of the reaction is fast. Unlike the micron powders, the particle size distribution has a significant effect on the shape of the oxidation curves of the nanopowders.

  18. Individual monitoring of internal exposure to uranium oxides in two fuel fabrication plants

    International Nuclear Information System (INIS)

    Bourdeix, F.; Achiary, J.; Berard, P.

    1989-01-01

    Individual monitoring of personal exposure to inhalation of uranium oxides throughout the manufacture of fuel for pressurized water reactor (PWR) includes lung gamma-spectrometry, fecal analysis and urine analysis. Examination of the results shows the following: internal exposure is the consequence of repeated intake incidents as revealed by early peaks of urinary and particularly fecal elimination; a shift is often observed with the results of aerosol concentration measured through air collectors; the measured variations of uranium lung incorporations are relatively fast (apparent mean period 165 d). Correct evaluation of the effective dose equivalent from inhalation requires further information concerning the aerosol size distribution at work stations, the physico-chemical characteristics of the product leading to an estimate of its actual biological solubility, and the measurement of the fraction of aerosol liable to intake with an individual portable collector [fr

  19. Tungsten oxide coatings deposited by plasma spray using powder and solution precursor for detection of nitrogen dioxide gas

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chao, E-mail: zhangc@yzu.edu.cn [College of Mechanical Engineering, Yangzhou University, Yangzhou 225127 (China); Wang, Jie [College of Mechanical Engineering, Yangzhou University, Yangzhou 225127 (China); Geng, Xin [College of Mechanical Engineering, Yangzhou University, Yangzhou 225127 (China); College of Chemistry and Chemical Engineering, Yangzhou University, Yangzhou 225002 (China)

    2016-05-25

    Increasing attention has been paid on preparation methods for resistive-type gas sensors based on semiconductor metal oxides. In this work, tungsten oxide (WO{sub 3}) coatings were prepared on alumina substrates and used as gas sensitive layers. The coatings were deposited by atmospheric plasma spray using powder, solution precursor, or a combination of both. Tungsten oxide powder through a powder port and ammonium tungstate aqueous solution through a liquid port were injected into plasma stream respectively or together to deposit WO{sub 3} coatings. Phase structures in the coatings were characterized by X-ray diffraction analyzer. The field-emission scanning electron microscopy images confirmed that the coatings were in microstructure, nanostructure or micro-nanostructure. The sensing properties of the sensors based on the coatings exposed to 1 ppm nitrogen dioxide gas were characterized in a home-made instrument. Sensing properties of the coatings were compared and discussed. The influences of gas humidity and working temperature on the sensor responses were further studied. - Highlights: • Porous gas sensitive coatings were deposited by plasma spray using powder and solution precursor. • Crystallized WO{sub 3} were obtained through hybrid plasma spray plus a pre-conditioned step. • Plasma power had an important influence on coating microstructure. • The particle size of atmospheric plasma-sprayed microstructured coating was stable. • Solution precursor plasma-sprayed WO{sub 3} coatings had nanostructure and showed good responses to 1 ppm NO{sub 2}.

  20. Selective oxidation of propene on bismuth molybdate and mixed oxides of tin and antimony and of uranium and antimony

    International Nuclear Information System (INIS)

    Pendleton, P.; Taylor, D.

    1976-01-01

    Propene + 18 0 2 reactions have been studied in a static reaction system on bismuth molybdate and mixed oxides of tin and antimony and of uranium and antimony. The [ 16 0] acrolein content of the total acrolein formed and the proportion of 16 0 in the oxygen of the carbon dioxide by-product have been determined. The results indicate that for each catalyst the lattice is the only direct source of the oxygen in the aldehyde, and that lattice and/or gas phase oxygen is used in carbon dioxide formation. Oxygen anion mobility appears to be greater in the molybdate catalyst than in the other two. (author)

  1. Direct dissolution and supercritical fluid extraction of uranium from UO2 powder, granule, green pellet and sintered pellet

    International Nuclear Information System (INIS)

    Rao, Ankita; Kumar, Pradeep; Ramakumar, K.L.

    2009-01-01

    In the present work, direct dissolution and extraction of UO 2 from the solid rejects various stages of fuel fabrication viz. powder granules green pellet and, sintered pellet has been studied. Powder and granules could be easily dissolved in TBP-HNO 3 complex at 50 deg C., whereas in case of green and sintered pellets at elevated temperature at raised to 80 deg C in TBP-HNO 3 complex. With supercritical (SC) CO 2 alone the efficiency was ∼70%. But with SC CO 2 +2.5% TBP, the efficiency was ∼95% for powder and granules, and ∼60% for green and sintered pellets. Nearly complete extraction (∼99%) was achievable for SC CO 2 + 2.5 % TTA in all cases. The method has distinct advantage of elimination of acid usage and minimization of liquid waste generation. (author)

  2. Contribution to the study of gaseous Carburization of Uranium

    International Nuclear Information System (INIS)

    Esteban Hernandez, J. A.; Jimenez Moreno, J. M.; Villota Ruiz, P. de

    1966-01-01

    Thermal decomposition of uranium hydride powder obtained by hydrogenation of uranium turnings is studied on the first part of this paper. Carburization of the uranium hydride or metallic uranium powder with methane is studied in the second part. A method of uranium monocarbide fabrication under static atmosphere is described. On this method hydrogen is removed by means of an uranium getter. (Author) 6 refs

  3. A procedure for oxidation during the acid leaching of non-ferrous ores, particularly uranium ores

    International Nuclear Information System (INIS)

    Zubcek, L.; Baloun, S.; Martinek, K.; Vebr, Z.; Krepelka, J.; Lasica, S.

    1989-01-01

    It is suggested that dust from the production of ferroalloys of manganese, particularly ferrosilicomanganese and ferromanganese, can be conveniently used for oxidation during the acid leaching of non-ferrous ores, particularly uranium ores. This dust contains 30 to 40% oxides of manganese, about one-half of this is MnO 2 . Iron in the dust is present in the trivalent form, and the dust is pefectly dry. The conventional grinding of oxidants for the ore processing is eliminated, the dust being available in particle size below 0.2 mm. The dust is added in amounts of 5 to 100 kg per ton of the ore, and the suspension is typically heated at 115 degC for 3.5 hr. (P.A.)

  4. Growth characteristics of a strain of iron-oxidizing bacterium and its application in bioleaching of uranium ores

    International Nuclear Information System (INIS)

    Zhang Rui; Liu Yajie; Gao Feng; Xu Lingling

    2008-01-01

    05B is a strain of iron-oxidizing bacterium which separated from a uranium ore. The effect of temperature, initial pH, inoculation amount and initial total iron concentration on the strain's growth and activities in bioleaching of uranium ores are studied. The results show that the optimum growth temperature is 40-45 degree C, the optimum inoculation pH value being 1.5-1.7, the optimum initial inoculation amount being 10%-20%, and the initial total iron concentration being not more than 5 g/L. 05B is fit for leaching of low grade uranium ores. (authors)

  5. Evaluating chemical extraction techniques for the determination of uranium oxidation state in reduced aquifer sediments

    Science.gov (United States)

    Stoliker, Deborah L.; Campbell, Kate M.; Fox, Patricia M.; Singer, David M.; Kaviani, Nazila; Carey, Minna; Peck, Nicole E.; Barger, John R.; Kent, Douglas B.; Davis, James A.

    2013-01-01

    Extraction techniques utilizing high pH and (bi)carbonate concentrations were evaluated for their efficacy in determining the oxidation state of uranium (U) in reduced sediments collected from Rifle, CO. Differences in dissolved concentrations between oxic and anoxic extractions have been proposed as a means to quantify the U(VI) and U(IV) content of sediments. An additional step was added to anoxic extractions using a strong anion exchange resin to separate dissolved U(IV) and U(VI). X-ray spectroscopy showed that U(IV) in the sediments was present as polymerized precipitates similar to uraninite and/or less ordered U(IV), referred to as non-uraninite U(IV) species associated with biomass (NUSAB). Extractions of sediment containing both uraninite and NUSAB displayed higher dissolved uranium concentrations under oxic than anoxic conditions while extractions of sediment dominated by NUSAB resulted in identical dissolved U concentrations. Dissolved U(IV) was rapidly oxidized under anoxic conditions in all experiments. Uraninite reacted minimally under anoxic conditions but thermodynamic calculations show that its propensity to oxidize is sensitive to solution chemistry and sediment mineralogy. A universal method for quantification of U(IV) and U(VI) in sediments has not yet been developed but the chemical extractions, when combined with solid-phase characterization, have a narrow range of applicability for sediments without U(VI).

  6. Determination of oxygen in ternary uranium oxides by a gravimetric alkaline earth addition method

    International Nuclear Information System (INIS)

    Fujino, T.; Tagawa, H.

    1979-01-01

    The applicability of a gravimetric method based on alkaline earth metal addition for the determination of oxygen in ternary uranium oxides of the tupe M-U-O (M=La, Ce and Th) is described. The oxide sample is mixed with MgO or Basub(2.8)UOsub(5.8) and heated in air under suitable conditions. Because uranium is completely oxidized to the hexavalent state during the reaction, oxygen can be determined from the weight change. Oxygen in Lasub(y)Usub(1-y)Osub(2+x) is determined up to y = 0.8 with a standard deviation for x of +- 0.006 with MgO. For Thsub(y)Usub(1-y)Osub(2+x) the value of x is determined with Basub(2.8)UOsub(5.8) with a standard deviation of +- 0.01 at y = 0.8. For Cesub(y)Usub(1-y)Osub(2+x), the method can be applied only for low cerium concentrations where y = 0-0.2; the value for x with Basub(2.8)UOsub(5.8) at y = 0.2 showed a standard deviation of +- 0.002. (Auth.)

  7. Manufacturing And High Temperature Oxidation Properties Of Electro-Sprayed Fe-24.5% Cr-5%Al Powder Porous Metal

    Directory of Open Access Journals (Sweden)

    Lee Kee-Ahn

    2015-06-01

    Full Text Available Fe-Cr-Al based Powder porous metals were manufactured using a new electro-spray process, and the microstructures and high-temperature oxidation properties were examined. The porous materials were obtained at different sintering temperatures (1350°C, 1400°C, 1450°C, and 1500°C and with different pore sizes (500 μm, 450 μm, and 200 μm. High-temperature oxidation experiments (TGA, Thermal Gravimetry Analysis were conducted for 24 hours at 1000°C in a 79% N2+ 21% O2, 100 mL/min. atmosphere. The Fe-Cr-Al powder porous metals manufactured through the electro-spray process showed more-excellent oxidation resistance as sintering temperature and pore size increased. In addition, the fact that the densities and surface areas of the abovementioned powder porous metals had the largest effects on the metal’s oxidation properties could be identified.

  8. Study for electrochemical behavior of uranium oxide in a molten LiCl-Li2O system

    International Nuclear Information System (INIS)

    Park, Sung Bin; Park, Byung Heung; Seo, Chung Seok; Jung, Ki Jung; Park, Seong Won

    2005-01-01

    Interest in the electrolytic reduction of uranium oxide is increasing in the treatment of spent fuel oxides. With complicated and expensive procedures many reactive metals can be prepared in a pure metal form, the electrochemical reduction of a metal oxide has been recently proposed in metallurgy. The electrochemical reduction process is simple and rapid when compared to the conventional processes. The process can reduce the production costs and be applicable to a wide range of metal oxides. Chen et al. proposed the direct electrochemical reduction of titanium dioxide to titanium in a molten calcium chloride. Argonne National Laboratory (ANL) has reported the experimental results of an electrochemical reduction of the uranium oxide fuel in a bench-scale apparatus with a cyclic voltammetry, and has designed high-capacity reduction (HCR) cells and conducted three kg-scale UO 2 reduction runs. Gourishankar et al. classified the mechanisms of the electrolytic reduction of the metal oxides in a LiCl-Li 2 O molten salt system into two types; the simultaneous reduction and the direct electrochemical reduction. The uranium oxide in LiCl-Li 2 O molten salt was converted to uranium metal according to two mechanisms. Korea Atomic Energy Research Institute (KAERI) has developed the Advanced Spent Fuel Conditioning Process (ACP) to be an innovative technology in handling the PWR spent fuel. As part of ACP, the electrolytic reduction process (ER process) is the electrochemical reduction process of uranium oxide to uranium metal in molten salt. The ER process has advantages in a technical stability, an economic potential and a good proliferation resistance. KAERI has reported on the good experimental results of an electrochemical reduction of the uranium oxide in a 20 kg HM/batch lab-scale. In this work, cyclic voltammograms for a LiCl-3 wt% Li 2 O system and an U 3 O 8 -LiCl-3 wt% Li 2 O system with the integrated cathode assembly have been obtained. From the cyclic

  9. Transformations of highly enriched uranium into metal or oxide; Etudes des procedes de transformation des composes d'uranium a fort enrichissement isotopique

    Energy Technology Data Exchange (ETDEWEB)

    Nollet, P; Sarrat, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    The enriched uranium workshops in Cadarache have a double purpose on the one hand to convert uranium hexafluoride into metal or oxide, and on the other hand to recover the uranium contained in scrap materials produced in the different metallurgical transformations. The principles that have been adopted for the design and safety of these workshops are reported. The nuclear safety is based on the geometrical limitations of the processing vessels. To establish the processes and the technology of these workshops, many studies have been made since 1960, some of which have led to original achievements. The uranium hexafluoride of high isotopic enrichment is converted either by injection of the gas into ammonia or by an original process of direct hydrogen reduction to uranium tetrafluoride. The uranium contained m uranium-zirconium metal scrap can be recovered by combustion with hydrogen chloride followed treatment of the uranium chloride by fluorine in order to obtain the uranium in the hexafluoride state. Recovery of the uranium contained m various scrap materials is obtained by a conventional refining process combustion of metallic scrap, nitric acid dissolution of the oxide, solvent purification by tributyl phosphate, ammonium diuranate precipitation, calcining, reduction and hydro fluorination into uranium tetrafluoride, bomb reduction by calcium and slag treatment. Two separate workshops operate along these lines one takes care of the uranium with an isotopic enrichment of up to 3 p. 100, the other handles the high enrichments. The handling of each step of this process, bearing in mind the necessity for nuclear safety, has raised some special technological problems and has led to the conception of new apparatus, in particular the roasting furnace for metal turnings, the nitric acid dissolution unit, the continuous precipitator and ever safe filter and dryer for ammonium diuranate, the reduction and hydro fluorination furnace and the slag recovery apparatus These are

  10. Calculation of oxygen distribution in uranium-plutonium oxide fuels during irradiation (programme CODIF)

    International Nuclear Information System (INIS)

    Moreno, A.; Sari, C.

    1978-01-01

    Radial gradients of oxygen to metal ratio, O/M, in uranium-plutonium oxide fuel pins, during irradiation and at the end of life, have been calculated on the basis of solid-state thermal diffusion using measured values of the heat of transport. A detailed computer model which includes the calculation of temperature profiles and the variation of the average O/M ratio as a function of burn-up is given. Calculations show that oxygen profiles are affected by the isotopic composition of the fuel, by the temperature profiles and by fuel-cladding interactions

  11. Oxidative destruction of ammonia for restoration of uranium solution mining sites

    International Nuclear Information System (INIS)

    Humenick, M.J.; Garwacka, K.

    1984-01-01

    A laboratory experimental research project was conducted to evaluate the use of chlorine for the oxidative destruction of residual ammonia that may remain in ground water after in-situ uranium solution mining operations. The work tested the idea of injecting high strength calcium hypochlorite solution into the mining zone to convert ammonia to nitrogen gas as a final cleanup process for ammonia removal from the ground water system. This paper details ammonia removal efficiency as a function of chlorine dose, reactant, and product material balances, and how the concept may be used as a final ground water restoration process

  12. Adsorption of uranium composites onto saltrock oxides - experimental and theoretical study.

    Science.gov (United States)

    Ivanova, Bojidarka; Spiteller, Michael

    2014-09-01

    The study encompassed experimental mass spectrometric and theoretical quantum chemical studies on adsorption of uranium species in different oxidation states of the metal ion, and oxides of UxOy(n+) type, where x = 1 or 3, y = 2 or 8, and n = 0, 1 or 2 onto nanosize-particles of saltrock oxides MO (M = Mg(II), Ca(II), Ni(II), Co(II), Sr(II) or Ba(II)), M2Oy (M = Au(III) or Ag(I), y = 3 or 1) silicates 3Al2O3.2SiO2, natural kaolinite (Al2O2·2SiO2·2H2O), illite (K0.78Ca0.02Na0.02(Mg0.34Al1.69Fe(III)0.02)[Si3.35Al0.65]O10(OH)2·nH2O), CaSiO3, 3MgO·4SiO2,H2O, and M(1)M(2)(SiO4)X2 (M(1) = M(2) = Al or M(1) = K, M(2) = Al, X = F or Cl), respectively. The UV-MALDI-Orbitrap mass spectrometry was utilized in solid-state and semi-liquid colloidal state, involving the laser ablation at λex = 337.2 nm. The theoretical modeling and experimental design was based on chemical-, physico-chemical, physical and biological processes involving uranium species under environmental conditions. Therefore, the results reported are crucial for quality control and monitoring programs for assessment of radionuclide migration. They impact significantly the methodology for evaluation of human health risk from radioactive contamination. The study has importance for understanding the coordination and red-ox chemistry of uranium compounds as well. Due to the double nature of uranium between rare element and superconductivity like materials as well as variety of oxidation states ∈ (+1)-(+6), the there remain challenging areas for theoretical and experimental research, which are of significant importance for management of nuclear fuel cycles and waste storage. Copyright © 2014 Elsevier Ltd. All rights reserved.

  13. Preliminary testing of a planar converter with uranium oxide pellets in the emitter

    International Nuclear Information System (INIS)

    Miskolczy, G.; Lieb, D.P.; Hatch, G.L.

    1992-01-01

    Nuclear reactor thermionic space power systems incorporating thermionic fuel element generally use refractory metal emitters, which contain the nuclear fuel. The purpose of the current work is to determine the effect, if any, of the diffusion of uranium oxide fuel through chemically vapor deposited (CVD) tungsten on converter performance. This paper describes the preliminary testing of the converter to assess the converter performance before any significant diffusion takes place. In testing, the emitter temperature was 1800 K and the collector temperature was varied from 1000 K to 1070 K. Experiments also examined pressure versus loading characteristics of the graphite

  14. Chlorination of uranium oxides in melts of alkali metal chlorides and their mixtures

    International Nuclear Information System (INIS)

    Vorobej, M.P.; Bevz, A.S.; Skiba, O.V.

    1978-01-01

    Chlorination of UO 2 , U 3 O 8 , and UO 3 in melts of chlorides of alkali metals and of their mixtures has been studied by thermogravimetric, X-ray phase, and spectrophotometric methods. The thermogravimetric method has been proposed for evaluating the state of uranylcation in the melt; the effect of the composition of the oxide being chlorinated and of the salt-solvent on the composition of the chlorination products has been studied. The effect of the composition of the chlorination products on the stoichiometry of the electrolytic uranium dioxide has been shown

  15. Oxidative destruction of ammonia for restoration of uranium solution mining sites

    Energy Technology Data Exchange (ETDEWEB)

    Humenick, M.J.; Garwacka, K.

    1984-02-01

    A laboratory experimental research project was conducted to evaluate the use of chlorine for the oxidative destruction of residual ammonia that may remain in ground water after in-situ uranium solution mining operations. The work tested the idea of injecting high strength calcium hypochlorite solution into the mining zone to convert ammonia to nitrogen gas as a final cleanup process for ammonia removal from the ground water system. This paper details ammonia removal efficiency as a function of chlorine dose, reactant, and product material balances, and how the concept may be used as a final ground water restoration process.

  16. Oxidative destruction of ammonia for restoration of uranium solution mining sites

    Energy Technology Data Exchange (ETDEWEB)

    Humenick, M.J.; Garwacka, K.

    1984-01-01

    A laboratory experimental research project was conducted to evaluate the use of chlorine for the oxidative destruction of residual ammonia that may remain in ground water after in-situ uranium solution mining operations. The work tested the idea of injecting high strength calcium hypochlorite solution into the mining zone to convert ammonia to nitrogen gas as a final cleanup process for ammonia removal from the ground water system. This paper details ammonia removal efficiency as a function of chlorine dose, reactant, and product material balances, and how the concept may be used as a final ground water restoration process.

  17. Standard guide for the determination of iodine-129 In uranium oxide

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 This method covers the determination of iodine-129 (129I) in uranium oxide by gamma-ray spectrometry. The method could also be applicable to the determination of 129I in aqueous matrices. 1.2 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and to determine the applicability of regulatory limitations prior to use.

  18. Electrical conductivity of molten carbonate and carbonate-chloride systems coexisting with aluminium oxide powder

    Energy Technology Data Exchange (ETDEWEB)

    Nikolaeva, Elena V. [Institute of High Temperature Electrochemistry, Yekaterinburg (Russian Federation); Ural Federal Univ., Yekaterinburg (Russian Federation); Ural State Economic Univ., Yekaterinburg (Russian Federation); Bovet, Andrey L.; Zakiryanova, Irina D. [Institute of High Temperature Electrochemistry, Yekaterinburg (Russian Federation); Ural Federal Univ., Yekaterinburg (Russian Federation)

    2018-04-01

    The electrical properties of composite electrolytes (suspensions) composed of α-Al{sub 2}O{sub 3} powder and molten carbonate eutectic (Li{sub 2}CO{sub 3}-Na{sub 2}CO{sub 3}-K{sub 2}CO{sub 3}){sub eut} or molten carbonate-chloride mixture 0.72(Li{sub 2}CO{sub 3}-Na{sub 2}CO{sub 3}-K{sub 2}CO{sub 3}){sub eut}-0.28NaCl have been investigated by AC impedance method. This system shows a dependence of the electrical conductivity upon the temperature and the α-Al{sub 2}O{sub 3} content. The specific electrical conductivity of the α-Al{sub 2}O{sub 3}/(Li{sub 2}CO{sub 3}-Na{sub 2}CO{sub 3}-K{sub 2}CO{sub 3}){sub eut} system can be adequately described by the Maxwell equation for two-phase heterogeneous materials. The regression equation for the dependence of the specific conductivity of the α-Al{sub 2}O{sub 3}/(Li{sub 2}CO{sub 3}-Na{sub 2}CO{sub 3}-K{sub 2}CO{sub 3}){sub eut} composite on the aluminium oxide concentration and temperature was obtained.

  19. Droplet size prediction in ultrasonic nebulization for non-oxide ceramic powder synthesis.

    Science.gov (United States)

    Muñoz, Mariana; Goutier, Simon; Foucaud, Sylvie; Mariaux, Gilles; Poirier, Thierry

    2018-03-01

    Spray pyrolysis process has been used for the synthesis of non-oxide ceramic powders from liquid precursors in the Si/C/N system. Particles with a high thermal stability and with variable composition and size distribution have been obtained. In this process, the mechanisms involved in precursor decomposition and gas phase recombination of species are still unknown. The final aim of this work consists in improving the whole process comprehension by an experimental/modelling approach that helps to connect the synthesized particles characteristics to the precursor properties and process operating parameters. It includes the following steps: aerosol formation by a piezoelectric nebulizer, its transport and the chemical-physical phenomena involved in the reaction processes. This paper focuses on the aerosol characterization to understand the relationship between the liquid precursor properties and the liquid droplet diameter distribution. Liquids with properties close to the precursor of interest (hexamethyldisilazane) have been used. Experiments have been performed using a shadowgraphy technique to determine the drop size distribution of the aerosol. For all operating parameters of the nebulizer device and liquids used, bimodal droplet size distributions have been obtained. Correlations proposed in the literature for the droplet size prediction by ultrasonic nebulization were used and adapted to the specific nebulizer device used in this study, showing rather good agreement with experimental values. Copyright © 2017 Elsevier B.V. All rights reserved.

  20. Bioaccessibility of micron-sized powder particles of molybdenum metal, iron metal, molybdenum oxides and ferromolybdenum--Importance of surface oxides.

    Science.gov (United States)

    Mörsdorf, Alexander; Odnevall Wallinder, Inger; Hedberg, Yolanda

    2015-08-01

    The European chemical framework REACH requires that hazards and risks posed by chemicals, including alloys and metals, that are manufactured, imported or used in different products (substances or articles) are identified and proven safe for humans and the environment. Metals and alloys need hence to be investigated on their extent of released metals (bioaccessibility) in biologically relevant environments. Read-across from available studies may be used for similar materials. This study investigates the release of molybdenum and iron from powder particles of molybdenum metal (Mo), a ferromolybdenum alloy (FeMo), an iron metal powder (Fe), MoO2, and MoO3 in different synthetic body fluids of pH ranging from 1.5 to 7.4 and of different composition. Spectroscopic tools and cyclic voltammetry have been employed to characterize surface oxides, microscopy, light scattering and nitrogen absorption for particle characterization, and atomic absorption spectroscopy to quantify released amounts of metals. The release of molybdenum from the Mo powder generally increased with pH and was influenced by the fluid composition. The mixed iron and molybdenum surface oxide of the FeMo powder acted as a barrier both at acidic and weakly alkaline conditions. These findings underline the importance of the surface oxide characteristics for the bioaccessibility of metal alloys. Copyright © 2015 The Authors. Published by Elsevier Inc. All rights reserved.

  1. Uranium oxidation kinetics monitored by in-situ X-ray diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Zalkind, S., E-mail: shimonzl@nrcn.org.il; Rafailov, G.; Halevy, I.; Livneh, T.; Rubin, A.; Maimon, H.; Schweke, D.

    2017-03-15

    The oxidation kinetics of U-0.1 wt%Cr at oxygen pressures of 150 Torr and the temperature range of 90–150 °C was studied by means of in-situ X-ray diffraction (XRD). A “breakaway” in the oxidation kinetics is found at ∼0.25 μm, turning from a parabolic to a linear rate law. At the initial stage of oxidation the growth plane of UO{sub 2}(111) is the prominent one. As the oxide thickens, the growth rate of UO{sub 2}(220) plane increases and both planes grow concurrently. The activation energies obtained for the oxide growth are Q{sub parabolic} = 17.5 kcal/mol and Q{sub linear} = 19 kcal/mol. Enhanced oxidation around uranium carbide (UC) inclusions is clearly observed by scanning electron microscopy (SEM).

  2. High quality aluminium doped zinc oxide target synthesis from nanoparticulate powder and characterisation of sputtered thin films

    Energy Technology Data Exchange (ETDEWEB)

    Isherwood, P.J.M., E-mail: P.J.M.Isherwood@lboro.ac.uk [Centre for Renewable Energy Systems Technology, Loughborough University, Loughborough, Leicestershire LE11 3TU (United Kingdom); Neves, N. [Innovnano, S. A., Rua Coimbra Inovação Parque, IParque Lote 13, 3040-570 Antanhol, Coimbra (Portugal); Bowers, J.W. [Centre for Renewable Energy Systems Technology, Loughborough University, Loughborough, Leicestershire LE11 3TU (United Kingdom); Newbatt, P. [Innovnano, S. A., Rua Coimbra Inovação Parque, IParque Lote 13, 3040-570 Antanhol, Coimbra (Portugal); Walls, J.M. [Centre for Renewable Energy Systems Technology, Loughborough University, Loughborough, Leicestershire LE11 3TU (United Kingdom)

    2014-09-01

    Nanoparticulate aluminium-doped zinc oxide powder was synthesised through detonation and subsequent rapid quenching of metallic precursors. This technique allows for precise compositional control and rapid nanoparticle production. The resulting powder was used to form sputter targets, which were used to deposit thin films by radio frequency sputtering. These films show excellent sheet resistance and transmission values for a wide range of deposition temperatures. Crystal structure analysis shows that crystals in the target have a random orientation, whereas the crystals in the films grow perpendicular to the substrate surface and propagate preferentially along the (002) axis. Higher temperature deposition reduces crystal quality with a corresponding decrease in refractive index and an increase in sheet resistance. Films deposited between room temperature and 300 °C were found to have sheet resistances equivalent to or better than indium tin oxide films for a given average transmission value. - Highlights: • Nanoparticulate AZO powder was used to produce sputter targets. • The powder synthesis technique allows for precise compositional control. • Sputtered films show excellent optical, electronic and structural properties. • High temperature films show reduced electrical and structural quality. • For a given transmission, films show equivalent sheet resistances to ITO.

  3. Uranium Immobilization through Fe(II) bio-oxidation: A Column study

    Energy Technology Data Exchange (ETDEWEB)

    Coates, John D.

    2009-09-14

    Current research on the bioremediation of heavy metals and radionuclides is focused on the ability of reducing organisms to use these metals as alternative electron acceptors in the absence of oxygen and thus precipitate them out of solution. However, many aspects of this proposed scheme need to be resolved, not the least of which is the time frame of the treatment process. Once treatment is complete and the electron donor addition is halted, the system will ultimately revert back to an oxic state and potentially result in the abiotic reoxidation and remobilization of the immobilized metals. In addition, the possibility exists that the presence of more electropositive electron acceptors such as nitrate or oxygen will also stimulate the biological oxidation and remobilization of these contaminants. The selective nitrate-dependent biooxidation of added Fe(II) may offer an effective means of “capping off” and completing the attenuation of these contaminants in a reducing environment making the contaminants less accessible to abiotic and biotic reactions and allowing the system to naturally revert to an oxic state. Our previous DOE-NABIR funded studies demonstrated that radionuclides such as uranium and cobalt are rapidly removed from solution during the biogenic formation of Fe(III)-oxides. In the case of uranium, X-ray spectroscopy analysis indicated that the uranium was in the hexavalent form (normally soluble) and was bound to the precipitated Fe(III)-oxides thus demonstrating the bioremediative potential of this process. We also demonstrated that nitrate-dependent Fe(II)- oxidizing bacteria are prevalent in the sediment and groundwater samples collected from sites 1 and 2 and the background site of the NABIR FRC in Oakridge, TN. However, all of these studies were performed in batch experiments in the laboratory with pure cultures and although a significant amount was learned about the microbiology of nitrate-dependent bio-oxidation of Fe(II), the effects of

  4. Unraveling uranium induced oxidative stress related responses in Arabidopsis thaliana seedlings. Part II: responses in the leaves and general conclusions

    Energy Technology Data Exchange (ETDEWEB)

    Vanhoudt, Nathalie, E-mail: nvanhoud@sckcen.be [Belgian Nuclear Research Center (SCK-CEN), Biosphere Impact Studies, Boeretang 200, 2400 Mol (Belgium); Hasselt University, Environmental Biology, Centre for Environmental Sciences, Agoralaan Building D, 3590 Diepenbeek (Belgium); Cuypers, Ann [Hasselt University, Environmental Biology, Centre for Environmental Sciences, Agoralaan Building D, 3590 Diepenbeek (Belgium); Horemans, Nele [Belgian Nuclear Research Center (SCK-CEN), Biosphere Impact Studies, Boeretang 200, 2400 Mol (Belgium); Remans, Tony; Opdenakker, Kelly; Smeets, Karen [Hasselt University, Environmental Biology, Centre for Environmental Sciences, Agoralaan Building D, 3590 Diepenbeek (Belgium); Bello, Daniel Martinez [Hasselt University, Interuniversity Institute for Biostatistics and Statistical Bioinformatics, Agoralaan Building D, 3590 Diepenbeek (Belgium); Havaux, Michel [Commissariat a l' Energie Atomique (CEA)/Cadarache, Direction des Sciences du Vivant, Departement d' Ecophysiologie Vegetale et de Microbiologie, Laboratoire d' Ecophysiologie de la Photosynthese, 13108 Saint-Paul-lez-Durance (France); Wannijn, Jean; Van Hees, May [Belgian Nuclear Research Center (SCK-CEN), Biosphere Impact Studies, Boeretang 200, 2400 Mol (Belgium); Vangronsveld, Jaco [Hasselt University, Environmental Biology, Centre for Environmental Sciences, Agoralaan Building D, 3590 Diepenbeek (Belgium); Vandenhove, Hildegarde [Belgian Nuclear Research Center (SCK-CEN), Biosphere Impact Studies, Boeretang 200, 2400 Mol (Belgium)

    2011-06-15

    The cellular redox balance seems an important modulator under heavy metal stress. While for other heavy metals these processes are well studied, oxidative stress related responses are also known to be triggered under uranium stress but information remains limited. This study aimed to further unravel the mechanisms by which plants respond to uranium stress. Seventeen-day-old Arabidopsis thaliana seedlings, grown on a modified Hoagland solution under controlled conditions, were exposed to 0, 0.1, 1, 10 and 100 {mu}M uranium for 1, 3 and 7 days. While in Part I of this study oxidative stress related responses in the roots were discussed, this second Part II discusses oxidative stress related responses in the leaves and general conclusions drawn from the results of the roots and the leaves will be presented. As several responses were already visible following 1 day exposure, when uranium concentrations in the leaves were negligible, a root-to-shoot signaling system was suggested in which plastids could be important sensing sites. While lipid peroxidation, based on the amount of thiobarbituric acid reactive compounds, was observed after exposure to 100 {mu}M uranium, affecting membrane structure and function, a transient concentration dependent response pattern was visible for lipoxygenase initiated lipid peroxidation. This transient character of uranium stress responses in leaves was emphasized by results of lipoxygenase (LOX2) and antioxidative enzyme transcript levels, enzyme capacities and glutathione concentrations both in time as with concentration. The ascorbate redox balance seemed an important modulator of uranium stress responses in the leaves as in addition to the previous transient responses, the total ascorbate concentration and ascorbate/dehydroascorbate redox balance increased in a concentration and time dependent manner. This could represent either a slow transient response or a stable increase with regard to plant acclimation to uranium stress

  5. Unraveling uranium induced oxidative stress related responses in Arabidopsis thaliana seedlings. Part II: responses in the leaves and general conclusions.

    Science.gov (United States)

    Vanhoudt, Nathalie; Cuypers, Ann; Horemans, Nele; Remans, Tony; Opdenakker, Kelly; Smeets, Karen; Bello, Daniel Martinez; Havaux, Michel; Wannijn, Jean; Van Hees, May; Vangronsveld, Jaco; Vandenhove, Hildegarde

    2011-06-01

    The cellular redox balance seems an important modulator under heavy metal stress. While for other heavy metals these processes are well studied, oxidative stress related responses are also known to be triggered under uranium stress but information remains limited. This study aimed to further unravel the mechanisms by which plants respond to uranium stress. Seventeen-day-old Arabidopsis thaliana seedlings, grown on a modified Hoagland solution under controlled conditions, were exposed to 0, 0.1, 1, 10 and 100 μM uranium for 1, 3 and 7 days. While in Part I of this study oxidative stress related responses in the roots were discussed, this second Part II discusses oxidative stress related responses in the leaves and general conclusions drawn from the results of the roots and the leaves will be presented. As several responses were already visible following 1 day exposure, when uranium concentrations in the leaves were negligible, a root-to-shoot signaling system was suggested in which plastids could be important sensing sites. While lipid peroxidation, based on the amount of thiobarbituric acid reactive compounds, was observed after exposure to 100 μM uranium, affecting membrane structure and function, a transient concentration dependent response pattern was visible for lipoxygenase initiated lipid peroxidation. This transient character of uranium stress responses in leaves was emphasized by results of lipoxygenase (LOX2) and antioxidative enzyme transcript levels, enzyme capacities and glutathione concentrations both in time as with concentration. The ascorbate redox balance seemed an important modulator of uranium stress responses in the leaves as in addition to the previous transient responses, the total ascorbate concentration and ascorbate/dehydroascorbate redox balance increased in a concentration and time dependent manner. This could represent either a slow transient response or a stable increase with regard to plant acclimation to uranium stress. Copyright

  6. Effect of salt, kinnow and pomegranate fruit by-product powders on color and oxidative stability of raw ground goat meat during refrigerated storage.

    Science.gov (United States)

    Devatkal, Suresh K; Naveena, B M

    2010-06-01

    Effects of salt, kinnow and pomegranate fruit by-product powders on color and oxidative stability of raw ground goat meat stored at 4+/-1 degrees C was evaluated. Five treatments evaluated include: control (only meat), MS (meat+2% salt), KRP (meat+2% salt+2% kinnow rind powder), PRP (meat+2% salt+2% pomegranate rind powder) and PSP (meat+2% salt+2% pomegranate seed powder). Addition of salt resulted in reduction of redness scores. Lightness increased in control and unchanged in others during storage. Redness scores declined and yellowness showed inconsistent changes during storage. Thiobarbituric acid reactive substances (TBARS) values were higher (PPRP>KRP>control>MS. Therefore, these powders have potential to be used as natural antioxidants to minimize the auto-oxidation and salt induced lipid oxidation in raw ground goat meat. Crown Copyright 2010. Published by Elsevier Ltd. All rights reserved.

  7. Influent of Carbonization of Sol Solution at the External Gelation Process on the Quality of Uranium Oxide Kernel

    International Nuclear Information System (INIS)

    Damunir; Sukarsono

    2007-01-01

    The influent of carbonization of sol solution at the external gelation process on the quality of uranium oxide kernel was done. Variables observed are the influent of carbon, temperature and time of reduction process of U 3 O 8 kernel resulted from carbonization of sol solution. First of all, uranyl nitrate was reacted with 1 M NH 4 OH solution, producing the colloid of UO 3 . Then by mixing and heating up to the temperature of 60-80 °C, the colloid solution was reacted with PVA, mono sorbitol oleate and paraffin producing of uranium-PVA sol. Then sol solution was carbonized with carbon black of mol ratio of carbon to uranium =2.32-6.62, produce of carbide gel. Gel then washed, dried and calcined at 800 °C for 4 hours to produce of U 3 O 8 kernel containing carbon. Then the kernel was reduced by H 2 gas in the medium of N 2 gas at 500-800 °C, 50 mmHg pressure for 3 hours. The process was repeated at 700 °C, 50 mmHg pressure for 1-4 hours. The characterization of chemical properties of the gel grains and uranium oxide kernel using FTIR covering the analysis of absorption band of infra red spectrum of UO 3 , C-OH, NH 3 , C-C, C-H and OH functional group. The physical properties of uranium oxide covering specific surface area, void volume, mean diameter using surface area meter Nova-1000 and as N 2 gas an absorbent. And O/U ratio of uranium dioxide kernel by gravimetry method. The result of experiment showed that carbonization of sol solution at the external gelation process give influencing the quality of uranium oxide kernel. (author)

  8. Characteristics of interlayer oxidation zone and uranium metallogenetic prospect of Zhiluo formation in Daliuta area, Ordos Basin

    International Nuclear Information System (INIS)

    Yang Jianxin; Li Xide; Zhang Zhaolin

    2006-01-01

    Ordos Basin is a large down-warping basin in the northwest of North-China Platform; Daliuta area is located in the northeast of Ordos Basin. In this area, sand bodies of fluvial facies developed well in the submember of the lower member of the target Zhiluo Formation of Middle Jurassic and several sand belt of large scale occurred. Yellow interlayer oxidation zone have been discovered in belt I and belt III by the drilling and it is of a certain scale. Due to the young age of interlayer oxidation and unsatisfied uranium sources, uranium metallogenic prospect of this area need more research and exploration. (authors)

  9. Towards uranium dose-effects relationships for bio markers of oxidative stress in fish

    International Nuclear Information System (INIS)

    Buet, A.; Camilleri, V.; Simon, O.; Garnier-Laplace, J.

    2004-01-01

    Within the Envirhom program, the bioavailability of uranium, a widely spread metal in fresh waters, has been studied in various organisms to gain understanding of metal-organisms interactions. Experiments are still in progress to establish a comprehensive basis of early and delayed involved toxicological mechanisms. Uranium is known to be both radio-toxic and chemo-toxic. The cellular damage of radiation, but also of heavy metal exposure, is mainly associated with an oxidative injury due to the formation of reactive oxygen species. Moreover, the intensity of oxidative damages is dependant on the efficiency of antioxidant defense systems of the organism. In this context, short-term experiments were performed with juvenile rainbow trout (Onchorynchus mykiss) in order to (1) assess the response of some antioxidant parameters and other potential bio-markers and (2) to better characterize the kinetics of the responses in relation with U concentrations and exposure duration. Trout were exposed by direct pathway to a range of U concentrations in water (low, medium and high: 20, 100 and 500 μg U.L -1 respectively) during 10 days. Several antioxidant parameters were measured: the rate of reduced glutathione (GSH) that plays a major role in cellular detoxification and antioxidant defense, and the activities of superoxide dismutases (SOD) and catalase that are involved in the detoxification of oxygen reactive species. The activity of glutathione reductase (GR), that restores the pool of GSH was measured, as well as the activity of acetylcholinesterase (AChE), a marker of neurotoxicity. In parallel, U analysis were performed in gill, skin, muscle, skeleton, intestine, liver, kidneys and body residues of exposed trout in order to assess the dependence of biological responses with a potential uranium bioaccumulation in fish tissues. (author)

  10. Tables of thermodynamic functions for gaseous thorium, uranium, and plutonium oxides

    International Nuclear Information System (INIS)

    Green, D.W.

    1980-03-01

    Measured and estimated spectroscopic data for thorium, uranium, and plutonium oxide vapor species have been used with the methods of statistical mechanics to calculate thermodynamic functions. Some inconsistencies between spectroscopic data and some thermodynamic data have been resolved by recalculating ΔH 0 /sub f/ (298.15 0 K) values for the vapor species of these oxides. Evaluation of the uncertainties in data, methods of estimating molecular parameters, and effects of assumptions have been discussed elsewhere. The tables of thermodynamic functions that were reported earlier have been revised principally because the low-frequency vibrational modes of UO 2 and UO 3 have now been measured. These new empirical data resulted in changes in the electronic contributions to the calculated thermodynamic functions of UO 2 and the estimated vibrational contributions for PuO 2 . In addition, some minor changes have been made in the methods of calculation of the electronic contributions for all molecules

  11. Surface Area, and Oxidation Effects on Nitridation Kinetics of Silicon Powder Compacts

    Science.gov (United States)

    Bhatt, R. T.; Palczer, A. R.

    1998-01-01

    Commercially available silicon powders were wet-attrition-milled from 2 to 48 hr to achieve surface areas (SA's) ranging from 1.3 to 70 sq m/g. The surface area effects on the nitridation kinetics of silicon powder compacts were determined at 1250 or 1350 C for 4 hr. In addition, the influence of nitridation environment, and preoxidation on nitridation kinetics of a silicon powder of high surface area (approximately equals 63 sq m/g) was investigated. As the surface area increased, so did the percentage nitridation after 4 hr in N2 at 1250 or 1350 C. Silicon powders of high surface area (greater than 40 sq m/g) can be nitrided to greater than 70% at 1250 C in 4 hr. The nitridation kinetics of the high-surface-area powder compacts were significantly delayed by preoxidation treatment. Conversely, the nitridation environment had no significant influence on the nitridation kinetics of the same powder. Impurities present in the starting powder, and those accumulated during attrition milling, appeared to react with the silica layer on the surface of silicon particles to form a molten silicate layer, which provided a path for rapid diffusion of nitrogen and enhanced the nitridation kinetics of high surface area silicon powder.

  12. Absolute determination by X-ray diffraction of a binary or ternary mixture: nickel oxide and fluoride in a nickel powder (1960)

    International Nuclear Information System (INIS)

    Charpin, P.; Hauptman, A.

    1960-01-01

    The method employed is based upon the comparison between computed and measured intensities for conveniently selected X-Ray diffraction lines of each component of the powder. Care must be taken to allow for absorption, both inside each grain and in overall sample. This method has been applied to the determination of nickel oxide and fluoride in a nickel powder. (author) [fr

  13. The cathodic reduction of dioxygen on uranium oxide in dilute alkaline aqueous solution

    International Nuclear Information System (INIS)

    Hocking, W.H.; Betteridge, J.S.; Shoesmith, D.W.

    1991-09-01

    The cathodic reduction of dioxygen on uranium oxide in dilute alkaline aqueous solutions has been investigated within the context of a program to develop a comprehensive model to predict the behaviour of used CANDU (Canada Deuterium Uranium) nuclear fuel under disposal-vault conditions. Two different kinds of ceramic UO 2 were studied: reactor-grade CANDU fuel with normal p-type electrical conductivity and low-resistance material that exhibits n-type photoelectrochemical behaviour. The transport of electroactive species in solution was controlled by varying the rotation rate of rotating disc electrodes (RDE) and rotating ring-disc electrodes (RRDE). Steady-state polarization measurements were made using the current-interrupt method to compensate for the potential drop caused by ohmic resistance. Any release of peroxide to solution from the UO 2 (disc) surface could be monitored by oxidizing it at the Au ring of an RRDE. The existing theory for the cathodic 0 2 -reduction process as applied to RDE and RRDE experiments has been reviewed as a starting point for the interpretation of the results obtained in our work. (37 figs., 2 tabs., 170 refs.)

  14. Implication of volume changes in uranium oxides: A density functional study

    International Nuclear Information System (INIS)

    Szpunar, B.; Szpunar, J.A.; Milman, V.; Goldberg, A.

    2013-01-01

    In severe nuclear accident scenarios (in air environments and high temperatures) UO 2 fuel pellets oxidise to produce uranium oxides with higher oxygen content, e.g., U 4 O 9 or U 3 O 8 . As a first step in investigating the microstructural changes following UO 2 oxidation to hexagonal high temperature phase of U 3 O 8 , density functional quantum mechanical calculations of the structure, elastic properties and electronic structure of U 3 O 8 have been performed. The calculated properties of hexagonal phase of U 3 O 8 are compared to those of the orthorhombic pseudo-hexagonal phase which is stable at room temperature. The total energy technique based on the local density approximation plus Hubbard U as implemented in the CASTEP code is used to investigate changes in the lattice constants. The first-principles calculations predict a 35-42% increase in volume per uranium atom as a result of the transformation from UO 2 to U 3 O 8 , in agreement with experimental data. The implications of this prediction on the linear expansion and fragmentation of fuel are discussed. (authors)

  15. Inorganic oxides as alternative in the separation of non fissioned residual uranium

    International Nuclear Information System (INIS)

    Baca G, A.

    1997-01-01

    The Al 2 O 3 , SiO 2 and SnO 2 as well as vegetable carbon have been studied for its possible use as sorbent in the concentration and separation of non fissioned residual uranium of some fission products such as: 141 Ce, 134 Cs, 125 Sb, 103 Ru, 95 Zr, 95 Nb of alkaline aqueous systems. The separation efficiency has been evaluated using natural uranium and radionuclides in static and dynamic processes, through liquid scintillation and gamma spectrometry. Therefore Al 2 O 3 , SiO 2 , SnO 2 and carbon were pre-treated thermic and chemically and characterized through the technique of Nitrogen absorption analysis, X-ray diffraction and IR spectroscopy. By means of the p H determination and the aqueous system potential the present hydrolysis products were determined. The inorganic oxides show structural and surface changes due to the treatment. The adsorption process is realized by different mechanism depending of the sorbent. The results show that the retention capacity is a dependence of the oxides pre-treatment and of the hydrolysis products in the aqueous system, as well as of the experimental conditions. Not in this way for carbon in which the results show the treatment and the experimental conditions significantly have not influence in its adsorption capacity. (Author)

  16. Guideline tests on the corrosion of sintered uranium oxide by lead and sodium about 450 deg. C; Essais d'orientation sur la corrosion de l'oxyde d'uranium fritte par le plomb et le sodium aux environs de 450 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Portnoff, A.; Pointud, R.

    1958-05-15

    Within the frame of the investigation of behaviour of a fuel element (such as UO{sub 2}) under irradiation, the authors report the study of the physical-chemical action of the coolant at different temperatures on the body to be irradiated. Thus, sintered uranium oxide has been submitted to corrosion by lead and by sodium during 250 hours at temperatures between 400 and 500 C. The physical characteristics of the UO{sub 2} powder and of different sintered UO{sub 2} pellets produced from this powder under different sintering processes are indicated, as well as the results of a spectrographic analysis of the sintered UO{sub 2}. Corrosion devices, treatments and obtained results are reported for corrosion by lead and by sodium. In the latter case, extraction processes are discussed (using butyl alcohol, or melting in vaseline oil)

  17. Mineralogical-geochemical specificity of the uranium mineralization superposed on the oxidized rocks

    International Nuclear Information System (INIS)

    Bulatov, S.G.; Shchetochkin, V.N.

    1975-01-01

    Taking as an example a uranium deposit connected with oxidation zones developing along the strata, the author examines the mineralogical and geochemical features of a pitchblende-sooty uraninite mineralization superimposed on limonitized sandstones. The typical relations between ore mineralization with new formations of the infiltration oxidation process and the changes caused by the action of rising thermal solutions on the rocks are given. Based on these relations, two generations of different ages of rich pitchblende-sooty uraninite ores are distinguished, separated by the time of development of the oxidation processes. The typical change around the ore is a reduction of limonitized rocks, accompanied by their pyritization, clarification and hematitization. The ore concentrations were formed as a result of the action of rising thermal solutions that had interacted with oxidized rocks. The development of late oxidation processes caused the redistribution of these ore concentrations and their downward shift along the stratum slope following the limonitization boundary. On the basis of the data presented, comments of a forecasting and prospecting nature are made. (author)

  18. A study of integrated cathode assembly for electrolytic reduction of uranium oxide in LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Park, Sung Bin; Seo, Jung Seok; Kang, Dae Seung; Kwon, Sun Kil; Park, Seong Won

    2004-01-01

    Interest of electrolytic reduction of uranium oxide is increasing in treatment of spent metal fuels. Argonne National Laboratory (ANL) has reported the experimental results of electrochemical reduction of uranium oxide fuel in bench-scale apparatus with cyclic voltammetry, and has designed high-capacity reduction (HCR) cells and conducted three kg-scale UO 2 reduction runs. From the cyclic voltammograms, the mechanism of electrolytic reduction of metal oxides is analyzed. The uranium oxide in LiCl-Li 2 O is converted to uranium metal according to the two mechanism; direct and indirect electrolytic reduction. In this study, cyclic voltammograms for LiCl-3wt% Li 2 O system and U 3 O 8 -LiCl-3wt% Li 2 O system using the 325-mesh stainless steel screen in cathode assembly have been obtained. Direct electrolytic reduction of uranium oxide in LiCl-3wt% Li 2 O molten salt has been conducted

  19. Characteristics of an integrated cathode assembly for the electrolytic reduction of uranium oxide in a LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Sung Bin Park; Byung Heung Park; Sang Mun Jeong; Jin Mok Hur; Chung Seok Seo; Seong Won Park; Seung-Hoon Choi

    2006-01-01

    Electrochemical behavior of the reduction of uranium oxide was studied in a LiCl-Li 2 O molten salt system with an integrated cathode assembly. The mechanism for the electrolytic reduction of uranium oxide was studied through cyclic voltammetry. By means of a chronopotentiometry, the effects of the thickness of the uranium oxide, the thickness of the MgO membrane and the material of the conductor of an integrated cathode assembly on the overpotential of the cathode were investigated. From the voltamograms, the reduction potential of the uranium oxide and Li 2 O was obtained and the two mechanisms of the electrolytic reduction were considered with regard to the applied cathode potential. From the chronopotentiograms, the exchange current, the transfer coefficient and the maximum allowable current based on the Tafel behavior were obtained with regard to the thickness of the uranium oxide, and of the MgO membrane and the material of the conductor of an integrated cathode assembly. (author)

  20. The combined effect of uranium and gamma radiation on biological responses and oxidative stress induced in Arabidopsis thaliana

    International Nuclear Information System (INIS)

    Vanhoudt, Nathalie; Vandenhove, Hildegarde; Horemans, Nele; Wannijn, Jean; Van Hees, May; Vangronsveld, Jaco; Cuypers, Ann

    2010-01-01

    Uranium never occurs as a single pollutant in the environment, but always in combination with other stressors such as ionizing radiation. As effects induced by multiple contaminants can differ markedly from the effects induced by the individual stressors, this multiple pollution context should not be neglected. In this study, effects on growth, nutrient uptake and oxidative stress induced by the single stressors uranium and gamma radiation are compared with the effects induced by the combination of both stressors. By doing this, we aim to better understand the effects induced by the combined stressors but also to get more insight in stressor-specific response mechanisms. Eighteen-day-old Arabidopsis thaliana seedlings were exposed for 3 days to 10 μM uranium and 3.5 Gy gamma radiation. Gamma radiation interfered with uranium uptake, resulting in decreased uranium concentrations in the roots, but with higher transport to the leaves. This resulted in a better root growth but increased leaf lipid peroxidation. For the other endpoints studied, effects under combined exposure were mostly determined by uranium presence and only limited influenced by gamma presence. Furthermore, an important role is suggested for CAT1/2/3 gene expression under uranium and mixed stressor conditions in the leaves.