National Library of Energy BETA

Sample records for uranium oxide conversion

  1. CRAD, Training- 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 Training Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

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

    National Nuclear Security Administration (NNSA)

    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

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

    Broader source: Energy.gov [DOE]

    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 Paducah site; transportation of all cylinders (DUF6, enriched, and empty) currently stored at the East Tennessee Technology Park (ETTP) near Oak Ridge, Tennessee, to Portsmouth; construction of a new cylinder storage yard at Portsmouth (if required) for ETTP cylinders; transportation of depleted uranium conversion products and waste materials to a disposal facility; transportation and sale of the hydrogen fluoride (HF) produced as a conversion coproduct; and neutralization of HF to calcium fluoride and its sale or disposal in the event that the HF product is not sold.

  12. 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.

  13. Uranium Mining, Conversion, and Enrichment Industries

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

    Analysis of Potential Impacts of Uranium Transfers on the Domestic Uranium Mining, Conversion, and Enrichment Industries May 1, 2015 ii EXECUTIVE SUMMARY: The Department of Energy ("Department" or "DOE") plans to transfer the equivalent of up to 2,100 metric tons ("MTU") of natural uranium per year (with a higher total for calendar year 2015, mainly because of transfers already executed or under way before today's determination). These transfers would include 1,600

  14. Method for converting uranium oxides to uranium metal

    DOE Patents [OSTI]

    Duerksen, Walter K. (Norris, TN)

    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.

  15. DOE Extends Contract to Operate Depleted Uranium Hexafluoride Conversion

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

    Plants | Department of Energy Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants DOE Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants December 24, 2015 - 11:06am Addthis LEXINGTON, Ky. (Dec. 24, 2015) - The U.S. Department of Energy's Office of Environmental Management (EM) today announced it is extending its contract for Operations of Depleted Uranium Hexafluoride (DUF6) Conversion Facilities at Paducah, Kentucky and Portsmouth, Ohio for a

  16. DOE Extends Contract to Operate Depleted Uranium Hexafluoride Conversion

    Office of Environmental Management (EM)

    Plants | Department of Energy Contract to Operate Depleted Uranium Hexafluoride Conversion Plants DOE Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants December 24, 2015 - 10:00am Addthis Media Contact Brad Mitzelfelt, 859-219-4035 brad.mitzelfelt@lex.doe.gov LEXINGTON, Ky. - The U.S. Department of Energy's Office of Environmental Management (EM) today announced it is extending its contract for Operations of Depleted Uranium Hexafluoride (DUF6) Conversion Facilities

  17. 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)

  18. Conversion of depleted uranium hexafluoride to a solid uranium compound

    DOE Patents [OSTI]

    Rothman, Alan B. (Willowbrook, IL); Graczyk, Donald G. (Lemont, IL); Essling, Alice M. (Elmhurst, IL); Horwitz, E. Philip (Naperville, IL)

    2001-01-01

    A process for converting UF.sub.6 to a solid uranium compound such as UO.sub.2 and CaF. The UF.sub.6 vapor form is contacted with an aqueous solution of NH.sub.4 OH at a pH greater than 7 to precipitate at least some solid uranium values as a solid leaving an aqueous solution containing NH.sub.4 OH and NH.sub.4 F and remaining uranium values. The solid uranium values are separated from the aqueous solution of NH.sub.4 OH and NH.sub.4 F and remaining uranium values which is then diluted with additional water precipitating more uranium values as a solid leaving trace quantities of uranium in a dilute aqueous solution. The dilute aqueous solution is contacted with an ion-exchange resin to remove substantially all the uranium values from the dilute aqueous solution. The dilute solution being contacted with Ca(OH).sub.2 to precipitate CaF.sub.2 leaving dilute NH.sub.4 OH.

  19. 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.

  20. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    SciTech Connect (OSTI)

    1995-07-05

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

  1. 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.

  2. Uranium-233 purification and conversion to stabilized ceramic grade urania for LWBR fuel fabrication (LWBR Development Program)

    SciTech Connect (OSTI)

    Lloyd, R.

    1980-10-01

    High purity ceramic grade urania (/sup 233/UO/sub 2/) used in manufacturing the fuel for the Light Water Breeder Reactor (LWBR) core was made from uranium-233 that was obtained by irradiating thoria under special conditions to result in not more than 10 ppM of uranium-232 in the recovered uranium-233 product. A developmental study established the operating parameters of the conversion process for transforming the uranium-233 into urania powder with the appropriate chemical and physical attributes for use in fabricating the LWBR core fuel. This developmental study included the following: (a) design of an ion exchange purification process for removing the gamma-emitting alpha-decay daughters of uranium-232, to reduce the gamma-radiation field of the uranium-233 during LWBR fuel manufacture; (b) definition of the parameters for precipitating the uranium-233 as ammonium uranate (ADU) and for reducing the ADU with hydrogen to yield a urania conversion product of the proper particle size, surface area and sinterability for use in manufacturing the LWBR fuel; (c) establishment of parameters and design of equipment for stabilizing the urania conversion product to prevent it from undergoing excessive oxidation on exposure to the air during LWBR fuel manufacturing operations; and (d) development of a procedure and a facility to reprocess the unirradiated thoria-urania fuel scrap from the LWBR core manufacturing operations to recover the uranium-233 and convert it into high purity ceramic grade urania for LWBR core fabrication.

  3. 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.

  4. EIS-0329: Proposed Construction, Operation, Decontamination/Decommissioning of Depleted Uranium Hexafluoride Conversion Facilities

    Broader source: Energy.gov [DOE]

    This EIS analyzes DOE's proposal to construct, operate, maintain, and decontaminate and decommission two depleted uranium hexafluoride (DUF 6) conversion facilities, at Portsmouth, Ohio, and Paducah, Kentucky.

  5. 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.

  6. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    SciTech Connect (OSTI)

    Renfro, David G; Cook, David Howard; Freels, James D; Griffin, Frederick P; Ilas, Germina; Sease, John D; Chandler, David

    2012-03-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  7. Conversion of actinide and RE oxides into nitrates and their recovery into fluids

    SciTech Connect (OSTI)

    Bondin, V.V.; Bychkov, S.I.; Efremov, I.G.; Revenko, Y.A.; Babain, V.A.; Murzin, A.A.; Romanovsky, V.N; Fedorov, Y.S.; Shadrin, A.Y.; Ryabkova, N.V.; Li, E.N.

    2007-07-01

    The conditions for uranium oxides completely convert into uranyl nitrate hexahydrate in nitrogen tetra-oxide media (75 deg. C, 0,5-3,0 MPa, [UO{sub x}]:[H{sub 2}O]:[NO{sub 2}]=1:8:6) were found out. The conversion of Pu contained simulator of oxide spent nuclear fuel of thermal reactors was successfully demonstrated. The possibility of uranium recovery up to 95% from TR SNF without plutonium separation from FP is practically showed, what corresponds with Non-proliferation Treaty. (authors)

  8. EIS-0359: Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky Site

    Broader source: Energy.gov [DOE]

    This site-specific EIS considers the construction, operation, maintenance, and decontamination and decommissioning of the proposed depleted uranium hexafluoride (DUF6) conversion facility at three locations within the Paducah site; transportation of depleted uranium conversion products and waste materials to a disposal facility; transportation and sale of the hydrogen fluoride (HF) produced as a conversion co-product; and neutralization of HF to calcium fluoride and its sale or disposal in the event that the HF product is not sold.

  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. 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. ;...

  11. Simultaneous constraint and phase conversion processing of oxide superconductors

    DOE Patents [OSTI]

    Li, Qi (Marlborough, MA); Thompson, Elliott D. (Coventry, RI); Riley, Jr., Gilbert N. (Marlborough, MA); Hellstrom, Eric E. (Madison, WI); Larbalestier, David C. (Madison, WI); DeMoranville, Kenneth L. (Jefferson, MA); Parrell, Jeffrey A. (Roselle Park, NJ); Reeves, Jodi L. (Madison, WI)

    2003-04-29

    A method of making an oxide superconductor article includes subjecting an oxide superconductor precursor to a texturing operation to orient grains of the oxide superconductor precursor to obtain a highly textured precursor; and converting the textured oxide superconducting precursor into an oxide superconductor, while simultaneously applying a force to the precursor which at least matches the expansion force experienced by the precursor during phase conversion to the oxide superconductor. The density and the degree of texture of the oxide superconductor precursor are retained during phase conversion. The constraining force may be applied isostatically.

  12. Structural Sequestration of Uranium in Bacteriogenic Manganese Oxides

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

    Sequestration of Uranium in Bacteriogenic Manganese Oxides Samuel M. Webb (Stanford Synchrotron Radiation Laboratory), Bradley M. Tebo (Oregon Health and Science University), and John Bargar (Stanford Synchrotron Radiation Laboratory). Microbial Respiration Figure 1. Manganese oxides precipitated around a spore (cell) of the marine Mn(II)-oxidizing bacterium, Bacillus sp., strain SG-1. This cell is about 0.5 µm diameter (small axis). Manganese oxides are formed in soils, watersheds, and sea

  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. DOE Issues Request for Quotations for Depleted Uranium Hexafluoride Conversion Technical Services

    Broader source: Energy.gov [DOE]

    Cincinnati – The U.S. Department of Energy (DOE) today issued a Request for Quotation (RFQ) for engineering and operations technical services to support the Portsmouth Paducah Project Office and the oversight of operations of the Depleted Uranium Hexafluoride (DUF6) Conversion Project located in Paducah KY, and Portsmouth OH.

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

    Office of Scientific and Technical Information (OSTI)

    Covalency in oxidized uranium Citation Details In-Document Search This content will become publicly available on July 28, 2016 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. ; Weng, T.-C. Publication Date: 2015-07-29 OSTI Identifier: 1204690 Grant/Contract Number: AC52-07NA27344; AC02-05CH11231 Type: Publisher's Accepted Manuscript Journal Name: Physical Review B Additional

  16. Use of the UNCLE Facility to Assess Integrated Online Monitoring Systems for Detection of Diversions at Uranium Conversion Facilities

    SciTech Connect (OSTI)

    Dewji, Shaheen A; Chapman, Jeffrey Allen; Lee, Denise L; Rauch, Eric; Hertel, Nolan

    2011-01-01

    Historically, the approach to safeguarding nuclear material in the front end of the fuel cycle was implemented only at the stage when UF6 was declared as feedstock for enrichment plants. Recent International Atomic Energy Agency (IAEA) circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exist. Oak Ridge National Laboratory has developed the Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility to simulate the full-scale operating conditions for a purified uranium-bearing aqueous stream exiting the solvent extraction process conducted in a natural uranium conversion plant (NUCP) operating at 6000 MTU/year. Monitoring instruments, including the 3He passive neutron detector developed at Los Alamos National Laboratory and the Endress+Hauser Promass 83F Coriolis meter, have been tested at UNCLE and field tested at Springfields. The field trials demonstrated the need to perform full-scale equipment testing under controlled conditions prior to field deployment of operations and safeguards monitoring at additional plants. Currently, UNCLE is testing neutron-based monitoring for detection of noncompliant activities; however, gamma-ray source term monitoring is currently being explored complementary to the neutron detector in order to detect undeclared activities in a more timely manner. The preliminary results of gamma-ray source term modeling and monitoring at UNCLE are being analyzed as part of a comprehensive source term and detector benchmarking effort. Based on neutron source term detection capabilities, alternative gamma-based detection and monitoring methods will be proposed to more effectively monitor NUCP operations in verifying or detecting deviations from declared conversion activities.

  17. Natural uranium/conversion services/enrichment services

    SciTech Connect (OSTI)

    1993-12-31

    This article is the 1993 uranium market summary. During this reporting period, there were 50 deals in the concentrates market, 26 deals in the UF6 market, and 14 deals for enrichment services. In the concentrates market, the restricted value closed $0.15 higher at $9.85, and the unrestricted value closed down $0.65 at $7.00. In the UF6 market, restricted prices fluctuated and closed higher at $31.00, and unrestricted prices closed at their initial value of $24.75. The restricted transaction value closed at $10.25 and the unrestricted value closed at $7.15. In the enrichment services market, the restricted value moved steadily higher to close at $84.00 per SWU, and the unrestricted value closed at its initial value of $68.00 per SWU.

  18. Complex oxides useful for thermoelectric energy conversion

    DOE Patents [OSTI]

    Majumdar, Arunava (Orinda, CA); Ramesh, Ramamoorthy (Moraga, CA); Yu, Choongho (College Station, TX); Scullin, Matthew L. (Berkeley, CA); Huijben, Mark (Enschede, NL)

    2012-07-17

    The invention provides for a thermoelectric system comprising a substrate comprising a first complex oxide, wherein the substrate is optionally embedded with a second complex oxide. The thermoelectric system can be used for thermoelectric power generation or thermoelectric cooling.

  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. Development of Integrated Online Monitoring Systems for Detection of Diversion at Natural Uranium Conversion Facilities

    SciTech Connect (OSTI)

    Dewji, Shaheen A; Lee, Denise L; Croft, Stephen; McElroy, Robert Dennis; Hertel, Nolan; Chapman, Jeffrey Allen; Cleveland, Steven L

    2013-01-01

    Recent work at Oak Ridge National Laboratory (ORNL) has focused on some source term modeling of uranyl nitrate (UN) as part of a comprehensive validation effort employing gamma-ray detector instrumentation for the detection of diversion from declared conversion activities. Conversion, the process by which natural uranium ore (yellowcake) is purified and converted through a series of chemical processes into uranium hexafluoride gas (UF6), has historically been excluded from the nuclear safeguards requirements of the 235U-based nuclear fuel cycle. The undeclared diversion of this product material could potentially provide feedstock for a clandestine weapons program for state or non-state entities. Given the changing global political environment and the increased availability of dual-use nuclear technology, the International Atomic Energy Agency has evolved its policies to emphasize safeguarding this potential feedstock material in response to dynamic and evolving potential diversion pathways. To meet the demand for instrumentation testing at conversion facilities, ORNL developed the Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in a natural uranium conversion plant. This work investigates gamma-ray signatures of UN circulating in the UNCLE facility and evaluates detector instrumentation sensitivity to UN for safeguards applications. These detector validation activities include assessing detector responses to the UN gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10-90g U/L of naturally enriched UN will be presented. A range of gamma-ray lines was examined and self-attenuation factors were calculated, in addition to attenuation for transmission measurement of density, concentration and enrichment. A detailed uncertainty analysis will be presented providing insights into instrumentation limitations to spoofing.

  1. HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

    SciTech Connect (OSTI)

    1995-09-01

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation.

  2. 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 these materials.

  3. Transition Metal Oxide Alloys as Potential Solar Energy Conversion Materials

    SciTech Connect (OSTI)

    Toroker, Maytal; Carter, Emily A.

    2013-02-21

    First-row transition metal oxides (TMOs) are inexpensive potentia alternative materials for solar energy conversion devices. However, some TMOs, such as manganese(II) oxide, have band gaps that are too large for efficiently absorbing solar energy. Other TMOs, such as iron(II) oxide, have conduction and valence band edges with the same orbital character that may lead to unfavorably high electronhole recombination rates. Another limitation of iron(II) oxide is that the calculated valence band edge is not positioned well for oxidizing water. We predict that key properties, including band gaps, band edge positions, and possibly electronhole recombination rates, may be improved by alloying TMOs that have different band alignments. A new metric, the band gap center offset, is introduced for simple screening of potential parent materials. The concept is illustrated by calculating the electronic structure of binary oxide alloys that contain manganese, nickel, iron, zinc, and/or magnesium, within density functional theory (DFT)+U and hybrid DFT theories. We conclude that alloys of iron(II) oxide are worth evaluating further as solar energy conversion materials.

  4. 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].

  5. 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

  6. 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 (UO32H2O) 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 21C and 50C. 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 UO2.0040.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21C than the particles prepared at 50C. 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.

  7. Enterprise Assessments Targeted Review of the Paducah Depleted Uranium Hexafluoride Conversion Facility Fire Protection Program … September 2015

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

    Targeted Review of the Paducah Depleted Uranium Hexafluoride Conversion Facility Fire Protection Program September 2015 Office of Nuclear Safety and Environmental Assessments Office of Environment, Safety and Health Assessments Office of Enterprise Assessments U.S. Department of Energy i Table of Contents Acronyms ..................................................................................................................................................... ii Executive Summary

  8. DOE Issues Final Request for Proposal for the Operation of Depleted Uranium Hexafluoride (DUF6) Conversion Facilities

    Broader source: Energy.gov [DOE]

    Cincinnati -- The U.S. Department of Energy (DOE) today issued a Final Request for Proposal (RFP), for the Operation of Depleted Uranium Hexafluoride (DUF6) Conversion Facilities at Paducah, Kentucky and Portsmouth, Ohio. A cost-plus award fee and firm-fixed-price contract line item contract will be awarded from this Final RFP.

  9. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    SciTech Connect (OSTI)

    Stillman, J.; Feldman, E.; Foyto, L; Kutikkad, K; McKibben, J C; Peters, N.; Stevens, J.

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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...

  15. 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.

  16. Draft Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Portsmouth, Ohio, Site

    SciTech Connect (OSTI)

    N /A

    2003-11-28

    This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF{sub 6}) conversion facility at the U.S. Department of Energy (DOE) Portsmouth site in Ohio (Figure S-1). The proposed facility would convert the DUF{sub 6} stored at Portsmouth to a more stable chemical form suitable for use or disposal. The facility would also convert the DUF{sub 6} from the East Tennessee Technology Park (ETTP) site near Oak Ridge, Tennessee. In a Notice of Intent (NOI) published in the Federal Register on September 18, 2001 (Federal Register, Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF{sub 6} conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (United States Code, Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (Code of Federal Regulations, Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF{sub 6} conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a Federal Register Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (D&D) of the proposed conversion facility at three alternative locations within the Portsmouth site; from the transportation of all ETTP cylinders (DUF{sub 6}, low-enriched UF6 [LEU-UF{sub 6}], and empty) to Portsmouth; from the transportation of depleted uranium conversion products to a disposal facility; and from the transportation, sale, use, or disposal of the fluoride-containing conversion products (hydrogen fluoride [HF] or calcium fluoride [CaF{sub 2}]). An option of shipping the ETTP cylinders to Paducah is also considered. In addition, this EIS evaluates a no action alternative, which assumes continued storage of DUF{sub 6} in cylinders at the Portsmouth and ETTP sites. A separate EIS (DOE/EIS-0359) evaluates potential environmental impacts for the proposed Paducah conversion facility.

  17. Draft Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky, Site

    SciTech Connect (OSTI)

    N /A

    2003-11-28

    This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF{sub 6}) conversion facility at the U.S. Department of Energy (DOE) Paducah site in northwestern Kentucky (Figure S-1). The proposed facility would convert the DUF{sub 6} stored at Paducah to a more stable chemical form suitable for use or disposal. In a Notice of Intent (NOI) published in the ''Federal Register'' (FR) on September 18, 2001 (''Federal Register'', Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF{sub 6} conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (''United States Code'', Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (''Code of Federal Regulations'', Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF{sub 6} conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a ''Federal Register'' Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (D&D) of the proposed conversion facility at three alternative locations within the Paducah site; from the transportation of depleted uranium conversion products to a disposal facility; and from the transportation, sale, use, or disposal of the fluoride-containing conversion products (hydrogen fluoride [HF] or calcium fluoride [CaF{sub 2}]). Although not part of the proposed action, an option of shipping all cylinders (DUF{sub 6}, low-enriched UF{sub 6} [LEU-UF{sub 6}], and empty) stored at the East Tennessee Technology Park (ETTP) near Oak Ridge, Tennessee, to Paducah rather than to Portsmouth is also considered. In addition, this EIS evaluates a no action alternative, which assumes continued storage of DUF{sub 6} in cylinders at the Paducah site. A separate EIS (DOE/EIS-0360) evaluates the potential environmental impacts for the proposed Portsmouth conversion facility.

  18. Study on Evaluation of Project Management Data for Decommissioning of Uranium Refining and Conversion Plant - 12234

    SciTech Connect (OSTI)

    Usui, Hideo; Izumo, Sari; Tachibana, Mitsuo; Shibahara, Yuji; Morimoto, Yasuyuki; Tokuyasu, Takashi; Takahashi, Nobuo; Tanaka, Yoshio; Sugitsue, Noritake

    2012-07-01

    Some of nuclear facilities that would no longer be required have been decommissioned in JAEA (Japan Atomic Energy Agency). A lot of nuclear facilities have to be decommissioned in JAEA in near future. To implement decommissioning of nuclear facilities, it was important to make a rational decommissioning plan. Therefore, project management data evaluation system for dismantling activities (PRODIA code) has been developed, and will be useful for making a detailed decommissioning plan for an object facility. Dismantling of dry conversion facility in the uranium refining and conversion plant (URCP) at Ningyo-toge began in 2008. During dismantling activities, project management data such as manpower and amount of waste generation have been collected. Such collected project management data has been evaluated and used to establish a calculation formula to calculate manpower for dismantling equipment of chemical process and calculate manpower for using a green house (GH) which was a temporary structure for preventing the spread of contaminants during dismantling. In the calculation formula to calculate project management data related to dismantling of equipment, the relation of dismantling manpower to each piece of equipment was evaluated. Furthermore, the relation of dismantling manpower to each chemical process was evaluated. The results showed promise for evaluating dismantling manpower with respect to each chemical process. In the calculation formula to calculate project management data related to use of the GH, relations of GH installation manpower and removal manpower to GH footprint were evaluated. Furthermore, the calculation formula for secondary waste generation was established. In this study, project management data related to dismantling of equipment and use of the GH were evaluated and analyzed. The project management data, manpower for dismantling of equipment, manpower for installation and removal of GH, and secondary waste generation from GH were considered. Establishment of the calculation formula for dismantling of each kind of equipment makes it possible to evaluate manpower for dismantling the whole facility. However, it is not easy to prepare calculation formula for all kinds of equipment that exist in the facility. Therefore, a simpler evaluation method was considered to calculate manpower based on facility characteristics. The results showed promise for evaluating dismantling manpower with respect to each chemical process. For dismantling of contaminated equipment, a GH has been used for protection of the spread of contamination. The use of a GH increases manpower for installation and removal of GH etc. Moreover, structural materials of the GH such as plastic sheets, adhesive tape become a burnable secondary waste. To create an effective dismantling plan, it is necessary to carefully consider use of a GH preliminarily. Thus, an evaluation method of project management data such as manpower and secondary waste generation was considered. The results showed promise for evaluating project management data of GH by using established calculation formula. (authors)

  19. Biocatalytic conversion of ethylene to ethylene oxide using an engineered toluene monooxygenase

    SciTech Connect (OSTI)

    Carlin, DA; Bertolani, SJ; Siegel, JB

    2015-01-01

    Mutants of toluene o-xylene monooxygenase are demonstrated to oxidize ethylene to ethylene oxide in vivo at yields of >99%. The best mutant increases ethylene oxidation activity by >5500-fold relative to the native enzyme. This is the first report of a recombinant enzyme capable of carrying out this industrially significant chemical conversion.

  20. Excess Uranium Management

    Broader source: Energy.gov [DOE]

    The Department's Notice of Issues for Public Comment on the effects of DOE transfers of excess uranium on domestic uranium mining, conversion, and enrichment industries.

  1. 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.

  2. 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.

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

    DOE Patents [OSTI]

    McLean, II, William (Oakland, CA); Miller, Philip E. (Livermore, CA)

    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.

  4. Feasibility Study on the Use of On-line Multivariate Statistical Process Control for Safeguards Applications in Natural Uranium Conversion Plants

    SciTech Connect (OSTI)

    Ladd-Lively, Jennifer L

    2014-01-01

    The objective of this work was to determine the feasibility of using on-line multivariate statistical process control (MSPC) for safeguards applications in natural uranium conversion plants. Multivariate statistical process control is commonly used throughout industry for the detection of faults. For safeguards applications in uranium conversion plants, faults could include the diversion of intermediate products such as uranium dioxide, uranium tetrafluoride, and uranium hexafluoride. This study was limited to a 100 metric ton of uranium (MTU) per year natural uranium conversion plant (NUCP) using the wet solvent extraction method for the purification of uranium ore concentrate. A key component in the multivariate statistical methodology is the Principal Component Analysis (PCA) approach for the analysis of data, development of the base case model, and evaluation of future operations. The PCA approach was implemented through the use of singular value decomposition of the data matrix where the data matrix represents normal operation of the plant. Component mole balances were used to model each of the process units in the NUCP. However, this approach could be applied to any data set. The monitoring framework developed in this research could be used to determine whether or not a diversion of material has occurred at an NUCP as part of an International Atomic Energy Agency (IAEA) safeguards system. This approach can be used to identify the key monitoring locations, as well as locations where monitoring is unimportant. Detection limits at the key monitoring locations can also be established using this technique. Several faulty scenarios were developed to test the monitoring framework after the base case or normal operating conditions of the PCA model were established. In all of the scenarios, the monitoring framework was able to detect the fault. Overall this study was successful at meeting the stated objective.

  5. 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.

  6. Assessment of the Portsmouth/Paducah Project Office Conduct of Operations Oversight of the Depleted Uranium Hexafluoride Conversion Plants, May 2012

    Office of Environmental Management (EM)

    Assessment of the Portsmouth/Paducah Project Office Conduct of Operations Oversight of the Depleted Uranium Hexafluoride Conversion Plants May 2012 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy i Table of Contents 1.0 Purpose ................................................................................................................................................... 1 2.0 Background

  7. 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.

  8. Conversion of hazardous materials using supercritical water oxidation

    DOE Patents [OSTI]

    Rofer, Cheryl K. (Los Alamos, NM); Buelow, Steven J. (Los Alamos, NM); Dyer, Richard B. (Los Alamos, NM); Wander, Joseph D. (Parker, FL)

    1992-01-01

    A process for destruction of hazardous materials in a medium of supercritical water without the addition of an oxidant material. The harzardous material is converted to simple compounds which are relatively benign or easily treatable to yield materials which can be discharged into the environment. Treatment agents may be added to the reactants in order to bind certain materials, such as chlorine, in the form of salts or to otherwise facilitate the destruction reactions.

  9. 2011 Final Report - Nano-Oxide Photocatalysis for Solar Energy Conversion

    Office of Scientific and Technical Information (OSTI)

    (Technical Report) | SciTech Connect Technical Report: 2011 Final Report - Nano-Oxide Photocatalysis for Solar Energy Conversion Citation Details In-Document Search Title: 2011 Final Report - Nano-Oxide Photocatalysis for Solar Energy Conversion We have very recently discovered a new hydrogen-producing photocatalyst is BiNbO4. BiNbO4 powders prepared by solid state reaction were tested for photocatalytic activity in methanol solutions under UV irradiation. When the material is tested without

  10. 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. The nitrate-dependent U(IV) oxidizing microbial population in groundwaters is less numerous ranging from 0 cells mL-1 (Well FW300, Uncontaminated Background NABIR FRC) to 4.3 x 102 cells mL-1 (Well TPB16, Contaminated Area 2 NABIR FRC). The presence of nitrate-dependent U(IV) oxidizing bacteria supports our hypothesis that bacteria capable of anaerobic U(IV) oxidation are ubiquitous and indigenous to sedimentary and groundwater environments.

  11. 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-calcium-carbonato complexes, and ferrihydrite on the rate and extent of uranium reduction in complex geochemical systems.

  12. 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.

  13. 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 accelerated the Li2O dissolution rate by more than a factor of six.

  14. 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)

  15. Novel Solar Energy Conversion Materials by Design of Mn(II) Oxides

    SciTech Connect (OSTI)

    Lany, S.; Peng, H.; Ndione, P.; Zakutayev, A.; Ginley, D. S.

    2013-01-01

    Solar energy conversion materials need to fulfill simultaneously a number of requirements in regard of their band-structure, optical properties, carrier transport, and doping. Despite their desirable chemical properties, e.g., for photo-electrocatalysis, transition-metal oxides usually do not have desirable semiconducting properties. Instead, oxides with open cation d-shells are typically Mott or charge-transfer insulators with notoriously poor transport properties, resulting from large effective electron/hole masses or from carrier self-trapping. Based on the notion that the electronic structure features (p-d interaction) supporting the p-type conductivity in d10 oxides like Cu2O and CuAlO2 occurs in a similar fashion also in the d5 (high-spin) oxides, we recently studied theoretically the band-structure and transport properties of the prototypical binary d5 oxides MnO and Fe2O3 [PRB 85, 201202(R)]. We found that MnO tends to self-trap holes by forming Mn+III, whereas Fe2O3 self-traps electrons by forming Fe+II. However, the self-trapping of holes is suppressed by when Mn is tetrahedrally coordinated, which suggests specific routes to design novel solar conversion materials by considering ternary Mn(II) oxides or oxide alloys. We are presenting theory, synthesis, and initial characterization for these novel energy materials.

  16. Safeguards Options for Natural Uranium Conversion Facilities ? A Collaborative Effort between the U.S. Department of Energy (DOE) and the National Nuclear Energy Commission of Brazil (CNEN)

    SciTech Connect (OSTI)

    Raffo-Caiado, Ana Claudia; Begovich, John M; Ferrada, Juan J

    2008-01-01

    In 2005, the National Nuclear Energy Commission of Brazil (CNEN) and the U.S. Department of Energy (DOE) agreed on a collaborative effort to evaluate measures that can strengthen the effectiveness of international safeguards at a natural uranium conversion plant (NUCP). The work was performed by DOE's Oak Ridge National Laboratory and CNEN. A generic model of an NUCP was developed and typical processing steps were defined. The study, completed in early 2007, identified potential safeguards measures and evaluated their effectiveness and impacts on operations. In addition, advanced instrumentation and techniques for verification purposes were identified and investigated. The scope of the work was framed by the International Atomic Energy Agency's (IAEA's) 2003 revised policy concerning the starting point of safeguards at uranium conversion facilities. Before this policy, only the final products of the uranium conversion plant were considered to be of composition and purity suitable for use in the nuclear fuel cycle and, therefore, subject to AEA safeguards control. DOE and CNEN have explored options for implementing the IAEA policy, although Brazil understands that the new policy established by the IAEA is beyond the framework of the Quadripartite Agreement of which it is one of the parties, together with Argentina, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials, and the IAEA. This paper highlights the findings of this joint collaborative effort and identifies technical measures to strengthen international safeguards in NUCPs.

  17. 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 complex processes (such as advective flow) present in the natural environment are unknown. The objective of the current studies was to address some of these short-comings in an attempt to develop this bioremediative strategy into a robust, field applicable technology. This objective was approached by both pure culture studies investigating the mechanism of Fe(II) oxidation by nitrate reducing bacteria and examining the flow dynamics and microbial processes in advective flow columns amended with Fe(II) and nitrate over an extended period.

  18. 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.

  19. 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.

  20. Compilation of Requirements for Safe Handling of Fluorine and Fluorine-Containing Products of Uranium Hexafluoride Conversion

    SciTech Connect (OSTI)

    Ferrada, J.J.

    2000-04-03

    Public Law (PL) 105-204 requires the U.S. Department of Energy to develop a plan for inclusion in the fiscal year 2000 budget for conversion of the Department's stockpile of depleted uranium hexafluoride (DUF{sub 6}) to a more stable form over an extended period. The conversion process into a more stable form will produce fluorine compounds (e.g., elemental fluorine or hydrofluoric acid) that need to be handled safely. This document compiles the requirements necessary to handle these materials within health and safety standards, which may apply in order to ensure protection of the environment and the safety and health of workers and the public. Fluorine is a pale-yellow gas with a pungent, irritating odor. It is the most reactive nonmetal and will react vigorously with most oxidizable substances at room temperature, frequently with ignition. Fluorine is a severe irritant of the eyes, mucous membranes, skin, and lungs. In humans, the inhalation of high concentrations causes laryngeal spasm and broncospasms, followed by the delayed onset of pulmonary edema. At sublethal levels, severe local irritation and laryngeal spasm will preclude voluntary exposure to high concentrations, unless the individual is trapped or incapacitated. A blast of fluorine gas on the shaved skin of a rabbit causes a second degree burn. Lower concentrations cause severe burns of insidious onset, resulting in ulceration, similar to the effects produced by hydrogen fluoride. Hydrofluoric acid is a colorless, fuming liquid or gas with a pungent odor. It is soluble in water with release of heat. Ingestion of an estimated 1.5 grams produced sudden death without gross pathological damage. Repeated ingestion of small amounts resulted in moderately advanced hardening of the bones. Contact of skin with anhydrous liquid produces severe burns. Inhalation of AHA or aqueous hydrofluoric acid mist or vapors can cause severe respiratory tract irritation that may be fatal. Based on the extreme chemical properties of these chemicals as noted above, fluorine or fluorine compounds must be handled appropriately within the boundaries of many safety requirements for the protection of the environment and the public. This report analyzes the safety requirements that regulatory agencies have issued to handle fluorine or fluorine compounds and lists them in Table 1. Table 1 lists the source of the requirements, the specific section of the source document, and a brief description of the requirements.

  1. Strategies to Suppress Cation Vacancies in Metal Oxide Alloys: Consequences for Solar Energy Conversion

    SciTech Connect (OSTI)

    Toroker, Maytal; Carter, Emily A.

    2015-09-01

    First-row transition metal oxides (TMOs) are promising alternative materials for inexpensive and efficient solar energy conversion. However, their conversion efficiency can be deleteriously affected by material imperfections, such as atomic vacancies. In this work, we provide examples showing that in some iron-containing TMOs, iron cation vacancy formation can be suppressed via alloying. We calculate within density functional theory+U theory the iron vacancy formation energy in binary rock-salt oxide alloys that contain iron, manganese, nickel, zinc, and/or magnesium. We demonstrate that formation of iron vacancies is less favorable if we choose to alloy iron(II) oxide with metals that cannot readily accept vacancy-generated holes, e.g., magnesium, manganese, nickel, or zinc. Since there are less available sites for holes and the holes are forced to reside on iron cations, the driving force for iron vacancy formation decreases. These results are consistent with an experiment observing a sharp drop in cation vacancy concentration upon alloying iron(II) oxide with manganese.

  2. 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.

  3. DOE Extends Contract to Operate Depleted Uranium Hexafluoride...

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

    Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants DOE Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants December 24, 2015 -...

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

    DOE Patents [OSTI]

    Poa, David S. (Naperville, IL); Burris, Leslie (Naperville, IL); Steunenberg, Robert K. (Naperville, IL); Tomczuk, Zygmunt (Orland Park, IL)

    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.

  5. 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.

  6. 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 ...

  7. 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.

  8. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? RETURN TO 100 MW

    SciTech Connect (OSTI)

    Smith, Kevin Arthur [ORNL; Primm, Trent [ORNL

    2009-01-01

    The feasibility of low-enriched uranium (LEU) fuel as a replacement for the current, high enriched uranium (HEU) fuel for the High Flux Isotope Reactor (HFIR) has been under study since 2006. Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting to LEU fuel requires returning the reactor power to 100 MW from 85 MW. The analyses required to up-rate the reactor power and the methods to perform these analyses are discussed. Comments regarding the regulatory approval process are provided along with a conceptual schedule.

  9. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    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.

  10. 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.

  11. Sodiation kinetics of metal oxide conversion electrodes: A comparative study with lithiation

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

    He, Kai; Lin, Feng; Zhu, Yizhou; Yu, Xiqian; Li, Jing; Lin, Ruoqian; Nordlund, Dennis; Weng, Tsu Chien; Richards, Ryan M.; Yang, Xiao -Qing; et al

    2015-08-19

    The development of sodium ion batteries (NIBs) can provide an alternative to lithium ion batteries (LIBs) for sustainable, low-cost energy storage. However, due to the larger size and higher m/e ratio of the sodium ion compared to lithium, sodiation reactions of candidate electrodes are expected to differ in significant ways from the corresponding lithium ones. In this work, we investigated the sodiation mechanism of a typical transition metal-oxide, NiO, through a set of correlated techniques, including electrochemical and synchrotron studies, real-time electron microscopy observation, and ab initio molecular dynamics (MD) simulations. We found that a crystalline Na₂O reaction layer thatmore » was formed at the beginning of sodiation plays an important role in blocking the further transport of sodium ions. In addition, sodiation in NiO exhibits a “shrinking-core” mode that results from a layer-by-layer reaction, as identified by ab initio MD simulations. For lithiation, however, the formation of Li anti-site defects significantly distorts the local NiO lattice that facilitates Li insertion, thus enhancing the overall reaction rate. These observations delineate the mechanistic difference between sodiation and lithiation in metal-oxide conversion materials. More importantly, our findings identify the importance of understanding the role of reaction layers on the functioning of electrodes and thus provide critical insights into further optimizing NIB materials through surface engineering.« less

  12. Sodiation kinetics of metal oxide conversion electrodes: A comparative study with lithiation

    SciTech Connect (OSTI)

    He, Kai; Lin, Feng; Zhu, Yizhou; Yu, Xiqian; Li, Jing; Lin, Ruoqian; Nordlund, Dennis; Weng, Tsu Chien; Richards, Ryan M.; Yang, Xiao -Qing; Doeff, Marca M.; Stach, Eric A.; Mo, Yifei; Xin, Huolin L.; Su, Dong

    2015-08-19

    The development of sodium ion batteries (NIBs) can provide an alternative to lithium ion batteries (LIBs) for sustainable, low-cost energy storage. However, due to the larger size and higher m/e ratio of the sodium ion compared to lithium, sodiation reactions of candidate electrodes are expected to differ in significant ways from the corresponding lithium ones. In this work, we investigated the sodiation mechanism of a typical transition metal-oxide, NiO, through a set of correlated techniques, including electrochemical and synchrotron studies, real-time electron microscopy observation, and ab initio molecular dynamics (MD) simulations. We found that a crystalline Na?O reaction layer that was formed at the beginning of sodiation plays an important role in blocking the further transport of sodium ions. In addition, sodiation in NiO exhibits a shrinking-core mode that results from a layer-by-layer reaction, as identified by ab initio MD simulations. For lithiation, however, the formation of Li anti-site defects significantly distorts the local NiO lattice that facilitates Li insertion, thus enhancing the overall reaction rate. These observations delineate the mechanistic difference between sodiation and lithiation in metal-oxide conversion materials. More importantly, our findings identify the importance of understanding the role of reaction layers on the functioning of electrodes and thus provide critical insights into further optimizing NIB materials through surface engineering.

  13. Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program

    SciTech Connect (OSTI)

    Nguyen Minh

    2006-07-31

    This report summarizes the work performed for Phase I (October 2001 - August 2006) under Cooperative Agreement DE-FC26-01NT41245 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled 'Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program'. The program focuses on the development of a low-cost, high-performance 3-to-10-kW solid oxide fuel cell (SOFC) system suitable for a broad spectrum of power-generation applications. During Phase I of the program significant progress has been made in the area of SOFC technology. A high-efficiency low-cost system was designed and supporting technology developed such as fuel processing, controls, thermal management, and power electronics. Phase I culminated in the successful demonstration of a prototype system that achieved a peak efficiency of 41%, a high-volume cost of $724/kW, a peak power of 5.4 kW, and a degradation rate of 1.8% per 500 hours. . An improved prototype system was designed, assembled, and delivered to DOE/NETL at the end of the program. This prototype achieved an extraordinary peak efficiency of 49.6%.

  14. Monte Carlo modeling and analyses of YALINA booster subcritical assembly, Part III : low enriched uranium conversion analyses.

    SciTech Connect (OSTI)

    Talamo, A.; Gohar, Y. (Nuclear Engineering Division) [Nuclear Engineering Division

    2011-05-12

    This study investigates the performance of the YALINA Booster subcritical assembly, located in Belarus, during operation with high (90%), medium (36%), and low (21%) enriched uranium fuels in the assembly's fast zone. The YALINA Booster is a zero-power, subcritical assembly driven by a conventional neutron generator. It was constructed for the purpose of investigating the static and dynamic neutronics properties of accelerator driven subcritical systems, and to serve as a fast neutron source for investigating the properties of nuclear reactions, in particular transmutation reactions involving minor-actinides. The first part of this study analyzes the assembly's performance with several fuel types. The MCNPX and MONK Monte Carlo codes were used to determine effective and source neutron multiplication factors, effective delayed neutron fraction, prompt neutron lifetime, neutron flux profiles and spectra, and neutron reaction rates produced from the use of three neutron sources: californium, deuterium-deuterium, and deuterium-tritium. In the latter two cases, the external neutron source operates in pulsed mode. The results discussed in the first part of this report show that the use of low enriched fuel in the fast zone of the assembly diminishes neutron multiplication. Therefore, the discussion in the second part of the report focuses on finding alternative fuel loading configurations that enhance neutron multiplication while using low enriched uranium fuel. It was found that arranging the interface absorber between the fast and the thermal zones in a circular rather than a square array is an effective method of operating the YALINA Booster subcritical assembly without downgrading neutron multiplication relative to the original value obtained with the use of the high enriched uranium fuels in the fast zone.

  15. 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

  16. 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.

  17. U.S. transparency monitoring of HEU oxide conversion and blending to LEU hexafluoride at three Russian blending plants

    SciTech Connect (OSTI)

    Leich, D., LLNL

    1998-07-27

    The down-blending of Russian highly enriched uranium (HEU) takes place at three Russian gaseous centrifuge enrichment plants. The fluorination of HEU oxide and down-blending of HEU hexafluoride began in 1994, and shipments of low enriched uranium (LEU) hexafluoride product to the United States Enrichment Corporation (USEC) began in 1995 US transparency monitoring under the HEU Purchase Agreement began in 1996 and includes a permanent monitoring presence US transparency monitoring at these facilities is intended to provide confidence that HEU is received and down-blended to LEU for shipment to USEC The monitoring begins with observation of the receipt of HEU oxide shipments, including confirmation of enrichment using US nondestructive assay equipment The feeding of HEU oxide to the fluorination process and the withdrawal of HEU hexafluoride are monitored Monitoring is also conducted where the blending takes place and where shipping cylinders are filled with LEU product. A series of process and material accountancy documents are provided to US monitors.

  18. A study of ZnxZryOz mixed oxides for direct conversion of ethanol to isobutene

    SciTech Connect (OSTI)

    Liu, Changjun; Sun, Junming; Smith, Colin; Wang, Yong

    2013-10-02

    ZnxZryOz mixed oxides were studied for direct conversion of ethanol to isobutene. Reaction conditions (temperature, residence time, ethanol molar fraction, steam to carbon ratio), catalyst composition, and pretreatment conditions were investigated, aiming at high-yield production of isobutene under industrially relevant conditions. An isobutene yield of 79% was achieved with an ethanol molar fraction of 8.3% at 475 C on fresh Zn1Zr8O17 catalysts. Further durability and regeneration tests revealed that the catalyst exhibited very slow deactivation via coking formation with isobutene yield maintained above 75% for more than 10 h time-on-stream. More importantly, the catalysts activity in terms of isobutene yield can be readily recovered after in situ calcination in air at 550 C for 2.5 h. XRD, TPO, IR analysis of adsorbed pyridine (IR-Py), and nitrogen sorption have been used to characterize the surface physical/chemical properties to correlate the structure and performance of the catalysts.

  19. 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.

  20. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (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 ...

  1. Compact reaction cell for homogenizing and down-blending highly enriched uranium metal

    DOE Patents [OSTI]

    McLean, W. II; Miller, P.E.; Horton, J.A.

    1995-05-02

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gases into the reaction chamber, the upper port allowing for the exit of gases from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gases into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell. 4 figs.

  2. Compact reaction cell for homogenizing and down-blanding highly enriched uranium metal

    DOE Patents [OSTI]

    McLean, II, William (Oakland, CA); Miller, Philip E. (Livermore, CA); Horton, James A. (Livermore, CA)

    1995-01-01

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gasses into the reaction chamber, the upper port allowing for the exit of gasses from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gasses into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell.

  3. 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.

  4. Integrated Biomass Gasification with Catalytic Partial Oxidation for Selective Tar Conversion

    SciTech Connect (OSTI)

    Zhang, Lingzhi; Wei, Wei; Manke, Jeff; Vazquez, Arturo; Thompson, Jeff; Thompson, Mark

    2011-05-28

    Biomass gasification is a flexible and efficient way of utilizing widely available domestic renewable resources. Syngas from biomass has the potential for biofuels production, which will enhance energy security and environmental benefits. Additionally, with the successful development of low Btu fuel engines (e.g. GE Jenbacher engines), syngas from biomass can be efficiently used for power/heat co-generation. However, biomass gasification has not been widely commercialized because of a number of technical/economic issues related to gasifier design and syngas cleanup. Biomass gasification, due to its scale limitation, cannot afford to use pure oxygen as the gasification agent that used in coal gasification. Because, it uses air instead of oxygen, the biomass gasification temperature is much lower than well-understood coal gasification. The low temperature leads to a lot of tar formation and the tar can gum up the downstream equipment. Thus, the biomass gasification tar removal is a critical technology challenge for all types of biomass gasifiers. This USDA/DOE funded program (award number: DE-FG36-O8GO18085) aims to develop an advanced catalytic tar conversion system that can economically and efficiently convert tar into useful light gases (such as syngas) for downstream fuel synthesis or power generation. This program has been executed by GE Global Research in Irvine, CA, in collaboration with Professor Lanny Schmidt's group at the University of Minnesota (UoMn). Biomass gasification produces a raw syngas stream containing H2, CO, CO2, H2O, CH4 and other hydrocarbons, tars, char, and ash. Tars are defined as organic compounds that are condensable at room temperature and are assumed to be largely aromatic. Downstream units in biomass gasification such as gas engine, turbine or fuel synthesis reactors require stringent control in syngas quality, especially tar content to avoid plugging (gum) of downstream equipment. Tar- and ash-free syngas streams are a critical requirement for commercial deployment of biomass-based power/heat co-generation and biofuels production. There are several commonly used syngas clean-up technologies: (1) Syngas cooling and water scrubbing has been commercially proven but efficiency is low and it is only effective at small scales. This route is accompanied with troublesome wastewater treatment. (2) The tar filtration method requires frequent filter replacement and solid residue treatment, leading to high operation and capital costs. (3) Thermal destruction typically operates at temperatures higher than 1000oC. It has slow kinetics and potential soot formation issues. The system is expensive and materials are not reliable at high temperatures. (4) In-bed cracking catalysts show rapid deactivation, with durability to be demonstrated. (5) External catalytic cracking or steam reforming has low thermal efficiency and is faced with problematic catalyst coking. Under this program, catalytic partial oxidation (CPO) is being evaluated for syngas tar clean-up in biomass gasification. The CPO reaction is exothermic, implying that no external heat is needed and the system is of high thermal efficiency. CPO is capable of processing large gas volume, indicating a very compact catalyst bed and a low reactor cost. Instead of traditional physical removal of tar, the CPO concept converts tar into useful light gases (eg. CO, H2, CH4). This eliminates waste treatment and disposal requirements. All those advantages make the CPO catalytic tar conversion system a viable solution for biomass gasification downstream gas clean-up. This program was conducted from October 1 2008 to February 28 2011 and divided into five major tasks. - Task A: Perform conceptual design and conduct preliminary system and economic analysis (Q1 2009 ~ Q2 2009) - Task B: Biomass gasification tests, product characterization, and CPO tar conversion catalyst preparation. This task will be conducted after completing process design and system economics analysis. Major milestones include identification of syngas cleaning requirements for proposed system

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

    DOE Patents [OSTI]

    Hayden, Jr., Howard W. (Oakridge, TN); Horton, James A. (Livermore, CA); Elliott, Guy R. B. (Los Alamos, NM)

    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.

  6. 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.

  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.; Maty, 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.

  8. 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

  9. 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 in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

  10. Floodplain/wetland assessment of the effects of construction and operation ofa depleted uranium hexafluoride conversion facility at the Paducah, Kentucky,site.

    SciTech Connect (OSTI)

    Van Lonkhuyzen, R.

    2005-09-09

    The U.S. Department of Energy (DOE) Depleted Uranium Hexafluoride (DUF{sub 6}) Management Program evaluated alternatives for managing its inventory of DUF{sub 6} and issued the ''Programmatic Environmental Impact Statement for Alternative Strategies for the Long-Term Management and Use of Depleted Uranium Hexafluoride'' (DUF{sub 6} PEIS) in April 1999 (DOE 1999). The DUF{sub 6} inventory is stored in cylinders at three DOE sites: Paducah, Kentucky; Portsmouth, Ohio; and East Tennessee Technology Park (ETTP), near Oak Ridge, Tennessee. In the Record of Decision for the DUF{sub 6} PEIS, DOE stated its decision to promptly convert the DUF{sub 6} inventory to a more stable chemical form. Subsequently, the U.S. Congress passed, and the President signed, the ''2002 Supplemental Appropriations Act for Further Recovery from and Response to Terrorist Attacks on the United States'' (Public Law No. 107-206). This law stipulated in part that, within 30 days of enactment, DOE must award a contract for the design, construction, and operation of a DUF{sub 6} conversion plant at the Department's Paducah, Kentucky, and Portsmouth, Ohio, sites, and for the shipment of DUF{sub 6} cylinders stored at ETTP to the Portsmouth site for conversion. This floodplain/wetland assessment has been prepared by DOE, pursuant to Executive Order 11988 (''Floodplain Management''), Executive Order 11990 (Protection of Wetlands), and DOE regulations for implementing these Executive Orders as set forth in Title 10, Part 1022, of the ''Code of Federal Regulations'' (10 CFR Part 1022 [''Compliance with Floodplain and Wetland Environmental Review Requirements'']), to evaluate potential impacts to floodplains and wetlands from the construction and operation of a conversion facility at the DOE Paducah site. Reconstruction of the bridge crossing Bayou Creek would occur within the Bayou Creek 100-year floodplain. Replacement of bridge components, including the bridge supports, however, would not be expected to result in measurable long-term changes to the floodplain. Approximately 0.16 acre (0.064 ha) of palustrine emergent wetlands would likely be eliminated by direct placement of fill material within Location A. Some wetlands that are not filled may be indirectly affected by an altered hydrologic regime, due to the proximity of construction, possibly resulting in a decreased frequency or duration of inundation or soil saturation and potential loss of hydrology necessary to sustain wetland conditions. Indirect impacts could be minimized by maintaining a buffer near adjacent wetlands. Wetlands would likely be impacted by construction at Location B; however, placement of a facility in the northern portion of this location would minimize wetland impacts. Construction at Location C could potentially result in impacts to wetlands, however placement of a facility in the southeastern portion of this location may best avoid direct impacts to wetlands. The hydrologic characteristics of nearby wetlands could be indirectly affected by adjacent construction. Executive Order 11990, ''Protection of Wetlands'', requires federal agencies to minimize the destruction, loss, or degradation of wetlands, and to preserve and enhance the natural and beneficial uses of wetlands. DOE regulations for implementing Executive Order 11990 as well as Executive Order 11988, ''Floodplain Management'', are set forth in 10 CFR Part 1022. Mitigation for unavoidable impacts may be developed in coordination with the appropriate regulatory agencies. Unavoidable impacts to wetlands that are within the jurisdiction of the USACE may require a CWA Section 404 Permit, which would trigger the requirement for a CWA Section 401 Water Quality Certification from the Commonwealth of Kentucky. A mitigation plan may be required prior to the initiation of construction. Cumulative impacts to floodplains and wetlands are anticipated to be negligible to minor under the proposed action, in conjunction with the effects of existing conditions and other activities. Habitat disturbance would involve settings commonly found i

  11. 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.

  12. Uranium dioxide electrolysis

    DOE Patents [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.

  13. Assessment of Preferred Depleted Uranium Disposal Forms

    SciTech Connect (OSTI)

    Croff, A.G.; Hightower, J.R.; Lee, D.W.; Michaels, G.E.; Ranek, N.L.; Trabalka, J.R.

    2000-06-01

    The Department of Energy (DOE) is in the process of converting about 700,000 metric tons (MT) of depleted uranium hexafluoride (DUF6) containing 475,000 MT of depleted uranium (DU) to a stable form more suitable for long-term storage or disposal. Potential conversion forms include the tetrafluoride (DUF4), oxide (DUO2 or DU3O8), or metal. If worthwhile beneficial uses cannot be found for the DU product form, it will be sent to an appropriate site for disposal. The DU products are considered to be low-level waste (LLW) under both DOE orders and Nuclear Regulatory Commission (NRC) regulations. The objective of this study was to assess the acceptability of the potential DU conversion products at potential LLW disposal sites to provide a basis for DOE decisions on the preferred DU product form and a path forward that will ensure reliable and efficient disposal.

  14. 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, reductant or complexant used for selectivity, and ionic liquids used as supportive media. To complete the extraction and recovery cycle, we then demonstrate uranium back extraction from the TBP loaded sc-CO2 phase into an aqueous phase and the characterization of the uranium complex formed at the end of this process. Another aspect of this project was to limit proliferation risks by either co-extracting uranium and plutonium, or by leaving plutonium behind by selectively extracting uranium. We report that the former is easily achieved, since plutonium is in the tetravalent or hexavalent oxidation state in the oxidizing environment created by the TBP-nitric acid complex, and is therefore co-extracted. The latter is more challenging, as a reductant or complexant to plutonium has to be used to selectively extract uranium. After undertaking experiments on different reducing or complexing systems (e.g., AcetoHydroxamic Acid (AHA), Fe(II), ascorbic acid), oxalic acid was chosen as it can complex tetravalent actinides (Pu, Np, Th) in the aqueous phase while allowing the extraction of hexavalent uranium in the sc-CO2 phase. Finally, we show results using an alternative media to commonly used aqueous phases: ionic liquids. We show the dissolution of uranium in ionic liquids and its extraction using sc-CO2 with and without the presence of AHA. The possible separation of trivalent actinides from uranium is also demonstrated in ionic liquids using neodymium as a surrogate and diglycolamides as the extractant.

  15. Selective aerobic alcohol oxidation method for conversion of lignin into simple aromatic compounds

    DOE Patents [OSTI]

    Stahl, Shannon S; Rahimi, Alireza

    2015-03-03

    Described is a method to oxidize lignin or lignin sub-units. The method includes oxidation of secondary benzylic alcohol in the lignin or lignin sub-unit to a corresponding ketone in the presence of unprotected primarily aliphatic alcohol in the lignin or lignin sub-unit. The optimal catalyst system consists of HNO.sub.3 in combination with another Bronsted acid, in the absence of a metal-containing catalyst, thereby yielding a selectively oxidized lignin or lignin sub-unit. The method may be carried out in the presence or absence of additional reagents including TEMPO and TEMPO derivatives.

  16. 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.

  17. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    SciTech Connect (OSTI)

    Chandler, David; Freels, James D; Ilas, Germina; Miller, James Henry; Primm, Trent; Sease, John D; Guida, Tracey; Jolly, Brian C

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  18. 2011 Final Report - Nano-Oxide Photocatalysis for Solar Energy Conversion

    SciTech Connect (OSTI)

    Eckstein, James N.; Suslick, Kenneth S.

    2011-10-19

    We have very recently discovered a new hydrogen-producing photocatalyst is BiNbO4. BiNbO4 powders prepared by solid state reaction were tested for photocatalytic activity in methanol solutions under UV irradiation. When the material is tested without the presence of a Pt co-catalyst, photocatalytic activity for H2 evolution is superior to that of TiO2. It was also found that BiNbO4 photodegrades into metallic Bi and reduced Nb oxides after use; materials were characterized by SEM, XRD, and XPS. Adding Pt to the surface of the photocatalyst increases photocatalytic activity and importantly, helps to prevent photodegradation of the oxide material. With 1 wt. % Pt loading, photodegradation is essentially absent. BiNbO4 photodegrades into metallic Bi and reduced Nb oxides after use; materials were characterized by SEM, XRD, and XPS. Adding Pt to the surface of the photocatalyst increases photocatalytic activity and importantly, helps to prevent photodegradation of the oxide material. With 1 wt. % Pt loading, photodegradation is essentially absent.

  19. 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.

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

    DOE Patents [OSTI]

    Kreuzmann, Alvin B. (Cincinnati, OH)

    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.

  1. Catalytic hydrodechlorination of chlorocarbons. 2: Ternary oxide supports for catalytic conversions of 1,2-dichlorobenzene

    SciTech Connect (OSTI)

    Gampine, A.; Eyman, D.P.

    1998-10-01

    Ternary oxides of Ti-Zr-Al and Ti-Zr-Si were prepared by coating commercial Al{sub 2}O{sub 3} and SiO{sub 2} with a THF solution of Ti(OPr{sup i}){sub 4} and Zr(OPr{sup 1}){sub 4} under controlled conditions. Nitrogen adsorption and X-ray powder diffraction indicate that the structure of the base supports, Al{sub 2}O{sub 3} and SiO{sub 2}, were not significantly altered upon coating and that TiO{sub 2} and ZrO{sub 2} were quite uniformly spread on them. The acid resistance of alumina was found to be increased upon coating. Palladium supported catalysts, Pd/TiZrAlO{sub x}, Pd/TiZrSiO{sub x}, Pd/TiO{sub 2}, Pd/ZrO{sub 2}, Pd/SiO{sub 2}, and Pd/Al{sub 2}O{sub 3} were prepared to evaluate the ternary oxides relative to the component single oxide supports. Palladium dispersion was determined using hydrogen chemisorption and the catalysts were evaluated for hydrodechlorination of 1,2-dichlorobenzene. The experimental runs were carried out in a microflow reactor system at atmospheric pressure, in the gas phase. The catalysts were oxidized and then reduced, prior to reaction. The kinetic studies showed that the ternary oxide-based catalyst, Pd/TiZrAlO{sub x} exhibited an improved stability and activity much higher than the arithmetic sum of the activities of the component single oxide based palladium catalysts. Comparison of the specific activities of the catalysts expressed as TOF, indicate that the observed differences in activity may be related to the chemical nature of the supports. The best catalyst had an initial specific activity of 16.6 s{sup {minus}1}. The authors observed that the pretreatment of the catalyst has a profound effect on its stability and activity. Also, the experimental results indicated that the major factors of the catalyst deactivation are agglomeration of palladium particles and HCl poisoning. Prospects for optimization of these catalysts are discussed in light of the results of this work.

  2. 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.

  3. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    SciTech Connect (OSTI)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace several obsolete components of the current analytical tool set used for ATR neutronics support. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

  4. 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.

  5. Revisiting Photoemission and Inverse Photoemission Spectra of Nickel Oxide from First Principles: Implications for Solar Energy Conversion

    SciTech Connect (OSTI)

    Alidoust, Nima; Toroker, Maytal; Carter, Emily A.

    2014-07-17

    We use two different ab initio quantum mechanics methods, complete active space self-consistent field theory applied to electrostatically embedded clusters and periodic many-body G?W? calculations, to reanalyze the states formed in nickel(II) oxide upon electron addition and ionization. In agreement with interpretations of earlier measurements, we find that the valence and conduction band edges consist of oxygen and nickel states, respectively. However, contrary to conventional wisdom, we find that the oxygen states of the valence band edge are localized whereas the nickel states at the conduction band edge are delocalized. We argue that these characteristics may lead to low electron-hole recombination and relatively efficient electron transport, which, coupled with band gap engineering, could produce higher solar energy conversion efficiency compared to that of other transition-metal oxides. Both methods find a photoemission/inverse-photoemission gap of 3.6-3.9 eV, in good agreement with the experimental range, lending credence to our analysis of the electronic structure of NiO.

  6. Enterprise Assessments Targeted Review of the Paducah Depleted Uranium

    Office of Environmental Management (EM)

    Hexafluoride Conversion Facility Fire Protection Program - September 2015 | Department of Energy Review of the Paducah Depleted Uranium Hexafluoride Conversion Facility Fire Protection Program - September 2015 Enterprise Assessments Targeted Review of the Paducah Depleted Uranium Hexafluoride Conversion Facility Fire Protection Program - September 2015 September 2015 Targeted Review of the Fire Protection Program at the Paducah Depleted Uranium Hexafluoride Conversion Facility The U.S.

  7. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    SciTech Connect (OSTI)

    Cook, David Howard; Freels, James D; Ilas, Germina; Jolly, Brian C; Miller, James Henry; Primm, Trent; Renfro, David G; Sease, John D; Pinkston, Daniel

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  8. 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.

  9. 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.

  10. Report on the Effect the Low Enriched Uranium Delivered Under the Highly

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

    Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Ops of the Gaseous Diffusion | Department of Energy on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Ops of the Gaseous Diffusion Report on the Effect the Low Enriched Uranium

  11. 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.

  12. Conversion of Worcester Polytechnic Institute Reactor to low enriched uranium (LEU) fuel: Technical progress report for period August 15, 1987-February 15, 1988

    SciTech Connect (OSTI)

    Newton, T.H. Jr.

    1988-02-01

    An HEU fuel element was removed from the WPI core and tested in a Babcock-Wilcox 6M shipping container on August 27, 1987, for radiation level adequacy in shipping. Levels were found to be adequate so that use of the 6M container can be made in shipping the HEU fuel after a few weeks of decay time. A final submittal of the SAR technical specification changes relating to the fuel conversion was made on September 17, 1987. Questions regarding this submittal were received on January 25, 1988, and responses to these questions were made on February 10, 1988.

  13. Corrosion-resistant uranium

    DOE Patents [OSTI]

    Hovis, Jr., Victor M.; Pullen, William C.; Kollie, Thomas G.; Bell, Richard T.

    1983-01-01

    The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

  14. Corrosion-resistant uranium

    DOE Patents [OSTI]

    Hovis, V.M. Jr.; Pullen, W.C.; Kollie, T.G.; Bell, R.T.

    1981-10-21

    The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

  15. Uranium Marketing Annual Report - Release Date: May 31, 2011

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

    and does not include the conversion service and enrichment service components. Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual Survey" (20...

  16. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  17. Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

    SciTech Connect (OSTI)

    S. D. Herrmann; S. X. Li

    2010-09-01

    A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

  18. Highly Enriched Uranium Transparency Program | National Nuclear...

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

    Program reduces nuclear risk by monitoring the conversion of 500 metric tons (MT) of Russian HEU, enough material for 20,000 nuclear weapons, into low enriched uranium (LEU). ...

  19. Inherently safe in situ uranium recovery

    DOE Patents [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.

  20. Treatment of Uranium and Plutonium Solutions Generated in the Atalante Facility, France - 12004

    SciTech Connect (OSTI)

    Lagrave, Herve

    2012-07-01

    The Atalante complex operated by the French Alternative Energies and Atomic Energy Commission (CEA) at the Rhone Valley Research Center consolidates research programs on actinide chemistry, especially separation chemistry, processing for recycling spent fuel, and fabrication of actinide targets for innovative concepts in future nuclear systems. The design of future systems (Generation IV reactors, material recycling) will increase the uranium and plutonium flows in the facility, making it important to anticipate the stepped-up activity and provide Atalante with equipment dedicated to processing these solutions to obtain a mixed uranium-plutonium oxide that will be stored pending reuse. Ongoing studies for integral recycling of the actinides have highlighted the need for reserving equipment to produce actinides mixed oxide powder and also minor actinides bearing oxide for R and D purpose. To meet this double objective a new shielded line should be built in the facility and should be operational 6 years after go decision. The main functions of the new unit would be to receive, concentrate and store solutions, purify them, ensure group conversion of actinides and conversion of excess uranium. This new unit will be constructed in a completely refurbished building devoted to subcritical and safe geometry of the process equipments. (author)

  1. Artificial layered perovskite oxides A(B{sub 0.5}B?{sub 0.5})O{sub 3} as potential solar energy conversion materials

    SciTech Connect (OSTI)

    Chen, Hungru; Umezawa, Naoto

    2015-02-07

    Perovskite oxides with a d{sup 0} electronic configuration are promising photocatalysts and exhibit high electron mobilities. However, their band gaps are too large for efficient solar energy conversion. On the other hand, transition metal cations with partially filled d{sup n} electronic configurations give rise to visible light absorption. In this study, by using hybrid density functional theory calculations, it is demonstrated that the virtues of the two categories of materials can be combined in perovskite oxide A(B{sub 0.5}B?{sub 0.5})O{sub 3} with a layered B-site ordering along the [001] direction. The electronic structures of the four selected perovskite oxide compounds, La(Ti{sub 0.5}Ni{sub 0.5})O{sub 3}, La(Ti{sub 0.5}Zn{sub 0.5})O{sub 3}, Sr(Nb{sub 0.5}Cr{sub 0.5})O{sub 3}, and Sr(Nb{sub 0.5}Fe{sub 0.5})O{sub 3} are calculated and discussed.

  2. 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.

  3. 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.

  4. Uranium Mining, Conversion, and Enrichment Industries

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

    ... Thus, increased demand in China and Russia will consume capacity with which URENCO would ... recent years has been from the Asian markets, specifically Japan, Taiwan, and South Korea. ...

  5. EIS-0359: DOE Notice of Availability of the Draft Supplement Analysis

    Broader source: Energy.gov [DOE]

    Disposal of Depleted Uranium Oxide Conversion Produce Generated from DOE's Inventory of Depleted Uranium Hexafluoride

  6. 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 have measured elevated U concentrations in oxic iron-rich sediments. To translate experimental results into numerical analysis of U fate and transport, a reaction network was developed based on Sani et al. (2004) to simulate U(VI) bioreduction with concomitant UO2 reoxidation in the presence of hematite or ferrihydrite. The reduction phase considers SRB reduction (using lactate) with the reductive dissolution of Fe(III) solids, which is set to be microbially mediated as well as abiotically driven by sulfide. Model results show the oxidation of HS by Fe(III) directly competes with UO2 reoxidation as Fe(III) oxidizes HS preferentially over UO2. The majority of Fe reduction is predicted to be abiotic, with ferrihydrite becoming fully consumed by reaction with sulfide. Predicted total dissolved carbonate concentrations from the degradation of lactate are elevated (log(pCO2) ~ 1) and, in the hematite system, yield close to two orders-of-magnitude higher U(VI) concentrations than under initial carbonate concentrations of 3 mM. Modeling of U(VI) bioreduction with concomitant reoxidation of UO2 in the presence of ferrihydrite was also extended to a two-dimensional field-scale groundwater flow and biogeochemically reactive transport model for the South Oyster site in eastern Virginia. This model was developed to simulate the field-scale immobilization and subsequent reoxidation of U by a biologically mediated reaction network.

  7. Formation of alcohol conversion catalysts

    DOE Patents [OSTI]

    Wachs, Israel E. (Bridgewater, NJ); Cai, Yeping (Louisville, KY)

    2001-01-01

    The method of the present invention involves a composition containing an intimate mixture of (a) metal oxide support particles and (b) a catalytically active metal oxide from Groups VA, VIA, or VIIA, its method of manufacture, and its method of use for converting alcohols to aldehydes. During the conversion process, catalytically active metal oxide from the discrete catalytic metal oxide particles migrates to the oxide support particles and forms a monolayer of catalytically active metal oxide on the oxide support particle to form a catalyst composition having a higher specific activity than the admixed particle composition.

  8. Disposition of DOE Excess Depleted Uranium, Natural Uranium, and

    Office of Environmental Management (EM)

    Low-Enriched Uranium | Department of Energy Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium 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 depleted uranium hexafluoride (DUF6), natural uranium hexafluoride (NUF6), and

  9. 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.

  10. 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.

  11. Alcohol conversion

    DOE Patents [OSTI]

    Wachs, Israel E. (Bridgewater, NJ); Cai, Yeping (Louisville, KY)

    2002-01-01

    Preparing an aldehyde from an alcohol by contacting the alcohol in the presence of oxygen with a catalyst prepared by contacting an intimate mixture containing metal oxide support particles and particles of a catalytically active metal oxide from Groups VA, VIA, or VIIA, with a gaseous stream containing an alcohol to cause metal oxide from the discrete catalytically active metal oxide particles to migrate to the metal oxide support particles and to form a monolayer of catalytically active metal oxide on said metal oxide support particles.

  12. The National Conversion Pilot Project

    SciTech Connect (OSTI)

    Roberts, A.V.

    1995-12-31

    The National Conversion Pilot Project (NCPP) is a recycling project under way at the U.S. Department of Energy (DOE) Rocky Flats Environmental Technology Site (RFETS) in Colorado. The recycling aim of the project is threefold: to reuse existing nuclear weapon component production facilities for the production of commercially marketable products, to reuse existing material (uranium, beryllium, and radioactively contaminated scrap metals) for the production of these products, and to reemploy former Rocky Flats workers in this process.

  13. Thermochemical Conversion

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

    Thermochemical Conversion - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs

  14. Pyroprocessing of Oxidized Sodium-Bonded Fast Reactor Fuel -- an Experimental Study of Treatment Options for Degraded EBR-II Fuel

    SciTech Connect (OSTI)

    S. D. Herrmann; L. A. Wurth; N. J. Gese

    2013-09-01

    An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electrometallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li2O at 650 C with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. The experimental study illustrated how zirconium oxide and sodium oxide present different challenges to a lithium-based electrolytic reduction system for conversion of select metal oxides to metal.

  15. 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.

  16. Performance and Fabrication Status of TREAT LEU Conversion Conceptual Design Concepts

    SciTech Connect (OSTI)

    IJ van Rooyen; SR Morrell; AE Wright; E. P Luther; K Jamison; AL Crawford; HT III Hartman

    2014-10-01

    Resumption of transient testing at the TREAT facility was approved in February 2014 to meet U.S. Department of Energy (DOE) objectives. The National Nuclear Security Administrations Global Threat Reduction Initiative Convert Program is evaluating conversion of TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU). This paper describes briefly the initial pre-conceptual designs screening decisions with more detailed discussions on current feasibility, qualification and fabrication approaches. Feasible fabrication will be shown for a LEU fuel element assembly that can meet TREAT design, performance, and safety requirements. The statement of feasibility recognizes that further development, analysis, and testing must be completed to refine the conceptual design. Engineering challenges such as cladding oxidation, high temperature material properties, and fuel block fabrication along with neutronics performance, will be highlighted. Preliminary engineering and supply chain evaluation provided confidence that the conceptual designs can be achieved.

  17. 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.

  18. Microsoft Word - L15 01-22 Uranium Tranfers

    Office of Environmental Management (EM)

    To: Office of Nuclear Energy Department of Energy 1000 Independence Ave., SW Washington, DC 20585 From: Nan Swift Federal Affairs Manager National Taxpayers Union 108 N. Alfred Street Alexandria, VA 22314 Subject: Request for Information: Excess Uranium Management: Effects of DOE Transfers of Excess Uranium on Domestic Uranium Mining, Conversion, and Enrichment Industries To whom it may concern: On behalf of the members of the National Taxpayers Union (NTU), I write to express our concerns

  19. 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.

  20. Oxide

    Energy Science and Technology Software Center (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.

  1. Depleted Uranium Hexafluoride (DUF6) Fully Operational at the Portsmouth

    Office of Environmental Management (EM)

    and Paducah Gaseous Diffusion Sites | Department of Energy Depleted Uranium Hexafluoride (DUF6) Fully Operational at the Portsmouth and Paducah Gaseous Diffusion Sites Depleted Uranium Hexafluoride (DUF6) Fully Operational at the Portsmouth and Paducah Gaseous Diffusion Sites October 20, 2011 - 9:16am Addthis When Babcock & Wilcox Conversion Services took over the DUF6 Project on March 29 of this year, the company had one thing in mind: Bring all seven conversion lines at both plants to

  2. 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)

  3. Recovery and Blend-Down Uranium for Beneficial use in Commercial Reactors - 13373

    SciTech Connect (OSTI)

    Magoulas, Virginia [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)] [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

    In April 2001 the Department of Energy (DOE) and the Tennessee Valley Authority (TVA) signed an Interagency Agreement to transfer approximately 33 MT of off-specification (off-spec) highly enriched uranium (HEU) from DOE to TVA for conversion to commercial reactor fuel. Since that time additional surplus off-spec HEU material has been added to the program, making the total approximately 46 MT off-spec HEU. The disposition path for approximately half (23 MT) of this 46 MT of surplus HEU material, was down blending through the H-canyon facility at the Savannah River Site (SRS). The HEU is purified through the H-canyon processes, and then blended with natural uranium (NU) to form low enriched uranium (LEU) solution with a 4.95% U-235 isotopic content. This material was then transported to a TVA subcontractor who converted the solution to uranium oxide and then fabricated into commercial light water reactor (LWR) fuel. This fuel is now powering TVA reactors and supplying electricity to approximately 1 million households in the TVA region. There is still in excess of approximately 10 to 14 MT of off-spec HEU throughout the DOE complex or future foreign and domestic research reactor returns that could be recovered and down blended for use in either currently designed light water reactors, ?5% enriched LEU, or be made available for use in subsequent advanced 'fast' reactor fuel designs, ?19% LEU. (authors)

  4. 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.

  5. 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.

  6. Process for recovering uranium from waste hydrocarbon oils containing the same. [Uranium contaminated lubricating oils from gaseous diffusion compressors

    DOE Patents [OSTI]

    Conrad, M.C.; Getz, P.A.; Hickman, J.E.; Payne, L.D.

    1982-06-29

    The invention is a process for the recovery of uranium from uranium-bearing hydrocarbon oils containing carboxylic acid as a degradation product. In one aspect, the invention comprises providing an emulsion of water and the oil, heating the same to a temperature effecting conversion of the emulsion to an organic phase and to an acidic aqueous phase containing uranium carboxylate, and recovering the uranium from the aqueous phase. The process is effective, simple and comparatively inexpensive. It avoids the use of toxic reagents and the formation of undesirable intermediates.

  7. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    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.

  8. Correlation of radioactive-waste-treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: conversion of yellow cake to uranium hexafluoride. Part II. The solvent extraction-fluorination process

    SciTech Connect (OSTI)

    Sears, M.B.; Etnier, E.L.; Hill, G.S.; Patton, B.D.; Witherspoon, J.P.; Yen, S.N.

    1983-03-01

    A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials and chemicals from a model uranium hexafluoride (UF/sub 6/) production plant using the solvent extraction-fluorination process, and to evaluate the radiological impact (dose commitment) of the release materials on the environment. The model plant processes 10,000 metric tons of uranium per year. Base-case waste treatment is the minimum necessary to operate the process. Effluents meet the radiological requirements listed in the Code of Federal Regulations, Title 10, Part 20 (10 CFR 20), Appendix B, Table II, but may not be acceptable chemically at all sites. Additional radwaste treatment techniques are applied to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The costs for the added waste treatment operations and the corresponding dose committment are correlated with the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases will require development and demonstration, or else is proprietary and unavailable for immediate use. The methodology and assumptions for the radiological doses are found in ORNL-4992.

  9. 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.

  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. 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.

  12. 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.

  13. 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.

  14. Safeguards on uranium ore concentrate? the impact of modern mining and milling process

    SciTech Connect (OSTI)

    Francis, Stephen

    2013-07-01

    Increased purity in uranium ore concentrate not only raises the question as to whether Safeguards should be applied to the entirety of uranium conversion facilities, but also as to whether some degree of coverage should be moved back to uranium ore concentrate production at uranium mining and milling facilities. This paper looks at uranium ore concentrate production across the globe and explores the extent to which increased purity is evident and the underlying reasons. Potential issues this increase in purity raises for IAEA's strategy on the Starting Point of Safeguards are also discussed.

  15. 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.

  16. Uranium Marketing Annual Report -

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

    3. Inventories of uranium by owner as of end of year, 2010-14 thousand pounds U3O8 equivalent Inventories at the end of the year Owner of uranium inventory 2010 2011 2012 2013...

  17. Uranium Marketing Annual Report -

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

    2. Inventories of natural and enriched uranium by material type as of end of year, 2010-14 thousand pounds U3O8 equivalent Inventories at the end of the year Type of uranium...

  18. 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 of twin boundaries formed during the martensitic transformation. Deformation actually results in a coarsening of the microstructure making EBSD more practical following a limited amount of strain. Figure 2 shows the microstructure resulting from 6% compression. Casting of U-10Mo results in considerable chemical segregation as is apparent in Figure 2a. The segregation subsists through rolling and heat treatment processes as shown in Figure 2b. EBSD was used to study the effects of homogenization time and temperature on chemical heterogeneity. It was found that times and temperatures that result in a chemically homogeneous microstructure also result in a significant increase in grain size. U-0.75Ti forms an acicular martinsite as shown in Figure 4. This microstructure prevails through cycling into the higher temperature solid uranium phases.

  19. 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.

  20. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    4. Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2012-14 2012 2013 2014 Advance Uranium Asset Management Ltd. (was Uranium Asset Management) American Fuel Resources, LLC Advance Uranium Asset Management Ltd. American Fuel Resources, LLC AREVA NC, Inc. AREVA / AREVA NC, Inc. AREVA NC, Inc. BHP Billiton Olympic Dam Corporation Pty Ltd ARMZ (AtomRedMetZoloto) BHP Billiton Olympic Dam Corporation Pty Ltd CAMECO BHP Billiton Olympic Dam Corporation Pty Ltd CAMECO

  1. 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.

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

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

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

  3. Table 9.3 Uranium Overview, 1949-2011

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

    3 Uranium Overview, 1949-2011 Year Domestic Concentrate Production 1 Purchased Imports 2 Export 2 Sales Electric Plant Purchases From Domestic Suppliers Loaded Into U.S. Nuclear Reactors 3 Inventories Average Price Domestic Suppliers Electric Plants Total Purchased Imports Domestic Purchases Million Pounds Uranium Oxide Dollars 4 per Pound Uranium Oxide 1949 0.36 4.3 0.0 NA NA NA NA NA NA NA 1950 .92 5.5 .0 NA NA NA NA NA NA NA 1951 1.54 6.1 .0 NA NA NA NA NA NA NA 1952 1.74 5.7 .0 NA NA NA NA

  4. Symposium on the Physical Chemistry of Solar Energy Conversion...

    Office of Scientific and Technical Information (OSTI)

    session): (1) Quantum Dots and Nanorods for Solar Energy Conversion (2 half-day sessions); (2) Artificial Photosynthesis: Water Oxidation; (3) Artificial Photosynthesis: Solar ...

  5. 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.

  6. Conversion and enrichment in the Soviet Union

    SciTech Connect (OSTI)

    1991-04-01

    In the Soviet Union, just as in the West, the civilian nuclear industry emerged from research work undertaken for nuclear weapons development. At first, researchers tried various techniques for physical separation of uranium isotopes: electromagnetic and molecular-kinetic thermo-diffusion methods; gaseous diffusion; and centrifuge methods. All of those methods, which are based primarily on differences in the atomic mass of uranium isotopes, called for extensive research and the development of new, technically unprecedented equipment. Gradually gaseous diffusion and gas centrifuge technology became recognized as most feasible for industrial use, so research on other methods was terminated. Industrial-scale uranium enrichment in the Soviet Union began in 1949 using the gaseous diffusion method; by the early 1960s, centrifuge technology was in use on an industrial scale. All Soviet production of highly-enriched, weapons-grade uranium was halted in 1987. The Soviet Union now has four enrichment plants in operation (at classified locations), solely for civilian nuclear power needs. All four enrichment plants have centrifuge modules, and enrichment provided by gaseous diffusion accounts for less than 5% of their total output. Two of the four enrichment plants also incorporate facilities for conversion to uranium hexafluoride (UF{sub 6}).

  7. Depleted uranium storage and disposal trade study: Summary report

    SciTech Connect (OSTI)

    Hightower, J.R.; Trabalka, J.R.

    2000-02-01

    The objectives of this study were to: identify the most desirable forms for conversion of depleted uranium hexafluoride (DUF6) for extended storage, identify the most desirable forms for conversion of DUF6 for disposal, evaluate the comparative costs for extended storage or disposal of the various forms, review benefits of the proposed plasma conversion process, estimate simplified life-cycle costs (LCCs) for five scenarios that entail either disposal or beneficial reuse, and determine whether an overall optimal form for conversion of DUF6 can be selected given current uncertainty about the endpoints (specific disposal site/technology or reuse options).

  8. EIS-0360: DOE Notice of Availability of the Draft Supplement Analysis

    Broader source: Energy.gov [DOE]

    Disposal of Depleted Uranium Oxide Conversion Produce Generated from DOE's Inventory of Depleted Uranium Hexafluoride, Clive, Utah

  9. 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.

  10. About the Uranium Mine Team | Department of Energy

    Energy Savers [EERE]

    Uranium Mine Team About the Uranium Mine Team Text coming

  11. 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.

  12. 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.

  13. 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.

  14. 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, 1994-2014 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

  15. 2014 Uranium Marketing Annual Report

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

    Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Table S3a. Foreign purchases, foreign sales, and uranium ...

  16. 2014 Uranium Marketing Annual Report

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

    5 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Table S2. Uranium feed deliveries, enrichment services, and uranium loaded by owners ...

  17. 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.

  18. 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...

  19. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    Note: Totals may not equal sum of components because of independent rounding. Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual Survey" (20...

  20. Uranium Marketing Annual Report -

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

    Note: Totals may not equal sum of components because of independent rounding. Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual Survey" (20...

  1. Uranium Marketing Annual Report -

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

    rounding. Weighted-average prices are not adjusted for inflation. Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual Survey" (2010-14)....

  2. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    of the United States. Weighted-average prices are not adjusted for inflation. Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual Survey" (2010...

  3. Uranium Marketing Annual Report -

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

    Totals may not equal sum of components because of independent rounding. Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual Survey" (2011...

  4. Uranium Marketing Annual Report -

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

    independent rounding. Weighted-average prices are not adjusted for inflation. Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual Survey" (2013...

  5. Uranium Marketing Annual Report -

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

    independent rounding. Weighted-average prices are not adjusted for inflation. Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual Survey" (2010-...

  6. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    Note: Totals may not equal sum of components because of independent rounding. Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual Survey" (2013...

  7. 2015 Solid Oxide Fuel Cells Project Portfolio

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

    2015 Solid Oxide Fuel Cells Project Portfolio Solid Oxide Fuel Cells are energy conversion devices that produce electric power through an electrochemical reaction rather than by...

  8. final ERI-2142 18-1501 Analysis of Potential Effects on Domestic Industries of DOE Excess Uranium Inventory 2015-2024.docx

    Broader source: Energy.gov (indexed) [DOE]

    ERI-2142.18-1501 Analysis of the Potential Effects on the Domestic Uranium Mining, Conversion and Enrichment Industries of the Introduction of DOE Excess Uranium Inventory During CY 2015 Through 2024 ENERGY RESOURCES INTERNATIONAL, INC. 1015 18 th Street, NW, Suite 650 Washington, DC 20036 USA Telephone: (202) 785-8833 Facsimile: (202) 785-8834 ERI-2142.18-1501 Analysis of the Potential Effects on the Domestic Uranium Mining, Conversion and Enrichment Industries of the Introduction of DOE

  9. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 Table 9. Summary production statistics of the U.S. uranium industry, 1993-2014 Exploration and Development Surface Drilling Exploration and Development Drilling Expenditures 1 Mine Production of Uranium Uranium Concentrate Production Uranium Concentrate Shipments Employment Year (million feet) (million dollars) (million pounds U 3 O 8 ) (million pounds U 3 O 8 )

  10. Vapor phase modifiers for oxidative coupling

    DOE Patents [OSTI]

    Warren, Barbara K. (Charleston, WV)

    1991-01-01

    Volatilized metal compounds retard vapor phase alkane conversion reactions in oxidative coupling processes that convert lower alkanes to higher hydrocarbons.

  11. 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...

  12. 2014 Uranium Marketing Annual Report

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

    Uranium Marketing Annual Report May 2015 Independent ... DC 20585 U.S. Energy Information Administration | 2014 ... Team, Office of Electricity, Renewables, and Uranium ...

  13. 2014 Domestic Uranium Production Report

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

    Domestic Uranium Production Report April 2015 Independent ... by the U.S. Energy Information Administration (EIA), ... Team, Office of Electricity, Renewables, and Uranium ...

  14. 2014 Uranium Marketing Annual Report

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

    U.S. Energy Information Administration 2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 thousand ...

  15. 2014 Uranium Marketing Annual Report

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

    U.S. Energy Information Administration 2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Minimum ...

  16. 2014 Uranium Marketing Annual Report

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

    ... Energy Information Administration, Form EIA-858 ""Uranium Marketing Annual Survey"" (2012-14)." "32 U.S. Energy Information Administration 2014 Uranium Marketing Annual Report

  17. 2014 Uranium Marketing Annual Report

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

    U.S. Energy Information Administration 2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Origin of ...

  18. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    SciTech Connect (OSTI)

    Diamond, D. J.; Baek, J. S.; Hanson, A. L.; Cheng, L-Y; Brown, N.; Cuadra, A.

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors, Chapter 18, Highly Enriched to Low-Enriched Uranium Conversions. The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  19. SINGLE-STEP CONVERSION OF UO$sub 3$ TO UF$sub 4$

    DOE Patents [OSTI]

    Moore, J.E.

    1960-07-12

    A description is given of the preparation of uranium tetrafluoride by reacting a hexavalent uranium compound with a pclysaccharide and gaseous hydrogen fluoride at an elevated temperature. Uranium trioxide and starch are combined with water to form a doughy mixture. which is extruded into pellets and dried. The pellets are then contacted with HF at a temperature from 500 to 700 deg C in a moving bed reactor to prcduce UF/sub 4/. Reduction of the hexavalent uranium to UO/sub 2/ and conversion of the UO/sub 2/ to UF/sub 4/ are accomplished simultaneously in this process.

  20. Thermochemical Conversion Processes | Department of Energy

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

    Processes Thermochemical Conversion Processes Gasification In gasification conversion, lignocellulosic feedstocks such as wood and forest products are broken down to synthesis gas, primarily carbon monoxide and hydrogen, using heat. The feedstock is then partially oxidized, or reformed with a gasifying agent (air, oxygen, or steam), which produces synthesis gas (syngas). The makeup of syngas will vary due to the different types of feedstocks, their moisture content, the type of gasifier used,

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

    DOE Patents [OSTI]

    Willit, James L. (Batavia, IL)

    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.

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

    DOE Patents [OSTI]

    Willit, James L. (Ratavia, IL)

    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.

  3. Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States and the Government of the Russian Federation has on the

    Energy Savers [EERE]

    Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States of America and the Government of the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Operation of the Gaseous Diffusion Plant 2008 Information Date: December 31, 2008 1 Introduction The Agreement Between the Government of the United States of America and the Government of the Russian Federation

  4. 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 Commissions Mid-Term Appraisal of the countrys 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 Indias 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.

  5. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    2 U.S. Energy Information Administration / 2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 2012 2013 2014 Advance Uranium Asset Management Ltd. (was Uranium Asset Management) American Fuel Resources, LLC Advance Uranium Asset Management Ltd. American Fuel Resources, LLC AREVA NC, Inc. AREVA / AREVA NC, Inc. AREVA NC, Inc. BHP Billiton Olympic Dam Corporation Pty Ltd ARMZ (AtomRedMetZoloto) BHP Billiton Olympic Dam

  6. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    a. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors, 1994-2014 million pounds U3O8 equivalent Delivery year Total purchased Purchased from U.S. producers Purchased from U.S. brokers and traders Purchased from other owners and operators of U.S. civilian nuclear power reactors, other U.S. suppliers, (and U.S. government for 2007)1 Purchased from foreign suppliers U.S.-origin uranium Foreign-origin uranium Spot contracts2 Short, medium, and long-term contracts3 1994

  7. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    . Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by supplier and delivery year, 2010-14 thousand pounds U3O8 equivalent, dollars per pound U3O8 equivalent Deliveries 2010 2011 2012 2013 2014 Purchased from U.S. producers Purchases of U.S.-origin and foreign-origin uranium 350 550 W W W Weighted-average price 47.13 58.12 W W W Purchased from U.S. brokers and traders Purchases of U.S.-origin and foreign-origin uranium 11,745 14,778 11,545 12,835 17,111

  8. 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 amendment arrival response data indicate some degree of overlap between the reactive species and thus potential for the formation of calcium-phosphate mineral phases (i.e., apatite formation), the efficiency of this treatment approach was relatively poor. In general, uranium performance monitoring results support the hypothesis that limited long-term treatment capacity (i.e., apatite formation) was established during the injection test. Two separate overarching issues affect the efficacy of apatite remediation for uranium sequestration within the 300 Area: 1) the efficacy of apatite for sequestering uranium under the present geochemical and hydrodynamic conditions, and 2) the formation and emplacement of apatite via polyphosphate technology. In addition, the long-term stability of uranium sequestered via apatite is dependent on the chemical speciation of uranium, surface speciation of apatite, and the mechanism of retention, which is highly susceptible to dynamic geochemical conditions. It was expected that uranium sequestration in the presence of hydroxyapatite would occur by sorption and/or surface complexation until all surface sites have been depleted, but that the high carbonate concentrations in the 300 Area would act to inhibit the transformation of sorbed uranium to chernikovite and/or autunite. Adsorption of uranium by apatite was never considered a viable approach for in situ uranium sequestration in and of itself, because by definition, this is a reversible reaction. The efficacy of uranium sequestration by apatite assumes that the adsorbed uranium would subsequently convert to autunite, or other stable uranium phases. Because this appears to not be the case in the 300 Area aquifer, even in locations near the river, apatite may have limited efficacy for the retention and long-term immobilization of uranium at the 300 Area site..

  9. BETO Conversion Program

    Broader source: Energy.gov [DOE]

    Breakout Session 2A—Conversion Technologies II: Bio-Oils, Sugar Intermediates, Precursors, Distributed Models, and Refinery Co-Processing BETO Conversion Program Bryna Berendzen, Technology Manager, Bioenergy Technologies Office, U.S. Department of Energy

  10. Domestic Uranium Production Report

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

    9. Summary production statistics of the U.S. uranium industry, 1993-2014 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

  11. Uranium Marketing Annual Report -

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

    b. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by purchaser, 2012-14 deliveries thousand pounds U3O8...

  12. Uranium Marketing Annual Report -

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

    1. Foreign sales of uranium from U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors by origin and delivery year, 2010-14 thousands pounds U3O8...

  13. 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 2014, by delivery year, 2015-24 thousand pounds U3O8 equivalent Year...

  14. Uranium Marketing Annual Report -

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

    0. U.S. broker and trader purchases of uranium by origin, supplier, and delivery year, 2010-14 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2010...

  15. Uranium Marketing Annual Report

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

    a. Foreign purchases, foreign sales, and uranium inventories owned by U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors, 1994-2014 million pounds U3O8...

  16. Uranium Marketing Annual Report -

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

    7. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by contract type and material type, 2014 deliveries thousand pounds U3O8 equivalent; dollars...

  17. Uranium Marketing Annual Report -

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

    4. Deliveries of uranium feed for enrichment by owners and operators of U.S. civilian nuclear power reactors by origin country and delivery year, 2012-14 thousand pounds U3O8...

  18. Uranium Marketing Annual Report -

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

    a. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by quantity, 2012-14 deliveries thousand pounds U3O8...

  19. 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)

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

    SciTech Connect (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.

  1. Sandia Energy - Wavelength Conversion Materials

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

    Wavelength Conversion Materials Home Energy Research EFRCs Solid-State Lighting Science EFRC Overview Wavelength Conversion Materials Wavelength Conversion MaterialsTara...

  2. 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.

  3. 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.

  4. Planning Document for an NBSR Conversion Safety Analysis Report

    SciTech Connect (OSTI)

    Diamond D. J.; Baek J.; Hanson, A.L.; Cheng, L-Y.; Brown, N.; Cuadra, A.

    2013-09-25

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the National Bureau of Standards Reactor (NBSR). The NBSR is a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a planning document for the conversion Safety Analysis Report (SAR) that would be submitted to, and approved by, the Nuclear Regulatory Commission (NRC) before the reactor could be converted.This report follows the recommended format and content from the NRC codified in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors, Chapter 18, Highly Enriched to Low-Enriched Uranium Conversions. The emphasis herein is on the SAR chapters that require significant changes as a result of conversion, primarily Chapter 4, Reactor Description, and Chapter 13, Safety Analysis. The document provides information on the proposed design for the LEU fuel elements and identifies what information is still missing. This document is intended to assist ongoing fuel development efforts, and to provide a platform for the development of the final conversion SAR. This report contributes directly to the reactor conversion pillar of the GTRI program, but also acts as a boundary condition for the fuel development and fuel fabrication pillars.

  5. 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

  6. file://\\\\fs-f1\\shared\\uranium\\uranium.html

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

    Glossary Home > Nuclear > U.S. Uranium Reserves Estimates U.S. Uranium Reserves Estimates Data for: 2008 Report Released: July 2010 Next Release Date: 2012 Summary The U.S. Energy...

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

    DOE Patents [OSTI]

    Hurst, Fred J. (Oak Ridge, TN); Crouse, David J. (Oak Ridge, TN)

    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.

  8. Uranium-titanium-niobium alloy

    DOE Patents [OSTI]

    Ludtka, Gail M. (Oak Ridge, TN); Ludtka, Gerard M. (Oak Ridge, TN)

    1990-01-01

    A uranium alloy having small additions of Ti and Nb shows improved strength and ductility in cross section of greater than one inch over prior uranium alloy having only Ti as an alloying element.

  9. 2014 Uranium Marketing Annual Report

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

    Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 2013 2014 2013 2014 2013 2014 Weighted-average price ...

  10. 2014 Uranium Marketing Annual Report

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

    2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 thousand pounds U 3 O 8 equivalent Year Maximum ...

  11. 2014 Uranium Marketing Annual Report

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

    1 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Deliveries Uranium concentrate Natural UF 6 Enriched UF 6 Total Purchases 2,004 1,312 ...

  12. 2014 Uranium Marketing Annual Report

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

    2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Table S1a. Uranium purchased by owners and operators of U.S. civilian nuclear power ...

  13. METHOD OF SINTERING URANIUM DIOXIDE

    DOE Patents [OSTI]

    Henderson, C.M.; Stavrolakis, J.A.

    1963-04-30

    This patent relates to a method of sintering uranium dioxide. Uranium dioxide bodies are heated to above 1200 nif- C in hydrogen, sintered in steam, and then cooled in hydrogen. (AEC)

  14. 2014 Domestic Uranium Production Report

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

    9. Summary production statistics of the U.S. uranium industry, 1993-2014" ,"Exploration and Development Surface ","Exploration and Development Drilling","Mine Production of Uranium ","Uranium Concentrate Production ","Uranium Concentrate Shipments ","Employment " "Year","Drilling (million feet)"," Expenditures 1 (million dollars)","Mine Production (million pounds U3O8)","(million pounds

  15. Domestic Uranium Production Report

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

    3. U.S. uranium concentrate production, shipments, and sales, 2003-14 Activity at U.S. mills and In-Situ-Leach plants 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 Estimated contained U3O8 (thousand pounds) Ore from Mines and Stockpiles Fed to Mills1 0 W W W 0 W W W W W W W Other Feed Materials 2 W W W W W W W W W W W W Total Mill Feed W W W W W W W W W W W W Uranium Concentrate Produced at U.S. Mills (thousand pounds U3O8) W W W W W W W W W W W W Uranium Concentrate Produced at

  16. Advanced Conversion Roadmap Workshop

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

    ... Deliverable: Roadmap for public dissemination which will guide Biomass program out-year R&D directions Workshop Objective Advanced Conversion Technology Roadmap Energy ...

  17. PROCESS OF PREPARING URANIUM CARBIDE

    DOE Patents [OSTI]

    Miller, W.E.; Stethers, H.L.; Johnson, T.R.

    1964-03-24

    A process of preparing uranium monocarbide is de scribed. Uranium metal is dissolved in cadmium, zinc, cadmium-- zinc, or magnesium-- zinc alloy and a small quantity of alkali metal is added. Addition of stoichiometric amounts of carbon at 500 to 820 deg C then precipitates uranium monocarbide. (AEC)

  18. Uranium from Seawater Program Review; Fuel Resources Uranium from Seawater Program DOE Office of Nuclear Energy

    SciTech Connect (OSTI)

    2013-07-01

    For nuclear energy to remain sustainable in the United States, economically viable sources of uranium beyond terrestrial ores must be developed. The goal of this program is to develop advanced adsorbents that can extract uranium from seawater at twice the capacity of the best adsorbent developed by researchers at the Japan Atomic Energy Agency (JAEA), 1.5 mg U/g adsorbent. A multidisciplinary team from Oak Ridge National Laboratory, Lawrence Berkeley National Laboratory, Pacific Northwest National Laboratory, and the University of Texas at Austin was assembled to address this challenging problem. Polymeric adsorbents, based on the radiation grafting of acrylonitrile and methacrylic acid onto high surface-area polyethylene fibers followed by conversion of the nitriles to amidoximes, have been developed. These poly(acrylamidoxime-co-methacrylic acid) fibers showed uranium adsorption capacities for the extraction of uranium from seawater that exceed 3 mg U/g adsorbent in testing at the Pacific Northwest National Laboratory Marine Sciences Laboratory. The essence of this novel technology lies in the unique high surface-area trunk material that considerably increases the grafting yield of functional groups without compromising its mechanical properties. This technology received an R&D100 Award in 2012. In addition, high surface area nanomaterial adsorbents are under development with the goal of increasing uranium adsorption capacity by taking advantage of the high surface areas and tunable porosity of carbon-based nanomaterials. Simultaneously, de novo structure-based computational design methods are being used to design more selective and stable ligands and the most promising candidates are being synthesized, tested and evaluated for incorporation onto a support matrix. Fundamental thermodynamic and kinetic studies are being carried out to improve the adsorption efficiency, the selectivity of uranium over other metals, and the stability of the adsorbents. Understanding the rate-limiting step of uranium uptake from seawater is also essential in designing an effective uranium recovery system. Finally, economic analyses have been used to guide these studies and highlight what parameters, such as capacity, recyclability, and stability, have the largest impact on the cost of extraction of uranium from seawater. Initially, the cost estimates by the JAEA for extraction of uranium from seawater with braided polymeric fibers functionalized with amidoxime ligands were evaluated and updated. The economic analyses were subsequently updated to reflect the results of this project while providing insight for cost reductions in the adsorbent development through “cradle-to-grave” case studies for the extraction process. This report highlights the progress made over the last three years on the design, synthesis, and testing of new materials to extract uranium for seawater. This report is organized into sections that highlight the major research activities in this project: (1) Chelate Design and Modeling, (2) Thermodynamics, Kinetics and Structure, (3) Advanced Polymeric Adsorbents by Radiation Induced Grafting, (4) Advanced Nanomaterial Adsorbents, (5) Adsorbent Screening and Modeling, (6) Marine Testing, and (7) Cost and Energy Assessment. At the end of each section, future research directions are briefly discussed to highlight the challenges that still remain to reduce the cost of extractions of uranium for seawater. Finally, contributions from the Nuclear Energy University Programs (NEUP), which complement this research program, are included at the end of this report.

  19. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    b. Weighted-average price of uranium purchased by owners and operators of U.S. civilian nuclear power reactors, 1994-2014 dollars per pound U3O8 equivalent Delivery year Total purchased (weighted-average price) Purchased from U.S. producers Purchased from U.S. brokers and traders Purchased from other owners and operators of U.S. civilian nuclear power reactors, other U.S. suppliers, (and U.S. government for 2007)1 Purchased from foreign suppliers U.S.-origin uranium (weighted-average price)

  20. Domestic Uranium Production Report

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

    10. Uranium reserve estimates at the end of 2013 and 2014 million pounds U3O8 End of 2013 End of 2014 Forward Cost2 Uranium Reserve Estimates1 by Mine and Property Status, Mining Method, and State(s) $0 to $30 per pound $0 to $50 per pound $0 to $100 per pound $0 to $30 per pound $0 to $50 per pound $0 to $100 per pound Properties with Exploration Completed, Exploration Continuing, and Only Assessment Work W W 130.7 W W 154.6 Properties Under Development for Production and Development Drilling W

  1. Uranium Transport Modeling

    SciTech Connect (OSTI)

    Bostick, William D.

    2008-01-15

    Uranium contamination is prevalent at many of the U.S. DOE facilities and at several civilian sites that have supported the nuclear fuel cycle. The potential off-site mobility of uranium depends on the partitioning of uranium between aqueous and solid (soil and sediment) phases. Hexavalent U (as uranyl, UO{sub 2}{sup 2+}) is relatively mobile, forming strong complexes with ubiquitous carbonate ion which renders it appreciably soluble even under mild reducing conditions. In the presence of carbonate, partition of uranyl to ferri-hydrate and select other mineral phases is usually maximum in the near-neutral pH range {approx} 5-8. The surface complexation reaction of uranyl with iron-containing minerals has been used as one means to model subsurface migration, used in conjunction with information on the site water chemistry and hydrology. Partitioning of uranium is often studied by short-term batch 'equilibrium' or long-term soil column testing ; MCLinc has performed both of these methodologies, with selection of method depending upon the requirements of the client or regulatory authority. Speciation of uranium in soil may be determined directly by instrumental techniques (e.g., x-ray photoelectron spectroscopy, XPS; x-ray diffraction, XRD; etc.) or by inference drawn from operational estimates. Often, the technique of choice for evaluating low-level radionuclide partitioning in soils and sediments is the sequential extraction approach. This methodology applies operationally-defined chemical treatments to selectively dissolve specific classes of macro-scale soil or sediment components. These methods recognize that total soil metal inventory is of limited use in understanding bioavailability or metal mobility, and that it is useful to estimate the amount of metal present in different solid-phase forms. Despite some drawbacks, the sequential extraction method can provide a valuable tool to distinguish among trace element fractions of different solubility related to mineral phases. Four case studies are presented: Water and Soil Characterization, Subsurface Stabilization of Uranium and other Toxic Metals, Reductive Precipitation (in situ bioremediation) of Uranium, and Physical Transport of Particle-bound Uranium by Erosion.

  2. High loading uranium fuel plate

    DOE Patents [OSTI]

    Wiencek, Thomas C.; Domagala, Robert F.; Thresh, Henry R.

    1990-01-01

    Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pair of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat having a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process.

  3. Can Ionic Liquids Be Used As Templating Agents For Controlled Design of Uranium-Containing Nanomaterials?

    SciTech Connect (OSTI)

    Visser, A.; Bridges, N.; Tosten, M.

    2013-04-09

    Nanostructured uranium oxides have been prepared in ionic liquids as templating agents. Using the ionic liquids as reaction media for inorganic nanomaterials takes advantage of the pre-organized structure of the ionic liquids which in turn controls the morphology of the inorganic nanomaterials. Variation of ionic liquid cation structure was investigated to determine the impact on the uranium oxide morphologies. For two ionic liquid cations, increasing the alkyl chain length increases the aspect ratio of the resulting nanostructured oxides. Understanding the resulting metal oxide morphologies could enhance fuel stability and design.

  4. Extraction of uranium from spent fuels using liquefied gases

    SciTech Connect (OSTI)

    Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi

    2007-07-01

    For reprocessing of spent nuclear fuels, a novel method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. As a fundamental study, the nitrate conversion with liquefied nitrogen dioxide and the nitrate extraction with supercritical carbon dioxide were demonstrated by using uranium dioxide powder, uranyl nitrate and tri-n-butylphosphate complex in the present study. (authors)

  5. Uranium Reduction by Clostridia

    SciTech Connect (OSTI)

    Francis, A.J.; Dodge, Cleveland J.; Gillow, Jeffrey B.

    2006-04-05

    The FRC groundwater and sediment contain significant concentrations of U and Tc and are dominated by low pH, and high nitrate and Al concentrations where dissimilatory metal reducing bacterial activity may be limited. The presence of Clostridia in Area 3 at the FRC site has been confirmed and their ability to reduce uranium under site conditions will be determined. Although the phenomenon of uranium reduction by Clostridia has been firmly established, the molecular mechanisms underlying such a reaction are not very clear. The authors are exploring the hypothesis that U(VI) reduction occurs through hydrogenases and other enzymes (Matin and Francis). Fundamental knowledge of metal reduction using Clostridia will allow us to exploit naturally occurring processes to attenuate radionuclide and metal contaminants in situ in the subsurface. The outline for this report are as follows: (1) Growth of Clostridium sp. under normal culture conditions; (2) Fate of metals and radionuclides in the presence of Clostridia; (3) Bioreduction of uranium associated with nitrate, citrate, and lepidocrocite; and (4) Utilization of Clostridium sp. for immobilization of uranium at the FRC Area 3 site.

  6. GLOBAL THREAT REDUCTION INITIATIVE REACTOR CONVERSION PROGRAM: STATUS AND CURRENT PLANS

    SciTech Connect (OSTI)

    Staples, Parrish A.; Leach, Wayne; Lacey, Jennifer M.

    2009-10-07

    The U.S. Department of Energys National Nuclear Security Administration (NNSA) Reactor Conversion Program supports the minimization, and to the extent possible, elimination of the use of high enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors and radioisotope production processes to the use of low enriched uranium (LEU). The Reactor Conversion Program is a technical pillar of the NNSA Global Threat Reduction Initiative (GTRI) which is a key organization for implementing U.S. HEU minimization policy and works to reduce and protect vulnerable nuclear and radiological material domestically and abroad.

  7. The IMCA: A field instrument for uranium enrichment measurements

    SciTech Connect (OSTI)

    Gardner, G.H.; Koskelo, M.; Moeslinger, M.; Mayer, R.L. II; McGinnis, B.R.; Wishard, B.

    1996-12-31

    The IMCA (Inspection Multi-Channel Analyzer) is a portable gamma-ray spectrometer designed to measure the enrichment of uranium either in a laboratory or in the field. The IMCA consists of a Canberra InSpector Multi-Channel Analyzer, sodium iodide or a planar germanium detector, and special application software. The system possesses a high degree of automation. The IMCA uses the uranium enrichment meter principle, and is designed to meet the International Atomic Energy Agency (IAEA) requirements for the verification of enriched uranium materials. The IMCA is available with MGA plutonium isotopic analysis software or MGAU uranium analysis software as well. In this paper, the authors present a detailed description of the hardware and software of the IMCA system, as well as results from preliminary measurements testing compliance of IMCA with IAEA requirements using uranium standards and UF6 cylinders. Measurements performed on UF6 cylinders in the field under variable environmental conditions (temperatures ranging from 0 to 35 C) have shown that good results can be achieved. The enrichment of UF6 contained in the cylinder is determined by using calibration constants generated from an instrument calibration, using traceable uranium oxide standards, performed in the laboratory under controlled environmental conditions. The IMCA software is designed to make the necessary matrix and container corrections to ensure that accurate results are achieved in the field.

  8. Biochemical Conversion | Department of Energy

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

    Conversion Biochemical Conversion This area focuses on the research, development and demonstration of biological processes that convert biomass to biofuels, chemicals, and power. Biochemical processes also complement thermochemical conversion by providing residual materials for further processing. Biochemical conversion will advance in the future by enhancing fuel yields in integrated biorefineries which combine conversion types with heat and power efficiencies to produce fuel and products.

  9. Colloids generation from metallic uranium fuel

    SciTech Connect (OSTI)

    Metz, C.; Fortner, J.; Goldberg, M.; Shelton-Davis, C.

    2000-07-20

    The possibility of colloid generation from spent fuel in an unsaturated environment has significant implications for storage of these fuels in the proposed repository at Yucca Mountain. Because colloids can act as a transport medium for sparingly soluble radionuclides, it might be possible for colloid-associated radionuclides to migrate large distances underground and present a human health concern. This study examines the nature of colloidal materials produced during corrosion of metallic uranium fuel in simulated groundwater at elevated temperature in an unsaturated environment. Colloidal analyses of the leachates from these corrosion tests were performed using dynamic light scattering and transmission electron microscopy. Results from both techniques indicate a bimodal distribution of small discrete particles and aggregates of the small particles. The average diameters of the small, discrete colloids are {approximately}3--12 nm, and the large aggregates have average diameters of {approximately}100--200 nm. X-ray diffraction of the solids from these tests indicates a mineral composition of uranium oxide or uranium oxy-hydroxide.

  10. Extraction of uranium: comparison of stripping with ammonia vs. strong acid

    SciTech Connect (OSTI)

    Moldovan, B.; Grinbaum, B.; Efraim, A.

    2008-07-01

    Following extraction of uranium in the first stage of solvent extraction using a tertiary amine, typically Alamine 336, the stripping of the extracted uranium is accomplished either by use of an aqueous solution of (NH{sub 4}){sub 2}SO{sub 4} /NH{sub 4}OH or by strong-acid stripping using 400-500 g/L H{sub 2}SO{sub 4}. Both processes have their merits and determine the downstream processing. The classical stripping with ammonia is followed by addition of strong base, to precipitate ammonium uranyl sulfate (NH{sub 4}){sub 2}UO{sub 2}(SO{sub 4}){sub 2}, which yields finally the yellow cake. Conversely, stripping with H{sub 2}SO{sub 4}, followed by oxidation with hydrogen peroxide yields uranyl oxide as product. At the Cameco Key Lake operation, both processes were tested on a pilot scale, using a Bateman Pulsed Column (BPC). The BPC proved to be applicable to both processes. It met the process criteria both for extraction and stripping, leaving less than 1 mg/L of U{sub 3}O{sub 8} in the raffinate, and product solution had the required concentration of U{sub 3}O{sub 8} at high flux and reasonable height of transfer unit. In the Key Lake mill, each operation can be carried out in a single column. The main advantages of the strong-acid stripping over ammonia stripping are: (1) 60% higher flux in the extraction, (2) tenfold higher concentration of the uranium in the product solution, and (3) far more robust process, with no need of pH control in the stripping and no need to add acid to the extraction in order to keep the pH above the point of precipitation of iron compounds. The advantages of the ammoniacal process are easier stripping, that is, less stages needed to reach equilibrium and lower concentration of modifier needed to prevent the creation of a third phase. (authors)

  11. Solid Fuels Conversion

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

    Solid Fuels Conversion - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced

  12. Algal Biomass Conversion

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

    BETO 2015 Project Peer Review Algal Biomass Conversion WBS 1.3.4.201 Philip T. Pienkos National Renewable Energy Laboratory March 24 th , 2015 This presentation does not contain any proprietary, confidential, or otherwise restricted information 2 Goal Statement Reduce algal biofuel production cost by developing advanced process options for the conversion of algal biomass into biofuels and bioproducts based on the three major biomass components: lipids, carbohydrates, and proteins. 3 Quad Chart

  13. power conversion efficiency

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

    power conversion efficiency - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs

  14. thermal energy power conversion

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

    thermal energy power conversion - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs

  15. Energy Conversion Efficiency

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

    Conversion Efficiency - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced

  16. Wavelength Conversion Materials

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

    Wavelength Conversion Materials - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs

  17. IL conversion technology

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

    conversion technology - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced

  18. Structured luminescence conversion layer

    DOE Patents [OSTI]

    Berben, Dirk; Antoniadis, Homer; Jermann, Frank; Krummacher, Benjamin Claus; Von Malm, Norwin; Zachau, Martin

    2012-12-11

    An apparatus device such as a light source is disclosed which has an OLED device and a structured luminescence conversion layer deposited on the substrate or transparent electrode of said OLED device and on the exterior of said OLED device. The structured luminescence conversion layer contains regions such as color-changing and non-color-changing regions with particular shapes arranged in a particular pattern.

  19. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    SciTech Connect (OSTI)

    A K Wertsching

    2012-09-01

    As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.

  20. Sandia Energy - Solid Fuels Conversion

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

    Solid Fuels Conversion Home Transportation Energy Predictive Simulation of Engines Clean FuelsPower Solid Fuels Conversion Solid Fuels ConversionAshley Otero2015-10-28T02:40:48+00...

  1. Method for fabricating uranium foils and uranium alloy foils

    DOE Patents [OSTI]

    Hofman, Gerard L.; Meyer, Mitchell K.; Knighton, Gaven C.; Clark, Curtis R.

    2006-09-05

    A method of producing thin foils of uranium or an alloy. The uranium or alloy is cast as a plate or sheet having a thickness less than about 5 mm and thereafter cold rolled in one or more passes at substantially ambient temperatures until the uranium or alloy thereof is in the shape of a foil having a thickness less than about 1.0 mm. The uranium alloy includes one or more of Zr, Nb, Mo, Cr, Fe, Si, Ni, Cu or Al.

  2. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    5. Enrichment service sellers to owners and operators of U.S. civilian nuclear power reactors, 2012-14 2012 2013 2014 Advance Uranium Asset Management Ltd. AREVA NC, Inc AREVA Enrichment Services, LLC / AREVA NC, Inc. AREVA NC, Inc. .CNEIC (China Nuclear Energy Industry Corporation) CNEIC (China Nuclear Energy Industry Corporation) CNEIC (China Nuclear Energy Industry Corporation) LES, LLC (Louisiana Energy Services) LES, LLC (Louisiana Energy Services) LES, LLC (Louisiana Energy Services)

  3. recycled_uranium.cdr

    Office of Legacy Management (LM)

    Recycled Uranium and Transuranics: Their Relationship to Weldon Spring Site Remedial Action Project Introduction Historical Perspective On August 8, 1999, Energy Secretary Bill Richardson announced a comprehensive set of actions to address issues raised at the Paducah, Kentucky, Gaseous Diffusion Plant that may have had the potential to affect the health of the workers. One of the issues addressed the need to determine the extent and significance of radioactive fission products and transuranic

  4. ELECTROLYSIS OF THORIUM AND URANIUM

    DOE Patents [OSTI]

    Hansen, W.N.

    1960-09-01

    An electrolytic method is given for obtaining pure thorium, uranium, and thorium-uranium alloys. The electrolytic cell comprises a cathode composed of a metal selected from the class consisting of zinc, cadmium, tin, lead, antimony, and bismuth, an anode composed of at least one of the metals selected from the group consisting of thorium and uranium in an impure state, and an electrolyte composed of a fused salt containing at least one of the salts of the metals selected from the class consisting of thorium, uranium. zinc, cadmium, tin, lead, antimony, and bismuth. Electrolysis of the fused salt while the cathode is maintained in the molten condition deposits thorium, uranium, or thorium-uranium alloys in pure form in the molten cathode which thereafter may be separated from the molten cathode product by distillation.

  5. Vapor phase modifiers for oxidative coupling

    DOE Patents [OSTI]

    Warren, B.K.

    1991-12-17

    Volatilized metal compounds are described which are capable of retarding vapor phase alkane conversion reactions in oxidative coupling processes that convert lower alkanes to higher hydrocarbons.

  6. VANE Uranium One JV | Open Energy Information

    Open Energy Info (EERE)

    VANE Uranium One JV Jump to: navigation, search Name: VANE-Uranium One JV Place: London, England, United Kingdom Zip: EC4V 6DX Product: JV between VANE Minerals Plc & Uranium One....

  7. SEPARATION OF THORIUM FROM URANIUM

    DOE Patents [OSTI]

    Bane, R.W.

    1959-09-01

    A description is given for the separation of thorium from uranium by forming an aqueous acidic solution containing ionic species of thorium, uranyl uranium, and hydroxylamine, flowing the solution through a column containing the phenol-formaldehyde type cation exchange resin to selectively adsorb substantially all the thorium values and a portion of the uranium values, flowing a dilute solution of hydrochloric acid through the column to desorb the uranium values, and then flowing a dilute aqueous acidic solution containing an ion, such as bisulfate, which has a complexing effect upon thortum through the column to desorb substantially all of the thorium.

  8. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Delivery year Total purchased (weighted- average price) Purchased from U.S. producers Purchased from U.S. brokers and traders Purchased from other owners and operators of U.S. civilian nuclear power reactors, other U.S. suppliers, (and U.S. government for 2007) 1 Purchased from foreign suppliers U.S.-origin uranium (weighted- average price) Foreign-origin uranium (weighted-

  9. 2014 Domestic Uranium Production Report

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

    or dissolving-out from mined rock, of the soluble uranium constituents by the natural action of percolating a prepared chemical solution through mounded (heaped) rock material. ...

  10. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

    Appropriations Subcommittee, is shown some of the technology in the Highly Enriched Uranium Materials Facility by Warehousing and Transportation Operations Manager Byron...

  11. 2014 Domestic Uranium Production Report

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

    Jab and Antelope Sweetwater, Wyoming 2,000,000 Developing Developing Developing Developing Developing Uranium One Americas, Inc. Moore Ranch Campbell, Wyoming 500,000 Permitted And ...

  12. 2014 Domestic Uranium Production Report

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

    ...ing","Developing","Developing","Developing","Developing" "Uranium One Americas, Inc.","Moore Ranch","Campbell, Wyoming",500000,"Permitted And Licensed","Permitted And ...

  13. 2014 Uranium Marketing Annual Survey

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

    5 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Quantity with reported price Weighted-average price Quantity with reported price ...

  14. Geochemical Evaluation of Uranium Fate and Transport Guterl Specialty Steel Site, New York - 12077

    SciTech Connect (OSTI)

    Frederick, Bill; Tandon, Vikas

    2012-07-01

    Between 1948 and 1952, up to 15,875 metric tons (35 million pounds) of natural uranium metal (U) were processed at the former Guterl Specialty Steel Corporation site in Lockport, New York. The resulting dust, thermal scale, mill shavings and associated land disposal contaminated both the facility and on-site soils. Uranium subsequently impacted groundwater and a fully developed plume exists below the site. Site soils are composed of anthropogenic fill and re-worked, glacially-derived native soil. This overburden is underlain by the weathered and fractured Lockport Dolostone bedrock. Shallow groundwater levels fluctuate seasonally and allow groundwater to contact U contaminated soil, which promotes transport. This condition is exemplified through coincident increases in specific conductivity and groundwater levels, which flush soluble constituents in the fill/soil to groundwater during recharge events. In addition, water-level fluctuations affect reduction-oxidation (redox) conditions at the site. The U in soils is subject to wetting and drying cycles that promote oxidation more than stable redox conditions (e.g., dry soil or fully saturated conditions). This oxidizing mechanism increases uranium solubility and mobility. Site groundwater also receives uranium via leaching from near-surface contaminated fill. The strong correlation between nitrate and uranium in groundwater indicates that uranium is mobile where oxidizing conditions occur. Analytical models of contaminant leaching determined that multiple pathways and transport mechanisms govern site risk. Uranium transport to groundwater involves three mechanisms: 1) direct contact of contaminated soil with groundwater, 2) the oxidation-state or chemical valence of uranium, and 3) the leaching of near-surface contamination to groundwater. These mechanisms require an integrated remedial solution that is sustainable and cost effective. (authors)

  15. Conversion Technologies for Advanced Biofuels - Carbohydrates...

    Broader source: Energy.gov (indexed) [DOE]

    Advanced Conversion Roadmap Workshop Workshop on Conversion Technologies for Advanced Biofuels - Carbohydrates Conversion Technologies for Advanced Biofuels - Carbohydrates...

  16. Method for providing uranium with a protective copper coating

    DOE Patents [OSTI]

    Waldrop, Forrest B. (Powell, TN); Jones, Edward (Knoxville, TN)

    1981-01-01

    The present invention is directed to a method for providing uranium metal with a protective coating of copper. Uranium metal is subjected to a conventional cleaning operation wherein oxides and other surface contaminants are removed, followed by etching and pickling operations. The copper coating is provided by first electrodepositing a thin and relatively porous flash layer of copper on the uranium in a copper cyanide bath. The resulting copper-layered article is then heated in an air or inert atmosphere to volatilize and drive off the volatile material underlying the copper flash layer. After the heating step an adherent and essentially non-porous layer of copper is electro-deposited on the flash layer of copper to provide an adherent, multi-layer copper coating which is essentially impervious to corrosion by most gases.

  17. Calculating Atomic Number Densities for Uranium

    Energy Science and Technology Software Center (OSTI)

    1993-01-01

    Provides method to calculate atomic number densities of selected uranium compounds and hydrogenous moderators for use in nuclear criticality safety analyses at gaseous diffusion uranium enrichment facilities.

  18. Nuclear radiation cleanup and uranium prospecting (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Nuclear radiation cleanup and uranium prospecting Citation Details In-Document Search Title: Nuclear radiation cleanup and uranium prospecting Apparatus, systems, and methods for...

  19. Nuclear radiation cleanup and uranium prospecting (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Nuclear radiation cleanup and uranium prospecting Citation Details In-Document Search Title: Nuclear radiation cleanup and uranium prospecting You are accessing a document from...

  20. Multiple Mechanisms of Uranium Immobilization by Cellulomonas...

    Office of Scientific and Technical Information (OSTI)

    Journal Article: Multiple Mechanisms of Uranium Immobilization by Cellulomonas sp. Strain ES6 Citation Details In-Document Search Title: Multiple Mechanisms of Uranium ...

  1. Uranium Resources Inc URI | Open Energy Information

    Open Energy Info (EERE)

    exploring, developing and mining uranium properties using the in situ recovery (ISR) or solution mining process. References: Uranium Resources, Inc. (URI)1 This article...

  2. Uranium Enrichment Decontamination and Decommissioning Fund's...

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

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2008 and 2007 Financial Statement Audit, OAS-FS-10-05 Uranium Enrichment Decontamination and...

  3. Direct conversion technology

    SciTech Connect (OSTI)

    Massier, P.F.; Back, L.H.; Ryan, M.A.; Fabris, G.

    1992-01-07

    The overall objective of the Direct Conversion Technology task is to develop an experimentally verified technology base for promising direct conversion systems that have potential application for energy conservation in the end-use sectors. This report contains progress of research on the Alkali Metal Thermal-to-Electric Converter (AMTEC) and on the Two-Phase Liquid-Metal MHD Electrical Generator (LMMHD) for the period January 1, 1991 through December 31, 1991. Research on AMTEC and on LMMHD was initiated during October 1987. Reports prepared on previous occasions (Refs. 1--5) contain descriptive and performance discussions of the following direct conversion concepts: thermoelectric, pyroelectric, thermionic, thermophotovoltaic, thermoacoustic, thermomagnetic, thermoelastic (Nitionol heat engine); and also, more complete descriptive discussions of AMTEC and LMMHD systems.

  4. Digital optical conversion module

    DOE Patents [OSTI]

    Kotter, D.K.; Rankin, R.A.

    1988-07-19

    A digital optical conversion module used to convert an analog signal to a computer compatible digital signal including a voltage-to-frequency converter, frequency offset response circuitry, and an electrical-to-optical converter. Also used in conjunction with the digital optical conversion module is an optical link and an interface at the computer for converting the optical signal back to an electrical signal. Suitable for use in hostile environments having high levels of electromagnetic interference, the conversion module retains high resolution of the analog signal while eliminating the potential for errors due to noise and interference. The module can be used to link analog output scientific equipment such as an electrometer used with a mass spectrometer to a computer. 2 figs.

  5. Digital optical conversion module

    DOE Patents [OSTI]

    Kotter, Dale K. (North Shelley, ID); Rankin, Richard A. (Ammon, ID)

    1991-02-26

    A digital optical conversion module used to convert an analog signal to a computer compatible digital signal including a voltage-to-frequency converter, frequency offset response circuitry, and an electrical-to-optical converter. Also used in conjunction with the digital optical conversion module is an optical link and an interface at the computer for converting the optical signal back to an electrical signal. Suitable for use in hostile environments having high levels of electromagnetic interference, the conversion module retains high resolution of the analog signal while eliminating the potential for errors due to noise and interference. The module can be used to link analog output scientific equipment such as an electrometer used with a mass spectrometer to a computer.

  6. Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina; Primm, Trent

    2009-11-01

    An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

  7. Direct Conversion Technology

    SciTech Connect (OSTI)

    Back, L.H.; Fabris, G.; Ryan, M.A.

    1992-07-01

    The overall objective of the Direct Conversion Technology task is to develop an experimentally verified technology base for promising direct conversion systems that have potential application for energy conservation in the end-use sectors. Initially, two systems were selected for exploratory research and advanced development. These are Alkali Metal Thermal-to-Electric Converter (AMTEC) and Two-Phase Liquid Metal MD Generator (LMMHD). This report describes progress that has been made during the first six months of 1992 on research activities associated with these two systems. (GHH)

  8. March market review. [Spot market prices for uranium (1993)

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The spot market price for uranium in unrestricted markets weakened further during March, and at month end, the NUEXCO Exchange Value had fallen $0.15, to $7.45 per pound U3O8. The Restricted American Market Penalty (RAMP) for concentrates increased $0.15, to $2.55 per pound U3O8. Ample UF6 supplies and limited demand led to a $0.50 decrease in the UF6 Value, to $25.00 per kgU as UF6, while the RAMP for UF6 increased $0.75, to $5.25 per kgU. Nine near-term uranium transactions were reported, totalling almost 3.3 million pounds equivalent U3O8. This is the largest monthly spot market volume since October 1992, and is double the volume reported in January and February. The March 31 Conversion Value was $4.25 per kgU as UF6. Beginning with the March 31 Value, NUEXCO now reports its Conversion Value in US dollars per kilogram of uranium (US$/kgU), reflecting current industry practice. The March loan market was inactive with no transactions reported. The Loan Rate remained unchanged at 3.0 percent per annum. Low demand and increased competition among sellers led to a one-dollar decrease in the SWU Value, to $65 per SWU, and the RAMP for SWU declined one dollar, to $9 per SWU.

  9. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    0. Contracted purchases of uranium from suppliers by owners and operators of U.S. civilian nuclear power reactors, in effect at the end of 2014, by delivery year, 2015-24 thousand pounds U3O8 equivalent Contracted purchases from U.S. suppliers Contracted purchases from foreign suppliers Contracted purchases from all suppliers Year of delivery Minimum Maximum Minimum Maximum Minimum Maximum 2015 8,405 8,843 31,468 34,156 39,873 42,999 2016 7,344 7,757 29,660 31,787 37,004 39,544 2017 5,980 6,561

  10. Domestic Uranium Production Report

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

    7. Employment in the U.S. uranium production industry by state, 2003-14 person-years State(s) 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 Wyoming 134 139 181 195 245 301 308 348 424 512 531 416 Colorado and Texas 48 140 269 263 557 696 340 292 331 248 198 105 Nebraska and New Mexico 92 102 123 160 149 160 159 134 127 W W W Arizona, Utah, and Washington 47 40 75 120 245 360 273 281 W W W W Alaska, Michigan, Nevada, and South Dakota 0 0 0 16 25 30 W W W W W 0 California, Montana,

  11. Domestic Uranium Production Report

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

    4. U.S. uranium mills by owner, location, capacity, and operating status at end of the year, 2010-14 Owner Mill and Heap Leach1 Facility name County, state (existing and planned locations) Capacity (short tons of ore per day) Operating status at end of the year 2010 2011 2012 2013 2014 EFR White Mesa LLC White Mesa Mill San Juan, Utah 2,000 Operating Operating Operating Operating Processing Alternate Feed Operating-Processing Alternate Feed Energy Fuels Resources Corp Pinon Ridge Mill Montrose,

  12. Speciation of Uranium in Biologically Reduced Sediments in the Old Rifle

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

    Aquifer | Stanford Synchrotron Radiation Lightsource Speciation of Uranium in Biologically Reduced Sediments in the Old Rifle Aquifer Wednesday, May 16, 2012 - 1:30pm SSRL Conference Room 137-322 Juan S. Lezama Pacheco The speciation and dynamics of Uranium(IV) in naturally and artificially bioreduced sediments, as well as its local nanometer-to-millimeter scale physical and chemical environment, controls its stability, susceptibility to oxidation, and subsequent transport behavior in

  13. Laser isotope separation: Uranium. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect (OSTI)

    1995-12-01

    The bibliography contains citations concerning the technology and assessment of laser separation of uranium isotopes, compounds, oxides, and alloys. Topics include uranium enrichment plants, isotope enriched materials, gaseous diffusion, centrifuge enrichment, reliability and safety, and atomic vapor separation. Citations also discuss commercial enrichment, market trends, licensing, international competition, and waste management. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  14. Ocean thermal energy conversion

    SciTech Connect (OSTI)

    Avery, W.H.

    1983-03-17

    A brief explanation of the Ocean Thermal Energy Conversion (OTEC) concept and an estimate of the amount of energy that can be produced from the ocean resource without introducing environmental concerns are presented. Use of the OTEC system to generate electric power and products which can replace fossil fuels is shown. The OTEC program status and its prospects for the future are discussed.

  15. Uranium Downblending and Disposition Project Technology Readiness

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

    Assessment | Department of Energy Uranium Downblending and Disposition Project Technology Readiness Assessment Uranium Downblending and Disposition Project Technology Readiness Assessment Full Document and Summary Versions are available for download PDF icon Uranium Downblending and Disposition Project Technology Readiness Assessment PDF icon Summary - Uranium233 Downblending and Disposition Project More Documents & Publications Compilation of TRA Summaries EA-1574: Final Environmental

  16. PREPARATION OF URANIUM-ALUMINUM ALLOYS

    DOE Patents [OSTI]

    Moore, R.H.

    1962-09-01

    A process is given for preparing uranium--aluminum alloys from a solution of uranium halide in an about equimolar molten alkali metal halide-- aluminum halide mixture and excess aluminum. The uranium halide is reduced and the uranium is alloyed with the excess aluminum. The alloy and salt are separated from each other. (AEC)

  17. EIS-0283-SA-03: Supplement Analysis | Department of Energy

    Energy Savers [EERE]

    Analysis Transportation of Depleted Uranium Hexafluoride for Conversion to Depleted Uranium Oxide This SA evaluates a proposal to transport cylinders of DUF from the Paducah...

  18. Wind energy conversion system

    DOE Patents [OSTI]

    Longrigg, Paul (Golden, CO)

    1987-01-01

    The wind energy conversion system includes a wind machine having a propeller connected to a generator of electric power, the propeller rotating the generator in response to force of an incident wind. The generator converts the power of the wind to electric power for use by an electric load. Circuitry for varying the duty factor of the generator output power is connected between the generator and the load to thereby alter a loading of the generator and the propeller by the electric load. Wind speed is sensed electro-optically to provide data of wind speed upwind of the propeller, to thereby permit tip speed ratio circuitry to operate the power control circuitry and thereby optimize the tip speed ratio by varying the loading of the propeller. Accordingly, the efficiency of the wind energy conversion system is maximized.

  19. Uranium Processing Facility | Y-12 National Security Complex

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

    Y-12 Uranium Processing Facility Uranium Processing Facility UPF will be a state-of-the-art, consolidated facility for enriched uranium operations including assembly,...

  20. Uranium hexafluoride bibliography

    SciTech Connect (OSTI)

    Burnham, S.L.

    1988-01-01

    This bibliography is a compilation of reports written about the transportation, handling, safety, and processing of uranium hexafluoride. An on-line literature search was executed using the DOE Energy files and the Nuclear Science Abstracts file to identify pertinent reports. The DOE Energy files contain unclassified information that is processed at the Office of Scientific and Technical Information of the US Department of Energy. The reports selected from these files were published between 1974 and 1983. Nuclear Science Abstracts contains unclassified international nuclear science and technology literature published from 1948 to 1976. In addition, scientific and technical reports published by the US Atomic Energy Commission and the US Energy Research and Development Administration, as well as those published by other agencies, universities, and industrial and research organizations, are included in the Nuclear Science Abstracts file. An alphabetical listing of the acronyms used to denote the corporate sponsors follows the bibliography.

  1. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    3. Deliveries of uranium feed by owners and operators of U.S. civilian nuclear power reactors by enrichment country and delivery year, 2012-14 thousand pounds U3O8 equivalent Feed deliveries in 2012 Feed deliveries in 2013 Feed deliveries in 2014 Enrichment country U.S.-origin Foreign-origin Total U.S.-origin Foreign-origin Total U.S.-origin Foreign-origin Total China 0 W W 0 W W W W W France 0 4,578 4,578 0 1,606 1,606 0 3.055 3,055 Germany W W 1,904 W W W W W 2,140 Netherlands W W 2,674 1,058

  2. Domestic Uranium Production Report

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

    1. U.S. uranium drilling activities, 2003-14 Exploration drilling Development drilling Exploration and development drilling Year Number of holes Feet (thousand) Number of holes Feet (thousand) Number of holes Feet (thousand) 2003 NA NA NA NA W W 2004 W W W W 2,185 1,249 2005 W W W W 3,143 1,668 2006 1,473 821 3,430 1,892 4,903 2,713 2007 4,351 2,200 4,996 2,946 9,347 5,146 2008 5,198 2,543 4,157 2,551 9,355 5,093 2009 1,790 1,051 3,889 2,691 5,679 3,742 2010 2,439 1,460 4,770 3,444 7,209 4,904

  3. Domestic Uranium Production Report

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

    6. Employment in the U.S. uranium production industry by category, 2003-14 person-years Year Exploration Mining Milling Processing Reclamation Total 2003 W W W W 117 321 2004 18 108 W W 121 420 2005 79 149 142 154 124 648 2006 188 121 W W 155 755 2007 375 378 107 216 155 1,231 2008 457 558 W W 154 1,563 2009 175 441 W W 162 1,096 2010 211 400 W W 125 1,073 2011 208 462 W W 102 1,191 2012 161 462 W W 179 1,196 2013 149 392 W W 199 1,156 2014 86 246 W W 161

  4. Domestic Uranium Production Report

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

    2. U.S. uranium mine production and number of mines and sources, 2003-14 Production / Mining method 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 Underground (estimated contained thousand pounds U3O8) W W W W W W W W W W W W Open Pit (estimated contained thousand pounds U3O8) 0 0 0 0 0 0 0 0 0 0 0 0 In-Situ Leaching (thousand pounds U3O8) W W 2,681 4,259 W W W W W W W W Other1 (thousand pounds U3O8) W W W W W W W W W W W W Total Mine Production (thousand pounds U3O8) E2,200 2,452

  5. Conversion of Questionnaire Data

    SciTech Connect (OSTI)

    Powell, Danny H; Elwood Jr, Robert H

    2011-01-01

    During the survey, respondents are asked to provide qualitative answers (well, adequate, needs improvement) on how well material control and accountability (MC&A) functions are being performed. These responses can be used to develop failure probabilities for basic events performed during routine operation of the MC&A systems. The failure frequencies for individual events may be used to estimate total system effectiveness using a fault tree in a probabilistic risk analysis (PRA). Numeric risk values are required for the PRA fault tree calculations that are performed to evaluate system effectiveness. So, the performance ratings in the questionnaire must be converted to relative risk values for all of the basic MC&A tasks performed in the facility. If a specific material protection, control, and accountability (MPC&A) task is being performed at the 'perfect' level, the task is considered to have a near zero risk of failure. If the task is performed at a less than perfect level, the deficiency in performance represents some risk of failure for the event. As the degree of deficiency in performance increases, the risk of failure increases. If a task that should be performed is not being performed, that task is in a state of failure. The failure probabilities of all basic events contribute to the total system risk. Conversion of questionnaire MPC&A system performance data to numeric values is a separate function from the process of completing the questionnaire. When specific questions in the questionnaire are answered, the focus is on correctly assessing and reporting, in an adjectival manner, the actual performance of the related MC&A function. Prior to conversion, consideration should not be given to the numeric value that will be assigned during the conversion process. In the conversion process, adjectival responses to questions on system performance are quantified based on a log normal scale typically used in human error analysis (see A.D. Swain and H.E. Guttmann, 'Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications,' NUREG/CR-1278). This conversion produces the basic event risk of failure values required for the fault tree calculations. The fault tree is a deductive logic structure that corresponds to the operational nuclear MC&A system at a nuclear facility. The conventional Delphi process is a time-honored approach commonly used in the risk assessment field to extract numerical values for the failure rates of actions or activities when statistically significant data is absent.

  6. Melting of Uranium Metal Powders with Residual Salts

    SciTech Connect (OSTI)

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

    2007-07-01

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

  7. Thermochemical Conversion | Department of Energy

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

    Conversion Thermochemical Conversion The Bioenergy Technologies Office conducts research on heat-, pressure-, and catalyst-based conversion of various biomass feedstocks to biofuels, chemicals, and power. These conversion processes, most notably fast pyrolysis (as well as other forms of direct liquefaction) and gasification, are described in detail in the links on the left. The Thermochemical Platform aims to efficiently produce biobased fuels and co-products via liquefaction and pyrolysis,

  8. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Deliveries 2010 2011 2012 2013 2014 Purchases of U.S.-origin and foreign-origin uranium 350 550 W W W Weighted-average price 47.13 58.12 W W W Purchases of U.S.-origin and foreign-origin uranium 11,745 14,778 11,545 12,835 17,111 Weighted-average price 44.98 53.29 54.44 50.44 42.90 Purchases 0 0 0 0 0 Weighted-average price -- -- -- -- -- Purchases of U.S.-origin and

  9. Y-12 and uranium history

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

    did happen six days after he was given the assignment. The history of uranium at Y-12 began with that decision, which will be commemorated on September 19, 2012, at...

  10. 2014 Uranium Marketing Annual Report

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

    9 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Annual Cumulative Annual Cumulative 2014 2,494 2,494 - -- 2015 6,014 8,507 3,496 3,496 ...

  11. 2014 Uranium Marketing Annual Report

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

    1 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 thousand pounds U 3 O 8 equivalent U.S.-origin Foreign- origin Total U.S.-origin ...

  12. 2014 Uranium Marketing Annual Report

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

    9 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Deliveries to foreign suppliers and utilities 2010 2011 2012 2013 2014 Foreign sales ...

  13. 2014 Uranium Marketing Annual Report

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

    9 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 thousand pounds U 3 O 8 equivalent; dollars per pound U 3 O 8 equivalent Deliveries ...

  14. Uranium Mobility During In Situ Redox Manipulation of the 100 Areas of the Hanford Site

    SciTech Connect (OSTI)

    Szecsody, James E.; Krupka, Kenneth M.; Williams, Mark D.; Cantrell, Kirk J.; Resch, Charles T.; Fruchter, Jonathan S.

    1998-12-03

    A series of laboratory experiments and computer simulations was conducted to assess the extent of uranium remobilization that is likely to occur at the end of the life cycle of an in situ sediment reduction process. The process is being tested for subsurface remediation of chromate and chlorinated solvent-contaminated sediments at the Hanford Site in southeastern Washington. Uranium species that occur naturally in the +6 valence state [U(VI)] at 10 ppb in groundwater at Hanford will accumulate as U(IV) through the reduction and subsequent precipitation conditions of the permeable barrier created by in situ redox manipulation. The precipitated uranium will be remobilized when the reductive capacity of the barrier is exhausted and the sediment is oxidized by the groundwater containing dissolved oxygen and other oxidants such as chromate. Although U(IV) accumulates from years or decades of reduction/precipitation within the reduced zone, U(VI) concentrations in solution are only somewhat elevated during aquifer oxidation because oxidation and dissolution reactions that release U(IV) precipitate to solution are slow. The release rate of uranium into solution was found to be controlled mainly by the oxidation/dissolution rate of the U(IV) precipitate (half-life 200 hours) and partially by the fast oxidation of adsorbed Fe(II) (halflife 5 hours) and the slow oxidation of Fe(II)CO3 (half-life 120 hours) in the reduced sediment. Simulations of uranium transport that incorporated these and other reactions under site-relevant conditions indicated that 35 ppb U(VI) is the maximum concentration likely to result from mobilization of the precipitated U(IV) species. Experiments also indicated that increasing the contact time between the U(IV) precipitates and the reduced sediment, which is likely to occur in the field, results in a slower U(IV) oxidation rate, which, in turn, would lower the maximum concentration of mobilized U(VI)...

  15. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, B.A.

    1983-06-10

    A method is described for measuring the uranium content of aqueous solutions wherein a uranyl phosphate complex is irradiated with a 5 nanosecond pulse of 425 nanometer laser light and resultant 520 nanometer emissions are observed for a period of 50 to 400 microseconds after the pulse. Plotting the natural logarithm of emission intensity as a function of time yields an intercept value which is proportional to uranium concentration.

  16. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    3 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 2012 2013 2014 Advance Uranium Asset Management Ltd. AREVA NC, Inc. AREVA Enrichment Services, LLC / AREVA NC, Inc. AREVA NC, Inc. CNEIC (China Nuclear Energy Industry Corporation) CNEIC (China Nuclear Energy Industry Corporation) CNEIC (China Nuclear Energy Industry Corporation) LES, LLC (Louisiana Energy Services) LES, LLC (Louisiana Energy Services) LES, LLC (Louisiana Energy Services) NUKEM, Inc.

  17. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, Bruce A. (Kennewick, WA)

    1986-01-01

    A method is described for measuring the uranium content of aqueous solutions wherein a uranyl phosphate complex is irradiated with a 5 nanosecond pulse of 425 nanometer laser light and resultant 520 nanometer emissions are observed for a period of 50 to 400 microseconds after the pulse. Plotting the natural logarithm of emission intensity as a function of time yields an intercept value which is proportional to uranium concentration.

  18. Indium oxide/n-silicon heterojunction solar cells

    DOE Patents [OSTI]

    Feng, Tom (Morris Plains, NJ); Ghosh, Amal K. (New Providence, NJ)

    1982-12-28

    A high photo-conversion efficiency indium oxide/n-silicon heterojunction solar cell is spray deposited from a solution containing indium trichloride. The solar cell exhibits an Air Mass One solar conversion efficiency in excess of about 10%.

  19. Zinc phosphate conversion coatings

    DOE Patents [OSTI]

    Sugama, Toshifumi

    1997-01-01

    Zinc phosphate conversion coatings for producing metals which exhibit enhanced corrosion prevention characteristics are prepared by the addition of a transition-metal-compound promoter comprising a manganese, iron, cobalt, nickel, or copper compound and an electrolyte such as polyacrylic acid, polymethacrylic acid, polyitaconic acid and poly-L-glutamic acid to a phosphating solution. These coatings are further improved by the incorporation of Fe ions. Thermal treatment of zinc phosphate coatings to generate .alpha.-phase anhydrous zinc phosphate improves the corrosion prevention qualities of the resulting coated metal.

  20. Zinc phosphate conversion coatings

    DOE Patents [OSTI]

    Sugama, T.

    1997-02-18

    Zinc phosphate conversion coatings for producing metals which exhibit enhanced corrosion prevention characteristics are prepared by the addition of a transition-metal-compound promoter comprising a manganese, iron, cobalt, nickel, or copper compound and an electrolyte such as polyacrylic acid, polymethacrylic acid, polyitaconic acid and poly-L-glutamic acid to a phosphating solution. These coatings are further improved by the incorporation of Fe ions. Thermal treatment of zinc phosphate coatings to generate {alpha}-phase anhydrous zinc phosphate improves the corrosion prevention qualities of the resulting coated metal. 33 figs.

  1. Atlantic Biomass Conversions Inc | Open Energy Information

    Open Energy Info (EERE)

    Biomass Conversions Inc Jump to: navigation, search Name: Atlantic Biomass Conversions Inc Place: Frederick, Maryland Sector: Biomass Product: Atlantic Biomass Conversions is...

  2. Energy conversion system

    DOE Patents [OSTI]

    Murphy, L.M.

    1985-09-16

    The energy conversion system includes a photo-voltaic array for receiving solar radiation and converting such radiation to electrical energy. The photo-voltaic array is mounted on a stretched membrane that is held by a frame. Tracking means for orienting the photo-voltaic array in predetermined positions that provide optimal exposure to solar radiation cooperate with the frame. An enclosure formed of a radiation transmissible material includes an inside containment space that accommodates the photo-voltaic array on the stretched membrane, the frame and the tracking means, and forms a protective shield for all such components. The enclosure is preferably formed of a flexible inflatable material and maintains its preferred form, such as a dome, under the influence of a low air pressure furnished to the dome. Under this arrangement the energy conversion system is streamlined for minimizing wind resistance, sufficiently weathproof for providing protection against weather hazards such as hail, capable of using diffused light, lightweight for low-cost construction and operational with a minimal power draw.

  3. Energy conversion system

    DOE Patents [OSTI]

    Murphy, Lawrence M.

    1987-01-01

    The energy conversion system includes a photo-voltaic array for receiving solar radiation and converting such radiation to electrical energy. The photo-voltaic array is mounted on a stretched membrane that is held by a frame. Tracking means for orienting the photo-voltaic array in predetermined positions that provide optimal exposure to solar radiation cooperate with the frame. An enclosure formed of a radiation transmissible material includes an inside containment space that accommodates the photo-voltaic array on the stretched membrane, the frame and the tracking means, and forms a protective shield for all such components. The enclosure is preferably formed of a flexible inflatable material and maintains its preferred form, such as a dome, under the influence of a low air pressure furnished to the dome. Under this arrangement the energy conversion system is streamlined for minimizing wind resistance, sufficiently weatherproof for providing protection against weather hazards such as hail, capable of using diffused light, lightweight for low-cost construction, and operational with a minimal power draw.

  4. Uranium Mobility During In Situ Redox Manipulation of the 100 Areas of the Hanford Site

    SciTech Connect (OSTI)

    CT Resch; JE Szecsody; JS Fruchter; KJ Cantrell; KM Krupka; MD Williams

    1998-12-03

    A series of laboratory experiments and computer simulations was conducted to assess the extent of uranium remobilization that is likely to occur at the end of the life cycle of an in situ sediment reduction process. The process is being tested for subsurface remediation of chromate- and chlorinated solvent-contaminated sediments at the Hanford Site in southeastern Washington. Uranium species that occur naturally in the +6 valence state {approximately}(VI) at 10 ppb in groundwater at Hanford will accumulate as U(N) through the reduction and subsequent precipitation conditions of the permeable barrier created by in situ redox manipulation. The precipitated uranium will W remobilized when the reductive capacity of the barrier is exhausted and the sediment is oxidized by the groundwater containing dissolved oxygen and other oxidants such as chromate. Although U(N) accumulates from years or decades of reduction/precipitation within the reduced zone, U(W) concentrations in solution are only somewhat elevated during aquifer oxidation because oxidation and dissolution reactions that release U(N) precipitate to solution are slow. The release rate of uranium into solution was found to be controlled mainly by the oxidation/dissolution rate of the U(IV) precipitate (half-life 200 hours) and partially by the fast oxidation of adsorbed Fe(II) (half- life 5 hours) and the slow oxidation of Fe(II)CO{sub 3} (half-life 120 hours) in the reduced sediment. Simulations of uranium transport that incorporated these and other reactions under site-relevant conditions indicated that 35 ppb U(VI) is the maximum concentration likely to result from mobilization of the precipitated U(IV) species. Experiments also indicated that increasing the contact time between the U(IV) precipitates and the reduced sediment, which is likely to occur in the field, results in a slower U(IV) oxidation rate, which, in turn, would lower the maximum concentration of mobilized U(W). A six-month-long column experiment confirmed that uranium accumulated in reduced sediment was released slowly into solution with U(W) concentrations at only slightly greater than influent U(W) concentrations. This experiment also demonstrated that dissolved chromate, another oxidant likely to be present in some field systems, did not increase the release rate of uranium into solution.

  5. Domestic Uranium Production Report

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

    8. U.S. uranium expenditures, 2003-14 million dollars Year Drilling1 Production2 Land and other 3 Total expenditures Total land and other Land Exploration Reclamation 2003 W W 31.3 NA NA NA W 2004 10.6 27.8 48.4 NA NA NA 86.9 2005 18.1 58.2 59.7 NA NA NA 136.0 2006 40.1 65.9 115.2 41.0 23.3 50.9 221.2 2007 67.5 90.4 178.2 77.7 50.3 50.2 336.2 2008 81.9 221.2 164.4 65.2 50.2 49.1 467.6 2009 35.4 141.0 104.0 17.3 24.2 62.4 280.5 2010 44.6 133.3 99.5 20.2 34.5 44.7 277.3 2011 53.6 168.8 96.8 19.6

  6. Uranium chloride extraction of transuranium elements from LWR fuel

    DOE Patents [OSTI]

    Miller, William E.; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Pierce, R. Dean

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

  7. Uranium chloride extraction of transuranium elements from LWR fuel

    DOE Patents [OSTI]

    Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

    1992-08-25

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure.

  8. Gyroharmonic conversion experiments

    SciTech Connect (OSTI)

    Hirshfield, J.L.; LaPointe, M.A.; Ganguly, A.K. [Omega-P, Inc., New Haven, Connecticut 06520 (United States); LaPointe, M.A. [Yale University, New Haven, Connecticut 06511 (United States)

    1999-05-01

    Generation of high power microwaves has been observed in experiments where a 250{endash}350 kV, 20{endash}30 A electron beam accelerated in a cyclotron autoresonance accelerator (CARA) passes through a cavity tuned gyroharmonic) and at 8.6 GHz (3rd harmonic) will be described. Theory indicates that high conversion efficiency can be obtained for a high quality beam injected into CARA, and when mode competition can be controlled. Comparisons will be made between the experiments and theory. Planned 7th harmonic experiments will also be described, in which phase matching between the TE-72 mode at 20 GHz, and the TE-11 mode at 2.86 GHz, allows efficient 20 GHz co-generation within the CARA waveguide itself. {copyright} {ital 1999 American Institute of Physics.}

  9. Gyroharmonic conversion experiments

    SciTech Connect (OSTI)

    Hirshfield, J. L.; LaPointe, M. A. [Omega-P, Inc., New Haven, Connecticut 06520 (United States); Yale University, New Haven, Connecticut 06511 (United States); Ganguly, A. K. [Omega-P, Inc., New Haven, Connecticut 06520 (United States)

    1999-05-07

    Generation of high power microwaves has been observed in experiments where a 250-350 kV, 20-30 A electron beam accelerated in a cyclotron autoresonance accelerator (CARA) passes through a cavity tuned gyroharmonic) and at 8.6 GHz (3rd harmonic) will be described. Theory indicates that high conversion efficiency can be obtained for a high quality beam injected into CARA, and when mode competition can be controlled. Comparisons will be made between the experiments and theory. Planned 7th harmonic experiments will also be described, in which phase matching between the TE-72 mode at 20 GHz, and the TE-11 mode at 2.86 GHz, allows efficient 20 GHz co-generation within the CARA waveguide itself.

  10. Plutonium (III) and uranium (III) nitrile complexes

    SciTech Connect (OSTI)

    Enriquez, A. E.; Matonic, J. H.; Scott, B. L.; Neu, M. P.

    2002-01-01

    Iodine oxidation of uranium and plutonium metals in tetrahydrofuran and pyridine form AnI{sub 3}(THF){sub 4} and AnI{sub 3}(py){sub 4} (An = Pu, U). These compounds represent convenient entries Into solution An(III) chemistry in organic solvents. Extensions of the actinide metal oxidation methodology in nitrile solvents by I{sub 2}, AgPF{sub 6}, and TIPF{sub 6} are presented here. Treatment of Pu{sup 0} in acetonitrile with iodine yields a putative PuI{sub 3}(NCMe){sub x} intermediate which can be trapped with the tripodal nitrogen donor ligand tpza (tpza = (tris[(2-pyrazinyl)methyl]amine)) and forms the eight-coordinate complex (tpza)PuI{sub 3}(NCMe). Treatment of excess U{sup 0} metal by iodine in acetonitrile afforded a brown crystalline mixed valence complex, [U(NCMe){sub 9}][UI{sub 6}][I], instead of UI{sub 3}(NCMe){sub 4}. The analogous reaction in bezonitrile forms red crystalline UI{sub 4}(NCPh){sub 4}. In contrast, treatment of UI{sub 3}(THF){sub 4} with excess acetonitrile cleanly generates [U(NCMe){sub 9}][I]{sub 3}. Oxidation of Pu{sup 0} by either TI(I) or Ag(I) hexafluorophosphate salts generates a nine-coordinate homoleptic acetonitrile adduct [Pu(NCMe){sub 9}][PF{sub 6}]{sub 3}. Attempts to oxidize U{sub 0} with these salts were unsuccessful.

  11. Progress Report for the grant "Hight-Resolution Mineralogical Charaterization and Biogeochemical Modeling of Uranium Reduction Pathways at the NABIR Field-Research Center"

    SciTech Connect (OSTI)

    Veblen, David R.

    2006-06-15

    We have successfully completed a proof-of-concept, one-year grant on a three-year proposal from the former NABIR program. Using a state-of-the-art 300-kV, atomic resolution, Field Emission Gun Transmission Electron Microscope (TEM), we have successfully identified three categories of mineral hosts for uranium in contaminated soils from the Oak Ridge FRC: (1) iron oxides: (2) mixed manganese-iron oxides; and (3) uranium phosphates.

  12. Conversion Technologies for Advanced Biofuels - Carbohydrates...

    Broader source: Energy.gov (indexed) [DOE]

    More Documents & Publications Conversion Technologies for Advanced Biofuels - Carbohydrates Production Advanced Conversion Roadmap Workshop Innovative Topics for...

  13. Solar Thermoelectric Energy Conversion | Department of Energy

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

    Solar Thermoelectric Energy Conversion Solar Thermoelectric Energy Conversion Efficiencies of different types of solar thermoelectric generators were predicted using theoretical ...

  14. Uranium Biomineralization By Natural Microbial Phosphatase Activities...

    Office of Scientific and Technical Information (OSTI)

    Uranium Biomineralization By Natural Microbial Phosphatase Activities in the Subsurface ... Country of Publication: United States Language: English Subject: 54 ENVIRONMENTAL ...

  15. Uranium Management and Policy | Department of Energy

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

    Uranium Management and Policy Uranium Management and Policy The Paducah Gaseous Diffusion Plant is located 3 miles south of the Ohio River and is 12 miles west of Paducah, Kentucky. Paducah remains the only operating gaseous diffusion uranium enrichment plant in the United States. The Paducah Gaseous Diffusion Plant is located 3 miles south of the Ohio River and is 12 miles west of Paducah, Kentucky. Paducah remains the only operating gaseous diffusion uranium enrichment plant in the United

  16. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    U.S. Energy Information Administration / 2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 thousand pounds U 3 O 8 equivalent 2010 2011 2012 2013 P2014 Owners and operators of U.S. civilian nuclear power reactors inventories 86,527 89,835 97,647 113,077 116,047 Uranium concentrate (U 3 O 8 ) 13,076 14,718 15,963 18,131 20,501 Natural UF 6 35,767 35,883 29,084 38,332 40,972 Enriched UF 6 25,392 19,596 38,428 40,841

  17. Domestic Uranium Production Report - Quarterly

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

    2. Number of uranium mills and plants producing uranium concentrate in the United States Uranium concentrate processing facilities End of Mills - conventional milling 1 Mills - other operations 2 In-situ-leach plants 3 Byproduct recovery plants 4 Total 1996 0 2 5 2 9 1997 0 3 6 2 11 1998 0 2 6 1 9 1999 1 2 4 0 7 2000 1 2 3 0 6 2001 0 1 3 0 4 2002 0 1 2 0 3 2003 0 0 2 0 2 2004 0 0 3 0 3 2005 0 1 3 0 4 2006 0 1 5 0 6 2007 0 1 5 0 6 2008 1 0 6 0 7 2009 0 1 3 0 4 2010 1 0 4 0 5 2011 1 0 5 0 6 2012 1

  18. Continuous reduction of uranium tetrafluoride

    SciTech Connect (OSTI)

    DeMint, A.L.; Maxey, A.W.

    1993-10-21

    Operation of a pilot-scale system for continuous metallothermic reduction of uranium tetrafluoride (UF{sub 4} or green salt) has been initiated. This activity is in support of the development of a cost- effective process to produce uranium-iron (U-Fe) alloy feed for the Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) program. To date, five runs have been made to reduce green salt (UF{sub 4}) with magnesium. During this quarter, three runs were made to perfect the feeding system, examine feed rates, and determine the need for a crust breaker/stirrer. No material was drawn off in any of the runs; both product metal and by-product salt were allowed to accumulate in the reactor.

  19. High strength uranium-tungsten alloys

    DOE Patents [OSTI]

    Dunn, Paul S. (Santa Fe, NM); Sheinberg, Haskell (Los Alamos, NM); Hogan, Billy M. (Los Alamos, NM); Lewis, Homer D. (Bayfield, CO); Dickinson, James M. (Los Alamos, NM)

    1991-01-01

    Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

  20. High strength uranium-tungsten alloy process

    DOE Patents [OSTI]

    Dunn, Paul S. (Santa Fe, NM); Sheinberg, Haskell (Los Alamos, NM); Hogan, Billy M. (Los Alamos, NM); Lewis, Homer D. (Bayfield, CO); Dickinson, James M. (Los Alamos, NM)

    1990-01-01

    Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

  1. Occupational safety data and casualty rates for the uranium fuel cycle. [Glossaries

    SciTech Connect (OSTI)

    O'Donnell, F.R.; Hoy, H.C.

    1981-10-01

    Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10/sup 12/ Btu of energy output, and per other appropriate units of output.

  2. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 State(s) 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 Wyoming 134 139 181 195 245 301 308 348 424 512 531 416 Colorado and Texas 48 140 269 263 557 696 340 292 331 248 198 105 Nebraska and New Mexico 92 102 123 160 149 160 159 134 127 W W W Arizona, Utah, and Washington 47 40 75 120 245 360 273 281 W W W W Alaska, Michigan, Nevada, and South Dakota 0

  3. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 million pounds U 3 O 8 $0 to $30 per pound $0 to $50 per pound $0 to $100 per pound $0 to $30 per pound $0 to $50 per pound $0 to $100 per pound Properties with Exploration Completed, Exploration Continuing, and Only Assessment Work W W 130.7 W W 154.6 Properties Under Development for Production and Development Drilling W 31.8 W W 38.2 W Mines in Production W 19.6 W

  4. 2014 Uranium Market Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    U.S. Energy Information Administration / 2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Purchase contract type (Signed in 2014) Quantity of deliveries received in 2014 Weighted-average price Number of purchase contracts for deliveries in 2014 Spot W W 67 Long-term W W 2 Total 12,263 34.83 69 Table 8. Contracts signed in 2014 by owners and operators of U.S. civilian nuclear power reactors by contract type thousand

  5. METHOD OF PROTECTIVELY COATING URANIUM

    DOE Patents [OSTI]

    Eubank, L.D.; Boller, E.R.

    1959-02-01

    A method is described for protectively coating uranium with zine comprising cleaning the U for coating by pickling in concentrated HNO/sub 3/, dipping the cleaned U into a bath of molten zinc between 430 to 600 C and containing less than 0 01% each of Fe and Pb, and withdrawing and cooling to solidify the coating. The zinccoated uranium may be given a; econd coating with another metal niore resistant to the corrosive influences particularly concerned. A coating of Pb containing small proportions of Ag or Sn, or Al containing small proportions of Si may be applied over the zinc coatings by dipping in molten baths of these metals.

  6. 2014 Domestic Uranium Production Report

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

    10. Uranium reserve estimates at the end of 2013 and 2014" "million pounds U3O8" ,"End of 2013",,,"End of 2014" "Uranium Reserve Estimates1 by Mine and Property Status, Mining Method, and State(s)","Forward Cost 2" ,"$0 to $30 per pound","$0 to $50 per pound","$0 to $100 per pound","$0 to $30 per pound","$0 to $50 per pound","$0 to $100 per pound" "Properties with Exploration

  7. Domestic Uranium Production Report - Quarterly

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

    3. U.S. uranium mills and heap leach facilities by owner, location, capacity, and operating status Operating status at the end of Owner Mill and Heap Leach1 Facility name County, state (existing and planned locations) Capacity (short tons of ore per day) 2014 1st quarter 2015 2nd quarter 2015 3rd quarter 2015 4th Quarter 2015 Anfield Resources Shootaring Canyon Uranium Mill Garfield, Utah 750 Standby Standby Standby Standby Standby EFR White Mesa LLC White Mesa Mill San Juan, Utah 2,000

  8. 230Th-234U Age-Dating Uranium by Mass Spectrometry

    SciTech Connect (OSTI)

    Williams, R W; Gaffney, A M

    2012-04-18

    This is the standard operating procedure used by the Isotope Ratio Mass Spectrometry Group of the Chemical Sciences Division at LLNL for the preparation of a sample of uranium oxide or uranium metal for {sup 230}Th-{sup 234}U age-dating. The method described here includes the dissolution of a sample of uranium oxide or uranium metal, preparation of a secondary dilution, spiking of separate aliquots for uranium and thorium isotope dilution measurements, and purification of uranium and thorium aliquots for mass spectrometry. This SOP may be applied to uranium samples of unknown purity as in a nuclear forensic investigation, and also to well-characterized samples such as, for example, U{sub 3}O{sub 8} and U-metal certified reference materials. The sample of uranium is transferred to a quartz or PFA vial, concentrated nitric acid is added and the sample is heated on a hotplate at approximately 100 C for several hours until it dissolves. The sample solution is diluted with water to make the solution approximately 4 M HNO{sub 3} and hydrofluoric acid is added to make it 0.05 M HF. A secondary dilution of the primary uranium solution is prepared. Separate aliquots for uranium and thorium isotope dilution measurements are taken and spiked with {sup 233}U and {sup 229}Th, respectively. The spiked aliquot for uranium isotope dilution analysis is purified using EiChrom UTEVA resin. The spiked aliquot for thorium isotope dilution analysis is purified by, first, a 1.8 mL AG1x8 resin bed in 9 M HCl on which U adsorbs and Th passes through; second, adsorbing Th on a 1 mL AG1x8 resin bed in 8 M HNO{sub 3} and then eluting it with 9 M HCl followed by 0.1 M HCl + 0.005 M HF; and third, by passing the Th through a final 1.0 mL AG1x8 resin bed in 9 M HCl. The mass spectrometry is performed using the procedure 'Th and U Mass Spectrometry for {sup 230}Th-{sup 234}U Age Dating'.

  9. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, Vladimir (Skokie, IL)

    1986-01-01

    This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

  10. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, V.

    1984-06-13

    This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

  11. NEPTUNIUM OXIDE PROCESSING

    SciTech Connect (OSTI)

    Jordan, J; Watkins, R; Hensel, S

    2009-05-27

    The Savannah River Site's HB-Line Facility completed a campaign in which fifty nine cans of neptunium oxide were produced and shipped to the Idaho National Laboratory in the 9975 shipping container. The neptunium campaign was divided into two parts: Part 1 which consisted of oxide made from H-Canyon neptunium solution which did not require any processing prior to conversion into an oxide, and Part 2 which consisted of oxide made from additional H-Canyon neptunium solutions which required processing to purify the solution prior to conversion into an oxide. The neptunium was received as a nitrate solution and converted to oxide through ion-exchange column extraction, precipitation, and calcination. Numerous processing challenges were encountered in order make a final neptunium oxide product that could be shipped in a 9975 shipping container. Among the challenges overcome was the issue of scale: translating lab scale production into full facility production. The balance between processing efficiency and product quality assurance was addressed during this campaign. Lessons learned from these challenges are applicable to other processing projects.

  12. PDB to AMPL Conversion

    Energy Science and Technology Software Center (OSTI)

    2002-09-01

    PDB to AMPL Conversion was written to convert protein data base files to AMPL files. The protein data bases on the internet contain a wealth of information about the structue and makeup of proteins. Each file contains information derived by one or more experiments and contains information on how the experiment waw performed, the amino acid building blocks of each chain, and often the three-dimensional structure of the protein extracted from the experiments. The waymore » a protein folds determines much about its function. Thus, studying the three-dimensional structure of the protein is of great interest. Analysing the contact maps is one way to examine the structure. A contact map is a graph which has a linear back bone of amino acids for nodes (i.e., adjacent amino acids are always connected) and vertices between non-adjacent nodes if they are close enough to be considered in contact. If the graphs are similar then the folds of the protein and their function should also be similar. This software extracts the contact maps from a protein data base file and puts in into AMPL data format. This format is designed for use in AMPL, a programming language for simplifying linear programming formulations.« less

  13. Static Scale Conversion (SSC)

    Energy Science and Technology Software Center (OSTI)

    2007-01-19

    The Static Scale Conversion (SSC) software is a unique enhancement to the AIMVEE system. It enables a SSC to weigh and measure vehicles and cargo dynamically (i.e., as they pass over the large scale. Included in the software is the AIMVEE computer code base. The SSC and AIMVEE computer system electronically continue to retrieve deployment information, identify vehicle automatically and determine total weight, individual axle weights, axle spacing and center-of-balance for any wheeled vehicle inmore » motion. The AIMVEE computer code system can also perform these functions statically for both wheel vehicles and cargo with information. The AIMVEE computer code system incorporates digital images and applies cubing algorithms to determine length, width, height for cubic dimensions of both vehicle and cargo. Once all this information is stored, it electronically links to data collection and dissemination systems to provide “actual” weight and measurement information for planning, deployment, and in-transit visibility.« less

  14. Y-12 fulfills major milestone in fuel conversion commitment for Jamaican

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

    research reactor | Y-12 National Security Complex fulfills major ... Y-12 fulfills major milestone in fuel conversion commitment for Jamaican research reactor Posted: June 3, 2014 - 4:42pm The Y-12 National Security Complex recently completed its final shipment of uranium dioxide powder that will be used as the feedstock for fueling the Safe LOW-POwer Kritical Experiment (SLOWPOKE) research reactor in Kingston, Jamaica. This project is one of many involving Y-12's expertise at creating

  15. Development of pulsed neutron uranium logging instrument

    SciTech Connect (OSTI)

    Wang, Xin-guang; Liu, Dan; Zhang, Feng

    2015-03-15

    This article introduces a development of pulsed neutron uranium logging instrument. By analyzing the temporal distribution of epithermal neutrons generated from the thermal fission of {sup 235}U, we propose a new method with a uranium-bearing index to calculate the uranium content in the formation. An instrument employing a D-T neutron generator and two epithermal neutron detectors has been developed. The logging response is studied using Monte Carlo simulation and experiments in calibration wells. The simulation and experimental results show that the uranium-bearing index is linearly correlated with the uranium content, and the porosity and thermal neutron lifetime of the formation can be acquired simultaneously.

  16. Process for alloying uranium and niobium

    DOE Patents [OSTI]

    Holcombe, Cressie E. (Farragut, TN); Northcutt, Jr., Walter G. (Oak Ridge, TN); Masters, David R. (Knoxville, TN); Chapman, Lloyd R. (Knoxville, TN)

    1991-01-01

    Alloys such as U-6Nb are prepared by forming a stacked sandwich array of uraniun sheets and niobium powder disposed in layers between the sheets, heating the array in a vacuum induction melting furnace to a temperature such as to melt the uranium, holding the resulting mixture at a temperature above the melting point of uranium until the niobium dissolves in the uranium, and casting the uranium-niobium solution. Compositional uniformity in the alloy product is enabled by use of the sandwich structure of uranium sheets and niobium powder.

  17. Domestic Uranium Production Report - Energy Information Administration

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report - Annual With Data for 2014 | Release Date: April 30, 2015 | Next Release Date: May 2016 | full report Previous domestic uranium production reports Year: 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 Go Drilling Figure 1. U.S. Uranium drilling by number of holes, 2004-14 Total uranium drilling was 1,752 holes covering 1.3 million feet, 67% fewer holes than in 2013 and the lowest since 2004. Expenditures for uranium drilling in the United States were $28

  18. Characterization of options and their analysis requirements for the long-term management of depleted uranium hexafluoride

    SciTech Connect (OSTI)

    Dubrin, J.W.; Rosen, R.S.; Zoller, J.N.; Harri, J.W.; Schwertz, N.L.

    1995-12-01

    The Department of Energy (DOE) is examining alternative strategies for the long-term management of depleted uranium hexafluoride (UF{sub 6}) currently stored at the gaseous diffusion plants at Portsmouth, Ohio, and Paducah, Kentucky, and on the Oak Ridge Reservation in Oak Ridge, Tennessee. This paper describes the methodology for the comprehensive and ongoing technical analysis of the options being considered. An overview of these options, along with several of the suboptions being considered, is presented. The long-term management strategy alternatives fall into three broad categories: use, storage, or disposal. Conversion of the depleted UF6 to another form such as oxide or metal is needed to implement most of these alternatives. Likewise, transportation of materials is an integral part of constructing the complete pathway between the current storage condition and ultimate disposition. The analysis of options includes development of pre-conceptual designs; estimates of effluents, wastes, and emissions; specification of resource requirements; and preliminary hazards assessments. The results of this analysis will assist DOE in selecting a strategy by providing the engineering information necessary to evaluate the environmental impacts and costs of implementing the management strategy alternatives.

  19. 2014 Uranium Marketing Annual Report

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

    7 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Year of delivery Minimum Maximum 2015 2,838 2,838 2016 3,573 3,573 2017 2,718 2,818 ...

  20. Neutronic Analyses for HEU to LEU fuel conversion of the Massachusetts Institute of Technology.

    SciTech Connect (OSTI)

    Wilson, E. H.; Newton, T. H.; Bergeron, A.; Horelik, N.; Stevens, J. G

    2011-03-02

    The Massachusetts Institute of Technology (MIT) reactor (MITR-II), based in Cambridge, Massachusetts, is a research reactor designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the MITR-II. This report presents the results of steady state neutronic safety analyses for conversion of MITR-II from the use of HEU fuel to the use of U-Mo LEU fuel. The objective of this work was to demonstrate that the safety analyses meet current requirements for an LEU core replacement of MITR-II.

  1. Alternative Fuels Data Center: Vehicle Conversions

    Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

    Conversions Printable Version Share this resource Send a link to Alternative Fuels Data Center: Vehicle Conversions to someone by E-mail Share Alternative Fuels Data Center: Vehicle Conversions on Facebook Tweet about Alternative Fuels Data Center: Vehicle Conversions on Twitter Bookmark Alternative Fuels Data Center: Vehicle Conversions on Google Bookmark Alternative Fuels Data Center: Vehicle Conversions on Delicious Rank Alternative Fuels Data Center: Vehicle Conversions on Digg Find More

  2. Alternative Fuels Data Center: Conversion Regulations

    Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

    Conversion Regulations to someone by E-mail Share Alternative Fuels Data Center: Conversion Regulations on Facebook Tweet about Alternative Fuels Data Center: Conversion Regulations on Twitter Bookmark Alternative Fuels Data Center: Conversion Regulations on Google Bookmark Alternative Fuels Data Center: Conversion Regulations on Delicious Rank Alternative Fuels Data Center: Conversion Regulations on Digg Find More places to share Alternative Fuels Data Center: Conversion Regulations on

  3. Chemical and Radiochemical Composition of Thermally Stabilized Plutonium Oxide from the Plutonium Finishing Plant Considered as Alternate Feedstock for the Mixed Oxide Fuel Fabrication Facility

    SciTech Connect (OSTI)

    Tingey, Joel M.; Jones, Susan A.

    2005-07-01

    Eighteen plutonium oxide samples originating from the Plutonium Finishing Plant (PFP) on the Hanford Site were analyzed to provide additional data on the suitability of PFP thermally stabilized plutonium oxides and Rocky Flats oxides as alternate feedstock to the Mixed Oxide Fuel Fabrication Facility (MFFF). Radiochemical and chemical analyses were performed on fusions, acid leaches, and water leaches of these 18 samples. The results from these destructive analyses were compared with nondestructive analyses (NDA) performed at PFP and the acceptance criteria for the alternate feedstock. The plutonium oxide materials considered as alternate feedstock at Hanford originated from several different sources including Rocky Flats oxide, scrap from the Remote Mechanical C-Line (RMC) and the Plutonium Reclamation Facility (PRF), and materials from other plutonium conversion processes at Hanford. These materials were received at PFP as metals, oxides, and solutions. All of the material considered as alternate feedstock was converted to PuO2 and thermally stabilized by heating the PuO2 powder at 950 C in an oxidizing environment. The two samples from solutions were converted to PuO2 by precipitation with Mg(OH)2. The 18 plutonium oxide samples were grouped into four categories based on their origin. The Rocky Flats oxide was divided into two categories, low- and high-chloride Rocky Flats oxides. The other two categories were PRF/RMC scrap oxides, which included scrap from both process lines and oxides produced from solutions. The two solution samples came from samples that were being tested at Pacific Northwest National Laboratory because all of the plutonium oxide from solutions at PFP had already been processed and placed in 3013 containers. These samples originated at the PFP and are from plutonium nitrate product and double-pass filtrate solutions after they had been thermally stabilized. The other 16 samples originated from thermal stabilization batches before canning at PFP. Samples varied in appearance depending on the original source of material. Rocky Flats items were mostly dark olive green with clumps that crushed easily with a mortar and pestle. PRF/RMC items showed more variability. These items were mostly rust colored. One sample contained white particles that were difficult to crush, and another sample was a dark grey with a mixture of fines and large, hard fragments. The appearance and feel of the fragments indicated they might be an alloy. The color of the solution samples was indicative of the impurities in the sample. The double-pass filtrate solution was a brown color indicative of the iron impurities in the sample. The other solution sample was light gray in color. Radiochemical analyses, including thermal ionization mass spectrometry (TIMS), alpha and gamma energy analysis (AEA and GEA), and kinetic phosphorescence analysis (KPA), indicate that these materials are all weapons-grade plutonium with consistent plutonium isotopics. A small amount of uranium (<0.14 wt%) is also present in these samples. The isotopic composition of the uranium varied widely but was consistent among each category of material. The primary water-soluble anions in these samples were Cl-, NO3-, SO42-, and PO43-. The only major anion observed in the Rocky Flats materials was Cl-, but the PRF/RMC samples had significant quantities of all of the primary anions observed. Prompt gamma measurements provide a representative analysis of the Cl- concentration in the bulk material. The primary anions observed in the solution samples were NO3-, and PO43-. The concentration of these anions did not exceed the mixed oxide (MOX) specification limits. Cations that exceeded the MOX specification limits included Cr, Fe, Ni, Al, Cu, and Si. All of the samples exceeded at least the 75% specification limit in one element.

  4. Uranium isotopes fingerprint biotic reduction

    SciTech Connect (OSTI)

    Stylo, Malgorzata; Neubert, Nadja; Wang, Yuheng; Monga, Nikhil; Romaniello, Stephen J.; Weyer, Stefan; Bernier-Latmani, Rizlan

    2015-04-20

    Knowledge of paleo-redox conditions in the Earths history provides a window into events that shaped the evolution of life on our planet. The role of microbial activity in paleo-redox processes remains unexplored due to the inability to discriminate biotic from abiotic redox transformations in the rock record. The ability to deconvolute these two processes would provide a means to identify environmental niches in which microbial activity was prevalent at a specific time in paleo-history and to correlate specific biogeochemical events with the corresponding microbial metabolism. Here, we demonstrate that the isotopic signature associated with microbial reduction of hexavalent uranium (U), i.e., the accumulation of the heavy isotope in the U(IV) phase, is readily distinguishable from that generated by abiotic uranium reduction in laboratory experiments. Thus, isotope signatures preserved in the geologic record through the reductive precipitation of uranium may provide the sought-after tool to probe for biotic processes. Because uranium is a common element in the Earths crust and a wide variety of metabolic groups of microorganisms catalyze the biological reduction of U(VI), this tool is applicable to a multiplicity of geological epochs and terrestrial environments. The findings of this study indicate that biological activity contributed to the formation of many authigenic U deposits, including sandstone U deposits of various ages, as well as modern, Cretaceous, and Archean black shales. In addition, engineered bioremediation activities also exhibit a biotic signature, suggesting that, although multiple pathways may be involved in the reduction, direct enzymatic reduction contributes substantially to the immobilization of uranium.

  5. Uranium isotopes fingerprint biotic reduction

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

    Stylo, Malgorzata; Neubert, Nadja; Wang, Yuheng; Monga, Nikhil; Romaniello, Stephen J.; Weyer, Stefan; Bernier-Latmani, Rizlan

    2015-04-20

    Knowledge of paleo-redox conditions in the Earth’s history provides a window into events that shaped the evolution of life on our planet. The role of microbial activity in paleo-redox processes remains unexplored due to the inability to discriminate biotic from abiotic redox transformations in the rock record. The ability to deconvolute these two processes would provide a means to identify environmental niches in which microbial activity was prevalent at a specific time in paleo-history and to correlate specific biogeochemical events with the corresponding microbial metabolism. Here, we demonstrate that the isotopic signature associated with microbial reduction of hexavalent uranium (U),more » i.e., the accumulation of the heavy isotope in the U(IV) phase, is readily distinguishable from that generated by abiotic uranium reduction in laboratory experiments. Thus, isotope signatures preserved in the geologic record through the reductive precipitation of uranium may provide the sought-after tool to probe for biotic processes. Because uranium is a common element in the Earth’s crust and a wide variety of metabolic groups of microorganisms catalyze the biological reduction of U(VI), this tool is applicable to a multiplicity of geological epochs and terrestrial environments. The findings of this study indicate that biological activity contributed to the formation of many authigenic U deposits, including sandstone U deposits of various ages, as well as modern, Cretaceous, and Archean black shales. In addition, engineered bioremediation activities also exhibit a biotic signature, suggesting that, although multiple pathways may be involved in the reduction, direct enzymatic reduction contributes substantially to the immobilization of uranium.« less

  6. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Goldmann, Andrew S.; Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  7. HIGHLY ENRICHED URANIUM BLEND DOWN PROGRAM AT THE SAVANNAH RIVER SITE PRESENT AND FUTURE

    SciTech Connect (OSTI)

    Magoulas, V; Charles Goergen, C; Ronald Oprea, R

    2008-06-05

    The Department of Energy (DOE) and Tennessee Valley Authority (TVA) entered into an Interagency Agreement to transfer approximately 40 metric tons of highly enriched uranium (HEU) to TVA for conversion to fuel for the Browns Ferry Nuclear Power Plant. Savannah River Site (SRS) inventories included a significant amount of this material, which resulted from processing spent fuel and surplus materials. The HEU is blended with natural uranium (NU) to low enriched uranium (LEU) with a 4.95% 235U isotopic content and shipped as solution to the TVA vendor. The HEU Blend Down Project provided the upgrades needed to achieve the product throughput and purity required and provided loading facilities. The first blending to low enriched uranium (LEU) took place in March 2003 with the initial shipment to the TVA vendor in July 2003. The SRS Shipments have continued on a regular schedule without any major issues for the past 5 years and are due to complete in September 2008. The HEU Blend program is now looking to continue its success by dispositioning an additional approximately 21 MTU of HEU material as part of the SRS Enriched Uranium Disposition Project.

  8. The Sorption/Desorption Behavior of Uranium in Transport Studies Using Yucca Mountain Alluvium

    SciTech Connect (OSTI)

    C. D. Scism

    2006-02-15

    Yucca Mountain, Nevada is the proposed site of a geologic repository for the disposal of spent nuclear fuel and high-level radioactive waste in the United States. In the event repository engineered barriers fail, the saturated alluvium located south of Yucca Mountain is expected to serve as a natural barrier to the migration of radionuclides to the accessible environment. The purpose of this study is to improve the characterization of uranium retardation in the saturated zone at Yucca Mountain to support refinement of an assessment model. The distribution of uranium desorption rates from alluvium obtained from Nye County bore holes EWDP-19IM1, EWDP-10SA, EWDP-22SA were studied to address inconsistencies between results from batch sorption and column transport experiments. The alluvium and groundwater were characterized to better understand the underlying mechanisms of the observed behavior. Desorption rate constants were obtained using an activity based mass balance equation and column desorption experiments were analyzed using a mathematical model utilizing multiple sorption sites with different first-order forward and reverse reaction rates. The uranium desorption rate constants decreased over time, suggesting that the alluvium has multiple types of active sorption sites with different affinities for uranium. While a significant fraction of the initially sorbed uranium desorbed from the alluvium quite rapidly, a roughly equivalent amount remained sorbed after several months of testing. The information obtained through this research suggests that uranium may experience greater effective retardation in the alluvium than simple batch sorption experiments would suggest. Electron Probe Microanalysis shows that uranium is associated with both clay minerals and iron oxides after sorption to alluvial material. These results provide further evidence that the alluvium contains multiple sorption sites for uranium.

  9. Advanced Conversion Roadmap Workshop | Department of Energy

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

    Advanced Conversion Roadmap Workshop Advanced Conversion Roadmap Workshop DOE introduction slides to the Advanced Conversion Roadmap Workshop webinar. PDF icon ctab_webinar_doe.pdf More Documents & Publications Conversion Technologies for Advanced Biofuels - Bio-Oil Production Conversion Technologies for Advanced Biofuels - Bio-Oil Upgrading Technology Pathway Selection Effort

  10. Prospects for the recovery of uranium from seawater

    SciTech Connect (OSTI)

    Best, F.R.; Driscoll, M.

    1986-04-01

    A computer program entitled URPE (Uranium Recovery Performance and Economics) has been developed to simulate the engineering performance and provide an economic analysis of a plant recovering uranium from seawater. The conceptual system design used as the focal point for the more general analysis consists of a floating oil-rig type of platform single-point moored in an open ocean current, using either high-volume-low-head axial pumps or the velocity head of the ambient ocean current to force seawater through a mass transfer medium (hydrous titanium oxide (HTO) coated onto particle beds or stacked tubes). Uranium is recovered from the seawater by an adsorption process, and later eluted from the adsober by an ammonium carbonate solution. A multiproduct cogenerating plant on board the platform burns coal to raise steam for electricity generation, desalination, and process heat requirements. Scrubbed stack gas from the plant is processed to recover carbon dioxide for chemical make-up needs. The equilibrium isotherm and the diffusion constant for the uranyl-HTO system, which are needed for bed performance calculations, have been calculated based on the data reported in the literature. In addition, a technique for calculating the rate constant of a fixed-bed adsoorbing system has been developed for use with Thomas' solution for predicting fixed-bed performance.

  11. Inherently safe in situ uranium recovery.

    SciTech Connect (OSTI)

    Krumhansl, James Lee; Beauheim, Richard Louis; Brady, Patrick Vane; Arnold, Bill Walter; Kanney, Joseph F.; McKenna, Sean Andrew

    2009-05-01

    Expansion of uranium mining in the United States is a concern to some environmental groups and sovereign Native American Nations. An approach which may alleviate some problems is to develop inherently safe in situ uranium recovery ('ISR') technologies. Current ISR technology relies on chemical extraction of trace levels of uranium from aquifers that, once mined, can still contain dissolved uranium and other trace metals that are a health concern. Existing ISR operations are few in number; however, high uranium prices are driving the industry to consider expanding operations nation-wide. Environmental concerns and enforcement of the new 30 ppb uranium drinking water standard may make opening new mining operations more difficult and costly. Here we propose a technological fix: the development of inherently safe in situ recovery (ISISR) methods. The four central features of an ISISR approach are: (1) New 'green' leachants that break down predictably in the subsurface, leaving uranium, and associated trace metals, in an immobile form; (2) Post-leachant uranium/metals-immobilizing washes that provide a backup decontamination process; (3) An optimized well-field design that increases uranium recovery efficiency and minimizes excursions of contaminated water; and (4) A combined hydrologic/geochemical protocol for designing low-cost post-extraction long-term monitoring. ISISR would bring larger amounts of uranium to the surface, leave fewer toxic metals in the aquifer, and cost less to monitor safely - thus providing a 'win-win-win' solution to all stakeholders.

  12. Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance

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

    Agreement | Department of Energy Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act (TSCA) Uranium Enrichment Federal Facility Compliance Agreement establishes a plan to bring DOE's Uranium Enrichment Plants (and support facilities) located in Portsmouth, Ohio and Paducah, Kentucky and DOE's former Uranium Enrichment Plant (and support

  13. Y-12 and the super enriched Uranium 235Ž

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

    "super enriched Uranium 235" Ken Bernander called me to say that he had read in the newspaper about the 100 milligrams of uranium oxide that is 99.999% U-235. He was chuckling when he called, so I knew he was going to tell me something good. Ken calls me when he reads my columns in The Oak Ridger and often gives me insights I would not otherwise get. So, getting a call from Ken is a delight. This time he wanted me to talk to Bill Harmon. It seems that Bill had told Ken that he recalled

  14. Catalytic oxidation of light alkanes in presence of a base

    DOE Patents [OSTI]

    Bhinde, Manoj V.; Bierl, Thomas W.

    1998-01-01

    The presence of a base in the reaction mixture in a metal-ligand catalyzed partial oxidation of alkanes results in sustained catalyst activity, and in greater percent conversion as compared with oxidation in the absence of base, while maintaining satisfactory selectivity for the desired oxidation, for example the oxidation of isobutane to isobutanol.

  15. Catalytic oxidation of light alkanes in presence of a base

    DOE Patents [OSTI]

    Bhinde, M.V.; Bierl, T.W.

    1998-03-03

    The presence of a base in the reaction mixture in a metal-ligand catalyzed partial oxidation of alkanes results in sustained catalyst activity, and in greater percent conversion as compared with oxidation in the absence of base, while maintaining satisfactory selectivity for the desired oxidation, for example the oxidation of isobutane to isobutanol. 1 fig.

  16. Enzymes for improved biomass conversion

    DOE Patents [OSTI]

    Brunecky, Roman; Himmel, Michael E.

    2016-02-02

    Disclosed herein are enzymes and combinations of the enzymes useful for the hydrolysis of cellulose and the conversion of biomass. Methods of degrading cellulose and biomass using enzymes and cocktails of enzymes are also disclosed.

  17. Start Your Energy Conversion Devices

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

    Start Your Energy Conversion Devices "Start Your Energy Conversion Devices" Sunrayce 97 Begins Day 1 - Indianapolis to Terre Haute For more information contact: Patrick Booher, Program Manager Sunrayce 97 586-9275 Cathy Short 586-9302 Terre Haute, Ind., June 19, 1997 -- Manta GT, the Massachusetts Institute of Technology's entry in Sunrayce 97 made the quickest trip Thursday from the Indianapolis Motor Speedway to Terre Haute in the first leg North America's largest solar car event.

  18. EPA Redesigns Conversion Certification Policies

    Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

    EPA Redesigns Conversion Certification Policies At a recent meeting held in Washington, DC, officials from the U.S. Environmental Protection Agency (EPA) opened dialogue about proposed changes to its emission certification policies that affect alternative fuel vehicles (AFVs). "We are trying to accommo- date the Energy Policy Act (EPAct) and Executive Order requirements while trying to change enforce- ment policies and guidance with respect to conversions," said Rich Ackerman of EPA's

  19. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 Number of Holes Feet (thousand) Number of Holes Feet (thousand) Number of Holes Feet (thousand) 2003 NA NA NA NA W W 2004 W W W W 2,185 1,249 2005 W W W W 3,143 1,668 2006 1,473 821 3,430 1,892 4,903 2,713 2007 4,351 2,200 4,996 2,946 9,347 5,146 2008 5,198 2,543 4,157 2,551 9,355 5,093 2009 1,790 1,051 3,889 2,691 5,679 3,742 2010 2,439 1,460 4,770 3,444 7,209

  20. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 Activity at U.S. Mills and In-Situ-Leach Plants 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 Ore from Underground Mines and Stockpiles Fed to Mills 1 0 W W W 0 W W W W W W W Other Feed Materials 2 W W W W W W W W W W W W Total Mill Feed W W W W W W W W W W W W (thousand pounds U 3 O 8 ) W W W W W W W W W W W W (thousand pounds U 3 O 8 ) W W W W W W W

  1. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    U.S. Energy Information Administration / 2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 thousand separative work units (SWU) Country of enrichment service (SWU-origin) 2010 2011 2012 2013 2014 China 0 W W W 636 France W W 0 0 0 Germany 681 1,539 1,075 753 1,005 Netherlands 2,292 1,506 1,496 2,112 1,801 Russia 5,055 5,308 6,560 2,491 3,083 United Kingdom 2,119 2,813 2,648 2,674 2,435 Europe 1 W 670 W 0 W Other 2 W

  2. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    8 U.S. Energy Information Administration / 2014 Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Deliveries 2010 2011 2012 2013 2014 Purchases 2,226 1,668 1,194 W 410 Weighted-average price 43.36 54.85 51.78 W 33.55 Purchases 27,186 24,695 24,606 W 28,743 Weighted-average price 41.42 49.69 47.75 W 38.42 Purchases 29,412 26,363 25,800 30,191 29,153 Weighted-average price 41.57 50.02 47.94 42.95 38.35 Purchases 24,693

  3. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Purchases Weighted- average price Purchases Weighted- average price Purchases Weighted- average price Purchases Weighted- average price Purchases Weighted- average price Australia 7,112 51.35 6,001 57.47 6,724 51.17 10,741 49.92 10,511 48.03 Brazil W W W W W W W W W W Canada 10,238 50.35 10,832 56.08 13,584 56.75 7,808 52.61 9,789 45.87 China 0 -- W W W W W W W W Czech

  4. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Marketing Annual Report 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Number of purchasers Quantity with reported price Weighted- average price Number of purchasers Quantity with reported price Weighted- average price Number of purchasers Quantity with reported price Weighted- average price First 8 10,981 45.58 8 12,328 42.01 8 11,681 37.64 Second 7 11,659 53.03 8 13,143 49.94 7 8,493 42.68 Third 7 21,146 57.22 7 18,057 53.43 7 21,805 48.04

  5. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    7 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 Capacity (short tons of ore per day) 2010 2011 2012 2013 2014 EFR White Mesa LLC White Mesa Mill San Juan, Utah 2,000 Operating Operating Operating Operating-Processing Alternate Feed Operating-Processing Alternate Feed Energy Fuels Resources Corp Pinon Ridge Mill Montrose, Colorado 500 Developing Permitted And Licensed Partially Permitted And Licensed Permitted And Licensed Permitted And Licensed

  6. uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    uranium | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog Home

  7. :- : DRILLING URANIUM BILLETS ON A

    Office of Legacy Management (LM)

    'Xxy";^ ...... ' '. .- -- Metals, Ceramics, and Materials. : . - ,.. ; - . _ : , , ' z . , -, .- . >. ; . .. :- : DRILLING URANIUM BILLETS ON A .-... r .. .. i ' LEBLOND-CARLSTEDT RAPID BORER 4 r . _.i'- ' ...... ' -'".. :-'' ,' :... : , '.- ' ;BY R.' J. ' ANSEN .AEC RESEARCH AND DEVELOPMENT REPORT PERSONAL PROPERTY OF J. F. Schlltz .:- DECLASSIFIED - PER AUTHORITY OF (DAlE) (NhTI L (DATE)UE) FEED MATERIALS PRODUCTION CENTER NATIONAL LFE A COMPANY OF OHIO 26 1 3967 3035406 NLCO -

  8. Alternative Fuels Data Center: Propane Vehicle Conversions

    Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

    Conversions to someone by E-mail Share Alternative Fuels Data Center: Propane Vehicle Conversions on Facebook Tweet about Alternative Fuels Data Center: Propane Vehicle Conversions on Twitter Bookmark Alternative Fuels Data Center: Propane Vehicle Conversions on Google Bookmark Alternative Fuels Data Center: Propane Vehicle Conversions on Delicious Rank Alternative Fuels Data Center: Propane Vehicle Conversions on Digg Find More places to share Alternative Fuels Data Center: Propane Vehicle

  9. Alternative Fuels Data Center: Vehicle Conversion Basics

    Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

    Vehicle Conversion Basics to someone by E-mail Share Alternative Fuels Data Center: Vehicle Conversion Basics on Facebook Tweet about Alternative Fuels Data Center: Vehicle Conversion Basics on Twitter Bookmark Alternative Fuels Data Center: Vehicle Conversion Basics on Google Bookmark Alternative Fuels Data Center: Vehicle Conversion Basics on Delicious Rank Alternative Fuels Data Center: Vehicle Conversion Basics on Digg Find More places to share Alternative Fuels Data Center: Vehicle

  10. Uranium Leasing Program | Department of Energy

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

    » Uranium Leasing Program Uranium Leasing Program Abandoned Mine Reclamation, Uravan Mineral Belt, Colorado Abandoned Mine Reclamation, Uravan Mineral Belt, Colorado LM currently manages the Uranium Leasing Program and continues to administer 31 lease tracts, all located within the Uravan Mineral Belt in southwestern Colorado. Twenty-nine of these lease tracts are actively held under lease and two tracts have been placed in inactive status indefinitely. Administrative duties include ongoing

  11. Conversation with Paul Brown | Open Energy Information

    Open Energy Info (EERE)

    Conversation with Paul Brown Jump to: navigation, search OpenEI Reference LibraryAdd to library Personal Communication: Conversation with Paul Brown Author Paul Brown Recipient...

  12. Biological Conversion of Sugars To Hydrocarbons | Department...

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

    To Hydrocarbons Biological Conversion of Sugars To Hydrocarbons PDF explaining the biological process of bioenergy PDF icon Biological Conversion of Sugars To Hydrocarbons More...

  13. Nanostructured High Temperature Bulk Thermoelectric Energy Conversion...

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

    High Temperature Bulk Thermoelectric Energy Conversion for Efficient Waste Heat Recovery Nanostructured High Temperature Bulk Thermoelectric Energy Conversion for Efficient Waste ...

  14. Biochemical Conversion: Using Hydrolysis, Fermentation, and Catalysis...

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

    Biochemical Conversion: Using Hydrolysis, Fermentation, and Catalysis to Make Fuels and Chemicals Biochemical Conversion: Using Hydrolysis, Fermentation, and Catalysis to Make...

  15. Consent Order, Uranium Disposition Services, LLC - NCO-2010-01...

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

    Uranium Disposition Services, LLC - NCO-2010-01 Consent Order, Uranium Disposition Services, LLC - NCO-2010-01 March 26, 2010 Issued to Uranium Disposition Services, LLC related to ...

  16. Uranium metal reactions with hydrogen and water vapour and the reactivity of the uranium hydride produced

    SciTech Connect (OSTI)

    Godfrey, H.; Broan, C.; Goddard, D.; Hodge, N.; Woodhouse, G.; Diggle, A.; Orr, R.

    2013-07-01

    Within the nuclear industry, metallic uranium has been used as a fuel. If this metal is stored in a hydrogen rich environment then the uranium metal can react with the hydrogen to form uranium hydride which can be pyrophoric when exposed to air. The UK National Nuclear Laboratory has been carrying out a programme of research for Sellafield Limited to investigate the conditions required for the formation and persistence of uranium hydride and the reactivity of the material formed. The experimental results presented here have described new results characterising uranium hydride formed from bulk uranium at 50 and 160 C. degrees and measurements of the hydrolysis kinetics of these materials in liquid water. It has been shown that there is an increase in the proportion of alpha-uranium hydride in material formed at lower temperatures and that there is an increase in the rate of reaction with water of uranium hydride formed at lower temperatures. This may at least in part be attributable to a difference in the reaction rate between alpha and beta-uranium hydride. A striking observation is the strong dependence of the hydrolysis reaction rate on the temperature of preparation of the uranium hydride. For example, the reaction rate of uranium hydride prepared at 50 C. degrees was over ten times higher than that prepared at 160 C. degrees at 20% extent of reaction. The decrease in reaction rate with the extent of reaction also depended on the temperature of uranium hydride preparation.

  17. Method for direct conversion of gaseous hydrocarbons to liquids

    DOE Patents [OSTI]

    Kong, Peter C.; Lessing, Paul A.

    2006-03-07

    A chemical reactor for direct conversion of hydrocarbons includes a dielectric barrier discharge plasma cell and a solid oxide electrochemical cell in fluid communication therewith. The discharge plasma cell comprises a pair of electrodes separated by a dielectric material and passageway therebetween. The electrochemical cell comprises a mixed-conducting solid oxide electrolyte membrane tube positioned between a porous cathode and a porous anode, and a gas inlet tube for feeding oxygen containing gas to the porous cathode. An inlet is provided for feeding hydrocarbons to the passageway of the discharge plasma cell, and an outlet is provided for discharging reaction products from the reactor. A packed bed catalyst may optionally be used in the reactor to increase efficiency of conversion. The reactor can be modified to allow use of a light source for directing ultraviolet light into the discharge plasma cell and the electrochemical cell.

  18. Method for regeneration and activity improvement of syngas conversion catalyst

    DOE Patents [OSTI]

    Lucki, Stanley J. (Runnemede, NJ); Brennan, James A. (Cherry Hill, NJ)

    1980-01-01

    A method is disclosed for the treatment of single particle iron-containing syngas (synthes.s gas) conversion catalysts comprising iron, a crystalline acidic aluminosilicate zeolite having a silica to alumina ratio of at least 12, a pore size greater than about 5 Angstrom units and a constraint index of about 1-12 and a matrix. The catalyst does not contain promoters and the treatment is applicable to either the regeneration of said spent single particle iron-containing catalyst or for the initial activation of fresh catalyst. The treatment involves air oxidation, hydrogen reduction, followed by a second air oxidation and contact of the iron-containing single particle catalyst with syngas prior to its use for the catalytic conversion of said syngas. The single particle iron-containing catalysts are prepared from a water insoluble organic iron compound.

  19. Potentiometric determination of uranium in organic extracts

    SciTech Connect (OSTI)

    Bodnar, L.Z.

    1980-05-01

    The potentimetric determination of uranium in organic extracts was studied. A mixture of 30% TBP, (tributylphosphate), in carbon tetrachloride was used, with the NBL (New Brunswick Laboratory) titrimetric procedure. Results include a comparative analysis performed on organic extracts of fissium alloys vs those performed on aqueous samples of the same alloys which had been treated to remove interfering elements. Also comparative analyses were performed on sample solutions from a typical scrap recovery operation common in the uranium processing industry. A limited number of residue type materials, calciner products, and presscakes were subjected to analysis by organic extraction. The uranium extraction was not hindered by 30% TBP/CCl/sub 4/. To fully demonstrate the capabilities of the extraction technique and its compatibility with the NBL potentiometric uranium determination, a series of uranium standards was subjected to uranium extraction with 30% TBP/CCl/sub 4/. The uranium was then stripped out of the organic phase with 40 mL of H/sub 3/PO/sub 4/, 15 mL of H/sub 2/0, and 1 mL of 1M FeSO/sub 4/ solution. The uranium was then determined in the aqueous phosphoric phase by the regular NBL potentiometric method, omitting only the addition of another 40 mL of H/sub 3/PO/sub 4/. Uranium determinations ranging from approximately 20 to 150 mg of U were successfully made with the same accuracy and precision normally achieved. 8 tables. (DP)

  20. 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 Posted: July 18, 2014 - 4:39pm Front row, left to right: Bill Priest, Consolidated Nuclear...

  1. Colorimetric detection of uranium in water

    DOE Patents [OSTI]

    DeVol, Timothy A.; Hixon, Amy E.; DiPrete, David P.

    2012-03-13

    Disclosed are methods, materials and systems that can be used to determine qualitatively or quantitatively the level of uranium contamination in water samples. Beneficially, disclosed systems are relatively simple and cost-effective. For example, disclosed systems can be utilized by consumers having little or no training in chemical analysis techniques. Methods generally include a concentration step and a complexation step. Uranium concentration can be carried out according to an extraction chromatographic process and complexation can chemically bind uranium with a detectable substance such that the formed substance is visually detectable. Methods can detect uranium contamination down to levels even below the MCL as established by the EPA.

  2. Radiological Safety Training for Uranium Facilities

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

    ... Continued on Next Page * Stein, F., Instructor Competencies: the Standards. International ... and acute exposures to significant amounts of uranium may result in kidney damage. ...

  3. Uranium Enrichment Decontamination and Decommissioning Fund's...

    Broader source: Energy.gov (indexed) [DOE]

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2011 Financial ... Dear Mr. Friedman: We have audited the financial statements of the Department of Energy's ...

  4. Plutonium Uranium Extraction Plant (PUREX) - Hanford Site

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

    Site. The Plutonium Uranium Extraction Plant is massive. It is longer than three football fields, stands 64 feet above the ground, and extends another 40 feet below ground....

  5. Method for providing uranium articles with a corrosion resistant anodized coating

    DOE Patents [OSTI]

    Waldrop, Forrest B. (Powell, TN); Washington, Charles A. (Oak Ridge, TN)

    1982-01-01

    Uranium articles are provided with anodized oxide coatings in an aqueous solution of an electrolyte selected from the group consisting of potassium phosphate, potassium hydroxide, ammonium hydroxide, and a mixture of potassium tetraborate and boric acid. The uranium articles are anodized at a temperature greater than about 75.degree. C. with a current flow of less than about 0.036 A/cm.sup.2 of surface area while the pH of the solution is maintained in a range of about 2 to 11.5. The pH values of the aqueous solution and the low current density utilized during the electrolysis prevent excessive dissolution of the uranium and porosity in the film or watering. The relatively high temperature of the electrolyte bath inhibits hydration and the attendant deleterious pitting so as to enhance corrosion resistance of the anodized coating.

  6. Method for providing uranium articles with a corrosion-resistant anodized coating

    DOE Patents [OSTI]

    Waldrop, F.B.; Washington, C.A.

    1981-01-07

    Uranium articles are provided with anodized oxide coatings in an aqueous solution of an electrolyte selected from the group consisting of potassium phosphate, potassium hydroxide, ammonium hydroxide, and a mixture of potassium tetraborate and boric acid. The uranium articles are anodized at a temperature greater than about 75/sup 0/C with a current flow of less than about 0.036 A/cm/sup 2/ of surface area while the pH of the solution is maintained in a range of about 2 to 11.5. The pH values of the aqueous solution and the low current density utilized during the electrolysis prevent excessive dissolution of the uranium and porosity in the film or watering. The relatively high temperature of the electrolyte bath inhibits hydration and the attendant deleterious pitting so as to enhance corrosion resistance of the anodized coating.

  7. Highly Enriched Uranium Materials Facility | Y-12 National Security Complex

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

    Highly Enriched Uranium ... Highly Enriched Uranium Materials Facility HEUMF The Highly Enriched Uranium Materials Facility is our nation's central repository for highly enriched uranium, a vital national security asset. HEUMF is a massive concrete and steel structure that provides maximum security for the highly enriched uranium material that it protects. Approximately 300 feet by 475 feet, HEUMF has areas for receiving, shipping and providing long-term storage of the enriched uranium, as well

  8. High strength and density tungsten-uranium alloys

    DOE Patents [OSTI]

    Sheinberg, Haskell (Los Alamos, NM)

    1993-01-01

    Alloys of tungsten and uranium and a method for making the alloys. The amount of tungsten present in the alloys is from about 55 vol % to about 85 vol %. A porous preform is made by sintering consolidated tungsten powder. The preform is impregnated with molten uranium such that (1) uranium fills the pores of the preform to form uranium in a tungsten matrix or (2) uranium dissolves portions of the preform to form a continuous uranium phase containing tungsten particles.

  9. Millisecond Oxidation of Alkanes

    Broader source: Energy.gov [DOE]

    This factsheet describes a project whose goal is to commercialize a production process for propylene and acrylic acid from propane using a catalytic auto-thermal oxydehydrogenation process operating at short contact times. Auto-thermal oxidation for conversion of propane to propylene and acrylic acid promises energy savings of 20 trillion Btu per year by 2020. In addition to reducing energy consumption, this technology can reduce manufacturing costs by up to 25 percent, and reduce a variety of greenhouse gas emissions.

  10. lithium cobalt oxide cathode

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

    lithium cobalt oxide cathode - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs

  11. Composite catalyst for carbon monoxide and hydrocarbon oxidation

    DOE Patents [OSTI]

    Liu, Wei (Cambridge, MA); Flytzani-Stephanopoulos, Maria (Winchester, MA)

    1996-01-01

    A method and composition for the complete oxidation of carbon monoxide and/or hydrocarbon compounds. The method involves reacting the carbon monoxide and/or hydrocarbons with an oxidizing agent in the presence of a metal oxide composite catalyst. The catalyst is prepared by combining fluorite-type oxygen ion conductors with active transition metals. The fluorite oxide, selected from the group consisting of cerium oxide, zirconium oxide, thorium oxide, hafnium oxide, and uranium oxide, and may be doped by alkaline earth and rare earth oxides. The transition metals, selected from the group consisting of molybdnum, copper, cobalt, maganese, nickel, and silver, are used as additives. The atomic ratio of transition metal to fluorite oxide is less than one.

  12. Composite catalyst for carbon monoxide and hydrocarbon oxidation

    DOE Patents [OSTI]

    Liu, W.; Flytzani-Stephanopoulos, M.

    1996-03-19

    A method and composition are disclosed for the complete oxidation of carbon monoxide and/or hydrocarbon compounds. The method involves reacting the carbon monoxide and/or hydrocarbons with an oxidizing agent in the presence of a metal oxide composite catalyst. The catalyst is prepared by combining fluorite-type oxygen ion conductors with active transition metals. The fluorite oxide, selected from the group consisting of cerium oxide, zirconium oxide, thorium oxide, hafnium oxide, and uranium oxide, and may be doped by alkaline earth and rare earth oxides. The transition metals, selected from the group consisting of molybdenum, copper, cobalt, manganese, nickel, and silver, are used as additives. The atomic ratio of transition metal to fluorite oxide is less than one.

  13. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect (OSTI)

    Brown N. R.; Brown,N.R.; Baek,J.S; Hanson, A.L.; Cuadra,A.; Cheng,L.Y.; Diamond, D.J.

    2013-03-31

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. . The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). In addition, a summary of the methodology to obtain these results is presented.

  14. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect (OSTI)

    Brown, N. R.; Brown, N. R.; Baek, J. S; Hanson, A. L.; Cuadra, A.; Cheng, L. Y.; Diamond, D. J.

    2014-04-30

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

  15. Method for forming indium oxide/n-silicon heterojunction solar cells

    DOE Patents [OSTI]

    Feng, Tom (Morris Plains, NJ); Ghosh, Amal K. (New Providence, NJ)

    1984-03-13

    A high photo-conversion efficiency indium oxide/n-silicon heterojunction solar cell is spray deposited from a solution containing indium trichloride. The solar cell exhibits an Air Mass One solar conversion efficiency in excess of about 10%.

  16. Impact of HFIR LEU Conversion on Beryllium Reflector Degradation Factors

    SciTech Connect (OSTI)

    Ilas, Dan

    2013-10-01

    An assessment of the impact of low enriched uranium (LEU) conversion on the factors that may cause the degradation of the beryllium reflector is performed for the High Flux Isotope Reactor (HFIR). The computational methods, models, and tools, comparisons with previous work, along with the results obtained are documented and discussed in this report. The report documents the results for the gas and neutronic poison production, and the heating in the beryllium reflector for both the highly enriched uranium (HEU) and LEU HFIR configurations, and discusses the impact that the conversion to LEU may have on these quantities. A time-averaging procedure was developed to calculate the isotopic (gas and poisons) production in reflector. The sensitivity of this approach to different approximations is gauged and documented. The results show that the gas is produced in the beryllium reflector at a total rate of 0.304 g/cycle for the HEU configuration; this rate increases by ~12% for the LEU case. The total tritium production rate in reflector is 0.098 g/cycle for the HEU core and approximately 11% higher for the LEU core. A significant increase (up to ~25%) in the neutronic poisons production in the reflector during the operation cycles is observed for the LEU core, compared to the HEU case, for regions close to the core s horizontal midplane. The poisoning level of the reflector may increase by more than two orders of magnitude during long periods of downtime. The heating rate in the reflector is estimated to be approximately 20% lower for the LEU core than for the HEU core. The decrease is due to a significantly lower contribution of the heating produced by the gamma radiation for the LEU core. Both the isotopic (gas and neutronic poisons) production and the heating rates are spatially non-uniform throughout the beryllium reflector volume. The maximum values typically occur in the removable reflector and close to the midplane.

  17. DOE Evaluates Environmental Impacts of Uranium Mining on Government...

    Energy Savers [EERE]

    Evaluates Environmental Impacts of Uranium Mining on Government Land in Western Colorado DOE Evaluates Environmental Impacts of Uranium Mining on Government Land in Western...

  18. Record of Decision for the Uranium Leasing Program Programmatic...

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

    Record of Decision for the Uranium Leasing Program Programmatic Environmental Impact Statement Record of Decision for the Uranium Leasing Program Programmatic Environmental Impact...

  19. Toxic Substances Control Act Uranium Enrichment Federal Facility...

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

    Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic...

  20. DOE/NNSA Successfully Establishes Uranium Lease and Takeback...

    National Nuclear Security Administration (NNSA)

    Apply for Our Jobs Our Jobs Working at NNSA Blog Home NNSA Blog DOENNSA Successfully Establishes Uranium Lease and Takeback ... DOENNSA Successfully Establishes Uranium Lease ...

  1. Decommissioning of U.S. Uranium Production Facilities

    Reports and Publications (EIA)

    1995-01-01

    This report analyzes the uranium production facility decommissioning process and its potential impact on uranium supply and prices. 1995 represents the most recent publication year.

  2. DOE Seeks Contractor for Depleted Uranium Hexafluoride (DUF6...

    Office of Environmental Management (EM)

    Seeks Contractor for Depleted Uranium Hexafluoride (DUF6) Operations at Ohio and Kentucky Facilities DOE Seeks Contractor for Depleted Uranium Hexafluoride (DUF6) Operations at...

  3. Legacy Management Work Progresses on Defense-Related Uranium...

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

    defense-related legacy uranium mine sites located within 11 uranium mining districts in 6 western states. At these sites, photographs and global positioning location data were...

  4. Highly Enriched Uranium Materials Facility, Major Design Changes...

    Energy Savers [EERE]

    Highly Enriched Uranium Materials Facility, Major Design Changes Late...Lessons Learned Report, NNSA, Dec 2010 Highly Enriched Uranium Materials Facility, Major Design Changes...

  5. Sequestering Uranium from Seawater: Binding Strength and Modes...

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

    Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl Complexes with Glutarimidedioxime Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl...

  6. 3rd Quarter 2015 Domestic Uranium Production Report

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

    U.S. Energy Information Administration: Form EIA-851A and Form EIA-851Q, ""Domestic Uranium Production Report.""" " U.S. Energy Information Administration Domestic Uranium...

  7. Domestic Uranium Production Report 4th Quarter 2015

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

    Domestic Uranium Production Report 4th Quarter 2015 February ... DC 20585 U.S. Energy Information Administration | ... Team, Office of Electricity, Renewables, and Uranium ...

  8. Thermal conversion of biomass to valuable fuels, chemical feedstocks and chemicals

    DOE Patents [OSTI]

    Peters, William A. (Lexington, MA); Howard, Jack B. (Winchester, MA); Modestino, Anthony J. (Hanson, MA); Vogel, Fredreric (Villigen PSI, CH); Steffin, Carsten R. (Herne, DE)

    2009-02-24

    A continuous process for the conversion of biomass to form a chemical feedstock is described. The biomass and an exogenous metal oxide, preferably calcium oxide, or metal oxide precursor are continuously fed into a reaction chamber that is operated at a temperature of at least 1400.degree. C. to form reaction products including metal carbide. The metal oxide or metal oxide precursor is capable of forming a hydrolizable metal carbide. The reaction products are quenched to a temperature of 800.degree. C. or less. The resulting metal carbide is separated from the reaction products or, alternatively, when quenched with water, hydolyzed to provide a recoverable hydrocarbon gas feedstock.

  9. Enhancing uranium uptake by amidoxime adsorbent in seawater: An investigation for optimum alkaline conditioning parameters

    SciTech Connect (OSTI)

    Das, Sadananda; Tsouris, Costas; Zhang, Chenxi; Brown, Suree; Janke, Christopher James; Mayes, Richard T.; Kuo, Li -Jung; Gill, Gary; Dai, Sheng; Kim, J.; Oyola, Y.; Wood, J. R.

    2015-09-07

    A high-surface-area polyethylene-fiber adsorbent (AF160-2) has been developed at the Oak Ridge National Laboratory by radiation-induced graft polymerization of acrylonitrile and itaconic acid. The grafted nitriles were converted to amidoxime groups by treating with hydroxylamine. The amidoximated adsorbents were then conditioned with potassium hydroxide (KOH) by varying different reaction parameters such as KOH concentration (0.2, 0.44, and 0.6 M), duration (1, 2, and 3 h), and temperature (60, 70, and 80 C). Adsorbent screening was then performed with simulated seawater solutions containing sodium chloride and sodium bicarbonate, at concentrations found in seawater, and uranium nitrate at a uranium concentration of ~78 ppm and pH 8. Fourier transform infrared spectroscopy and solid-state NMR analyses indicated that a fraction of amidoxime groups was hydrolyzed to carboxylate during KOH conditioning. The uranium adsorption capacity in the simulated seawater screening solution gradually increased with conditioning time and temperature for all KOH concentrations. It was also observed that the adsorption capacity increased with an increase in concentration of KOH for all the conditioning times and temperatures. AF160-2 adsorbent samples were also tested with natural seawater using flow-through experiments to determine uranium adsorption capacity with varying KOH conditioning time and temperature. Based on uranium loading capacity values of several AF160-2 samples, it was observed that changing KOH conditioning time from 3 to 1 h at 60, 70, and 80 C resulted in an increase of the uranium loading capacity in seawater, which did not follow the trend found in laboratory screening with stimulated solutions. Longer KOH conditioning times lead to significantly higher uptake of divalent metal ions, such as calcium and magnesium, which is a result of amidoxime conversion into less selective carboxylate. The scanning electron microscopy showed that long conditioning times may also lead to adsorbent degradation.

  10. Enhancing uranium uptake by amidoxime adsorbent in seawater: An investigation for optimum alkaline conditioning parameters

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

    Das, Sadananda; Tsouris, Costas; Zhang, Chenxi; Brown, Suree; Janke, Christopher James; Mayes, Richard T.; Kuo, Li -Jung; Gill, Gary; Dai, Sheng; Kim, J.; et al

    2015-09-07

    A high-surface-area polyethylene-fiber adsorbent (AF160-2) has been developed at the Oak Ridge National Laboratory by radiation-induced graft polymerization of acrylonitrile and itaconic acid. The grafted nitriles were converted to amidoxime groups by treating with hydroxylamine. The amidoximated adsorbents were then conditioned with potassium hydroxide (KOH) by varying different reaction parameters such as KOH concentration (0.2, 0.44, and 0.6 M), duration (1, 2, and 3 h), and temperature (60, 70, and 80 °C). Adsorbent screening was then performed with simulated seawater solutions containing sodium chloride and sodium bicarbonate, at concentrations found in seawater, and uranium nitrate at a uranium concentration ofmore » ~7–8 ppm and pH 8. Fourier transform infrared spectroscopy and solid-state NMR analyses indicated that a fraction of amidoxime groups was hydrolyzed to carboxylate during KOH conditioning. The uranium adsorption capacity in the simulated seawater screening solution gradually increased with conditioning time and temperature for all KOH concentrations. It was also observed that the adsorption capacity increased with an increase in concentration of KOH for all the conditioning times and temperatures. AF160-2 adsorbent samples were also tested with natural seawater using flow-through experiments to determine uranium adsorption capacity with varying KOH conditioning time and temperature. Based on uranium loading capacity values of several AF160-2 samples, it was observed that changing KOH conditioning time from 3 to 1 h at 60, 70, and 80 °C resulted in an increase of the uranium loading capacity in seawater, which did not follow the trend found in laboratory screening with stimulated solutions. Longer KOH conditioning times lead to significantly higher uptake of divalent metal ions, such as calcium and magnesium, which is a result of amidoxime conversion into less selective carboxylate. The scanning electron microscopy showed that long conditioning times may also lead to adsorbent degradation.« less

  11. Corrosion of Uranium in Desert Soil, with Application to GCD Source Term M

    SciTech Connect (OSTI)

    ANDERSON, HOWARD L.; BACA, JULIANNE; KRUMHANSL, JAMES L.; STOCKMAN, HARLAN W.; THOMPSON, MOLLIE E.

    1999-09-01

    Uranium fragments from the Sandia Sled Track were studied as analogues for weapons components and depleted uranium buried at the Greater Confinement Disposal (GCD) site in Nevada. The Sled Track uranium fragments originated as weapons mockups and counterweights impacted on concrete and soil barriers, and experienced heating and fragmentation similar to processes thought to affect the Nuclear Weapons Accident Residues (NWAR) at GCD. Furthermore, the Sandia uranium was buried in unsaturated desert soils for 10 to 40 years, and has undergone weathering processes expected to affect the GCD wastes. Scanning electron microscopy, X-ray diffraction and microprobe analyses of the fragments show rapid alteration from metals to dominantly VI-valent oxy-hydroxides. Leaching studies of the samples give results consistent with published U-oxide dissolution rates, and suggest longer experimental periods (ca. 1 year) would be required to reach equilibrium solution concentrations. Thermochemical modeling with the EQ3/6 code indicates that the uranium concentrations in solutions saturated with becquerelite could increase as the pore waters evaporate, due to changes in carbonate equilibria and increased ionic strength.

  12. Biochemical Conversion - Biorefinery Integration | Department of Energy

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

    Research & Development » Conversion Technologies » Biochemical Conversion » Biochemical Conversion - Biorefinery Integration Biochemical Conversion - Biorefinery Integration One of the essential elements in the economical and efficient production of cellulosic biofuels is the development of biorefineries. Similar in concept to traditional petroleum refineries, biorefineries convert various types of biomass feedstock into marketable chemicals, fuels, and products. By taking advantage of

  13. Biomass Program 2007 Accomplishments - Biochemical Conversion Platform

    SciTech Connect (OSTI)

    none,

    2009-10-27

    This document details accomplishments of the Biomass Program Biochemical Conversion Platform accomplishments in 2007.

  14. Biochemical Conversion Pilot Plant (Fact Sheet)

    SciTech Connect (OSTI)

    Not Available

    2012-06-01

    This fact sheet provides information about Biochemical Conversion Pilot Plant capabilities and resources at NREL.

  15. Biomass Program 2007 Accomplishments - Thermochemical Conversion Platform

    SciTech Connect (OSTI)

    none,

    2009-10-27

    This document details the accomplishments of the Biomass Program Thermochemical Conversion Platform in 2007.

  16. Conversion Technologies for Advanced Biofuels - Carbohydrates Production

    Office of Environmental Management (EM)

    | Department of Energy Production Conversion Technologies for Advanced Biofuels - Carbohydrates Production Purdue University report-out presentation at the CTAB webinar on Carbohydrates Production. PDF icon ctab_webinar_carbohydrates_production.pdf More Documents & Publications Advanced Conversion Roadmap Workshop Workshop on Conversion Technologies for Advanced Biofuels - Carbohydrates Conversion Technologies for Advanced Biofuels - Carbohydrates Upgrading

  17. Recirculation in multiple wave conversions

    SciTech Connect (OSTI)

    Kaufman, A. N.; Brizard, A.J.; Kaufman, A.N.; Tracy, E.R.

    2008-07-30

    A one-dimensional multiple wave-conversion model is constructed that allows energy recirculation in ray phase space. Using a modular eikonal approach, the connection coefficients for this model are calculated by ray phase-space methods. Analytical results (confirmed numerically) show that all connection coefficients exhibit interference effects that depend on an interference phase, calculated from the coupling constants and the area enclosed by the intersecting rays. This conceptual model, which focuses on the topology of intersecting rays in phase space, is used to investigate how mode conversion between primary and secondary waves is modified by the presence of a tertiary wave.

  18. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    5 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Enrichment service contract type U.S. enrichment Foreign enrichment Total Spot W W 628 Long-term W W 12,310 Total 3,773 9,165 12,939 Table 17. Purchases of enrichment services by owners and operators of U.S. civilian nuclear power reactors by contract type in delivery year, 2014 thousand separative work units (SWU) W = Data withheld to avoid disclosure of individual company data. Note: Totals may not

  19. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    7 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Table S3b. Weighted-average price of foreign purchases and foreign sales by U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors, 1994-2014 Delivery year Foreign purchases by U.S. suppliers Foreign purchases by owners and operators of U.S. civilian nuclear power reactors Total foreign purchases (weighted-average price) U.S. broker and trader purchases from foreign suppliers

  20. 2014 Domestic Uranium Production Report

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

    7. Employment in the U.S. uranium production industry by state, 2003-14" "person-years" "State(s)",2003,2004,2005,2006,2007,2008,2009,2010,2011,2012,2013,2014 "Wyoming",134,139,181,195,245,301,308,348,424,512,531,416 "Colorado and Texas",48,140,269,263,557,696,340,292,331,248,198,105 "Nebraska and New Mexico",92,102,123,160,149,160,159,134,127,"W","W","W" "Arizona, Utah, and

  1. 2014 Domestic Uranium Production Report

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

    2. U.S. uranium mine production and number of mines and sources, 2003-14" "Production / Mining Method",2003,2004,2005,2006,2007,2008,2009,2010,2011,2012,2013,2014 "Underground" "(estimated contained thousand pounds U3O8)","W","W","W","W","W","W","W","W","W","W","W","W" "Open Pit" "(estimated contained thousand pounds

  2. 2014 Domestic Uranium Production Report

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

    3. U.S. uranium concentrate production, shipments, and sales, 2003-14" "Activity at U.S. Mills and In-Situ-Leach Plants",2003,2004,2005,2006,2007,2008,2009,2010,2011,2012,2013,2014 "Estimated contained U3O8 (thousand pounds)" "Ore from Underground Mines and Stockpiles Fed to Mills 1",0,"W","W","W",0,"W","W","W","W","W","W","W" "Other Feed Materials

  3. 2014 Uranium Marketing Annual Report

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

    5. Enrichment service sellers to owners and operators of U.S. civilian nuclear power reactors, 2012-14" 2012,2013,2014 "Advance Uranium Asset Management Ltd.","AREVA NC, Inc.","AREVA Enrichment Services, LLC / AREVA NC, Inc." "AREVA NC, Inc.","CNEIC (China Nuclear Energy Industry Corporation)","CNEIC (China Nuclear Energy Industry Corporation)" "CNEIC (China Nuclear Energy Industry Corporation)","LES, LLC (Louisiana

  4. 2014 Uranium Marketing Annual Report

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

    6a. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by quantity, 2012-14 deliveries" "thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent" "Quantity distribution 1","Deliveries in 2012",,"Deliveries in 2013",,"Deliveries in 2014" ,"Quantity with reported price","Weighted-average price","Quantity with reported

  5. 2014 Uranium Marketing Annual Report

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

    b. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by purchaser, 2012-14 deliveries" "thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent" "Distribution of purchasers","Deliveries in 2012",,,"Deliveries in 2013",,,"Deliveries in 2014" ,"Number of purchasers","Quantity with reported price","Weighted-average price","Number of

  6. 2014 Uranium Marketing Annual Survey

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

    7. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by contract type and material type, 2014 deliveries" "thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent" "Material Type","Spot Contracts 1",,"Long-Term Contracts 2",,"Total" ,"Quantity with reported price","Weighted-average price","Quantity with reported price","Weighted-average price","Quantity

  7. Biofuel Conversion Basics | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    The conversion of biomass solids into liquid or gaseous biofuels is a complex process. Today, the most common conversion processes are biochemical- and thermochemical-based. However, researchers are also exploring photobiological conversion processes. Biochemical Conversion Processes In biochemical conversion processes, enzymes and microorganisms are used as biocatalysts to convert biomass or biomass-derived compounds into desirable products. Cellulase and hemicellulase enzymes break down the

  8. Biochemical Conversion Related Links | Department of Energy

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

    Conversion » Biochemical Conversion Related Links Biochemical Conversion Related Links Further reading about current Bioenergy Technologies Office R&D in the Biochemical Platform can be found in this website's Information Resources section. Key publications will also be provided on this page. Using Fermentation and Catalysis to Make Fuels and Products: Biochemical Conversion (January 2011) Biochemical Conversion 2009 Peer Review Biochemical Production of Ethanol from Corn Stover: 2007 State

  9. Thermochemical Conversion Related Links | Department of Energy

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

    Conversion » Thermochemical Conversion Related Links Thermochemical Conversion Related Links Further reading about current Bioenergy Technologies Office R&D in the Thermochemical Platform can be found in this website's Information Resources section. Some key publications are: Using Heat and Chemistry to Make Fuel and Power: Thermochemical Conversion (January 2011) Thermochemical Conversion 2009 Peer Review Design Case Summary: Production of Gasoline and Diesel from Biomass via Fast

  10. Ocean thermal energy conversion (OTEC)

    SciTech Connect (OSTI)

    Lockerby, R.W.

    1981-01-01

    Ocean thermal energy conversion (OTEC) is reviewed briefly. The two types of OTEC system (open and closed) are described and limitations are pointed out. A bibliography of 148 references on OTEC is given for the time period 1975 to 1980. Entries are arranged alphabetically according to the author's name. (MJJ)

  11. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006

    SciTech Connect (OSTI)

    Primm, R. T.; Ellis, R. J.; Gehin, J. C.; Clarno, K. T.; Williams, K. A.; Moses, D. L.

    2006-11-01

    Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, continuously graded fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 ?m is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457?m. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

  12. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  13. Uranium Pyrophoricity Phenomena and Prediction

    SciTech Connect (OSTI)

    DUNCAN, D.R.

    2000-04-20

    We have compiled a topical reference on the phenomena, experiences, experiments, and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel Project (SNFP) with specific applications to SNFP process and situations. The purpose of the compilation is to create a reference to integrate and preserve this knowledge. Decades ago, uranium and zirconium fires were commonplace at Atomic Energy Commission facilities, and good documentation of experiences is surprisingly sparse. Today, these phenomena are important to site remediation and analysis of packaging, transportation, and processing of unirradiated metal scrap and spent nuclear fuel. Our document, bearing the same title as this paper, will soon be available in the Hanford document system [Plys, et al., 2000]. This paper explains general content of our topical reference and provides examples useful throughout the DOE complex. Moreover, the methods described here can be applied to analysis of potentially pyrophoric plutonium, metal, or metal hydride compounds provided that kinetic data are available. A key feature of this paper is a set of straightforward equations and values that are immediately applicable to safety analysis.

  14. Uranium Leasing Program Environmental Documents | Department of Energy

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

    Environmental Documents Uranium Leasing Program Environmental Documents Uranium Leasing Program Mitigation Action Plan for the Final Uranium Leasing Program Programmatic Environmental Impact Statement DOE/EIS-0472 (November 2014) Record of Decision Final Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS)

  15. Uranium Lease Tracts Location Map | Department of Energy

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

    Uranium Lease Tracts Location Map Uranium Lease Tracts Location Map Uranium Lease Tracts Location Map PDF icon Uranium Lease Tracts Location Map More Documents & Publications EA-1037: Final Environmental Assessment EA-1535: Final Programmatic Environmental Assessment EIS-0472: Notice of Intent to Prepare a Programmatic Environmental Impact Statement

  16. Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

    SciTech Connect (OSTI)

    Pinkston, Daniel [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL

    2010-10-01

    The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

  17. MHD compressor---expander conversion system integrated with GCR inside a deployable reflector

    SciTech Connect (OSTI)

    Tuninetti, G. . Research Div.); Botta, E.; Criscuolo, C.; Riscossa, P. . Nuclear Div.); Giammanco, F. . Dipt. di Fisica); Rosa-Clot, M. . Dipt. di Fisica)

    1989-04-20

    This work originates from the proposal MHD Compressor-Expander Conversion System Integrated with a GCR Inside a Deployable Reflector''. The proposal concerned an innovative concept of nuclear, closed-cycle MHD converter for power generation on space-based systems in the multi-megawatt range. The basic element of this converter is the Power Conversion Unit (PCU) consisting of a gas core reactor directly coupled to an MHD expansion channel. Integrated with the PCU, a deployable reflector provides reactivity control. The working fluid could be either uranium hexafluoride or a mixture of uranium hexafluoride and helium, added to enhance the heat transfer properties. The original Statement of Work, which concerned the whole conversion system, was subsequently redirected and focused on the basic mechanisms of neutronics, reactivity control, ionization and electrical conductivity in the PCU. Furthermore, the study was required to be inherently generic such that the study was required to be inherently generic such that the analysis an results can be applied to various nuclear reactor and/or MHD channel designs''.

  18. Thermophotovoltaic conversion using selective infrared line emitters and large band gap photovoltaic devices

    DOE Patents [OSTI]

    Brandhorst, Jr., Henry W. (Auburn, AL); Chen, Zheng (Auburn, AL)

    2000-01-01

    Efficient thermophotovoltaic conversion can be performed using photovoltaic devices with a band gap in the 0.75-1.4 electron volt range, and selective infrared emitters chosen from among the rare earth oxides which are thermally stimulated to emit infrared radiation whose energy very largely corresponds to the aforementioned band gap. It is possible to use thermovoltaic devices operating at relatively high temperatures, up to about 300.degree. C., without seriously impairing the efficiency of energy conversion.

  19. Study of Chemical Changes in Uranium Oxyfluoride Particles Progress Report March - October 2009

    SciTech Connect (OSTI)

    Kips, R; Kristo, M; Hutcheon, I

    2009-11-22

    Nuclear forensics relies on the analysis of certain sample characteristics to determine the origin and history of a nuclear material. In the specific case of uranium enrichment facilities, it is the release of trace amounts of uranium hexafluoride (UF{sub 6}) gas - used for the enrichment of uranium - that leaves a process-characteristic fingerprint. When UF{sub 6} gas interacts with atmospheric moisture, uranium oxyfluoride particles or particle agglomerates are formed with sizes ranging from several microns down to a few tens of nanometers. These particles are routinely collected by safeguards organizations, such as the International Atomic Energy Agency (IAEA), allowing them to verify whether a facility is compliant with its declarations. Spectrometric analysis of uranium particles from UF{sub 6} hydrolysis has revealed the presence of both particles that contain fluorine, and particles that do not. It is therefore assumed that uranium oxyfluoride is unstable, and decomposes to form uranium oxide. Understanding the rate of fluorine loss in uranium oxyfluoride particles, and the parameters that control it, may therefore contribute to placing boundaries on the particle's exposure time in the environment. Expressly for the purpose of this study, we prepared a set of uranium oxyfluoride particles at the Institute for Reference Materials and Measurements (EU-JRC-IRMM) from a static release of UF{sub 6} in a humid atmosphere. The majority of the samples was stored in controlled temperature, humidity and lighting conditions. Single particles were characterized by a suite of micro-analytical techniques, including NanoSIMS, micro-Raman spectrometry (MRS), scanning (SEM) and transmission (TEM) electron microscopy, energy-dispersive X-ray spectrometry (EDX) and focused ion beam (FIB). The small particle size was found to be the main analytical challenge. The relative amount of fluorine, as well as the particle chemical composition and morphology were determined at different stages in the ageing process, and immediately after preparation. This report summarizes our most recent findings for each of the analytical techniques listed above, and provides an outlook on what remains to be resolved. Additional spectroscopic and mass spectrometric measurements were carried out at Pacific Northwest National Laboratory, but are not included in this summary.

  20. uranium

    National Nuclear Security Administration (NNSA)

    a>

    NNSA Removes U.S.-Origin HEU from Jamaica, Makes the Caribbean HEU Free http:nnsa.energy.govmediaroompressreleasesnnsa-removes-u.s.-origin-heu-jamaica-mak...

  1. Technical Basis for Assessing Uranium Bioremediation Performance

    SciTech Connect (OSTI)

    PE Long; SB Yabusaki; PD Meyer; CJ Murray; AL NGuessan

    2008-04-01

    In situ bioremediation of uranium holds significant promise for effective stabilization of U(VI) from groundwater at reduced cost compared to conventional pump and treat. This promise is unlikely to be realized unless researchers and practitioners successfully predict and demonstrate the long-term effectiveness of uranium bioremediation protocols. Field research to date has focused on both proof of principle and a mechanistic level of understanding. Current practice typically involves an engineering approach using proprietary amendments that focuses mainly on monitoring U(VI) concentration for a limited time period. Given the complexity of uranium biogeochemistry and uranium secondary minerals, and the lack of documented case studies, a systematic monitoring approach using multiple performance indicators is needed. This document provides an overview of uranium bioremediation, summarizes design considerations, and identifies and prioritizes field performance indicators for the application of uranium bioremediation. The performance indicators provided as part of this document are based on current biogeochemical understanding of uranium and will enable practitioners to monitor the performance of their system and make a strong case to clients, regulators, and the public that the future performance of the system can be assured and changes in performance addressed as needed. The performance indicators established by this document and the information gained by using these indicators do add to the cost of uranium bioremediation. However, they are vital to the long-term success of the application of uranium bioremediation and provide a significant assurance that regulatory goals will be met. The document also emphasizes the need for systematic development of key information from bench scale tests and pilot scales tests prior to full-scale implementation.

  2. CONVERSION EXTRACTION DESULFURIZATION (CED) PHASE III

    SciTech Connect (OSTI)

    James Boltz

    2005-03-01

    This project was undertaken to refine the Conversion Extraction Desulfurization (CED) technology to efficiently and economically remove sulfur from diesel fuel to levels below 15-ppm. CED is considered a generic term covering all desulfurization processes that involve oxidation and extraction. The CED process first extracts a fraction of the sulfur from the diesel, then selectively oxidizes the remaining sulfur compounds, and finally extracts these oxidized materials. The Department of Energy (DOE) awarded Petro Star Inc. a contract to fund Phase III of the CED process development. Phase III consisted of testing a continuous-flow process, optimization of the process steps, design of a pilot plant, and completion of a market study for licensing the process. Petro Star and the Degussa Corporation in coordination with Koch Modular Process Systems (KMPS) tested six key process steps in a 7.6-centimeter (cm) (3.0-inch) inside diameter (ID) column at gas oil feed rates of 7.8 to 93.3 liters per hour (l/h) (2.1 to 24.6 gallons per hour). The team verified the technical feasibility with respect to hydraulics for each unit operation tested and successfully demonstrated pre-extraction and solvent recovery distillation. Test operations conducted at KMPS demonstrated that the oxidation reaction converted a maximum of 97% of the thiophenes. The CED Process Development Team demonstrated that CED technology is capable of reducing the sulfur content of light atmospheric gas oil from 5,000-ppm to less than 15-ppm within the laboratory scale. In continuous flow trials, the CED process consistently produced fuel with approximately 20-ppm of sulfur. The process economics study calculated an estimated process cost of $5.70 per product barrel. The Kline Company performed a marketing study to evaluate the possibility of licensing the CED technology. Kline concluded that only 13 refineries harbored opportunity for the CED process. The Kline study and the research team's discussions with prospective refineries led to the conclusion that there were not likely prospects for the licensing of the CED process.

  3. Nanostructured Water Oxidation Catalysts - Energy Innovation Portal

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

    Industrial Technologies Industrial Technologies Advanced Materials Advanced Materials Find More Like This Return to Search Nanostructured Water Oxidation Catalysts Lawrence Berkeley National Laboratory Contact LBL About This Technology Technology Marketing SummaryHeinz Frei and Feng Jiao of Berkeley Lab have developed a visible light driven catalytic system for oxidizing water. Efficient catalytic water oxidation is a critical step for any artificial sunlight-to-fuel conversion system.

  4. Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Primm, Trent; Guida, Tracey

    2010-02-01

    Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

  5. Electrocatalysts for carbon dioxide conversion

    DOE Patents [OSTI]

    Masel, Richard I; Salehi-Khojin, Amin

    2015-04-21

    Electrocatalysts for carbon dioxide conversion include at least one catalytically active element with a particle size above 0.6 nm. The electrocatalysts can also include a Helper Catalyst. The catalysts can be used to increase the rate, modify the selectivity or lower the overpotential of electrochemical conversion of CO.sub.2. Chemical processes and devices using the catalysts also include processes to produce CO, HCO.sup.-, H.sub.2CO, (HCO.sub.2).sup.-, H.sub.2CO.sub.2, CH.sub.3OH, CH.sub.4, C.sub.2H.sub.4, CH.sub.3CH.sub.2OH, CH.sub.3COO.sup.-, CH.sub.3COOH, C.sub.2H.sub.6, (COOH).sub.2, or (COO.sup.-).sub.2, and a specific device, namely, a CO.sub.2 sensor.

  6. Secretarial Determination for the Sale or Transfer of Uranium | Department

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

    of Energy Secretarial Determination for the Sale or Transfer of Uranium Secretarial Determination for the Sale or Transfer of Uranium Secretarial Determination for the Sale or Transfer of Uranium, May 15, 2012 PDF icon Secretarial Determination for the Sale or Transfer of Uranium.pdf More Documents & Publications Secretarial Determination Pursuant to USEC Privatization Act for the Sale or Transfer of Low-Enriched Uranium Before the House Committee on Oversight and Government Reform

  7. Record of Decision for the Uranium Leasing Program Programmatic

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

    Environmental Impact Statement | Department of Energy Record of Decision for the Uranium Leasing Program Programmatic Environmental Impact Statement Record of Decision for the Uranium Leasing Program Programmatic Environmental Impact Statement The U.S. Department of Energy (DOE) issued its Record of Decision for the Uranium Leasing Program on May 6, 2014, announcing that it will continue managing the Uranium Leasing Program for another 10 years. PDF icon Record of Decision for the Uranium

  8. Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium

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

    and Uranium-233 | Department of Energy Waste Management » Nuclear Materials & Waste » Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 105-K building houses the K-Area Material Storage (KAMS) facility, designated for the consolidated storage of surplus plutonium at Savannah River Site pending disposition. The plutonium shipped to KAMS is sealed inside a

  9. Uranium Leasing Program: Program Summary | Department of Energy

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

    Uranium Leasing Program » Uranium Leasing Program: Program Summary Uranium Leasing Program: Program Summary Uranium Leasing Program: Program Summary The Atomic Energy Act and other legislative actions authorized the U.S. Atomic Energy Commission (AEC), predecessor agency to the DOE, to withdraw lands from the public domain and then lease them to private industry for mineral exploration and for development and mining of uranium and vanadium ore. A total of 25,000 acres of land in southwestern

  10. Mined Land Reclamation on DOE's Uranium Lease Tracts, Southwestern

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

    Colorado | Department of Energy Mined Land Reclamation on DOE's Uranium Lease Tracts, Southwestern Colorado Mined Land Reclamation on DOE's Uranium Lease Tracts, Southwestern Colorado Mined Land Reclamation on DOE's Uranium Lease Tracts, Southwestern Colorado PDF icon Mined Land Reclamation on DOE's Uranium Lease Tracts, Southwestern Colorado More Documents & Publications EA-1535: Final Programmatic Environmental Assessment EA-1037: Final Environmental Assessment Final Uranium Leasing

  11. Final Uranium Leasing Program Programmatic Environmental Impact Statement

    Office of Environmental Management (EM)

    (PEIS) | Department of Energy Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) Final Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) Uranium Leasing Program-Mesa, Montrose, and San Miguel Counties, Colorado EIS-0472 evaluated the environmental impacts of management alternatives for DOE's Uranium Leasing Program, under which DOE administers tracts of land in western Colorado for exploration, development, and the extraction of uranium and

  12. RECOVERY OF URANIUM FROM CARBONATE LEACH LIQUORS

    DOE Patents [OSTI]

    Wilson, H.F.

    1958-07-01

    An improved process is described for the recovery of uranium from vanadifrous ores. In the prior art such ores have been digested with alkali carbonate solutions at a pH of less than 10 and then contacted with a strong base anion exchange resin to separate uranium from vanadium. It has been found that if the exchamge resin feed solution has its pH adjusted to the range 10.8 to 11.8, that vanadium adsorption on the resin is markedly decreased and the separation of uranium from the vanadium is thereby improved.

  13. Algal Polyculture Conversion & Analysis

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

    Algal Polyculture Conversion & Analysis Ron Pate This presentation does not contain any proprietary, confidential, or otherwise restricted information 24 March 2015 Algae Technology Area DOE Bioenergy Technologies Office (BETO) 2015 Project Peer Review Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. 2 | Bioenergy Technologies

  14. Conversion | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Conversion | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog

  15. Inherently safe in situ uranium recovery (Patent) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Inherently safe in situ uranium recovery Citation Details In-Document Search Title: Inherently safe in situ uranium recovery An in situ recovery of uranium operation involves...

  16. Thermodynamic properties of uranium dioxide

    SciTech Connect (OSTI)

    Fink, J.K.; Chasanov, M.G.; Leibowitz, L.

    1981-04-01

    In order to provide reliable and consistent data on the thermophysical properties of reactor materials for reactor safety studies, this revision is prepared for the thermodynamic properties of the uranium dioxide portion of the fuel property section of the report Properties for LMFBR Safety Analysis. Since the original report was issued in 1976, there has been international agreement on a vapor pressure equation for the total pressure over UO/sub 2/, new methods have been suggested for the calculation of enthalpy and heat capacity, and a phase change at 2670 K has been proposed. In this report, an electronic term is used in place of the Frenkel defect term in the enthalpy and heat capacity equation and the phase transition is accepted.

  17. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    9 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 Year Exploration Mining Milling Processing Reclamation Total 2003 W W W W 117 321 2004 18 108 W W 121 420 2005 79 149 142 154 124 648 2006 188 121 W W 155 755 2007 375 378 107 216 155 1,231 2008 457 558 W W 154 1,563 2009 175 441 W W 162 1,096 2010 211 400 W W 125 1,073 2011 208 462 W W 102 1,191 2012 161 462 W W 179 1,196 2013 149 392 W W 199 1,156 2014 86 246 W W 161 787 Figure 3. Employment in

  18. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    5 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 Production / Mining Method 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 (estimated contained thousand pounds U 3 O 8 ) W W W W W W W W W W W W (estimated contained thousand pounds U 3 O 8 ) 0 0 0 0 0 0 0 0 0 0 0 0 (thousand pounds U 3 O 8 ) W W 2,681 4,259 W W W W W W W W (thousand pounds U 3 O 8 ) W W W W W W W W W W W W (thousand pounds U 3 O 8 ) E2,200 2,452 3,045 4,692 4,541

  19. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    3 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 As of As of December 31, 2013 December 31, 2014 2015 45,498 48,206 2,708 2,708 2016 48,693 46,529 -2,164 544 2017 47,005 49,924 2,919 3,463 2018 52,138 51,169 -969 2,494 2019 50,041 46,184 -3,857 -1,363 2020 49,726 49,598 -128 -1,491 2021 50,455 51,793 1,338 -153 2022 49,320 50,286 966 813 2023 49,688 49,118 -570 243 2024 - 51,829 -- -- thousand pounds U 3 O 8 equivalent Cumulative Figure 14. Shipments

  20. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    7 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Deliveries 2010 2011 2012 2013 2014 Foreign purchases 24,985 19,318 20,196 23,233 24,199 Weighted-average price 41.30 48.80 46.80 43.25 39.13 Foreign purchases 30,362 35,071 36,037 34,095 34,404 Weighted-average price 51.69 56.87 54.08 51.64 47.62 Foreign purchases 55,347 54,388 56,233 57,328 58,603 Weighted-average price 47.01 54.00 51.44 48.24 44.11 thousand pounds U 3 O 8 equivalent Figure 17.