Sample records for uranium uranium fuel

  1. High loading uranium fuel plate

    DOE Patents [OSTI]

    Wiencek, Thomas C. (Bolingbrook, IL); Domagala, Robert F. (Indian Head Park, IL); Thresh, Henry R. (Palos Heights, IL)

    1990-01-01T23:59:59.000Z

    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.

  2. URANIUM MILLING ACTIVITIES AT SEQUOYAH FUELS CORPORATION

    E-Print Network [OSTI]

    unknown authors

    Sequoyah Fuels Corporation (SFC) describes previous operations at its Gore, Oklahoma, uranium conversion facility as: (1) the recovery of uranium by concentration and purification processes; and (2) the conversion of concentrated and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these

  3. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    SciTech Connect (OSTI)

    A K Wertsching

    2012-09-01T23:59:59.000Z

    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 1970’s. 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.

  4. Fuel cycle optimization of thorium and uranium fueled PWR systems

    E-Print Network [OSTI]

    Garel, Keith Courtnay

    1977-01-01T23:59:59.000Z

    The burnup neutronics of uniform PWR lattices are examined with respect to reduction of uranium ore requirements with an emphasis on variation of the fuel-to-moderator ratio

  5. Measurement of enriched uranium and uranium-aluminum fuel materials with the AWCC

    SciTech Connect (OSTI)

    Krick, M.S.; Menlove, H.O.; Zick, J.; Ikonomou, P.

    1985-05-01T23:59:59.000Z

    The active well coincidence counter (AWCC) was calibrated at the Chalk River Nuclear Laboratories (CRNL) for the assay of 93%-enriched fuel materials in three categories: (1) uranium-aluminum billets, (2) uranium-aluminum fuel elements, and (3) uranium metal pieces. The AWCC was a standard instrument supplied to the International Atomic Energy Agency under the International Safeguards Project Office Task A.51. Excellent agreement was obtained between the CRNL measurements and previous Los Alamos National Laboratory measurements on similar mockup fuel material. Calibration curves were obtained for each sample category. 2 refs., 8 figs., 15 tabs.

  6. Uranium and cesium diffusion in fuel cladding of electrogenerating channel

    SciTech Connect (OSTI)

    Vasil’ev, I. V., E-mail: fnti@mail.ru; Ivanov, A. S.; Churin, V. A. [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15T23:59:59.000Z

    The results of reactor tests of a carbonitride fuel in a single-crystal cladding from a molybdenum-based alloy can be used in substantiating the operational reliability of fuels in developing a project of a megawatt space nuclear power plant. The results of experimental studies of uranium and cesium penetration into the single-crystal cladding of fuel elements with a carbonitride fuel are interpreted. Those fuel elements passed nuclear power tests in the Ya-82 pilot plant for 8300 h at a temperature of about 1500°C. It is shown that the diffusion coefficients for uranium diffusion into the cladding are virtually coincident with the diffusion coefficients measured earlier for uranium diffusion into polycrystalline molybdenum. It is found that the penetration of uranium into the cladding is likely to occur only in the case of a direct contact between the cladding and fuel. The experimentally observed nonmonotonic uranium-concentration profiles are explained in terms of predominant uranium diffusion along grain boundaries. It is shown that a substantially nonmonotonic behavior observed in our experiment for the uranium-concentration profile may be explained by the presence of a polycrystalline structure of the cladding in the surface region from its inner side. The diffusion coefficient is estimated for the grain-boundary diffusion of uranium. The diffusion coefficients for cesium are estimated on the basis of experimental data obtained in the present study.

  7. Uranium industry annual 1996

    SciTech Connect (OSTI)

    NONE

    1997-04-01T23:59:59.000Z

    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.

  8. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03T23:59:59.000Z

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  9. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1998-01-01T23:59:59.000Z

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  10. Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications 

    E-Print Network [OSTI]

    Helmreich, Grant

    2012-02-14T23:59:59.000Z

    The sintering behavior of uranium and uranium-zirconium alloys in the alpha phase were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding...

  11. Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications

    E-Print Network [OSTI]

    Helmreich, Grant

    2012-02-14T23:59:59.000Z

    The sintering behavior of uranium and uranium-zirconium alloys in the alpha phase were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding...

  12. Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

    DOE Patents [OSTI]

    Ackerman, John P. (Downers Grove, IL); Miller, William E. (Naperville, IL)

    1989-01-01T23:59:59.000Z

    An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

  13. Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

    DOE Patents [OSTI]

    Ackerman, J.P.; Miller, W.E.

    1987-11-05T23:59:59.000Z

    An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

  14. New Tool for Proliferation Resistance Evaluation Applied to Uranium and Thorium Fueled Fast Reactor Fuel Cycles

    E-Print Network [OSTI]

    Metcalf, Richard R.

    2010-07-14T23:59:59.000Z

    reactor cycle and one scenario involves theft from a PUREX facility in a LWR cycle. The FBRFC was evaluated with uranium-plutonium fuel and a second time using thorium-uranium fuel. These diversion scenarios were tested with both uniform and expert weights...

  15. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, J.P.

    1992-03-17T23:59:59.000Z

    A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  16. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, John P. (Downers Grove, IL)

    1992-01-01T23:59:59.000Z

    A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  17. Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems 

    E-Print Network [OSTI]

    Szakaly, Frank Joseph

    2004-09-30T23:59:59.000Z

    The purpose of this work is to investigate the implementation of nitride fuels containing little or no uranium in a fast-spectrum nuclear reactor to reduce the amount of plutonium and minor actinides in spent nuclear fuel ...

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

    SciTech Connect (OSTI)

    Dr. Paul A. Lessing

    2012-03-01T23:59:59.000Z

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

  19. Uranium industry annual 1998

    SciTech Connect (OSTI)

    NONE

    1999-04-22T23:59:59.000Z

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

  20. Uranium chloride extraction of transuranium elements from LWR fuel

    DOE Patents [OSTI]

    Miller, William E. (Naperville, IL); Ackerman, John P. (Downers Grove, IL); Battles, James E. (Oak Forest, IL); Johnson, Terry R. (Wheaton, IL); Pierce, R. Dean (Naperville, IL)

    1992-01-01T23:59:59.000Z

    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.

  1. 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.

    1992-08-25T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2011-11-01T23:59:59.000Z

    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.

  3. Empirical modeling of uranium nitride fuels

    E-Print Network [OSTI]

    Brozak, Daniel Edward

    1987-01-01T23:59:59.000Z

    of this work was to develop an irradiation performance data base for nitride fuels and to provide empirical modeling capabilities for fuel swelling and fission gas release in nitride fuels. The nitride fuels data base represents the most extensive effort... to date to systematically collect, evaluate and analyze irradiation performance data for nitride fuels. The data base will be a valuable tool for all researchers in the advanced fuel modeling field in that it provides a foundation for the construction...

  4. Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

    SciTech Connect (OSTI)

    Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y. [RPA - V.G.Khlopin Radium Institute, St-Petersburg (Russian Federation); Baryshnikov, M.V.; Kryukov, O.V.; Khaperskaya, A.V. [State Corporation ROSATOM, Moscow (Russian Federation)

    2013-07-01T23:59:59.000Z

    REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

  5. Nuclear Fuel Facts: Uranium | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreaking ofOilNEWResponse to Time-Based Rates from the ConsumerNuclearCycle

  6. Uranium industry annual 1995

    SciTech Connect (OSTI)

    NONE

    1996-05-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    none,

    2013-07-01T23:59:59.000Z

    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.

  8. Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

    SciTech Connect (OSTI)

    Pesic, Milan P

    2003-10-15T23:59:59.000Z

    The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments', contributing to the preservation of a rather small number of heavy water benchmark critical experiments.

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

    DOE Patents [OSTI]

    Herrmann, Steven Douglas

    2014-05-27T23:59:59.000Z

    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.

  10. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Pruett, D.J.; McTaggart, D.R.

    1983-08-31T23:59:59.000Z

    Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

  11. Uranium industry annual 1997

    SciTech Connect (OSTI)

    NONE

    1998-04-01T23:59:59.000Z

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

  12. URANIUM IN ALKALINE ROCKS

    E-Print Network [OSTI]

    Murphy, M.

    2011-01-01T23:59:59.000Z

    chemical elements uranium zirconium niobium beryllium rarerare earths, niobium, zirconium, uranium, and thorium.respect, uranium and thorium are niobium in carbonatitcs.

  13. The study of material accountancy procedures for uranium in a whole nuclear fuel cycle

    SciTech Connect (OSTI)

    Nakano, Hiromasa; Akiba, Mitsunori [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1995-07-01T23:59:59.000Z

    Material accountancy procedures for uranium under a whole nuclear fuel cycle were studied by taking into consideration the material accountancy capability associated with realistic measurement uncertainties. The significant quantity used by the International Atomic Energy Agency (IAEA) for low-enriched uranium is 75 kg U-235 contained. A loss of U-235 contained in uranium can be detected by either of the following two procedures: one is a traditional U-235 isotope balance, and the other is a total uranium element balance. Facility types studied in this paper were UF6 conversion, gas centrifuge uranium enrichment, fuel fabrication, reprocessing, plutonium conversion, and MOX fuel production in Japan, where recycled uranium is processed in addition to natural uranium. It was found that the material accountancy capability of a total uranium element balance was almost always higher than that of a U-235 isotope balance under normal accuracy of weight, concentration, and enrichment measurements. Changing from the traditional U-235 isotope balance to the total uranium element balance for these facilities would lead to a gain of U-235 loss detection capability through material accountancy and to a reduction in the required resources of both the IAEA and operators.

  14. Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems

    E-Print Network [OSTI]

    Szakaly, Frank Joseph

    2004-09-30T23:59:59.000Z

    The purpose of this work is to investigate the implementation of nitride fuels containing little or no uranium in a fast-spectrum nuclear reactor to reduce the amount of plutonium and minor actinides in spent nuclear fuel destined for the Yucca...

  15. Determination of Uranium Metal Concentration in Irradiated Fuel Storage Basin Sludge Using Selective Dissolution

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Sinkov, Sergey I.; Chenault, Jeffrey W.; Schmidt, Andrew J.; Welsh, Terri L.; Pool, Karl N.

    2014-03-01T23:59:59.000Z

    Uranium metal corroding in water-saturated sludges now held in the US Department of Energy Hanford Site K West irradiated fuel storage basin can create hazardous hydrogen atmospheres during handling, immobilization, or subsequent transport and storage. Knowledge of uranium metal concentration in sludge thus is essential to safe sludge management and process design, requiring an expeditious routine analytical method to detect uranium metal concentrations as low as 0.03 wt% in sludge even in the presence of 30 wt% or higher total uranium concentrations.

  16. Engineering analysis of low enriched uranium fuel using improved zirconium hydride cross sections 

    E-Print Network [OSTI]

    Candalino, Robert Wilcox

    2006-10-30T23:59:59.000Z

    A neutronic and thermal hydraulic analysis of the 1-MW TRIGA research reactor at the Texas A&M University Nuclear Science Center using a new low enriched uranium fuel (named 30/20 fuel) was completed. This analysis provides ...

  17. DUSCOBS - a depleted-uranium silicate backfill for transport, storage, and disposal of spent nuclear fuel

    SciTech Connect (OSTI)

    Forsberg, C.W.; Pope, R.B.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

    1995-11-30T23:59:59.000Z

    A Depleted Uranium Silicate COntainer Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

  18. Uranium dioxide electrolysis

    DOE Patents [OSTI]

    Willit, James L. (Batavia, IL); Ackerman, John P. (Prescott, AZ); Williamson, Mark A. (Naperville, IL)

    2009-12-29T23:59:59.000Z

    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.

  19. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    DOE Patents [OSTI]

    Travelli, A.

    1985-10-25T23:59:59.000Z

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  20. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    DOE Patents [OSTI]

    Travelli, Armando (Hinsdale, IL)

    1988-01-01T23:59:59.000Z

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  1. Evaluation of health effects in Sequoyah Fuels Corporation workers from accidental exposure to uranium hexafluoride

    SciTech Connect (OSTI)

    Fisher, D.R. (Pacific Northwest Lab., Richland, WA (USA)); Swint, M.J.; Kathren, R.L. (Hanford Environmental Health Foundation, Richland, WA (USA))

    1990-05-01T23:59:59.000Z

    Urine bioassay measurements for uranium and medical laboratory results were studied to determine whether there were any health effects from uranium intake among a group of 31 workers exposed to uranium hexafluoride (UF{sub 6}) and hydrolysis products following the accidental rupture of a 14-ton shipping cylinder in early 1986 at the Sequoyah Fuels Corporation uranium conversion facility in Gore, Oklahoma. Physiological indicators studied to detect kidney tissue damage included tests for urinary protein, casts and cells, blood, specific gravity, and urine pH, blood urea nitrogen, and blood creatinine. We concluded after reviewing two years of follow-up medical data that none of the 31 workers sustained any observable health effects from exposure to uranium. The early excretion of uranium in urine showed more rapid systemic uptake of uranium from the lung than is assumed using the International Commission on Radiological Protection (ICRP) Publication 30 and Publication 54 models. The urinary excretion data from these workers were used to develop an improved systemic recycling model for inhaled soluble uranium. We estimated initial intakes, clearance rates, kidney burdens, and resulting radiation doses to lungs, kidneys, and bone surfaces. 38 refs., 10 figs., 7 tabs.

  2. Optimization of a seed and blanket thorium-uranium fuel cycle for pressurized water reactors

    E-Print Network [OSTI]

    Wang, Dean, 1971-

    2003-01-01T23:59:59.000Z

    A heterogeneous LWR core design, which employs a thorium/uranium once through fuel cycle, is optimized for good economics, wide safety margins, minimal waste burden and high proliferation resistance. The focus is on the ...

  3. Engineering analysis of low enriched uranium fuel using improved zirconium hydride cross sections

    E-Print Network [OSTI]

    Candalino, Robert Wilcox

    2006-10-30T23:59:59.000Z

    lifetime as the current high enriched uranium fuel and stay within the thermal and safety limits for the facility. It was also determined that the control rod worths and the temperature coefficient of reactivity would provide sufficient negative reactivity...

  4. THERMODYNAMIC MODEL FOR URANIUM DIOXIDE BASED NUCLEAR FUEL

    SciTech Connect (OSTI)

    Thompson, Dr. William T. [Royal Military College of Canada; Lewis, Dr. Brian J [Royal Military College of Canada; Corcoran, E. C. [Royal Military College of Canada; Kaye, Dr. Matthew H. [Royal Military College of Canada; White, S. J. [Royal Military College of Canada; Akbari, F. [Atomic Energy of Canada Limited, Chalk River Laboratories; Higgs, Jamie D. [Atomic Energy of Canada Limited, Point Lepreau; Thompson, D. M. [Praxair Inc.; Besmann, Theodore M [ORNL; Vogel, S. C. [Los Alamos National Laboratory (LANL)

    2007-01-01T23:59:59.000Z

    Many projects involving nuclear fuel rest on a quantitative understanding of the co-existing phases at various stages of burnup. Since the many fission products have considerably different abilities to chemically associate with oxygen, and the oxygen-to-metal molar ratio is slowly changing, the chemical potential of oxygen is a function of burnup. Concurrently, well-recognized small fractions of new phases such as inert gas, noble metals, zirconates, etc. also develop. To further complicate matters, the dominant UO2 fuel phase may be non-stoichiometric and most of the minor phases themselves have a variable composition dependent on temperature and possible contact with the coolant in the event of a sheathing breach. A thermodynamic fuel model to predict the phases in partially burned CANDU (CANada Deuterium Uranium) nuclear fuel containing many major fission products has been under development. The building blocks of the model are the standard Gibbs energies of formation of the many possible compounds expressed as a function of temperature. To these data are added mixing terms associated with the appearance of the component species in particular phases. In operational terms, the treatment rests on the ability to minimize the Gibbs energy in a multicomponent system, in our case using the algorithms developed by Eriksson. The model is capable of handling non-stoichiometry in the UO2 fluorite phase, dilute solution behaviour of significant solute oxides, noble metal inclusions, a second metal solid solution U(Pd-Rh-Ru)3, zirconate, molybdate, and uranate solutions as well as other minor solid phases, and volatile gaseous species.

  5. 300 AREA URANIUM CONTAMINATION

    SciTech Connect (OSTI)

    BORGHESE JV

    2009-07-02T23:59:59.000Z

    {sm_bullet} Uranium fuel production {sm_bullet} Test reactor and separations experiments {sm_bullet} Animal and radiobiology experiments conducted at the. 331 Laboratory Complex {sm_bullet} .Deactivation, decontamination, decommissioning,. and demolition of 300 Area facilities

  6. Influence of uranium hydride oxidation on uranium metal behaviour

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

    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)

  7. Depleted uranium management alternatives

    SciTech Connect (OSTI)

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

    1994-08-01T23:59:59.000Z

    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.

  8. URANIUM IN ALKALINE ROCKS

    E-Print Network [OSTI]

    Murphy, M.

    2011-01-01T23:59:59.000Z

    Greenland," in Uranium Exploration Geology, Int. AtomicMigration of Uranium and Thorium—Exploration Significance,"interesting for future uranium exploration. The c r i t e r

  9. Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-zirconium Alloys for Advanced Nuclear Fuel Applications 

    E-Print Network [OSTI]

    Garnetti, David J.

    2010-07-14T23:59:59.000Z

    The research in this thesis covers the design and implementation of a depleted uranium (DU) powder production system and the initial results of a DU-Zr-Mg alloy alpha phase sintering experiment where the Mg is a surrogate ...

  10. Fingerprinting Uranium | EMSL

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

    Fingerprinting Uranium Fingerprinting Uranium Researchers show how to use x-rays to identify mobile, stationary forms of atomic pollutant PNNL and University of North Texas...

  11. India's Worsening Uranium Shortage

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-01-15T23:59:59.000Z

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

  12. Uranium Industry Annual, 1992

    SciTech Connect (OSTI)

    Not Available

    1993-10-28T23:59:59.000Z

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

  13. Radiation measurements of uranium ingots from the electrometallurgical treatment of spent fuel.

    SciTech Connect (OSTI)

    Westphal, B. R.; Liaw, J. R.; Krsul, J. R.; Maddison, D. W.; Jensen, B. A.

    2003-03-24T23:59:59.000Z

    Radiation measurements and gamma spectroscopy analyses were made on numerous uranium ingots produced during the treatment of Experimental Breeder Reactor-II (EBR-II) spent nuclear fuel. The objective of these measurements was to provide background data for shielding concerns and potential process optimization. The uranium ingots resulted from the processing of both driver and blanket fuel by the electrometallurgical treatment process. The observed variation in the measurements was traced to the levels of certain fission product residues that remained in the uranium ingots produced during spent fuel treatment. A minor process change to hold the material at an elevated temperature for a specified length of time was found to significantly reduce concentrations of high-activity fission products and, thus the radiation field.

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

    SciTech Connect (OSTI)

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

    1981-10-01T23:59:59.000Z

    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.

  15. Electrochemical separation of aluminum from uranium for research reactor spent nuclear fuel applications.

    SciTech Connect (OSTI)

    Slater, S. A.; Willit, J. L.; Gay, E. C.; Chemical Engineering

    1999-01-01T23:59:59.000Z

    Researchers at Argonne National Laboratory (ANL) are developing an electrorefining process to treat aluminum-based spent nuclear fuel by electrochemically separating aluminum from uranium. The aluminum electrorefiner is modeled after the high-throughput electrorefiner developed at ANL. Aluminum is electrorefined, using a fluoride salt electrolyte, in a potential range of -0.1 V to -0.2 V, while uranium is electrorefined in a potential range of -0.3 V to -0.4 V; therefore, aluminum can be selectively separated electrochemically from uranium. A series of laboratory-scale experiments was performed to demonstrate the aluminum electrorefining concept. These experiments involved selecting an electrolyte (determining a suitable fluoride salt composition); selecting a crucible material for the electrochemical cell; optimizing the operating conditions; determining the effect of adding alkaline and rare earth elements to the electrolyte; and demonstrating the electrochemical separation of aluminum from uranium, using a U-Al-Si alloy as a simulant for aluminum-based spent nuclear fuel. Results of the laboratory-scale experiments indicate that aluminum can be selectively electrotransported from the anode to the cathode, while uranium remains in the anode basket.

  16. Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

    SciTech Connect (OSTI)

    Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

    2012-07-30T23:59:59.000Z

    Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium is more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.

  17. Uranium hexafluoride handling. Proceedings

    SciTech Connect (OSTI)

    Not Available

    1991-12-31T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Lloyd, R.

    1980-10-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Sean M. McDeavitt

    2011-04-29T23:59:59.000Z

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500şC to 600şC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Dis

  20. Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-zirconium Alloys for Advanced Nuclear Fuel Applications

    E-Print Network [OSTI]

    Garnetti, David J.

    2010-07-14T23:59:59.000Z

    The research in this thesis covers the design and implementation of a depleted uranium (DU) powder production system and the initial results of a DU-Zr-Mg alloy alpha phase sintering experiment where the Mg is a surrogate for Pu and Am. The powder...

  1. Environmental effects of the uranium fuel cycle: a review of data for technetium

    SciTech Connect (OSTI)

    Till, J.E.; Shor, R.W.; Hoffman, F.O.

    1984-01-01T23:59:59.000Z

    Sources of potential releases of /sup 99/Tc to the environment are reviewed for the uranium fuel cycle considering two options: the recycle of spent uranium fuel and no fuel recycling. In the no recycle option, the only source of /sup 99/Tc release is an extremely small amount associated with airborne emissions from the processing of high-level wastes. With recycling, /sup 99/Tc releases are associated with the operation of reprocessing facilities, UF/sub 6/ conversion plants, uranium enrichment plants, fuel fabrication facilities, and low- and high-level waste processing and storage facilities. Among these, the most prominent /sup 99/Tc releases are from the liquid effluents of uranium enrichment facilities (0.22 Ci per reference reactor year). A review of parameters of importance for predicting the environmental behavior and fate of /sup 99/Tc indicates a substantial reduction from earlier estimates of the radiological significance of exposure pathways involving the ingestion of milk and meat. More important routes of exposure to /sup 99/Tc will probably be associated with drinking water and the consumption of aquatic organisms, garden vegetables, and eggs. For each parameter reviewed in this study, a range of values is recommended for radiological assessment calculations. Where obvious discrepancies exist between these ranges and the default values listed in USNRC Regulatory Guide 1.109, consideration for revision of the USNRC default values is recommended.

  2. New method of uranium and plutonium extraction in reprocessing of the spent nuclear fuel

    SciTech Connect (OSTI)

    Volk, V.; Dvoeglazov, K.; Veslov, S.; Rubisov, V. [JSC - VNIINM Bochvar, Moscow (Russian Federation); Alekseenko, V. [FSUE - Federal Nuclear and Radiation Safety Center, Moscow (Russian Federation); Krivitsky, Y.; Alekseenko, S.; Bondin, V. [FSUE - Mining and Chemical Combine, Zheleznogorsk (Russian Federation)

    2013-07-01T23:59:59.000Z

    It is shown that a two-stage process of uranium and plutonium extraction during the reprocessing of spent nuclear fuel solves the problem of obtaining a high-concentrated extract without increasing the loss risk with raffinate and avoids the accumulation of plutonium in the unit. A possible further optimization of the process would be the creation of steps inside the stages.

  3. Uranium industry annual 1994

    SciTech Connect (OSTI)

    NONE

    1995-07-05T23:59:59.000Z

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

  4. Thermal analysis of uranium zirconium hydride fuel using a lead-bismuth gap at LWR operating temperatures

    E-Print Network [OSTI]

    Ensor, Brendan M. (Brendan Melvin)

    2012-01-01T23:59:59.000Z

    Next generation nuclear technology calls for more advanced fuels to maximize the effectiveness of new designs. A fuel currently being studied for use in advanced light water reactors (LWRs) is uranium zirconium hydride ...

  5. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect (OSTI)

    Chodak, P. III

    1996-05-01T23:59:59.000Z

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  6. Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

    SciTech Connect (OSTI)

    Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

    2003-08-01T23:59:59.000Z

    The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer (''blanket'') on the uranium metal corrosion rates were also evaluated.

  7. Use of depleted uranium silicate glass to minimize release of radionuclides from spent nuclear fuel waste packages

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-01-20T23:59:59.000Z

    A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

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

    SciTech Connect (OSTI)

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01T23:59:59.000Z

    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.

  9. Progress in developing very-high-density low-enriched-uranium fuels.

    SciTech Connect (OSTI)

    Hayes, S. L.; Hofman, G. L.; Meyer, M. K; Snelgrove, J. L.; Strain, R. V.; Wiencek, T. C.

    1999-02-19T23:59:59.000Z

    Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-MO alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm{sup 3}. Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 C in 8-g U/cm{sup 3} fuel.

  10. Initial assessment of radiation behavior of very-high-density low-enriched-uranium fuels.

    SciTech Connect (OSTI)

    Hofman, G. L.; Meyer, M. L.; Snelgrove, J. L.; Dietz, M. L.; Strain, R. V.; Kim, K. H.

    1999-10-01T23:59:59.000Z

    Results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm{sup 3}. Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 C)are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 C in 8-g U/cm{sup 3} fuel.

  11. Depleted Uranium Technical Brief

    E-Print Network [OSTI]

    Depleted Uranium Technical Brief United States Environmental Protection Agency Office of Air and Radiation Washington, DC 20460 EPA-402-R-06-011 December 2006 #12;#12;Depleted Uranium Technical Brief EPA of Radiation and Indoor Air Radiation Protection Division ii #12;iii #12;FOREWARD The Depleted Uranium

  12. Standard test method for atom percent fission in uranium and plutonium fuel (Neodymium-148 Method)

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    1996-01-01T23:59:59.000Z

    1.1 This test method covers the determination of stable fission product 148Nd in irradiated uranium (U) fuel (with initial plutonium (Pu) content from 0 to 50 %) as a measure of fuel burnup (1-3). 1.2 It is possible to obtain additional information about the uranium and plutonium concentrations and isotopic abundances on the same sample taken for burnup analysis. If this additional information is desired, it can be obtained by precisely measuring the spike and sample volumes and following the instructions in Test Method E267. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  13. Neutronics studies of uranium-based fully ceramic micro-encapsulated fuel for PWRs

    SciTech Connect (OSTI)

    George, N. M.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States); Terrani, K.; Godfrey, A.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-01T23:59:59.000Z

    This study evaluates the core neutronics and fuel cycle characteristics using uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR assembly designs with FCM fuel have been developed, which by virtue of their TRISO particle-based elements are expected to achieve higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software used to model the assembly designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities; however, the Reactivity-Equivalent Physical Transformation (RPT) method was used for lattice calculations due to the long run times associated with the SCALE DH capability. In order to understand the impact on reactivity and reactor operating cycle length, a parametric study was performed by varying TRISO particle design features, such as kernel diameter, coating layer thicknesses, and packing fraction. Also, other features such as the selection of matrix material (SiC, zirconium) and fuel rod dimensions were studied. After evaluating different uranium-based fuels, the higher compound density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime and temperature coefficients of reactivity, as well as pin cell and assembly peaking factors. (authors)

  14. Method for converting uranium oxides to uranium metal

    DOE Patents [OSTI]

    Duerksen, Walter K. (Norris, TN)

    1988-01-01T23:59:59.000Z

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

    DOE Patents [OSTI]

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

    1997-12-16T23:59:59.000Z

    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.

  16. 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)

    1997-01-01T23:59:59.000Z

    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.

  17. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1997-03-01T23:59:59.000Z

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs.

  18. Benefits/impacts of utilizing depleted uranium silicate glass as backfill for spent fuel waste packages

    SciTech Connect (OSTI)

    Pope, R.B.; Forsberg, C.W.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

    1996-05-01T23:59:59.000Z

    An assessment has been made of the benefits and impacts which can be derived by filling a spent nuclear fuel multi-purpose canister with depleted uranium silicate (DUS) glass at a reactor site. Although the primary purpose of the DUS glass fill would be to enhance repository performance assessment and control criticality of geologic times, a number of benefits to the waste management system can be derived from adding the DUS glass prior to shipment from the reactor site.

  19. Optical Constants ofOptical Constants of Uranium Nitride Thin FilmsUranium Nitride Thin Films

    E-Print Network [OSTI]

    Hart, Gus

    Optical Constants ofOptical Constants of Uranium Nitride Thin FilmsUranium Nitride Thin FilmsDelta--Beta Scatter Plot at 220 eVBeta Scatter Plot at 220 eV #12;Why Uranium Nitride?Why Uranium Nitride? UraniumUranium, uranium,Bombard target, uranium, with argon ionswith argon ions Uranium atoms leaveUranium atoms leave

  20. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect (OSTI)

    Bi, G.; Liu, C.; Si, S. [Shanghai Nuclear Engineering Research and Design Inst., No. 29, Hongcao Road, Shanghai, 200233 (China)

    2012-07-01T23:59:59.000Z

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)

  1. Uranium resource utilization improvements in the once-through PWR fuel cycle

    SciTech Connect (OSTI)

    Matzie, R A [ed.

    1980-04-01T23:59:59.000Z

    In support of the Nonproliferation Alternative Systems Assessment Program (NASAP), Combustion Engineering, Inc. performed a comprehensive analytical study of potential uranium utilization improvement options that can be backfit into existing PWRs operating on the once-through uranium fuel cycle. A large number of potential improvement options were examined as part of a preliminary survey of candidate options. The most attractive of these, from the standpoint of uranium utilization improvement, economic viability, and ease of implementation, were then selected for detailed analysis and were included in a single composite improvement case. This composite case represents an estimate of the total savings in U/sub 3/O/sub 8/ consumption that can be achieved in current-design PWRs by implementing improvements which can be developed and demonstrated in the near term. The improvement options which were evaluated in detail and included in the composite case were a new five-batch, extended-burnup fuel management scheme, low-leakage fuel management, modified lattice designs, axial blankets, reinsertion of initial core batches, and end-of-cycle stretchout.

  2. Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

    SciTech Connect (OSTI)

    Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M. [Candu Energy Inc., 2285 Speakman Drive, Mississauga, ON L5K 1B1 (Canada)

    2012-07-01T23:59:59.000Z

    The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

  3. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of uranium

  4. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of uranium4.

  5. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries2.5.3. Uranium

  6. NEUTRONICS STUDIES OF URANIUM-BASED FULLY CERAMIC MICRO-ENCAPSULATED FUEL FOR PWRs

    SciTech Connect (OSTI)

    George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Godfrey, Andrew T [ORNL

    2012-01-01T23:59:59.000Z

    This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature coefficients of reactivity, as well as pin cell and assembly peaking factors. Key Words: FCM, TRISO, Uranium Mononitride, PWR

  7. Production of small uranium dioxide microspheres for cermet nuclear fuel using the internal gelation process

    SciTech Connect (OSTI)

    Collins, Robert T [ORNL] [ORNL; Collins, Jack Lee [ORNL] [ORNL; Hunt, Rodney Dale [ORNL] [ORNL; Ladd-Lively, Jennifer L [ORNL] [ORNL; Patton, Kaara K [ORNL] [ORNL; Hickman, Robert [NASA Marshall Space Flight Center, Huntsville, AL] [NASA Marshall Space Flight Center, Huntsville, AL

    2014-01-01T23:59:59.000Z

    The U.S. National Aeronautics and Space Administration (NASA) is developing a uranium dioxide (UO2)/tungsten cermet fuel for potential use as the nuclear cryogenic propulsion stage (NCPS). The first generation NCPS is expected to be made from dense UO2 microspheres with diameters between 75 and 150 m. Previously, the internal gelation process and a hood-scale apparatus with a vibrating nozzle were used to form gel spheres, which became UO2 kernels with diameters between 350 and 850 m. For the NASA spheres, the vibrating nozzle was replaced with a custom designed, two-fluid nozzle to produce gel spheres in the desired smaller size range. This paper describes the operational methodology used to make 3 kg of uranium oxide microspheres.

  8. Uranium hexafluoride public risk

    SciTech Connect (OSTI)

    Fisher, D.R.; Hui, T.E.; Yurconic, M.; Johnson, J.R.

    1994-08-01T23:59:59.000Z

    The limiting value for uranium toxicity in a human being should be based on the concentration of uranium (U) in the kidneys. The threshold for nephrotoxicity appears to lie very near 3 {mu}g U per gram kidney tissue. There does not appear to be strong scientific support for any other improved estimate, either higher or lower than this, of the threshold for uranium nephrotoxicity in a human being. The value 3 {mu}g U per gram kidney is the concentration that results from a single intake of about 30 mg soluble uranium by inhalation (assuming the metabolism of a standard person). The concentration of uranium continues to increase in the kidneys after long-term, continuous (or chronic) exposure. After chronic intakes of soluble uranium by workers at the rate of 10 mg U per week, the concentration of uranium in the kidneys approaches and may even exceed the nephrotoxic limit of 3 {mu}g U per gram kidney tissue. Precise values of the kidney concentration depend on the biokinetic model and model parameters assumed for such a calculation. Since it is possible for the concentration of uranium in the kidneys to exceed 3 {mu}g per gram tissue at an intake rate of 10 mg U per week over long periods of time, we believe that the kidneys are protected from injury when intakes of soluble uranium at the rate of 10 mg U per week do not continue for more than two consecutive weeks. For long-term, continuous occupational exposure to low-level, soluble uranium, we recommend a reduced weekly intake limit of 5 mg uranium to prevent nephrotoxicity in workers. Our analysis shows that the nephrotoxic limit of 3 {mu}g U per gram kidney tissues is not exceeded after long-term, continuous uranium intake at the intake rate of 5 mg soluble uranium per week.

  9. Uranium Transport in a High-Throughput Electrorefiner for EBR-II Blanket Fuel

    SciTech Connect (OSTI)

    Ahluwalia, Rajesh K.; Hua, Thanh Q.; Vaden, DeeEarl [Argonne National Laboratory (United States)

    2004-01-15T23:59:59.000Z

    A unique high-throughput Mk-V electrorefiner is being used in the electrometallurgical treatment of the metallic sodium-bonded blanket fuel from the Experimental Breeder Reactor II. Over many cycles, it transports uranium back and forth between the anodic fuel dissolution baskets and the cathode tubes until, because of imperfect adherence of the dendrites, it all ends up in the product collector at the bottom. The transport behavior of uranium in the high-throughput electrorefiner can be understood in terms of the sticking coefficients for uranium adherence to the cathode tubes in the forward direction and to the dissolution baskets in the reverse direction. The sticking coefficients are inferred from the experimental voltage and current traces and are correlated in terms of a single parameter representing the ratio of the cell current to the limiting current at the surface acting as the cathode. The correlations are incorporated into an engineering model that calculates the transport of uranium in the different modes of operation. The model also uses the experimentally derived electrorefiner operating maps that describe the relationship between the cell voltage and the cell current for the three principal transport modes. It is shown that the model correctly simulates the cycle-to-cycle variation of the voltage and current profiles. The model is used to conduct a parametric study of electrorefiner throughput rate as a function of the principal operating parameters. The throughput rate is found to improve with lowering of the basket rotation speed, reduction of UCl{sub 3} concentration in salt, and increasing the maximum cell current or cut-off voltage. Operating conditions are identified that can improve the throughput rate by 60 to 70% over that achieved at present.

  10. The End of Cheap Uranium

    E-Print Network [OSTI]

    Michael Dittmar

    2011-06-21T23:59:59.000Z

    Historic data from many countries demonstrate that on average no more than 50-70% of the uranium in a deposit could be mined. An analysis of more recent data from Canada and Australia leads to a mining model with an average deposit extraction lifetime of 10+- 2 years. This simple model provides an accurate description of the extractable amount of uranium for the recent mining operations. Using this model for all larger existing and planned uranium mines up to 2030, a global uranium mining peak of at most 58 +- 4 ktons around the year 2015 is obtained. Thereafter we predict that uranium mine production will decline to at most 54 +- 5 ktons by 2025 and, with the decline steepening, to at most 41 +- 5 ktons around 2030. This amount will not be sufficient to fuel the existing and planned nuclear power plants during the next 10-20 years. In fact, we find that it will be difficult to avoid supply shortages even under a slow 1%/year worldwide nuclear energy phase-out scenario up to 2025. We thus suggest that a worldwide nuclear energy phase-out is in order. If such a slow global phase-out is not voluntarily effected, the end of the present cheap uranium supply situation will be unavoidable. The result will be that some countries will simply be unable to afford sufficient uranium fuel at that point, which implies involuntary and perhaps chaotic nuclear phase-outs in those countries involving brownouts, blackouts, and worse.

  11. Global terrestrial uranium supply and its policy implications : a probabilistic projection of future uranium costs

    E-Print Network [OSTI]

    Matthews, Isaac A

    2010-01-01T23:59:59.000Z

    An accurate outlook on long-term uranium resources is critical in forecasting uranium costresource relationships, and for energy policy planning as regards the development and deployment of nuclear fuel cycle alternatives. ...

  12. Uranium Mill Tailings Management

    SciTech Connect (OSTI)

    Nelson, J.D.

    1982-01-01T23:59:59.000Z

    This book presents the papers given at the Fifth Symposium on Uranium Mill Tailings Management. Advances made with regard to uranium mill tailings management, environmental effects, regulations, and reclamation are reviewed. Topics considered include tailings management and design (e.g., the Uranium Mill Tailings Remedial Action Project, environmental standards for uranium mill tailings disposal), surface stabilization (e.g., the long-term stability of tailings, long-term rock durability), radiological aspects (e.g. the radioactive composition of airborne particulates), contaminant migration (e.g., chemical transport beneath a uranium mill tailings pile, the interaction of acidic leachate with soils), radon control and covers (e.g., radon emanation characteristics, designing surface covers for inactive uranium mill tailings), and seepage and liners (e.g., hydrologic observations, liner requirements).

  13. Mr. William f. Crow, Acting Director . Uranium Fuel Licensing Branch

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling7 August 2008 Office7-TACi+J-UN 2DCT i* ,,45

  14. German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC) Environmental AssessmentsGeoffrey CampbelllongApplyingGeorgeGeothermal Potential

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

    DOE Patents [OSTI]

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

    1995-06-06T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1995-01-01T23:59:59.000Z

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

  17. Preparation of uranium compounds

    DOE Patents [OSTI]

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

    2013-02-19T23:59:59.000Z

    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.

  18. Uranium Ore Uranium is extracted

    E-Print Network [OSTI]

    be discharged to water. Radioactive Wastes--Wastes managed for their radioactive content. Spent Nuclear Fuels--Fuel plants with reactors that use water for moderating nuclear reactions and cooling. Spent Nuclear Fuel Used or"spent"nuclear fuel is stored in pools, or in specially designed dry storage casks. Fabrication

  19. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect (OSTI)

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01T23:59:59.000Z

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  20. CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS

    E-Print Network [OSTI]

    Hart, Gus

    CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS by David T. Oliphant. Woolley Dean, College of Physical and Mathematical Sciences #12;ABSTRACT CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS David T. Oliphant Department of Physics and Astronomy

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

    SciTech Connect (OSTI)

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

    2012-03-01T23:59:59.000Z

    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.

  2. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Friedman, H.A.

    1984-06-13T23:59:59.000Z

    A method for decontaminating uranium product from the Purex 5 process comprises addition of hydrazine to the product uranyl nitrate stream from the Purex process, which contains hexavalent (UO/sub 2//sup 2 +/) uranium and heptavalent technetium (TcO/sub 4/-). Technetium in the product stream is reduced and then complexed by the addition of oxalic acid (H/sub 2/C/sub 2/O/sub 4/), and the Tc-oxalate complex is readily separated from the 10 uranium by solvent extraction with 30 vol % tributyl phosphate in n-dodecane.

  3. EPA Update: NESHAP Uranium Activities

    E-Print Network [OSTI]

    EPA Update: NESHAP Uranium Activities Reid J. Rosnick Environmental Protection Agency Radiation Protection Division (6608J) Washington, DC 20460 NMA/NRC Uranium Recovery Workshop July 2, 2009 #12 for underground uranium mining operations (Subpart B) EPA regulatory requirements for operating uranium mill

  4. Improved Irradiation Performance of Uranium-Molybdenum/Aluminum Dispersion Fuel by Silicon Addition in Aluminum

    SciTech Connect (OSTI)

    Yeon Soo Kim; G. L. Hofman; A. B. Robinson; D. M. Wachs

    2013-10-01T23:59:59.000Z

    Uranium-molybdenum fuel particle dispersion in aluminum is a form of fuel under development for conversion of high-power research and test reactors from highly enriched to low-enriched uranium in the U.S. Global Threat Reduction Initiative program (also known as the Reduced Enrichment for Research and Test Reactors program). Extensive irradiation tests have been conducted to find a solution for problems caused by interaction layer growth and pore formation between U-Mo and Al. Adding a small amount of Si (up to [approximately]5 wt%) in the Al matrix was one of the proposed remedies. The effect of silicon addition in the Al matrix was examined using irradiation test results by comparing side-by-side samples with different Si additions. Interaction layer growth was progressively reduced with increasing Si addition to the matrix Al, up to 4.8 wt%. The Si addition also appeared to delay pore formation and growth between the U-Mo and Al.

  5. RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

    SciTech Connect (OSTI)

    Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

    2009-07-01T23:59:59.000Z

    In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

  6. WISE Uranium Project - Fact Sheet

    E-Print Network [OSTI]

    Hazards From Depleted

    t in the depleted uranium. For this purpose, we first need to calculate the mass balance of the enrichment process. We then calculate the inhalation doses from the depleted uranium and compare the dose contributions from the nuclides of interest. Mass balance for uranium enrichment at Paducah [DOE_1984, p.35] Feed Product Tails Other Mass [st] 758002 124718 621894 11390 Mass fraction 100.00% 16.45% 82.04% 1.50% Concentration of plutonium in tails (depleted uranium) from enrichment of reprocessed uranium, assuming that all plutonium were transfered to the tails: Concentration of neptunium in tails from enrichment of reprocessed uranium uranium, assuming that all neptunium were transfered to the tails: - 2 - Schematic of historic uranium enrichment process at Paducah [DOE_1999b] - -7 For comparison, we first calculate the inhalation dose from depleted uranium produced from natural uranium. We assume that the short-lived decay products have reached secular equilibrium with th

  7. Fabrication and Characterization of Uranium-based High Temperature...

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

    Fabrication and Characterization of Uranium-based High Temperature Reactor Fuel June 01, 2013 The Uranium Fuel Development Laboratory is a modern R&D scale lab for the fabrication...

  8. Comparison of potential radiological consequences from a spent-fuel repository and natural uranium deposits

    SciTech Connect (OSTI)

    Wick, O.J.; Cloninger, M.O.

    1980-09-01T23:59:59.000Z

    A general criterion has been suggested for deep geological repositories containing spent fuel - the repositories should impose no greater radiological risk than due to naturally occurring uranium deposits. The following analysis investigates the rationale of that suggestion and determines whether current expectations of spent-fuel repository performance are consistent with such a criterion. In this study, reference spent-fuel repositories were compared to natural uranium-ore deposits. Comparisons were based on intrinsic characteristics, such as radionuclide inventory, depth, proximity to aquifers, and regional distribution, and actual and potential radiological consequences that are now occurring from some ore deposits and that may eventually occur from repositories and other ore deposits. The comparison results show that the repositories are quite comparable to the natural ore deposits and, in some cases, present less radiological hazard than their natural counterparts. On the basis of the first comparison, placing spent fuel in a deep geologic repository apparently reduces the hazard from natural radioactive materials occurring in the earth's crust by locating the waste in impermeable strata without access to oxidizing conditions. On the basis of the second comparison, a repository constructed within reasonable constraints presents no greater hazard than a large ore deposit. It is recommended that if the naturally radioactive environment is to be used as a basis for a criterion regarding repositories, then this criterion should be carefully constructed. The criterion should be based on the radiological quality of the waters in the immediate region of a specific repository, and it should be in terms of an acceptable potential increase in the radiological content of those waters due to the existence of the repository.

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

    DOE Patents [OSTI]

    Kreuzmann, Alvin B. (Cincinnati, OH)

    1983-01-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Kreuzmann, A.B.

    1982-10-27T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1999-02-01T23:59:59.000Z

    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.

  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

    2011-05-01T23:59:59.000Z

    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 2009 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2010-01-01T23:59:59.000Z

    With several countries currently building nuclear power plants and planning the construction of more to meet long-term increases in electricity demand, uranium resources, production and demand remain topics of notable interest. In response to the projected growth in demand for uranium and declining inventories, the uranium industry – the first critical link in the fuel supply chain for nuclear reactors – is boosting production and developing plans for further increases in the near future. Strong market conditions will, however, be necessary to trigger the investments required to meet projected demand. The "Red Book", jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, is a recognised world reference on uranium. It is based on information compiled in 40 countries, including those that are major producers and consumers of uranium. This 23rd edition provides a comprehensive review of world uranium supply and demand as of 1 January 2009, as well as data on global ur...

  14. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

    2014-10-30T23:59:59.000Z

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.

  15. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    SciTech Connect (OSTI)

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01T23:59:59.000Z

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  16. Biogeochemical Processes In Ethanol Stimulated Uranium Contaminated...

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

    Processes In Ethanol Stimulated Uranium Contaminated Subsurface Sediments. Biogeochemical Processes In Ethanol Stimulated Uranium Contaminated Subsurface Sediments. Abstract: A...

  17. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1989-01-01T23:59:59.000Z

    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.

  18. Controlling uranium reactivity March 18, 2008

    E-Print Network [OSTI]

    Meyer, Karsten

    for the last decade. Most of their work involves depleted uranium, a more common form of uraniumMarch 2008 Controlling uranium reactivity March 18, 2008 Uranium is an often misunderstood metal uranium research. In reality, uranium presents a wealth of possibilities for funda- mental chemistry. Many

  19. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    SciTech Connect (OSTI)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal,; Monado, Fiber; Sekimoto, Hiroshi [Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134 (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Reserach of Laboratory for Nuclear Reactors, Tokyo Institute of Technology O-okayama, Meguro-ku, Tokyo 152 (Japan)

    2012-06-06T23:59:59.000Z

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

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

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Guida, Tracey [University of Pittsburgh

    2010-02-01T23:59:59.000Z

    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.

  1. Uranium resources: Issues and facts

    SciTech Connect (OSTI)

    Delene, J.G.

    1993-12-31T23:59:59.000Z

    Although there are several secondary issues, the most important uranium resource issue is, ``will there be enough uranium available at a cost which will allow nuclear power to be competitive in the future?`` This paper will attempt to answer this question by discussing uranium supply, demand, and economics from the perspective of the United States. The paper will discuss: how much uranium is available; the sensitivity of nuclear power costs to uranium price; the potential future demand for uranium in the Unites States, some of the options available to reduce this demand, the potential role of the Advanced Liquid Metal Cooled Reactor (ALMR) in reducing uranium demand; and potential alternative uranium sources and technologies.

  2. Uranium-titanium-niobium alloy

    DOE Patents [OSTI]

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

    1990-01-01T23:59:59.000Z

    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.

  3. Uranium deposits of Brazil

    SciTech Connect (OSTI)

    NONE

    1991-09-01T23:59:59.000Z

    Brazil is a country of vast natural resources, including numerous uranium deposits. In support of the country`s nuclear power program, Brazil has developed the most active uranium industry in South America. Brazil has one operating reactor (Angra 1, a 626-MWe PWR), and two under construction. The country`s economic challenges have slowed the progress of its nuclear program. At present, the Pocos de Caldas district is the only active uranium production. In 1990, the Cercado open-pit mine produced approximately 45 metric tons (MT) U{sub 3}O{sub 8} (100 thousand pounds). Brazil`s state-owned uranium production and processing company, Uranio do Brasil, announced it has decided to begin shifting its production from the high-cost and nearly depleted deposits at Pocos de Caldas, to lower-cost reserves at Lagoa Real. Production at Lagoa Real is schedules to begin by 1993. In addition to these two districts, Brazil has many other known uranium deposits, and as a whole, it is estimated that Brazil has over 275,000 MT U{sub 3}O{sub 8} (600 million pounds U{sub 3}O{sub 8}) in reserves.

  4. Uranium immobilization and nuclear waste

    SciTech Connect (OSTI)

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

    1982-02-01T23:59:59.000Z

    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.

  5. EA-1172: Sale of Surplus Natural and Low Enriched Uranium, Piketon, Ohio

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to sell uranium for subsequent enrichment and fabrication into commercial nuclear power reactor fuel.  The uranium is currently stored...

  6. Uranium Recovery from Seawater: Development of Fiber Adsorbents Prepared via Atom-Transfer Radical Polymerization

    SciTech Connect (OSTI)

    Saito, Tomonori; Brown, Suree; Chatterjee, Sabornie; Kim, Jungseung; Tsouris, Constantinos; Mayes, Richard; Kuo, Li-Jung; Gill, Gary A.; Oyola, Yatsandra; Janke, C.; Dai, Sheng

    2014-07-09T23:59:59.000Z

    Uranium exists uniformly at a concentration of ~3.3 ppb in seawater. The extraction of uranium from seawater presents a very attractive alternative source of uranium for nuclear fuel needs.

  7. Corrosion-resistant uranium

    DOE Patents [OSTI]

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

    1981-10-21T23:59:59.000Z

    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.

  8. Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors

    SciTech Connect (OSTI)

    Biswas, D; Mennerdahl, D

    2008-06-23T23:59:59.000Z

    The ANSI/ANS 8.12 standard was first approved in July 1978. At that time, this edition was applicable to operations with plutonium-uranium oxide (MOX) fuel mixtures outside reactors and was limited to subcritical limits for homogeneous systems. The next major revision, ANSI/ANS-8.12-1987, included the addition of subcritical limits for heterogeneous systems. The standard was subsequently reaffirmed in February 1993. During late 1990s, substantial work was done by the ANS 8.12 Standard Working Group to re-examine the technical data presented in the standard using the latest codes and cross section sets. Calculations performed showed good agreement with the values published in the standard. This effort resulted in the reaffirmation of the standard in March 2002. The standard is currently in a maintenance mode. After 2002, activities included discussions to determine the future direction of the standard and to follow the MOX standard development by the International Standard Organization (ISO). In 2007, the Working Group decided to revise the standard to extend the areas of applicability by providing a wider range of subcritical data. The intent is to cover a wider domain of MOX fuel fabrication and operations. It was also decided to follow the ISO MOX standard specifications (related to MOX density and isotopics) and develop a new set of subcritical limits for homogeneous systems. This has resulted in the submittal (and subsequent approval) of the project initiation notification system form (PINS) in 2007.

  9. Molten-Salt Depleted-Uranium Reactor

    E-Print Network [OSTI]

    Dong, Bao-Guo; Gu, Ji-Yuan

    2015-01-01T23:59:59.000Z

    The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

  10. Bioremediation of uranium contaminated soils and wastes

    SciTech Connect (OSTI)

    Francis, A.J.

    1998-12-31T23:59:59.000Z

    Contamination of soils, water, and sediments by radionuclides and toxic metals from uranium mill tailings, nuclear fuel manufacturing and nuclear weapons production is a major concern. Studies of the mechanisms of biotransformation of uranium and toxic metals under various microbial process conditions has resulted in the development of two treatment processes: (1) stabilization of uranium and toxic metals with reduction in waste volume and (2) removal and recovery of uranium and toxic metals from wastes and contaminated soils. Stabilization of uranium and toxic metals in wastes is accomplished by exploiting the unique metabolic capabilities of the anaerobic bacterium, Clostridium sp. The radionuclides and toxic metals are solubilized by the bacteria directly by enzymatic reductive dissolution, or indirectly due to the production of organic acid metabolites. The radionuclides and toxic metals released into solution are immobilized by enzymatic reductive precipitation, biosorption and redistribution with stable mineral phases in the waste. Non-hazardous bulk components of the waste volume. In the second process uranium and toxic metals are removed from wastes or contaminated soils by extracting with the complexing agent citric acid. The citric-acid extract is subjected to biodegradation to recover the toxic metals, followed by photochemical degradation of the uranium citrate complex which is recalcitrant to biodegradation. The toxic metals and uranium are recovered in separate fractions for recycling or for disposal. The use of combined chemical and microbiological treatment process is more efficient than present methods and should result in considerable savings in clean-up and disposal costs.

  11. Uranium from seawater

    SciTech Connect (OSTI)

    Gregg, D.; Folkendt, M.

    1982-09-21T23:59:59.000Z

    A novel process for recovering uranium from seawater is proposed and some of the critical technical parameters are evaluated. The process, in summary, consists of two different options for contacting adsorbant pellets with seawater without pumping the seawater. It is expected that this will reduce the mass handling requirements, compared to pumped seawater systems, by a factor of approximately 10/sup 5/, which should also result in a large reduction in initial capital investment. Activated carbon, possibly in combination with a small amount of dissolved titanium hydroxide, is expected to be the preferred adsorbant material instead of the commonly assumed titanium hydroxide alone. The activated carbon, after exposure to seawater, can be stripped of uranium with an appropriate eluant (probably an acid) or can be burned for its heating value (possible in a power plant) leaving the uranium further enriched in its ash. The uranium, representing about 1% of the ash, is then a rich ore and would be recovered in a conventional manner. Experimental results have indicated that activated carbon, acting alone, is not adequately effective in adsorbing the uranium from seawater. We measured partition coefficients (concentration ratios) of approximately 10/sup 3/ in seawater instead of the reported values of 10/sup 5/. However, preliminary tests carried out in fresh water show considerable promise for an extraction system that uses a combination of dissolved titanium hydroxide (in minute amounts) which forms an insoluble compound with the uranyl ion, and the insoluble compound then being sorbed out on activated carbon. Such a system showed partition coefficients in excess of 10/sup 5/ in fresh water. However, the system was not tested in seawater.

  12. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    SciTech Connect (OSTI)

    Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12T23:59:59.000Z

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  13. NEUTRALIZATIONS OF HIGH ALUMINUM LOW URANIUM USED NUCLEAR FUEL SOLUTIONS CONTAINING GADOLINIUM AS A NEUTRON POISON

    SciTech Connect (OSTI)

    Taylor-Pashow, K.

    2011-06-08T23:59:59.000Z

    H-Canyon will begin dissolving High Aluminum - Low Uranium (High Al/Low U) Used Nuclear Fuel (UNF) following approval by DOE which is anticipated in CY2011. High Al/Low U is an aluminum/enriched uranium UNF with small quantities of uranium relative to aluminum. The maximum enrichment level expected is 93% {sup 235}U. The High Al/Low U UNF will be dissolved in H-Canyon in a nitric acid/mercury/gadolinium solution. The resulting solution will be neutralized and transferred to Tank 39H in the Tank Farm. To confirm that the solution generated could be poisoned with Gd, neutralized, and discarded to the Savannah River Site (SRS) high level waste (HLW) system without undue nuclear safety concerns the caustic precipitation of simulant solutions was examined. Experiments were performed with three simulant solutions representative of the H-Canyon estimated concentrations in the final solutions after dissolution. The maximum U, Gd, and Al concentration were selected for testing from the range of solution compositions provided. Simulants were prepared in three different nitric acid concentrations, ranging from 0.5 to 1.5 M. The simulant solutions were neutralized to four different endpoints: (1) just before a solid phase was formed (pH 3.5-4), (2) the point where a solid phase was obtained, (3) 0.8 M free hydroxide, and (4) 1.2 M free hydroxide, using 50 wt % sodium hydroxide (NaOH). The settling behavior of the neutralized solutions was found to be slower compared to previous studies, with settling continuing over a one week period. Due to the high concentration of Al in these solutions, precipitation of solids was observed immediately upon addition of NaOH. Precipitation continued as additional NaOH was added, reaching a point where the mixture becomes almost completely solid due to the large amount of precipitate. As additional NaOH was added, some of the precipitate began to redissolve, and the solutions neutralized to the final two endpoints mixed easily and had expected densities of typical neutralized waste. Based on particle size and scanning electron microscopy analyses, the neutralized solids were found to be homogeneous and less than 20 microns in size. The majority of solids were less than 4 microns in size. Compared to previous studies, a larger percentage of the Gd was found to precipitate in the partially neutralized solutions (at pH 3.5-4). In addition the Gd:U mass ratio was found to be at least 1.0 in all of the solids obtained after partial or full neutralization. The hydrogen to U (H:U) molar ratios for two accident scenarios were also determined. The first was for transient neutralization and agitator failure. Experimentally this scenario was determined by measuring the H:U ratio of the settled solids. The minimum H:U molar ratio for solids from fully neutralized solutions was 388:1. The second accident scenario is for the solids drying out in an unagitiated pump box. Experimentally, this scenario was determined by measuring the H:U molar ratio in centrifuged solids. The minimum H:U atom ratios for centrifuged precipitated solids was 250:1. It was determined previously that a 30:1 H:Pu atom ratio was sufficient for a 1:1 Gd:Pu mass ratio. Assuming a 1:1 equivalence with {sup 239}Pu, the results of these experiments show Gd is a viable poison for neutralizing U/Gd solutions with the tested compositions.

  14. Material property correlations for uranium mononitride

    E-Print Network [OSTI]

    Hayes, Steven Lowe

    1989-01-01T23:59:59.000Z

    . 1 1770 - 2083 20. 7 - 34. 4 158, 1773 13-54 Test Environment Fuel Manafact- uring Route Test conducted in vaccuum (10~-5 ton) Cold pressed and sintered. Test conducted in 200 torr nitrogen atmosphere Isostatically Hot Pressed. Test... conductivity, high uranium density, stable irradiation behavior and compatibility with liquid metal coolants and refractory metal structural materials all combine to make uranium mononitride (UN) a very attractive nuclear fuel for use in high temperature...

  15. US-Russian collaboration in MPC & A enhancements at the Elektrostal Uranium Fuel-Fabrication Plant

    SciTech Connect (OSTI)

    Smith, H.; Murray, W.; Whiteson, R. [and others

    1997-11-01T23:59:59.000Z

    Enhancement of the nuclear materials protection, control, and accounting of (MPC&A) at the Elektrostal Machine-Building Plant (ELEMASH) has proceeded in two phases. Initially, Elektrostal served as the model facility at which to test US/Russian collaboration and to demonstrate MPC&A technologies available for safeguards enhancements at Russian facilities. This phase addressed material control and accounting (MC&A) in the low-enriched uranium (LEU) fuel-fabrication processes and the physical protection (PP) of part of the (higher-enrichment) breeder-fuel process. The second phase, identified later in the broader US/Russian agreement for expanded MPC&A cooperation. includes implementation of appropriate MC&A and PP systems in the breeder-fuel fabrication processes. Within the past year, an automated physical protection system has been installed and demonstrated in building 274, and an automated MC&A system has been designed and is being installed and will be tested in the LEU process. Attention has now turned to assuring longterm sustainability for the first phase and beginning MPC&A upgrades for the second phase. Sustainability measures establish the infrastructure for operation, maintenance, and repair of the installed systems-with US support for the lifetime of the US/Russian Agreement, but evolving toward full Russian operation of the system over the long term. For phase 2, which will address higher enrichments, projects have been identified to characterize the facilities, design MPC&A systems, procure appropriate equipment, and install and test final systems. One goal in phase 2 will be to build on initial work to create shared, plant-wide MPC&A assets for operation, maintenance, and evaluation of all safeguards systems.

  16. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium Production

  17. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium

  18. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium9 2014

  19. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium9

  20. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium911 2014

  1. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium911

  2. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium9117 2014

  3. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium9117 20145

  4. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomestic Uranium9117

  5. 2014 Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial andSeptember 25,9,1996 N Y MDomesticDomestic Uranium

  6. 2014 Domestic Uranium Production Report

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghurajiConventionalMississippi" ,"Plant","Primary1. TotalRevenueTotal97.10. Uranium

  7. 2014 Domestic Uranium Production Report

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghurajiConventionalMississippi" ,"Plant","Primary1. TotalRevenueTotal97.10. Uranium9.

  8. 2014 Uranium Marketing Annual Report

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghuraji Agro IndustriesTownDells,1 U.S. Department of Energygasoline4Residential17. Purchases of6a. Uranium

  9. 2014 Uranium Marketing Annual Report

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghuraji Agro IndustriesTownDells,1 U.S. Department of Energygasoline4Residential17. Purchases4. Uranium

  10. 2014 Uranium Marketing Annual Survey

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghuraji Agro IndustriesTownDells,1 U.S. Department of Energygasoline4Residential17. Purchases4. Uranium57.

  11. Method of preparation of uranium nitride

    DOE Patents [OSTI]

    Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

    2013-07-09T23:59:59.000Z

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

  12. LIBS Spectral Data for a Mixed Actinide Fuel Pellet Containing Uranium, Plutonium, Neptunium and Americium

    SciTech Connect (OSTI)

    Judge, Elizabeth J. [Los Alamos National Laboratory; Berg, John M. [Los Alamos National Laboratory; Le, Loan A. [Los Alamos National Laboratory; Lopez, Leon N. [Los Alamos National Laboratory; Barefield, James E. [Los Alamos National Laboratory

    2012-06-18T23:59:59.000Z

    Laser-induced breakdown spectroscopy (LIBS) was used to analyze a mixed actinide fuel pellet containing 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2}. The preliminary data shown here is the first report of LIBS analysis of a mixed actinide fuel pellet, to the authors knowledge. The LIBS spectral data was acquired in a plutonium facility at Los Alamos National Laboratory where the sample was contained within a glove box. The initial installation of the glove box was not intended for complete ultraviolet (UV), visible (VIS) and near infrared (NIR) transmission, therefore the LIBS spectrum is truncated in the UV and NIR regions due to the optical transmission of the window port and filters that were installed. The optical collection of the emission from the LIBS plasma will be optimized in the future. However, the preliminary LIBS data acquired is worth reporting due to the uniqueness of the sample and spectral data. The analysis of several actinides in the presence of each other is an important feature of this analysis since traditional methods must chemically separate uranium, plutonium, neptunium, and americium prior to analysis. Due to the historic nature of the sample fuel pellet analyzed, the provided sample composition of 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2} cannot be confirm without further analytical processing. Uranium, plutonium, and americium emission lines were abundant and easily assigned while neptunium was more difficult to identify. There may be several reasons for this observation, other than knowing the exact sample composition of the fuel pellet. First, the atomic emission wavelength resources for neptunium are limited and such techniques as hollow cathode discharge lamp have different dynamics than the plasma used in LIBS which results in different emission spectra. Secondly, due to the complex sample of four actinide elements, which all have very dense electronic energy levels, there may be reactions and interactions occurring within the plasma, such as collisional energy transfer, that might be a factor in the reduction in neptunium emission lines. Neptunium has to be analyzed alone using LIBS to further understand the dynamics that may be occurring in the plasma of the mixed actinide fuel pellet sample. The LIBS data suggests that the emission spectrum for the mixed actinide fuel pellet is not simply the sum of the emission spectra of the pure samples but is dependent on the species present in the plasma and the interactions and reactions that occur within the plasma. Finally, many of the neptunium lines are in the near infrared region which is drastically reduced in intensity by the current optical setup and possibly the sensitivity of the emission detector in the spectral region. Once the optics are replaced and the optical collection system is modified and optimized, the probability of observing emission lines for neptunium might be increased significantly. The mixed actinide fuel pellet was analyzed under the experimental conditions listed in Table 1. The LIBS spectra of the fuel pellet are shown in Figures 1-49. The spectra are labeled with the observed wavelength and atomic species (both neutral (I) and ionic (II)). Table 2 is a complete list of the observed and literature based emission wavelengths. The literature wavelengths have references including NIST Atomic Spectra Database (NIST), B.A. Palmer et al. 'An Atlas of Uranium Emission Intensities in a Hollow Cathode Discharge' taken at the Kitt Peak National Observatory (KPNO), R.L. Kurucz 1995 Atomic Line Data from the Smithsonian Astrophysical Observatory (SAO), J. Blaise et al. 'The Atomic Spectrum of Plutonium' from Argonne National Laboratory (BFG), and M. Fred and F.S. Tomkins, 'Preliminary Term Analysis of Am I and Am II Spectra' (FT). The dash (-) shown under Ionic State indicates that the ionic state of the transition was not available. In the spectra, the dash (-) is replaced with a question mark (?). Peaks that are not assigned are most likely real features and not noise but cannot be confidently assi

  13. Potential benefits and impacts on the CRWMS transportation system of filling spent fuel shipping casks with depleted uranium silicate glass

    SciTech Connect (OSTI)

    Pope, R.B.; Forsberg, C.W.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

    1996-06-01T23:59:59.000Z

    A new technology, the Depleted Uranium Silicate COntainer Fill System (DUSCOFS), is proposed to improve the performance and reduce the uncertainties of geological disposal of spent nuclear fuel (SNF), thus reducing both radionuclide release rates from the waste package and the potential for repository nuclear criticality events. DUSCOFS may also provide benefits for SNF storage and transport if it is loaded into the container early in the waste management cycle. Assessments have been made of the benefits to be derived by placing depleted uranium silicate (DUS) glass into SNF containers for enhancing repository performance assessment and controlling criticality over geologic times in the repository. Also, the performance, benefits, and impacts which can be derived if the SNF is loaded into a multi-purpose canister with DUS glass at a reactor site have been assessed. The DUSCOFS concept and the benefits to the waste management cycle of implementing DUSCOFS early in the cycle are discussed in this paper.

  14. Method for fabricating uranium foils and uranium alloy foils

    DOE Patents [OSTI]

    Hofman, Gerard L. (Downers Grove, IL); Meyer, Mitchell K. (Idaho Falls, ID); Knighton, Gaven C. (Moore, ID); Clark, Curtis R. (Idaho Falls, ID)

    2006-09-05T23:59:59.000Z

    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.

  15. Composition of the U.S. DOE Depleted Uranium Inventory

    E-Print Network [OSTI]

    Concentration Of Less

    about 2.75 wt% U-235. For further enrichment, the material was shipped to the Oak Ridge and Portsmouth plants. In addition to natural uranium, also uranium recycled from spent fuel was fed into the Paducah enrichment cascade (Table 2 and Fig. 2). The recycled uranium introduced various isotopes not found in natural uranium into the cascade: fission products, such as Technetium-99; transuranics, such as Neptunium-237 and Plutonium-239; and the artificial uranium isotope of Uranium-236. The spent fuel, from which uranium was recycled, originated from the Hanford and Savannah River military plutonium production reactors. This uranium was recycled, although its assay of U-235 was somewhat lower than in natural uranium (Table 2). This obviously must be seen in the context of the Cold War era, when uranium was a scarce resource. Due to the low burn-up of the military reactors, concentrations of artificial U-236 are comparatively low in this recycled uranium. The recycled uranium represents

  16. Recovery of uranium from seawater

    SciTech Connect (OSTI)

    Sugasaka, K. (Government Industrial Research Inst., Shikoku, Japan); Katoh, S.; Takai, N.; Takahashi, H.; Umezawa, Y.

    1981-01-01T23:59:59.000Z

    Seawater contains various elements in solution. Deuterium, lithium, and uranium are the important ingredients for energy application at present and in the future. This paper deals with the recovery of uranium from seawater, with emphasis on the development of an adsorbent with high selectivity and rate of adsorption for uranium. Polyacrylamidoxime chelating resins were synthesized from various co-polymers of acrylonitrile and cross-linking agents. The resulting resins with the chelating amidoxime group showed selective adsorption for uranium in seawater. The amount of uranium adsorbed from seawater at room temperature reached 3.2 mg/g resin after 180 days. Polyacrylamidoxime fiber, which was prepared from polyacrylonitrile fiber and hydroxylamine, showed a high rate of adsorption for uranium. The polyacrylamidoxime fiber conditioned with 1 M HC1 and 1 M NaOH adsorbed 4 mg U/g fiber from seawater in ten days. 9 figures, 6 tables.

  17. Depleted uranium disposal options evaluation

    SciTech Connect (OSTI)

    Hertzler, T.J.; Nishimoto, D.D.; Otis, M.D. [Science Applications International Corp., Idaho Falls, ID (United States). Waste Management Technology Div.

    1994-05-01T23:59:59.000Z

    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.

  18. Disposition of Surplus Highly Enriched Uranium

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

    four alternatives that would eliminate the weapons-usability of HEU by blending it with depleted uranium, natural uranium, or low-enriched uranium (LEU) to create LEU, either as...

  19. Uranium in prehistoric Indian pottery

    E-Print Network [OSTI]

    Filberth, Ernest William

    1976-01-01T23:59:59.000Z

    URANIUM IN PREHISTORIC INDIAN POTTERY A Thesis by ERNEST WILLIAM FILBERTH Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree of MASTER OF SCIENCE December 1976 Major Subject...: Chemistry URANIUM IN PREHISTORIC INDIAN POTTERY A Thesis by ERNEST WILLIAM FILBERTH Approved as to style and content by: (Chairman of Committee) (Head of Department) (Member) (Membe (Member) (Member) December 1976 ABSTRACT Uranium in Prehistoric...

  20. anthropogenic uranium enrichments: Topics by E-print Network

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

    Websites Summary: Flats Plutonium and Uranium Weapons-Grade Plutonium Enriched Uranium Depleted Uranium Plutonium-238 0.01 - 0.05% Uranium-234 0.1 - 1.02% Uranium-234...

  1. The concept of the use of recycled uranium for increasing the degree of security of export deliveries of fuel for light-water reactors

    SciTech Connect (OSTI)

    Alekseev, P. N.; Ivanov, E. A.; Nevinitsa, V. A.; Ponomarev-Stepnoi, N. N.; Rumyantsev, A. N.; Shmelev, V. M. [Russian Research Center Kurchatov Institute (Russian Federation); Borisevich, V. D.; Smirnov, A. Yu.; Sulaberidze, G. A. [National Nuclear Research University MEPhI (Russian Federation)

    2010-12-15T23:59:59.000Z

    The present paper deals with investigation of the possibilities for reducing the risk of proliferation of fissionable materials by means of increasing the degree of protection of fresh fuel intended for light-water reactors against unsanctioned use in the case of withdrawal of a recipient country of deliveries from IAEA safeguards. It is shown that the use of recycled uranium for manufacturing export nuclear fuel makes transfer of nuclear material removed from the fuel assemblies for weapons purposes difficult because of the presence of isotope {sup 232}U, whose content increases when one attempts to enrich uranium extracted from fresh fuel. In combination with restricted access to technologies for isotope separation by means of establishing international centers for uranium enrichment, this technical measure can significantly reduce the risk of proliferation associated with export deliveries of fuel made of low-enriched uranium. The assessment of a maximum level of contamination of nuclear material being transferred by isotope {sup 232}U for the given isotope composition of the initial fuel is obtained. The concept of further investigations of the degree of security of export deliveries of fuel assemblies with recycled uranium intended for light-water reactors is suggested.

  2. Disposition of Surplus Highly Enriched Uranium

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

    of Surplus Highly Enriched Uranium Environmental Impact Statement kternationd Atomic Energy Agency Idaho Nationrd Engineering Laborato low-enriched uranium low-level waste...

  3. Uranium Processing Facility Site Readiness Subproject Completed...

    National Nuclear Security Administration (NNSA)

    Field Offices Welcome to the NNSA Production Office NPO News Releases Uranium Processing Facility Site Readiness Subproject Completed ... Uranium Processing Facility Site...

  4. Unexpected, Stable Form of Uranium Detected | EMSL

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

    Unexpected, Stable Form of Uranium Detected Unexpected, Stable Form of Uranium Detected Insights on underappreciated reaction could shed light on environmental cleanup options...

  5. Uranium Weapons Components Successfully Dismantled | National...

    National Nuclear Security Administration (NNSA)

    Our Jobs Our Jobs Working at NNSA Blog Home About Us Our History NNSA Timeline Uranium Weapons Components Successfully Dismantled Uranium Weapons Components Successfully...

  6. Adsorptive Stripping Voltammetric Measurements of Trace Uranium...

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

    Adsorptive Stripping Voltammetric Measurements of Trace Uranium at the Bismuth Film Electrode. Adsorptive Stripping Voltammetric Measurements of Trace Uranium at the Bismuth Film...

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

    Office of Environmental Management (EM)

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

  8. 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)

    2001-01-01T23:59:59.000Z

    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.

  9. Elastic Properties of Rolled Uranium -- 10 wt.% Molybdenum Nuclear Fuel Foils

    SciTech Connect (OSTI)

    D. W. Brown; D. J. Alexander; K. D. Clarke; B. Clausen; M. A. Okuniewski; T. A. Sisneros

    2013-11-01T23:59:59.000Z

    In situ neutron diffraction data was collected during elastic loading of rolled foils of uranium-10 wt.% molybdenum bonded to a thin layer of zirconium. Lattice parameters were ascertained from the diffraction patterns to determine the elastic strain and, subsequently, the elastic moduli and Poisson’s ratio in the rolling and transverse directions. The foil was found to be elastically isotropic in the rolling plane with an effective modulus of 86 + / - 3 GPa and a Poisson’s ratio 0.39 + / - 0.04.

  10. Recycling Of Uranium- And Plutonium-Contaminated Metals From Decommissioning Of The Hanau Fuel Fabrication Plant

    SciTech Connect (OSTI)

    Kluth, T.; Quade, U.; Lederbrink, F. W.

    2003-02-26T23:59:59.000Z

    Decommissioning of a nuclear facility comprises not only actual dismantling but also, above all, management of the resulting residual materials and waste. Siemens Decommissioning Projects (DP) in Hanau has been involved in this task since 1995 when the decision was taken to decommission and dismantle the Hanau Fuel Fabrication Plant. Due to the decommissioning, large amounts of contaminated steel scrap have to be managed. The contamination of this metal scrap can be found almost exclusively in the form of surface contamination. Various decontamination technologies are involved, as there are blasting and wiping. Often these methods are not sufficient to meet the free release limits. In these cases, SIEMENS has decided to melt the scrap at Siempelkamp's melting plant. The plant is licensed according to the German Radiation Protection Ordinance Section 7 (issue of 20.07.2001). The furnace is a medium frequency induction type with a load capacity of 3.2 t and a throughput of 2 t/h for steel melting. For safety reasons, the furnace is widely operated by remote handling. A highly efficient filter system of cyclone, bag filter and HEPA-filter in two lines retains the dust and aerosol activity from the off-gas system. The slag is solidified at the surface of the melt and gripped before pouring the liquid iron into a chill. Since 1989, in total 15,000 t have been molten in the plant, 2,000 t of them having been contaminated steel scrap from the decommissioning of fuel fabrication plants. Decontamination factors could be achieved between 80 and 100 by the high affinity of the uranium to the slag former. The activity is transferred to the slag up to nearly 100 %. Samples taken from metal, slag and dust are analyzed by gamma measurements of the 186 keV line of U235 and the 1001 keV line of Pa234m for U238. All produced ingots showed a remaining activity less than 1 Bq/g and could be released for industrial reuse.

  11. US, Kazakhstan Cooperate to Eliminate Highly Enriched Uranium...

    National Nuclear Security Administration (NNSA)

    of 36 kilograms (approximately 80 pounds) of highly enriched uranium (HEU) spent fuel from the Institute of Nuclear Physics (INP) in Almaty, Kazakhstan. The HEU was...

  12. US-Russian collaboration for enhancing nuclear materials protection, control, and accounting at the Elektrostal uranium fuel-fabrication plant

    SciTech Connect (OSTI)

    Smith, H. [Los Alamos National Lab., NM (United States); Allentuck, J. [Brookhaven National Lab., Upton, NY (United States); Barham, M. [Oak Ridge National Lab., TN (United States); Bishop, M. [Sandia National Labs., Albuquerque, NM (United States); Wentz, D. [Lawrence Livermore National Lab., CA (United States); Steele, B.; Bricker, K. [Pacific Northwest National Lab., Richland, WA (United States); Cherry, R. [USDOE, Washington, DC (United States); Snegosky, T. [Dept. of Defense, Washington, DC (United States). Defense Nuclear Agency

    1996-09-01T23:59:59.000Z

    In September 1993, an implementing agreement was signed that authorized collaborative projects to enhance Russian national materials control and accounting, physical protection, and regulatory activities, with US assistance funded by the Nunn-Lugar Act. At the first US-Russian technical working group meeting in Moscow in February 1994, it was decided to identify a model facility where materials protection, control, and accounting (MPC and A) and regulatory projects could be carried out using proven technologies and approaches. The low-enriched uranium (LEU or RBMK and VVER) fuel-fabrication process at Elektrostal was selected, and collaborative work began in June 1994. Based on many factors, including initial successes at Elektrostal, the Russians expanded the cooperation by proposing five additional sites for MPC and A development: the Elektrostal medium-enriched uranium (MEU or BN) fuel-fabrication process and additional facilities at Podolsk, Dmitrovgrad, Obninsk, and Mayak. Since that time, multilaboratory teams have been formed to develop and implement MPC and A upgrades at the additional sites, and much new work is underway. This paper summarizes the current status of MPC and A enhancement projects in the LEU fuel-fabrication process and discusses the status of work that addresses similar enhancements in the MEU (BN) fuel processes at Elektrostal, under the recently expanded US-Russian MPC and A cooperation.

  13. Conversion and Evaluation of the University of Massachusetts Lowell Research Reactor From High-Enriched To Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Leo M. Bobek

    2003-11-19T23:59:59.000Z

    The process for converting the University of Massachusetts Lowell Research Reactor (UMLRR) from high-enrichment uranium (HEU) fuel to low-enrichment uranium (LEU) fuel began in 1988. Several years of design reviews, computational modeling, and thermal hydraulic analyses resulted in a preliminary reference core design and configuration based on 20 standard, MTR-type, flat-plate, 19.75% enriched, uranium silicide (u3Si2) fuel elements. A final safety analysis for the fuel conversion was submitted to the Nuclear Regulatory Commission (NRC) in 1993. The NRC made two additional requests for additional information and supplements were submitted in 1994 and 1997. The new UMLRR Reactor Supervisor initiated an effort to change the LEU reference core configuration to eliminate a complicated control rod modification needed for the smaller core.

  14. 2013 Domestic Uranium Production Report

    E-Print Network [OSTI]

    Administration (EIA), the statistical and analytical agency within the U.S. Department of Energy. By law, EIA.S. Energy Information Administration | 2013 Domestic Uranium Production Report iii Preface The U.S. Energy://www.eia.doe.gov/glossary/. #12;U.S. Energy Information Administration | 2013 Domestic Uranium Production Report iv Contents

  15. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    SciTech Connect (OSTI)

    SCHWINKENDORF, K.N.

    2006-05-12T23:59:59.000Z

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can supply useful information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions.

  16. Uranium isotopic composition and uranium concentration in special reference material SRM A (uranium in KCl/LiCl salt matrix)

    SciTech Connect (OSTI)

    Graczyk, D.G.; Essling, A.M.; Sabau, C.S.; Smith, F.P.; Bowers, D.L.; Ackerman, J.P.

    1997-07-01T23:59:59.000Z

    To help assure that analysis data of known quality will be produced in support of demonstration programs at the Fuel Conditioning Facility at Argonne National Laboratory-West (Idaho Falls, ID), a special reference material has been prepared and characterized. Designated SRM A, the material consists of individual units of LiCl/KCl eutectic salt containing a nominal concentration of 2.5 wt. % enriched uranium. Analyses were performed at Argonne National Laboratory-East (Argonne, IL) to determine the uniformity of the material and to establish reference values for the uranium concentration and uranium isotopic composition. Ten units from a batch of approximately 190 units were analyzed by the mass spectrometric isotope dilution technique to determine their uranium concentration. These measurements provided a mean value of 2.5058 {+-} 0.0052 wt. % U, where the uncertainty includes estimated limits to both random and systematic errors that might have affected the measurements. Evidence was found of a small, apparently random, non-uniformity in uranium content of the individual SRM A units, which exhibits a standard deviation of 0.078% of the mean uranium concentration. Isotopic analysis of the uranium from three units, by means of thermal ionization mass spectrometry with a special, internal-standard procedure, indicated that the uranium isotopy is uniform among the pellets with a composition corresponding to 0.1115 {+-} 0.0006 wt. % {sup 234}U, 19.8336 {+-} 0.0059 wt. % {sup 235}U, 0.1337 {+-} 0.0006 wt. % {sup 236}U, and 79.9171 {+-} 0.0057 wt. % {sup 238}U.

  17. SHEEP MOUNTAIN URANIUM PROJECT CROOKS GAP, WYOMING

    E-Print Network [OSTI]

    SHEEP MOUNTAIN URANIUM PROJECT CROOKS GAP, WYOMING US EPA Project Meeting April 7 2011April 7, 2011/Titan Uranium, VP Development · Deborah LebowAal/EPA Region 8 Air Program Introduction to Titan Uranium USA;PROJECT OVERVIEW ·Site Location·Site Location ·Fremont , Wyoming ·Existing Uranium Mine Permit 381C

  18. APPENDIX J Partition Coefficients For Uranium

    E-Print Network [OSTI]

    APPENDIX J Partition Coefficients For Uranium #12;Appendix J Partition Coefficients For Uranium J.1.0 Background The review of uranium Kd values obtained for a number of soils, crushed rock and their effects on uranium adsorption on soils are discussed below. The solution pH was also used as the basis

  19. Depleted uranium: A DOE management guide

    SciTech Connect (OSTI)

    NONE

    1995-10-01T23:59:59.000Z

    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.

  20. BIOREMEDIATION OF URANIUM CONTAMINATED SOILS AND WASTES.

    SciTech Connect (OSTI)

    FRANCIS,A.J.

    1998-09-17T23:59:59.000Z

    Contamination of soils, water, and sediments by radionuclides and toxic metals from uranium mill tailings, nuclear fuel manufacturing and nuclear weapons production is a major concern. Studies of the mechanisms of biotransformation of uranium and toxic metals under various microbial process conditions has resulted in the development of two treatment processes: (i) stabilization of uranium and toxic metals with reduction in waste volume and (ii) removal and recovery of uranium and toxic metals from wastes and contaminated soils. Stabilization of uranium and toxic metals in wastes is accomplished by exploiting the unique metabolic capabilities of the anaerobic bacterium, Clostridium sp. The radionuclides and toxic metals are solubilized by the bacteria directly by enzymatic reductive dissolution, or indirectly due to the production of organic acid metabolites. The radionuclides and toxic metals released into solution are immobilized by enzymatic reductive precipitation, biosorption and redistribution with stable mineral phases in the waste. Non-hazardous bulk components of the waste such as Ca, Fe, K, Mg and Na released into solution are removed, thus reducing the waste volume. In the second process uranium and toxic metals are removed from wastes or contaminated soils by extracting with the complexing agent citric acid. The citric-acid extract is subjected to biodegradation to recover the toxic metals, followed by photochemical degradation of the uranium citrate complex which is recalcitrant to biodegradation. The toxic metals and uranium are recovered in separate fractions for recycling or for disposal. The use of combined chemical and microbiological treatment process is more efficient than present methods and should result in considerable savings in clean-up and disposal costs.

  1. Opportunities to reduce consumption of natural uranium in reactor SVBR-75/100 when changing over to the closed fuel cycle

    SciTech Connect (OSTI)

    Toshinsky, G.I.; Komlev, O.G.; Mel'nikov, K.G.; Novikova, N.N. [FSUE SSC RF-IPPE, 1, Bondarenko sq., Obninsk, Kaluga rg., 249033 (Russian Federation)

    2007-07-01T23:59:59.000Z

    The design of reactor SVBR-75/100 allows it to operate using different types of fuel and in different fuel cycles without changing its design and deteriorating its safety characteristics. Fuel-at-once refueling adopted in the design (lack of partial refueling) makes it possible to change the core content at each refueling by using the type of fuel that is the most economically effective at the current stage of nuclear power (NP) development. In the nearest future use of mastered oxide uranium fuel and operating in the opened fuel cycle with postponed reprocessing will be the most economically effective. Changeover to the mixed uranium-plutonium fuel and closed nuclear fuel cycle (NFC) will be economically effective in an event of increase of natural uranium costs when the expenditures for construction of the enterprises on reprocessing the spent nuclear fuel (SNF), re-fabrication of new fuel with plutonium and their operating are less than the corresponding costs of natural uranium, its enrichment costs, the costs of manufacturing fresh uranium fuel and long temporary storage of SNF. At this, it is possible to use both MOX fuel with weapon or reactor plutonium and mixed nitride fuel in case its usage is more profitable. As fast reactors (FR) using uranium fuel and operating in the opened NFC consume much more natural uranium in comparison with thermal reactors (TR), and at the expected high paces of NP development the cheap resources of natural uranium will be exhausted prior to the middle of the century that will cause increase in the uranium cost, the period of FRs operating in the opened NFC must be maximally reduced. However, it should be mentioned that it is difficult to forecast reliably the date when because of the increased cost of natural uranium the NP will lose its competitiveness with electric power using fossil fuel. This is conditioned by the fact that the cost of the NPP produced electricity is less sensitive to the cost of natural uranium in contrast to the cost of electricity produced by thermal power plants using fossil fuel. At the same time, the available resources of natural uranium are increasing progressively with increase of its cost. The expenditure caused by changeover to the closed NFC will be less, if plutonium extracted from the own SNF of uranium loads is used in fabrication of the first MOX fuel loads. If the oxide uranium fuel is used, by the end of the lifetime a comparatively high breeding ratio (BR) ({approx}0.84) provides a sufficiently high content of plutonium in the SNF that may be used in the next fuel lifetimes when organizing the closed fuel cycle. Moreover, the own SNF of starting loads from oxide uranium fuel contains large quantity of unburned uranium-235 that is expedient to use for forming load for the next lifetime. From the very beginning of realization of the extended program on implementation of reactors SVBR-75/100 in the NP, use of plutonium extracted from the TRs' SNF for forming the starting loads of those reactors for the purpose of total elimination of natural uranium consumption will be more expensive as compared with the considered variant of changeover from the opened NFC to the closed NFC. This is conditioned by the fact that for the plutonium extracted from the TRs' SNF, the plutonium cost determined by a volume of SNF reprocessing per ton of plutonium will be several times higher as compared with its cost in case of using the own SNF because of considerably less content of plutonium in the TRs' SNF. It should be taken into account that the organization of the enterprise on large-scale reprocessing of TRs' SNF and MOX fuel fabrication must precede the construction of NPPs with FRs. Thus, the demands in investments are increased. At the same time, for the proposed changeover from the opened NFC to the closed one the construction of the closed NFC enterprise may be long postponed from FR launching that reduces the investment demands. At this, as the assessments have revealed, the investment fund for construction of such enterprise could be formed during abo ut t

  2. Safe Operating Procedure SAFETY PROTOCOL: URANIUM

    E-Print Network [OSTI]

    Farritor, Shane

    involve the use of natural or depleted uranium. Natural isotopes of uranium are U-238, U-235 and U-234 (see Table 1 for natural abundances). Depleted uranium contains less of the isotopes: U-235 and U-234. The specific activity of depleted uranium (5.0E-7 Ci/g) is less than that of natural uranium (7.1E-7 Ci

  3. DEPARTMENT OF ENERGY Excess Uranium Management: Effects of DOE...

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

    Excess Uranium Management: Effects of DOE Transfers of Excess Uranium on Domestic Uranium Mining, Conversion, and Enrichment Industries; Request for Information AGENCY: Office of...

  4. Uranium(VI) Diffusion in Low-Permeability Subsurface Materials...

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

    Uranium(VI) Diffusion in Low-Permeability Subsurface Materials. Uranium(VI) Diffusion in Low-Permeability Subsurface Materials. Abstract: Uranium(VI) diffusion was investigated in...

  5. Proteogenomic monitoring of Geobacter physiology during stimulated uranium bioremediation

    E-Print Network [OSTI]

    Wilkins, M.J.

    2010-01-01T23:59:59.000Z

    Phillips.  1992.  Bioremediation of  uranium contamination in situ uranium bioremediation.  Microbial Biotechnology 2:genes during in situ bioremediation of uranium?contaminated 

  6. adepleted uranium hexafluoride: Topics by E-print Network

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

    and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these unknown authors 15...

  7. active uranium americium: Topics by E-print Network

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

    and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these unknown authors 5...

  8. anthropogenic uranium concentration: Topics by E-print Network

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

    and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these unknown authors 12...

  9. abandoned uranium mill: Topics by E-print Network

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

    and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these unknown authors 3...

  10. anaconda uranium mill: Topics by E-print Network

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

    and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these unknown authors 3...

  11. THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE

    E-Print Network [OSTI]

    Olander, Donald R.

    2013-01-01T23:59:59.000Z

    1979) in "Uranium Enrichment", S. Villani, Ed. , Springer-E. (1973) "Uranium Enrichment by Gas Centrifuge" Mills andTHE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE

  12. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

    2006-02-01T23:59:59.000Z

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  13. The Uranium Institute 24th Annual Symposium

    E-Print Network [OSTI]

    Laughlin, Robert B.

    -239 for use in subsequent reactors. A fast neutron reactor is capable of producing more plutonium fuel than the uranium fuel it burns, leading to a breeder reactor. In addition, if the reactor is a fast with half lives of 30 years or less. The fast neutron reactor of preference was to be cooled with liquid

  14. The uranium cylinder assay system for enrichment plant safeguards

    SciTech Connect (OSTI)

    Miller, Karen A [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Marlow, Johnna B [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Rael, Carlos D [Los Alamos National Laboratory; Iwamoto, Tomonori [JNFL; Tamura, Takayuki [JNFL; Aiuchi, Syun [JNFL

    2010-01-01T23:59:59.000Z

    Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF{sub 6} cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF{sub 6} cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.

  15. Fabrication and Characterization of Uranium-Molybdenum-Zirconium Alloys

    E-Print Network [OSTI]

    Woolum, Connor

    2014-12-12T23:59:59.000Z

    As part of a global effort to convert reactors that require highly enriched uranium to instead operate with low enriched uranium, monolithic fuel plates consisting of a U-Mo fuel meat with a zirconium foil barrier layer and clad in aluminum...

  16. Spectroscopic Evidence for Uranium Bearing Precipitates in Vadose Zone

    E-Print Network [OSTI]

    Northwest Laboratory, Richland, Washington 99352 Uranium (U) solid-state speciation in vadose zone sediments of past nuclear fuel fabrication processes, uranium (U) has been recognized as one of the most widespreadHanfordsitesthatreceivedU-containingwastesduring its mission of Pu production between 1940 and 1990. Unirradiated fuel rod wastes were disposed

  17. Development of a low enrichment uranium core for the MIT reactor

    E-Print Network [OSTI]

    Newton, Thomas Henderson

    2006-01-01T23:59:59.000Z

    An investigation has been made into converting the MIT research reactor from using high enrichment uranium (HEU) to low enrichment uranium (LEU) with a newly developed fuel material. The LEU fuel introduces negative ...

  18. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, Bruce A. (Kennewick, WA)

    1986-01-01T23:59:59.000Z

    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.

  19. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, B.A.

    1983-06-10T23:59:59.000Z

    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.

  20. Beneficial Uses of Depleted Uranium

    SciTech Connect (OSTI)

    Brown, C. [U.S. Department of Energy, Germantown, MD (United States); Croff, A.G.; Haire, M. J. [Oak Ridge National Lab., TN (United States)

    1997-08-01T23:59:59.000Z

    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.

  1. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    SciTech Connect (OSTI)

    TOFFER, H.

    2006-07-18T23:59:59.000Z

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel and four (4) spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data, such as the uncertainty in fuel exposure impact on reactivity and the pulse neutron data evaluation methodology, failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements supply useful information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions.

  2. Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2007-01-01T23:59:59.000Z

    Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration

  3. Process for alloying uranium and niobium

    SciTech Connect (OSTI)

    Holcombe, C.E.; Northcutt, W.G.; Masters, D.R.; Chapman, L.R.

    1990-12-31T23:59:59.000Z

    Alloys such as U-6Nb are prepared by forming a stacked sandwich array of uranium 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.

  4. Process for alloying uranium and niobium

    SciTech Connect (OSTI)

    Holcombe, C.E.; Northcutt, W.G.; Masters, D.R.; Chapman, L.R.

    1991-04-09T23:59:59.000Z

    This patent describes alloys such as U-6Nb prepared by forming a stacked sandwich array of uranium 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.

  5. Demonstration of jackhammer incorporating depleted uranium

    SciTech Connect (OSTI)

    Fischer, L E; Hoard, R W; Carter, D L; Saculla, M D; Wilson, G V

    2000-04-01T23:59:59.000Z

    The United States Government currently has an abundance of depleted uranium (DU). This surplus of about 1 billion pounds is the result of an enrichment process using gaseous diffusion to produce enriched and depleted uranium. The enriched uranium has been used primarily for either nuclear weapons for the military or nuclear fuel for the commercial power industry. Most of the depleted uranium remains at the enrichment process plants in the form of depleted uranium hexafluoride (DUF{sub 6}). The Department of Energy (DOE) recently began a study to identify possible commercial applications for the surplus material. One of these potential applications is to use the DU in high-density strikers/hammers in pneumatically driven tools, such as jack hammers and piledrivers to improve their impulse performance. The use of DU could potentially increase tunneling velocity and excavation into target materials with improved efficiency. This report describes the efforts undertaken to analyze the particulars of using DU in two specific striking applications: the jackhammer and chipper tool.

  6. Depleted uranium hexafluoride: Waste or resource?

    SciTech Connect (OSTI)

    Schwertz, N.; Zoller, J.; Rosen, R.; Patton, S. [Lawrence Livermore National Lab., CA (United States); Bradley, C. [USDOE Office of Nuclear Energy, Science, Technology, Washington, DC (United States); Murray, A. [SAIC (United States)

    1995-07-01T23:59:59.000Z

    the US Department of Energy is evaluating technologies for the storage, disposal, or re-use of depleted uranium hexafluoride (UF{sub 6}). This paper discusses the following options, and provides a technology assessment for each one: (1) conversion to UO{sub 2} for use as mixed oxide duel, (2) conversion to UO{sub 2} to make DUCRETE for a multi-purpose storage container, (3) conversion to depleted uranium metal for use as shielding, (4) conversion to uranium carbide for use as high-temperature gas-cooled reactor (HTGR) fuel. In addition, conversion to U{sub 3}O{sub 8} as an option for long-term storage is discussed.

  7. Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed

    SciTech Connect (OSTI)

    Smirnov, A. Yu., E-mail: a.y.smirnoff@rambler.ru; Sulaberidze, G. A. [National Research Nuclear University MEPhI (Russian Federation); Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A., E-mail: neva@dhtp.kiae.ru; Proselkov, V. N.; Chibinyaev, A. V. [Russian Research Centre Kurchatov Institute (Russian Federation)

    2012-12-15T23:59:59.000Z

    A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

  8. U. S. forms uranium enrichment corporation

    SciTech Connect (OSTI)

    Seltzer, R.

    1993-07-12T23:59:59.000Z

    After almost 40 years of operation, the federal government is withdrawing from the uranium enrichment business. On July 1, the Department of Energy turned over to a new government-owned entity--the US Enrichment Corp. (USEC)--both the DOE enrichment plants at Paducah, Ky., and Portsmouth, Ohio, and domestic and international marketing of enriched uranium from them. Pushed by the inability of DOE's enrichment operations to meet foreign competition, Congress established USEC under the National Energy Policy Act of 1992, envisioning the new corporation as the first step to full privatization. With gross revenues of $1.5 billion in fiscal 1992, USEC would rank 275th on the Fortune 500 list of top US companies. USEC will lease from DOE the Paducah and Portsmouth facilities, built in the early 1950s, which use the gaseous diffusion process for uranium enrichment. USEC's stock is held by the US Treasury, to which it will pay annual dividends. Martin Marietta Energy Systems, which has operated Paducah since 1984 and Portsmouth since 1986 for DOE, will continue to operate both plants for USEC. Closing one of the two facilities will be studied, especially in light of a 40% world surplus of capacity over demand. USEC also will consider other nuclear-fuel-related ventures. USEC will produce only low-enriched uranium, not weapons-grade material. Indeed, USEC will implement a contract now being completed under which the US will purchase weapons-grade uranium from dismantled Russian nuclear weapons and convert it into low-enriched uranium for power reactor fuel.

  9. Standard practice for removal of uranium or plutonium, or both, for impurity assay in uranium or plutonium materials

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2006-01-01T23:59:59.000Z

    Standard practice for removal of uranium or plutonium, or both, for impurity assay in uranium or plutonium materials

  10. Depleted uranium disposal options.

    SciTech Connect (OSTI)

    Biwer, B. M.; Ranek, N. L.; Goldberg, M.; Avci, H. I.

    2000-04-01T23:59:59.000Z

    Depleted uranium hexafluoride (UF{sub 6}) has been produced in the United States since the 1940s as part of both the military program and the civilian nuclear energy program. The U.S. Department of Energy (DOE) is the agency responsible for managing most of the depleted UF{sub 6} that has been produced in the United States. The total quantity of depleted UF{sub 6} that DOE has to or will have to manage is approximately 700,000 Mg. Studies have been conducted to evaluate the various alternatives for managing this material. This paper evaluates and summarizes the alternative of disposal as low-level waste (LLW). Results of the analysis indicate that UF{sub 6} needs to be converted to a more stable form, such as U{sub 3}O{sub 8}, before disposal as LLW. Estimates of the environmental impacts of disposal in a dry environment are within the currently applicable standards and regulations. Of the currently operating LLW disposal facilities, available information indicates that either of two DOE facilities--the Hanford Site or the Nevada Test Site--or a commercial facility--Envirocare of Utah--would be able to dispose of up to the entire DOE inventory of depleted UF{sub 6}.

  11. Uranium 2007 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2008-01-01T23:59:59.000Z

    Based on official information received from 40 countries, Uranium 2007 provides a comprehensive review of world uranium supply and demand as of 1st January 2007, as well as data on global uranium exploration, resources, production and reactor-related requirements. It provides substantive new information from major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Projections of nuclear generating capacity and reactor-related uranium requirements through 2030 are also featured, along with an analysis of long-term uranium supply and demand issues. It finds that with rising demand and declining inventories, uranium prices have increased dramatically in recent years. As a result, the uranium industry is undergoing a significant revival, bringing to an end a period of over 20 years of underinvestment.

  12. A uranium-titanium-niobium alloy

    SciTech Connect (OSTI)

    Ludtka, G.M.; Ludtka, G.M.

    1990-02-23T23:59:59.000Z

    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.

  13. Inherently safe in situ uranium recovery

    DOE Patents [OSTI]

    Krumhansl, James L; Brady, Patrick V

    2014-04-29T23:59:59.000Z

    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.

  14. Uranium Acquisition | Y-12 National Security Complex

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

    of Interest (EOI) to acquire up to 6,800 metric tons of Uranium (MTU) of high purity depleted uranium metal (DU) and related material and services. This request for EOI does...

  15. EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany

    Broader source: Energy.gov [DOE]

    This environmental assessment (EA) will evaluate the potential environmental impacts of a DOE proposal to accept used nuclear fuel from the Federal Republic of Germany at DOE’s Savannah River Site (SRS) for processing and disposition. This used nuclear fuel is composed of kernels containing thorium and U.S.-origin highly enriched uranium (HEU) embedded in small graphite spheres that were irradiated in nuclear reactors used for research and development purposes.

  16. Conversion and Blending Facility Highly enriched uranium to low enriched uranium as uranium hexafluoride. Revision 1

    SciTech Connect (OSTI)

    NONE

    1995-07-05T23:59:59.000Z

    This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF{sub 6} and a (2) blend the pure HEU UF{sub 6} with diluent UF{sub 6} to produce LWR grade LEU-UF{sub 6}. The primary emphasis of this blending 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 chemical and isotopic concentrations of the blended LEU product will be held within the specifications required for LWR fuel. The blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry.

  17. Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center

    SciTech Connect (OSTI)

    Myers, Astasia [Stanford University, Stanford, CA 94305, USA and MonAme Scientific Research Center, Ulaanbaatar (Mongolia)

    2011-06-28T23:59:59.000Z

    The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

  18. D Riso-R-429 Automated Uranium

    E-Print Network [OSTI]

    -induced delayed-neutron coun- ting is applied preferably in large geochemical exploration pro- grammes. UraniumCM i D Riso-R-429 Automated Uranium Analysis by Delayed-Neutron Counting H. Kunzendorf, L. Lřvborg AUTOMATED URANIUM ANALYSIS BY DELAYED-NEUTRON COUNTING H. Kunzendorf, L. Lřvborg and E.M. Christiansen

  19. Remediation and Recovery of Uranium from Contaminated

    E-Print Network [OSTI]

    Lovley, Derek

    that Geobacter species can effectively remove uranium from contaminated groundwater by reducing soluble U emplaced in flow- through columns of uranium-contaminated sediments readily removed U(VI) from the groundwater, and 87% of the uranium that had been removed was recovered from the electrode surface after

  20. 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)

    1991-01-01T23:59:59.000Z

    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. 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)

    1990-01-01T23:59:59.000Z

    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.

  2. Uranium Watch REGULATORY CONFUSION: FEDERALAND STATE

    E-Print Network [OSTI]

    Uranium Watch Report REGULATORY CONFUSION: FEDERALAND STATE ENFORCEMENT OF 40 C.F.R. PART 61 SUBPART W INTRODUCTION 1. This Uranium Watch Report, Regulatory Confusion: Federal and State Enforcement at the White Mesa Uranium Mill, San Juan County, Utah. 2. The DAQ, a Division of the Utah Department

  3. Clean Air Act Requirements: Uranium Mill Tailings

    E-Print Network [OSTI]

    EPA'S Clean Air Act Requirements: Uranium Mill Tailings Radon Emissions Rulemaking Reid J. Rosnick requirements for operating uranium mill tailings (Subpart W) Status update on Subpart W activities Outreach/Communications #12;3 EPA Regulatory Requirements for Operating Uranium Mill Tailings (Clean Air Act) · 40 CFR 61

  4. URANIUM MILL TAILINGS RADON FLUX CALCULATIONS

    E-Print Network [OSTI]

    URANIUM MILL TAILINGS RADON FLUX CALCULATIONS PIĂ?ON RIDGE PROJECT MONTROSE COUNTY, COLORADO Inc. (Golder) was commissioned by EFRC to evaluate the operations of the uranium mill tailings storage in this report were conducted using the WISE Uranium Mill Tailings Radon Flux Calculator, as updated on November

  5. Methodology and a preliminary data base for examining the health risks of electricity generation from uranium and coal fuels

    SciTech Connect (OSTI)

    El-Bassioni, A.A.

    1980-08-01T23:59:59.000Z

    An analytical model was developed to assess and examine the health effects associated with the production of electricity from uranium and coal fuels. The model is based on a systematic methodology that is both simple and easy to check, and provides details about the various components of health risk. A preliminary set of data that is needed to calculate the health risks was gathered, normalized to the model facilities, and presented in a concise manner. Additional data will become available as a result of other evaluations of both fuel cycles, and they should be included in the data base. An iterative approach involving only a few steps is recommended for validating the model. After each validation step, the model is improved in the areas where new information or increased interest justifies such upgrading. Sensitivity analysis is proposed as the best method of using the model to its full potential. Detailed quantification of the risks associated with the two fuel cycles is not presented in this report. The evaluation of risks from producing electricity by these two methods can be completed only after several steps that address difficult social and technical questions. Preliminary quantitative assessment showed that several factors not considered in detail in previous studies are potentially important. 255 refs., 21 figs., 179 tabs.

  6. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01T23:59:59.000Z

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  7. Material property correlations for uranium mononitride 

    E-Print Network [OSTI]

    Hayes, Steven Lowe

    1989-01-01T23:59:59.000Z

    who have provided technical support for this project throughout its duration. I also express my sincere appreciation and thanks to the U. S. Department of Energy and Oak Ridge Associated Universities whose Nuclear Engineering and Health Physics... space nuclear reactors. Uranium mononitride is currently the reference fuel for the SP-100 space reactor system and will likely be considered for application in future multimegawatt space power systems as well. Although fuel modeling efforts have...

  8. Uranium mill tailings and radon

    SciTech Connect (OSTI)

    Hanchey, L A

    1981-01-01T23:59:59.000Z

    The major health hazard from uranium mill tailings is presumed to be respiratory cancer resulting from the inhalation of radon daughter products. A review of studies on inhalation of radon and its daughters indicates that the hazard from the tailings is extremely small. If the assumptions used in the studies are correct, one or two people per year in the US may develop cancer as a result of radon exhaled from all the Uranium Mill Tailings Remedial Action Program sites. The remedial action should reduce the hazard from the tailings by a factor of about 100.

  9. Physical and mechanical metallurgy of uranium and uranium alloys

    SciTech Connect (OSTI)

    Eckelmeyer, K.H. [Sandia National Labs. (United States)

    1998-12-31T23:59:59.000Z

    Engineering disadvantages of unalloyed uranium include relatively low strength, low ductility, and poor oxidation and corrosion resistance. As-cast uranium typically exhibits very large grains that cause nonuniform deformation and low tensile ductility. Uranium is often alloyed to improve its corrosion resistance and mechanical properties. Titanium is most commonly used to increase strength; niobium and molybdenum, to increase oxidation and corrosion resistance; and vanadium, to refine alpha grain size in castings. Under equilibrium conditions these elements are extensively soluble in the high-temperature gamma phase, slightly soluble in the intermediate temperature beta phase, and essentially insoluble in the low-temperature alpha phase. Uranium alloys are vacuum solution heat treated in the gamma range to dissolve the alloying elements and remove hydrogen. The subsequent microstructures and properties are determined by the cooling rate from the solution treatment temperature. Oxidation and corrosion resistance increases with increasing the amount of alloy in solid solution. As a result, alloys such as U-6%Nb and U-10%Mo are often used in applications requiring good corrosion resistance.

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

    SciTech Connect (OSTI)

    McCabe, Rodney J. [Los Alamos National Laboratory; Kelly, Ann Marie [Los Alamos National Laboratory; Clarke, Amy J. [Los Alamos National Laboratory; Field, Robert D. [Los Alamos National Laboratory; Wenk, H. R. [University of California, Berkeley

    2012-07-25T23:59:59.000Z

    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.

  11. Removal of uranium from aqueous HF solutions

    DOE Patents [OSTI]

    Pulley, Howard (West Paducah, KY); Seltzer, Steven F. (Paducah, KY)

    1980-01-01T23:59:59.000Z

    This invention is a simple and effective method for removing uranium from aqueous HF solutions containing trace quantities of the same. The method comprises contacting the solution with particulate calcium fluoride to form uranium-bearing particulates, permitting the particulates to settle, and separting the solution from the settled particulates. The CaF.sub.2 is selected to have a nitrogen surface area in a selected range and is employed in an amount providing a calcium fluoride/uranium weight ratio in a selected range. As applied to dilute HF solutions containing 120 ppm uranium, the method removes at least 92% of the uranium, without introducing contaminants to the product solution.

  12. 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)

    1991-01-01T23:59:59.000Z

    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.

  13. 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

    2012-07-31T23:59:59.000Z

    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.

  14. Uranium 2014 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2014-01-01T23:59:59.000Z

    Published every other year, Uranium Resources, Production, and Demand, or the "Red Book" as it is commonly known, is jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It is the recognised world reference on uranium and is based on official information received from 43 countries. It presents the results of a thorough review of world uranium supplies and demand and provides a statistical profile of the world uranium industry in the areas of exploration, resource estimates, production and reactor-related requirements. It provides substantial new information from all major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Long-term projections of nuclear generating capacity and reactor-related uranium requirements are provided as well as a discussion of long-term uranium supply and demand issues. This edition focuses on recent price and production increases that could signal major changes in the industry.

  15. Uranium 2011 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2012-01-01T23:59:59.000Z

    In the wake of the Fukushima Daiichi nuclear power plant accident, questions are being raised about the future of the uranium market, including as regards the number of reactors expected to be built in the coming years, the amount of uranium required to meet forward demand, the adequacy of identified uranium resources to meet that demand and the ability of the sector to meet reactor requirements in a challenging investment climate. This 24th edition of the “Red Book”, a recognised world reference on uranium jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, provides analyses and information from 42 producing and consuming countries in order to address these and other questions. It offers a comprehensive review of world uranium supply and demand as well as data on global uranium exploration, resources, production and reactor-related requirements. It also provides substantive new information on established uranium production centres around the world and in countri...

  16. Uranium 2005 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2006-01-01T23:59:59.000Z

    Published every other year, Uranium Resources, Production, and Demand, or the "Red Book" as it is commonly known, is jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It is the recognised world reference on uranium and is based on official information received from 43 countries. This 21st edition presents the results of a thorough review of world uranium supplies and demand as of 1st January 2005 and provides a statistical profile of the world uranium industry in the areas of exploration, resource estimates, production and reactor-related requirements. It provides substantial new information from all major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Projections of nuclear generating capacity and reactor-related uranium requirements through 2025 are provided as well as a discussion of long-term uranium supply and demand issues. This edition focuses on recent price and production increases that could signal major c...

  17. Development of Novel Sorbents for Uranium Extraction from Seawater

    SciTech Connect (OSTI)

    Lin, Wenbin; Taylor-Pashow, Kathryn

    2014-01-08T23:59:59.000Z

    As the uranium resource in terrestrial ores is limited, it is difficult to ensure a long-term sustainable nuclear energy technology. The oceans contain approximately 4.5 billion tons of uranium, which is one thousand times the amount of uranium in terrestrial ores. Development of technologies to recover the uranium from seawater would greatly improve the uranium resource availability, sustaining the fuel supply for nuclear energy. Several methods have been previously evaluated including solvent extraction, ion exchange, flotation, biomass collection, and adsorption; however, none have been found to be suitable for reasons such as cost effectiveness, long term stability, and selectivity. Recent research has focused on the amidoxime functional group as a promising candidate for uranium sorption. Polymer beads and fibers have been functionalized with amidoxime functional groups, and uranium adsorption capacities as high as 1.5 g U/kg adsorbent have recently been reported with these types of materials. As uranium concentration in seawater is only ~3 ppb, great improvements to uranium collection systems must be made in order to make uranium extraction from seawater economically feasible. This proposed research intends to develop transformative technologies for economic uranium extraction from seawater. The Lin group will design advanced porous supports by taking advantage of recent breakthroughs in nanoscience and nanotechnology and incorporate high densities of well-designed chelators into such nanoporous supports to allow selective and efficient binding of uranyl ions from seawater. Several classes of nanoporous materials, including mesoporous silica nanoparticles (MSNs), mesoporous carbon nanoparticles (MCNs), meta-organic frameworks (MOFs), and covalent-organic frameworks (COFs), will be synthesized. Selective uranium-binding liagnds such as amidoxime will be incorporated into the nanoporous materials to afford a new generation of sorbent materials that will be evaluated for their uranium extraction efficiency. The initial testing of these materials for uranium binding will be carried out in the Lin group, but more detailed sorption studies will be carried out by Dr. Taylor-Pashow of Savannah River National Laboratory in order to obtain quantitative uranyl sorption selectivity and kinetics data for the proposed materials. The proposed nanostructured sorbent materials are expected to have higher binding capacities, enhanced extraction kinetics, optimal stripping efficiency for uranyl ions, and enhanced mechanical and chemical stabilities. This transformative research will significantly impact uranium extraction from seawater as well as benefit DOE’s efforts on environmental remediation by developing new materials and providing knowledge for enriching and sequestering ultralow concentrations of other metals.

  18. armoring uranium-mill tailings: Topics by E-print Network

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

    December 15, 1989). Current Standards 18 URANIUM MILLING ACTIVITIES AT SEQUOYAH FUELS CORPORATION CiteSeer Summary: Sequoyah Fuels Corporation (SFC) describes previous...

  19. Uranium Mining, Conversion, and Enrichment Industries

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group current C3EDepartment of Energy OfficeFact Sheet Uranium Mill Tailingsi

  20. Uranium Marketing Annual Report - Energy Information Administration

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for On-Highway4,1,50022,3,,,,6,1,9,1,50022,3,,,,6,1,Decade Year-0E (2001)gasoline prices4 Oil demand expected toall Uranium

  1. Uranium Leasing Program Documents | Department of Energy

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual Siteof Energy 2, 2015 -Helicopter-Japan Joint Nuclear D.C. *ofUranium EnrichmentDocuments

  2. Summary - Uranium233 Downblending and Disposition Project

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium Transferon the Passing of AdmiraltheOil and LessOak Ridge,SRSTank

  3. Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties

    E-Print Network [OSTI]

    Chiang, Keng-Yen

    2012-01-01T23:59:59.000Z

    The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor ...

  4. Mixed uranium dicarbide and uranium dioxide microspheres and process of making same

    DOE Patents [OSTI]

    Stinton, David P. (Knoxville, TN)

    1983-01-01T23:59:59.000Z

    Nuclear fuel microspheres are made by sintering microspheres containing uranium dioxide and uncombined carbon in a 1 mole percent carbon monoxide/99 mole percent argon atmosphere at 1550.degree. C. and then sintering the microspheres in a 3 mole percent carbon monoxide/97 mole percent argon atmosphere at the same temperature.

  5. Examination of the conversion of the U.S. submarine fleet from highly enriched uranium to low enriched uranium

    E-Print Network [OSTI]

    McCord, Cameron (Cameron Liam)

    2014-01-01T23:59:59.000Z

    The nuclear reactors used by the U.S. Navy for submarine propulsion are currently fueled by highly enriched uranium (HEU), but HEU brings administrative and political challenges. This issue has been studied by the Navy ...

  6. Uncertainty clouds uranium enrichment corporation's plans

    SciTech Connect (OSTI)

    Lane, E.

    1993-03-24T23:59:59.000Z

    An expected windfall to the US Treasury from the sale of the Energy Dept.'s commercial fuel enrichment facilities may evaporate in the next few weeks when the Clinton administration submits its fiscal 1994 budget proposal to Congress, according to congressional and administration officials. Under the Energy Policy Act of 1992, DOE is required to lease two uranium enrichment facilities, Portsmouth, Ohio, and Paducah, KY., to the government-owned US Enrichment Corp. (USEC) by July 1. Estimates by OMB and Treasury indicate a potential yearly payoff of $300 million from the government-owned company's sale of fuel for commercial reactors. Those two facilities use a process of gaseous diffusion to enrich uranium to about 3 percent for use as fuel in commercial power plants. DOE has contracts through at least 1996 to provide about 12 million separative work units (SWUs) yearly to US utilities and others world-wide. But under an agreement signed between the US and Russia last August, at least 10 metric tons, or 1.5 million SWUs, of low-enriched uranium (LEU) blended down from Russia warheads is expected to be delivered to the US starting in 1994. It could be sold at $50 to $60 per SWU, far below what DOE currently charges for its SWUs - $135 per SWU for 70 percent of the contract price and $90 per SWU for the remaining 30 percent.

  7. Reports on investigations of uranium anomalies. National Uranium Resource Evaluation

    SciTech Connect (OSTI)

    Goodknight, C.S.; Burger, J.A. (comps.) [comps.

    1982-10-01T23:59:59.000Z

    During the National Uranium Resource Evaluation (NURE) program, conducted for the US Department of Energy (DOE) by Bendix Field Engineering Corporation (BFEC), radiometric and geochemical surveys and geologic investigations detected anomalies indicative of possible uranium enrichment. Data from the Aerial Radiometric and Magnetic Survey (ARMS) and the Hydrogeochemical and Stream-Sediment Reconnaissance (HSSR), both of which were conducted on a national scale, yielded numerous anomalies that may signal areas favorable for the occurrence of uranium deposits. Results from geologic evaluations of individual 1/sup 0/ x 2/sup 0/ quadrangles for the NURE program also yielded anomalies, which could not be adequately checked during scheduled field work. Included in this volume are individual reports of field investigations for the following six areas which were shown on the basis of ARMS, HSSR, and (or) geologic data to be anomalous: (1) Hylas zone and northern Richmond basin, Virginia; (2) Sischu Creek area, Alaska; (3) Goodman-Dunbar area, Wisconsin; (4) McCaslin syncline, Wisconsin; (5) Mt. Withington Cauldron, Socorro County, New Mexico; (6) Lake Tecopa, Inyo County, California. Field checks were conducted in each case to verify an indicated anomalous condition and to determine the nature of materials causing the anomaly. The ultimate objective of work is to determine whether favorable conditions exist for the occurrence of uranium deposits in areas that either had not been previously evaluated or were evaluated before data from recent surveys were available. Most field checks were of short duration (2 to 5 days). The work was done by various investigators using different procedures, which accounts for variations in format in their reports. All papers have been abstracted and indexed.

  8. On the possibility of using uranium-beryllium oxide fuel in a VVER reactor

    SciTech Connect (OSTI)

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D. [National Research Center Kurchatov Institute (Russian Federation); Stogov, Yu. V., E-mail: YVStogov@mephi.ru [National Research Nuclear University MEPhI (Russian Federation)

    2014-12-15T23:59:59.000Z

    The possibility of using UO{sub 2}-BeO fuel in a VVER reactor is considered with allowance for the thermophysical properties of this fuel. Neutron characteristics of VVER fuel assemblies with UO{sub 2}-BeO fuel pellets are estimated.

  9. Occupational exposures to uranium: processes, hazards, and regulations

    SciTech Connect (OSTI)

    Stoetzel, G.A.; Fisher, D.R.; McCormack, W.D.; Hoenes, G.R.; Marks, S.; Moore, R.H.; Quilici, D.G.; Breitenstein, B.D.

    1981-04-01T23:59:59.000Z

    The United States Uranium Registry (USUR) was formed in 1978 to investigate potential hazards from occupational exposure to uranium and to assess the need for special health-related studies of uranium workers. This report provides a summary of Registry work done to date. The history of the uranium industry is outlined first, and the current commercial uranium industry (mining, milling, conversion, enrichment, and fuel fabrication) is described. This description includes information on basic processes and areas of greatest potential radiological exposure. In addition, inactive commercial facilities and other uranium operations are discussed. Regulation of the commercial production industry for uranium fuel is reported, including the historic development of regulations and the current regulatory agencies and procedures for each phase of the industry. A review of radiological health practices in the industry - facility monitoring, exposure control, exposure evaluation, and record-keeping - is presented. A discussion of the nonradiological hazards of the industry is provided, and the final section describes the tissue program developed as part of the Registry.

  10. High temperature behavior of metallic inclusions in uranium dioxide

    SciTech Connect (OSTI)

    Yang, R.L.

    1980-08-01T23:59:59.000Z

    The object of this thesis was to construct a temperature gradient furnace to simulate the thermal conditions in the reactor fuel and to study the migration of metallic inclusions in uranium oxide under the influence of temperature gradient. No thermal migration of molybdenum and tungsten inclusions was observed under the experimental conditions. Ruthenium inclusions, however, dissolved and diffused atomically through grain boundaries in slightly reduced uranium oxide. An intermetallic compound (probably URu/sub 3/) was formed by reaction of Ru and UO/sub 2-x/. The diffusivity and solubility of ruthenium in uranium oxide were measured.

  11. Chelating polymers for recovery of uranium from seawater

    SciTech Connect (OSTI)

    Kabay, N. (Ege Univ., Izmir (Turkey)); Egawa, Hiroaki (Kumamoto Univ. (Japan))

    1994-01-01T23:59:59.000Z

    Despite the low concentration of uranium in seawater (3.3 ppb), a special emphasis has been placed on its recovery. Although the concentration is low, it has been estimated that the world's oceans contain about 4 x 10[sup 9] tons of uranium - theoretically an unlimited supply of nuclear fuel. Adsorption has been considered to be a technically feasible procedure for a uranium recovery process with regard to economic and environmental impacts. The present paper restricts its coverage to those applications using chelating polymeric resins containing amidoxime groups as the most promising adsorbent. 72 refs., 8 figs., 1 tab.

  12. Power Surge: Uranium alloy fuel for TerraPower | Y-12 National Security

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible forPortsmouth/Paducah Project OfficePower Electronics Power Electronics

  13. Bacterial Community Succession During in situ Uranium Bioremediation: Spatial Similarities Along Controlled Flow Paths

    E-Print Network [OSTI]

    Hwang, Chiachi

    2009-01-01T23:59:59.000Z

    problem, and the use of depleted uranium and other heavyenvironmental hazard. Depleted uranium is weakly radioactive

  14. L'URANIUM ET LES ARMES L'URANIUM APPAUVRI. Pierre Roussel*

    E-Print Network [OSTI]

    Boyer, Edmond

    L'URANIUM ET LES ARMES � L'URANIUM APPAUVRI. Pierre Roussel* Institut de Physique Nucléaire, CNRS massivement dans la guerre du Golfe, des obus anti- chars ont été utilisés, avec des "charges d'uranium, avec une charge de 300 g d'uranium et tiré par des avions, l'autre de 120 mm de diamètre avec une

  15. Uranium in prehistoric Indian pottery 

    E-Print Network [OSTI]

    Filberth, Ernest William

    1976-01-01T23:59:59.000Z

    . 2 to 25 ppm (Katz 1951). From thermal equilibrium calculations on the earth's core, mantle, and crust, and through actual analysis of samples, uranium was found to be concentrated in the earth's crust. According to modern geological thought..., as the uniformly molten earth cooled, its matter became separated into one vapor phase and three concentric condensed phases: the siderosphere (the earth's core, probably primarily molten iron), the chalcosphere forming the intermediate shell (the mantle...

  16. Dry process fluorination of uranium dioxide using ammonium bifluoride

    E-Print Network [OSTI]

    Yeamans, Charles Burnett, 1978-

    2003-01-01T23:59:59.000Z

    An experimental study was conducted to determine the practicality of various unit operations for fluorination of uranium dioxide. The objective was to prepare ammonium uranium fluoride double salts from uranium dioxide and ...

  17. albarrana uranium ores: Topics by E-print Network

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

    and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these unknown authors 7 A...

  18. The fuel cycle economics of improved uranium utilization in light water reactors

    E-Print Network [OSTI]

    Abbaspour, Ali Tehrani

    A simple fuel cycle cost model has been formulated, tested satisfactorily (within better than 3% for a wide range of cases)

  19. Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)IntegratedSpeedingTechnicalPurchase, Delivery, and Storage ofQuadrennial

  20. Review of uranium bioassay techniques

    SciTech Connect (OSTI)

    Bogard, J.S.

    1996-04-01T23:59:59.000Z

    A variety of analytical techniques is available for evaluating uranium in excreta and tissues at levels appropriate for occupational exposure control and evaluation. A few (fluorometry, kinetic phosphorescence analysis, {alpha}-particle spectrometry, neutron irradiation techniques, and inductively-coupled plasma mass spectrometry) have also been demonstrated as capable of determining uranium in these materials at levels comparable to those which occur naturally. Sample preparation requirements and isotopic sensitivities vary widely among these techniques and should be considered carefully when choosing a method. This report discusses analytical techniques used for evaluating uranium in biological matrices (primarily urine) and limits of detection reported in the literature. No cost comparison is attempted, although references are cited which address cost. Techniques discussed include: {alpha}-particle spectrometry; liquid scintillation spectrometry, fluorometry, phosphorometry, neutron activation analysis, fission-track counting, UV-visible absorption spectrophotometry, resonance ionization mass spectrometry, and inductively-coupled plasma mass spectrometry. A summary table of reported limits of detection and of the more important experimental conditions associated with these reported limits is also provided.

  1. Uranium Tris-aryloxide Derivatives Supported by Triazacyclononane: Engendering a Reactive Uranium(III)

    E-Print Network [OSTI]

    Meyer, Karsten

    , we are currently investigating the coordina- tion chemistry of uranium metal centers with classicalUranium Tris-aryloxide Derivatives Supported by Triazacyclononane: Engendering a Reactive Uranium, and Karsten Meyer* Contribution from the Department of Chemistry and Biochemistry, UniVersity of California

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2008-01-01T23:59:59.000Z

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

  3. Energy balance for uranium recovery from seawater

    SciTech Connect (OSTI)

    Schneider, E.; Lindner, H. [The University of Texas, 1 University Station C2200, Austin, TX 78712 (United States)

    2013-07-01T23:59:59.000Z

    The energy return on investment (EROI) of an energy resource is the ratio of the energy it ultimately produces to the energy used to recover it. EROI is a key viability measure for a new recovery technology, particularly in its early stages of development when financial cost assessment would be premature or highly uncertain. This paper estimates the EROI of uranium recovery from seawater via a braid adsorbent technology. In this paper, the energy cost of obtaining uranium from seawater is assessed by breaking the production chain into three processes: adsorbent production, adsorbent deployment and mooring, and uranium elution and purification. Both direct and embodied energy inputs are considered. Direct energy is the energy used by the processes themselves, while embodied energy is used to fabricate their material, equipment or chemical inputs. If the uranium is used in a once-through fuel cycle, the braid adsorbent technology EROI ranges from 12 to 27, depending on still-uncertain performance and system design parameters. It is highly sensitive to the adsorbent capacity in grams of U captured per kg of adsorbent as well as to potential economies in chemical use. This compares to an EROI of ca. 300 for contemporary terrestrial mining. It is important to note that these figures only consider the mineral extraction step in the fuel cycle. At a reference performance level of 2.76 g U recovered per kg adsorbent immersed, the largest energy consumers are the chemicals used in adsorbent production (63%), anchor chain mooring system fabrication and operations (17%), and unit processes in the adsorbent production step (12%). (authors)

  4. Atomic Diffusion in the Uranium-50wt% Zirconium Nuclear Fuel System

    E-Print Network [OSTI]

    Eichel, Daniel

    2013-06-17T23:59:59.000Z

    Atomic diffusion phenomena were examined in a metal-alloy nuclear fuel system composed of ?-phase U-50wt%Zr fuel in contact with either Zr-10wt%Gd or Zr-10wt%Er. Each alloy was fabricated from elemental feed material via melt-casting, and diffusion...

  5. Colorimetric detection of uranium in water

    DOE Patents [OSTI]

    DeVol, Timothy A. (Clemson, SC); Hixon, Amy E. (Piedmont, SC); DiPrete, David P. (Evans, GA)

    2012-03-13T23:59:59.000Z

    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.

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

    Office of Environmental Management (EM)

    Leasing Program, under which DOE administers tracts of land in western Colorado for exploration, development, and the extraction of uranium and vanadium ores. ULP PEIS...

  7. Statistical data of the uranium industry

    SciTech Connect (OSTI)

    none,

    1982-01-01T23:59:59.000Z

    Statistical Data of the Uranium Industry is a compendium of information relating to US uranium reserves and potential resources and to exploration, mining, milling, and other activities of the uranium industry through 1981. The statistics are based primarily on data provided voluntarily by the uranium exploration, mining, and milling companies. The compendium has been published annually since 1968 and reflects the basic programs of the Grand Junction Area Office (GJAO) of the US Department of Energy. The production, reserves, and drilling information is reported in a manner which avoids disclosure of proprietary information.

  8. Review The Toxicity of Depleted Uranium

    E-Print Network [OSTI]

    Wayne Briner

    Abstract: Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a clear and defined set of symptoms. Chronic low-dose, or subacute, exposure to depleted uranium alters the appearance of milestones in developing organisms. Adult animals that were exposed to depleted uranium during development display persistent alterations in behavior, even after cessation of depleted uranium exposure. Adult animals exposed to depleted uranium demonstrate altered behaviors and a variety of alterations to brain chemistry. Despite its reduced level of radioactivity evidence continues to accumulate that depleted uranium, if ingested, may pose a radiologic hazard. The current state of knowledge concerning DU is discussed.

  9. Distribution of uranium-bearing phases in soils from Fernald

    SciTech Connect (OSTI)

    Buck, E.C.; Brown, N.R.; Dietz, N.L.

    1993-12-31T23:59:59.000Z

    Electron beam techniques have been used to characterize uranium-contaminated soils and the Fernald Site, Ohio. Uranium particulates have been deposited on the soil through chemical spills and from the operation of an incinerator plant on the site. The major uranium phases have been identified by electron microscopy as uraninite, autunite, and uranium phosphite [U(PO{sub 3}){sub 4}]. Some of the uranium has undergone weathering resulting in the redistribution of uranium within the soil.

  10. High strength and density tungsten-uranium alloys

    DOE Patents [OSTI]

    Sheinberg, Haskell (Los Alamos, NM)

    1993-01-01T23:59:59.000Z

    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.

  11. Fundamental study on recovery uranium oxide from HEPA filters

    SciTech Connect (OSTI)

    Izumida, T. [Hitachi Ltd., Ibaraki (Japan). Hitachi Works; Matsumoto, H.; Tsuchiya, H.; Iba, H. [Hitachi Nuclear Engineering Co., Ltd., Ibaraki (Japan); Noguchi, Y. [Radioactive Waste Management Center, Tokyo (Japan)

    1993-12-31T23:59:59.000Z

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

  12. Beryllium Impregnation of Uranium Fuel: Thermal Modeling of Cylindrical Objects for Efficiency Evaluation

    E-Print Network [OSTI]

    Lynn, Nicholas

    2011-08-04T23:59:59.000Z

    With active research projects related to nuclear waste immobilization and high conductivity nuclear fuels, a thermal model has been developed to simulate the temperature profile within a heat generating cylinder in order to imitate the behavior...

  13. Beryllium Impregnation of Uranium Fuel: Thermal Modeling of Cylindrical Objects for Efficiency Evaluation 

    E-Print Network [OSTI]

    Lynn, Nicholas

    2011-08-04T23:59:59.000Z

    With active research projects related to nuclear waste immobilization and high conductivity nuclear fuels, a thermal model has been developed to simulate the temperature profile within a heat generating cylinder in order to imitate the behavior...

  14. Silicon carbide performance as cladding for advanced uranium and thorium fuels for light water reactors

    E-Print Network [OSTI]

    Sukjai, Yanin

    2014-01-01T23:59:59.000Z

    There has been an ongoing interest in replacing the fuel cladding zirconium-based alloys by other materials to reduce if not eliminate the autocatalytic and exothermic chemical reaction with water and steam at above 1,200 ...

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

    Office of Environmental Management (EM)

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

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

    Office of Environmental Management (EM)

    LM visited 84 defense-related legacy uranium mine sites located within 11 uranium mining districts in 6 western states. At these sites, photographs and global positioning...

  17. Uncertainty analysis of multi-rate kinetics of uranium desorption...

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

    Uncertainty analysis of multi-rate kinetics of uranium desorption from sediments. Uncertainty analysis of multi-rate kinetics of uranium desorption from sediments. Abstract: A...

  18. 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...

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

    Energy Savers [EERE]

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

  20. Geochemical Controls on Contaminant Uranium in Vadose Hanford...

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

    Controls on Contaminant Uranium in Vadose Hanford Formation Sediments at the 200 Area and 300 Area, Hanford Site, Geochemical Controls on Contaminant Uranium in Vadose Hanford...

  1. Microbial Reduction of Uranium under Iron- and Sulfate-reducing...

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

    Uranium under Iron- and Sulfate-reducing Conditions: Effect of Amended Goethite on Microbial Community Microbial Reduction of Uranium under Iron- and Sulfate-reducing Conditions:...

  2. Microscopic Reactive Diffusion of Uranium in the Contaminated...

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

    Reactive Diffusion of Uranium in the Contaminated Sediments at Hanford, United States. Microscopic Reactive Diffusion of Uranium in the Contaminated Sediments at Hanford, United...

  3. Y-12 uranium storage facility?a Ťdream come true?

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

    ranks and actually provides the first impedance for the just finished highly enriched uranium storage facility. Recently the Highly Enriched Uranium Material Facility was...

  4. Composition, stability, and measurement of reduced uranium phases...

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

    Composition, stability, and measurement of reduced uranium phases for groundwater bioremediation at Old Rifle, CO. Composition, stability, and measurement of reduced uranium phases...

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

    Energy Savers [EERE]

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

  6. 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...

  7. alloyed uranium transformation: Topics by E-print Network

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

    were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding... Helmreich, Grant...

  8. EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth...

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

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

  9. acute uranium intoxication: Topics by E-print Network

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

    consists of replacing the water with 20 Garland Jr., Theodore 8 Review The Toxicity of Depleted Uranium CiteSeer Summary: Abstract: Depleted uranium (DU) is an emerging...

  10. alloyed uranium sicral: Topics by E-print Network

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

    were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding... Helmreich, Grant...

  11. Depleted Uranium Hexafluoride (DUF6) Fully Operational at the...

    Office of Environmental Management (EM)

    Depleted Uranium Hexafluoride (DUF6) Fully Operational at the Portsmouth and Paducah Gaseous Diffusion Sites Depleted Uranium Hexafluoride (DUF6) Fully Operational at the...

  12. Uranium Nitride as LWR TRISO Fuel: Thermodynamic Modeling of U-C-N

    SciTech Connect (OSTI)

    Besmann, Theodore M [ORNL; Shin, Dongwon [ORNL

    2012-01-01T23:59:59.000Z

    TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will need to be UN. In support of the fuel development effort, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide and it will be in equilibrium with carbon within the TRISO particle. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Selected measurements were used to fit a first order model of the UC1-xNx phase, represented by the inter-solution of UN and UC. Fit to the data was significantly improved by also adjusting the heat of formation for UN by ~12 kJ/mol and the phase equilbria was best reproduced by also adjusting the heat for U2N3 by +XXX. The determined interaction parameters yielded a slightly positive deviation from ideality, which agrees with lattice parameter measurements which show positive deviation from Vegard s law. The resultant model together with reported values for other phases in the system were used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.

  13. Thermo-chemical Modelling of Uranium-free Nitride Fuels Mikael JOLKKONEN1;;y

    E-Print Network [OSTI]

    Haviland, David

    Stockholm, Sweden 2 Paul-Scherrer-Institut, 5232 Villigen PSI, Switzerland (Received October 10, 2003 that the practical difficulties with yield and purity are of a kinetic rather than a thermodynamical nature. We sintering under inert gas. Addition of nitrogen in small amounts to fuel pin filling gas also appears

  14. recycled_uranium.cdr

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling7 AugustAFRICAN3uj:'I,\ W:'. ' .r:lCM ' ~.

  15. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars per611. U.S.

  16. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars per611. U.S.6.

  17. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars per611.

  18. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars per611.8. U.S.

  19. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars per611.8.

  20. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars per611.8.4.

  1. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars per611.8.4.2.

  2. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars per611.8.4.2.3.

  3. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars

  4. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127by Local(Dollars9. Summary

  5. Uranium Marketing Annual Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N) COFeet)Updates to

  6. Uranium Marketing Annual Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N) COFeet)Updates tob.

  7. Uranium Marketing Annual Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N) COFeet)Updates tob.S2.

  8. Uranium Marketing Annual Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N) COFeet)Updates

  9. Uranium Marketing Annual Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N) COFeet)Updatesb.

  10. Uranium Marketing Annual Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N) COFeet)Updatesb.4.

  11. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

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  12. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

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  13. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)

  14. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of

  15. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of6.

  16. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of6.7.

  17. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of6.7.8.

  18. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of6.7.8.9.

  19. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of6.7.8.9..

  20. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries of6.7.8.9..0.

  1. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries

  2. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries2. Inventories

  3. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries2.

  4. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries2.5.

  5. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

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  6. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries2.5.3.a.

  7. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries2.5.3.a.b.

  8. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries2.5.3.a.b.7.

  9. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3. Deliveries2.5.3.a.b.7.8.

  10. Uranium Marketing Annual Report -

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3.

  11. alkaline-earth metal uranium: Topics by E-print Network

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

    predominantly... Sames, William 2011-08-08 59 Characterization of Thermal Properties of Depleted Uranium Metal Microspheres Texas A&M University - TxSpace Summary: Nuclear fuel...

  12. Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion

    E-Print Network [OSTI]

    Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

    2008-01-01T23:59:59.000Z

    The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the ...

  13. Estimate of radiation release from MIT reactor with low enriched uranium (LEU) core during maximum hypothetical accident

    E-Print Network [OSTI]

    Plumer, Kevin E. (Kevin Edward)

    2011-01-01T23:59:59.000Z

    In accordance with a 1986 NRC ruling, the MIT Research Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. A component of the conversion analysis ...

  14. Uranium at Y-12: Accountability | Y-12 National Security Complex

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron SpinPrinceton PlasmaAfternoon4. Uranium purchased by... Uranium at Y-12:

  15. Uranium at Y-12: Casting | Y-12 National Security Complex

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron SpinPrinceton PlasmaAfternoon4. Uranium purchased by... Uranium at

  16. Uranium at Y-12: Recovery | Y-12 National Security Complex

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron SpinPrinceton PlasmaAfternoon4. Uranium purchased by...Recovery Uranium

  17. Influence of Temperature on the Corrosion of Uranium Dioxide Nuclear Fuel

    SciTech Connect (OSTI)

    Broczkowski, Michael E.; Noel, Jamie J.; Shoesmith, David W. [Chemistry, The University of Western Ontario, 1151 Richmond St., London, N6A 5B7 (Canada)

    2007-07-01T23:59:59.000Z

    The anodic dissolution of UO{sub 2} has been studied at 60 deg. C and the results compared to previous observations at 22 deg. C. The rate of oxidation / dissolution was determined electrochemically at constant potentials in the range -500 mV to 500 mV (vs. SCE). The composition of the electrochemically oxidized surface was determined by X-Ray Photoelectron Spectroscopy (XPS). The onset of oxidation (UO{sub 2} {yields} UO{sub 2+x}) occurred at approximately the same potential (-400 mV) at both temperatures. However, the conversion of U{sub V} to U{sub VI}, and hence to soluble UO{sub 2}{sup 2+}, was accelerated by temperature. This acceleration of dissolution caused the development of acidity at localized sites on the fuel surface at lower (less oxidizing) potentials ({>=} 100 mV) at 60 deg. C than at 22 deg. C ({>=} 350 mV)

  18. Control of structure and reactivity by ligand design : applications to small molecule activation by low-valent uranium complexes

    E-Print Network [OSTI]

    Lam, Oanh Phi

    2010-01-01T23:59:59.000Z

    Coordination Chemistry of Uranium………………………………….11 1.4researchers from uranium chemistry. Fortunately, despiteclassical coordination chemistry of uranium has flourished

  19. An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

    SciTech Connect (OSTI)

    Bathke, Charles Gary [Los Alamos National Laboratory; Wallace, Richard K [Los Alamos National Laboratory; Hase, Kevin R [Los Alamos National Laboratory; Sleaford, Brad W [LLNL; Ebbinghaus, Bartley B [LLNL; Collins, Brian W [PNNL; Bradley, Keith S [LLNL; Prichard, Andrew W [PNNL; Smith, Brian W [PNNL

    2010-01-01T23:59:59.000Z

    This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled until consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.

  20. Uranium Management - Preservation of a National Asset

    SciTech Connect (OSTI)

    Jackson, J. D.; Stroud, J. C.

    2002-02-27T23:59:59.000Z

    The Uranium Management Group (UMG) was established at the Department of Energy's (DOE's) Oak Ridge Operations in 1999 as a mechanism to expedite the de-inventory of surplus uranium from the Fernald Environmental Management Project site. This successful initial venture has broadened into providing uranium material de-inventory and consolidation support to the Hanford site as well as retrieving uranium materials that the Department had previously provided to universities under the loan/lease program. As of December 31, 2001, {approx} 4,300 metric tons of uranium (MTU) have been consolidated into a more cost effective interim storage location at the Portsmouth site near Piketon, OH. The UMG continues to uphold its corporate support mission by promoting the Nuclear Materials Stewardship Initiative (NMSI) and the twenty-five (25) action items of the Integrated Nuclear Materials Management Plan (1). Before additional consolidation efforts may commence to remove excess inventory from Environmental Management closure sites and universities, a Programmatic Environmental Assessment (PEA) must be completed. Two (2) noteworthy efforts currently being pursued involve the investigation of re-use opportunities for surplus uranium materials and the recovery of usable uranium from the shutdown Portsmouth cascade. In summary, the UMG is available as a DOE complex-wide technical resource to promote the responsible management of surplus uranium.

  1. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect (OSTI)

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06T23:59:59.000Z

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are similar to those obtained with the traditional UO2 nuclear fuel.

  2. Recovery of uranium by immobilized polyhydroxyanthraquinone

    SciTech Connect (OSTI)

    Sakaguchi, T.; Nakajima, A.

    1986-01-01T23:59:59.000Z

    Nine species of polyhydroxyanthraquinone and two of polyhydroxynaphthoquinone were screened to determine which have the greatest ability to accumulate uranium. 1,2-Dihydroxyanthraquinone and 3-amino-1,2-dihydroxyanthraquinone have extremely high accumulation abilities. To improve the adsorbing characteristics of these compounds, the authors tried to immobilize these compounds by coupling with diazotized aminopolystyrene. The immobilized 1,2-dihydroxyanthraquinone has the most favorable features for uranium recovery; high selective adsorption ability to uranium, rapid adsorption rate, and applicability in both column and batch systems. This adsorbent can recover uranium almost quantitatively from natural seawater. Almost all uranium adsorbed is desorbed with a solution of 1 N HCl. Thus, immobilized 1,2-dihydroxyanthraquinone can be used repeatedly in the adsorption-desorption process.

  3. Scrap uranium recycling via electron beam melting

    SciTech Connect (OSTI)

    McKoon, R.

    1993-11-01T23:59:59.000Z

    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.

  4. Method for fabricating laminated uranium composites

    DOE Patents [OSTI]

    Chapman, L.R.

    1983-08-03T23:59:59.000Z

    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.

  5. Thermodynamic data for uranium fluorides

    SciTech Connect (OSTI)

    Leitnaker, J.M.

    1983-03-01T23:59:59.000Z

    Self-consistent thermodynamic data have been tabulated for uranium fluorides between UF/sub 4/ and UF/sub 6/, including UF/sub 4/ (solid and gas), U/sub 4/F/sub 17/ (solid), U/sub 2/F/sub 9/ (solid), UF/sub 5/ (solid and gas), U/sub 2/F/sub 10/ (gas), and UF/sub 6/ (solid, liquid, and gas). Included are thermal function - the heat capacity, enthalpy, and free energy function, heats of formation, and vaporization behavior.

  6. Methods for Investigating Gas Bubble Formation in Uranium-Zirconium Alloys

    E-Print Network [OSTI]

    Mews, Kathryn Ann Wright

    2013-05-06T23:59:59.000Z

    cycle. Liquid sodium fast spectrum reactors are an essential link in closing the fuel cycle with their ability to burn transuranics and depleted uranium, their transmutation possibilities, and their breeder applications. Development of metal fuels... uranium and plutonium also require certain permitting and handling precautions. In order to facilitate the experimental work in the facilities available at Texas A&M, the use of these materials was discounted. Instead, unirradiated, depleted uranium...

  7. Removal of uranium from uranium-contaminated soils -- Phase 1: Bench-scale testing. Uranium in Soils Integrated Demonstration

    SciTech Connect (OSTI)

    Francis, C. W.

    1993-09-01T23:59:59.000Z

    To address the management of uranium-contaminated soils at Fernald and other DOE sites, the DOE Office of Technology Development formed the Uranium in Soils Integrated Demonstration (USID) program. The USID has five major tasks. These include the development and demonstration of technologies that are able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from the soil, (3) treat the soil and dispose of any waste, (4) establish performance assessments, and (5) meet necessary state and federal regulations. This report deals with soil decontamination or removal of uranium from contaminated soils. The report was compiled by the USID task group that addresses soil decontamination; includes data from projects under the management of four DOE facilities [Argonne National Laboratory (ANL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), and the Savannah River Plant (SRP)]; and consists of four separate reports written by staff at these facilities. The fundamental goal of the soil decontamination task group has been the selective extraction/leaching or removal of uranium from soil faster, cheaper, and safer than current conventional technologies. The objective is to selectively remove uranium from soil without seriously degrading the soil`s physicochemical characteristics or generating waste forms that are difficult to manage and/or dispose of. Emphasis in research was placed more strongly on chemical extraction techniques than physical extraction techniques.

  8. Recovery of uranium by using new microorganisms isolated from North American uranium deposits

    SciTech Connect (OSTI)

    Sakaguchi, T.; Nakajima, A.; Tsuruta, T. [Miyazaki Medical College (Japan)

    1995-12-31T23:59:59.000Z

    Some attempts were made to remove uranium that may be present in refining effluents, mine tailings by using new microorganisms isolated from uranium deposits and peculiar natural environments. To screen microorganisms isolated from uranium deposits and peculiar natural environments in North America and Japan for maximal accumulation of uranium, hundreds of microorganisms were examined. Some microorganisms can accumulate about 500 mg (4.2 mEq) of uranium per gram of Microbial cells within 1 h. The uranium accumulation capacity of the cells exceeds that of commercially available chelating agents (2-3 mEq/g adsorbent). We attempted to recover uranium from uranium refining waste water by using new microorganisms. As a result, these microbial cells can recover trace amounts of uranium from uranium waste water with high efficiency. These strains also have a high accumulating ability for thorium. Thus, these new microorganisms can be used as an adsorbing agent for the removal of nuclear elements may be present in metallurgical effluents, mine tailings and other waste sources.

  9. A Geostatistical Study of the Uranium Deposit at Kvanefjeld,

    E-Print Network [OSTI]

    are identified by the discriminating effect of the individual variable. INIS descriptors; URANIUM ORES? RESERVES

  10. Uranium Cluster Chemistry DOI: 10.1002/anie.200906605

    E-Print Network [OSTI]

    Uranium Cluster Chemistry DOI: 10.1002/anie.200906605 Tetranuclear Uranium Clusters by Reductive in the coordination chemistry and small-molecule reactivity of uranium. Among the intriguing reactivity patterns of tetravalent uranium with 3,5-dimethylpyrazolate (Me2PzŔ ) led to forma- tion of an unprecedented homoleptic

  11. Diffusion model of the non-stoichiometric uranium dioxide

    SciTech Connect (OSTI)

    Moore, Emily, E-mail: emily.moore@cea.fr [CEA Saclay, DEN-DPC-SCCME, 91191 Gif-sur-Yvette Cedex (France); Guéneau, Christine, E-mail: christine.gueneau@cea.fr [CEA Saclay, DEN-DPC-SCCME, 91191 Gif-sur-Yvette Cedex (France); Crocombette, Jean-Paul, E-mail: jean-paul.crocombette@cea.fr [CEA Saclay, DEN DEN, Service de Recherches de Métallurgie Physique, 91191 Gif-sur-Yvette Cedex (France)

    2013-07-15T23:59:59.000Z

    Uranium dioxide (UO{sub 2}), which is used in light water reactors, exhibits a large range of non-stoichiometry over a wide temperature scale up to 2000 K. Understanding diffusion behavior of uranium oxides under such conditions is essential to ensure safe reactor operation. The current understanding of diffusion properties is largely limited by the stoichiometric deviations inherent to the fuel. The present DICTRA-based model considers diffusion across non-stoichiometric ranges described by experimentally available data. A vacancy and interstitial model of diffusion is applied to the U–O system as a function of its defect structure derived from CALPHAD-type thermodynamic descriptions. Oxygen and uranium self and tracer diffusion coefficients are assessed for the construction of a mobility database. Chemical diffusion coefficients of oxygen are derived with respect to the Darken relation and migration energies of defects are evaluated as a function of stoichiometric deviation. - Graphical abstract: Complete description of Oxygen–Uranium diffusion as a function of composition at various temperatures according to the developed Dictra model. - Highlights: • Assessment of a uranium–oxygen diffusion model with Dictra. • Complete description of U–O diffusion over wide temperature and composition range. • Oxygen model includes terms for interstitial and vacancy migration. • Interaction terms between defects help describe non-stoichiometric domain of UO{sub 2±x}. • Uranium model is separated into mobility terms for the cationic species.

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

    DOE Patents [OSTI]

    Willit, James L. (Batavia, IL)

    2010-09-21T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Willit, James L. (Ratavia, IL)

    2007-09-11T23:59:59.000Z

    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.

  14. Investigation of Uranium Polymorphs

    SciTech Connect (OSTI)

    Sweet, Lucas E.; Henager, Charles H.; Hu, Shenyang Y.; Johnson, Timothy J.; Meier, David E.; Peper, Shane M.; Schwantes, Jon M.

    2011-08-01T23:59:59.000Z

    The UO3-water system is complex and has not been fully characterized, even though these species are common throughout the nuclear fuel cycle. As an example, most production schemes for UO3 result in a mixture of up to six or more different polymorphic phases, and small differences in these conditions will affect phase genesis that ultimately result in measureable changes to the end product. As a result, this feature of the UO3-water system may be useful as a means for determining process history. This research effort attempts to better characterize the UO3-water system with a variety of optical techniques for the purpose of developing some predictive capability for estimating process history in polymorphic phases of unknown origin. Three commercially relevant preparation methods for the production of UO3 were explored. Previously unreported low temperature routes to ?- and ?-UO3 were discovered. Raman and fluorescence spectroscopic libraries were established for pure and mixed polymorphic forms of UO3 in addition to the common hydrolysis products of UO3. An advantage of the sensitivity of optical fluorescence microscopy over XRD has been demonstrated. Preliminary aging studies of the ? and ? forms of UO3 have been conducted. In addition, development of a 3-D phase field model used to predict phase genesis of the system was initiated. Thermodynamic and structural constants that will feed the model have been gathered from the literature for most of the UO3 polymorphic phases.

  15. Technical Basis for Assessing Uranium Bioremediation Performance

    SciTech Connect (OSTI)

    PE Long; SB Yabusaki; PD Meyer; CJ Murray; AL N’Guessan

    2008-04-01T23:59:59.000Z

    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.

  16. Electrochemistry, Spectroscopy, and Reactivity of Uranium Complexes Supported by Ferrocene Diamide Ligands

    E-Print Network [OSTI]

    Duhovic, Selma

    2012-01-01T23:59:59.000Z

    J. L. , Pentavalent Uranium Chemistry-Synthetic Pursuit of afor Trivalent Uranium Chemistry. Inorg. Chem. 1989, 28, (and High-Valent Uranium Chemistry. Organometallics 2011,

  17. Recent International R&D Activities in the Extraction of Uranium from Seawater

    E-Print Network [OSTI]

    Rao, Linfeng

    2011-01-01T23:59:59.000Z

    Uranium and Rare Earth Elements Using Biomass of Algae, Bioinorganic ChemistryRecovery of uranium from sea water. Chemistry & Industry (uranium recovery from seawater. Industrial & Engineering Chemistry

  18. Decolonizing cartographies : sovereignty, territoriality, and maps of meaning in the uranium landscape

    E-Print Network [OSTI]

    Voyles, Traci Brynne

    2010-01-01T23:59:59.000Z

    continued mining and uranium exploration on and near theand thereby open to uranium exploration, claims-staking, andbe used for uranium mining or exploration. One Hispano

  19. Recent International R&D Activities in the Extraction of Uranium from Seawater

    E-Print Network [OSTI]

    Rao, Linfeng

    2011-01-01T23:59:59.000Z

    for extracting uranium from seawater. Brit. (1978), 3 pp.Ger. ). Recovery of uranium from seawater. Ger. Offen. (Ger. ). Recovery of uranium from seawater. Ger. Offen. (

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

    E-Print Network [OSTI]

    Tian, Guoxin

    2013-01-01T23:59:59.000Z

    Sequestering uranium from seawater: binding strength andin sequestering uranium from seawater, forms strongExtraction of uranium from seawater is very challenging, not

  1. Electrodic voltages accompanying stimulated bioremediation of a uranium-contaminated aquifer

    E-Print Network [OSTI]

    Williams, K.H.

    2010-01-01T23:59:59.000Z

    2008), Sustained Removal of Uranium From ContaminatedR. T. Anderson (2007), Uranium removal from groundwater viasulfide and the removal of uranium from groundwater. The

  2. Sulfur isotopes as indicators of amended bacterial sulfate reduction processes influencing field scale uranium bioremediation

    E-Print Network [OSTI]

    Druhan, J.L.

    2009-01-01T23:59:59.000Z

    sulfate reduction and uranium removal. The samples for thisanism of Sulfate and Uranium Removal. In M-23, low acetatethe highest rates of uranium removal were observed at redox

  3. Field-based detection and monitoring of uranium in contaminated groundwater using two immunosensors

    E-Print Network [OSTI]

    Melton, S.J.

    2010-01-01T23:59:59.000Z

    an in situ uranium bioremediation field site. Appl. Environ.undergoing uranium bioremediation. Int. J. Systematicstimulated uranium bioremediation. Appl. Environ. Microbiol.

  4. Electrochemistry, Spectroscopy, and Reactivity of Uranium Complexes Supported by Ferrocene Diamide Ligands

    E-Print Network [OSTI]

    Duhovic, Selma

    2012-01-01T23:59:59.000Z

    J. L. , Pentavalent Uranium Chemistry-Synthetic Pursuit of aand High-Valent Uranium Chemistry. Organometallics 2011,for Trivalent Uranium Chemistry. Inorg. Chem. 1989, 28, (

  5. Bacterial Community Succession During in situ Uranium Bioremediation: Spatial Similarities Along Controlled Flow Paths

    E-Print Network [OSTI]

    Hwang, Chiachi

    2009-01-01T23:59:59.000Z

    problem, and the use of depleted uranium and other heavyenvironmental hazard. Depleted uranium is weakly radioactiveMB. (2004). Depleted and natural uranium: chemistry and

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

    SciTech Connect (OSTI)

    NONE

    1995-07-05T23:59:59.000Z

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

  7. Capstone Depleted Uranium Aerosols: Generation and Characterization

    SciTech Connect (OSTI)

    Parkhurst, MaryAnn; Szrom, Fran; Guilmette, Ray; Holmes, Tom; Cheng, Yung-Sung; Kenoyer, Judson L.; Collins, John W.; Sanderson, T. Ellory; Fliszar, Richard W.; Gold, Kenneth; Beckman, John C.; Long, Julie

    2004-10-19T23:59:59.000Z

    In a study designed to provide an improved scientific basis for assessing possible health effects from inhaling depleted uranium (DU) aerosols, a series of DU penetrators was fired at an Abrams tank and a Bradley fighting vehicle. A robust sampling system was designed to collect aerosols in this difficult environment and continuously monitor the sampler flow rates. Aerosols collected were analyzed for uranium concentration and particle size distribution as a function of time. They were also analyzed for uranium oxide phases, particle morphology, and dissolution in vitro. The resulting data provide input useful in human health risk assessments.

  8. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1990-01-01T23:59:59.000Z

    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.

  9. Measurements of Low-Enriched Uranium Holdup.

    SciTech Connect (OSTI)

    Belian, A. P. (Anthony P.); Reilly, T. D. (T. Douglas); Russo, P. A. (Phyllis A.); Tobin, S. J. (Stephen J.)

    2005-01-01T23:59:59.000Z

    A recent effort determined uranium holdup at a large fuel fabrication facility abroad where low enriched ({approx} 3%) uranium (LEU) oxide feeds the pellet manufacturing process. Measurements taken with both high- and low-resolution gamma-ray spectrometry systems include extensive data for the ventilation and vacuum systems. Equipment dimensions and the corresponding holdup deposit masses are large for LEU. Because deposits are infinitely thick to the 186 keV gamma ray in many locations in an LEU environment, measurements of both the 186 and 1001 keV gamma-rays were required, and self-attenuation was significant at 1001 keV in many cases. These wide-dynamic-range measruements used short count times, portable scintillator detectors, and portable MCAs. Because equipment is elevated above floor levels, most measurements were made with detectors mounted on extended telescoping poles. One of the main goals of this effort was to demonstrate and validate methods for measurement and quantitative analysis of LEU holdup using low-resolution detectors and the Generalized Geometry Holdup (GGH) techniques. The current GGH approach is applied elsewhere for holdup measurements of plutonium and high-enriched uranium. The recent experience is directly applicable to holdup measruements at LEU facilities such as the Paducah and Portmouth gaseous diffusion enrichment plants and elsewhere, including LEU sites where D and D is active. This report discusses the measurement methodology, calibration of the measurement equipment, measurement control, analysis of the data, and the global and local assay results including random and systematic uncertainties. It includes field-validation exercises (multiple calibrated systems that perform measruements on the same extended equipment) as well as quantitative validation results obtained on reference materials assembled to emulate the deposits in an extended vacuum line that was also measured by these techniques. The paper examines the differences in assay results between the low-resolution system using the GGH method and the high-resolution system utilizing the commercially available ISOCS analysis method.

  10. 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.

    2015-04-01T23:59:59.000Z

    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

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

    SciTech Connect (OSTI)

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

    2014-09-01T23:59:59.000Z

    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.

  12. Mica Surfaces Stabilize Pentavalent Uranium. | EMSL

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

    Ilton ES, A Haiduc, CL Cahill, and AR Felmy.2005."Mica Surfaces Stabilize Pentavalent Uranium."Inorganic Chemistry 44(9):2986-2988. Authors: ES Ilton A Haiduc CL Cahill AR...

  13. Process for reducing beta activity in uranium

    DOE Patents [OSTI]

    Briggs, Gifford G. (Cincinnatti, OH); Kato, Takeo R. (Cincinnatti, OH); Schonegg, Edward (Cleves, OH)

    1986-01-01T23:59:59.000Z

    This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which have undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed.

  14. Desorption of uranium from amidoxime fiber adsorbent

    SciTech Connect (OSTI)

    Goto, Akira; Morooka, Shigeharu; Fukamachi, Masakazu; Kusakabe, Katsuki (Kyushu Univ., Fukuoka (Japan)); Kago, Tokihiro (Towa Univ., Fukuoka (Japan))

    1993-10-01T23:59:59.000Z

    An amidoxime fibrous adsorbent is contacted with uranium-enriched seawater (10 ppm); about 10 mg uranium is loaded per 1 g dry fiber. Then the rate and yield of uranium desorption from the fiber are determined with various eluents. Acid solutions are superior to alkali carbonate solutions as eluents. With a 0.1 mol[center dot]L[sup [minus]1] HCl solution, desorption is completed in 2 hours regardless of the presence of uranium in the leaching solution up to 15 ppm ([approx]6 [times] 10[sup [minus]5]mol[center dot]L[sup [minus]1]). Serial operation of the adsorption-desorption cycle four times does not affect desorption efficiency, but the addition of heavy metal ions to the eluent at a level of 1.8 [times] 10[sup [minus]3]mol[center dot]L[sup [minus]1] significantly decreases desorption efficiency. 13 refs., 5 figs., 1 tab.

  15. Investigation of Trace Uranium in Biological Matrices

    E-Print Network [OSTI]

    Miller, James Christopher

    2013-05-31T23:59:59.000Z

    U.S. Department of Energy synthetic urine quality assurance standards from an inter-laboratory exercise in 2012. The separation apparatus was able to consistently separate uranium from the synthetic urine solutions with a consistent recovery between...

  16. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01T23:59:59.000Z

    Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

  17. In situ remediation of uranium contaminated groundwater

    SciTech Connect (OSTI)

    Dwyer, B.P.; Marozas, D.C. [Sandia National Labs., Albuquerque, NM (United States)

    1997-12-31T23:59:59.000Z

    In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment - various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field preliminary results are discussed with regard to other potential contaminated groundwater treatment applications.

  18. In situ remediation of uranium contaminated groundwater

    SciTech Connect (OSTI)

    Dwyer, B.P.; Marozas, D.C.

    1997-02-01T23:59:59.000Z

    In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field results are discussed with regard to other potential contaminated groundwater treatment applications.

  19. Process for reducing beta activity in uranium

    DOE Patents [OSTI]

    Briggs, G.G.; Kato, T.R.; Schonegg, E.

    1985-04-11T23:59:59.000Z

    This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed. 5 tabs.

  20. Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth

    2002-09-01T23:59:59.000Z

    The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: · Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, · Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, · Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, · Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and · Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.

  1. The ultimate disposition of depleted uranium

    SciTech Connect (OSTI)

    Lemons, T.R. [Uranium Enrichment Organization, Oak Ridge, TN (United States)

    1991-12-31T23:59:59.000Z

    Depleted uranium (DU) is produced as a by-product of the uranium enrichment process. Over 340,000 MTU of DU in the form of UF{sub 6} have been accumulated at the US government gaseous diffusion plants and the stockpile continues to grow. An overview of issues and objectives associated with the inventory management and the ultimate disposition of this material is presented.

  2. Investigation of Trace Uranium in Biological Matrices 

    E-Print Network [OSTI]

    Miller, James Christopher

    2013-05-31T23:59:59.000Z

    . This monitoring is often multi-faceted and typically involves an air sampling and biological sampling regime. The regime depends on the potential for exposures, the materials and chemical compounds being used, and the facility history. Specifically... Y-12 led the early US uranium enrichment programs, it also pioneered early uranium bioassay.[8] Likewise, the 5 Savannah River Site (SRS) pioneered plutonium bioassay techniques.[9] From these programs, techniques were developed to detect...

  3. Electrochemical method of producing eutectic uranium alloy and apparatus

    DOE Patents [OSTI]

    Horton, James A. (Livermore, CA); Hayden, H. Wayne (Oakridge, TN)

    1995-01-01T23:59:59.000Z

    An apparatus and method for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode.

  4. Characterization of uranium(VI) in seawater

    SciTech Connect (OSTI)

    Djogic, R.; Sipos, L.; Branica, M.

    1986-09-01T23:59:59.000Z

    The physicochemical characterization of uranium(VI) in seawater is described on the basis of species distribution calculations and experiments using polarography and spectrophotometry in artificial seawater at elevated uranium concentrations. Various dissolved uranium(VI) species are identified under different conditions of pH and carbonate concentration. Below pH 4, the hydrated uranyl ion is present in the free state (forming labile complexes). Above pH 4, a stepwise coordination of uranyl by the carbonate ion occurs. The monocarbonate complex is formed in the pH range 4-5, the bicarbonate uranyl complex between 5 and 6. Above pH 8, uranium is present predominately as the tricarbonate and to a smaller extent as a trihydroxide complex. There is satisfactory agreement between our experiments and the theoretically computed distribution of uranium(VI) in seawater based on published stability constants. The experiments done at higher concentrations are justified by theoretical distributions showing that there is no great difference in species distribution between the uranium at concentrations of 10/sup -4/ and /sup -8/ mol dm/sup -3/.

  5. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    SciTech Connect (OSTI)

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01T23:59:59.000Z

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  6. Method of removing niobium from uranium-niobium alloy

    SciTech Connect (OSTI)

    Pollock, E.N.; Schlier, D.S.; Shinopulos, G.

    1992-01-28T23:59:59.000Z

    This patent describes a method of removing niobium from a uranium-niobium alloy. It comprises dissolving the uranium-niobium alloy metal pieces in a first aqueous solution containing an acid selected from the group consisting of hydrochloric acid and sulfuric acid and fluoboric acid as a catalyst to provide a second aqueous solution, which includes uranium (U{sup +4}), acid radical ions, the acids insolubles including uranium oxides and niobium oxides; adding nitric acid to the insolubles to oxidize the niobium oxides to yield niobic acid and to complete the solubilization of any residual uranium; and separating the niobic acid from the nitric acid and solubilized uranium.

  7. Paragenesis and Geochronology of the Nopal I Uranium Deposit, Mexico

    SciTech Connect (OSTI)

    M. Fayek; M. Ren

    2007-02-14T23:59:59.000Z

    Uranium deposits can, by analogy, provide important information on the long-term performance of radioactive waste forms and radioactive waste repositories. Their complex mineralogy and variable elemental and isotopic compositions can provide important information, provided that analyses are obtained on the scale of several micrometers. Here, we present a structural model of the Nopal I deposit as well as petrography at the nanoscale coupled with preliminary U-Th-Pb ages and O isotopic compositions of uranium-rich minerals obtained by Secondary Ion Mass Spectrometry (SIMS). This multi-technique approach promises to provide ''natural system'' data on the corrosion rate of uraninite, the natural analogue of spent nuclear fuel.

  8. Fabrication options for depleted uranium components in shielded containers

    SciTech Connect (OSTI)

    Derrington, S.B.; Thompson, J.E.; Coates, C.W.

    1994-01-27T23:59:59.000Z

    Depleted uranium (DU) is an attractive material for the gamma-shielding components in containers designed for the storage, transport, and disposal of high-level radioactive wastes or spent nuclear fuel. The size and weight of these components present fabrication challenges. A broad range of technical expertise, capabilities, and facilities for uranium manufacturing and technology development exist at the Department of Energy laboratories and production facilities and within commercial industry. Several cast and wrought processes are available to fabricate the DU components. Integration of the DU fabrication capabilities and physical limitations for handling the DU components into the early design phase will ensure a fabricable product.

  9. Field-based detection and monitoring of uranium in contaminated groundwater using two immunosensors

    E-Print Network [OSTI]

    Melton, S.J.

    2010-01-01T23:59:59.000Z

    D. R. , Sustained removal of uranium from contaminated9. 18. Brina, R. , Uranium removal from contaminated water

  10. Detection of hexavalent uranium with inline and field-portable immunosensors

    E-Print Network [OSTI]

    Melton, Scott J.

    2009-01-01T23:59:59.000Z

    were able detect the removal of uranium from the groundwaterDR (2008) Sustained removal of uranium from contaminated

  11. Molecular analysis of phosphate limitation in Geobacteraceae during the bioremediation of a uranium-contaminated aquifer

    E-Print Network [OSTI]

    N'Guessan, L.A.

    2010-01-01T23:59:59.000Z

    DR (2008). Sustained Removal of Uranium From ContaminatedKomlos J et al (2007). Uranium removal from groundwater via

  12. Uranium Pyrophoricity Phenomena and Prediction (FAI/00-39)

    SciTech Connect (OSTI)

    PLYS, M.G.

    2000-10-10T23:59:59.000Z

    The purpose of this report is to provide a topical reference on the phenomena and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel (SNF) Project with specific applications to SNF Project processes and situations. Spent metallic uranium nuclear fuel is currently stored underwater at the K basins in the Hanford 100 area, and planned processing steps include: (1) At the basins, cleaning and placing fuel elements and scrap into stainless steel multi-canister overpacks (MCOs) holding about 6 MT of fuel apiece; (2) At nearby cold vacuum drying (CVD) stations, draining, vacuum drying, and mechanically sealing the MCOs; (3) Shipping the MCOs to the Canister Storage Building (CSB) on the 200 Area plateau; and (4) Welding shut and placing the MCOs for interim (40 year) dry storage in closed CSB storage tubes cooled by natural air circulation through the surrounding vault. Damaged fuel elements have exposed and corroded fuel surfaces, which can exothermically react with water vapor and oxygen during normal process steps and in off-normal situations, A key process safety concern is the rate of reaction of damaged fuel and the potential for self-sustaining or runaway reactions, also known as uranium fires or fuel ignition. Uranium metal and one of its corrosion products, uranium hydride, are potentially pyrophoric materials. Dangers of pyrophoricity of uranium and its hydride have long been known in the U.S. Department of Energy (Atomic Energy Commission/DOE) complex and will be discussed more below; it is sufficient here to note that there are numerous documented instances of uranium fires during normal operations. The motivation for this work is to place the safety of the present process in proper perspective given past operational experience. Steps in development of such a perspective are: (1) Description of underlying physical causes for runaway reactions, (2) Modeling physical processes to explain runaway reactions, (3) Validation of the method against experimental data, (4) Application of the method to plausibly explain operational experience, and (5) Application of the method to present process steps to demonstrate process safety and margin. Essentially, the logic above is used to demonstrate that runaway reactions cannot occur during normal SNF Project process steps, and to illustrate the depth of the technical basis for such a conclusion. Some off-normal conditions are identified here that could potentially lead to runaway reactions. However, this document is not intended to provide an exhaustive analysis of such cases. In summary, this report provides a ''toolkit'' of models and approaches for analysis of pyrophoricity safety issues at Hanford, and the technical basis for the recommended approaches. A summary of recommended methods appears in Section 9.0.

  13. Studies of epithermal neutrons in uranium, heavy water lattices

    E-Print Network [OSTI]

    D'Ardenne, Walter Herbert

    1964-01-01T23:59:59.000Z

    Measurements related to reactor physics parameters were made in three heavy water lattices. The three lattices studied consisted of 0.250-inch-diameter, 1.03 w/o U2 3 5 uranium fuel rods arranged in triangular arrays and ...

  14. Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

    SciTech Connect (OSTI)

    Gotovchikov, Vitaly; Seredenko, V.A.; Shatalov, V.V.; Mironov, B.S.; Kaplenkov, V.N.; Seredenko, A.V.; Saranchin, V.K.; Shulgin, A.S.; Kalmakov, Danila [All-Russian Research Institute of Chemical Technology (ARRICT), Kashirskoe Shosse 33, Moscow 115230 (Russian Federation); Haire, M.J.; Forsberg, C.W. [Oak Ridge National Laboratory - ORNL, 1 Bethel Valley Rd, Oak Ridge, TN 37830 (United States)

    2007-07-01T23:59:59.000Z

    Available in abstract form only. Full text of publication follows: This paper describes the results of joint research program of Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory to develop new materials for build spent nuclear fuel (SNF) storage, transport, and disposal casks using shielding made with depleted uranium dioxide (DUO{sub 2}) in a DUO{sub 2}-steel cermet or a DUCRETE with DUAGG (DUO{sub 2} aggregate) with selective additives in cement matrix. The preparation of DUO{sub 2} particles and aggregates for shielding could be produced from technologies that are extrapolated from the costly multi-step nuclear fuel pellet technologies. Melting the DUO{sub 2} and allowing it to freeze will produce a product near 100% theoretical density and assure that the product produces no volatile materials upon subsequent heating. Melting is a one step process that provides an opportunity to include additives in the DUO{sub 2} to modify its chemical or nuclear properties. The proposed work is directed to develop cold-wall induction heated melters (ICCM) for this specific application. Experiments on melting DUO{sub 2} were carried out in high frequency ICCM with cold crucible. It was experimentally proved an opportunity to produce molten DUO{sub 2} from mixed oxides (DU{sub 3}O{sub 8}) by reducing melting in ICCM. This will allow using DU{sub 3}O{sub 8} generated in direct conversion of depleted uranium hexafluoride as source material for melted and granulated DUO{sub 2} production. Experiments on the addition of alloying components - gadolinium oxide and others into DUO{sub 2} melt while in crucible to improve neutron and gamma radiation-shielding and operation properties of the final solids were carried out. (authors)

  15. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, Vladimir (Skokie, IL)

    1986-01-01T23:59:59.000Z

    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.

  16. Depleted uranium hexafluoride: The source material for advanced shielding systems

    SciTech Connect (OSTI)

    Quapp, W.J.; Lessing, P.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Cooley, C.R. [Department of Technology, Germantown, MD (United States)

    1997-02-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) has a management challenge and financial liability problem 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. DOE is evaluating several options for the disposition of this UF{sub 6}, including continued storage, disposal, and recycle into a product. Based on studies conducted to date, the most feasible recycle option for the depleted uranium is shielding in low-level waste, spent nuclear fuel, or vitrified high-level waste containers. Estimates for the cost of disposal, using existing technologies, range between $3.8 and $11.3 billion depending on factors such as the disposal site and the applicability of the Resource Conservation and Recovery Act (RCRA). Advanced technologies can reduce these costs, but UF{sub 6} disposal still represents large future costs. This paper describes an application for depleted uranium in which depleted uranium hexafluoride is converted into an oxide and then into a heavy aggregate. The heavy uranium aggregate is combined with conventional concrete materials to form an ultra high density concrete, DUCRETE, weighing more than 400 lb/ft{sup 3}. DUCRETE can be used as shielding in spent nuclear fuel/high-level waste casks at a cost comparable to the lower of the disposal cost estimates. Consequently, the case can be made that DUCRETE shielded casks are an alternative to disposal. In this case, a beneficial long term solution is attained for much less than the combined cost of independently providing shielded casks and disposing of the depleted uranium. Furthermore, if disposal is avoided, the political problems associated with selection of a disposal location are also avoided. Other studies have also shown cost benefits for low level waste shielded disposal containers.

  17. Analysis of strategies for improving uranium utilization in pressurized water reactors

    E-Print Network [OSTI]

    Sefcik, Joseph A.

    1981-01-01T23:59:59.000Z

    Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal ...

  18. The use of burnable poison to improve uranium utilization in PWRs

    E-Print Network [OSTI]

    Loh, Wee Tee

    1982-01-01T23:59:59.000Z

    A methodology based on the linear reactivity model of core behavior has been developed and employed to evaluate fuel management tactics for improving uranium utilization in Pressurized Water Reactors in a once-through fuel ...

  19. Uranium transformations in static microcosms.

    SciTech Connect (OSTI)

    Kelly, S. D.; Wu, W.; Yang, F.; Criddle, C.; Marsh, T. L.; O'Loughlin, E. J.; Ravel, B.; Watson, D.; Jardine, P. M.; Kemner, K. M.; Stanford Univ.; Michigan State Univ.; ORNL; BNL; EXAFS Analysis

    2010-01-01T23:59:59.000Z

    Elucidation of complex biogeochemical processes and their effects on speciation of U in the subsurface is critical for developing remediation strategies with an understanding of stability. We have developed static microcosms that are similar to bioreduction process studies in situ under laminar flow conditions or in sediment pores. Uranium L{sub 3}-edge X-ray absorption near-edge spectroscopy analysis with depth in the microcosms indicated that transformation of U{sup VI} to U{sup IV} occurred by at least two distinct processes. Extended X-ray absorption fine structure (EXAFS) analysis indicated that initial U{sup VI} species associated with C- and P-containing ligands were transformed to U{sup IV} in the form of uraninite and U associated with Fe-bound ligands. Microbial community analysis identified putative Fe{sup III} and sulfate reducers at two different depths in the microcosms. The slow reduction of U{sup VI} to U{sup IV} may contribute the stability of U{sup IV} within microcosms at 11 months after a decrease in bioreducing conditions due to limited electron donors.

  20. Potential incorporation of transuranics into uranium phases

    SciTech Connect (OSTI)

    Kim, C. W.; Wronkiewicz, D. J.; Buck, E. C.

    1999-12-07T23:59:59.000Z

    The UO{sub 2} in spent nuclear fuel is unstable under moist oxidizing conditions and will be altered to uranyl oxide hydrate phases. The transuranics released during the corrosion of spent fuel may also be incorporated into the structures of secondary U{sup 6+} phases. The incorporation of radionuclides into alteration products will affect their mobility. A series of precipitation tests were conducted at either 150 or 90 C for seven days to determine the potential incorporation of Ce{sup 4+} and Nd{sup 3+} (surrogates for Pu{sup 4+} and Am{sup 3+}, respectively) into uranium phases. Ianthinite ([U{sub 2}{sup 4+}(UO{sub 2}){sub 4}O{sub 6}(OH){sub 4}(H{sub 2}O){sub 4}](H{sub 2}O){sub 5}) was produced by dissolving uranium oxyacetate in a solution containing copper acetate monohydrate as a reductant. The leachant used in these tests were doped with either 2.1 ppm cerium or 399 ppm neodymium. Inductively coupled plasma-mass spectrometer (ICP-MS) analysis of the solid phase reaction products which were dissolved in a HNO{sub 3} solution indicates that about 306 ppm Ce (K{sub d} = 146) was incorporated into ianthinite, while neodymium contents were much higher, being approximately 24,800 ppm (K{sub d} = 62). Solid phase examinations using an analytical transmission electron microscope/electron energy-loss spectrometer (AEM/EELS) indicate a uniform distribution of Nd, while Ce contents were below detection. Becquerelite (Ca[(UO{sub 2}){sub 6}O{sub 4}(OH){sub 6}]{center_dot}8H{sub 2}O) was produced by dissolving uranium oxyacetate in a solution containing calcium acetate. The leachant in these tests was doped with either 2.1 ppm cerium or 277 ppm neodymium. ICP-MS results indicate that about 33 ppm Ce (K{sub d}=16) was incorporated into becquerelite, while neodymium contents were higher, being approximately 1,300 ppm (K{sub d}=5). Homogeneous distribution of Nd in the solid phase was noted during AEM/EELS examination, and Ce contents were also below detection.

  1. Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

    SciTech Connect (OSTI)

    Gotovchikov, V.T.; Seredenko, V.A.; Shatalov, V.V.; Mironov, B.S.; Kaplenkov, V.N.; Seredenko, A.V.; Saranchin, V.K.; Shulgin, A.S. [All-Russian Research Institute of Chemical Technology (ARRICT), Moscow (Russian Federation); Haire, M.J.; Forsberg, C.W. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    2007-07-01T23:59:59.000Z

    This paper describes the results of a joint research program between the Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory in the United States to develop new radiation shielding materials for use in the construction of casks for spent nuclear fuel (SNF) and radioactive wastes. Research and development is underway to develop SNF storage, transport, and disposal casks using shielding made with two new depleted uranium dioxide (DUO{sub 2}) materials: a DUO{sub 2}-steel cermet, and, DUCRETE with DUAGG (DUO{sub 2} aggregate). Melting the DUO{sub 2} and allowing it to freeze will produce a near 100% theoretical density product and assures that the product produces no volatile materials upon subsequent heating. Induction cold-crucible melters (ICCM) are being developed for this specific application. An ICCM is, potentially, a high throughput low-cost process. Schematics of a pilot facility were developed for the production of molten DUO{sub 2} from DU{sub 3}O{sub 8} to produce granules <1 mm in diameter in a continuous mode of operation. Thermodynamic analysis was conducted for uranium-oxygen system in the temperature range from 300 to 4000 K in various gas mediums. Temperature limits of stability for various uranium oxides were determined. Experiments on melting DUO{sub 2} were carried out in a high frequency ICCM in a cold crucible with a 120 mm in diameter. The microstructure of molten DUO{sub 2} was studied and lattice parameters were determined. It was experimentally proved, and validated by X-ray analysis, that an opportunity exists to produce molten DUO{sub 2} from mixed oxides (primarily DU{sub 3}O{sub 8}) by reduction melting in ICCM. This will allow using DU{sub 3}O{sub 8} directly to make DUO{sub 2}-a separate unit operation to produce UO{sub 2} feed material is not needed. Experiments were conducted concerning the addition of alloying components, gadolinium et al. oxides, into the DUO{sub 2} melt while in the crucible. These additives improve neutron and gamma radiation shielding and operation properties of the final solids. Cermet samples of 50 wt % DUO{sub 2} were produced. (authors)

  2. Molten uranium dioxide structure and dynamics

    SciTech Connect (OSTI)

    Skinner, L. B. [Argonne National Laboratory (ANL), Argonne, IL (United States); Stony Brook Univ., Stony Brook, NY (United States); Materials Development Inc., Arlington Heights, IL (United States); Parise, J. B. [Stony Brook Univ., Stony Brook, NY (United States); Benmore, C. J. [Argonne National Laboratory (ANL), Argonne, IL (United States); Weber, J. K.R. [Materials Development Inc., Arlington Heights, IL (United States); Williamson, M. A. [Argonne National Laboratory (ANL), Argonne, IL (United States); Tamalonis, A. [Materials Development Inc., Arlington Heights, IL (United States); Hebden, A. [Argonne National Laboratory (ANL), Argonne, IL (United States); Wiencek, T. [Argonne National Laboratory (ANL), Argonne, IL (United States); Alderman, O. L.G. [Materials Development Inc., Arlington Heights, IL (United States); Guthrie, M. [Carnegie Inst., Washington, DC (United States); Leibowitz, L. [Argonne National Laboratory (ANL), Argonne, IL (United States)

    2014-11-20T23:59:59.000Z

    Uranium dioxide (UO2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligible U-O coordination change. On melting, the average U-O coordination drops from 8 to 6.7 ± 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.

  3. Molten uranium dioxide structure and dynamics

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

    Skinner, L. B. [Argonne National Laboratory (ANL), Argonne, IL (United States); Stony Brook Univ., Stony Brook, NY (United States); Materials Development Inc., Arlington Heights, IL (United States); Parise, J. B. [Stony Brook Univ., Stony Brook, NY (United States); Benmore, C. J. [Argonne National Laboratory (ANL), Argonne, IL (United States); Weber, J. K.R. [Materials Development Inc., Arlington Heights, IL (United States); Williamson, M. A. [Argonne National Laboratory (ANL), Argonne, IL (United States); Tamalonis, A. [Materials Development Inc., Arlington Heights, IL (United States); Hebden, A. [Argonne National Laboratory (ANL), Argonne, IL (United States); Wiencek, T. [Argonne National Laboratory (ANL), Argonne, IL (United States); Alderman, O. L.G. [Materials Development Inc., Arlington Heights, IL (United States); Guthrie, M. [Carnegie Inst., Washington, DC (United States); Leibowitz, L. [Argonne National Laboratory (ANL), Argonne, IL (United States)

    2014-11-20T23:59:59.000Z

    Uranium dioxide (UO2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligible U-O coordination change. On melting, the average U-O coordination drops from 8 to 6.7 ± 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.

  4. Molten uranium dioxide structure and dynamics

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

    Skinner, L. B.; Parise, J. B.; Benmore, C. J.; Weber, J. K.R.; Williamson, M. A.; Tamalonis, A.; Hebden, A.; Wiencek, T.; Alderman, O. L.G.; Guthrie, M.; et al

    2014-11-20T23:59:59.000Z

    Uranium dioxide (UO2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligible U-O coordination change. Onmore »melting, the average U-O coordination drops from 8 to 6.7 ± 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.« less

  5. Prospects for the recovery of uranium from seawater

    E-Print Network [OSTI]

    Best, F. R.

    1980-01-01T23:59:59.000Z

    A computer program entitled URPE (Uranium Recovery Performance and Economics) has been developed to simulate the engineering performance and provide an economic analysis O of a plant recovering uranium from seawater. The ...

  6. EA-1290: Disposition of Russian Federation Titled Natural Uranium

    Broader source: Energy.gov [DOE]

    This EA evaluates the potential environmental impacts of a proposal to transport up to an average of 9,000 metric tons per year of natural uranium as uranium hexafluoride (UF6) from the United...

  7. Assessments of long-term uranium supply availability

    E-Print Network [OSTI]

    Zaterman, Daniel R

    2009-01-01T23:59:59.000Z

    The future viability of nuclear power will depend on the long-term availability of uranium. A two-form uranium supply model was used to estimate the date at which peak production will occur. The model assumes a constant ...

  8. Y-12 and the Ťsuper enriched Uranium 235?

    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...

  9. RESOLUTION OF URANIUM ISOTOPES WITH KINETIC PHOSPHORESCENCE ANALYSIS

    SciTech Connect (OSTI)

    Miley, Sarah M.; Hylden, Anne T.; Friese, Judah I.

    2013-04-01T23:59:59.000Z

    This study was conducted to test the ability of the Chemchek™ Kinetic Phosphorescence Analyzer Model KPA-11 with an auto-sampler to resolve the difference in phosphorescent decay rates of several different uranium isotopes, and therefore identify the uranium isotope ratios present in a sample. Kinetic phosphorescence analysis (KPA) is a technique that provides rapid, accurate, and precise determination of uranium concentration in aqueous solutions. Utilizing a pulsed-laser source to excite an aqueous solution of uranium, this technique measures the phosphorescent emission intensity over time to determine the phosphorescence decay profile. The phosphorescence intensity at the onset of decay is proportional to the uranium concentration in the sample. Calibration with uranium standards results in the accurate determination of actual concentration of the sample. Different isotopes of uranium, however, have unique properties which should result in different phosphorescence decay rates seen via KPA. Results show that a KPA is capable of resolving uranium isotopes.

  10. abandoned uranium mines: Topics by E-print Network

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

    residents. 3.1.1 On-Site Recreation Since most uranium locations are on federal lands 91 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

  11. Monitoring Uranium Transformations Determined by the Evolution of Biogeochemical Processes

    SciTech Connect (OSTI)

    Marsh, Terence L.

    2013-07-30T23:59:59.000Z

    Our contribution to the larger project (ANL) was the phylogenetic analysis of evolved communities capable of reducing metals including uranium.

  12. The radioactive Substances (Uranium and Thorium) Exemption Order 1962 

    E-Print Network [OSTI]

    Joseph, Keith

    1962-01-01T23:59:59.000Z

    STATUTORY INSTRUMENTS 1962 No.2710 ATOMIC ENERGY AND RADIOACTIVE SUBSTANCES The Radioactive Substances (Uranium and Thorium) Exemption Order 1962...

  13. Modeling Uranium-Proton Ion Exchange in Biosorption

    E-Print Network [OSTI]

    Volesky, Bohumil

    seaweed biomass was used to remove the heavy metal uranium from the aqueous solution. Uranium biosorptionModeling Uranium-Proton Ion Exchange in Biosorption J I N B A I Y A N G A N D B O H U M I L V O L E, Quebec, Canada H3A 2B2 Biosorption of uranium metal ions by a nonliving protonated Sargassum fluitans

  14. Depleted Uranium in Kosovo Post-Conflict Environmental Assessment

    E-Print Network [OSTI]

    Unep Scientific; Mission Kosovo

    2.1 UNEP’s role in post-conflict environmental assessment................................................9 2.2 Depleted uranium............................................................10

  15. TRIMOLECULAR REACTIONS OF URANIUM HEXAFLUORIDE WITH WATER

    SciTech Connect (OSTI)

    Westbrook, M.; Becnel, J.; Garrison, S.

    2010-02-25T23:59:59.000Z

    The hydrolysis reaction of uranium hexafluoride (UF{sub 6}) is a key step in the synthesis of uranium dioxide (UO{sub 2}) powder for nuclear fuels. Mechanisms for the hydrolysis reactions are studied here with density functional theory and the Stuttgart small-core scalar relativistic pseudopotential and associated basis set for uranium. The reaction of a single UF{sub 6} molecule with a water molecule in the gas phase has been previously predicted to proceed over a relatively sizeable barrier of 78.2 kJ {center_dot} mol{sup -1}, indicating this reaction is only feasible at elevated temperatures. Given the observed formation of a second morphology for the UO{sub 2} product coupled with the observations of rapid, spontaneous hydrolysis at ambient conditions, an alternate reaction pathway must exist. In the present work, two trimolecular hydrolysis mechanisms are studied with density functional theory: (1) the reaction between two UF{sub 6} molecules and one water molecule, and (2) the reaction of two water molecules with a single UF{sub 6} molecule. The predicted reaction of two UF{sub 6} molecules with one water molecule displays an interesting 'fluorine-shuttle' mechanism, a significant energy barrier of 69.0 kJ {center_dot} mol{sup -1} to the formation of UF{sub 5}OH, and an enthalpy of reaction ({Delta}H{sub 298}) of +17.9 kJ {center_dot} mol{sup -1}. The reaction of a single UF{sub 6} molecule with two water molecules displays a 'proton-shuttle' mechanism, and is more favorable, having a slightly lower computed energy barrier of 58.9 kJ {center_dot} mol{sup -1} and an exothermic enthalpy of reaction ({Delta}H{sub 298}) of -13.9 kJ {center_dot} mol{sup -1}. The exothermic nature of the overall UF{sub 6} + 2 {center_dot} H{sub 2}O trimolecular reaction and the lowering of the barrier height with respect to the bimolecular reaction are encouraging; however, the sizable energy barrier indicates further study of the UF{sub 6} hydrolysis reaction mechanism is warranted to resolve the remaining discrepancies between the predicted mechanisms and experimental observations.

  16. Candidate processes for diluting the {sup 235}U isotope in weapons-capable highly enriched uranium

    SciTech Connect (OSTI)

    Snider, J.D.

    1996-02-01T23:59:59.000Z

    The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile {sup 235}U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile {sup 235}U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel.

  17. Tables des principaux minerais d'uranium et de thorium

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    233 Tables des principaux minerais d'uranium et de thorium Par B. SZILARD [Faculté des Sciences de minerais d'uranium et de thorium avec leurs données les plus importantes, telles que la com- position, la teneur en uranium et en thorium, la provenance et quelques indications générales. La liste ne prétend pas

  18. Estimating terrestrial uranium and thorium by antineutrino flux measurements

    E-Print Network [OSTI]

    Mcdonough, William F.

    Estimating terrestrial uranium and thorium by antineutrino flux measurements Stephen T. Dye, and approved November 16, 2007 (received for review July 11, 2007) Uranium and thorium within the Earth produce of uranium and thorium concentrations in geological reservoirs relies largely on geochemi- cal model

  19. Microbial Janitors: Enabling natural microbes to clean up uranium contamination

    E-Print Network [OSTI]

    of Energy's Environmental Remediation Sciences Program. Q: How can uranium be removed or neutralized so in the contaminated subsurface and engineering the subsurface environment to stimulate nitrate removal and uraniumMicrobial Janitors: Enabling natural microbes to clean up uranium contamination Oak Ridge

  20. Appendix IV. Risks Associated with Conventional Uranium Milling Introduction

    E-Print Network [OSTI]

    ", uranium is removed from the processed ore with sulfuric acid. Sodium chlorate is also addedAppendix IV. Risks Associated with Conventional Uranium Milling Operations Introduction Although uranium mill tailings are considered byproduct materials under the AEA and not TENORM, EPA's Science

  1. EPA Uranium Program Update Loren W. Setlow and

    E-Print Network [OSTI]

    EPA Uranium Program Update Loren W. Setlow and Reid J. Rosnick Environmental Protection Agency Office of Radiation and Indoor Air (6608J) Washington, DC 20460 NMA/NRC Uranium Recovery Workshop April 30, 2008 #12;2 Overview EPA Radiation protection program Uranium reports and abandoned mine lands

  2. Standard Review Plan for In Situ Leach Uranium

    E-Print Network [OSTI]

    NUREG-1569 Standard Review Plan for In Situ Leach Uranium Extraction License Applications Final Washington, DC 20555-0001 #12;NUREG-1569 Standard Review Plan for In Situ Leach Uranium Extraction License OF A STANDARD REVIEW PLAN (NUREG­1569) FOR STAFF REVIEWS FOR IN SITU LEACH URANIUM EXTRACTION LICENSE

  3. Uranium Mill Tailings Remedial Action Project surface project management plan

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    This Project Management Plan describes the planning, systems, and organization that shall be used to manage the Uranium Mill Tailings Remedial Action Project (UMTRA). US DOE is authorized to stabilize and control surface tailings and ground water contamination at 24 inactive uranium processing sites and associated vicinity properties containing uranium mill tailings and related residual radioactive materials.

  4. Bioremediation of Uranium Plumes with Nano-scale

    E-Print Network [OSTI]

    Fay, Noah

    (IV) (UO2[s], uraninite) Anthropogenic · Release of mill tailings during uranium mining - MobilizationBioremediation of Uranium Plumes with Nano-scale Zero-valent Iron Angela Athey Advisers: Dr. Reyes Undergraduate Student Fellowship Program April 15, 2011 #12;Main Sources of Uranium Natural · Leaching from

  5. Depleted uranium plasma reduction system study

    SciTech Connect (OSTI)

    Rekemeyer, P.; Feizollahi, F.; Quapp, W.J.; Brown, B.W.

    1994-12-01T23:59:59.000Z

    A system life-cycle cost study was conducted of a preliminary design concept for a plasma reduction process for converting depleted uranium to uranium metal and anhydrous HF. The plasma-based process is expected to offer significant economic and environmental advantages over present technology. Depleted Uranium is currently stored in the form of solid UF{sub 6}, of which approximately 575,000 metric tons is stored at three locations in the U.S. The proposed system is preconceptual in nature, but includes all necessary processing equipment and facilities to perform the process. The study has identified total processing cost of approximately $3.00/kg of UF{sub 6} processed. Based on the results of this study, the development of a laboratory-scale system (1 kg/h throughput of UF6) is warranted. Further scaling of the process to pilot scale will be determined after laboratory testing is complete.

  6. Method for fluorination of uranium oxide

    DOE Patents [OSTI]

    Petit, George S. (Oak Ridge, TN)

    1987-01-01T23:59:59.000Z

    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.

  7. Reactor physics studies for assessment of tramp uranium methods

    SciTech Connect (OSTI)

    Grimm, P.; Vasiliev, A.; Wieselquist, W.; Ferroukhi, H. [Paul Scherrer Institut, CH 5232 Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, CH 5325 Leibstadt (Switzerland)

    2012-07-01T23:59:59.000Z

    This paper presents calculation studies towards validation of a methodology for estimations of the tramp uranium mass from water chemistry measurements. Particular emphasis is given to verify, from a reactor physics point of view, the justification basis for the so-called 'Pu-based model' versus the 'U-based model' as a key assumption for the methodology. The computational studies are carried out for a typical BWR fuel assembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclides and the fraction of {sup 235}U fissions to total fissions in different zones of a fuel rod, including tramp uranium on the clad surface, it is found that Pu gives the dominant contribution to fissions for tramp uranium after an irradiation on the outer clad surface of at least one cycle in a BWR. Thus, the use of the so-called Pu model for the determination of the tramp uranium mass (this means in particular using the yields for {sup 239}Pu fission) appears justified in the cases considered. On that basis, replacing the older U model by a Pu model is recommended. (authors)

  8. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, V.

    1984-06-13T23:59:59.000Z

    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.

  9. Floating plant can get uranium from seawater

    SciTech Connect (OSTI)

    Not Available

    1984-02-01T23:59:59.000Z

    A floating plant has been designed to extract uranium from seawater using solid adsorbents. Ore is removed from the adsorbent material by means of a solvent and concentrated in ion exchangers. Seawater is supplied to the adsorbent inside by wave energy and is based on the principle that waves will rush up a sloping plane that is partly submerged and fill a reservoir to a level higher than the still water level in the sea. The company projects that an offshore plant for recovering 600 tons of uranium/yr would comprise 22 floating concrete units, each measuring 430 x 75 meters.

  10. Decarburization of uranium via electron beam processing

    SciTech Connect (OSTI)

    McKoon, R H

    1998-10-23T23:59:59.000Z

    For many commercial and military applications, the successive Vacuum Induction Melting of uranium metal in graphite crucibles results in a product which is out of specification in carbon. The current recovery method involves dissolution of the metal in acid and chemical purification. This is both expensive and generates mixed waste. A study was undertaken at Lawrence Livermore National Laboratory to investigate the feasibility of reducing the carbon content of uranium metal using electron beam techniques. Results will be presented on the rate and extent of carbon removal as a function of various operating parameters.

  11. Progress toward uranium scrap recycling via EBCHR

    SciTech Connect (OSTI)

    McKoon, R.H.

    1994-11-01T23:59:59.000Z

    A 250 kW electron beam cold hearth refining (EBCHR) melt furnace at Lawrence Livermore National Laboratory (LLNL) has been in operation for over a year producing 5.5 in.-diameter ingots of various uranium alloys. Production of in-specification uranium-6%-niobium (U-6Nb) alloy ingots has been demonstrated using virgin feedstock. A vibratory scrap feeder has been installed on the system and the ability to recycle chopped U-6Nb scrap has been established. A preliminary comparison of vacuum arc remelted (VAR) and electron beam (EB) melted product is presented.

  12. Simplifying strong electronic correlations in uranium: Localized uranium heavy-fermion UM2Zn20 (M=Co,Rh) compounds

    E-Print Network [OSTI]

    Lawrence, Jon

    Simplifying strong electronic correlations in uranium: Localized uranium heavy-fermion UM2Zn20 (M Atómica, 8400 Bariloche, Argentina 6 Department of Chemistry and Biochemistry, University of Delaware-field effects corroborate an ionic-like uranium electronic configura- tion in UM2Zn20. DOI: 10.1103/PhysRevB.78

  13. BLENDING LOW ENRICHED URANIUM WITH DEPLETED URANIUM TO CREATE A SOURCE MATERIAL ORE THAT CAN BE PROCESSED FOR THE RECOVERY OF YELLOWCAKE AT A CONVENTIONAL URANIUM MILL

    SciTech Connect (OSTI)

    Schutt, Stephen M.; Hochstein, Ron F.; Frydenlund, David C.; Thompson, Anthony J.

    2003-02-27T23:59:59.000Z

    Throughout the United States Department of Energy (DOE) complex, there are a number of streams of low enriched uranium (LEU) that contain various trace contaminants. These surplus nuclear materials require processing in order to meet commercial fuel cycle specifications. To date, they have not been designated as waste for disposal at the DOE's Nevada Test Site (NTS). Currently, with no commercial outlet available, the DOE is evaluating treatment and disposal as the ultimate disposition path for these materials. This paper will describe an innovative program that will provide a solution to DOE that will allow disposition of these materials at a cost that will be competitive with treatment and disposal at the NTS, while at the same time recycling the material to recover a valuable energy resource (yellowcake) for reintroduction into the commercial nuclear fuel cycle. International Uranium (USA) Corporation (IUSA) and Nuclear Fuel Services, Inc. (NFS) have entered into a commercial relationship to pursue the development of this program. The program involves the design of a process and construction of a plant at NFS' site in Erwin, Tennessee, for the blending of contaminated LEU with depleted uranium (DU) to produce a uranium source material ore (USM Ore{trademark}). The USM Ore{trademark} will then be further processed at IUC's White Mesa Mill, located near Blanding, Utah, to produce conventional yellowcake, which can be delivered to conversion facilities, in the same manner as yellowcake that is produced from natural ores or other alternate feed materials. The primary source of feed for the business will be the significant sources of trace contaminated materials within the DOE complex. NFS has developed a dry blending process (DRYSM Process) to blend the surplus LEU material with DU at its Part 70 licensed facility, to produce USM Ore{trademark} with a U235 content within the range of U235 concentrations for source material. By reducing the U235 content to source material levels in this manner, the material will be suitable for processing at a conventional uranium mill under its existing Part 40 license to remove contaminants and enable the product to re-enter the commercial fuel cycle. The tailings from processing the USM Ore{trademark} at the mill will be permanently disposed of in the mill's tailings impoundment as 11e.(2) byproduct material. Blending LEU with DU to make a uranium source material ore that can be returned to the nuclear fuel cycle for processing to produce yellowcake, has never been accomplished before. This program will allow DOE to disposition its surplus LEU and DU in a cost effective manner, and at the same time provide for the recovery of valuable energy resources that would be lost through processing and disposal of the materials. This paper will discuss the nature of the surplus LEU and DU materials, the manner in which the LEU will be blended with DU to form a uranium source material ore, and the legal means by which this blending can be accomplished at a facility licensed under 10 CFR Part 70 to produce ore that can be processed at a conventional uranium mill licensed under 10 CFR Part 40.

  14. Recovery of uranium from seawater. 7; Concentration and separation of uranium in acidic eluate

    SciTech Connect (OSTI)

    Egawa, H.; Nonaka, T. (Dept. of Applied Chemistry, Faculty of Engineering, Kumamoto Univ., Kurokami 2-39-1, Kumamoto 860 (JP)); Nakayama, M. (Faculty of Pharmaceutical Sciences, Kumamoto Univ., Oe-Honmachi 5-1, Kumamoto 862 (JP))

    1990-11-01T23:59:59.000Z

    This paper reports on macroporous chelating resins (RSP, RSPO, RCSP, and RCSPO) containing dihydroxphosphino and/or -phosphono groups were examined for the concentration and separation of uranium from acidic eluates of macroporous chelating resin containing amidoxime groups. RSP and RSPO had a high adsorption capacity for uranium even in 0.25-0.50 mol {center dot} dm{sup {minus}3} H{sub 2}SO{sub 4}. Uranium adsorbed on the resins was eluted easily as a uranyl carbonate complex by use of 0.25 mol {center dot} dm{sup {minus}3} Na{sub 2}CO{sub 3}. In this effluent, other metal ions were hardly present. The use of RSP and RSPO was very effective in concentrating uranium from seawater and separating it from most other elements.

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

    SciTech Connect (OSTI)

    NONE

    1995-07-05T23:59:59.000Z

    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.

  16. Updated Conceptual Model for the 300 Area Uranium Groundwater Plume

    SciTech Connect (OSTI)

    Zachara, John M.; Freshley, Mark D.; Last, George V.; Peterson, Robert E.; Bjornstad, Bruce N.

    2012-11-01T23:59:59.000Z

    The 300 Area uranium groundwater plume in the 300-FF-5 Operable Unit is residual from past discharge of nuclear fuel fabrication wastes to a number of liquid (and solid) disposal sites. The source zones in the disposal sites were remediated by excavation and backfilled to grade, but sorbed uranium remains in deeper, unexcavated vadose zone sediments. In spite of source term removal, the groundwater plume has shown remarkable persistence, with concentrations exceeding the drinking water standard over an area of approximately 1 km2. The plume resides within a coupled vadose zone, groundwater, river zone system of immense complexity and scale. Interactions between geologic structure, the hydrologic system driven by the Columbia River, groundwater-river exchange points, and the geochemistry of uranium contribute to persistence of the plume. The U.S. Department of Energy (DOE) recently completed a Remedial Investigation/Feasibility Study (RI/FS) to document characterization of the 300 Area uranium plume and plan for beginning to implement proposed remedial actions. As part of the RI/FS document, a conceptual model was developed that integrates knowledge of the hydrogeologic and geochemical properties of the 300 Area and controlling processes to yield an understanding of how the system behaves and the variables that control it. Recent results from the Hanford Integrated Field Research Challenge site and the Subsurface Biogeochemistry Scientific Focus Area Project funded by the DOE Office of Science were used to update the conceptual model and provide an assessment of key factors controlling plume persistence.

  17. 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 [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01T23:59:59.000Z

    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.

  18. Extraction of uranium from seawater using magnetic adsorbents

    SciTech Connect (OSTI)

    Yamashita, H. (Hitachi Research Lab., Japan); Fujita, K.; Nakajima, F.; Ozawa, Y.; Murata, T.

    1981-01-01T23:59:59.000Z

    A new process for the extraction of uranium from seawater was developed. In the process, uranium adsorption is effected using powdered magnetic adsorbents; the adsorbents are then separated from seawater using magnetic separation technology. This process is superior to a column method using a granulated hydrous titanium oxide adsorber bed in the following ways: (1) a higher rate of adsorption is realized because smaller particles are used in the uranium adsorption; and (2) blocking, which is inevitable in an adsorber bed, is eliminated. The composite hydrous titanium-iron oxide as a magnetic adsorbent having high uranium adsorption capacity and magnetization can be prepared by adding urea to a mixed solution of titanium sulfate and ferrous sulfate. Adsorption and desoprtion of uranium and the removal of the adsorbent using a small-scale uranium extraction plant (about 15 m/sup 3//d) is reported, and the feasibility of uranium extraction from seawater by this process is demonstrated. 10 figures.

  19. New aspects of uranium recovery from seawater

    SciTech Connect (OSTI)

    Hetkamp, D.; Wagener, K.

    1982-10-01T23:59:59.000Z

    The properties of various adsorbents for uranium extraction from seawater are measured under standardized experimental conditions. It turns out that fractionated humic acids have exceptionally fast loading kinetics. This property leads to a substantial reduction of capital investments in conventional adsorbent bed techniques as well as in a procedure designed to avoid large adsorbent bed constructions by using carrier bodies in the open sea.

  20. Phosphate Barriers for Immobilization of Uranium Plumes

    SciTech Connect (OSTI)

    Burns, Peter C.

    2004-12-01T23:59:59.000Z

    Uranium contamination of the subsurface remains a persistent problem plaguing remedial design at sites across the U.S. that were involved with production, handling, storage, milling, and reprocessing of uranium for both civilian and defense related purposes. Remediation efforts to date have relied upon excavation, pump-and-treat, or passive remediation barriers (PRB?s) to remove or attenuate uranium mobility. Documented cases convincingly demonstrate that excavation and pump-and-treat methods are ineffective for a number of highly contaminated sites. There is growing concern that use of conventional PRB?s, such as zero-valent iron, may be a temporary solution to a problem that will persist for thousands of years. Alternatives to the standard treatment methods are therefore warranted. The core objective of our research is to demonstrate that a phosphorus amendment strategy will result in a reduction of dissolved uranium to below the proposed drinking water standard. Our hypothesis is that long-chain sodium polyphosphate compounds forestall precipitation of sparingly soluble uranyl phosphate compounds, which is paramount to preventing fouling of wells at the point of injection.

  1. The multiphoton ionization of uranium hexafluoride

    SciTech Connect (OSTI)

    Armstrong, D.P. (Oak Ridge K-25 Site, TN (United States). UEO Enrichment Technical Operations Div.)

    1992-05-01T23:59:59.000Z

    Multiphoton ionization (MPI) time-of-flight mass spectroscopy and photoelectron spectroscopy studies of UF{sub 6} have been conducted using focused light from the Nd:YAG laser fundamental ({lambda}=1064 nm) and its harmonics ({lambda}=532, 355, or 266 nm), as well as other wavelengths provided by a tunable dye laser. The MPI mass spectra are dominated by the singly and multiply charged uranium ions rather than by the UF{sub x}{sup +} fragment ions even at the lowest laser power densities at which signal could be detected. The laser power dependence of U{sup n+} ions signals indicates that saturation can occur for many of the steps required for their ionization. In general, the doubly-charged uranium ion (U{sup 2+}) intensity is much greater than that of the singly-charged uranium ion (U{sup +}). For the case of the tunable dye laser experiments, the U{sup n+} (n = 1- 4) wavelength dependence is relatively unstructured and does not show observable resonance enhancement at known atomic uranium excitation wavelengths. The dominance of the U{sup 2+} ion and the absence or very small intensities of UF{sub x}{sup +} fragments, along with the unsaturated wavelength dependence, indicate that mechanisms may exist other than ionization of bare U atoms after the stepwise photodissociation of F atoms from the parent molecule.

  2. The Quest for the Heaviest Uranium Isotope

    E-Print Network [OSTI]

    S. Schramm; D. Gridnev; D. V. Tarasov; V. N. Tarasov; W. Greiner

    2012-01-17T23:59:59.000Z

    We study Uranium isotopes and surrounding elements at very large neutron number excess. Relativistic mean field and Skyrme-type approaches with different parametrizations are used in the study. Most models show clear indications for isotopes that are stable with respect to neutron emission far beyond N=184 up to the range of around N=258.

  3. Geodatabase of the South Texas Uranium District

    E-Print Network [OSTI]

    Mark Beaman; William Wade Mcgee

    Uranium and its associated trace elements and radionuclides are ubiquitous in the South Texas Tertiary environment. Surface mining of this resource from the 1960s through the early 1980s at over sixty locations has left an extensive anthropological footprint (Fig. 1) in the lower Nueces and San Antonio river basins. Reclamation of mining initiated after 1975 has been under the regulatory authority of the Railroad Commission of Texas (RCT). However, mines that were active before the Texas Surface Mining Act of 1975 was enacted, and never reclaimed, are now considered abandoned. The Abandoned Mine Land Section of the RCT is currently reclaiming these pre-regulation uranium mines with funding from the federal government. The RCT monitors the overall effectiveness of this process through post-reclamation radiation and vegetative cover surveys, water quality testing, slope stability and erosion control monitoring. Presently a number of graduate and postgraduate students are completing research on the watershed and reservoir distribution of trace elements and radionuclides downstream of the South Texas Uranium District. The question remains as to whether the elevated levels of uranium, its associated trace elements and radiation levels in the South Texas environment are due to mining

  4. Uranium Sequestration by Aluminum Phosphate Minerals in Unsaturated Soils

    SciTech Connect (OSTI)

    Jerden, James L. Jr. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL, 60439 (United States)

    2007-07-01T23:59:59.000Z

    A mineralogical and geochemical study of soils developed from the unmined Coles Hill uranium deposit (Virginia) was undertaken to determine how phosphorous influences the speciation of uranium in an oxidizing soil/saprolite system typical of the eastern United States. This paper presents mineralogical and geochemical results that identify and quantify the processes by which uranium has been sequestered in these soils. It was found that uranium is not leached from the saturated soil zone (saprolites) overlying the deposit due to the formation of a sparingly soluble uranyl phosphate mineral of the meta-autunite group. The concentration of uranium in the saprolites is approximately 1000 mg uranium per kg of saprolite. It was also found that a significant amount of uranium was retained in the unsaturated soil zone overlying uranium-rich saprolites. The uranium concentration in the unsaturated soils is approximately 200 mg uranium per kg of soil (20 times higher than uranium concentrations in similar soils adjacent to the deposit). Mineralogical evidence indicates that uranium in this zone is sequestered by a barium-strontium-calcium aluminum phosphate mineral of the crandallite group (gorceixite). This mineral is intimately inter-grown with iron and manganese oxides that also contain uranium. The amount of uranium associated with both the aluminum phosphates (as much as 1.4 weight percent) has been measured by electron microprobe micro-analyses and the geochemical conditions under which these minerals formed has been studied using thermodynamic reaction path modeling. The geochemical data and modeling results suggest the meta-autunite group minerals present in the saprolites overlying the deposit are unstable in the unsaturated zone soils overlying the deposit due to a decrease in soil pH (down to a pH of 4.5) at depths less than 5 meters below the surface. Mineralogical observations suggest that, once exposed to the unsaturated environment, the meta-autunite group minerals react to form U(VI)- bearing aluminum phosphates. (author)

  5. Standard test method for determination of uranium or gadolinium (or both) in gadolinium oxide-uranium oxide pellets or by X-ray fluorescence (XRF)

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2008-01-01T23:59:59.000Z

    Standard test method for determination of uranium or gadolinium (or both) in gadolinium oxide-uranium oxide pellets or by X-ray fluorescence (XRF)

  6. Control of structure and reactivity by ligand design : applications to small molecule activation by low-valent uranium complexes

    E-Print Network [OSTI]

    Lam, Oanh Phi

    2010-01-01T23:59:59.000Z

    researchers from uranium chemistry. Fortunately, despitescarce in uranium coordination chemistry. A more detailedligands for uranium coordination chemistry. Figure 4-2.

  7. Microstructural Evolution and Radiation Effects of Uranium-Bearing Diffusion Couples

    E-Print Network [OSTI]

    Wei, Chao-Chen

    2014-12-12T23:59:59.000Z

    -cladding chemical interaction, or FCCI, is one of the primary material problems during reactor operations. A series of tests using uranium-bearing fuel alloys and various cladding materials were performed to assess the diffusional interactions. However...

  8. EA-1255: Project Partnership Transportation of Foreign-Owned Enriched Uranium from the Republic of Georgia

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to transport 5.26 kilograms of enriched uranium-23 5 in the form of nuclear fuel, from the Republic of Georgia to the United Kingdom.

  9. Extraction of Uranium from Seawater: Design and Testing of a Symbiotic System

    E-Print Network [OSTI]

    Wu, You

    Uranium present in low concentration in ocean water has the potential to greatly augment the current fuel reserve for nuclear power generation, but the challenge of extracting it economically remains. Two new designs of ...

  10. Comprehensive Investigation of the Uranium-Zirconium Alloy System: Thermophysical Properties, Phase Characterization and Ion Implantation Effects 

    E-Print Network [OSTI]

    Ahn, Sangjoon

    2013-07-31T23:59:59.000Z

    be solved even though consensus is difficult to achieve. Fresh nuclear fuel typically starts as pure uranium dioxide, but spent nuclear fuel is an complex compound comprised of about 95 % of valuable resources, uranium and plutonium, can be recycled.../Analyzer ix TRU Transuranic element U-Zr Uranium-Zirconium U-Pu-Zr Uranium-Plutonium-Zirconium XEDS X-ray Energy Dispersive Spectroscopy XRD X-ray Diffraction WDS Wavelength Dispersive Spectroscopy ZOLZ Zeroth Order Laue Zone x TABLE OF CONTENTS...

  11. Chemical Equilibrium of the Dissolved Uranium in Groundwaters From a Spanish Uranium-Ore Deposit

    SciTech Connect (OSTI)

    Garralon, Antonio; Gomez, Paloma; Turrero, Maria Jesus; Buil, Belen; Sanchez, Lorenzo [Departamento de Medio Ambiente, CIEMAT, Avda. Complutense 22. Edificio 19, Madrid, 28040 (Spain)

    2007-07-01T23:59:59.000Z

    The main objectives of this work are to determine the hydrogeochemical evolution of an uranium ore and identify the main water/rock interaction processes that control the dissolved uranium content. The Mina Fe uranium-ore deposit is the most important and biggest mine worked in Spain. Sageras area is located at the north part of the Mina Fe, over the same ore deposit. The uranium deposit was not mined in Sageras and was only perturbed by the exploration activities performed 20 years ago. The studied area is located 10 Km northeast of Ciudad Rodrigo (Salamanca) at an altitude over 650 m.a.s.l. The uranium mineralization is related to faults affecting the metasediments of the Upper Proterozoic to Lower Cambrian schist-graywacke complex (CEG), located in the Centro-Iberian Zone of the Hesperian Massif . The primary uranium minerals are uraninite and coffinite but numerous secondary uranium minerals have been formed as a result of the weathering processes: yellow gummite, autunite, meta-autunite, torbernite, saleeite, uranotile, ianthinite and uranopilite. The water flow at regional scale is controlled by the topography. Recharge takes place mainly in the surrounding mountains (Sierra Pena de Francia) and discharge at fluvial courses, mainly Agueda and Yeltes rivers, boundaries S-NW and NE of the area, respectively. Deep flows (lower than 100 m depth) should be upwards due to the river vicinity, with flow directions towards the W, NW or N. In Sageras-Mina Fe there are more than 100 boreholes drilled to investigate the mineral resources of the deposit. 35 boreholes were selected in order to analyze the chemical composition of groundwaters based on their depth and situation around the uranium ore. Groundwater samples come from 50 to 150 m depth. The waters are classified as calcium-bicarbonate type waters, with a redox potential that indicates they are slightly reduced (values vary between 50 to -350 mV). The TOC varies between <0.1 and 4.0 mgC/L and the dissolved uranium has a maximum value of 7.7 mg/L. According the analytical data of dissolved uranium, the mineral closest to equilibrium seems to be UO{sub 2}(am). The tritium contents in the groundwaters vary between 1.5 and 7.3 T.U. Considering that the mean value of tritium in rainwater from the studied area has a value of 4 T.U., it can be concluded that the residence times of the groundwaters are relatively short, not longer than 50 years in the oldest case. (authors)

  12. Uranium Mill Tailings Radiation Control Act Sites Fact Sheet

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group current C3EDepartment of Energy OfficeFact Sheet Uranium Mill Tailings

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

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3.1. Unfilled uranium

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

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at Commercial and InstitutionalArea:Mnt(N)3.1. Unfilled uranium4.

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron SpinPrinceton PlasmaAfternoon4. Uranium purchased by ownersAbout /

  16. Uranium Track Team | Y-12 National Security Complex

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron SpinPrinceton PlasmaAfternoon4. Uranium purchased by ownersAboutTrack

  17. Uranium at Y-12: Inspection | Y-12 National Security Complex

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron SpinPrinceton PlasmaAfternoon4. Uranium purchased by...

  18. Uranium concentrate production in the United States, 2004-13

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron SpinPrinceton PlasmaAfternoon4. Uranium purchased by...Recovery

  19. Uranium Leasing Program Environmental Documents | Department of Energy

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual Siteof Energy 2, 2015 -Helicopter-Japan Joint Nuclear D.C. *ofUraniumEnvironmental

  20. The Uranium Processing Facility Finite Element Meshing Discussion

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently AskedEnergyIssuesEnergy SolarRadioactive LiquidSavings for Specific MeasuresUranium

  1. Uranium Processing Facility Team Signs Partnering Agreement | Y-12 National

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

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  2. The Uranium Processing Facility (UPF) Finite Element Meshing Discussion |

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium Transferon theTed Donat AboutThe QuadrennialDepartment

  3. Uranium Downblending and Disposition Project Technology Readiness Assessment

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium TransferonUS-India EnergyUnlocking CustomerOutreachReport onEducation

  4. 1st Quarter 2015 Domestic Uranium Production Report

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghurajiConventionalMississippi" ,"Plant","Primary1. Total production of uranium

  5. 1st Quarter 2015 Domestic Uranium Production Report

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghurajiConventionalMississippi" ,"Plant","Primary1. Total production of uranium2.

  6. 1st Quarter 2015 Domestic Uranium Production Report

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghurajiConventionalMississippi" ,"Plant","Primary1. Total production of uranium2.3.

  7. 1st Quarter 2015 Domestic Uranium Production Report

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghurajiConventionalMississippi" ,"Plant","Primary1. Total production of uranium2.3.4.

  8. Assessing the environmental availability of uranium in soils and sediments

    SciTech Connect (OSTI)

    Amonette, J.E.; Holdren, G.R. Jr.; Krupa, K.M.; Lindenmeier, C.W. [Pacific Northwest Lab., Richland, WA (United States)

    1994-06-01T23:59:59.000Z

    Soils and sediments contaminated with uranium pose certain environmental and ecological risks. At low to moderate levels of contamination, the magnitude of these risks depends not only on the absolute concentrations of uranium in the material but also on the availability of the uranium to drinking water supplies, plants, or higher organisms. Rational approaches for regulating the clean-up of sites contaminated with uranium, therefore, should consider the value of assessing the environmental availability of uranium at the site before making decisions regarding remediation. The purpose of this work is to review existing approaches and procedures to determine their potential applicability for assessing the environmental availability of uranium in bulk soils or sediments. In addition to making the recommendations regarding methodology, the authors have tabulated data from the literature on the aqueous complexes of uranium and major uranium minerals, examined the possibility of predicting environmental availability of uranium based on thermodynamic solubility data, and compiled a representative list of analytical laboratories capable of performing environmental analyses of uranium in soils and sediments.

  9. New Prototype Safeguards Technology Offers Improved Confidence and Automation for Uranium Enrichment Facilities

    SciTech Connect (OSTI)

    Brim, Cornelia P.

    2013-03-04T23:59:59.000Z

    An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

  10. 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

    2013-02-01T23:59:59.000Z

    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.

  11. New Prototype Safeguards Technology Offers Improved Confidence and Automation for Uranium Enrichment Facilities

    SciTech Connect (OSTI)

    Brim, Cornelia P.

    2013-04-01T23:59:59.000Z

    An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

  12. Recovery of uranium from seawater by immobilized tannin

    SciTech Connect (OSTI)

    Sakaguchi, T.; Nakajima, A.

    1987-06-01T23:59:59.000Z

    Tannin compounds having multiple adjacent hydroxy groups have an extremely high affinity for uranium. To prevent the leaching of tannins into water and to improve the adsorbing characteristics of these compounds, the authors tried to immobilize tannins. The immobilized tannin has the most favorable features for uranium recovery; high selective adsorption ability to uranium, rapid adsorption rate, and applicability in both column and batch systems. The immobilized tannin can recover uranium from natural seawater with high efficiency. About 2530 ..mu..g uranium is adsorbed per gram of this adsorbent within 22 h. Depending on the concentration in seawater, an enrichment of up to 766,000-fold within the adsorbent is possible. Almost all uranium adsorbed is easily desorbed with a very dilute acid. Thus, the immobilized tannin can be used repeatedly in the adsorption-desorption process.

  13. In-line assay monitor for uranium hexafluoride

    DOE Patents [OSTI]

    Wallace, S.A.

    1980-03-21T23:59:59.000Z

    An in-line assay monitor for determining the content of uranium-235 in a uranium hexafluoride gas isotopic separation system is provided which removes the necessity of complete access to the operating parameters of the system for determining the uranium-235 content. The method and monitor for carrying out the method involve cooling of a radiation pervious chamber connected in fluid communication with the selected point in the system to withdraw a specimen and solidify the specimen in the chamber. The specimen is irradiated by means of an ionizing radiation source of energy different from that of the 185 keV gamma emissions from uranium-235. The uranium-235 content of the specimen is determined from comparison of the accumulated 185 keV energy counts and reference energy counts. The latter is used to measure the total uranium isotopic content of the specimen.

  14. 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.

    2000-06-01T23:59:59.000Z

    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.

  15. 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.

    2012-01-23T23:59:59.000Z

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

  16. Uranium enrichment export control guide: Gaseous diffusion

    SciTech Connect (OSTI)

    Not Available

    1989-09-01T23:59:59.000Z

    This document was prepared to serve as a guide for export control officials in their interpretation, understanding, and implementation of export laws that relate to the Zangger International Trigger List for gaseous diffusion uranium enrichment process components, equipment, and materials. Particular emphasis is focused on items that are especially designed or prepared since export controls are required for these by States that are party to the International Nuclear Nonproliferation Treaty.

  17. Uranium recovery from seawater by adsorption

    SciTech Connect (OSTI)

    Koske, P.H.; Ohlrogge, K.; Peinemann, K.V.

    1988-10-01T23:59:59.000Z

    Results are presented of a 10 weeks field experiment producing uranium from seawater by the so-called adsorber-loop-concept. For the adsorption process polyamidoxin (PAO) granulate has been used with grain sizes between 0.3 - 1.2 mm diameter. The performance of the adsorber and the efficiency of the adsorption process - especially with regard to high volume flows of seawater - are presented.

  18. The ultimate disposition of depleted uranium

    SciTech Connect (OSTI)

    Not Available

    1990-12-01T23:59:59.000Z

    Significant amounts of the depleted uranium (DU) created by past uranium enrichment activities have been sold, disposed of commercially, or utilized by defense programs. In recent years, however, the demand for DU has become quite small compared to quantities available, and within the US Department of Energy (DOE) there is concern for any risks and/or cost liabilities that might be associated with the ever-growing inventory of this material. As a result, Martin Marietta Energy Systems, Inc. (Energy Systems), was asked to review options and to develop a comprehensive plan for inventory management and the ultimate disposition of DU accumulated at the gaseous diffusion plants (GDPs). An Energy Systems task team, under the chairmanship of T. R. Lemons, was formed in late 1989 to provide advice and guidance for this task. This report reviews options and recommends actions and objectives in the management of working inventories of partially depleted feed (PDF) materials and for the ultimate disposition of fully depleted uranium (FDU). Actions that should be considered are as follows. (1) Inspect UF{sub 6} cylinders on a semiannual basis. (2) Upgrade cylinder maintenance and storage yards. (3) Convert FDU to U{sub 3}O{sub 8} for long-term storage or disposal. This will include provisions for partial recovery of costs to offset those associated with DU inventory management and the ultimate disposal of FDU. Another recommendation is to drop the term tails'' in favor of depleted uranium'' or DU'' because the tails'' label implies that it is waste.'' 13 refs.

  19. Phosphate Barriers for Immobilization of Uranium Plumes

    SciTech Connect (OSTI)

    Burns, Peter C.

    2005-06-01T23:59:59.000Z

    Uranium contamination of the subsurface has remained a persistent problem plaguing remedial design at sites across the U.S. that were involved with production, handling, storage, milling, and reprocessing of fissile uranium for both civilian and defense related purposes. Remediation efforts to date have relied upon excavation, pump-and-treat, or passive remediation barriers (PRB?s) to remove or attenuate uranium mobility. Documented cases convincingly demonstrate that excavation and pump-and-treat methods are ineffective for a number of highly contaminated sites. There is growing concern that use of conventional PRB?s, such as zero-valent iron, are a temporary solution to a problem that will persist for thousands of years. Alternatives to the standard treatment methods are therefore warranted. The core objective of our research is to demonstrate that a phosphorous amendment strategy will result in a reduction of dissolved uranium to below the proposed drinking water standard. Our hypothesis is that long-chain polyphosphate compounds forestall precipitation of sparingly soluble uranyl phosphate compounds, which is key to preventing fouling of wells at the point of injection. Our other fundamental objective is to synthesize and correctly characterize the uranyl phosphate phases that form in the geochemical conditions under consideration. This report summarizes work conducted at the University of Notre Dame through November of 2003 under DOE grant DE-FG07-02ER63489, which has been funded since September, 2002. The objectives at Notre Dame are development of synthesis techniques for uranyl phosphate phases, together with detailed structural and chemical characterization of the myriad of uranyl phosphate phases that may form under geochemical conditions under consideration.

  20. Uranio impoverito: perché? (Depleted uranium: why?)

    E-Print Network [OSTI]

    Germano D'Abramo

    2003-06-05T23:59:59.000Z

    In this paper we develop a simple model of the penetration process of a long rod through an uniform target. Applying the momentum and energy conservation laws, we derive an analytical relation which shows how the penetration depth depends upon the density of the rod, given a fixed kinetic energy. This work was sparked off by the necessity of understanding the effectiveness of high density penetrators (e.g. depleted uranium penetrators) as anti-tank weapons.

  1. Plutonium Uranium Extraction Facility Documented Safety Analysis

    SciTech Connect (OSTI)

    DODD, E.N.

    2003-10-08T23:59:59.000Z

    This document provides the documented safety analysis (DSA) and Central Plateau Remediation Project (CP) requirements that apply to surveillance and maintenance (S&M) activities at the Plutonium-Uranium Extraction (PUREX) facility. This DSA was developed in accordance with DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities''. Upon approval and implementation of this document, the current safety basis documents will be retired.

  2. Measurement of uranium enrichment by gamma spectroscopy: result of an experimental design

    E-Print Network [OSTI]

    , Gamma Spectrometry, uranium enrichment #12;PAPER Measurement of uranium enrichment by gamma spectroscopy: result of an experimental design Gamma spectroscopy is commonly used in nuclear safeguards to measure uranium enrichment. An experimental

  3. Magnetic Exchange Coupling and Single-Molecule Magnetism in Uranium Complexes

    E-Print Network [OSTI]

    Rinehart, Jeffrey Dennis

    2010-01-01T23:59:59.000Z

    in molecular uranium cluster chemistry. 13 Compound 2 ischemistry and small-molecule reactivity of uranium. AmongUranium Complexes by Jeffrey Dennis Rinehart Doctor of Philosophy in Chemistry

  4. Recent International R&D Activities in the Extraction of Uranium from Seawater

    E-Print Network [OSTI]

    Rao, Linfeng

    2011-01-01T23:59:59.000Z

    Uranium and Rare Earth Elements Using Biomass of Algae, Bioinorganic Chemistry andRecovery of uranium from sea water. Chemistry & Industry (of uranium from seawater. Turkish Journal of Chemistry, 17 (

  5. The geochemistry of uranium in the Orca Basin

    E-Print Network [OSTI]

    Weber, Frederick Fewell

    1979-01-01T23:59:59.000Z

    . , 1974). Substantial uranium enrichments have been reported by many investigators for samples taken from aroxic environments (Strom, 1948; Starik et al. , '1961; Swanson, 1961; Sackett and Cook, 1969; Kolodny and Kaplan, 1969; Bertine et al. , 1970...) ~ Degens et al. , (1977) report concentrations of uranium as high as 60ppm, more than an order of magnitude enrich- meut, for Black Sea sediments. If marine reducing environments are found with uranium concentrations apuroaching 100ppm, they will begin...

  6. The geochemistry of uranium in the Orca Basin 

    E-Print Network [OSTI]

    Weber, Frederick Fewell

    1979-01-01T23:59:59.000Z

    as uranium concentrations dzop to an average of 2. dppm, indicative of relatively low uranium bearing pelagic particle deposition. Furthermore, the 13 C values become heavier in this region, lacking a large terrest. rial component. This evidence suggests... 39 the basin walls. It may be possible that these particles at the brine/seawater interface incorporate any uranium reduced in this zone and carry it to the brine perimeter where it is deposited. Unfortunately, attempts to accurately core...

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't Your Destiny:Revised Finding of No53197E T ADRAFT ENVIRONMENTALCombustion SystemLow-Enriched

  8. National Uranium Resource Evaluation: Salina Quadrangle, Utah

    SciTech Connect (OSTI)

    Lupe, R.D.; Campbell, J.A.; Franczyk, K.J.; Luft, S.J.; Peterson, F.; Robinson, K.

    1982-09-01T23:59:59.000Z

    Two stratigraphic units, the Late Jurassic Salt Wash Member of the Morrison Formation and the Triassic Chinle Formation, were determined to be favorable for the occurrence of uranium deposits that meet the minimum size and grade requirements of the US Department of Energy in the Salina 1 x 2/sup 0/ Quadrangle, Utah. Three areas judged favorable for the Salt Wash Member are the Tidwell and Notom districts, and the Henry Mountains mineral belt. The criteria used to establish favorability were the presence of: (1) fluvial sandstone beds deposited by low-energy streams; (2) actively moving major and minor structures such as the Paradox basin and the many folds within it; (3) paleostream transport directions approximately perpendicular to the trend of many of the paleofolds; (4) presence of favorable gray lacustrine mudstone beds; and (5) known uranium occurrences associated with the favorable gray mudstones. Four favorable areas have been outlined for the Chinle Formation. These are the San Rafael Swell, Inter River, and the Orange Cliffs subareas and the Capitol Reef area. The criteria used to establish these areas are: the sandstone-to-mudstone ratios and the geographic distribution of the Petrified Forest Member of the Chinle Formation which is considered as the probable source for the uranium.

  9. Chapter 3. Volume and Characteristics of Uranium Mine Wastes Uranium has been found and mined in a wide variety of rocks, including sandstone, carbonates1

    E-Print Network [OSTI]

    3-1 Chapter 3. Volume and Characteristics of Uranium Mine Wastes Uranium has been found and mined conventional mining, solution extraction, and milling of uranium, a principal focus of this report is TENORM, or which may need future reclamation. When uranium mining first started, most of the ores were recovered

  10. Control of structure and reactivity by ligand design : applications to small molecule activation by low-valent uranium complexes

    E-Print Network [OSTI]

    Lam, Oanh Phi

    2010-01-01T23:59:59.000Z

    coordination chemistry is depleted uranium, a by-product innuclear reactors. Depleted uranium Figure 1-1. The periodic

  11. Uranium immobilization by sulfate-reducing biofilms grown on...

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

    of uranium-complexing carbonates. The biofilms were grown in three identically operated fixed bed reactors, filled with three types of minerals: one noncarbonate-bearing...

  12. Assessment of Controlling Processes for Field-Scale Uranium Reactive...

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

    at the 300A site. However, the model simulations also revealed that the groundwater chemistry was relatively stable during the uranium tracer experiment and therefore...

  13. Secretarial Determination of No Adverse Material Impact for Uranium...

    Energy Savers [EERE]

    set forth in the 2012 Secretarial Determination and the Department's Excess Uranium Inventory Management Plan released in July 2013. Secretarial Determination 5-15-14.pdf More...

  14. Assessment of Controlling Processes for Field-Scale Uranium Reactive...

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

    reactive transport model was employed to assess the key factors and processes that control the field-scale uranium reactive transport. Taking into consideration of relevant...

  15. Uranium-contaminated soils: Ultramicrotomy and electron beam analysis

    SciTech Connect (OSTI)

    Buck, E.C.; Dietz, N.L.; Bates, J.K.; Cunnane, J.C.

    1994-04-01T23:59:59.000Z

    Uranium contaminated soils from the Fernald Operation Site, Ohio, have been examined by a combination of optical microscopy, scanning electron microscopy with backscattered electron detection (SEM/BSE), and analytical electron microscopy (AEM). A method is described for preparing of transmission electron microscopy (TEM) thin sections by ultramicrotomy. By using these thin sections, SEM and TEM images can be compared directly. Uranium was found in iron oxides, silicates (soddyite), phosphates (autunites), and fluorite. Little uranium was associated with clays. The distribution of uranium phases was found to be inhomogeneous at the microscopic level.

  16. High grade uranium resources in the United States : an overview

    E-Print Network [OSTI]

    Graves, Richard E.

    1974-01-01T23:59:59.000Z

    A time analysis of uranium exploration, production and known reserves in the United States is employed to reveal industry trends. The

  17. The radioactive Substances (Prepared Uranium Thorium Compounds) Exemption Order 1962 

    E-Print Network [OSTI]

    Joseph, Keith

    1962-01-01T23:59:59.000Z

    STATUTORY INSTRUMENTS 1962 No. 2711 ATOMIC ENERGY AND RADIOACI1VE SUBSTANCES The Radioactive Substances (prepared Uranium and Thorium Compounds) Exemption Order 1962...

  18. EIS-0472: Uranium Leasing Program, Mesa, Montrose, and San Miguel...

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

    Leasing Program, under which DOE administers tracts of land in western Colorado for exploration, development, and the extraction of uranium and vanadium ores. The cooperating...

  19. Financial Assurance for In Situ Uranium Facilities (Texas)

    Broader source: Energy.gov [DOE]

    Owners or operators are required to provide financial assurance for in situ uranium sites. This money is required for: decommissioning, decontamination, demolition, and waste disposal for buildings...

  20. President Truman Increases Production of Uranium and Plutonium...

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

    Increases Production of Uranium and Plutonium | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing...