National Library of Energy BETA

Sample records for uranium oxide lb

  1. Effect of ammonia plasma treatment on graphene oxide LB monolayers

    SciTech Connect (OSTI)

    Singh, Gulbagh; Botcha, V. Divakar; Narayanam, Pavan K.; Sutar, D. S.; Talwar, S. S.; Major, S. S.; Srinivasa, R. S.

    2013-02-05

    Graphene oxide monolayer sheets were transferred on Si and SiO{sub 2}/Si substrates by Langmuir-Blodgett technique and were exposed to ammonia plasma at room temperature. The monolayer character of both graphene oxide and plasma treated graphene oxide sheets were ascertained by atomic force microscopy. X-ray photoelectron spectroscopy and Raman spectroscopy revealed that ammonia plasma treatment results in enhancement of graphitic carbon content along with the incorporation of nitrogen. The conductivity of graphene oxide monolayers, which was in the range of 10{sup -6}-10{sup -7} S/cm, increased to 10{sup -2}-10{sup -3} S/cm after the ammonia plasma treatment. These results indicate that the graphene oxide was simultaneously reduced and N-doped during ammonia plasma treatment, without affecting the morphological stability of sheets.

  2. Method for converting uranium oxides to uranium metal

    DOE Patents [OSTI]

    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.

  3. PROCESS OF PRODUCING REFRACTORY URANIUM OXIDE ARTICLES

    DOE Patents [OSTI]

    Hamilton, N.E.

    1957-12-01

    A method is presented for fabricating uranium oxide into a shaped refractory article by introducing a uranium halide fluxing reagent into the uranium oxide, and then mixing and compressing the materials into a shaped composite mass. The shaped mass of uranium oxide and uranium halide is then fired at an elevated temperature so as to form a refractory sintered article. It was found in the present invention that the introduction of a uraninm halide fluxing agent afforded a fluxing action with the uranium oxide particles and that excellent cohesion between these oxide particles was obtained. Approximately 90% of uranium dioxide and 10% of uranium tetrafluoride represent a preferred composition.

  4. Influence of uranium hydride oxidation on uranium metal behaviour

    SciTech Connect (OSTI)

    Patel, N.; Hambley, D.; Clarke, S.A.; 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)

  5. CATALYZED OXIDATION OF URANIUM IN CARBONATE SOLUTIONS

    DOE Patents [OSTI]

    Clifford, W.E.

    1962-05-29

    A process is given wherein carbonate solutions are employed to leach uranium from ores and the like containing lower valent uranium species by utilizing catalytic amounts of copper in the presence of ammonia therein and simultaneously supplying an oxidizing agent thereto. The catalysis accelerates rate of dissolution and increases recovery of uranium from the ore. (AEC)

  6. Oxidation and crystal field effects in uranium

    SciTech Connect (OSTI)

    Tobin, J. G.; Booth, C. H.; Shuh, D. K.; van der Laan, G.; Sokaras, D.; Weng, T. -C.; Yu, S. W.; Bagus, P. S.; Tyliszczak, T.; Nordlund, D.

    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.

  7. Method for fluorination of uranium oxide

    DOE Patents [OSTI]

    Petit, George S. (Oak Ridge, TN)

    1987-01-01

    Highly pure uranium hexafluoride is made from uranium oxide and fluorine. The uranium oxide, which includes UO.sub.3, UO.sub.2, U.sub.3 O.sub.8 and mixtures thereof, is introduced together with a small amount of a fluorine-reactive substance, selected from alkali chlorides, silicon dioxide, silicic acid, ferric oxide, and bromine, into a constant volume reaction zone. Sufficient fluorine is charged into the zone at a temperature below approximately 0.degree. C. to provide an initial pressure of at least approximately 600 lbs/sq. in. at the ambient atmospheric temperature. The temperature is then allowed to rise in the reaction zone until reaction occurs.

  8. Uranium Oxide Aerosol Transport in Porous Graphite

    SciTech Connect (OSTI)

    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 reactors lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

  9. Table 4.10 Uranium Reserves, 2008 (Million Pounds Uranium Oxide)

    U.S. Energy Information Administration (EIA) Indexed Site

    0 Uranium Reserves,1 2008 (Million Pounds Uranium Oxide) State Forward-Cost 2 Category (dollars 3 per pound) $50 or Less $100 or Less Total 539 1,227 Wyoming 220 446 New Mexico 179 390 Arizona, Colorado, Utah 63 198 Texas 27 40 Others 4 50 154 1The U.S. Energy Information Administration (EIA) category of uranium reserves is equivalent to the internationally reported category of "Reasonably Assured Resources" (RAR). Notes: * Estimates are at end of year. * See "Uranium Oxide"

  10. Analysis of uranium oxide weathering by molecular spectroscopy. Final report

    SciTech Connect (OSTI)

    Zickafoose, M.S.

    1997-11-01

    A preliminary study of the weathering of uranium oxide particles diluted in diamond dust at ambient environmental conditions is presented. The primary weathering reaction is oxidation of the uranium from the +4 to +6 oxidation state, although formation of compounds such as carbonates and hydroxides is possible. Identification of the state of uranium oxide has been attempted using luminescence spectroscopy and diffuse reflectance Fourier transform infrared spectroscopy (DRIFTS). Luminescence spectra of nominal samples of three common oxides, UO3, U3O8, and UO2, have been measured showing significant spectral differences in peaks at 494 nm, 507 nm, 529 nm, and 553 nm. DRIFTS spectra of the same three oxides show significant differences in peaks at 960 /cm, 856 /cm, and 754 /cm. The differences in these peaks allow determination of the oxidation to the +6 state in these compounds.

  11. Oxidation and crystal field effects in uranium (Journal Article...

    Office of Scientific and Technical Information (OSTI)

    5, 2016 Title: Oxidation and crystal field effects in uranium Authors: Tobin, J. G. ; Yu, S.-W. ; Booth, C. H. ; Tyliszczak, T. ; Shuh, D. K. ; van der Laan, G. ; Sokaras, D. ;...

  12. Covalency in oxidized uranium (Journal Article) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Title: Covalency in oxidized uranium Authors: Tobin, J. G. ; Yu, S.-W. ; Qiao, R. ; Yang, W. L. ; Booth, C. H. ; Shuh, D. K. ; Duffin, A. M. ; Sokaras, D. ; Nordlund, D. ;...

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

    SciTech Connect (OSTI)

    Graves, Christopher R; Kiplinger, Jaqueline L

    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.

  14. Oxidation Protection of Uranium Nitride Fuel using Liquid Phase Sintering

    SciTech Connect (OSTI)

    Dr. Paul A. Lessing

    2012-03-01

    Two methods are proposed to increase the oxidation resistance of uranium nitride (UN) nuclear fuel. These paths are: (1) Addition of USi{sub x} (e.g. U3Si2) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with various compounds (followed by densification via Spark Plasma Sintering or Liquid Phase Sintering) that will greatly increase oxidation resistance. The advantages (high thermal conductivity, very high melting point, and high density) of nitride fuel have long been recognized. The sodium cooled BR-10 reactor in Russia operated for 18 years on uranium nitride fuel (UN was used as the driver fuel for two core loads). However, the potential advantages (large power up-grade, increased cycle lengths, possible high burn-ups) as a Light Water Reactor (LWR) fuel are offset by uranium nitride's extremely low oxidation resistance (UN powders oxidize in air and UN pellets decompose in hot water). Innovative research is proposed to solve this problem and thereby provide an accident tolerant LWR fuel that would resist water leaks and high temperature steam oxidation/spalling during an accident. It is proposed that we investigate two methods to increase the oxidation resistance of UN: (1) Addition of USi{sub x} (e.g. U{sub 3}Si{sub 2}) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with compounds (followed by densification via Spark Plasma Sintering) that will greatly increase oxidation resistance.

  15. CRAD, Management- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Management program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  16. CRAD, Training- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Office of Energy Efficiency and Renewable Energy (EERE)

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Training Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  17. COLUMBIC OXIDE ADSORPTION PROCESS FOR SEPARATING URANIUM AND PLUTONIUM IONS

    DOE Patents [OSTI]

    Beaton, R.H.

    1959-07-14

    A process is described for separating plutonium ions from a solution of neutron irradiated uranium in which columbic oxide is used as an adsorbert. According to the invention the plutonium ion is selectively adsorbed by Passing a solution containing the plutonium in a valence state not higher than 4 through a porous bed or column of granules of hydrated columbic oxide. The adsorbed plutonium is then desorbed by elution with 3 N nitric acid.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  20. METHOD OF PREPARING URANIUM, THORIUM, OR PLUTONIUM OXIDES IN LIQUID BISMUTH

    DOE Patents [OSTI]

    Davidson, J.K.; Robb, W.L.; Salmon, O.N.

    1960-11-22

    A method is given for forming compositions, as well as the compositions themselves, employing uranium hydride in a liquid bismuth composition to increase the solubility of uranium, plutonium and thorium oxides in the liquid bismuth. The finely divided oxide of uranium, plutonium. or thorium is mixed with the liquid bismuth and uranium hydride, the hydride being present in an amount equal to about 3 at. %, heated to about 5OO deg C, agitated and thereafter cooled and excess resultant hydrogen removed therefrom.

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

    SciTech Connect (OSTI)

    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

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

    SciTech Connect (OSTI)

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

    1994-08-01

    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[sub 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[sub 4] to produce CF[sub 4] in addition to the reduction of UO[sub 2], but the fraction of metal from the reduction of UF[sub 4] can be decreased by increasing the concentration of dissolved UO[sub 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[sub 4].

  3. EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio

    Energy Savers [EERE]

    Site | Department of Energy 60: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio Site EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio Site Summary This site-specific EIS analyzes the construction, operation, maintenance, and decontamination and decommissioning of the proposed depleted uranium hexafluoride (DUF6) conversion facility at three alternative locations within the Portsmouth site; transportation of all cylinders (DUF6, enriched, and

  4. LB CRAC Workshops (rates/meetings)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Load-Based (LB) CRAC Power Rate Adjustment Workshop Materials Related Links: Power Rate Adjustments > Load-Based (LB) CRAC December 13, 2006 LB CRAC Workshop Materials (updated...

  5. Dissolution of uranium oxides from simulated environmental swipes using ammonium bifluoride

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Meyers, Lisa A.; Yoshida, Thomas M.; Chamberlin, Rebecca M.; Xu, Ning

    2016-04-09

    We developed an analytical chemistry method to quantitatively recover microgram quanties of solid uranium oxides from swipe media using ammonium bifluoride (ABF, NH4HF2) solution. Recovery of uranium from surrogate swipe media (filter paper) was demonstrated at initial uranium loading levels between 3 and 20 µg filter-1. Moreover, the optimal conditions for extracting U3O8 and UO2 are using 1 % ABF solution and incubating at 80 °C for one hour. The average uranium recoveries are 100 % for U3O8, and 90 % for UO2. Finally, with this method, uranium concentration as low as 3 µg filter-1 can be recovered for analysis.

  6. Oxidation behaviour of uranium and neptunium in stabilised zirconia

    SciTech Connect (OSTI)

    Walter, Marcus; Somers, Joseph; Bouexiere, Daniel; Gaczynski, Piotr; Brendebach, Boris

    2009-12-15

    Yttria stabilised zirconia (YSZ) based (Zr,Y,U)O{sub 2-x} and (Zr,Y,Np)O{sub 2-x} solid solutions with 6 and 20 mol% actinide were prepared with Y/Zr ratios ranging from 0.2 to 2.0 to investigate uranium and neptunium oxidation behaviour depending on the oxygen vacancies in the defect fluorite lattice. Sintering at 1600 deg. C in Ar/H{sub 2} yields a cubic, fluorite-type structure with U(IV) and Np(IV). Annealing (Zr,Y,U)O{sub 2-x} with Y/Zr=0.2 at 800 deg. C in air results in a tetragonal phase, whereas (Zr,Y,U)O{sub 2-x} with higher Y/Zr ratios and (Zr,Y,Np)O{sub 2-x} retain the cubic structure. XANES and O/M measurements indicate mixed U(V)-U(VI) and Np(IV)-Np(V) oxidation states after oxidation. Based on X-ray diffraction, O/M and EXAFS measurements, different oxidation mechanisms are identified for U- and Np-doped stabilised zirconia. In contrast to U, excess oxygen vacancies are needed to oxidise Np in (Zr,Y,Np)O{sub 2-x} as the oxidation process competes with Zr for oxygen vacancies. As a consequence, U(VI) and Np(V) can only be obtained in stabilised zirconia with Y/Zr=1 but not in YSZ with Y/Zr=0.2. - Graphical abstract: The O/U ratio in oxidised (Zr,Y,U)O{sub 2-x} depends on the Y/U ratio, whereas O/Np in (Zr,Y,Np)O{sub 2-x} correlates with the Y/(Zr+Np) ratio. This indicates that both Zr and Np compete for oxygen vacancies, which hinders the Np oxidation at low Y/Zr ratios. Display Omitted

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

    SciTech Connect (OSTI)

    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.

  8. Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

    DOE Patents [OSTI]

    Herrmann, Steven Douglas

    2014-05-27

    Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

  9. Stability of uranium incorporated into Fe(hydr)oxides under fluctuating redox conditions

    SciTech Connect (OSTI)

    Stewart, B.D.; Nico, P.S.; Fendorf, S.

    2009-04-01

    Reaction pathways resulting in uranium bearing solids that are stable (i.e., having limited solubility) under both aerobic and anaerobic conditions will limit dissolved concentrations and migration of this toxin. Here we examine the sorption mechanism and propensity for release of uranium reacted with Fe (hydr)oxides under cyclic oxidizing and reducing conditions. Upon reaction of ferrihydrite with Fe(II) under conditions where aqueous Ca-UO{sub 2}-CO{sub 3} species predominate (3 mM Ca and 3.8 mM CO{sub 3}-total), dissolved uranium concentrations decrease from 0.16 mM to below detection limit (BDL) after 5 to 15 d, depending on the Fe(II) concentration. In systems undergoing 3 successive redox cycles (15 d of reduction followed by 5 d of oxidation) and a pulsed decrease to 0.15 mM CO{sub 3}-total, dissolved uranium concentrations varied depending on the Fe(II) concentration during the initial and subsequent reduction phases - U concentrations resulting during the oxic 'rebound' varied inversely with the Fe(II) concentration during the reduction cycle. Uranium removed from solution remains in the oxidized form and is found both adsorbed on and incorporated into the structure of newly formed goethite and magnetite. Our 15 results reveal that the fate of uranium is dependent on anaerobic/aerobic conditions, aqueous uranium speciation, and the fate of iron.

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

    SciTech Connect (OSTI)

    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.

  11. uranium

    National Nuclear Security Administration (NNSA)

    to prepare surplus plutonium for disposition, and readiness to begin the Second Uranium Cycle, to start processing spent nuclear fuel.

    H Canyon is also being...

  12. Multiple reaction fronts in the oxidation-reduction of iron-rich uranium ores

    SciTech Connect (OSTI)

    Dewynne, J.N. . Faculty of Mathematical Studies); Fowler, A.C. . Mathematical Inst.); Hagan, P.S. )

    1993-08-01

    When a container of radioactive waste is buried underground, it eventually corrodes, and leakage of radioactive material to the surrounding rock occurs. Depending on the chemistry of the rock, many different reactions may occur. A particular case concerns the oxidation and reduction of uranium ores by infiltrating groundwater, since UO[sub 3] is relatively soluble (and hence potentially transportable to the water supply), whereas UO[sub 2] is essentially insoluble. It is therefore of concern to those involved with radioactive waste disposal to understand the mechanics of uranium transport through reduction and oxidation reactions. This paper describes the oxidation of iron-rich uranium-bearing rocks by infiltration of groundwater. A reaction-diffusion model is set up to describe the sequence of reactions involving iron oxidation, uranium oxidation and reduction, sulfuric acid production, and dissolution of the host rock that occur. On a geological timescale of millions of years, the reactions occur very fast in very thin reaction fronts. It is shown that the redox front that separates oxidized (orange) rock from reduced (black) rock must actually consist of two separate fronts that move together, at which the two separate processes of uranium oxidation and iron reduction occur, respectively. Between these fronts, a high concentration of uranium is predicted. The mechanics of this process are not specific to uranium-mediated redox reactions, but apply generally and may be used to explain the formation of concentrated ore deposits in extended veins. On the long timescales of relevance, a quasi-static response results, and the problem can be solved explicitly in one dimension. This provides a framework for studying more realistic two-dimensional problems in fissured rocks and also for the future study of uraninite nodule formation.

  13. Oxidation behavior and segregation of uranium in the intermetallic compound UFe/sub 2/

    SciTech Connect (OSTI)

    Erbudak, M.; Stucki, F.

    1985-08-15

    Ion scattering and Auger-electron spectroscopies show that there is a large segregation of uranium at the surface of UFe/sub 2/. Adsorbed oxygen reacts only with this uranium and forms a stable oxide layer at the surface, which prevents further oxygen diffusing into the solid. As a result of this process, the iron remains in metallic form even after prolonged oxygen exposures.

  14. Draft Supplement Analysis for Location(s) to Dispose of Depleted Uranium Oxide Conversion Product Generated from DOE'S Inventory of Depleted Uranium Hexafluoride

    Office of Environmental Management (EM)

    DRAFT SUPPLEMENT ANALYSIS FOR LOCATION(S) TO DISPOSE OF DEPLETED URANIUM OXIDE CONVERSION PRODUCT GENERATED FROM DOE'S INVENTORY OF DEPLETED URANIUM HEXAFLUORIDE (DOE/EIS-0359-SA1 AND DOE/EIS-0360-SA1) March 2007 March 2007 i CONTENTS NOTATION........................................................................................................................... iv 1 INTRODUCTION AND BACKGROUND ................................................................. 1 1.1 Why DOE Has Prepared This

  15. CRAD, Conduct of Operations- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January, 2005 assessment of Conduct of Operations program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  16. CRAD, Environmental Protection- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Environmental Compliance program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  17. CRAD, Occupational Safety & Health- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Industrial Safety and Industrial Health programs at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  18. EIS-0089: PUREX Plant and Uranium Oxide Plant Facilities, Hanford Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this statement to evaluate the environmental impacts of resumption of operations of the PUREX/Uranium Oxide facilities at the Hanford Site to produce plutonium and other special nuclear materials for national defense needs.

  19. CRAD, Emergency Management- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Emergency Management program at the Y-12 Enriched Uranium Operations Oxide Conversion Facility.

  20. CRAD, Safety Basis- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Safety Basis at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  1. CRAD, Radiological Controls- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Radiation Protection Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  2. CRAD, DOE Oversight- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a DOE independent oversight assessment of the Y-12 Site Office's programs for oversight of its contractors at the Y-12 Enriched Uranium Operations Oxide Conversion Facility.

  3. Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

    DOE Patents [OSTI]

    McLean, W. II; Miller, P.E.

    1997-12-16

    A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs.

  4. Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

    DOE Patents [OSTI]

    McLean, II, William; Miller, Philip E.

    1997-01-01

    A method for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction.

  5. Uranium

    SciTech Connect (OSTI)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-10-01

    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/sub 3/O/sub 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. (DP)

  6. Uranium Oxide as a Highly Reflective Coating from 100-400 eV

    SciTech Connect (OSTI)

    Sandberg, Richard L.; Allred, David D.; Bissell, Luke J.; Johnson, Jed E.; Turley, R. Steven

    2004-05-12

    We present the measured reflectances (Beamline 6.3.2, ALS at LBNL) of naturally oxidized uranium and naturally oxidized nickel thin films from 100-460 eV (2.7 to 11.6 nm) at 5 and 15 degrees grazing incidence. These show that uranium, as UO2, can fulfill its promise as the highest known single surface reflector for this portion of the soft x-ray region, being nearly twice as reflective as nickel in the 124-250 eV (5-10 nm) region. This is due to its large index of refraction coupled with low absorption. Nickel is commonly used in soft x-ray applications in astronomy and synchrotrons. (Its reflectance at 10 deg. exceeds that of Au and Ir for most of this range.) We prepared uranium and nickel thin films via DC-magnetron sputtering of a depleted U target and resistive heating evaporation respectively. Ambient oxidation quickly brought the U sample to UO2 (total thickness about 30 nm). The nickel sample (50 nm) also acquired a thin native oxide coating (<2nm). Though the density of U in UO2 is only half of the metal, its reflectance is high and it is relatively stable against further changes.

  7. Anaerobic U(IV) Bio-oxidation and the Resultant Remobilization of Uranium in Contaminated Sediments

    SciTech Connect (OSTI)

    Coates, John D.

    2005-06-01

    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

  8. Influence of uranyl speciation and iron oxides on uranium biogeochemical redox reactions

    SciTech Connect (OSTI)

    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

  9. Depleted uranium oxides and silicates as spent nuclear fuel waste package fill materials

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-09-10

    A new repository waste package (WP) concept for spent nuclear fuel (SNF) is being investigated that uses depleted uranium (DU) to improve performance and reduce the uncertainties of geological disposal of SNF. The WP would be filled with SNF and then filled with depleted uranium (DU) ({approximately}0.2 wt % {sup 235}U) dioxide (UO{sub 2}) or DU silicate-glass beads. Fission products and actinides can not escape the SNF UO{sub 2} crystals until the UO{sub 2} dissolves or is transformed into other chemical species. After WP failure, the DU fill material slows dissolution by three mechanisms: (1) saturation of AT groundwater with DU and suppression of SNF dissolution, (2) maintenance of chemically reducing conditions in the WP that minimize SNF solubility by sacrificial oxidation of DU from the +4 valence state, and (3) evolution of DU to lower-density hydrated uranium silicates. The fill expansion seals the WP from water flow. The DU also isotopically exchanges with SNF uranium as the SNF degrades to reduce long-term nuclear-criticality concerns.

  10. Observations of Oxygen Ion Behavior in the Lithium-Based Electrolytic Reduction of Uranium Oxide

    SciTech Connect (OSTI)

    Steven D. Herrmann; Shelly X. Li; Brenda E. Serrano-Rodriguez

    2009-09-01

    Parametric studies were performed on a lithium-based electrolytic reduction process at bench-scale to investigate the behavior of oxygen ions in the reduction of uranium oxide for various electrochemical cell configurations. Specifically, a series of eight electrolytic reduction runs was performed in a common salt bath of LiCl 1 wt% Li2O. The variable parameters included fuel basket containment material (i.e., stainless steel wire mesh and sintered stainless steel) and applied electrical charge (i.e., 75 150% of the theoretical charge for complete reduction of uranium oxide in a basket to uranium metal). Samples of the molten salt electrolyte were taken at regular intervals throughout each run and analyzed to produce a time plot of Li2O concentrations in the bulk salt over the course of the runs. Following each run, the fuel basket was sectioned and the fuel was removed. Samples of the fuel were analyzed for the extent of uranium oxide reduction to metal and for the concentration of salt constituents, i.e., LiCl and Li2O. Extents of uranium oxide reduction ranged from 43 70% in stainless steel wire mesh baskets and 8 33 % in sintered stainless steel baskets. The concentrations of Li2O in the salt phase of the fuel product from the stainless steel wire mesh baskets ranged from 6.2 9.2 wt%, while those for the sintered stainless steel baskets ranged from 26 46 wt%. Another series of tests was performed to investigate the dissolution of Li2O in LiCl at 650 C across various cathode containment materials (i.e., stainless steel wire mesh, sintered stainless steel and porous magnesia) and configurations (i.e., stationary and rotating cylindrical baskets). Dissolution of identical loadings of Li2O particulate reached equilibrium within one hour for stationary stainless steel wire mesh baskets, while the same took several hours for sintered stainless steel and porous magnesia baskets. Rotation of an annular cylindrical basket of stainless steel wire mesh

  11. DECONTAMINATION OF URANIUM

    DOE Patents [OSTI]

    Feder, H.M.; Chellew, N.R.

    1958-02-01

    This patent deals with the separation of rare earth and other fission products from neutron bombarded uranium. This is accomplished by melting the uranium in contact with either thorium oxide, maguesium oxide, alumnum oxide, beryllium oxide, or uranium dioxide. The melting is preferably carried out at from 1150 deg to 1400 deg C in an inert atmosphere, such as argon or helium. During this treatment a scale of uranium dioxide forms on the uranium whtch contains most of the fission products.

  12. Sulfurization behavior of cerium doped uranium oxides by CS{sub 2}

    SciTech Connect (OSTI)

    Sato, Nobuaki; Kato, Shintaro; Kirishima, Akira; Tochiyama, Osamu

    2007-07-01

    For the recovery of nuclear materials from the spent nuclear fuel, the sulfide process has been proposed and the voloxidation of spent fuel and selective sulfurization rare-earth elements has been proposed. In this paper, cerium was used as a stand-in of plutonium and sulfurization behavior of cerium doped uranium dioxide by CS{sub 2} was studied. UO{sub 2} was oxidized to U{sub 3}O{sub 8} in air, while the Ce doped UO{sub 2} solid solution was formed in the presence of CeO{sub 2} by the heat treatment in air. The effect of heating time, temperature and the ratio of uranium to cerium on the formation of solid solution was analyzed. The results were also compared with those of thermodynamic consideration. (authors)

  13. A two-dimensional, finite-difference model of the oxidation of a uranium carbide fuel pellet

    SciTech Connect (OSTI)

    Shepherd, James; Fairweather, Michael; Hanson, Bruce C.; Heggs, Peter J.

    2015-12-31

    The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

  14. Uranium Immobilization through Fe(II) bio-oxidation: A Column study

    SciTech Connect (OSTI)

    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

  15. Uranium in geologic fluids: Estimates of standard partial molal properties, oxidation potentials, and hydrolysis constants at high temperatures and pressures

    SciTech Connect (OSTI)

    Shock, E.L.; Sassani, D.C.; Betz, H.

    1997-10-01

    Theoretical methods are used with the available experimental data to provide estimates of parameters for the revised-HKF equations of state for aqueous uranium species. These parameters are used with standard state thermodynamic data at 25{degrees}C and 1 bar to calculate equilibrium constants for redox reactions among the four most common oxidation states of uranium (U(III), U(IV), U(V), and U(VI)), and their hydrolysis reactions at temperatures to 1000{degrees}C and pressures to 5 kb. A total of nineteen aqueous uranium species are included. The predicted equilibrium constants are used to construct oxidation potential-pH diagrams at elevated temperatures and pressures and to calculate the solubilities of uraninite as functions of temperature and pH, which are compared to experimental data. Oxidation potential-pH diagrams illustrate the relative stabilities of aqueous uranium species and indicate that U(IV) and U(VI) species predominate in aqueous solution in the U-O-H system. Increasing temperature stabilizes U(VI) and U(III) species relative to U(IV) species, but U(IV) species dominate at oxidation states consistent and mineral-buffer assemblages and near-neutral pH. At low pH, U(VI) is stabilized relative to U(IV) suggesting that uranium transport in hydrothermal systems requires either acidic solutions or potent complexes of U(IV). 40 refs., 15 figs., 3 tabs.

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

    DOE Patents [OSTI]

    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.

  17. Process for electroslag refining of uranium and uranium alloys

    DOE Patents [OSTI]

    Lewis, P.S. Jr.; Agee, W.A.; Bullock, J.S. IV; Condon, J.B.

    1975-07-22

    A process is described for electroslag refining of uranium and uranium alloys wherein molten uranium and uranium alloys are melted in a molten layer of a fluoride slag containing up to about 8 weight percent calcium metal. The calcium metal reduces oxides in the uranium and uranium alloys to provide them with an oxygen content of less than 100 parts per million. (auth)

  18. Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

    SciTech Connect (OSTI)

    Degueldre, Claude Cozzo, Cedric; Martin, Matthias; Grolimund, Daniel; Mieszczynski, Cyprian

    2013-06-01

    Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O? lattice in an irradiated (60 MW d kg?) MOX sample was performed employing micro-X-ray fluorescence (-XRF) and micro-X-ray absorption fine structure (-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (~0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am? species within an [AmO?]? coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix. - Graphical abstract: Americium LIII XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 ?m300 ?m beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Am is found as trivalent in the UO? matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. The americium redox state as determined from XAS data of irradiated fuel material was Am(III). In the sample, the Am? face an AmO??coordination environment in the (Pu,U)O? matrix. The americium dioxide is reduced by the uranium dioxide matrix.

  19. Apparatus and process for the electrolytic reduction of uranium and plutonium oxides

    DOE Patents [OSTI]

    Poa, David S.; Burris, Leslie; Steunenberg, Robert K.; Tomczuk, Zygmunt

    1991-01-01

    An apparatus and process for reducing uranium and/or plutonium oxides to produce a solid, high-purity metal. The apparatus is an electrolyte cell consisting of a first container, and a smaller second container within the first container. An electrolyte fills both containers, the level of the electrolyte in the first container being above the top of the second container so that the electrolyte can be circulated between the containers. The anode is positioned in the first container while the cathode is located in the second container. Means are provided for passing an inert gas into the electrolyte near the lower end of the anode to sparge the electrolyte and to remove gases which form on the anode during the reduction operation. Means are also provided for mixing and stirring the electrolyte in the first container to solubilize the metal oxide in the electrolyte and to transport the electrolyte containing dissolved oxide into contact with the cathode in the second container. The cell is operated at a temperature below the melting temperature of the metal product so that the metal forms as a solid on the cathode.

  20. Incorporation of oxidized uranium into Fe (hydr)oxides during Fe(II) catalyzed remineralization

    SciTech Connect (OSTI)

    Nico, Peter S.; Stewart, Brandy D.; Fendorf, Scott

    2009-07-01

    The form of solid phase U after Fe(II) induced anaerobic remineralization of ferrihydrite in the presence of aqueous and absorbed U(VI) was investigated under both abiotic batch and biotic flow conditions. Experiments were conducted with synthetic ground waters containing 0.168 mM U(VI), 3.8 mM carbonate, and 3.0 mM Ca{sup 2+}. In spite of the high solubility of U(VI) under these conditions, appreciable removal of U(VI) from solution was observed in both the abiotic and biotic systems. The majority of the removed U was determined to be substituted as oxidized U (U(VI) or U(V)) into the octahedral position of the goethite and magnetite formed during ferrihydrite remineralization. It is estimated that between 3% and 6% of octahedral Fe(III) centers in the new Fe minerals were occupied by U(VI). This site specific substitution is distinct from the non-specific U co-precipitation processes in which uranyl compounds, e.g. uranyl hydroxide or carbonate, are entrapped with newly formed Fe oxides. The prevalence of site specific U incorporation under both abiotic and biotic conditions and the fact that the produced solids were shown to be resistant to both extraction (30 mM KHCO{sub 3}) and oxidation (air for 5 days) suggest the potential importance of sequestration in Fe oxides as a stable and immobile form of U in the environment.

  1. Disposition of Uranium -233 (sup 233U) in Plutonium Metal and Oxide at the Rocky Flats Environmental Technology Site

    SciTech Connect (OSTI)

    Freiboth, Cameron J.; Gibbs, Frank E.

    2000-03-01

    This report documents the position that the concentration of Uranium-233 ({sup 233}U) in plutonium metal and oxide currently stored at the DOE Rocky Flats Environmental Technology Site (RFETS) is well below the maximum permissible stabilization, packaging, shipping and storage limits. The {sup 233}U stabilization, packaging and storage limit is 0.5 weight percent (wt%), which is also the shipping limit maximum. These two plutonium products (metal and oxide) are scheduled for processing through the Building 371 Plutonium Stabilization and Packaging System (PuSPS). This justification is supported by written technical reports, personnel interviews, and nuclear material inventories, as compiled in the ''History of Uranium-233 ({sup 233}U) Processing at the Rocky Flats Plant In Support of the RFETS Acceptable Knowledge Program'' RS-090-056, April 1, 1999. Relevant data from this report is summarized for application to the PuSPS metal and oxide processing campaigns.

  2. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A.

    1989-01-01

    A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

  3. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A.

    1989-08-01

    A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

  4. PRODUCTION OF URANIUM TETRACHLORIDE

    DOE Patents [OSTI]

    Calkins, V.P.

    1958-12-16

    A process is descrlbed for the production of uranium tetrachloride by contacting uranlum values such as uranium hexafluoride, uranlum tetrafluoride, or uranium oxides with either aluminum chloride, boron chloride, or sodium alumlnum chloride under substantially anhydrous condltlons at such a temperature and pressure that the chlorldes are maintained in the molten form and until the uranium values are completely converted to uranlum tetrachloride.

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

    SciTech Connect (OSTI)

    Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; Williams, Ross; Lindvall, Rachel; Eppich, Gary; Roberts, Sarah; Borg, Lars; Gaffney, Amy; Plaue, Jonathan; Knight, Kim; Loi, Elaine; Hotchkis, Michael; Moody, Kenton; Singleton, Michael; Robel, Martin; Hutcheon, Ian

    2015-04-13

    Abstract

    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

  6. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect (OSTI)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  7. Load-Based (LB) CRAC (rates/adjustments)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    absolutely will change. June 10, 2003 LB CRAC Workshop: Presentation on Impacts of the Enron Settlement on LB CRAC (PDF, 6 pages, 24 kb, posted June 11, 2003) Related follow-up...

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; Williams, Ross; Lindvall, Rachel; Eppich, Gary; Roberts, Sarah; Borg, Lars; Gaffney, Amy; Plaue, Jonathan; et al

    2015-04-13

    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 K2(UO2)3O4·4H2O. 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 UO3·2H2O, with minor phases ofmore » U3O8 and UO2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U 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.« less

  9. URANIUM LEACHING AND RECOVERY PROCESS

    DOE Patents [OSTI]

    McClaine, L.A.

    1959-08-18

    A process is described for recovering uranium from carbonate leach solutions by precipitating uranium as a mixed oxidation state compound. Uranium is recovered by adding a quadrivalent uranium carbon;te solution to the carbonate solution, adjusting the pH to 13 or greater, and precipitating the uranium as a filterable mixed oxidation state compound. In the event vanadium occurs with the uranium, the vanadium is unaffected by the uranium precipitation step and remains in the carbonate solution. The uranium-free solution is electrolyzed in the cathode compartment of a mercury cathode diaphragm cell to reduce and precipitate the vanadium.

  10. Influences of Organic Carbon Supply Rate on Uranium Bioreduction in Initially Oxidizing, Contaminated Sediment

    SciTech Connect (OSTI)

    Tokunaga, Tetsu K.; Wan, Jiamin; Kim, Yongman; Daly, Rebecca A.; Brodie, Eoin L.; Hazen, Terry C.; Herman, Don; Firestone, Mary K.

    2008-06-10

    Remediation of uranium (U) contaminated sediments through in-situ stimulation of bioreduction to insoluble UO{sub 2} is a potential treatment strategy under active investigation. Previously, we found that newly reduced U(IV) can be reoxidized under reducing conditions sustained by a continuous supply of organic carbon (OC) because of residual reactive Fe(III) and enhanced U(VI) solubility through complexation with carbonate generated through OC oxidation. That finding motivated this investigation directed at identifying a range of OC supply rates that is optimal for establishing U bioreduction and immobilization in initially oxidizing sediments. The effects of OC supply rate, from 0 to 580 mmol OC (kg sediment){sup -1} year{sup -1}, and OC form (lactate and acetate) on U bioreduction were tested in flow-through columns containing U-contaminated sediments. An intermediate supply rate on the order of 150 mmol OC (kg sediment){sup -1} year{sup -1} was determined to be most effective at immobilizing U. At lower OC supply rates, U bioreduction was not achieved, and U(VI) solubility was enhanced by complexation with carbonate (from OC oxidation). At the highest OC supply rate, resulting highly carbonate-enriched solutions also supported elevated levels of U(VI), even though strongly reducing conditions were established. Lactate and acetate were found to have very similar geochemical impacts on effluent U concentrations (and other measured chemical species), when compared at equivalent OC supply rates. While the catalysts of U(VI) reduction to U(IV) are presumably bacteria, the composition of the bacterial community, the Fe reducing community, and the sulfate reducing community had no direct relationship with effluent U concentrations. The OC supply rate has competing effects of driving reduction of U(VI) to low solubility U(IV) solids, as well as causing formation of highly soluble U(VI)-carbonato complexes. These offsetting influences will require careful control of OC

  11. Method of preparation of uranium nitride

    DOE Patents [OSTI]

    Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

    2013-07-09

    Method for producing terminal uranium nitride complexes comprising providing a suitable starting material comprising uranium; oxidizing the starting material with a suitable oxidant to produce one or more uranium(IV)-azide complexes; and, sufficiently irradiating the uranium(IV)-azide complexes to produce the terminal uranium nitride complexes.

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

    SciTech Connect (OSTI)

    Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; Williams, Ross; Lindvall, Rachel; Eppich, Gary; Roberts, Sarah; Borg, Lars; Gaffney, Amy; Plaue, Jonathan; Knight, Kim; Loi, Elaine; Hotchkis, Michael; Moody, Kenton; Singleton, Michael; Robel, Martin; Hutcheon, Ian

    2015-04-13

    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 K2(UO2)3O4·4H2O. 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 UO3·2H2O, with minor phases of U3O8 and UO2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U 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.

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

    SciTech Connect (OSTI)

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

    1999-03-01

    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.

  14. Structural Sequestration of Uranium in Bacteriogenic Manganese...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Sequestration of Uranium in Bacteriogenic Manganese Oxides Samuel M. Webb (Stanford ... Uranium is a key contaminant of concern at US DOE sites and shuttered mining and ore ...

  15. Process for continuous production of metallic uranium and uranium alloys

    DOE Patents [OSTI]

    Hayden, H.W. Jr.; Horton, J.A.; Elliott, G.R.B.

    1995-06-06

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

  16. Process for continuous production of metallic uranium and uranium alloys

    DOE Patents [OSTI]

    Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

    1995-01-01

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

  17. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  18. Market overview: Increase in uranium prices continues

    SciTech Connect (OSTI)

    1996-04-01

    Spot market activity totaled just over 200,000 lbs of U308 equivalent. The restricted uranium spot market price range increased from a high last month of $14.75/lb U308 to a low this month of $15.25/lb U308. There was also an increase in the unrestricted range this month with the upper end of the range increasing by $0.50/lb U308. The lower end of the spot conversion price range increased by R0.35/kg U while the upper end of the separative work price range increased by $2.00/SWU.

  19. Efficacy of a solution-based approach for making sodalite waste forms for an oxide reduction salt utilized in the reprocessing of used uranium oxide fuel

    SciTech Connect (OSTI)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyáš, Josef; Burns, Carolyne A.

    2015-04-01

    This paper describes the various approaches attempted to make solution-derived sodalite with a LiCl-Li2O oxide reduction salt used to dissolve used uranium oxide fuel so the uranium can be recovered and recycled. The approaches include modified sol-gel and solutionbased synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3- SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2O and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions in the crystalline product (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt but that the incorporation of Li into the sodalite is low.

  20. Efficacy of a solution-based approach for making sodalite waste forms for an oxide reduction salt utilized in the reprocessing of used uranium oxide fuel

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyáš, Josef; Burns, Carolyne A.

    2015-04-01

    This paper describes the various approaches attempted to make solution-derived sodalite with a LiCl-Li2O oxide reduction salt used to dissolve used uranium oxide fuel so the uranium can be recovered and recycled. The approaches include modified sol-gel and solutionbased synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3- SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2Omore » and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions in the crystalline product (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt but that the incorporation of Li into the sodalite is low.« less

  1. PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

    DOE Patents [OSTI]

    Holden, R.B.; Powers, R.M.; Blaber, O.J.

    1959-09-22

    The preparation of uranium metal by the carbon reduction of an oxide of uranium is described. In a preferred embodiment of the invention a charge composed of carbon and uranium oxide is heated to a solid mass after which it is further heated under vacuum to a temperature of about 2000 deg C to produce a fused uranium metal. Slowly ccoling the fused mass produces a dendritic structure of uranium carbide in uranium metal. Reacting the solidified charge with deionized water hydrolyzes the uranium carbide to finely divide uranium dioxide which can be separated from the coarser uranium metal by ordinary filtration methods.

  2. Incorporation of uranium in pyrochlore oxides and pressure-induced phase transitions

    SciTech Connect (OSTI)

    Zhang, F.X.; Lang, M.; Tracy, C.; Ewing, R.C.; Gregg, D.J.; Lumpkin, G.R.

    2014-11-15

    Uranium-doped gadolinium zirconates with pyrochlore structure were studied at ambient and high-pressure conditions up to 40 GPa. The bonding environment of uranium in the structure was determined by x-ray photoelectron and Raman spectroscopies and x-ray diffraction. The uranium valence for samples prepared in air is mainly U{sup 6+}, but U{sup 4+} is present in pyrochlores fabricated in an argon atmosphere. Rietveld refinement of the XRD pattern suggests that uranium ions in pyrochlores are on the 16d site in 6-fold coordination with oxygen. At pressures greater than 22 GPa, the pyrochlore structure transformed to a cotunnite-type phase. The cotunnite high-pressure phase transformed to a defect fluorite structure on the release of pressure. - Graphical abstract: In U-bearing pyrochlore, U ions mainly occupy the 16d site and replace the smaller Zr{sup 4+}, part of the oxygen will occupy the 8b site, which is empty to most pyrochlores. At pressure of 22 GPa, the pyrochlore lattice is not stable and transforms to a cotunnite-type structure. The high-pressure structure is not stable and transform to a fluorite or back to the pyrochlore structure when pressure is released. - Highlights: • We found that U ions mainly occupy the smaller cation site in U-bearing pyrochlore. • Pyrochlore structure is not stable at pressure of more than 20 GPa. • The quenched sample has a pyrochlore or a disordered fluorite structure.

  3. CRAD, Criticality Safety- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Criticality Safety program at the Y-12 - Enriched Uranium Facility.

  4. Uranium dioxide electrolysis

    SciTech Connect (OSTI)

    Willit, James L.; Ackerman, John P.; Williamson, Mark A.

    2009-12-29

    This is a single stage process for treating spent nuclear fuel from light water reactors. The spent nuclear fuel, uranium oxide, UO.sub.2, is added to a solution of UCl.sub.4 dissolved in molten LiCl. A carbon anode and a metallic cathode is positioned in the molten salt bath. A power source is connected to the electrodes and a voltage greater than or equal to 1.3 volts is applied to the bath. At the anode, the carbon is oxidized to form carbon dioxide and uranium chloride. At the cathode, uranium is electroplated. The uranium chloride at the cathode reacts with more uranium oxide to continue the reaction. The process may also be used with other transuranic oxides and rare earth metal oxides.

  5. Simulation of uranium transport with variable temperature and oxidation potential: The computer program THCC (Thermo-Hydro-Chemical Coupling)

    SciTech Connect (OSTI)

    Carnahan, C.L.

    1986-12-01

    A simulator of reactive chemical transport has been constructed with the capabilities of treating variable temperatures and variable oxidation potentials within a single simulation. Homogeneous and heterogeneous chemical reactions are simulated at temperature-dependent equilibrium, and changes of oxidation states of multivalent elements can be simulated during transport. Chemical mass action relations for formation of complexes in the fluid phase are included explicitly within the partial differential equations of transport, and a special algorithm greatly simplifies treatment of reversible precipitation of solid phases. This approach allows direct solution of the complete set of governing equations for concentrations of all aqueous species and solids affected simultaneously by chemical and physical processes. Results of example simulations of transport, along a temperature gradient, of uranium solution species under conditions of varying pH and oxidation potential and with reversible precipitation of uraninite and coffinite are presented. The examples illustrate how inclusion of variable temperature and oxidation potential in numerical simulators can enhance understanding of the chemical mechanisms affecting migration of multivalent waste elements.

  6. METHOD FOR RECOVERING URANIUM FROM OILS

    DOE Patents [OSTI]

    Gooch, L.H.

    1959-07-14

    A method is presented for recovering uranium from hydrocarbon oils, wherein the uranium is principally present as UF/sub 4/. According to the invention, substantially complete removal of the uranium from the hydrocarbon oil may be effected by intimately mixing one part of acetone to about 2 to 12 parts of the hydrocarbon oil containing uranium and separating the resulting cake of uranium from the resulting mixture. The uranium in the cake may be readily recovered by burning to the oxide.

  7. Efficacy of a Solution-Based Approach for Making Sodalite Waste Forms for an Oxide Reduction Salt Utilized in the Reprocessing of Used Uranium Oxide Fuel

    SciTech Connect (OSTI)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyas, Josef; Burns, Carolyn A.

    2015-04-01

    This paper describes various approaches for making sodalite with a LiCl-Li2O oxide reduction salt used to recover uranium from used oxide fuel. The approaches include sol-gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3-SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2O and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt.

  8. Method for the recovery of uranium values from uranium tetrafluoride

    DOE Patents [OSTI]

    Kreuzmann, Alvin B.

    1983-01-01

    The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions wherein the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

  9. Method for the recovery of uranium values from uranium tetrafluoride

    DOE Patents [OSTI]

    Kreuzmann, A.B.

    1982-10-27

    The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions whereas the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

  10. METHOD OF ROLLING URANIUM

    DOE Patents [OSTI]

    Smith, C.S.

    1959-08-01

    A method is described for rolling uranium metal at relatively low temperatures and under non-oxidizing conditions. The method involves the steps of heating the uranium to 200 deg C in an oil bath, withdrawing the uranium and permitting the oil to drain so that only a thin protective coating remains and rolling the oil coated uranium at a temperature of 200 deg C to give about a 15% reduction in thickness at each pass. The operation may be repeated to accomplish about a 90% reduction without edge cracking, checking or any appreciable increase in brittleness.

  11. Experimental studies and thermodynamic modelling of volatilities of uranium, plutonium, and americium from their oxides and from their oxides interacted with ash

    SciTech Connect (OSTI)

    Krikorian, O.H.; Ebbinghaus, B.B.; Adamson, M.G.; Fontes, A.S. Jr.; Fleming, D.L.

    1993-09-15

    The purpose of this study is to identify the types and amounts of volatile gaseous species of U, Pu, and Am that are produced in the combustion chamber offgases of mixed waste oxidation processors. Primary emphasis is on the Rocky Flats Plant Fluidized Bed Incinerator. Transpiration experiments have been carried out on U{sub 3}O{sub 8}(s), U{sub 3}O{sub 8} interacted with various ash materials, PuO{sub 2}(s), PuO{sub 2} interacted with ash materials, and a 3%PuO{sub 2}/0.06%AmO{sub 2}/ash material, all in the presence of steam and oxygen, and at temperatures in the vicinity of 1,300 K. UO{sub 3}(g) and UO{sub 2}(OH){sub 2}(g) have been identified as the uranium volatile species and thermodynamic data established for them. Pu and Am are found to have very low volatilities, and carryover of Pu and Am as fine dust particulates is found to dominate over vapor transport. The authors are able to set upper limits on Pu and Am volatilities. Very little lowering of U volatility is found for U{sub 3}O{sub 8} interacted with typical ashes. However, ashes high in Na{sub 2}O (6.4 wt %) or in CaO (25 wt %) showed about an order of magnitude reduction in U volatility. Thermodynamic modeling studies were carried out that show that for aluminosilicate ash materials, it is the presence of group I and group II oxides that reduces the activity of the actinide oxides. K{sub 2}O is the most effective followed by Na{sub 2}O and CaO for common ash constituents. A more major effect in actinide activity lowering could be achieved by adding excess group I or group II oxides to exceed their interaction with the ash and lead to direct formation of alkali or alkaline earth uranates, plutonates, and americates.

  12. A coupled transport and solid mechanics formulation with improved reaction kinetics parameters for modeling oxidation and decomposition in a uranium hydride bed.

    SciTech Connect (OSTI)

    Salloum, Maher N.; Shugard, Andrew D.; Kanouff, Michael P.; Gharagozloo, Patricia E.

    2013-03-01

    Modeling of reacting flows in porous media has become particularly important with the increased interest in hydrogen solid-storage beds. An advanced type of storage bed has been proposed that utilizes oxidation of uranium hydride to heat and decompose the hydride, releasing the hydrogen. To reduce the cost and time required to develop these systems experimentally, a valid computational model is required that simulates the reaction of uranium hydride and oxygen gas in a hydrogen storage bed using multiphysics finite element modeling. This SAND report discusses the advancements made in FY12 (since our last SAND report SAND2011-6939) to the model developed as a part of an ASC-P&EM project to address the shortcomings of the previous model. The model considers chemical reactions, heat transport, and mass transport within a hydride bed. Previously, the time-varying permeability and porosity were considered uniform. This led to discrepancies between the simulated results and experimental measurements. In this work, the effects of non-uniform changes in permeability and porosity due to phase and thermal expansion are accounted for. These expansions result in mechanical stresses that lead to bed deformation. To describe this, a simplified solid mechanics model for the local variation of permeability and porosity as a function of the local bed deformation is developed. By using this solid mechanics model, the agreement between our reacting bed model and the experimental data is improved. Additionally, more accurate uranium hydride oxidation kinetics parameters are obtained by fitting the experimental results from a pure uranium hydride oxidation measurement to the ones obtained from the coupled transport-solid mechanics model. Finally, the coupled transport-solid mechanics model governing equations and boundary conditions are summarized and recommendations are made for further development of ARIA and other Sandia codes in order for them to sufficiently implement the model.

  13. Absorption of Thermal Neutrons in Uranium

    DOE R&D Accomplishments [OSTI]

    Creutz, E. C.; Wilson, R. R.; Wigner, E. P.

    1941-09-26

    A knowledge of the absorption processes for neutrons in uranium is important for planning a chain reaction experiment. The absorption of thermal neutrons in uranium and uranium oxide has been studied. Neutrons from the cyclotron were slowed down by passage through a graphite block. A uranium or uranium oxide sphere was placed at various positions in the block. The neutron intensity at different points in the sphere and in the graphite was measured by observing the activity induced in detectors or uranium oxide or manganese. It was found that both the fission activity in the uranium oxide and the activity induced in manganese was affected by non-thermal neutrons. An experimental correction for such effects was made by making measurements with the detectors surrounded by cadmium. After such corrections the results from three methods of procedure with the uranium oxide detectors and from the manganese detectors were consistent to within a few per cent.

  14. Computation Results from a Parametric Study to Determine Bounding Critical Systems of Homogeneously Water-Moderated Mixed Plutonium--Uranium Oxides

    SciTech Connect (OSTI)

    Shimizu, Y.

    2001-01-11

    This report provides computational results of an extensive study to examine the following: (1) infinite media neutron-multiplication factors; (2) material bucklings; (3) bounding infinite media critical concentrations; (4) bounding finite critical dimensions of water-reflected and homogeneously water-moderated one-dimensional systems (i.e., spheres, cylinders of infinite length, and slabs that are infinite in two dimensions) that were comprised of various proportions and densities of plutonium oxides and uranium oxides, each having various isotopic compositions; and (5) sensitivity coefficients of delta k-eff with respect to critical geometry delta dimensions were determined for each of the three geometries that were studied. The study was undertaken to support the development of a standard that is sponsored by the International Standards Organization (ISO) under Technical Committee 85, Nuclear Energy (TC 85)--Subcommittee 5, Nuclear Fuel Technology (SC 5)--Working Group 8, Standardization of Calculations, Procedures and Practices Related to Criticality Safety (WG 8). The designation and title of the ISO TC 85/SC 5/WG 8 standard working draft is WD 14941, ''Nuclear energy--Fissile materials--Nuclear criticality control and safety of plutonium-uranium oxide fuel mixtures outside of reactors.'' Various ISO member participants performed similar computational studies using their indigenous computational codes to provide comparative results for analysis in the development of the standard.

  15. Photoelectron spectroscopy and theoretical studies of gaseous uranium hexachlorides in different oxidation states: UCl{sub 6}{sup q?} (q = 02)

    SciTech Connect (OSTI)

    Su, Jing; Dau, Phuong D.; Huang, Dao-Ling; Wang, Lai-Sheng; Liu, Hong-Tao; Wei, Fan; Schwarz, W. H. E.; Li, Jun

    2015-04-07

    Uranium chlorides are important in actinide chemistry and nuclear industries, but their chemical bonding and many physical and chemical properties are not well understood yet. Here, we report the first experimental observation of two gaseous uranium hexachloride anions, UCl{sub 6}{sup ?} and UCl{sub 6}{sup 2?}, which are probed by photoelectron spectroscopy in conjunction with quantum chemistry calculations. The electron affinity of UCl{sub 6} is measured for the first time as +5.3 eV; its second electron affinity is measured to be +0.60 eV from the photoelectron spectra of UCl{sub 6}{sup 2?}. We observe that the detachment cross sections of the 5f electrons are extremely weak in the visible and UV energy ranges. It is found that the one-electron one-determinental molecular orbital picture and Koopmans theorem break down for the strongly internally correlated U-5f{sup 2} valence shell of tetravalent U{sup +4} in UCl{sub 6}{sup 2?}. The calculated adiabatic and vertical electron detachment energies from ab initio calculations agree well with the experimental observations. Electronic structure and chemical bonding in the uranium hexachloride species UCl{sub 6}{sup 2?} to UCl{sub 6} are discussed as a function of the oxidation state of U.

  16. URANIUM EXTRACTION PROCESS

    DOE Patents [OSTI]

    Baldwin, W.H.; Higgins, C.E.

    1958-12-16

    A process is described for recovering uranium values from acidic aqueous solutions containing hexavalent uranium by contacting the solution with an organic solution comprised of a substantially water-immiscible organlc diluent and an organic phosphate to extract the uranlum values into the organic phase. Carbon tetrachloride and a petroleum hydrocarbon fraction, such as kerosene, are sultable diluents to be used in combination with organlc phosphates such as dibutyl butylphosphonate, trlbutyl phosphine oxide, and tributyl phosphate.

  17. Uranium immobilization and nuclear waste

    SciTech Connect (OSTI)

    Duffy, C.J.; Ogard, A.E.

    1982-02-01

    Considerable information useful in nuclear waste storage can be gained by studying the conditions of uranium ore deposit formation. Further information can be gained by comparing the chemistry of uranium to nuclear fission products and other radionuclides of concern to nuclear waste disposal. Redox state appears to be the most important variable in controlling uranium solubility, especially at near neutral pH, which is characteristic of most ground water. This is probably also true of neptunium, plutonium, and technetium. Further, redox conditions that immobilize uranium should immobilize these elements. The mechanisms that have produced uranium ore bodies in the Earth's crust are somewhat less clear. At the temperatures of hydrothermal uranium deposits, equilibrium models are probably adequate, aqueous uranium (VI) being reduced and precipitated by interaction with ferrous-iron-bearing oxides and silicates. In lower temperature roll-type uranium deposits, overall equilibrium may not have been achieved. The involvement of sulfate-reducing bacteria in ore-body formation has been postulated, but is uncertain. Reduced sulfur species do, however, appear to be involved in much of the low temperature uranium precipitation. Assessment of the possibility of uranium transport in natural ground water is complicated because the system is generally not in overall equilibrium. For this reason, Eh measurements are of limited value. If a ground water is to be capable of reducing uranium, it must contain ions capable of reducing uranium both thermodynamically and kinetically. At present, the best candidates are reduced sulfur species.

  18. Uranium hexafluoride handling. Proceedings

    SciTech Connect (OSTI)

    Not Available

    1991-12-31

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF{sub 6} from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

  19. URANIUM ALLOYS

    DOE Patents [OSTI]

    Colbeck, E.W.

    1959-12-29

    A uranium alloy is reported containing from 0.1 to 5 per cent by weight of molybdenum and from 0.1 to 5 per cent by weight of silicon, the balance being uranium.

  20. MELTING AND PURIFICATION OF URANIUM

    DOE Patents [OSTI]

    Spedding, F.H.; Gray, C.F.

    1958-09-16

    A process is described for treating uranium ingots having inner metal portions and an outer oxide skin. The method consists in partially supporting such an ingot on the surface of a grid or pierced plate. A sufficient weight of uranium is provided so that when the mass becomes molten, the oxide skin bursts at the unsupported portions of its bottom surface, allowing molten urantum to flow through the burst skin and into a container provided below.

  1. SURFACE TREATMENT OF METALLIC URANIUM

    DOE Patents [OSTI]

    Gray, A.G.; Schweikher, E.W.

    1958-05-27

    The treatment of metallic uranium to provide a surface to which adherent electroplates can be applied is described. Metallic uranium is subjected to an etchant treatment in aqueous concentrated hydrochloric acid, and the etched metal is then treated to dissolve the resulting black oxide and/or chloride film without destroying the etched metal surface. The oxide or chloride removal is effected by means of moderately concentrated nitric acid in 3 to 20 seconds.

  2. METHOD OF PRODUCING URANIUM METAL BY ELECTROLYSIS

    DOE Patents [OSTI]

    Piper, R.D.

    1962-09-01

    A process is given for making uranium metal from oxidic material by electrolytic deposition on the cathode. The oxidic material admixed with two moles of carbon per one mole of uranium dioxide forms the anode, and the electrolyte is a mixture of from 40 to 75% of calcium fluoride or barium fluoride, 15 to 45% of uranium tetrafluoride, and from 10 to 20% of lithium fluoride or magnesium fluoride; the temperature of the electrolyte is between 1150 and 1175 deg C. (AEC)

  3. PRODUCTION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-08-27

    A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method, the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ converted to UF/sub 6/ by reaction with a fluorinating agent, such as CoF/sub 3/. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reac tion chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. After nitrogen gas is used to sweep out the hydrogen and the water vapor formed, and while continuing to inaintain the temperature between 400 deg C and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion of UO/sub 2/ to UF/sub 4/ the temperature of the reaction chamber is lowered to about 400 deg C or less, the UF/sub 4/ is mixed with the requisite quantity of CoF/sub 3/, and after evacuating the chamber, the mixture is heated to 300 to 400 deg C, and the resulting UF/sub 6/ is led off and delivered to a condenser.

  4. METHOD OF SEPARATING URANIUM VALUES, PLUTONIUM VALUES AND FISSION PRODUCTS BY CHLORINATION

    DOE Patents [OSTI]

    Brown, H.S.; Seaborg, G.T.

    1959-02-24

    The separation of plutonium and uranium from each other and from other substances is described. In general, the method comprises the steps of contacting the uranium with chlorine in the presence of a holdback material selected from the group consisting of lanthanum oxide and thorium oxide to form a uranium chloride higher than uranium tetrachloride, and thereafter heating the uranium chloride thus formed to a temperature at which the uranium chloride is volatilized off but below the volatilizalion temperature of plutonium chloride.

  5. TREATMENT OF URANIUM SURFACES

    DOE Patents [OSTI]

    Slunder, C.J.

    1959-02-01

    An improved process is presented for prcparation of uranium surfaces prior to electroplating. The surfacc of the uranium to be electroplated is anodized in a bath comprising a solution of approximately 20 to 602 by weight of phosphoric acid which contains about 20 cc per liter of concentrated hydrochloric acid. Anodization is carried out for approximately 20 minutes at a current density of about 0.5 amperes per square inch at a temperature of about 35 to 45 C. The oxidic film produced by anodization is removed by dipping in strong nitric acid, followed by rinsing with water just prior to electroplating.

  6. Inherently safe in situ uranium recovery

    SciTech Connect (OSTI)

    Krumhansl, James L; Brady, Patrick V

    2014-04-29

    An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

  7. Engineering assessment of inactive uranium mill tailings

    SciTech Connect (OSTI)

    Not Available

    1981-07-01

    The Grand Junction site has been reevaluated in order to revise the October 1977 engineering assessment of the problems resulting from the existence of radioactive uranium mill tailings at Grand Junction, Colorado. This engineering assessment has included the preparation of topographic maps, the performance of core drillings and radiometric measurements sufficient to determine areas and volumes of tailings and radiation exposures of individuals and nearby populations, the investigations of site hydrology and meteorology, and the evaluation and costing of alternative corrective actions. Radon gas released from the 1.9 million tons of tailings at the Grand Junction site constitutes the most significant environmental impact, although windblown tailings and external gamma radiation are also factors. The eight alternative actions presented herein range from millsite and off-site decontamination with the addition of 3 m of stabilization cover material (Option I), to removal of the tailings to remote disposal sites and decontamination of the tailings site (Options II through VIII). Cost estimates for the eight options range from about $10,200,000 for stabilization in-place to about $39,500,000 for disposal in the DeBeque area, at a distance of about 35 mi, using transportation by rail. If transportation to DeBeque were by truck, the cost estimated to be about $41,900,000. Three principal alternatives for the reprocessing of the Grand Junction tailings were examined: (a) heap leaching; (b) treatment at an existing mill; and (c) reprocessing at a new conventional mill constructed for tailings reprocessing. The cost of the uranium recovered would be about $200/lb by heap leach and $150/lb by conventional plant processes. The spot market price for uranium was $25/lb early in 1981. Therefore, reprocessing the tailings for uranium recovery appears not to be economically attractive.

  8. Separation of uranium from (Th,U)O.sub.2 solid solutions

    DOE Patents [OSTI]

    Chiotti, Premo; Jha, Mahesh Chandra

    1976-09-28

    Uranium is separated from mixed oxides of thorium and uranium by a pyrometallurgical process in which the oxides are mixed with a molten chloride salt containing thorium tetrachloride and thorium metal which reduces the uranium oxide to uranium metal which can then be recovered from the molten salt. The process is particularly useful for the recovery of uranium from generally insoluble high-density sol-gel thoria-urania nuclear reactor fuel pellets.

  9. EXTRACTION OF URANIUM

    DOE Patents [OSTI]

    Kesler, R.D.; Rabb, D.D.

    1959-07-28

    An improved process is presented for recovering uranium from a carnotite ore. In the improved process U/sub 2/O/sub 5/ is added to the comminuted ore along with the usual amount of NaCl prior to roasting. The amount of U/sub 2/O/ sub 5/ is dependent on the amount of free calcium oxide and the uranium in the ore. Specifically, the desirable amount of U/sub 2/O/sub 5/ is 3.2% for each 1% of CaO, and 5 to 6% for each 1% of uranium. The mixture is roasted at about 1560 deg C for about 30 min and then leached with a 3 to 9% aqueous solution of sodium carbonate.

  10. SOLVENT EXTRACTION PROCESS FOR URANIUM FROM CHLORIDE SOLUTIONS

    DOE Patents [OSTI]

    Blake, C.A. Jr.; Brown, K.B.; Horner, D.E.

    1960-05-24

    An improvement was made in a uranium extraction process wherein the organic extractant is a phosphine oxide. An aqueous solution containing phosphate ions or sulfate ions together with uranium is provided with a source of chloride ions during the extraction step. The presence of the chloride ions enables a phosphine oxide to extract uranium in the presence of strong uranium- complexing ions such as phosphate or sulfate ions.

  11. METHOD OF ELECTROPLATING ON URANIUM

    DOE Patents [OSTI]

    Rebol, E.W.; Wehrmann, R.F.

    1959-04-28

    This patent relates to a preparation of metallic uranium surfaces for receiving coatings, particularly in order to secure adherent electroplated coatings upon uranium metal. In accordance with the invention the uranium surface is pretreated by degreasing in trichloroethylene, followed by immersion in 25 to 50% nitric acid for several minutes, and then rinsed with running water, prior to pickling in trichloroacetic acid. The last treatment is best accomplished by making the uranium the anode in an aqueous solution of 50 per cent by weight trichloroacetic acid until work-distorted crystals or oxide present on the metal surface have been removed and the basic crystalline structure of the base metal has been exposed. Following these initial steps the metallic uranium is rinsed in dilute nitric acid and then electroplated with nickel. Adnerent firmly-bonded coatings of nickel are obtained.

  12. URANIUM COMPOSITIONS

    DOE Patents [OSTI]

    Allen, N.P.; Grogan, J.D.

    1959-05-12

    This patent relates to high purity uranium alloys characterized by improved stability to thermal cycling and low thermal neutron absorption. The high purity uranium alloy contains less than 0.1 per cent by weight in total amount of any ore or more of the elements such as aluminum, silicon, phosphorous, tin, lead, bismuth, niobium, and zinc.

  13. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    1. U.S. Forward-Cost Uranium Reserves by State, Year-End 2008 State $50/lb $100/lb Ore (million tons) Gradea (%) U3O8 (million lbs) Ore (million tons) Gradea (%) U3O8 (million lbs) Wyoming 145 0.076% 220 398 0.056% 446 New Mexico 64 0.140% 179 186 0.105% 390 Arizona, Colorado, Utah 22 0.145% 63 117 0.084% 198 Texas 15 0.089% 27 32 0.062% 40 Otherb 28 0.090% 50 95 0.081% 154 Total 275 0.098% 539 828 0.074% 1,227 a Average percent U3O8 per ton of ore. b Includes Alaska, California, Idaho, Montana,

  14. High-Quality Draft Genome Sequence of Desulfovibrio carbinoliphilus FW-101-2B, an Organic Acid-Oxidizing Sulfate-Reducing Bacterium Isolated from Uranium(VI)-Contaminated Groundwater

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Ramsay, Bradley D.; Hwang, Chiachi; Woo, Hannah L.; Carroll, Sue L.; Lucas, Susan; Han, James; Lapidus, Alla L.; Cheng, Jan-Fang; Goodwin, Lynne A.; Pitluck, Samuel; et al

    2015-03-12

    Desulfovibrio carbinoliphilus subsp. oakridgensis FW-101-2B is an anaerobic, organic acid/alcohol-oxidizing, sulfate-reducing δ-proteobacterium. FW-101-2B was isolated from contaminated groundwater at The Field Research Center at Oak Ridge National Lab after in situ stimulation for heavy metal-reducing conditions. The genome will help elucidate the metabolic potential of sulfate-reducing bacteria during uranium reduction.

  15. High-Quality Draft Genome Sequence of Desulfovibrio carbinoliphilus FW-101-2B, an Organic Acid-Oxidizing Sulfate-Reducing Bacterium Isolated from Uranium(VI)-Contaminated Groundwater

    SciTech Connect (OSTI)

    Ramsay, Bradley D.; Hwang, Chiachi; Woo, Hannah L.; Carroll, Sue L.; Lucas, Susan; Han, James; Lapidus, Alla L.; Cheng, Jan-Fang; Goodwin, Lynne A.; Pitluck, Samuel; Peters, Lin; Chertkov, Olga; Held, Brittany; Detter, John C.; Han, Cliff S.; Tapia, Roxanne; Land, Miriam L.; Hauser, Loren J.; Kyrpides, Nikos C.; Ivanova, Natalia N.; Mikhailova, Natalia; Pagani, Loanna; Woyke, Tanja; Arkin, Adam P.; Dehal, Paramvir; Chivian, Dylan; Criddle, Craig S.; Wu, Weimin; Chakraborty, Romy; Hazen, Terry C.; Fields, Matthew W.

    2015-03-12

    Desulfovibrio carbinoliphilus subsp. oakridgensis FW-101-2B is an anaerobic, organic acid/alcohol-oxidizing, sulfate-reducing ?-proteobacterium. FW-101-2B was isolated from contaminated groundwater at The Field Research Center at Oak Ridge National Lab after in situ stimulation for heavy metal-reducing conditions. The genome will help elucidate the metabolic potential of sulfate-reducing bacteria during uranium reduction.

  16. Uranium Processing Facility | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Uranium Processing Facility

  17. Uranium industry annual 1997

    SciTech Connect (OSTI)

    1998-04-01

    This report provides statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing.

  18. Economics of large-scale thorium oxide production: assessment of domestic resources

    SciTech Connect (OSTI)

    Young, J.K.; Bloomster, C.H.; Enderlin, W.I.; Morgenstern, M.H.; Ballinger, M.Y.; Drost, M.K.; Weakley, S.A.

    1980-02-01

    The supply curve illustrates that sufficient amounts of thorium exist supply a domestic thorium-reactor economy. Most likely costs of production range from $3 to $60/lb ThO/sub 2/. Near-term thorium oxide resources include the stockpiles in Ohio, Maryland, and Tennessee and the thorite deposits at Hall Mountain, Idaho. Costs are under $10/lb thorium oxide. Longer term economic deposits include Wet Mountain, Colorado; Lemhi Pass, Idaho; and Palmer, Michigan. Most likely costs are under $20/lb thorium oxide. Long-term deposits include Bald Mountain, Wyoming; Bear Lodge, Wyoming; and Conway, New Hampshire. Costs approximately equal or exceed $50/lb thorium oxide.

  19. JACKETING URANIUM

    DOE Patents [OSTI]

    Saller, H.A.; Keeler, J.R.

    1959-07-14

    The bonding to uranium of sheathing of iron or cobalt, or nickel, or alloys thereof is described. The bonding is accomplished by electro-depositing both surfaces to be joined with a coating of silver and amalgamating or alloying the silver layer with mercury or indium. Then the silver alloy is homogenized by exerting pressure on an assembly of the uranium core and the metal jacket, reducing the area of assembly and heating the assembly to homogenize by diffusion.

  20. TERNARY ALLOYS OF URANIUM, COLUMBIUM, AND ZIRCONIUM

    DOE Patents [OSTI]

    Foote, F.G.

    1960-08-01

    Ternary alloys of uranium are described which are useful as neutron- reflecting materials in a fast neutron reactor. They are especially resistant to corrosion caused by oxidative processes of gascous or aqueous origin and comprise uranium as the predominant metal with zirconiunn and niobium wherein the total content of the minor alloying elements is between 2 and 8% by weight.

  1. Subsurface Uranium Fate and Transport: Integrated Experiments and Modeling of Coupled Biogeochemical Mechanisms of Nanocrystalline Uraninite Oxidation by Fe(III)-(hydr)oxides - Project Final Report

    SciTech Connect (OSTI)

    Peyton, Brent M. [Montana State University; Timothy, Ginn R. [University of California Davis; Sani, Rajesh K. [South Dakota School of Mines and Technology

    2013-08-14

    Subsurface bacteria including sulfate reducing bacteria (SRB) reduce soluble U(VI) to insoluble U(IV) with subsequent precipitation of UO2. We have shown that SRB reduce U(VI) to nanometer-sized UO2 particles (1-5 nm) which are both intra- and extracellular, with UO2 inside the cell likely physically shielded from subsequent oxidation processes. We evaluated the UO2 nanoparticles produced by Desulfovibrio desulfuricans G20 under growth and non-growth conditions in the presence of lactate or pyruvate and sulfate, thiosulfate, or fumarate, using ultrafiltration and HR-TEM. Results showed that a significant mass fraction of bioreduced U (35-60%) existed as a mobile phase when the initial concentration of U(VI) was 160 M. Further experiments with different initial U(VI) concentrations (25 - 900 ?M) in MTM with PIPES or bicarbonate buffers indicated that aggregation of uraninite depended on the initial concentrations of U(VI) and type of buffer. It is known that under some conditions SRB-mediated UO2 nanocrystals can be reoxidized (and thus remobilized) by Fe(III)-(hydr)oxides, common constituents of soils and sediments. To elucidate the mechanism of UO2 reoxidation by Fe(III) (hydr)oxides, we studied the impact of Fe and U chelating compounds (citrate, NTA, and EDTA) on reoxidation rates. Experiments were conducted in anaerobic batch systems in PIPES buffer. Results showed EDTA significantly accelerated UO2 reoxidation with an initial rate of 9.5?M day-1 for ferrihydrite. In all cases, bicarbonate increased the rate and extent of UO2 reoxidation with ferrihydrite. The highest rate of UO2 reoxidation occurred when the chelator promoted UO2 and Fe(III) (hydr)oxide dissolution as demonstrated with EDTA. When UO2 dissolution did not occur, UO2 reoxidation likely proceeded through an aqueous Fe(III) intermediate as observed for both NTA and citrate. To complement to these laboratory studies, we collected U-bearing samples from a surface seep at the Rifle field site and

  2. PROCESS FOR PRODUCING URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-10-22

    A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ convented to UF/sub 6/ by reaction with a fluorinating agent. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reaction chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. The oven is then swept clean of hydrogen and the water vapor formed by means of nitrogen and then while continuing to maintain the temperature between 400 and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion to uranium tetrafluoride, the temperature of the reaction chamber is lowered to ahout 400 deg C, and elemental fluorine is used as the fluorinating agent for the conversion of UF/sub 4/ into UF/sub 6/. The fluorine gas is passed into the chamber, and the UF/sub 6/ formed passes out and is delivered to a condenser.

  3. Y-12 Knows Uranium | Y-12 National Security Complex

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Y-12 Knows Uranium Y-12 Knows Uranium Posted: July 22, 2013 - 3:45pm | Y-12 Report | Volume 10, Issue 1 | 2013 Y-12 produces many forms of uranium. They may be used in chemical processing steps on-site or shipped elsewhere to serve as raw materials for nuclear fuel or as research tools. All of uranium's uses, defense related and otherwise, are critical to the nation. Y-12's understanding of uranium, coupled with the site's work with enriched uranium metal, alloys, oxides, compounds and

  4. Uranium Metal Analysis via Selective Dissolution

    SciTech Connect (OSTI)

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

    2008-09-10

    Uranium metal, which is present in sludge held in the Hanford Site K West Basin, can create hazardous hydrogen atmospheres during sludge handling, immobilization, or subsequent transport and storage operations by its oxidation/corrosion in water. A thorough knowledge of the uranium metal concentration in sludge therefore is essential to successful sludge management and waste process design. The goal of this work was to establish a rapid routine analytical method to determine uranium metal concentrations as low as 0.03 wt% in sludge even in the presence of up to 1000-fold higher total uranium concentrations (i.e., up to 30 wt% and more uranium) for samples to be taken during the upcoming sludge characterization campaign and in future analyses for sludge handling and processing. This report describes the experiments and results obtained in developing the selective dissolution technique to determine uranium metal concentration in K Basin sludge.

  5. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    and Development Drilling","Mine Production of Uranium ","Uranium Concentrate Production ","Uranium Concentrate Shipments ","Employment " "Year","Drilling (million feet)"," ...

  6. Method for fabricating laminated uranium composites

    DOE Patents [OSTI]

    Chapman, L.R.

    1983-08-03

    The present invention is directed to a process for fabricating laminated composites of uranium or uranium alloys and at least one other metal or alloy. The laminated composites are fabricated by forming a casting of the molten uranium with the other metal or alloy which is selectively positioned in the casting and then hot-rolling the casting into a laminated plate in or around which the casting components are metallurgically bonded to one another to form the composite. The process of the present invention provides strong metallurgical bonds between the laminate components primarily since the bond disrupting surface oxides on the uranium or uranium alloy float to the surface of the casting to effectively remove the oxides from the bonding surfaces of the components.

  7. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    This book presents the GAO's views on the Department of Energy's (DOE) program to develop a new uranium enrichment technology, the atomic vapor laser isotope separation process (AVLIS). Views are drawn from GAO's ongoing review of AVLIS, in which the technical, program, and market issues that need to be addressed before an AVLIS plant is built are examined.

  8. Beneficial Uses of Depleted Uranium

    SciTech Connect (OSTI)

    Brown, C.; Croff, A.G.; Haire, M. J.

    1997-08-01

    Naturally occurring uranium contains 0.71 wt% {sup 235}U. In order for the uranium to be useful in most fission reactors, it must be enriched the concentration of the fissile isotope {sup 235}U must be increased. Depleted uranium (DU) is a co-product of the processing of natural uranium to produce enriched uranium, and DU has a {sup 235}U concentration of less than 0.71 wt%. In the United States, essentially all of the DU inventory is in the chemical form of uranium hexafluoride (UF{sub 6}) and is stored in large cylinders above ground. If this co-product material were to be declared surplus, converted to a stable oxide form, and disposed, the costs are estimated to be several billion dollars. Only small amounts of DU have at this time been beneficially reused. The U.S. Department of Energy (DOE) has begun the Beneficial Uses of DU Project to identify large-scale uses of DU and encourage its reuse for the primary purpose of potentially reducing the cost and expediting the disposition of the DU inventory. This paper discusses the inventory of DU and its rate of increase; DU disposition options; beneficial use options; a preliminary cost analysis; and major technical, institutional, and regulatory issues to be resolved.

  9. RECOVERY OF URANIUM VALUES FROM URANIUM BEARING RAW MATERIALS

    DOE Patents [OSTI]

    Michal, E.J.; Porter, R.R.

    1959-06-16

    Uranium leaching from ground uranium-bearing raw materials using MnO/sub 2/ in H/sub 2/SO/sub 4/ is described. The MnO/sub 2/ oxidizes U to the leachable hexavalent state. The MnO/sub 2/ does not replace Fe normally added, because the Fe complexes P and catalyzes the MnO/sub 2/ reaction. Three examples of continuous processes are given, but batch operation is also possible. The use of MnO/sub 2/ makes possible recovery of very low U values. (T.R.H.)

  10. Uranium industry annual 1996

    SciTech Connect (OSTI)

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

  11. Synthesis of uranium nitride and uranium carbide powder by carbothermic reduction

    SciTech Connect (OSTI)

    Dunwoody, J.T.; Stanek, C.R.; McClellan, K.J.; Voit, S.L.; Volz, H.M.; Hickman, R.R.

    2007-07-01

    Uranium nitride and uranium carbide are being considered as high burnup fuels in next generation nuclear reactors and accelerated driven systems for the transmutation of nuclear waste. The same characteristics that make nitrides and carbides candidates for these applications (i.e. favorable thermal properties, mutual solubility of nitrides, etc.), also make these compositions candidate fuels for space nuclear reactors. In this paper, we discuss the synthesis and characterization of depleted uranium nitride and carbide for a space nuclear reactor program. Importantly, this project emphasized that to synthesize high quality uranium nitride and carbide, it is necessary to understand the exact stoichiometry of the oxide feedstock. (authors)

  12. Disposition of Depleted Uranium Oxide

    SciTech Connect (OSTI)

    Crandall, J.L.

    2001-08-13

    This document summarizes environmental information which has been collected up to June 1983 at Savannah River Plant. Of particular interest is an updating of dose estimates from changes in methodology of calculation, lower cesium transport estimates from Steel Creek, and new sports fish consumption data for the Savannah River. The status of various permitting requirements are also discussed.

  13. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-08-01

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector.

  14. Oxide

    SciTech Connect (OSTI)

    2014-07-15

    Oxide is a modular framework for feature extraction and analysis of executable files. Oxide is useful in a variety of reverse engineering and categorization tasks relating to executable content.

  15. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    2. Maximum anticipated uranium market requirements of owners and operators of U.S. ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  16. COPPER COATED URANIUM ARTICLE

    DOE Patents [OSTI]

    Gray, A.G.

    1958-10-01

    Various techniques and methods for obtaining coppercoated uranium are given. Specifically disclosed are a group of complex uranium coatings having successive layers of nickel, copper, lead, and tin.

  17. PROCESSES OF RECOVERING URANIUM FROM A CALUTRON

    DOE Patents [OSTI]

    Baird, D.O.; Zumwalt, L.R.

    1958-07-15

    An improved process is described for recovering the residue of a uranium compound which has been subjected to treatment in a calutron, from the parts of the calutron disposed in the source region upon which the residue is deposited. The process may be utilized when the uranium compound adheres to a surface containing metals of the group consisting of copper, iron, chromium, and nickel. The steps comprise washing the surface with an aqueous acidic oxidizing solvent for the uranium whereby there is obtained an acidic aqueous Solution containing uranium as uranyl ions and metals of said group as impurities, treating the acidic solution with sodium acetate in the presenee of added sodium nitrate to precipitate the uranium as sodium uranyl acetate away from the impurities in the solution, and separating the sodium uranyl acetate from the solution.

  18. Scrap uranium recycling via electron beam melting

    SciTech Connect (OSTI)

    McKoon, R.

    1993-11-01

    A program is underway at the Lawrence Livermore National Laboratory (LLNL) to recycle scrap uranium metal. Currently, much of the material from forging and machining processes is considered radioactive waste and is disposed of by oxidation and encapsulation at significant cost. In the recycling process, uranium and uranium alloys in various forms will be processed by electron beam melting and continuously cast into ingots meeting applicable specifications for virgin material. Existing vacuum processing facilities at LLNL are in compliance with all current federal and state environmental, safety and health regulations for the electron beam melting and vaporization of uranium metal. One of these facilities has been retrofitted with an auxiliary electron beam gun system, water-cooled hearth, crucible and ingot puller to create an electron beam melt furnace. In this furnace, basic process R&D on uranium recycling will be performed with the goal of eventual transfer of this technology to a production facility.

  19. Uranium Industry Annual, 1992

    SciTech Connect (OSTI)

    Not Available

    1993-10-28

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

  20. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A.

    1990-01-01

    An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.

  1. Method of Making Uranium Dioxide Bodies

    DOE Patents [OSTI]

    Wilhelm, H. A.; McClusky, J. K.

    1973-09-25

    Sintered uranium dioxide bodies having controlled density are produced from U.sub.3 O.sub.8 and carbon by varying the mole ratio of carbon to U.sub.3 O.sub.8 in the mixture, which is compressed and sintered in a neutral or slightly oxidizing atmosphere to form dense slightly hyperstoichiometric uranium dioxide bodies. If the bodies are to be used as nuclear reactor fuel, they are subsequently heated in a hydrogen atmosphere to achieve stoichiometry. This method can also be used to produce fuel elements of uranium dioxide -- plutonium dioxide having controlled density.

  2. URANIUM EXTRACTION

    DOE Patents [OSTI]

    Harrington, C.D.; Opie, J.V.

    1958-07-01

    The recovery of uranium values from uranium ore such as pitchblende is described. The ore is first dissolved in nitric acid, and a water soluble nitrate is added as a salting out agent. The resulting feed solution is then contacted with diethyl ether, whereby the bulk of the uranyl nitrate and a portion of the impurities are taken up by the ether. This acid ether extract is then separated from the aqueous raffinate, and contacted with water causing back extractioa of the uranyl nitrate and impurities into the water to form a crude liquor. After separation from the ether extract, this crude liquor is heated to about 118 deg C to obtain molten uranyl nitrate hexahydratc. After being slightly cooled the uranyl nitrate hexahydrate is contacted with acid free diethyl ether whereby the bulk of the uranyl nitrate is dissolved into the ethcr to form a neutral ether solution while most of the impurities remain in the aqueous waste. After separation from the aqueous waste, the resultant ether solution is washed with about l0% of its volume of water to free it of any dissolved impurities and is then contacted with at least one half its volume of water whereby the uranyl nitrate is extracted into the water to form an aqueous product solution.

  3. PRODUCTION OF URANIUM MONOCARBIDE

    DOE Patents [OSTI]

    Powers, R.M.

    1962-07-24

    A method of making essentially stoichiometric uranium monocarbide by pelletizing a mixture of uranium tetrafluoride, silicon, and carbon and reacting the mixture at a temperature of approximately 1500 to 1700 deg C until the reaction goes to completion, forming uranium monocarbide powder and volatile silicon tetrafluoride, is described. The powder is then melted to produce uranium monocarbide in massive form. (AEC)

  4. FORMING TUBES AND RODS OF URANIUM METAL BY EXTRUSION

    DOE Patents [OSTI]

    Creutz, E.C.

    1959-01-27

    A method and apparatus are presented for the extrusion of uranium metal. Since uranium is very brittle if worked in the beta phase, it is desirable to extrude it in the gamma phase. However, in the gamma temperature range thc uranium will alloy with the metal of the extrusion dic, and is readily oxidized to a great degree. According to this patent, uranium extrusion in thc ganmma phase may be safely carried out by preheating a billet of uranium in an inert atmosphere to a trmperature between 780 C and 1100 C. The heated billet is then placed in an extrusion apparatus having dies which have been maintained at an elevated temperature for a sufficient length of time to produce an oxide film, and placing a copper disc between the uranium billet and the die.

  5. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Ross, Kyle W.; Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  6. URANIUM DECONTAMINATION

    DOE Patents [OSTI]

    Buckingham, J.S.; Carroll, J.L.

    1959-12-22

    A process is described for reducing the extractability of ruthenium, zirconium, and niobium values into hexone contained in an aqueous nitric acid uranium-containing solution. The solution is made acid-deficient, heated to between 55 and 70 deg C, and at that temperature a water-soluble inorganic thiosulfate is added. By this, a precipitate is formed which carries the bulk of the ruthenium, and the remainder of the ruthenium as well as the zirconium and niobium are converted to a hexone-nonextractable form. The rutheniumcontaining precipitate can either be removed from the solu tion or it can be dissolved as a hexone-non-extractable compound by the addition of sodium dichromate prior to hexone extraction.

  7. Review of the NURE Assessment of the U.S. Gulf Coast Uranium Province

    SciTech Connect (OSTI)

    Hall, Susan M.

    2013-09-15

    Historic exploration and development were used to evaluate the reliability of domestic uranium reserves and potential resources estimated by the U.S. Department of Energy national uranium resource evaluation (NURE) program in the U.S. Gulf Coast Uranium Province. NURE estimated 87 million pounds of reserves in the $30/lb U{sub 3}O{sub 8} cost category in the Coast Plain uranium resource region, most in the Gulf Coast Uranium Province. Since NURE, 40 million pounds of reserves have been mined, and 38 million pounds are estimated to remain in place as of 2012, accounting for all but 9 million pounds of U{sub 3}O{sub 8} in the reserve or production categories in the NURE estimate. Considering the complexities and uncertainties of the analysis, this study indicates that the NURE reserve estimates for the province were accurate. An unconditional potential resource of 1.4 billion pounds of U{sub 3}O{sub 8}, 600 million pounds of U{sub 3}O{sub 8} in the forward cost category of $30/lb U{sub 3}O{sub 8} (1980 prices), was estimated in 106 favorable areas by the NURE program in the province. Removing potential resources from the non-productive Houston embayment, and those reserves estimated below historic and current mining depths reduces the unconditional potential resource 33% to about 930 million pounds of U{sub 3}O{sub 8}, and that in the $30/lb cost category 34% to 399 million pounds of U{sub 3}O{sub 8}. Based on production records and reserve estimates tabulated for the region, most of the production since 1980 is likely from the reserves identified by NURE. The potential resource predicted by NURE has not been developed, likely due to a variety of factors related to the low uranium prices that have prevailed since 1980.

  8. Uranium industry annual 1994

    SciTech Connect (OSTI)

    1995-07-05

    The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ``Uranium Industry Annual Survey.`` Data collected on the ``Uranium Industry Annual Survey`` (UIAS) provide a comprehensive statistical characterization of the industry`s activities for the survey year and also include some information about industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ``Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,`` is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2.

  9. Uranium industry annual 1998

    SciTech Connect (OSTI)

    1999-04-22

    The Uranium Industry Annual 1998 (UIA 1998) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. It contains data for the period 1989 through 2008 as collected on the Form EIA-858, ``Uranium Industry Annual Survey.`` Data provides a comprehensive statistical characterization of the industry`s activities for the survey year and also include some information about industry`s plans and commitments for the near-term future. Data on uranium raw materials activities for 1989 through 1998, including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment, are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2008, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, and uranium inventories, are shown in Chapter 2. The methodology used in the 1998 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ``Uranium Industry Annual Survey`` is provided in Appendix C. The Form EIA-858 ``Uranium Industry Annual Survey`` is shown in Appendix D. For the readers convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix E along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 24 figs., 56 tabs.

  10. Adsorption study for uranium in Rocky Flats groundwater

    SciTech Connect (OSTI)

    Laul, J.C.; Rupert, M.C.; Harris, M.J.; Duran, A.

    1995-01-01

    Six adsorbents were studied to determine their effectiveness in removing uranium in Rocky Flats groundwater. The bench column and batch (Kd) tests showed that uranium can be removed (>99.9%) by four adsorbents. Bone Charcoal (R1O22); F-1 Alumina (granular activated alumina); BIOFIX (immobilized biological agent); SOPBPLUS (mixed metal oxide); Filtrasorb 300 (granular activated carbon); and Zeolite (clinoptilolite).

  11. Identification and In-vivo Characterization of a Novel OhrR Transcriptional Regulator in Burkholderia xenovorans LB400

    SciTech Connect (OSTI)

    Nguyen, Tinh T.; Mart-Arbona, Ricardo; Hall, Richard S.; Maity, Tuhin; Valdez, Yolanda E.; Dunbar, John M.; Unkefer, Clifford J.; Unkefer, Pat J.

    2013-05-21

    Transcriptional regulators (TRs) are an important and versatile group of proteins, yet very little progress has been achieved towards the discovery and annotation of their biological functions. We have characterized a previously unknown organic hydroperoxide resistance regulator from Burkholderia xenovoransLB400, Bxe_B2842, which is homologous to E. colis OhrR. Bxe_B2842 regulates the expression of an organic hydroperoxide resistance protein (OsmC). We utilized frontal affinity chromatography coupled with mass spectrometry (FAC-MS) and electrophoretic mobility gel shift assays (EMSA) to identify and characterize the possible effectors of the regulation by Bxe_B2842. Without an effector, Bxe_B2842 binds a DNA operator sequence (DOS) upstream of osmC. FAC-MS results suggest that 2-aminophenol binds to the protein and is potentially an effector molecule. EMSA analysis shows that 2-aminophenol also attenuates the Bxe_B2842s affinity for its DOS. EMSA analysis also shows that organic peroxides attenuate Bxe_B2842/DOS affinity, suggesting that binding of the TR to its DOS is regulated by the two-cysteine mechanism, common to TRs in this family. Bxe_B2842 is the first OhrR TR to have both oxidative and effector-binding mechanisms of regulation. Our paper reveals further mechanistic diversity TR mediated gene regulation and provides insights into methods for function discovery of TRs.

  12. Identification and In-vivo Characterization of a Novel OhrR Transcriptional Regulator in Burkholderia xenovorans LB400

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Nguyen, Tinh T.; Martí-Arbona, Ricardo; Hall, Richard S.; Maity, Tuhin; Valdez, Yolanda E.; Dunbar, John M.; Unkefer, Clifford J.; Unkefer, Pat J.

    2013-05-21

    Transcriptional regulators (TRs) are an important and versatile group of proteins, yet very little progress has been achieved towards the discovery and annotation of their biological functions. We have characterized a previously unknown organic hydroperoxide resistance regulator from Burkholderia xenovoransLB400, Bxe_B2842, which is homologous to E. coli’s OhrR. Bxe_B2842 regulates the expression of an organic hydroperoxide resistance protein (OsmC). We utilized frontal affinity chromatography coupled with mass spectrometry (FAC-MS) and electrophoretic mobility gel shift assays (EMSA) to identify and characterize the possible effectors of the regulation by Bxe_B2842. Without an effector, Bxe_B2842 binds a DNA operator sequence (DOS) upstream ofmore » osmC. FAC-MS results suggest that 2-aminophenol binds to the protein and is potentially an effector molecule. EMSA analysis shows that 2-aminophenol also attenuates the Bxe_B2842’s affinity for its DOS. EMSA analysis also shows that organic peroxides attenuate Bxe_B2842/DOS affinity, suggesting that binding of the TR to its DOS is regulated by the two-cysteine mechanism, common to TRs in this family. Bxe_B2842 is the first OhrR TR to have both oxidative and effector-binding mechanisms of regulation. Our paper reveals further mechanistic diversity TR mediated gene regulation and provides insights into methods for function discovery of TRs.« less

  13. NGSI FY15 Final Report. Innovative Sample Preparation for in-Field Uranium Isotopic Determinations

    SciTech Connect (OSTI)

    Yoshida, Thomas M.; Meyers, Lisa

    2015-11-10

    Our FY14 Final Report included an introduction to the project, background, literature search of uranium dissolution methods, assessment of commercial off the shelf (COTS) automated sample preparation systems, as well as data and results for dissolution of bulk quantities of uranium oxides, and dissolution of uranium oxides from swipe filter materials using ammonium bifluoride (ABF). Also, discussed were reaction studies of solid ABF with uranium oxide that provided a basis for determining the ABF/uranium oxide dissolution mechanism. This report details the final experiments for optimizing dissolution of U3O8 and UO2 using ABF and steps leading to development of a Standard Operating Procedure (SOP) for dissolution of uranium oxides on swipe filters.

  14. PRODUCTION OF PURIFIED URANIUM

    DOE Patents [OSTI]

    Burris, L. Jr.; Knighton, J.B.; Feder, H.M.

    1960-01-26

    A pyrometallurgical method for processing nuclear reactor fuel elements containing uranium and fission products and for reducing uranium compound; to metallic uranium is reported. If the material proccssed is essentially metallic uranium, it is dissolved in zinc, the sulution is cooled to crystallize UZn/sub 9/ , and the UZn/sub 9/ is distilled to obtain uranium free of fission products. If the material processed is a uranium compound, the sollvent is an alloy of zinc and magnesium and the remaining steps are the same.

  15. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Bailes, R.H.; Long, R.S.; Olson, R.S.; Kerlinger, H.O.

    1959-02-10

    A method is described for recovering uranium values from uranium bearing phosphate solutions such as are encountered in the manufacture of phosphate fertilizers. The solution is first treated with a reducing agent to obtain all the uranium in the tetravalent state. Following this reduction, the solution is treated to co-precipitate the rcduced uranium as a fluoride, together with other insoluble fluorides, thereby accomplishing a substantially complete recovery of even trace amounts of uranium from the phosphate solution. This precipitate usually takes the form of a complex fluoride precipitate, and after appropriate pre-treatment, the uranium fluorides are leached from this precipitate and rccovered from the leach solution.

  16. Neutron Reference Benchmark Field Specification: ACRR 44 Inch Lead-Boron (LB44) Bucket Environment (ACRR-LB44-CC-32-CL).

    SciTech Connect (OSTI)

    Vega, Richard Manuel; Parma, Edward J.; Griffin, Patrick J.; Vehar, David W.

    2015-07-01

    This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the 44 inch Lead-Boron (LB44) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 31 integral dosimetry measurements in the neutron field are reported.

  17. Method of recovering uranium hexafluoride

    DOE Patents [OSTI]

    Schuman, S.

    1975-12-01

    A method of recovering uranium hexafluoride from gaseous mixtures which comprises adsorbing said uranium hexafluoride on activated carbon is described.

  18. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    5. Shipments of uranium feed by owners and operators of U.S. civilian nuclear power ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  19. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    Inventories of uranium by owner as of end of year, 2011-15 thousand pounds U3O8 equivalent Inventories at the end of the year Owner of uranium inventory 2011 2012 2013 2014 P2015 ...

  20. Uranium Marketing Annual Report

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2013-15 2013 2014 2015 American Fuel Resources, LLC Advance Uranium Asset Management Ltd. AREVA ...

  1. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    0. Contracted purchases of uranium from suppliers by owners and operators of U.S. civilian ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  2. Uranium Marketing Annual Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    a. Foreign purchases, foreign sales, and uranium inventories owned by U.S. suppliers and ... Foreign sales U.S. supplier owned uranium inventories Owners and operators of U.S. ...

  3. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors by year, 2011-15 thousand pounds U3O8 equivalent Origin of uranium 2011 2012 2013 2014 P2015 ...

  4. METHOD FOR PURIFYING URANIUM

    DOE Patents [OSTI]

    Knighton, J.B.; Feder, H.M.

    1960-04-26

    A process is given for purifying a uranium-base nuclear material. The nuclear material is dissolved in zinc or a zinc-magnesium alloy and the concentration of magnesium is increased until uranium precipitates.

  5. NICKEL COATED URANIUM ARTICLE

    DOE Patents [OSTI]

    Gray, A.G.

    1958-10-01

    Nickel coatings on uranium and various methods of obtaining such coatings are described. Specifically disclosed are such nickel or nickel alloy layers as barriers between uranium and aluminum- silicon, chromium, or copper coatings.

  6. Electron Backscatter Diffraction (EBSD) Characterization of Uranium and Uranium Alloys

    SciTech Connect (OSTI)

    McCabe, Rodney J.; Kelly, Ann Marie; Clarke, Amy J.; Field, Robert D.; Wenk, H. R.

    2012-07-25

    Electron backscatter diffraction (EBSD) was used to examine the microstructures of unalloyed uranium, U-6Nb, U-10Mo, and U-0.75Ti. For unalloyed uranium, we used EBSD to examine the effects of various processes on microstructures including casting, rolling and forming, recrystallization, welding, and quasi-static and shock deformation. For U-6Nb we used EBSD to examine the microstructural evolution during shape memory loading. EBSD was used to study chemical homogenization in U-10Mo, and for U-0.75Ti, we used EBSD to study the microstructure and texture evolution during thermal cycling and deformation. The studied uranium alloys have significant microstructural and chemical differences and each of these alloys presents unique preparation challenges. Each of the alloys is prepared by a sequence of mechanical grinding and polishing followed by electropolishing with subtle differences between the alloys. U-6Nb and U-0.75Ti both have martensitic microstructures and both require special care in order to avoid mechanical polishing artifacts. Unalloyed uranium has a tendency to rapidly oxidize when exposed to air and a two-step electropolish is employed, the first step to remove the damaged surface layer resulting from the mechanical preparation and the second step to passivate the surface. All of the alloying additions provide a level of surface passivation and different one and two step electropolishes are employed to create good EBSD surfaces. Because of its low symmetry crystal structure, uranium exhibits complex deformation behavior including operation of multiple deformation twinning modes. EBSD was used to observe and quantify twinning contributions to deformation and to examine the fracture behavior. Figure 1 shows a cross section of two mating fracture surfaces in cast uranium showing the propensity of deformation twinning and intergranular fracture largely between dissimilarly oriented grains. Deformation of U-6Nb in the shape memory regime occurs by the motion

  7. PROCESS OF PURIFYING URANIUM

    DOE Patents [OSTI]

    Seaborg, G.T.; Orlemann, E.F.; Jensen, L.H.

    1958-12-23

    A method of obtaining substantially pure uranium from a uranium composition contaminated with light element impurities such as sodium, magnesium, beryllium, and the like is described. An acidic aqueous solution containing tetravalent uranium is treated with a soluble molybdate to form insoluble uranous molybdate which is removed. This material after washing is dissolved in concentrated nitric acid to obtaln a uranyl nitrate solution from which highly purified uranium is obtained by extraction with ether.

  8. PREPARATION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Lawroski, S.; Jonke, A.A.; Steunenberg, R.K.

    1959-10-01

    A process is described for preparing uranium hexafluoride from carbonate- leach uranium ore concentrate. The briquetted, crushed, and screened concentrate is reacted with hydrogen fluoride in a fluidized bed, and the uranium tetrafluoride formed is mixed with a solid diluent, such as calcium fluoride. This mixture is fluorinated with fluorine and an inert diluent gas, also in a fluidized bed, and the uranium hexafluoride obtained is finally purified by fractional distillation.

  9. Uranium industry annual 1995

    SciTech Connect (OSTI)

    1996-05-01

    The Uranium Industry Annual 1995 (UIA 1995) provides current statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1995 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the period 1986 through 2005 as collected on the Form EIA-858, ``Uranium Industry Annual Survey``. Data collected on the ``Uranium Industry Annual Survey`` provide a comprehensive statistical characterization of the industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1995, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. Data on uranium raw materials activities for 1986 through 1995 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2005, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. The methodology used in the 1995 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ``Uranium Industry Annual Survey`` is provided in Appendix C. For the reader`s convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix D along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 14 figs., 56 tabs.

  10. SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

    DOE Patents [OSTI]

    Spence, R.; Lister, M.W.

    1958-12-16

    Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

  11. TRACE ELEMENT ANALYSES OF URANIUM MATERIALS

    SciTech Connect (OSTI)

    Beals, D; Charles Shick, C

    2008-06-09

    The Savannah River National Laboratory (SRNL) has developed an analytical method to measure many trace elements in a variety of uranium materials at the high part-per-billion (ppb) to low part-per-million (ppm) levels using matrix removal and analysis by quadrapole ICP-MS. Over 35 elements were measured in uranium oxides, acetate, ore and metal. Replicate analyses of samples did provide precise results however none of the materials was certified for trace element content thus no measure of the accuracy could be made. The DOE New Brunswick Laboratory (NBL) does provide a Certified Reference Material (CRM) that has provisional values for a series of trace elements. The NBL CRM were purchased and analyzed to determine the accuracy of the method for the analysis of trace elements in uranium oxide. These results are presented and discussed in the following paper.

  12. Depleted uranium disposal options evaluation

    SciTech Connect (OSTI)

    Hertzler, T.J.; Nishimoto, D.D.; Otis, M.D.

    1994-05-01

    The Department of Energy (DOE), Office of Environmental Restoration and Waste Management, has chartered a study to evaluate alternative management strategies for depleted uranium (DU) currently stored throughout the DOE complex. Historically, DU has been maintained as a strategic resource because of uses for DU metal and potential uses for further enrichment or for uranium oxide as breeder reactor blanket fuel. This study has focused on evaluating the disposal options for DU if it were considered a waste. This report is in no way declaring these DU reserves a ``waste,`` but is intended to provide baseline data for comparison with other management options for use of DU. To PICS considered in this report include: Retrievable disposal; permanent disposal; health hazards; radiation toxicity and chemical toxicity.

  13. PRODUCTION OF URANIUM TETRAFLUORIDE

    DOE Patents [OSTI]

    Shaw, W.E.; Spenceley, R.M.; Teetzel, F.M.

    1959-08-01

    A method is presented for producing uranium tetrafluoride from the gaseous hexafluoride by feeding the hexafluoride into a high temperature zone obtained by the recombination of molecularly dissociated hydrogen. The molal ratio of hydrogen to uranium hexnfluoride is preferably about 3 to 1. Uranium tetrafluoride is obtained in a finely divided, anhydrous state.

  14. Final Uranium Leasing Program Programmatic Environmental Impact...

    Energy Savers [EERE]

    Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) Final Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) Uranium Leasing ...

  15. PROCESS FOR THE RECOVERY AND PURIFICATION OF URANIUM DEPOSITS

    DOE Patents [OSTI]

    Carter, J.M.; Kamen, M.D.

    1958-10-14

    A process is presented for recovering uranium values from UCl/sub 4/ deposits formed on calutrons. Such deposits are removed from the calutron parts by an aqueous wash solution which then contains the uranium values in addition to the following impurities: Ni, Cu, Fe, and Cr. This impurity bearing wash solution is treated with an oxidizing agent, and the oxidized solution is then treated with ammonia in order to precipitate the uranium as ammonium diuranate. The metal impurities of iron and chromium, which form insoluble hydroxides, are precipitated along with the uranium values. The precipitate is separated from the solution, dissolved in acid, and the solution again treated with ammonia and ammonium carbonate, which results in the precipitation of the metal impurities as hydroxides while the uranium values remain in solution.

  16. URANIUM EXTRACTION PROCESS USING SYNERGISTIC REAGENTS

    DOE Patents [OSTI]

    Schmitt, J.M.; Blake, C.A. Jr.; Brown, K.B.; Coleman, C.F.

    1958-11-01

    Improved methods are presented for recovering uranium values from aqueous solutions by organic solvent extraction. The improvement lies in the use, in combination, of two classes of organic compounds so that their extracting properties are enhanced synergistically. The two classes of organic compounds are dialkylphosphoric acid and certain neutral organophosphorus compounds such as trialkylphosphates, trialkylphosphonates, trlalkylphosphinates and trialkylphosphine oxides.

  17. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    Major U.S. Uranium Reserves

  18. PRODUCTION OF URANIUM

    DOE Patents [OSTI]

    Spedding, F.H.; Wilhelm, H.A.; Keller, W.H.

    1958-04-15

    The production of uranium metal by the reduction of uranium tetrafluoride is described. Massive uranium metal of high purily is produced by reacting uranium tetrafluoride with 2 to 20% stoichiometric excess of magnesium at a temperature sufficient to promote the reaction and then mantaining the reaction mass in a sealed vessel at temperature in the range of 1150 to 2000 d C, under a superatomospheric pressure of magnesium for a period of time sufficient 10 allow separation of liquid uranium and liquid magnesium fluoride into separate layers.

  19. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

    Carter, J.M.; Larson, C.E.

    1958-10-01

    A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

  20. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Yeager, J.H.

    1958-08-12

    In the prior art processing of uranium ores, the ore is flrst digested with nitric acid and filtered, and the uranium values are then extracted tom the filtrate by contacting with an organic solvent. The insoluble residue has been processed separately in order to recover any uranium which it might contain. The improvement consists in contacting a slurry, composed of both solution and residue, with the organic solvent prior to filtration. Tbe result is that uranium values contained in the residue are extracted along with the uranium values contained th the solution in one step.

  1. URANIUM SEPARATION PROCESS

    DOE Patents [OSTI]

    Hyde, E.K.; Katzin, L.I.; Wolf, M.J.

    1959-07-14

    The separation of uranium from a mixture of uranium and thorium by organic solvent extraction from an aqueous solution is described. The uranium is separrted from an aqueous mixture of uranium and thorium nitrates 3 N in nitric acid and containing salting out agents such as ammonium nitrate, so as to bring ihe total nitrate ion concentration to a maximum of about 8 N by contacting the mixture with an immiscible aliphatic oxygen containing organic solvent such as diethyl carbinol, hexone, n-amyl acetate and the like. The uranium values may be recovered from the organic phase by back extraction with water.

  2. Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide

    SciTech Connect (OSTI)

    Farawila, Anne F.; O'Hara, Matthew J.; Wai, Chien M.; Taylor, Harry Z.; Liao, Yu-Jung

    2012-07-31

    Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc -CO2) system was designed to mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc -CO2 and are capable of extracting uranium directly from UO2, UO3 and U3O8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO3 loaded Sc-CO2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc -CO2 pressure and temperature, TBP/HNO3 complex used

  3. Depleted uranium: A DOE management guide

    SciTech Connect (OSTI)

    1995-10-01

    The U.S. Department of Energy (DOE) has a management challenge and financial liability in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. The annual storage and maintenance cost is approximately $10 million. This report summarizes several studies undertaken by the DOE Office of Technology Development (OTD) to evaluate options for long-term depleted uranium management. Based on studies conducted to date, the most likely use of the depleted uranium is for shielding of spent nuclear fuel (SNF) or vitrified high-level waste (HLW) containers. The alternative to finding a use for the depleted uranium is disposal as a radioactive waste. Estimated disposal costs, utilizing existing technologies, range between $3.8 and $11.3 billion, depending on factors such as applicability of the Resource Conservation and Recovery Act (RCRA) and the location of the disposal site. The cost of recycling the depleted uranium in a concrete based shielding in SNF/HLW containers, although substantial, is comparable to or less than the cost of disposal. Consequently, the case can be made that if DOE invests in developing depleted uranium shielded containers instead of disposal, a long-term solution to the UF{sub 6} problem is attained at comparable or lower cost than disposal as a waste. Two concepts for depleted uranium storage casks were considered in these studies. The first is based on standard fabrication concepts previously developed for depleted uranium metal. The second converts the UF{sub 6} to an oxide aggregate that is used in concrete to make dry storage casks.

  4. About the Uranium Mine Team | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Mine Team About the Uranium Mine Team Text coming

  5. New Technique for Speciation of Uranium in Sediments Following Acetate-Stimulated Bioremediation

    SciTech Connect (OSTI)

    Not Available

    2011-06-22

    Acetate-stimulated bioremediation is a promising new technique for sequestering toxic uranium contamination from groundwater. The speciation of uranium in sediments after such bioremediation attempts remains unknown as a result of low uranium concentration, and is important to analyzing the stability of sequestered uranium. A new technique was developed for investigating the oxidation state and local molecular structure of uranium from field site sediments using X-Ray Absorption Spectroscopy (XAS), and was implemented at the site of a former uranium mill in Rifle, CO. Glass columns filled with bioactive Rifle sediments were deployed in wells in the contaminated Rifle aquifer and amended with a hexavalent uranium (U(VI)) stock solution to increase uranium concentration while maintaining field conditions. This sediment was harvested and XAS was utilized to analyze the oxidation state and local molecular structure of the uranium in sediment samples. Extended X-Ray Absorption Fine Structure (EXAFS) data was collected and compared to known uranium spectra to determine the local molecular structure of the uranium in the sediment. Fitting was used to determine that the field site sediments did not contain uraninite (UO{sub 2}), indicating that models based on bioreduction using pure bacterial cultures are not accurate for bioremediation in the field. Stability tests on the monomeric tetravalent uranium (U(IV)) produced by bioremediation are needed in order to assess the efficacy of acetate-stimulation bioremediation.

  6. Synthesis of Uranium Trichloride for the Pyrometallurgical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; J.C. Price; R.D. Mariani

    2011-11-01

    The pyroprocessing of used nuclear fuel via electrorefining requires the continued addition of uranium trichloride to sustain operations. Uranium trichloride is utilized as an oxidant in the system to allow separation of uranium metal from the minor actinides and fission products. The inventory of uranium trichloride had diminished to a point that production was necessary to continue electrorefiner operations. Following initial experimentation, cupric chloride was chosen as a reactant with uranium metal to synthesize uranium trichloride. Despite the variability in equipment and charge characteristics, uranium trichloride was produced in sufficient quantities to maintain operations in the electrorefiner. The results and conclusions from several experiments are presented along with a set of optimized operating conditions for the synthesis of uranium trichloride.

  7. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    S2. Uranium feed deliveries, enrichment services, and uranium loaded by owners and operators of U.S. civilian nuclear power reactors, 1994-2015 million pounds U3O8 equivalent million separative work units (SWU) Year Feed deliveries by owners and operators of U.S. civilian nuclear power reactors Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors U.S.-origin enrichment services purchased Foreign-origin enrichment services purchased Total purchased enrichment services

  8. Preparation of uranium compounds

    DOE Patents [OSTI]

    Kiplinger, Jaqueline L; Montreal, Marisa J; Thomson, Robert K; Cantat, Thibault; Travia, Nicholas E

    2013-02-19

    UI.sub.3(1,4-dioxane).sub.1.5 and UI.sub.4(1,4-dioxane).sub.2, were synthesized in high yield by reacting turnings of elemental uranium with iodine dissolved in 1,4-dioxane under mild conditions. These molecular compounds of uranium are thermally stable and excellent precursor materials for synthesizing other molecular compounds of uranium including alkoxide, amide, organometallic, and halide compounds.

  9. Summary of the engineering assessment of inactive uranium mill tailings, Shiprock Site, Shiprock, New Mexico

    SciTech Connect (OSTI)

    1981-07-01

    Ford, Bacon and Davis Utah Inc. has reevaluated the Shiprock site in order to revise the March 1977 engineering assessment of the problems resulting from the existence of radioactive uranium mill tailings at Shiprock, New Mexico. This engineering assessment has included the preparation of topographic maps, the performance of core drillings and radiometric measurements sufficient to determine areas and volumes of tailings and radiation exposures of individuals and nearby populations, the investigations of site hydrology and meteorology, and the evaluation and costing of alternative corrective actions. Radon gas released from the 1.5 million dry tons of tailings at the Shiprock site constitutes the most significant environmental impact, although windblown tailings and external gamma radiation also are factors. The eight alternative actions presented in this engineering assessment range from millsite decontamination with the addition of 3 m of the stabilization cover material (Option I), to removal of the tailings to remote disposal sites and decontamination of the tailings site (Options II through VIII). Cost estimates for the eight options range from about $13,400,000 for stabilization in place to about $37,900,000 for disposal at a distance of about 16 miles. Three principal alternatives for the reprocessing of the Shiprock tailings were examined: (a) heap leaching; (b) treatment at an existing mill; and(c) reprocessing at a new conventional mill constructed for tailings reprocessing. The cost of the uranium recovered would be about $230/lb by heap leach and $250/lb by conventional plant processes. The spot market price for uranium was $25/lb early in 1981. Therefore, reprocessing the tailings for uranium recovery is not economically attractive.

  10. PRODUCTION OF URANIUM

    DOE Patents [OSTI]

    Ruehle, A.E.; Stevenson, J.W.

    1957-11-12

    An improved process is described for the magnesium reduction of UF/sub 4/ to produce uranium metal. In the past, there have been undesirable premature reactions between the Mg and the bomb liner or the UF/sub 4/ before the actual ignition of the bomb reaction. Since these premature reactions impair the yield of uranium metal, they have been inhibited by forming a protective film upon the particles of Mg by reacting it with hydrated uranium tetrafluoride, sodium bifluoride, uranyl fluoride, or uranium trioxide. This may be accomplished by adding about 0.5 to 2% of the additive to the bomb charge.

  11. COATING URANIUM FROM CARBONYLS

    DOE Patents [OSTI]

    Gurinsky, D.H.; Storrs, S.S.

    1959-07-14

    Methods are described for making adherent corrosion resistant coatings on uranium metal. According to the invention, the uranium metal is heated in the presence of an organometallic compound such as the carbonyls of nickel, molybdenum, chromium, niobium, and tungsten at a temperature sufficient to decompose the metal carbonyl and dry plate the resultant free metal on the surface of the uranium metal body. The metal coated body is then further heated at a higher temperature to thermally diffuse the coating metal within the uranium bcdy.

  12. Uranium Dispersion & Dosimetry Model.

    Energy Science and Technology Software Center (OSTI)

    2002-03-22

    The Uranium Dispersion and Dosimetry (UDAD) program provides estimates of potential radiation exposure to individuals and to the general population in the vicinity of a uranium processing facility such as a uranium mine or mill. Only transport through the air is considered. Exposure results from inhalation, external irradiation from airborne and ground-deposited activity, and ingestion of foodstuffs. Individual dose commitments, population dose commitments, and environmental dose commitments are computed. The program was developed for applicationmore » to uranium mining and milling; however, it may be applied to dispersion of any other pollutant.« less

  13. Uranium Purchases Report

    Reports and Publications (EIA)

    1996-01-01

    Final issue. This report details natural and enriched uranium purchases as reported by owners and operators of commercial nuclear power plants. 1996 represents the most recent publication year.

  14. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Kaufman, D.

    1958-04-15

    A process of recovering uranium from very low-grade ore residues is described. These low-grade uraniumcontaining hydroxide precipitates, which also contain hydrated silica and iron and aluminum hydroxides, are subjected to multiple leachings with aqueous solutions of sodium carbonate at a pH of at least 9. This leaching serves to selectively extract the uranium from the precipitate, but to leave the greater part of the silica, iron, and aluminum with the residue. The uranium is then separated from the leach liquor by the addition of an acid in sufficient amount to destroy the carbonate followed by the addition of ammonia to precipitate uranium as ammonium diuranate.

  15. highly enriched uranium

    National Nuclear Security Administration (NNSA)

    and radioisotope supply capabilities of MURR and Nordion with General Atomics' selective gas extraction technology-which allows their low-enriched uranium (LEU) targets to remain...

  16. Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Resources, Inc., dba Cameco Resources Smith Ranch-Highland Operation Converse, Wyoming ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  17. MOA-2013-BLG-220Lb: Massive planetary companion to galactic-disk host

    SciTech Connect (OSTI)

    Yee, J. C.; Gould, A.; Gaudi, B. S.; Han, C.; Choi, J.-Y.; Hwang, K.-H.; Jung, Y. K.; Skowron, J.; Udalski, A.; Bond, I. A.; Hundertmark, M.; Monard, L. A. G.; Porritt, I.; Nelson, P.; Bozza, V.; Albrow, M. D.; Christie, G. W.; DePoy, D. L.; Lee, C.-U.; McCormick, J.; Collaboration: μFUN Collaboration),; MOA Collaboration),; OGLE Collaboration),; RoboNet Collaboration),; and others

    2014-07-20

    We report the discovery of MOA-2013-BLG-220Lb, which has a super-Jupiter mass ratio q = 3.01 ± 0.02 × 10{sup –3} relative to its host. The proper motion, μ = 12.5 ± 1 mas yr{sup –1}, is one of the highest for microlensing planets yet discovered, implying that it will be possible to separately resolve the host within ∼7 yr. Two separate lines of evidence imply that the planet and host are in the Galactic disk. The planet could have been detected and characterized purely with follow-up data, which has important implications for microlensing surveys, both current and into the Large Synoptic Survey Telescope (LSST) era.

  18. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    data set of uranium reserves that were published in the July 2010 report U.S. Uranium Reserves Estimates at http:www.eia.govcneafnuclearpagereservesures.html. ...

  19. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Number of Holes Feet (thousand) Number of Holes ...

  20. 2015 Uranium Market Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium ... received in 2015 Weighted-average price Number of purchase contracts for ...

  1. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Number of purchasers Quantity with reported price ...

  2. U.S.Uranium Reserves

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Reserves Data for: 2003 Release Date: June 2004 Next Release: Not determined Uranium Reserves Estimates The Energy Information Administration (EIA) has reported the...

  3. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    1 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  4. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Minimum ...

  5. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Origin of ...

  6. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    7 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  7. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    3 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  8. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    9 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  9. PROCESS FOR MAKING URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Rosen, R.

    1959-07-14

    A process is described for producing uranium hexafluoride by reacting uranium hexachloride with hydrogen fluoride at a temperature below about 150 deg C, under anhydrous conditions.

  10. Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) owns and manages an inventory of depleted uranium (DU), natural uranium (NU), and low-enriched uranium (LEU) that is currently stored in large cylinders as...

  11. RECOVERY OF URANIUM AND THORIUM FROM AQUEOUS SOLUTIONS

    DOE Patents [OSTI]

    Calkins, G.D.

    1958-06-10

    >A process is described for the recovery of uranium and thorium from monazite sand, which is frequently processed by treating it with a hot sodium hydroxide solution whereby a precipitate forms consisting mainly of oxides or hydroxides of the rare earths, thorium and uranium. The precipitate is dissolved in mineral acid, and the acid solution is then neutralized to a pH value of between 5.2 and 6.2 whereby both the uranium and thorium precipitate as the hydroxides, while substantially all the rare earth metal values present remain in the solution. The uranium and thoriunn can then be separated by dissolving the precipitate in a solution containing a mixture of alkali carbonate and alkali bicarbonate: and contacting the carbonate solution with a strong-base anion exchange resin whereby the uranium values are adsorbed on the resin while the thorium remains in solution.

  12. Record production on Gary No. 13 blast furnace with 450 lb./THM co-injection rates

    SciTech Connect (OSTI)

    Schuett, K.J.; White, D.G.

    1996-12-31

    Coal injection was initiated on No. 13 Blast Furnace in 1993 with 400 lb/THM achieved in 9 months. In early 1994, cold weather and coal preparation upsets led to the use of a second injectant, oil atomized by natural gas, to supplement the coal. Various combinations of coal and oil were investigated as total injection was increased to 450 lb/THM. Beginning in the last half of 1994, a continuing effort has been made to increase furnace production while maintaining this high co-injection level. Typical furnace production is now in excess of 10,000 THM/day compared with about 8500 THM/day in late 1993.

  13. Disposition of Uranium Oxide From Conversion of Depleted Uranium

    Energy Savers [EERE]

    Disability Employment Program Disability Employment Program The Department of Energy is committed to fostering a culture of diversity. We recognize that individuals with disabilities are an untapped talent pool and possess the skills and competencies that the Department needs to remain competitive. On July 26, 2010, President Obama issued Executive Order 13548. The Order emphasizes the Government's role as a catalyst in becoming a model employer for individuals with disabilities to include

  14. URANIUM SEPARATION PROCESS

    DOE Patents [OSTI]

    McVey, W.H.; Reas, W.H.

    1959-03-10

    The separation of uranium from an aqueous solution containing a water soluble uranyl salt is described. The process involves adding an alkali thiocyanate to the aqueous solution, contacting the resulting solution with methyl isobutyl ketons and separating the resulting aqueous and organic phase. The uranium is extracted in the organic phase as UO/sub 2/(SCN)/sub/.

  15. DECONTAMINATION OF URANIUM

    DOE Patents [OSTI]

    Spedding, F.H.; Butler, T.A.

    1962-05-15

    A process is given for separating fission products from uranium by extracting the former into molten aluminum. Phase isolation can be accomplished by selectively hydriding the uranium at between 200 and 300 deg C and separating the hydride powder from coarse particles of fissionproduct-containing aluminum. (AEC)

  16. Melting characteristics of the stainless steel generated from the uranium conversion plant

    SciTech Connect (OSTI)

    Choi, W.K.; Song, P.S.; Oh, W.Z.; Jung, C.H.; Min, B.Y.

    2007-07-01

    The partition ratio of cerium (Ce) and uranium (U) in the ingot, slag and dust phases has been investigated for the effect of the slag type, slag concentration and basicity in an electric arc melting process. An electric arc furnace (EAF) was used to melt the stainless steel wastes, simulated by uranium oxide and the real wastes from the uranium conversion plant in Korea Atomic Energy Research Institute (KAERI). The composition of the slag former used to capture the contaminants such as uranium, cerium, and cesium during the melt decontamination process generally consisted of silica (SiO{sub 2}), calcium oxide (CaO) and aluminum oxide (Al{sub 2}O{sub 3}). Also, Calcium fluoride (CaF{sub 2} ), nickel oxide (NiO), and ferric oxide (Fe{sub 2}O{sub 3}) were added to provide an increase in the slag fluidity and oxidative potential. Cerium was used as a surrogate for the uranium because the thermochemical and physical properties of cerium are very similar to those of uranium. Cerium was removed from the ingot phase to slag phase by up to 99% in this study. The absorption ratio of cerium was increased with an increase of the amount of the slag former. And the maximum removal of cerium occurred when the basicity index of the slag former was 0.82. The natural uranium (UO{sub 2}) was partitioned from the ingot phase to the slag phase by up to 95%. The absorption of the natural uranium was considerably dependent on the basicity index of the slag former and the composition of the slag former. The optimum condition for the removal of the uranium was about 1.5 for the basicity index and 15 wt% of the slag former. According to the increase of the amount of slag former, the absorption of uranium oxide in the slag phase was linearly increased due to an increase of its capacity to capture uranium oxide within the slag phase. Through experiments with various slag formers, we verified that the slag formers containing calcium fluoride (CaF{sub 2}) and a high amount of silica were more

  17. U. S. Attic Grsrgy 'bmiss~o3 U. S. Atomic lb-al-w CamLf3sion December 27, S.957

    Office of Legacy Management (LM)

    a . - 3 U. S. Attic Grsrgy 'bmiss~o3 U. S. Atomic lb-al-w CamLf3sion December 27, S.957

  18. METHOD FOR PURIFYING URANIUM

    DOE Patents [OSTI]

    Kennedy, J.W.; Segre, E.G.

    1958-08-26

    A method is presented for obtaining a compound of uranium in an extremely pure state and in such a condition that it can be used in determinations of the isotopic composition of uranium. Uranium deposited in calutron receivers is removed therefrom by washing with cold nitric acid and the resulting solution, coataining uranium and trace amounts of various impurities, such as Fe, Ag, Zn, Pb, and Ni, is then subjected to various analytical manipulations to obtain an impurity-free uranium containing solution. This solution is then evaporated on a platinum disk and the residue is ignited converting it to U2/sub 3//sub 8/. The platinum disk having such a thin film of pure U/sub 2/O/sub 8/ is suitable for use with isotopic determination techaiques.

  19. URANIUM PRECIPITATION PROCESS

    DOE Patents [OSTI]

    Thunaes, A.; Brown, E.A.; Smith, H.W.; Simard, R.

    1957-12-01

    A method for the recovery of uranium from sulfuric acid solutions is described. In the present process, sulfuric acid is added to the uranium bearing solution to bring the pH to between 1 and 1.8, preferably to about 1.4, and aluminum metal is then used as a reducing agent to convert hexavalent uranium to the tetravalent state. As the reaction proceeds, the pH rises amd a selective precipitation of uranium occurs resulting in a high grade precipitate. This process is an improvement over the process using metallic iron, in that metallic aluminum reacts less readily than metallic iron with sulfuric acid, thus avoiding consumption of the reducing agent and a raising of the pH without accomplishing the desired reduction of the hexavalent uranium in the solution. Another disadvantage to the use of iron is that positive ferric ions will precipitate with negative phosphate and arsenate ions at the pH range employed.

  20. Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

    DOE Patents [OSTI]

    Willit, James L.

    2010-09-21

    An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

  1. Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

    DOE Patents [OSTI]

    Willit, James L.

    2007-09-11

    An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

  2. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

    Kilner, S.B.

    1959-12-29

    A method is presented for separating and recovering uranium from a complex mixure of impurities. The uranium is dissolved to produce an aqueous acidic solution including various impurities. In accordance with one method, with the uranium in the uranyl state, hydrogen cyanide is introduced into the solution to complex the impurities. Subsequently, ammonia is added to the solution to precipitate the uraniunn as ammonium diuranate away from the impurities in the solution. Alternatively, the uranium is precipitated by adding an alkaline metal hydroxide. In accordance with the second method, the uranium is reduced to the uranous state in the solution. The reduced solution is then treated with solid alkali metal cyanide sufficient to render the solution about 0.1 to 1.0 N in cyanide ions whereat cyanide complex ions of the metal impurities are produced and the uranium is simultaneously precipituted as uranous hydroxide. Alternatively, hydrogen cyanide may be added to the reduced solution and the uranium precipitated subsequently by adding ammonium hydroxide or an alkali metal hydroxide. Other refinements of the method are also disclosed.

  3. India's Worsening Uranium Shortage

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-01-15

    As a result of NSG restrictions, India cannot import the natural uranium required to fuel its Pressurized Heavy Water Reactors (PHWRs); consequently, it is forced to rely on the expediency of domestic uranium production. However, domestic production from mines and byproduct sources has not kept pace with demand from commercial reactors. This shortage has been officially confirmed by the Indian Planning Commission’s Mid-Term Appraisal of the country’s current Five Year Plan. The report stresses that as a result of the uranium shortage, Indian PHWR load factors have been continually decreasing. The Uranium Corporation of India Ltd (UCIL) operates a number of underground mines in the Singhbhum Shear Zone of Jharkhand, and it is all processed at a single mill in Jaduguda. UCIL is attempting to aggrandize operations by establishing new mines and mills in other states, but the requisite permit-gathering and development time will defer production until at least 2009. A significant portion of India’s uranium comes from byproduct sources, but a number of these are derived from accumulated stores that are nearing exhaustion. A current maximum estimate of indigenous uranium production is 430t/yr (230t from mines and 200t from byproduct sources); whereas, the current uranium requirement for Indian PHWRs is 455t/yr (depending on plant capacity factor). This deficit is exacerbated by the additional requirements of the Indian weapons program. Present power generation capacity of Indian nuclear plants is 4350 MWe. The power generation target set by the Indian Department of Atomic Energy (DAE) is 20,000 MWe by the year 2020. It is expected that around half of this total will be provided by PHWRs using indigenously supplied uranium with the bulk of the remainder provided by breeder reactors or pressurized water reactors using imported low-enriched uranium.

  4. Depleted uranium management alternatives

    SciTech Connect (OSTI)

    Hertzler, T.J.; Nishimoto, D.D.

    1994-08-01

    This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process.

  5. Variations of Stream and Groundwater Uranium Concentrations with Rainfall in the North Carolina Piedmont

    SciTech Connect (OSTI)

    Price, V.

    2001-03-15

    This paper discusses how uranium contents of stream and groundwater samples from the North Carolina Piedmont vary systematically with rainfall. Data for three well-stream pairs sampled at three-week intervals for one year shows that uranium content of streams and groundwater correlates with the amount of local precipitation. The phenomenon is believed to be caused by oxidizing rainfall flushing uranium from the vadose zone.

  6. 300 Area Uranium Stabilization Through Polyphosphate Injection: Final Report

    SciTech Connect (OSTI)

    Vermeul, Vincent R.; Bjornstad, Bruce N.; Fritz, Brad G.; Fruchter, Jonathan S.; Mackley, Rob D.; Newcomer, Darrell R.; Mendoza, Donaldo P.; Rockhold, Mark L.; Wellman, Dawn M.; Williams, Mark D.

    2009-06-30

    The objective of the treatability test was to evaluate the efficacy of using polyphosphate injections to treat uranium-contaminated groundwater in situ. A test site consisting of an injection well and 15 monitoring wells was installed in the 300 Area near the process trenches that had previously received uranium-bearing effluents. This report summarizes the work on the polyphosphate injection project, including bench-scale laboratory studies, a field injection test, and the subsequent analysis and interpretation of the results. Previous laboratory tests have demonstrated that when a soluble form of polyphosphate is injected into uranium-bearing saturated porous media, immobilization of uranium occurs due to formation of an insoluble uranyl phosphate, autunite [Ca(UO2)2(PO4)2•nH2O]. These tests were conducted at conditions expected for the aquifer and used Hanford soils and groundwater containing very low concentrations of uranium (10-6 M). Because autunite sequesters uranium in the oxidized form U(VI) rather than forcing reduction to U(IV), the possibility of re-oxidation and subsequent re-mobilization is negated. Extensive testing demonstrated the very low solubility and slow dissolution kinetics of autunite. In addition to autunite, excess phosphorous may result in apatite mineral formation, which provides a long-term source of treatment capacity. Phosphate arrival response data indicate that, under site conditions, the polyphosphate amendment could be effectively distributed over a relatively large lateral extent, with wells located at a radial distance of 23 m (75 ft) reaching from between 40% and 60% of the injection concentration. Given these phosphate transport characteristics, direct treatment of uranium through the formation of uranyl-phosphate mineral phases (i.e., autunite) could likely be effectively implemented at full field scale. However, formation of calcium-phosphate mineral phases using the selected three-phase approach was problematic. Although

  7. Occurrence of Metastudtite (Uranium Peroxide Dihydrate) at a FUSRAP Site

    SciTech Connect (OSTI)

    Young, C.M.; Nelson, K.A.; Stevens, G.T.; Grassi, V.J.

    2006-07-01

    Uranium concentrations in groundwater in a localized area of a site exceed the USEPA Maximum Contaminant Level (MCL) by a factor of one thousand. Although the groundwater seepage velocity ranges up to 0.7 meters per day (m/day), data indicate that the uranium is not migrating in groundwater. We believe that the uranium is not mobile because of local geochemical conditions and the unstable nature of the uranium compound present at the site; uranium peroxide dihydrate (metastudtite). Metastudtite [UO{sub 4}.2(H{sub 2}O) or (U(O{sub 2})|O|(OH){sub 2}).3H{sub 2}O] has been identified at other sites as an alteration product in casks of spent nuclear fuel, but neither enriched nor depleted uranium were present at this site. Metastudtite was first identified as a natural mineral in 1983, although documented occurrences in the environment are uncommon. The U.S. Army Corps of Engineers (USACE) is conducting a remedial investigation at the DuPont Chambers Works in Deep water New Jersey under the Formerly Utilized Sites Remedial Action Program (FUSRAP) to evaluate radioactive contamination resulting from historical activities conducted in support of Manhattan Engineering District operations. From 1942 to 1947, Chambers Works converted uranium oxides to uranium tetrafluoride and uranium metal. More than half of the production at this facility resulted from the recovery process, where uranium-bearing dross and scrap were reacted with hydrogen peroxide to produce uranium peroxide dihydrate. The 280-hectare Chambers Works has produced some 600 products, including petrochemicals, aromatics, fluoro-chemicals, polymers, and elastomers. Contaminants resulting from these processes, including separate-phase petrochemicals, have also been detected within the boundaries of the FUSRAP investigation. USACE initiated remedial investigation field activities in 2002. The radionuclides of concern are natural uranium (U{sub nat}) and its short-lived progeny. Areas of impacted soil generally

  8. Fire testing of bare uranium hexafluoride cylinders

    SciTech Connect (OSTI)

    Pryor, W.A.

    1991-12-31

    In 1965, the Oak Ridge Gaseous Diffusion Plant (ORGDP), now the K-25 Site, conducted a series of tests in which bare cylinders of uranium hexafluoride (UF{sub 6}) were exposed to engulfing oil fires for the US Atomic Energy Commission (AEC), now the US Department of Energy (DOE). The tests are described and the results, conclusions, and observations are presented. Two each of the following types of cylinders were tested: 3.5-in.-diam {times} 7.5-in.-long cylinders of Monel (Harshaw), 5.0-in.-diam {times} 30-in.-long cylinders of Monel, and 8-in.-diam {times} 48-in.-long cylinders of nickel. The cylinders were filled approximately to the standard UF{sub 6} fill limits of 5, 55, and 250 lb, respectively, with a U-235 content of 0.22%. The 5-in.- and 8-in.-diam cylinders were tested individually with and without their metal valve covers. For the 3.5-in.-diam Harshaw cylinders and the 5.0-in.-diam cylinder without a valve cover the valves failed and UF{sub 6} was released. The remaining cylinders ruptured explosively in time intervals ranging from about 8.5 to 11 min.

  9. Fire testing of bare uranium hexafluoride cylinders

    SciTech Connect (OSTI)

    Pryor, W.A.

    1991-12-31

    In 1965, the Oak Ridge Gaseous Diffusion Plant (ORGDP), now the K-25 Site, conducted a series of tests in which bare cylinders of uranium hexafluoride (UF{sub 6}) were exposed to engulfing oil fires for the US Atomic Energy Commission (AEC), now the US Department of Energy (DOE). The tests are described and the results, conclusions, and observations are presented. Two each of the following types of cylinders were tested: 3.5-in.-diam {times} 7.5-in.-long cylinders of Monel (Harshaw), 5.0-in.-diam {times} x 30-in.-long cylinders of Monel, and 8-in.-diam {times} 48-in.-long cylinders of nickel. The cylinders were filled approximately to the standard UF{sub 6} fill limits of 5, 55, and 250 lb, respectively, with a U-235 content of 0.22%. The 5-in.- and 8-in.-diam cylinders were tested individually with and without their metal valve covers. For the 3.5-in.-diam Harshaw cylinders and the 5.0-in.-diam cylinder without a valve cover, the valves failed and UF{sub 6} was released. The remaining 6 cylinders ruptured explosively in time intervals ranging from about 8.5 to 11 min.

  10. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    4. Deliveries of uranium feed for enrichment by owners and operators of U.S. civilian nuclear power reactors by origin country and delivery year, 2013-15 thousand pounds U3O8 ...

  11. Uranium Marketing Annual Report -

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    9. Contracted purchases of uranium by owners and operators of U.S. civilian nuclear power reactors, signed in 2015, by delivery year, 2016-25 thousand pounds U3O8 equivalent Year ...

  12. PURIFICATION OF URANIUM FUELS

    DOE Patents [OSTI]

    Niedrach, L.W.; Glamm, A.C.

    1959-09-01

    An electrolytic process of refining or decontaminating uranium is presented. The impure uranium is made the anode of an electrolytic cell. The molten salt electrolyte of this cell comprises a uranium halide such as UF/sub 4/ or UCl/sub 3/ and an alkaline earth metal halide such as CaCl/sub 2/, BaF/sub 2/, or BaCl/sub 2/. The cathode of the cell is a metal such as Mn, Cr, Co, Fe, or Ni which forms a low melting eutectic with U. The cell is operated at a temperature below the melting point of U. In operation the electrodeposited uranium becomes alloyed with the metal of the cathode, and the low melting alloy thus formed drips from the cathode.

  13. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Stevenson, J.W.; Werkema, R.G.

    1959-07-28

    The recovery of uranium from magnesium fluoride slag obtained as a by- product in the production of uranium metal by the bomb reduction prccess is presented. Generally the recovery is accomplished by finely grinding the slag, roasting ihe ground slag air, and leaching the roasted slag with a hot, aqueous solution containing an excess of the sodium bicarbonate stoichiometrically required to form soluble uranium carbonate complex. The roasting is preferably carried out at between 425 and 485 deg C for about three hours. The leaching is preferably done at 70 to 90 deg C and under pressure. After leaching and filtration the uranium may be recovered from the clear leach liquor by any desired method.

  14. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    9. Summary production statistics of the U.S. uranium industry, 1993-2015 Year Exploration and development surface drilling (million feet) Exploration and development drilling expenditures 1 (million dollars) Mine production of uranium (million pounds U3O8) Uranium concentrate production (million pounds U3O8) Uranium concentrate shipments (million pounds U3O8) Employment (person-years) 1993 1.1 5.7 2.1 3.1 3.4 871 1994 0.7 1.1 2.5 3.4 6.3 980 1995 1.3 2.6 3.5 6.0 5.5 1,107 1996 3.0 7.2 4.7 6.3

  15. ANODIC TREATMENT OF URANIUM

    DOE Patents [OSTI]

    Kolodney, M.

    1959-02-01

    A method is presented for effecting eloctrolytic dissolution of a metallic uranium article at a uniform rate. The uranium is made the anode in an aqueous phosphoric acid solution containing nitrate ions furnished by either ammonium nitrate, lithium nitrate, sodium nitrate, or potassium nitrate. A stainless steel cathode is employed and electrolysls carried out at a current density of about 0.1 to 1 ampere per square inch.

  16. PROCESS FOR PRODUCING URANIUM HALIDES

    DOE Patents [OSTI]

    Murphree, E.V.

    1957-10-29

    A process amd associated apparatus for producing UF/sub 4/ from U/sub 3/ O/sub 8/ by a fluidized'' technique are reported. The U/sub 3/O/sub 8/ is first reduced to UO/sub 2/ by reaction with hydrogen, and the lower oxide of uranium is then reacted with gaseous HF to produce UF/sub 4/. In each case the reactant gas is used, alone or in combination with inert gases, to fluidize'' the finely divided reactant solid. The complete setup of the plant equipment including bins, reactor and the associated piping and valving, is described. An auxiliary fluorination reactor allows for the direct production of UF/sub 6/ from UF/sub 4/ and fluorine gas, or if desired, UF/sub 4/ may be collected as the product.

  17. Ionic Liquids as templating agents in formation of uranium-containing nanomaterials

    DOE Patents [OSTI]

    Visser, Ann E; Bridges, Nicholas J

    2014-06-10

    A method for forming nanoparticles containing uranium oxide is described. The method includes combining a uranium-containing feedstock with an ionic liquid to form a mixture and holding the mixture at an elevated temperature for a period of time to form the product nanoparticles. The method can be carried out at low temperatures, for instance less than about 300.degree. C.

  18. Statistical design of a uranium corrosion experiment

    SciTech Connect (OSTI)

    Wendelberger, Joanne R; Moore, Leslie M

    2009-01-01

    This work supports an experiment being conducted by Roland Schulze and Mary Ann Hill to study hydride formation, one of the most important forms of corrosion observed in uranium and uranium alloys. The study goals and objectives are described in Schulze and Hill (2008), and the work described here focuses on development of a statistical experiment plan being used for the study. The results of this study will contribute to the development of a uranium hydriding model for use in lifetime prediction models. A parametric study of the effect of hydrogen pressure, gap size and abrasion on hydride initiation and growth is being planned where results can be analyzed statistically to determine individual effects as well as multi-variable interactions. Input to ESC from this experiment will include expected hydride nucleation, size, distribution, and volume on various uranium surface situations (geometry) as a function of age. This study will also address the effect of hydrogen threshold pressure on corrosion nucleation and the effect of oxide abrasion/breach on hydriding processes. Statistical experiment plans provide for efficient collection of data that aids in understanding the impact of specific experiment factors on initiation and growth of corrosion. The experiment planning methods used here also allow for robust data collection accommodating other sources of variation such as the density of inclusions, assumed to vary linearly along the cast rods from which samples are obtained.

  19. PROCESSING OF NEUTRON-IRRADIATED URANIUM

    DOE Patents [OSTI]

    Hopkins, H.H. Jr.

    1960-09-01

    An improved "Purex" process for separating uranium, plutonium, and fission products from nitric acid solutions of neutron-irradiated uranium is offered. Uranium is first extracted into tributyl phosphate (TBP) away from plutonium and fission products after adjustment of the acidity from 0.3 to 0.5 M and heating from 60 to 70 deg C. Coextracted plutonium, ruthenium, and fission products are fractionally removed from the TBP by three scrubbing steps with a 0.5 M nitric acid solution of ferrous sulfamate (FSA), from 3.5 to 5 M nitric acid, and water, respectively, and the purified uranium is finally recovered from the TBP by precipitation with an aqueous solution of oxalic acid. The plutonium in the 0.3 to 0.5 M acid solution is oxidized to the tetravalent state with sodium nitrite and extracted into TBP containing a small amount of dibutyl phosphate (DBP). Plutonium is then back-extracted from the TBP-DBP mixture with a nitric acid solution of FSA, reoxidized with sodium nitrite in the aqueous strip solution obtained, and once more extracted with TBP alone. Finally the plutonium is stripped from the TBP with dilute acid, and a portion of the strip solution thus obtained is recycled into the TBPDBP for further purification.

  20. PREPARATION OF REFRACTORY OXIDE CRYSTALS

    DOE Patents [OSTI]

    Grimes, W.R.; Shaffer, J.H.; Watson, G.M.

    1962-11-13

    A method is given for preparing uranium dioxide, thorium oxide, and beryllium oxide in the form of enlarged individual crystals. The surface of a fused alkali metal halide melt containing dissolved uranium, thorium, or beryllium values is contacted with a water-vapor-bearing inert gas stream at a rate of 5 to 10 cubic centimeters per minute per square centimeter of melt surface area. Growth of individual crystals is obtained by prolonged contact. Beryllium oxide-coated uranium dioxide crystals are prepared by disposing uranium dioxide crystals 5 to 20 microns in diameter in a beryllium-containing melt and contacting the melt with a water-vapor-bearing inert gas stream in the same manner. (AEC)

  1. PROCESSES FOR SEPARATING AND RECOVERING CONSTITUENTS OF NEUTRON IRRADIATED URANIUM

    DOE Patents [OSTI]

    Connick, R.E.; Gofman, J.W.; Pimentel, G.C.

    1959-11-10

    Processes are described for preparing plutonium, particularly processes of separating plutonium from uranium and fission products in neutron-irradiated uraniumcontaining matter. Specifically, plutonium solutions containing uranium, fission products and other impurities are contacted with reducing agents such as sulfur dioxide, uranous ion, hydroxyl ammonium chloride, hydrogen peroxide, and ferrous ion whereby the plutoninm is reduced to its fluoride-insoluble state. The reduced plutonium is then carried out of solution by precipitating niobic oxide therein. Uranium and certain fission products remain behind in the solution. Certain other fission products precipitate along with the plutonium. Subsequently, the plutonium and fission product precipitates are redissolved, and the solution is oxidized with oxidizing agents such as chlorine, peroxydisulfate ion in the presence of silver ion, permanganate ion, dichromate ion, ceric ion, and a bromate ion, whereby plutonium is oxidized to the fluoride-soluble state. The oxidized solution is once again treated with niobic oxide, thus precipitating the contamirant fission products along with the niobic oxide while the oxidized plutonium remains in solution. Plutonium is then recovered from the decontaminated solution.

  2. Uranium (VI) solubility in carbonate-free ERDA-6 brine

    SciTech Connect (OSTI)

    Lucchini, Jean-francois; Khaing, Hnin; Reed, Donald T

    2010-01-01

    When present, uranium is usually an element of importance in a nuclear waste repository. In the Waste Isolation Pilot Plant (WIPP), uranium is the most prevalent actinide component by mass, with about 647 metric tons to be placed in the repository. Therefore, the chemistry of uranium, and especially its solubility in the WIPP conditions, needs to be well determined. Long-term experiments were performed to measure the solubility of uranium (VI) in carbonate-free ERDA-6 brine, a simulated WIPP brine, at pC{sub H+} values between 8 and 12.5. These data, obtained from the over-saturation approach, were the first repository-relevant data for the VI actinide oxidation state. The solubility trends observed pointed towards low uranium solubility in WIPP brines and a lack of amphotericity. At the expected pC{sub H+} in the WIPP ({approx} 9.5), measured uranium solubility approached 10{sup -7} M. The objective of these experiments was to establish a baseline solubility to further investigate the effects of carbonate complexation on uranium solubility in WIPP brines.

  3. OGLE-2008-BLG-355Lb: A massive planet around a late-type star

    SciTech Connect (OSTI)

    Koshimoto, N.; Sumi, T.; Fukagawa, M.; Shibai, H.; Udalski, A.; Bennett, D. P.; Bond, I. A.; Ling, C. H.; Rattenbury, N.; Botzler, C. S.; Freeman, M.; Abe, F.; Furusawa, K.; Itow, Y.; Masuda, K.; Matsubara, Y.; Fukui, A.; Muraki, Y.; Ohnishi, K.; Saito, To.; Collaboration: MOA Collaboration; OGLE Collaboration; and others

    2014-06-20

    We report the discovery of a massive planet, OGLE-2008-BLG-355Lb. The light curve analysis indicates a planet:host mass ratio of q = 0.0118 ± 0.0006 at a separation of 0.877 ± 0.010 Einstein radii. We do not measure a significant microlensing parallax signal and do not have high angular resolution images that could detect the planetary host star. Therefore, we do not have a direct measurement of the host star mass. A Bayesian analysis, assuming that all host stars have equal probability to host a planet with the measured mass ratio, implies a host star mass of M{sub h}=0.37{sub −0.17}{sup +0.30} M{sub ⊙} and a companion of mass M{sub P}=4.6{sub −2.2}{sup +3.7}M{sub J}, at a projected separation of r{sub ⊥}=1.70{sub −0.30}{sup +0.29} AU. The implied distance to the planetary system is D {sub L} = 6.8 ± 1.1 kpc. A planetary system with the properties preferred by the Bayesian analysis may be a challenge to the core accretion model of planet formation, as the core accretion model predicts that massive planets are far more likely to form around more massive host stars. This core accretion model prediction is not consistent with our Bayesian prior of an equal probability of host stars of all masses to host a planet with the measured mass ratio. Thus, if the core accretion model prediction is right, we should expect that follow-up high angular resolution observations will detect a host star with a mass in the upper part of the range allowed by the Bayesian analysis. That is, the host would probably be a K or G dwarf.

  4. D0 Decomissioning : Storage of Depleted Uranium Modules Inside D0 Calorimeters after the Termination of D0 Experiment

    SciTech Connect (OSTI)

    Sarychev, Michael; /Fermilab

    2011-09-21

    Dzero liquid Argon calorimeters contain hadronic modules made of depleted uranium plates. After the termination of DO detector's operation, liquid Argon will be transferred back to Argon storage Dewar, and all three calorimeters will be warmed up. At this point, there is no intention to disassemble the calorimeters. The depleted uranium modules will stay inside the cryostats. Depleted uranium is a by-product of the uranium enrichment process. It is slightly radioactive, emits alpha, beta and gamma radiation. External radiation hazards are minimal. Alpha radiation has no external exposure hazards, as dead layers of skin stop it; beta radiation might have effects only when there is a direct contact with skin; and gamma rays are negligible - levels are extremely low. Depleted uranium is a pyrophoric material. Small particles (such as shavings, powder etc.) may ignite with presence of Oxygen (air). Also, in presence of air and moisture it can oxidize. Depleted uranium can absorb moisture and keep oxidizing later, even after air and moisture are excluded. Uranium oxide can powder and flake off. This powder is also pyrographic. Uranium oxide may create health problems if inhaled. Since uranium oxide is water soluble, it may enter the bloodstream and cause toxic effects.

  5. Uranium Processing Facility Team Signs Partnering Agreement ...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Processing Facility ... Uranium Processing Facility Team Signs Partnering Agreement ... Nuclear Security, LLC; John Eschenberg, Uranium Processing Facility Project Office; Brian ...

  6. Nuclear Fuel Facts: Uranium | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Facts: Uranium Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Uranium is a silvery-white metallic chemical element in the periodic table, with atomic number 92. It is assigned the chemical symbol U. A uranium atom has 92 protons and 92 electrons, of which 6 are valence electrons. Uranium has the highest atomic weight (19 kg m) of all naturally occurring elements. Uranium occurs naturally in low concentrations in soil, rock and water, and is commercially extracted from uranium-bearing

  7. Assessment of the Group 5-6 (LB C2, LB S2, LV S1) Stack Sampling Probe Locations for Compliance with ANSI/HPS N13.1 1999

    SciTech Connect (OSTI)

    Glissmeyer, John A.; Flaherty, Julia E.; Piepel, Gregory F.

    2011-03-11

    This document reports on a series of tests to assess the proposed air sampling locations for the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Group 5-6 exhaust stacks with respect to the applicable criteria regarding the placement of an air sampling probe. The LB-C2, LV-S1, and LB S2 exhaust stacks were tested together as a group (Test Group 5-6) because the common factor in their design is that the last significant flow disturbance upstream of the air sampling probe is a reduction in duct diameter. Federal regulations( ) require that a sampling probe be located in the exhaust stack according to the criteria of the American National Standards Institute/Health Physics Society (ANSI/HPS) N13.1-1999, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities. These criteria address the capability of the sampling probe to extract a sample that represents the effluent stream. The testing on scale models of the stacks conducted for this project was part of the River Protection Project—Waste Treatment Plant Support Program under Contract No. DE-AC05-76RL01830 according to the statement of work issued by Bechtel National Inc. (BNI, 24590-QL-SRA-W000-00101, N13.1-1999 Stack Monitor Scale Model Testing and Qualification, Revision 1, 9/12/2007) and Work Authorization 09 of Memorandum of Agreement 24590-QL-HC9-WA49-00001. The internal Pacific Northwest National Laboratory (PNNL) project for this task is 53024, Work for Hanford Contractors Stack Monitoring. The testing described in this document was further guided by the Test Plan Scale Model Testing the Waste Treatment Plant LB-C2, LB-S2, and LV-S1 (Test Group 5-6) Stack Air Sampling Positions (TP-RPP-WTP-594). The tests conducted by PNNL during 2009 and 2010 on the Group 5-6 scale model systems are described in this report. The series of tests consists of various measurements taken over a grid of points in the duct cross-section at the designed sampling

  8. The development of an integrated multistage fluid bed retorting process. [Kentort II process--50-lb/hr

    SciTech Connect (OSTI)

    Carter, S.; Stehn, J.; Vego, A.; Taulbee, D.

    1992-05-01

    This report summarizes the progress made on the development of an integrated multistage fluidized bed retorting process (KENTORT II) during the period of January 1, 1992 through March 31, 1992. The KENTORT II process includes integral fluidized bed zones for pyrolysis, gasification, and combustion of the oil shale. The purpose of this program is to design and test the KENTORT II process at the 50-lb/hr scale. The design of the 50-lb/hr KENTORT II retort was completed and fabrication is ready to begin. Data from the cold-flow model of the system and operating experience from the 5-lb/hr unit were used as the basis for the design. In another aspect of the program, a study of the cracking and coking kinetics of shale oil vapors was continued. A mathematical model was implemented to characterize the important mass transfer effects of the system. This model will be eventually broadened to become a general fluidized bed coking model. In addition, experiments were performed to examine the effects of surface area, initial carbon content and steam treatment on coking activity. From the data that has been collected to-date, it appears that the coking activity of the tested substrates can be explained in terms of porosity (surface area and pore volume) and the initial carbon content of the solid.

  9. Reductive stripping process for the recovery of uranium from wet-process phosphoric acid

    DOE Patents [OSTI]

    Hurst, Fred J.; Crouse, David J.

    1984-01-01

    A reductive stripping flow sheet for recovery of uranium from wet-process phosphoric acid is described. Uranium is stripped from a uranium-loaded organic phase by a redox reaction converting the uranyl to uranous ion. The uranous ion is reoxidized to the uranyl oxidation state to form an aqueous feed solution highly concentrated in uranium. Processing of this feed through a second solvent extraction cycle requires far less stripping reagent as compared to a flow sheet which does not include the reductive stripping reaction.

  10. SEPARATION OF URANIUM AND PLUTONIUM OXIDES

    DOE Patents [OSTI]

    Benedict, G.E.; Lyon, W.L.

    1961-12-01

    ABS>A method of separating a mixture of UO/sub 2/ and PuO/sub 2/ is given which comprises immersing the mixture in a fused NaCl-KCl bath, chlorinating with chlorine or phosgene, and preferentially electrolytically or chemically reducing the UO/sub 2/Cl/sub 2/ so produced to UO/sub 2/ and filtering it out. (AEC)