National Library of Energy BETA

Sample records for uranium dioxide uo

  1. Uranium dioxide electrolysis

    DOE Patents [OSTI]

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

    2009-12-29

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

  2. Los Alamos probes mysteries of uranium dioxide's thermal conductivity

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

    Mysteries of uranium dioxide's thermal conductivity Los Alamos probes mysteries of uranium dioxide's thermal conductivity New research is showing that the thermal conductivity of cubic uranium dioxide is strongly affected by interactions between phonons carrying heat and magnetic spins. August 4, 2014 Illustration of anisotropic thermal conductivity in uranium dioxide (UO2). Scientists are studying the thermal conductivity related to the material's different crystallographic directions, hoping

  3. Molten uranium dioxide structure and dynamics

    SciTech Connect (OSTI)

    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.; Leibowitz, L.

    2014-11-21

    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

    2014-11-21

    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. METHOD OF SINTERING URANIUM DIOXIDE

    DOE Patents [OSTI]

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

    1963-04-30

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

  6. Thermodynamic properties of uranium dioxide

    SciTech Connect (OSTI)

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

    1981-04-01

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

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

    SciTech Connect (OSTI)

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

    2008-07-15

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

  8. Molecular Dynamics Simulation of Thermodynamic Properties in Uranium Dioxide

    SciTech Connect (OSTI)

    Wang, Xiangyu; Wu, Bin; Gao, Fei; Li, Xin; Sun, Xin; Khaleel, Mohammad A.; Akinlalu, Ademola V.; Liu, L.

    2014-03-01

    In the present study, we investigated the thermodynamic properties of uranium dioxide (UO2) by molecular dynamics (MD) simulations. As for solid UO2, the lattice parameter, density, and enthalpy obtained by MD simulations were in good agreement with existing experimental data and previous theoretical predictions. The calculated thermal conductivities matched the experiment results at the midtemperature range but were underestimated at very low and very high temperatures. The calculation results of mean square displacement represented the stability of uranium at all temperatures and the high mobility of oxygen toward 3000 K. By fitting the diffusivity constant of oxygen with the Vogel-Fulcher-Tamman law, we noticed a secondary phase transition near 2006.4 K, which can be identified as a ‘‘strong’’ to ‘‘fragile’’ supercooled liquid or glass phase transition in UO2. By fitting the oxygen diffusion constant with the Arrhenius equation, activation energies of 2.0 and 2.7 eV that we obtained were fairly close to the recommended values of 2.3 to 2.6 eV. Xiangyu Wang, Bin Wu, Fei Gao, Xin Li, Xin Sun, Mohammed A. Khaleel, Ademola V. Akinlalu and Li Liu

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

    SciTech Connect (OSTI)

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

    2012-07-31

    Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc -CO2) system was designed to mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc -CO2 and are capable of extracting uranium directly from UO2, UO3 and U3O8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO3 loaded Sc-CO2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc -CO2 pressure and temperature, TBP/HNO3 complex used, reductant or complexant used for selectivity, and ionic liquids used as supportive media. To complete the extraction and recovery cycle, we then demonstrate uranium back extraction from the TBP loaded sc-CO2 phase into an aqueous phase and the characterization of the uranium complex formed at the end of this process. Another aspect of this project was to limit proliferation risks by either co-extracting uranium and plutonium, or by leaving plutonium behind by selectively extracting uranium. We report that the former is easily achieved, since plutonium is in the tetravalent or hexavalent oxidation state in the oxidizing environment created by the TBP-nitric acid complex, and is therefore co-extracted. The latter is more challenging, as a reductant or complexant to plutonium has to be used to selectively extract uranium. After undertaking experiments on different reducing or complexing systems (e.g., AcetoHydroxamic Acid (AHA), Fe(II), ascorbic acid), oxalic acid was chosen as it can complex tetravalent actinides (Pu, Np, Th) in the aqueous phase while allowing the extraction of hexavalent uranium in the sc-CO2 phase. Finally, we show results using an alternative media to commonly used aqueous phases: ionic liquids. We show the dissolution of uranium in ionic liquids and its extraction using sc-CO2 with and without the presence of AHA. The possible separation of trivalent actinides from uranium is also demonstrated in ionic liquids using neodymium as a surrogate and diglycolamides as the extractant.

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

    SciTech Connect (OSTI)

    Collins, Robert T; Collins, Jack Lee; Hunt, Rodney Dale; Ladd-Lively, Jennifer L; Patton, Kaara K; Hickman, Robert

    2014-01-01

    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.

  11. Thermal Reactions of Uranium Metal, UO2, U3O8, UF4, and UO2F2 with NF3 to Produce UF6

    SciTech Connect (OSTI)

    McNamara, Bruce K.; Scheele, Randall D.; Kozelisky, Anne E.; Edwards, Matthew K.

    2009-11-01

    he objective of this paper is to demonstrate that NF3 fluorinates uranium metal, UO2, UF4, UO3, U3O8, and UO2F2•2H2O to produce the volatile UF6 at temperatures between 100 and 500?C. Thermogravimetric reaction profiles are described that reflect changes in the uranium oxidation state and discrete chemical speciation. Differences in the onset temperatures for each system indicate that NF3-substrate interactions are important for the temperature at which NF3 reacts: U metal > UO3 > UO2 > UO2F2 > UF4 and in fact may indicate different fluorination mechanisms for these various substrates. These studies demonstrate that NF3 is a potential replacement fluorinating agent in the existing nuclear fuel cycle and in oft-proposed actinide volatility reprocessing.

  12. Spin-lattice coupling in uranium dioxide probed by magnetostriction measurements at high magnetic fields (P08358-E001-PF)

    SciTech Connect (OSTI)

    Gofryk, K.; Jaime, M.

    2014-12-01

    Our preliminary magnetostriction measurements have already shown a strong interplay of lattice dynamic and magnetism in both antiferromagnetic and paramagnetic states, and give unambiguous evidence of strong spin- phonon coupling in uranium dioxide. Further studies are planned to address the puzzling behavior of UO2 in magnetic and paramagnetic states and details of the spin-phonon coupling.

  13. Incidence of High Nitrogen in Sintered Uranium Dioxide: A Case Study

    SciTech Connect (OSTI)

    Balakrishna, Palanki; Murty, B. Narasimha; Anuradha, M.; Yadav, R.B.; Jayaraj, R.N

    2005-05-15

    Nitrogen content, above the specified limit of 75 {mu}g(gU){sup -1}, was encountered in sintered uranium dioxide in the course of its manufacture. The cause was traced to the sintering process, wherein carbon, a degradation product of the die wall or admixed lubricant, was retained in the compact as a result of inadvertent reversal of gas flow in the sintering furnace. In the presence of carbon, the uranium dioxide reacted with nitrogen from the furnace atmosphere to form nitride. The compacts with high nitrogen were also those with low sintered density, arising from low green density. The low green density was due to filling problems of an inhomogeneous powder. The experiments carried out establish the causes of high nitrogen to be the carbon residue from lubricant when the UO{sub 2} is sintered in a cracked ammonia atmosphere.

  14. Oxygen transport in off-stoichiometric uranium dioxide mediated by defect clustering dynamics

    SciTech Connect (OSTI)

    Yu, Jianguo Bai, Xian-Ming; El-Azab, Anter; Allen, Todd R.

    2015-03-07

    Oxygen transport is central to many properties of oxides such as stoichiometric changes, phase transformation, and ionic conductivity. In this paper, we report a mechanism for oxygen transport in uranium dioxide (UO{sub 2}) in which the kinetics is mediated by defect clustering dynamics. In particular, the kinetic Monte Carlo method has been used to investigate the kinetics of oxygen transport in UO{sub 2} under the condition of creation and annihilation of oxygen vacancies and interstitials as well as oxygen interstitial clustering, with variable off-stoichiometry and temperature conditions. It is found that in hypo-stoichiometric UO{sub 2?x}, oxygen transport is well described by the vacancy diffusion mechanism while in hyper-stoichiometric UO{sub 2+x}, oxygen interstitial cluster diffusion contributes significantly to oxygen transport kinetics, particularly at high temperatures and high off-stoichiometry levels. It is also found that di-interstitial clusters and single interstitials play dominant roles in oxygen diffusion while other larger clusters have negligible contributions. However, the formation, coalescence, and dissociation of these larger clusters indirectly affects the overall oxygen diffusion due to their interactions with mono and di-interstitials, thus providing an explanation of the experimental observation of saturation or even drop of oxygen diffusivity at high off-stoichiometry.

  15. Role of uranium(VI) in the ThO/sub 2/-UO/sub 3/ sol-gel process

    SciTech Connect (OSTI)

    Tewari, P.H.; Campbell, A.B.

    1980-11-01

    Increases in pH and temperature of U(VI) solutions enhance adsorption of uranium on ThO/sub 2/ through hydrolysis of U(VI) as evidenced by absorption spectra changes of the solution. Sols of ThO/sub 2/-UO/sub 3/ are formed by adsorption of uranium on ThO/sub 2/. At low pH's (approx. pH 3.0), the sols behave as Newtonian fluids but at higher pH's the sols (especially the concentrated ones) transform into thixotropic gels. The increased adsorption of uranium by ThO/sub 2/ and the increased viscosity of the ThO/sub 2/-UO/sub 3/ sols with pH are related. Increased adsorption of uranium produces rod-shaped UO/sub 3/.2H/sub 2/O on the ThO/sub 2/ surface. These UO/sub 3/ nuclei link ThO/sub 2/ particles to form long rodlike particles. With further increased adsorption of uranium at higher pH's (less than or equal to 3.7), the particles crosslink to produce a structured network giving a thixotropic gel. Adsorption, electron microscopic, electrophoetic mobility, X-ray diffraction, and X-ray photoelectron spectroscopic data are presented to explain the role of U(VI) in the sol-gel process. 6 figures, 1 table.

  16. Uranium vacancy mobility at the ?5 symmetric tilt and ?5 twist grain boundaries in UO?

    SciTech Connect (OSTI)

    Uberuaga, Blas Pedro; Andersson, David A.

    2015-10-01

    Ionic transport at grain boundaries in oxides dictates a number of important phenomena, from ionic conductivity to sintering to creep. For nuclear fuels, it also influences fission gas bubble nucleation and growth. Here, using a combination of atomistic calculations and object kinetic Monte Carlo (okMC) simulations, we examine the kinetic pathways associated with uranium vacancies at two model grain boundaries in UO2. The barriers for vacancy motion were calculated using the nudged elastic band method at all uranium sites at each grain boundary and were used as the basis of the okMC simulations. For both boundaries considered – a simple tilt and a simple twist boundary – the mobility of uranium vacancies is significantly higher than in the bulk. For the tilt boundary, there is clearly preferred migration along the tilt axis as opposed to in the perpendicular direction while, for the twist boundary, migration is essentially isotropic within the boundary plane. These results show that cation defect mobility in fluorite-structured materials is enhanced at certain types of grain boundaries and is dependent on the boundary structure with the tilt boundary exhibiting higher rates of migration than the twist boundary.

  17. Migration of defect clusters and xenon-vacancy clusters in uranium dioxide

    SciTech Connect (OSTI)

    Chen, Dong; Gao, Fei; Deng, Huiqiu; Hu, Wangyu; Sun, Xin

    2014-07-01

    The possible transition states, minimum energy paths and migration mechanisms of defect clusters and xenon-vacancy defect clusters in uranium dioxide have been investigated using the dimer and the nudged elastic-band methods. The nearby O atom can easily hop into the oxygen vacancy position by overcoming a small energy barrier, which is much lower than that for the migration of a uranium vacancy. A simulation for a vacancy cluster consisting of two oxygen vacancies reveals that the energy barrier of the divacancy migration tends to decrease with increasing the separation distance of divacancy. For an oxygen interstitial, the migration barrier for the hopping mechanism is almost three times larger than that for the exchange mechanism. Xe moving between two interstitial sites is unlikely a dominant migration mechanism considering the higher energy barrier. A net migration process of a Xe-vacancy pair containing an oxygen vacancy and a xenon interstitial is identified by the NEB method. We expect the oxygen vacancy-assisted migration mechanism to possibly lead to a long distance migration of the Xe interstitials in UO2. The migration of defect clusters involving Xe substitution indicates that Xe atom migrating away from the uranium vacancy site is difficult.

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

    DOE Patents [OSTI]

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

    1995-01-01

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

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

    DOE Patents [OSTI]

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

    1995-06-06

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

  20. Uranium vacancy mobility at the sigma 5 symmetric tilt grain boundary in UO2

    SciTech Connect (OSTI)

    Uberuaga, Blas P.

    2012-05-02

    An important consequence of the fissioning process occurring during burnup is the formation of fission products. These fission products alter the thermo-mechanical properties of the fuel. They also lead to macroscopic changes in the fuel structure, including the formation of bubbles that are connected to swelling of the fuel. Subsequent release of fission gases increase the pressure in the plenum and can cause changes in the properties of the fuel pin itself. It is thus imperative to understand how fission products, and fission gases in particular, behave within the fuel in order to predict the performance of the fuel under operating conditions. Fission gas redistribution within the fuel is governed by mass transport and the presence of sinks such as impurities, dislocations, and grain boundaries. Thus, to understand how the distribution of fission gases evolves in the fuel, we must understand the underlying transport mechanisms, tied to the concentrations and mobilities of defects within the material, and how these gases interact with microstructural features that might act as sinks. Both of these issues have been addressed in previous work under NEAMS. However, once a fission product has reached a sink, such as a grain boundary, its mobility may be different there than in the grain interior and predicting how, for example, bubbles nucleate within grain boundaries necessitates an understanding of how fission gases diffuse within boundaries. That is the goal of the present work. In this report, we describe atomic level simulations of uranium vacancy diffusion in the pressence of a {Sigma}5 symmetric tilt boundary in urania (UO{sub 2}). This boundary was chosen as it is the simplest of the boundaries we considered in previous work on segregation and serves as a starting point for understanding defect mobility at boundaries. We use a combination of molecular statics calculations and kinetic Monte Carlo (kMC) to determine how the mobility of uranium vacancies is altered at this particular grain boundary. Given that the diffusion of fission gases such as Xe are tied to the mobility of uranium vacancies, these results given insight into how fission gas mobility differs at grain boundaries compared to bulk urania.

  1. Impact of homogeneous strain on uranium vacancy diffusion in uranium dioxide

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

    Goyal, Anuj; Phillpot, Simon R.; Subramanian, Gopinath; Andersson, David A.; Stanek, Chris R.; Uberuaga, Blas P.

    2015-03-03

    We present a detailed mechanism of, and the effect of homogeneous strains on, the migration of uranium vacancies in UO2. Vacancy migration pathways and barriers are identified using density functional theory and the effect of uniform strain fields are accounted for using the dipole tensor approach. We report complex migration pathways and noncubic symmetry associated with the uranium vacancy in UO2 and show that these complexities need to be carefully accounted for to predict the correct diffusion behavior of uranium vacancies. We show that under homogeneous strain fields, only the dipole tensor of the saddle with respect to the minimummore » is required to correctly predict the change in the energy barrier between the strained and the unstrained case. Diffusivities are computed using kinetic Monte Carlo simulations for both neutral and fully charged state of uranium single and divacancies. We calculate the effect of strain on migration barriers in the temperature range 800–1800 K for both vacancy types. Homogeneous strains as small as 2% have a considerable effect on diffusivity of both single and divacancies of uranium, with the effect of strain being more pronounced for single vacancies than divacancies. In contrast, the response of a given defect to strain is less sensitive to changes in the charge state of the defect. Further, strain leads to anisotropies in the mobility of the vacancy and the degree of anisotropy is very sensitive to the nature of the applied strain field for strain of equal magnitude. Our results indicate that the influence of strain on vacancy diffusivity will be significantly greater when single vacancies dominate the defect structure, such as sintering, while the effects will be much less substantial under irradiation conditions where divacancies dominate.« less

  2. Impact of homogeneous strain on uranium vacancy diffusion in uranium dioxide

    SciTech Connect (OSTI)

    Goyal, Anuj; Phillpot, Simon R.; Subramanian, Gopinath; Andersson, David A.; Stanek, Chris R.; Uberuaga, Blas P.

    2015-03-03

    We present a detailed mechanism of, and the effect of homogeneous strains on, the migration of uranium vacancies in UO2. Vacancy migration pathways and barriers are identified using density functional theory and the effect of uniform strain fields are accounted for using the dipole tensor approach. We report complex migration pathways and noncubic symmetry associated with the uranium vacancy in UO2 and show that these complexities need to be carefully accounted for to predict the correct diffusion behavior of uranium vacancies. We show that under homogeneous strain fields, only the dipole tensor of the saddle with respect to the minimum is required to correctly predict the change in the energy barrier between the strained and the unstrained case. Diffusivities are computed using kinetic Monte Carlo simulations for both neutral and fully charged state of uranium single and divacancies. We calculate the effect of strain on migration barriers in the temperature range 800–1800 K for both vacancy types. Homogeneous strains as small as 2% have a considerable effect on diffusivity of both single and divacancies of uranium, with the effect of strain being more pronounced for single vacancies than divacancies. In contrast, the response of a given defect to strain is less sensitive to changes in the charge state of the defect. Further, strain leads to anisotropies in the mobility of the vacancy and the degree of anisotropy is very sensitive to the nature of the applied strain field for strain of equal magnitude. Our results indicate that the influence of strain on vacancy diffusivity will be significantly greater when single vacancies dominate the defect structure, such as sintering, while the effects will be much less substantial under irradiation conditions where divacancies dominate.

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

    DOE Patents [OSTI]

    Stinton, David P. (Knoxville, TN)

    1983-01-01

    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.

  4. Simulation of xenon, uranium vacancy and interstitial diffusion and grain boundary segregation in UO2

    SciTech Connect (OSTI)

    Andersson, Anders D.; Tonks, Michael R.; Casillas, Luis; Nerikar, Pankaj; Vyas, Shyam; Uberuaga, Blas P.; Stanek, Christopher R.

    2014-10-31

    In light water reactor fuel, gaseous fission products segregate to grain boundaries, resulting in the nucleation and growth of large intergranular fission gas bubbles. Based on the mechanisms established from density functional theory (DFT) and empirical potential calculations 1, continuum models for diffusion of xenon (Xe), uranium (U) vacancies and U interstitials in UO2 have been derived for both intrinsic conditions and under irradiation. Segregation of Xe to grain boundaries is described by combining the bulk diffusion model with a model for the interaction between Xe atoms and three different grain boundaries in UO2 ( ?5 tilt, ?5 twist and a high angle random boundary),as derived from atomistic calculations. All models are implemented in the MARMOT phase field code, which is used to calculate effective Xe and U diffusivities as well as redistribution for a few simple microstructures.

  5. Influence of instrument conditions on the evaporation behavior of uranium dioxide with UV laser-assisted atom probe tomography

    SciTech Connect (OSTI)

    Valderrama, B.; Henderson, H.B.; Gan, J.; Manuel, M.V.

    2015-04-01

    Atom probe tomography (APT) provides the ability to detect subnanometer chemical variations spatially, with high accuracy. However, it is known that compositional accuracy can be affected by experimental conditions. A study of the effect of laser energy, specimen base temperature, and detection rate is performed on the evaporation behavior of uranium dioxide (UO2). In laser-assisted mode, tip geometry and standing voltage also contribute to the evaporation behavior. In this investigation, it was determined that modifying the detection rate and temperature did not affect the evaporation behavior as significantly as laser energy. It was also determined that three laser evaporation regimes are present in UO2. Very low laser energy produces a behavior similar to DC-field evaporation, moderate laser energy produces the desired laser-assisted field evaporation characteristic and high laser energy induces thermal effects, negatively altering the evaporation behavior. The need for UO2 to be analyzed under moderate laser energies to produce accurate stoichiometry distinguishes it from other oxides. The following experimental conditions providing the best combination of mass resolving power, accurate stoichiometry, and uniform evaporation behavior: 50 K, 10 pJ laser energy, a detection rate of 0.003 atoms per pulse, and a 100 kHz repetition rate.

  6. Oxidation and crystal field effects in uranium

    SciTech Connect (OSTI)

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

    2015-07-06

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

  7. Thermophysical properties of uranium dioxide - Version 0 for peer review

    SciTech Connect (OSTI)

    Fink, J.K.; Petri, M.C.

    1997-02-01

    Data on thermophysical properties of solid and liquid UO{sub 2} have been reviewed and critically assessed to obtain consistent thermophysical property recommendations for inclusion in the International Nuclear Safety Center Database on the World Wide Web (http://www.insc.anl.gov.). Thermodynamic properties that have been assessed are enthalpy, heat capacity, melting point, enthalpy of fusion, thermal expansion, density, surface tension, and vapor pressure. Transport properties that have been assessed are thermal conductivity, thermal diffusivity, viscosity, and emissivity. Summaries of the recommendations with uncertainties and detailed assessments for each property are included in this report and in the International Nuclear Safety Center Database for peer review. The assessments includes a review of the experiments and data, an examination of previous recommendations, the basis for selecting recommendations, a determination of uncertainties, and a comparison of recommendations with data and with previous recommendations. New data and research that have led to new recommendations include thermal expansion and density measurements of solid and liquid UO{sub 2}, derivation of physically-based equations for the thermal conductivity of solid UO{sub 2}, measurements of the heat capacity of liquid UO{sub 2}, and measurements and analysis of the thermal conductivity of liquid UO{sub 2}.

  8. Thermal Conductivity Measurement of Xe-Implanted Uranium Dioxide Thick Films using Multilayer Laser Flash Analysis

    SciTech Connect (OSTI)

    Nelson, Andrew T.

    2012-08-30

    The Fuel Cycle Research and Development program's Advanced Fuels campaign is currently pursuing use of ion beam assisted deposition to produce uranium dioxide thick films containing xenon in various morphologies. To date, this technique has provided materials of interest for validation of predictive fuel performance codes and to provide insight into the behavior of xenon and other fission gasses under extreme conditions. In addition to the structural data provided by such thick films, it may be possible to couple these materials with multilayer laser flash analysis in order to measure the impact of xenon on thermal transport in uranium dioxide. A number of substrate materials (single crystal silicon carbide, molybdenum, and quartz) containing uranium dioxide films ranging from one to eight microns in thickness were evaluated using multilayer laser flash analysis in order to provide recommendations on the most promising substrates and geometries for further investigation. In general, the uranium dioxide films grown to date using ion beam assisted deposition were all found too thin for accurate measurement. Of the substrates tested, molybdenum performed the best and looks to be the best candidate for further development. Results obtained within this study suggest that the technique does possess the necessary resolution for measurement of uranium dioxide thick films, provided the films are grown in excess of fifty microns. This requirement is congruent with the material needs when viewed from a fundamental standpoint, as this length scale of material is required to adequately sample grain boundaries and possible second phases present in ceramic nuclear fuel.

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

    SciTech Connect (OSTI)

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

    1991-11-01

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

  10. Local structure in solid solutions of stabilised zirconia with actinide dioxides (UO{sub 2}, NpO{sub 2})

    SciTech Connect (OSTI)

    Walter, Marcus; Somers, Joseph; Bouexiere, Daniel; Rothe, Joerg

    2011-04-15

    The local structure of (Zr,Lu,U)O{sub 2-x} and (Zr,Y,Np)O{sub 2-x} solid solutions has been investigated by extended X-ray absorption fine structure (EXAFS). Samples were prepared by mixing reactive (Zr,Lu)O{sub 2-x} and (Zr,Y)O{sub 2-x} precursor materials with the actinide oxide powders, respectively. Sintering at 1600 {sup o}C in Ar/H{sub 2} yields a fluorite structure with U(IV) and Np(IV). As typical for stabilised zirconia the metal-oxygen and metal-metal distances are characteristic for the different metal ions. The bond lengths increase with actinide concentration, whereas highest adaptation to the bulk stabilised zirconia structure was observed for U---O and Np---O bonds. The Zr---O bond shows only a slight increase from 2.14 A at 6 mol% actinide to 2.18 A at infinite dilution in UO{sub 2} and NpO{sub 2}. The short interatomic distance between Zr and the surrounding oxygen and metal atoms indicate a low relaxation of Zr with respect to the bulk structure, i.e. a strong Pauling behaviour. -- Graphical abstract: Metal-oxygen bond distances in (Zr,Lu,U)O{sub 2-x} solid solutions with different oxygen vacancy concentrations (Lu/Zr=1 and Lu/Zr=0.5). Display Omitted Research Highlights: {yields} EXAFS indicates high U and Np adaption to the bulk structure of stabilised zirconia. {yields} Zr---O bond length is 2.18 A at infinite Zr dilution in UO{sub 2} and NpO{sub 2}. {yields} Low relaxation (strong Pauling behaviour) of Zr explains its low solubility in UO{sub 2}.

  11. PRODUCTION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-08-27

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

  12. Comparison of Spectroscopic Data with Cluster Calculations of Plutonium, Plutonium Dioxide and Uranium Dioxide

    SciTech Connect (OSTI)

    Tobin, J G; Yu, S W; Chung, B W; Ryzhkov, M V; Mirmelstein, A

    2012-05-15

    Using spectroscopic data produced in the experimental investigations of bulk systems, including X-Ray Absorption Spectroscopy (XAS), Photoelectron Spectroscopy (PES) and Bremstrahlung Isochromat Spectroscopy (BIS), the theoretical results within for UO{sub 2}{sup 6}, PuO{sub 2}{sup 6} and Pu{sup 7} clusters have been evaluated. The calculations of the electronic structure of the clusters have been performed within the framework of the Relativistic Discrete-Variational Method (RDV). The comparisons between the LLNL experimental data and the Russian calculations are quite favorable. The cluster calculations may represent a new and useful avenue to address unresolved questions within the field of actinide electron structure, particularly that of Pu. Observation of the changes in the Pu electronic structure as a function of size suggests interesting implications for bulk Pu electronic structure.

  13. High Field Magnetization measurements of uranium dioxide single crystals (P08358- E003-PF)

    SciTech Connect (OSTI)

    Gofryk, K.; Harrison, N.; Jaime, M.

    2014-12-01

    Our preliminary high field magnetic measurements of UO2 are consistent with a complex nature of the magnetic ordering in this material, compatible with the previously proposed non-collinear 3-k magnetic structure. Further extensive magnetic studies on well-oriented (<100 > and <111>) UO2 crystals are planned to address the puzzling behavior of UO2 in both antiferromagnetic and paramagnetic states at high fields.

  14. Method for fluorination of uranium oxide

    DOE Patents [OSTI]

    Petit, George S. (Oak Ridge, TN)

    1987-01-01

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

  15. Magnetization measurements of uranium dioxide single crystals (P08358-E002-PF)

    SciTech Connect (OSTI)

    Gofryk, K.; Zapf, V.; Jaime, M.

    2014-12-01

    Our preliminary magnetic susceptibility measurements of UO2 point to complex nature of the magnetic ordering in this material, consistent with the proposed non-collinear 3-k magnetic structure. Further extensive magnetic studies are planned to address the puzzling behavior of UO2 in both antiferromagnetic and paramagnetic states.

  16. Atomistic study of porosity impact on phonon driven thermal conductivity: Application to uranium dioxide

    SciTech Connect (OSTI)

    Colbert, Mehdi; Ribeiro, Fabienne; Tréglia, Guy

    2014-01-21

    We present here an analytical method, based on the kinetic theory, to determine the impact of defects such as cavities on the thermal conductivity of a solid. This approach, which explicitly takes into account the effects of internal pore surfaces, will be referred to as the Phonon Interface THermal cONductivity (PITHON) model. Once exposed in the general case, this method is then illustrated in the case of uranium dioxide. It appears that taking properly into account these interface effects significantly modifies the temperature and porosity dependence of thermal conductivity with respect to that issued from either micromechanical models or more recent approaches, in particular, for small cavity sizes. More precisely, it is found that if the mean free path appears to have a major effect in this system in the temperature and porosity distribution range of interest, the variation of the specific heat at the surface of the cavity is predicted to be essential at very low temperature and small sizes for sufficiently large porosity.

  17. Kinetics of laser pulse vaporization of uranium dioxide by mass spectrometry

    SciTech Connect (OSTI)

    Tsai, C.

    1981-11-01

    Safety analyses of nuclear reactors require knowledge of the evaporation behavior of UO/sub 2/ at temperatures well above the melting point of 3140 K. In this study, rapid transient heating of a small spot on a UO/sub 2/ specimen was accomplished by a laser pulse, which generates a surface temperature excursion. This in turn vaporizes the target surface and the gas expands into vacuum. The surface temperature transient was monitored by a fast-response automatic optical pyrometer. The maximum surface temperatures investigated range from approx. 3700 K to approx. 4300 K. A computer program was developed to simulate the laser heating process and calculate the surface temperature evolution. The effect of the uncertainties of the high temperature material properties on the calculation was included in a sensitivity study for UO/sub 2/ vaporization. The measured surface temperatures were in satisfactory agreements.

  18. Fabrication of Natural Uranium UO2 Disks (Phase II): Texas A&M Work for Others Summary Document

    SciTech Connect (OSTI)

    Gerczak, Tyler J.; Baldwin, Charles A.; Schmidlin, Joshua E.; Henry, Jr, John James

    2015-08-28

    The steps to fabricate natural UO2 disks for an irradiation campaign led by Texas A&M University are outlined. The process was initiated with stoichiometry adjustment of parent, U3O8 powder. The next stage of sample preparation involved exploratory pellet pressing and sintering to achieve the desired natural UO2 pellet densities. Ideal densities were achieved through the use of a bimodal powder size blend. The steps involved with disk fabrication are also presented, describing the coring and thinning process executed to achieve final dimensionality.

  19. Effect of Grain Boundaries on Krypton Segregation Behavior in Irradiated Uranium Dioxide

    SciTech Connect (OSTI)

    Valderrama, Billy; He, Lingfeng; Henderson, Hunter B.; Pakarinen, Janne; Jaques, Brian; Gan, Jian; Butt, Darryl P.; Allen, Todd R.; Manuel, Michele V.

    2014-11-01

    Fission products, such as krypton (Kr), are known to be insoluble within UO2, segregating towards grain boundaries, eventually leading to a lowering of the thermal conductivity and fuel swelling. Recent computational studies have identified that differences in grain boundary structure have a significant effect on the segregation behavior of fission products. However, experimental work supporting these simulations is lacking. Atom probe tomography was used to measure the Kr distribution across grain boundaries in UO2. Polycrystalline depleted-UO2 samples was irradiated with 0.7 and 1.8 MeV Kr-ions and annealed to 1000ºC, 1300ºC, and 1600°C for 1 hour to produce a Kr-bubble dominated microstructure. The results of this work indicate a strong dependence of Kr concentration as a function of grain boundary structure. Temperature also influences grain boundary chemistry with greater Kr concentration evident at higher temperatures, resulting in a reduced Kr concentration in the bulk. While Kr migration is active at elevated temperatures, no changes in grain size or texture were observed in the irradiated UO2 samples.

  20. uranium

    National Nuclear Security Administration (NNSA)

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

    H Canyon is also being...

  1. Effect of point defects on the thermal conductivity of UO2: molecular dynamics simulations

    SciTech Connect (OSTI)

    Liu, Xiang-Yang; Stanek, Christopher Richard; Andersson, Anders David Ragnar

    2015-07-21

    The thermal conductivity of uranium dioxide (UO2) fuel is an important materials property that affects fuel performance since it is a key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. [1] The thermal conductivity of UO2 nuclear fuel is also affected by fission gas, fission products, defects, and microstructural features such as grain boundaries. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of irradiation induced point defects on the thermal conductivity of UO2, as a function of defect concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel [2].

  2. [Ni(H{sub 2}O){sub 4}]{sub 3}[U(OH,H{sub 2}O)(UO{sub 2}){sub 8}O{sub 12}(OH){sub 3}], crystal structure and comparison with uranium minerals with U{sub 3}O{sub 8}-type sheets

    SciTech Connect (OSTI)

    Rivenet, Murielle; Vigier, Nicolas; Roussel, Pascal; Abraham, Francis

    2009-04-15

    The new U(VI) compound, [Ni(H{sub 2}O){sub 4}]{sub 3}[U(OH,H{sub 2}O)(UO{sub 2}){sub 8}O{sub 12}(OH){sub 3}], was synthesized by mild hydrothermal reaction of uranyl and nickel nitrates. The crystal-structure was solved in the P-1 space group, a=8.627(2), b=10.566(2), c=12.091(4) A and alpha=110.59(1), beta=102.96(2), gamma=105.50(1){sup o}, R=0.0539 and wR=0.0464 from 3441 unique observed reflections and 151 parameters. The structure of the title compound is built from sheets of uranium polyhedra closely related to that in beta-U{sub 3}O{sub 8}. Within the sheets [(UO{sub 2})(OH)O{sub 4}] pentagonal bipyramids share equatorial edges to form chains, which are cross-linked by [(UO{sub 2})O{sub 4}] and [UO{sub 4}(H{sub 2}O)(OH)] square bipyramids and through hydroxyl groups shared between [(UO{sub 2})(OH)O{sub 4}] pentagonal bipyramids. The sheets are pillared by sharing the apical oxygen atoms of the [(UO{sub 2})(OH)O{sub 4}] pentagonal bipyramids with the oxygen atoms of [NiO{sub 2}(H{sub 2}O){sub 4}] octahedral units. That builds a three-dimensional framework with water molecules pointing towards the channels. On heating [Ni(H{sub 2}O){sub 4}]{sub 3}[U(OH,H{sub 2}O)(UO{sub 2}){sub 8}O{sub 12}(OH){sub 3}] decomposes into NiU{sub 3}O{sub 10}. - Graphical abstract: The framework of [Ni(H{sub 2}O){sub 4}]{sub 3}[U(OH,H{sub 2}O)(UO{sub 2}){sub 8}O{sub 12}(OH){sub 3}] built from uranium polyhedra sheets pillared by Ni-centered octahedra.

  3. Oxidative Dissolution of UO2 in a Simulated Groundwater Containing

    Office of Scientific and Technical Information (OSTI)

    Synthetic Nanocrystalline Mackinawite (Journal Article) | SciTech Connect Journal Article: Oxidative Dissolution of UO2 in a Simulated Groundwater Containing Synthetic Nanocrystalline Mackinawite Citation Details In-Document Search Title: Oxidative Dissolution of UO2 in a Simulated Groundwater Containing Synthetic Nanocrystalline Mackinawite The long-term success of in situ reductive immobilization of uranium (U) depends on the stability of U(IV) precipitates (e.g., uraninite) under oxic

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

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

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-09-10

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

  6. Development of an Innovative High-Thermal Conductivity UO2 Ceramic Composites Fuel Pellets with Carbon Nano-Tubes Using Spark Plasma Sintering

    SciTech Connect (OSTI)

    Subhash, Ghatu; Wu, Kuang-Hsi; Tulenko, James

    2014-03-10

    Uranium dioxide (UO2) is the most common fuel material in commercial nuclear power reactors. Despite its numerous advantages such as high melting point, good high-temperature stability, good chemical compatibility with cladding and coolant, and resistance to radiation, it suffers from low thermal conductivity that can result in large temperature gradients within the UO2 fuel pellet, causing it to crack and release fission gases. Thermal swelling of the pellets also limits the lifetime of UO2 fuel in the reactor. To mitigate these problems, we propose to develop novel UO2 fuel with uniformly distributed carbon nanotubes (CNTs) that can provide high-conductivity thermal pathways and can eliminate fuel cracking and fission gas release due to high temperatures. CNTs have been investigated extensively for the past decade to explore their unique physical properties and many potential applications. CNTs have high thermal conductivity (6600 W/mK for an individual single- walled CNT and >3000 W/mK for an individual multi-walled CNT) and high temperature stability up to 2800°C in vacuum and about 750°C in air. These properties make them attractive candidates in preparing nano-composites with new functional properties. The objective of the proposed research is to develop high thermal conductivity of UO2–CNT composites without affecting the neutronic property of UO2 significantly. The concept of this goal is to utilize a rapid sintering method (5–15 min) called spark plasma sintering (SPS) in which a mixture of CNTs and UO2 powder are used to make composites with different volume fractions of CNTs. Incorporation of these nanoscale materials plays a fundamentally critical role in controlling the performance and stability of UO2 fuel. We will use a novel in situ growth process to grow CNTs on UO2 particles for rapid sintering and develop UO2-CNT composites. This method is expected to provide a uniform distribution of CNTs at various volume fractions so that a high thermally conductive UO2-CNT composite is obtained with a minimal volume fraction of CNTs. The mixtures are sintered in the SPS facility at a range of temperatures, pressures, and time durations so as to identify the optimal processing conditions to obtain the desired microstructure of sintered UO2-CNT pellets. The second objective of the proposed work is to identify the optimal volume fraction of CNTs in the microstructure of the composites that provides the desired high thermal conductivity yet retaining the mechanical strength required for efficient function as a reactor fuel. We will systematically study the resulting microstructure (grain size, porosity, distribution of CNTs, etc.) obtained at various SPS processing conditions using optical microscopy, scanning electron microscopy (SEM), and transmission electron microscope (TEM). We will conduct indentation hardness measurements and uniaxial strength measurements as a function of volume fraction of CNTs to determine the mechanical strength and compare them to the properties of UO2. The fracture surfaces will be studied to determine the fracture characteristics that may relate to the observed cracking during service. Finally, we will perform thermal conductivity measurements on all the composites up to 1000° C. This study will relate the microstructure, mechanical properties, and thermal properties at various volume fractions of CNTs. The overall intent is to identify optimal processing conditions that will provide a well-consolidated compact with optimal microstructure and thermo-mechanical properties. The deliverables include: (1) fully characterized UO2-CNT composite with optimal CNT volume fraction and high thermal conductivity and (2) processing conditions for production of UO2-CNT composite pellets using SPS method.

  7. A Validation Study of Pin Heat Transfer for UO2 Fuel Based on the IFA-432 Experiments

    SciTech Connect (OSTI)

    Phillippe, Aaron M; Clarno, Kevin T; Banfield, James E; Ott, Larry J; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Hamilton, Steven P

    2014-01-01

    The IFA-432 (Integrated Fuel Assessment) experiments from the International Fuel Performance Experiments (IFPE) database were designed to study the effects of gap size, fuel density, and fuel densification on fuel centerline temperature in light-water-reactor fuel. An evaluation of nuclear fuel pin heat transfer in the FRAPCON-3.4 and Exnihilo codes for uranium dioxide (UO$_2$) fuel systems was performed, with a focus on the densification stage (2.2 \\unitfrac{GWd}{MT(UO$_{2}$)}). In addition, sensitivity studies were performed to evaluate the effect of the radial power shape and approximations to the geometry to account for the thermocouple hole. The analysis demonstrated excellent agreement for rods 1, 2, 3, and 5 (varying gap thicknesses and density with traditional fuel), demonstrating the accuracy of the codes and their underlying material models for traditional fuel. For rod 6, which contained unstable fuel that densified an order of magnitude more than traditional, stable fuel, the magnitude of densification was over-predicted and the temperatures were outside of the experimental uncertainty. The radial power shape within the fuel was shown to significantly impact the predicted centerline temperatures, whereas modeling the fuel at the thermocouple location as either annular or solid was relatively negligible. This has provided an initial benchmarking of the pin heat transfer capability of Exnihilo for UO$_2$ fuel with respect to a well-validated nuclear fuel performance code.

  8. PUREX/UO{sub 3} deactivation project management plan

    SciTech Connect (OSTI)

    Washenfelder, D.J.

    1993-12-01

    From 1955 through 1990, the Plutonium-Uranium Extraction Plant (PUREX) provided the United States Department of Energy Hanford Site with nuclear fuel reprocessing capability. It operated in sequence with the Uranium Trioxide (UO{sub 3}) Plant, which converted the PUREX liquid uranium nitrate product to solid UO{sub 3} powder. Final UO{sub 3} Plant operation ended in 1993. In December 1992, planning was initiated for the deactivation of PUREX and UO{sub 3} Plant. The objective of deactivation planning was to identify the activities needed to establish a passively safe, environmentally secure configuration at both plants, and ensure that the configuration could be retained during the post-deactivation period. The PUREX/UO{sub 3} Deactivation Project management plan represents completion of the planning efforts. It presents the deactivation approach to be used for the two plants, and the supporting technical, cost, and schedule baselines. Deactivation activities concentrate on removal, reduction, and stabilization of the radioactive and chemical materials remaining at the plants, and the shutdown of the utilities and effluents. When deactivation is completed, the two plants will be left unoccupied and locked, pending eventual decontamination and decommissioning. Deactivation is expected to cost $233.8 million, require 5 years to complete, and yield $36 million in annual surveillance and maintenance cost savings.

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

    DOE Patents [OSTI]

    Moore, J.E.

    1960-07-12

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

  10. Chemical reactivity of CVC and CVD SiC with UO2 at high temperatures

    SciTech Connect (OSTI)

    Silva, Chinthaka M; Katoh, Yutai; Voit, Stewart L; Snead, Lance Lewis

    2015-01-01

    Two types of silicon carbide (SiC) synthesized using two different vapor deposition processes were embedded in UO2 pellets and evaluated for their potential chemical reaction with UO2. While minor reactivity between chemical-vapor-composited (CVC) SiC and UO2 was observed at comparatively low temperatures of 1100 and 1300 C, chemical-vapor-deposited (CVD) SiC did not show any such reactivity, according to microstructural investigations. However, both CVD and CVC SiCs showed some reaction with UO2 at a higher temperature (1500 C). Elemental maps supported by phase maps obtained using electron backscatter diffraction indicated that CVC SiC was more reactive than CVD SiC at 1500 C. Furthermore, this investigation indicated the formation of uranium carbides and uranium silicide chemical phases such as UC, USi2, and U3Si2 as a result of SiC reaction with UO2.

  11. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1990-01-01

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

  12. AVLIS modified direct denitration: UO{sub 3} powder evaluation

    SciTech Connect (OSTI)

    Slagle, O.D.; Davis, N.C.; Parchen, L.J.

    1994-02-01

    The evaluation study demonstrated that AVLIS-enriched uranium converted to UO{sub 3} can be used to prepare UO{sub 3} pellets having densities in the range required for commercial power reactor fuel. Specifically, the program has demonstrated that MDD (Modified Direct Denitration)-derived UO{sub 2} powders can be reduced to sinterable UO{sub 2} powder using reduction techniques that allow control of the final powder characteristics; the resulting UO{sub 2} powders can be processed/sintered using standard powder preparation and pellet fabrication techniques to yield pellets with densities greater than 96% TD; pellet microstructures appear similar to those of power reactor fuel, and because of the high final pellet densities, it is expected that they would remain stable during in-reactor operation; the results of the present study confirm the results of a similar study carried out in 1982 (Davis and Griffin 1992). The laboratory processes were selected on the basis that they could be scaled up to standard commercial fuel processing. However, larger scale testing may be required to establish techniques compatible with commercial fuel fabrication techniques.

  13. Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

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

    George, Nathan M.; Maldonado, G. Ivan; Terrani, Kurt A.; Godfrey, Andrew T.; Gehin, Jess C.; Powers, Jeffrey J.

    2014-12-01

    Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resultingmore » operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.« less

  14. Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

    SciTech Connect (OSTI)

    George, Nathan M.; Maldonado, G. Ivan; Terrani, Kurt A.; Godfrey, Andrew T.; Gehin, Jess C.; Powers, Jeffrey J.

    2014-12-01

    Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resulting operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.

  15. Mitigation of Hydrogen Gas Generation from the Reaction of Water with Uranium Metal in K Basins Sludge

    SciTech Connect (OSTI)

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

    2010-01-29

    Means to decrease the rate of hydrogen gas generation from the chemical reaction of uranium metal with water were identified by surveying the technical literature. The underlying chemistry and potential side reactions were explored by conducting 61 principal experiments. Several methods achieved significant hydrogen gas generation rate mitigation. Gas-generating side reactions from interactions of organics or sludge constituents with mitigating agents were observed. Further testing is recommended to develop deeper knowledge of the underlying chemistry and to advance the technology aturation level. Uranium metal reacts with water in K Basin sludge to form uranium hydride (UH3), uranium dioxide or uraninite (UO2), and diatomic hydrogen (H2). Mechanistic studies show that hydrogen radicals (H·) and UH3 serve as intermediates in the reaction of uranium metal with water to produce H2 and UO2. Because H2 is flammable, its release into the gas phase above K Basin sludge during sludge storage, processing, immobilization, shipment, and disposal is a concern to the safety of those operations. Findings from the technical literature and from experimental investigations with simple chemical systems (including uranium metal in water), in the presence of individual sludge simulant components, with complete sludge simulants, and with actual K Basin sludge are presented in this report. Based on the literature review and intermediate lab test results, sodium nitrate, sodium nitrite, Nochar Acid Bond N960, disodium hydrogen phosphate, and hexavalent uranium [U(VI)] were tested for their effects in decreasing the rate of hydrogen generation from the reaction of uranium metal with water. Nitrate and nitrite each were effective, decreasing hydrogen generation rates in actual sludge by factors of about 100 to 1000 when used at 0.5 molar (M) concentrations. Higher attenuation factors were achieved in tests with aqueous solutions alone. Nochar N960, a water sorbent, decreased hydrogen generation by no more than a factor of three while disodium phosphate increased the corrosion and hydrogen generation rates slightly. U(VI) showed some promise in attenuating hydrogen but only initial testing was completed. Uranium metal corrosion rates also were measured. Under many conditions showing high hydrogen gas attenuation, uranium metal continued to corrode at rates approaching those observed without additives. This combination of high hydrogen attenuation with relatively unabated uranium metal corrosion is significant as it provides a means to eliminate uranium metal by its corrosion in water without the accompanying hazards otherwise presented by hydrogen generation.

  16. Surface reactions of ethanol over UO2(100) thin film

    SciTech Connect (OSTI)

    S. D. Senanayake; Mudiyanselage, K.; Burrell, A. K.; Sadowski, J. T.; Idriss, H.

    2015-10-08

    The study of the reactions of oxygenates on well-defined oxide surfaces is important for the fundamental understanding of heterogeneous chemical pathways that are influenced by atomic geometry, electronic structure, and chemical composition. In this work, an ordered uranium oxide thin film surface terminated in the (100) orientation is prepared on a LaAlO3 substrate and studied for its reactivity with a C-2 oxygenate, ethanol (CH3CH2OH). With the use of synchrotron X-ray photoelectron spectroscopy (XPS), we have probed the adsorption and desorption processes observed in the valence band, C 1s, O 1s, and U 4f to investigate the bonding mode, surface composition, electronic structure, and probable chemical changes to the stoichiometric-UO2(100) [smooth-UO2(100)] and Ar+-sputtered UO2(100) [rough-UO2(100)] surfaces. Unlike UO2(111) single crystal and UO2 thin film, Ar-ion-sputtering of this UO2(100) did not result in noticeable reduction of U cations. Upon ethanol adsorption (saturation occurred at 0.5 ML), only the ethoxy (CH3CH2O) species is formed on smooth-UO2(100) whereas initially formed ethoxy species are partially oxidized to surface acetate (CH3COO–) on the Ar+-sputtered UO2(100) surface. Furthermore, all ethoxy and acetate species are removed from the surface between 600 and 700 K.

  17. Surface reactions of ethanol over UO2(100) thin film

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

    S. D. Senanayake; Mudiyanselage, K.; Burrell, A. K.; Sadowski, J. T.; Idriss, H.

    2015-10-08

    The study of the reactions of oxygenates on well-defined oxide surfaces is important for the fundamental understanding of heterogeneous chemical pathways that are influenced by atomic geometry, electronic structure, and chemical composition. In this work, an ordered uranium oxide thin film surface terminated in the (100) orientation is prepared on a LaAlO3 substrate and studied for its reactivity with a C-2 oxygenate, ethanol (CH3CH2OH). With the use of synchrotron X-ray photoelectron spectroscopy (XPS), we have probed the adsorption and desorption processes observed in the valence band, C 1s, O 1s, and U 4f to investigate the bonding mode, surface composition,more » electronic structure, and probable chemical changes to the stoichiometric-UO2(100) [smooth-UO2(100)] and Ar+-sputtered UO2(100) [rough-UO2(100)] surfaces. Unlike UO2(111) single crystal and UO2 thin film, Ar-ion-sputtering of this UO2(100) did not result in noticeable reduction of U cations. Upon ethanol adsorption (saturation occurred at 0.5 ML), only the ethoxy (CH3CH2O–) species is formed on smooth-UO2(100) whereas initially formed ethoxy species are partially oxidized to surface acetate (CH3COO–) on the Ar+-sputtered UO2(100) surface. Furthermore, all ethoxy and acetate species are removed from the surface between 600 and 700 K.« less

  18. Chemical reactivity of CVC and CVD SiC with UO2 at high temperatures

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

    Silva, Chinthaka M; Katoh, Yutai; Voit, Stewart L; Snead, Lance Lewis

    2015-01-01

    Two types of silicon carbide (SiC) synthesized using two different vapor deposition processes were embedded in UO2 pellets and evaluated for their potential chemical reaction with UO2. While minor reactivity between chemical-vapor-composited (CVC) SiC and UO2 was observed at comparatively low temperatures of 1100 and 1300 C, chemical-vapor-deposited (CVD) SiC did not show any such reactivity, according to microstructural investigations. However, both CVD and CVC SiCs showed some reaction with UO2 at a higher temperature (1500 C). Elemental maps supported by phase maps obtained using electron backscatter diffraction indicated that CVC SiC was more reactive than CVD SiC at 1500more » C. Furthermore, this investigation indicated the formation of uranium carbides and uranium silicide chemical phases such as UC, USi2, and U3Si2 as a result of SiC reaction with UO2.« less

  19. PUREX/UO{sub 3} facilities deactivation lessons learned history

    SciTech Connect (OSTI)

    Hamrick, D.G.; Gerber, M.S.

    1995-01-01

    The Plutonium-Uranium Extraction (PUREX) Facility operated from 1956-1972, from 1983-1988, and briefly during 1989-1990 to produce for national defense at the Hanford Site in Washington State. The Uranium Trioxide (UO{sub 3}) Facility operated at the Hanford Site from 1952-1972, 1984-1988, and briefly in 1993. Both plants were ordered to permanent shutdown by the U.S. Department of Energy (DOE) in December 1992, thus initiating their deactivation phase. Deactivation is that portion of a facility`s life cycle that occurs between operations and final decontamination and decommissioning (D&D). This document details the history of events, and the lessons learned, from the time of the PUREX Stabilization Campaign in 1989-1990, through the end of the first full fiscal year (FY) of the deactivation project (September 30, 1994).

  20. TRIMOLECULAR REACTIONS OF URANIUM HEXAFLUORIDE WITH WATER

    SciTech Connect (OSTI)

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

    2010-02-25

    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.

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

    SciTech Connect (OSTI)

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

    2013-06-01

    Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O? lattice in an irradiated (60 MW d kg?¹) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (~0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am³? species within an [AmO?]¹³? coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix. - Graphical abstract: Americium LIII XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 ?m×300 ?m beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Am is found as trivalent in the UO? matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: • Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. • The americium redox state as determined from XAS data of irradiated fuel material was Am(III). • In the sample, the Am³? face an AmO?¹³?coordination environment in the (Pu,U)O? matrix. • The americium dioxide is reduced by the uranium dioxide matrix.

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

    SciTech Connect (OSTI)

    Icenhour, A.S.

    2003-09-10

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

  3. Migration Mechanisms of Oxygen Interstitial Clusters in UO2 ...

    Office of Scientific and Technical Information (OSTI)

    Migration Mechanisms of Oxygen Interstitial Clusters in UO2 Citation Details In-Document Search Title: Migration Mechanisms of Oxygen Interstitial Clusters in UO2 Understanding the ...

  4. Sensitivity of UO2 Stability in a Reducing Environment on Radiolysis Model Parameters

    SciTech Connect (OSTI)

    Wittman, Richard S.; Buck, Edgar C.

    2012-09-01

    Results for a radiolysis model sensitivity study of radiolytically produced H2O2 are presented as they relate to Spent (or Used) Light Water Reactor uranium oxide (UO2) nuclear fuel (UNF) oxidation in a low oxygen environment. The model builds on previous reaction kinetic studies to represent the radiolytic processes occurring at the nuclear fuel surface. Hydrogen peroxide (H2O2) is the dominant oxidant for spent nuclear fuel in an O2-depleted water environment.

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

    DOE Patents [OSTI]

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

    2001-01-01

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

  6. Thermal Stabilization of {sup 233}UO{sub 2}, {sup 233}UO{sub 3}, and {sup 233}U{sub 3}O{sub 8}

    SciTech Connect (OSTI)

    Thein, S.M.

    2000-07-26

    This report identifies an appropriate thermal stabilization temperature for {sup 233}U oxides. The temperature is chosen principally on the basis of eliminating moisture and other residual volatiles. This report supports the U. S. Department of Energy (DOE) Standard for safe storage of {sup 233}U (DOE 2000), written as part of the response to Recommendation 97-1 of the Defense Nuclear Facilities Safety Board (DNFSB), addressing safe storage of {sup 233}U. The primary goals in choosing a stabilization temperature are (1) to ensure that the residual volatiles content is less than 0.5 wt % including moisture, which might produce pressurizing gases via radiolysis during long-term sealed storage; (2) to minimize potential for water readsorption above the 0.5 wt % threshold; and (3) to eliminate reactive uranium species. The secondary goals are (1) to reduce potential future chemical reactivity and (2) to increase the particle size thereby reducing the potential airborne release fraction (ARF) under postulated accident scenarios. The prevalent species of uranium oxide are the chemical forms UO{sub 2}, UO{sub 3}, and U{sub 3}O{sub 8}. Conversion to U{sub 3}O{sub 8} is sufficient to accomplish all of the desired goals. The preferred storage form is U{sub 3}O{sub 8} because it is more stable than UO{sub 2} or UO{sub 3} in oxidizing atmospheres. Heating in an oxidizing atmosphere at 750 C for at least one hour will achieve the thermal stabilization desired.

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

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

  8. Extraction of uranium from spent fuels using liquefied gases

    SciTech Connect (OSTI)

    Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi

    2007-07-01

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

  9. Bisphosphine dioxides

    DOE Patents [OSTI]

    Moloy, K.G.

    1990-02-20

    A process is described for the production of organic bisphosphine dioxides from organic bisphosphonates. The organic bisphosphonate is reacted with a Grignard reagent to give relatively high yields of the organic bisphosphine dioxide.

  10. Bisphosphine dioxides

    DOE Patents [OSTI]

    Moloy, Kenneth G.

    1990-01-01

    A process for the production of organic bisphosphine dioxides from organic bisphosphonates. The organic bisphosphonate is reacted with a Grignard reagent to give relatively high yields of the organic bisphosphine dioxide.

  11. Some effects of data base variations on numerical simulations of uranium migration

    SciTech Connect (OSTI)

    Carnahan, C.L.

    1987-12-01

    Numerical simulations of migration of chemicals in the geosphere depend on knowledge of identities of chemical species and on values of chemical equilibrium constants supplied to the simulators. In this work, some effects of variability in assumed speciation and in equilibrium constants were examined, using migration of uranium as an example. Various simulations were done of uranium migration in systems with varying oxidation potential, pH, and mator component content. A simulation including formation of aqueous species UO/sub 2//sup 2 +/, UO/sub 2/CO/sub 3//sup 0/, UO/sub 2/(CO/sub 3/)/sub 2//sup 2 -/, UO/sub 2/(CO/sub 3/)/sub 3//sup 4 -/, (UO/sub 2/)/sub 2/CO/sub 3/(OH)/sub 3//sup -/, UO/sub 2//sup +/, U(OH)/sub 4//sup 0/, and U(OH)/sub 5//sup -/ is compared to simulation excluding formation of UO/sub 2//sup +/ and U(OH)/sub 5//sup -/. These simulations relied on older data bases, and they are compared to a further simulation using recently published data on formation of U(OH)/sub 4//sup 0/, (UO/sub 2/)/sub 2/CO/sub 3/(OH)/sub 3//sup -/, UO/sub 2/(CO/sub 3/)/sub 5//sup 5 -/, and U(CO/sub 3/)/sub 5//sup 6 -/. Significant differences in dissolved uranium concentrations are noted among the simulations. Differences are noted also in precipitation of two solids, USiO/sub 4/(c) (coffinite) and CaUO/sub 4/(c) (calcium uranate), although the solubility products of the solids were not varied in the simulations. 18 refs., 9 figs., 2 tabs.

  12. Uranium Transport Modeling

    SciTech Connect (OSTI)

    Bostick, William D.

    2008-01-15

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

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

    SciTech Connect (OSTI)

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

    2008-07-08

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

  14. Thorium dioxide: properties and nuclear applications

    SciTech Connect (OSTI)

    Belle, J.; Berman, R.M.

    1984-01-01

    This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core.

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

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

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

    2015-04-13

    Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K2(UO2)3O4·4H2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarily UO3·2H2O, with minor phases ofmore » U3O8 and UO2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.« less

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

    SciTech Connect (OSTI)

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

    1994-08-01

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

  17. Microstructure changes and thermal conductivity reduction in UO2 following

    Office of Scientific and Technical Information (OSTI)

    3.9 MeV He2+ ion irradiation (Journal Article) | SciTech Connect Microstructure changes and thermal conductivity reduction in UO2 following 3.9 MeV He2+ ion irradiation Citation Details In-Document Search Title: Microstructure changes and thermal conductivity reduction in UO2 following 3.9 MeV He2+ ion irradiation The microstructural changes and associated effects on thermal conductivity were examined in UO2 after irradiation using 3.9 MeV He2+ ions. Lattice expansion of UO2 was observed in

  18. PUREX/UO3 Facilities deactivation lessons learned history

    SciTech Connect (OSTI)

    Gerber, M.S.

    1996-09-19

    Disconnecting the criticality alarm permanently in June 1996 signified that the hazards in the PUREX (plutonium-uranium extraction) plant had been so removed and reduced that criticality was no longer a credible event. Turning off the PUREX criticality alarm also marked a salient point in a historic deactivation project, 1 year before its anticipated conclusion. The PUREX/UO3 Deactivation Project began in October 1993 as a 5-year, $222.5- million project. As a result of innovations implemented during 1994 and 1995, the project schedule was shortened by over a year, with concomitant savings. In 1994, the innovations included arranging to send contaminated nitric acid from the PUREX Plant to British Nuclear Fuels, Limited (BNFL) for reuse and sending metal solutions containing plutonium and uranium from PUREX to the Hanford Site tank farms. These two steps saved the project $36.9- million. In 1995, reductions in overhead rate, work scope, and budget, along with curtailed capital equipment expenditures, reduced the cost another $25.6 million. These savings were achieved by using activity-based cost estimating and applying technical schedule enhancements. In 1996, a series of changes brought about under the general concept of ``reengineering`` reduced the cost approximately another $15 million, and moved the completion date to May 1997. With the total savings projected at about $75 million, or 33.7 percent of the originally projected cost, understanding how the changes came about, what decisions were made, and why they were made becomes important. At the same time sweeping changes in the cultural of the Hanford Site were taking place. These changes included shifting employee relations and work structures, introducing new philosophies and methods in maintaining safety and complying with regulations, using electronic technology to manage information, and, adopting new methods and bases for evaluating progress. Because these changes helped generate cost savings and were accompanied by and were an integral part of sweeping ``culture changes,`` the story of the lessons learned during the PUREX Deactivation Project are worth recounting. Foremost among the lessons is recognizing the benefits of ``right to left`` project planning. A deactivation project must start by identifying its end points, then make every task, budget, and organizational decision based on reaching those end points. Along with this key lesson is the knowledge that project planning and scheduling should be tied directly to costing, and the project status should be checked often (more often than needed to meet mandated reporting requirements) to reflect real-time work. People working on a successful project should never be guessing about its schedule or living with a paper schedule that does not represent the actual state of work. Other salient lessons were learned in the PUREX/UO3 Deactivation Project that support these guiding principles. They include recognizing the value of independent review, teamwork, and reengineering concepts; the need and value of cooperation between the DOE, its contractors, regulators, and stakeholders; and the essential nature of early and ongoing communication. Managing a successful project also requires being willing to take a fresh look at safety requirements and to apply them in a streamlined and sensible manner to deactivating facilities; draw on the enormous value of resident knowledge acquired by people over years and sometimes decades of working in old plants; and recognize the value of bringing in outside expertise for certain specialized tasks.This approach makes possible discovering the savings that can come when many creative options are pursued persistently and the wisdom of leaving some decisions to the future. The essential job of a deactivation project is to place a facility in a safe, stable, low-maintenance mode, for an interim period. Specific end points are identified to recognize and document this state. Keeping the limited objectives of the project in mind can guide decisions that reduce risks with minimal manipul

  19. Validation of the WATEQ4 geochemical model for uranium

    SciTech Connect (OSTI)

    Krupka, K.M.; Jenne, E.A.; Deutsch, W.J.

    1983-09-01

    As part of the Geochemical Modeling and Nuclide/Rock/Groundwater Interactions Studies Program, a study was conducted to partially validate the WATEQ4 aqueous speciation-solubility geochemical model for uranium. The solubility controls determined with the WATEQ4 geochemical model were in excellent agreement with those laboratory studies in which the solids schoepite (UO/sub 2/(OH)/sub 2/ . H/sub 2/O), UO/sub 2/(OH)/sub 2/, and rutherfordine ((UO/sub 2/CO/sub 3/) were identified as actual solubility controls for uranium. The results of modeling solution analyses from laboratory studies of uranyl phosphate solids, however, identified possible errors in the characterization of solids in the original solubility experiments. As part of this study, significant deficiencies in the WATEQ4 thermodynamic data base for uranium solutes and solids were corrected. Revisions included recalculation of selected uranium reactions. Additionally, thermodynamic data for the hydroxyl complexes of U(VI), including anionic (VI) species, were evaluated (to the extent permitted by the available data). Vanadium reactions were also added to the thermodynamic data base because uranium-vanadium solids can exist in natural ground-water systems. This study is only a partial validation of the WATEQ4 geochemical model because the available laboratory solubility studies do not cover the range of solid phases, alkaline pH values, and concentrations of inorganic complexing ligands needed to evaluate the potential solubility of uranium in ground waters associated with various proposed nuclear waste repositories. Further validation of this or other geochemical models for uranium will require careful determinations of uraninite solubility over the pH range of 7 to 10 under highly reducing conditions and of uranyl hydroxide and phosphate solubilities over the pH range of 7 to 10 under oxygenated conditions.

  20. Fuel Performance Experiments on the Atomistic Level, Studying Fuel Through Engineered Single Crystal UO2

    SciTech Connect (OSTI)

    Burgett, Eric; Deo, Chaitanya; Phillpot, Simon

    2015-05-08

    Fuel Performance Experiments on the Atomistic Level, Studying Fuel Through Engineered Single Crystal UO2

  1. plutonium dioxide

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

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

  2. Oxidative Dissolution of UO2 in a Simulated Groundwater Containing

    Office of Scientific and Technical Information (OSTI)

    Synthetic Nanocrystalline Mackinawite (Journal Article) | SciTech Connect Journal Article: Oxidative Dissolution of UO2 in a Simulated Groundwater Containing Synthetic Nanocrystalline Mackinawite Citation Details In-Document Search Title: Oxidative Dissolution of UO2 in a Simulated Groundwater Containing Synthetic Nanocrystalline Mackinawite Authors: Bi, Yuqiang ; Hyuna, Sung Pil ; Kukkadapu, Ravi K. ; Hayes, Kim F. ; , Publication Date: 2014-03-18 OSTI Identifier: 1124154 Report Number(s):

  3. Effect of Co-solutes on the Products and Solubility of Uranium(VI) Precipitated with Phosphate

    SciTech Connect (OSTI)

    Mehta, Vrajesh; Maillot, Fabien; Wang, Zheming; Catalano, Jeffrey G.; Giammar, Daniel E.

    2014-01-22

    Uranyl phosphate solids are often found with uranium ores, and their low solubility makes them promising target phases for in situ remediation of uranium-contaminated subsurface environments. The products and solubility of uranium(VI) precipitated with phosphate can be affected by the pH, dissolved inorganic carbon (DIC) concentration, and co-solute composition (e.g. Na+/Ca2+) of the groundwater. Batch experiments were performed to study the effect of these parameters on the products and extent of uranium precipitation induced by phosphate addition. In the absence of co-solute cations, chernikovite [H3O(UO2)(PO4)•3H2O] precipitated despite uranyl orthophosphate [(UO2)3(PO4)2•4H2O] being thermodynamically more favorable under certain conditions. As determined using X-ray diffraction, electron microscopy, and laser induced fluorescence spectroscopy, the presence of Na+ or Ca2+ as a co-solute led to the precipitation of sodium autunite ([Na2(UO2)2(PO4)2] and autunite [Ca(UO2)2(PO4)2]), which are structurally similar to chernikovite. In the presence of sodium, the dissolved U(VI) concentrations were generally in agreement with equilibrium predictions of sodium autunite solubility. However, in the calcium-containing systems, the observed concentrations were below the predicted solubility of autunite, suggesting the possibility of uranium adsorption to or incorporation in a calcium phosphate precipitate in addition to the precipitation of autunite.

  4. Dissolution Kinetics of Synthetic and Natural Meta-Autunite Minerals, X??n????[(UO?)(PO?)]? ? xH?O, Under Acidic Conditions

    SciTech Connect (OSTI)

    Wellman, Dawn M.; Gunderson, Katie M.; Icenhower, Jonathan P.; Forrester, Steven W.

    2007-11-01

    Mass transport within the uranium geochemical cycle is impacted by the availability of phosphorous. In oxidizing environments, in which the uranyl ionic species is typically mobile, formation of sparingly soluble uranyl phosphate minerals exert a strong influence on uranium transport. Autunite group minerals have been identified as the long-term uranium controlling phases in many systems of geochemical interest. Anthropogenic operations related to uranium mining operations have created acidic environments, exposing uranyl phosphate minerals to low pH groundwaters. Investigations regarding the dissolution behavior of autunite group minerals under acidic conditions have not been reported; consequently, knowledge of the longevity of uranium controlling solids is incomplete. The purpose of this investigation was to: 1) quantify the dissolution kinetics of natural calcium and synthetic sodium meta-autunite, under acidic conditions, 2) measure the effect of temperature and pH on meta-autunite mineral dissolution, and 3) investigate the formation of secondary uranyl phosphate phases as long-term controls on uranium migration. Single-pass flow-through (SPFT) dissolution tests were conducted over the pH range of 2 to 5 and from 5° to 70°C. Results presented here illustrate meta-autunite dissolution kinetics are strongly dependent on pH, but are relatively insensitive to temperature variations. In addition, the formation of secondary uranyl-phosphate phases such as, uranyl phosphate, (UO2)3(PO4)2 ? 4 H2O, may serve as a secondary phase limiting the migration of uranium in the environment.

  5. Uranium diphosphonates templated by interlayer organic amines

    SciTech Connect (OSTI)

    Nelson, Anna-Gay D.; Alekseev, Evgeny V.; Institut fuer Kristallographie, RWTH Aachen University, D-52066 Aachen ; Albrecht-Schmitt, Thomas E.; Department of Chemistry and Biochemistry, University of Notre Dame, IN 46556 ; Ewing, Rodney C.

    2013-02-15

    The hydrothermal treatment of uranium trioxide and methylenediphosphonic acid with a variety of amines (2,2-dipyridyl, triethylenediamine, ethylenediamine, and 1,10-phenanthroline) at 200 Degree-Sign C results in the crystallization of a series of layered uranium diphosphonate compounds, [C{sub 10}H{sub 9}N{sub 2}]{l_brace}UO{sub 2}(H{sub 2}O)[CH{sub 2}(PO{sub 3})(PO{sub 3}H)]{r_brace} (Ubip2), [C{sub 6}H{sub 14}N{sub 2}]{l_brace}(UO{sub 2}){sub 2}[CH{sub 2}(PO{sub 3})(PO{sub 3}H)]{sub 2}{center_dot}2H{sub 2}O{r_brace} (UDAB), [C{sub 2}H{sub 10}N{sub 2}]{sub 2}{l_brace}(UO{sub 2}){sub 2}(H{sub 2}O){sub 2}[CH{sub 2}(PO{sub 3}){sub 2}]{sub 2}{center_dot}0.5H{sub 2}O{r_brace} (Uethyl), and [C{sub 12}H{sub 9}N{sub 2}]{l_brace}UO{sub 2}(H{sub 2}O)[CH{sub 2}(PO{sub 3})(PO{sub 3}H)]{r_brace} (Uphen). The crystal structures of the compounds are based on UO{sub 7} units linked by methylenediphosphonate molecules to form two-dimensional anionic sheets in Ubip2 and UDAB, and one-dimensional anionic chains in Uethyl and Uphen, which are charge balanced by protonated amine molecules. Interaction of the amine molecules with phosphonate oxygens and water molecules results in extensive hydrogen bonding in the interlayer. These amine molecules serve both as structure-directing agents and charge-balancing cations for the anionic uranium phosphonate sheets and chains in the formation of the different coordination geometries and topologies of each structure. Reported herein are the syntheses, structural and spectroscopic characterization of the synthesized compounds. - Graphical abstract: The Raman spectra of the synthesized compounds and an illustration of the stacking of the layers with the diprotonated triethylenediamine molecules in [C{sub 6}H{sub 14}N{sub 2}]{l_brace}(UO{sub 2}){sub 2}[CH{sub 2}(PO{sub 3})(PO{sub 3}H)]{sub 2}{center_dot}2H{sub 2}O{r_brace} UDAB. Solvent water molecules are removed for clarity. The corresponding Raman spectra for the complexes synthesized is also shown. The structure is constructed from UO{sub 7} pentagonal bipyramids (yellow), oxygen=red, phosphorus=magenta, carbon=black, and nitrogen=blue. Highlights: Black-Right-Pointing-Pointer Organic amines act both as charge-balancing and as structure-directing agents. Black-Right-Pointing-Pointer Extensive hydrogen bonding interactions with solvent water molecules and amines. Black-Right-Pointing-Pointer Altering the organic amine (size or flexibility) affects structure formation.

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

    SciTech Connect (OSTI)

    Yoshida, Thomas M.; Meyers, Lisa

    2015-11-10

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

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

    SciTech Connect (OSTI)

    Not Available

    2011-06-22

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

  8. Synthesis and crystal structure of (NH{sub 4}){sub 3}[UO{sub 2}(CH{sub 3}COO){sub 3}]{sub 2}[UO{sub 2}(CH{sub 3}COO)(NCS){sub 2}(H{sub 2}O)

    SciTech Connect (OSTI)

    Serezhkina, L. B.; Peresypkina, E. V.; Virovets, A. V.; Karasev, M. O.

    2010-01-15

    Single crystals of the compound (NH{sub 4}){sub 3}[UO{sub 2}(CH{sub 3}COO){sub 3}]{sub 2}[UO{sub 2}(CH{sub 3}COO)(NCS){sub 2}(H{sub 2}O)] (I) are synthesized, and their structure is investigated using X-ray diffraction. Compound I crystallizes in the monoclinic system with the unit cell parameters a = 18.3414(6) A, b = 16.3858(7) A, c = 12.4183(5) A, {beta} = 92.992(1){sup o}, space group C2/c, Z = 4, V = 3727.1(3) A{sup 3}, and R = 0.0253. The uranium-containing structural units of crystals I are mononuclear complexes of two types with an island structure, i.e., the [UO{sub 2}(CH{sub 3}COO){sub 3}]{sup -} anionic complexes belonging to the crystal-chemical group (AB{sub 3}{sup 01} = UO{sub 2}{sup 2+}, B{sup 01} = CH{sub 3}COO{sup -}) of the uranyl complexes and the [UO{sub 2}(CH{sub 3}COO)(NCS){sub 2}(H{sub 2}O)]{sup -} anionic complexes belonging to the crystal-chemical group AB{sup 01}M{sub 3}{sup 1} (A = UO{sub 2}{sup 2+}, B{sup 01} = CH{sub 3}COO{sup -}, M{sup 1} = NCS{sup -} or H{sub 2}O).

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

    DOE Patents [OSTI]

    Friedman, Horace A. (Oak Ridge, TN)

    1985-01-01

    A method for decontaminating uranium product from the Purex 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 uranium by solvent extraction with 30 vol. % tributyl phosphate in n-dodecane.

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

    DOE Patents [OSTI]

    Friedman, H.A.

    1984-06-13

    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.

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

    Office of Environmental Management (EM)

    Low-Enriched Uranium | Department of Energy Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium The U.S. Department of Energy (DOE) owns and manages an inventory of depleted uranium (DU), natural uranium (NU), and low-enriched uranium (LEU) that is currently stored in large cylinders as depleted uranium hexafluoride (DUF6), natural uranium hexafluoride (NUF6), and

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

    SciTech Connect (OSTI)

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

    2015-04-13

    Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K2(UO2)3O4·4H2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarily UO3·2H2O, with minor phases of U3O8 and UO2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.

  13. [La(UO{sub 2})V{sub 2}O{sub 7}][(UO{sub 2})(VO{sub 4})] the first lanthanum uranyl-vanadate with structure built from two types of sheets based upon the uranophane anion-topology

    SciTech Connect (OSTI)

    Mer, A.; Obbade, S.; Rivenet, M.; Renard, C.; Abraham, F.

    2012-01-15

    The new lanthanum uranyl vanadate divanadate, [La(UO{sub 2})V{sub 2}O{sub 7}][(UO{sub 2})(VO{sub 4})] was obtained by reaction at 800 Degree-Sign C between lanthanum chloride, uranium oxide (U{sub 3}O{sub 8}) and vanadium oxide (V{sub 2}O{sub 5}) and the structure was determined from single-crystal X-ray diffraction data. This compound crystallizes in the orthorhombic system with space group P2{sub 1}2{sub 1}2{sub 1} and unit-cell parameters a=6.9470(2) A, b=7.0934(2) A, c=25.7464(6) A, V=1268.73(5) A{sup 3}, Z=4. A full matrix least-squares refinement yielded R{sub 1}=0.0219 for 5493 independent reflections. The crystal structure is characterized by the stacking of uranophane-type sheets {sup 2}{sub {infinity}}[(UO{sub 2})(VO{sub 4})]{sup -} and double layers {sup 2}{sub {infinity}}[La(UO{sub 2})(V{sub 2}O{sub 7})]{sup +} connected through La-O bonds involving the uranyl oxygen of the uranyl-vanadate sheets. The double layers result from the connection of two {sup 2}{sub {infinity}}[La(UO{sub 2})(VO{sub 4}){sub 2}]{sup -} sheets derived from the uranophane anion-topology by replacing half of the uranyl ions by lanthanum atoms and connected through the formation of divanadate entities. - Graphical abstract: A view of the three-dimensional structure of [La(UO{sub 2})V{sub 2}O{sub 7}][(UO{sub 2})(VO{sub 4})]. Highlights: Black-Right-Pointing-Pointer New lanthanum uranyl vanadate divanadate has been synthesized. Black-Right-Pointing-Pointer Structure was determined from single-crystal X-ray diffraction data. Black-Right-Pointing-Pointer Structure is characterized by uranophane-type sheets and double layers {sup 2}{sub {infinity}}[La(UO{sub 2})(V{sub 2}O{sub 7})]{sup +}.

  14. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

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

  15. Uranium industry annual 1996

    SciTech Connect (OSTI)

    1997-04-01

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

  16. Communication: Relativistic Fock-space coupled cluster study of small building blocks of larger uranium complexes

    SciTech Connect (OSTI)

    Tecmer, Pawe? Visscher, Lucas; Severo Pereira Gomes, André; Knecht, Stefan

    2014-07-28

    We present a study of the electronic structure of the [UO{sub 2}]{sup +}, [UO{sub 2}]{sup 2} {sup +}, [UO{sub 2}]{sup 3} {sup +}, NUO, [NUO]{sup +}, [NUO]{sup 2} {sup +}, [NUN]{sup ?}, NUN, and [NUN]{sup +} molecules with the intermediate Hamiltonian Fock-space coupled cluster method. The accuracy of mean-field approaches based on the eXact-2-Component Hamiltonian to incorporate spin–orbit coupling and Gaunt interactions are compared to results obtained with the Dirac–Coulomb Hamiltonian. Furthermore, we assess the reliability of calculations employing approximate density functionals in describing electronic spectra and quantities useful in rationalizing Uranium (VI) species reactivity (hardness, electronegativity, and electrophilicity)

  17. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-08-01

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

  18. Excess Uranium Management

    Broader source: Energy.gov [DOE]

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

  19. Density Functional Theory Calculations of Mass Transport in UO2

    SciTech Connect (OSTI)

    Andersson, Anders D.; Dorado, Boris; Uberuaga, Blas P.; Stanek, Christopher R.

    2012-06-26

    In this talk we present results of density functional theory (DFT) calculations of U, O and fission gas diffusion in UO{sub 2}. These processes all impact nuclear fuel performance. For example, the formation and retention of fission gas bubbles induce fuel swelling, which leads to mechanical interaction with the clad thereby increasing the probability for clad breach. Alternatively, fission gas can be released from the fuel to the plenum, which increases the pressure on the clad walls and decreases the gap thermal conductivity. The evolution of fuel microstructure features is strongly coupled to diffusion of U vacancies. Since both U and fission gas transport rates vary strongly with the O stoichiometry, it is also important to understand O diffusion. In order to better understand bulk Xe behavior in UO{sub 2{+-}x} we first calculate the relevant activation energies using DFT techniques. By analyzing a combination of Xe solution thermodynamics, migration barriers and the interaction of dissolved Xe atoms with U, we demonstrate that Xe diffusion predominantly occurs via a vacancy-mediated mechanism. Since Xe transport is closely related to diffusion of U vacancies, we have also studied the activation energy for this process. In order to explain the low value of 2.4 eV found for U migration from independent damage experiments (not thermal equilibrium) the presence of vacancy clusters must be included in the analysis. Next we investigate species transport on the (111) UO{sub 2} surface, which is motivated by the formation of small voids partially filled with fission gas atoms (bubbles) in UO{sub 2} under irradiation. Surface diffusion could be the rate-limiting step for diffusion of such bubbles, which is an alternative mechanism for mass transport in these materials. As expected, the activation energy for surface diffusion is significantly lower than for bulk transport. These results are further discussed in terms of engineering-scale fission gas release models. Finally, oxidation of UO{sub 2} and the importance of cluster formation for understanding thermodynamic and kinetic properties of UO{sub 2+x} are investigated.

  20. Nitrogen dioxide detection

    DOE Patents [OSTI]

    Sinha, Dipen N. (Los Alamos, NM); Agnew, Stephen F. (Los Alamos, NM); Christensen, William H. (Buena Park, CA)

    1993-01-01

    Method and apparatus for detecting the presence of gaseous nitrogen dioxide and determining the amount of gas which is present. Though polystyrene is normally an insulator, it becomes electrically conductive in the presence of nitrogen dioxide. Conductance or resistance of a polystyrene sensing element is related to the concentration of nitrogen dioxide at the sensing element.

  1. {gamma}-Radiolysis of NaCl Brine in the Presence of UO{sub 2}(s): Effects of Hydrogen and Bromide

    SciTech Connect (OSTI)

    Metz, Volker; Bohnert, Elke; Kelm, Manfred; Schild, Dieter; Kienzler, Bernhard

    2007-07-01

    A concentrated NaCl solution was {gamma}-irradiated in autoclaves under a pressure of 25 MPa. A set of experiments were conducted in 6 mol (kg H{sub 2}O){sup -1} NaCl solution in the presence of UO{sub 2}(s) pellets; in a second set of experiments, {gamma}-radiolysis of the NaCl brine was studied without UO{sub 2}(s). Hydrogen, oxygen and chlorate were formed as long-lived radiolysis products. Due to the high external pressure, all radiolysis products remained dissolved. H{sub 2} and O{sub 2} reached steady state concentrations in the range of 5.10{sup -3} to 6.10{sup -2} mol (kg H{sub 2}O){sup -1} corresponding to a partial gas pressure of {approx}2 to {approx}20 MPa. Radiolytic formation of hydrogen and oxygen increased with the concentration of bromide added to solution. Both, in the presence of bromide, resulting in a relatively high radiolytic yield, and in the absence of bromide surfaces of the UO{sub 2}(s) samples were oxidized, and concentration of dissolved uranium reached the solubility limit of the schoepite / NaUO{sub 2}O(OH)(cr) transition. At the end of the experiments, the pellets were covered by a surface layer of a secondary solid phase having a composition close to Na{sub 2}U{sub 2}O{sub 7}. The experimental results demonstrate that bromide counteracts an H{sub 2} inhibition effect on radiolysis gas production, even at a concentration ratio of [H{sub 2}] / [Br{sup -}] > 100. The present observations are related to the competitive reactions of OH radicals with H{sub 2}, Br{sup -} and Cl{sup -}. A similar competition of hydrogen and bromide, controlling the yield of {gamma}-radiolysis products, is expected for solutions of lower Cl{sup -} concentration. (authors)

  2. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2 +/- x: Implications for nuclear fuel performance modeling

    SciTech Connect (OSTI)

    Giovanni Pastore; Michael R. Tonks; Derek R. Gaston; Richard L. Williamson; David Andrs; Richard Martineau

    2014-03-01

    Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2x nonstoichiometrywere used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Risø fuel rod irradiation experiment was simulated. 2014 Elsevier B.V. All rights

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

    SciTech Connect (OSTI)

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

    2004-05-12

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

  4. Measurement of the Auger parameter and Wagner plot for uranium compounds

    SciTech Connect (OSTI)

    Holliday, Kiel S.; Siekhaus, Wigbert; Nelson, Art J.

    2013-05-15

    In this study, the photoemission from the U 4f{sub 7/2} and 4d{sub 5/2} states and the U N{sub 6}O{sub 45}O{sub 45} and N{sub 67}O{sub 45}V x-ray excited Auger transitions were measured for a range of uranium compounds. The data are presented in Wagner plots and the Auger parameter is calculated to determine the utility of this technique in the analysis of uranium materials. It was demonstrated that the equal core-level shift assumption holds for uranium. It was therefore possible to quantify the relative relaxation energies, and uranium was found to have localized core-hole shielding. The position of compounds within the Wagner plot made it possible to infer information on bonding character and local electron density. The relative ionicity of the uranium compounds studied follows the trend UF{sub 4} > UO{sub 3} > U{sub 3}O{sub 8} > U{sub 4}O{sub 9}/U{sub 3}O{sub 7} Almost-Equal-To UO{sub 2} > URu{sub 2}Si{sub 2}.

  5. 300 Area Uranium Stabilization Through Polyphosphate Injection: Final Report

    SciTech Connect (OSTI)

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

    2009-06-30

    The objective of the treatability test was to evaluate the efficacy of using polyphosphate injections to treat uranium-contaminated groundwater in situ. A test site consisting of an injection well and 15 monitoring wells was installed in the 300 Area near the process trenches that had previously received uranium-bearing effluents. This report summarizes the work on the polyphosphate injection project, including bench-scale laboratory studies, a field injection test, and the subsequent analysis and interpretation of the results. Previous laboratory tests have demonstrated that when a soluble form of polyphosphate is injected into uranium-bearing saturated porous media, immobilization of uranium occurs due to formation of an insoluble uranyl phosphate, autunite [Ca(UO2)2(PO4)2•nH2O]. These tests were conducted at conditions expected for the aquifer and used Hanford soils and groundwater containing very low concentrations of uranium (10-6 M). Because autunite sequesters uranium in the oxidized form U(VI) rather than forcing reduction to U(IV), the possibility of re-oxidation and subsequent re-mobilization is negated. Extensive testing demonstrated the very low solubility and slow dissolution kinetics of autunite. In addition to autunite, excess phosphorous may result in apatite mineral formation, which provides a long-term source of treatment capacity. Phosphate arrival response data indicate that, under site conditions, the polyphosphate amendment could be effectively distributed over a relatively large lateral extent, with wells located at a radial distance of 23 m (75 ft) reaching from between 40% and 60% of the injection concentration. Given these phosphate transport characteristics, direct treatment of uranium through the formation of uranyl-phosphate mineral phases (i.e., autunite) could likely be effectively implemented at full field scale. However, formation of calcium-phosphate mineral phases using the selected three-phase approach was problematic. Although amendment arrival response data indicate some degree of overlap between the reactive species and thus potential for the formation of calcium-phosphate mineral phases (i.e., apatite formation), the efficiency of this treatment approach was relatively poor. In general, uranium performance monitoring results support the hypothesis that limited long-term treatment capacity (i.e., apatite formation) was established during the injection test. Two separate overarching issues affect the efficacy of apatite remediation for uranium sequestration within the 300 Area: 1) the efficacy of apatite for sequestering uranium under the present geochemical and hydrodynamic conditions, and 2) the formation and emplacement of apatite via polyphosphate technology. In addition, the long-term stability of uranium sequestered via apatite is dependent on the chemical speciation of uranium, surface speciation of apatite, and the mechanism of retention, which is highly susceptible to dynamic geochemical conditions. It was expected that uranium sequestration in the presence of hydroxyapatite would occur by sorption and/or surface complexation until all surface sites have been depleted, but that the high carbonate concentrations in the 300 Area would act to inhibit the transformation of sorbed uranium to chernikovite and/or autunite. Adsorption of uranium by apatite was never considered a viable approach for in situ uranium sequestration in and of itself, because by definition, this is a reversible reaction. The efficacy of uranium sequestration by apatite assumes that the adsorbed uranium would subsequently convert to autunite, or other stable uranium phases. Because this appears to not be the case in the 300 Area aquifer, even in locations near the river, apatite may have limited efficacy for the retention and long-term immobilization of uranium at the 300 Area site..

  6. Carbon Dioxide Utilization Summit

    Broader source: Energy.gov [DOE]

    The 6th Carbon Dioxide Utilization Summit will be held in Newark, New Jersey, from Feb. 24–26, 2016. The conference will look at the benefits and challenges of carbon dioxide utilization. Advanced Algal Systems Program Manager Alison Goss Eng and Technology Manager Devinn Lambert will be in attendance. Dr. Goss Eng will be chairing a round table on Fuels and Chemicals during the Carbon Dioxide Utilization: From R&D to Commercialization discussion session.

  7. ARM - Carbon Dioxide

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

    Carbon Dioxide Outreach Home Room News Publications Traditional Knowledge Kiosks Barrow, Alaska Tropical Western Pacific Site Tours Contacts Students Study Hall About ARM Global Warming FAQ Just for Fun Meet our Friends Cool Sites Teachers Teachers' Toolbox Lesson Plans Carbon Dioxide Atmospheric concentrations of carbon dioxide have ranged from 200 to 280 ppm over the last 160,000 years. During the 1,000 years before the industrial revolution, in a time of stable global climate, the range was

  8. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

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

    1959-02-10

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

  9. PRODUCTION OF PURIFIED URANIUM

    DOE Patents [OSTI]

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

    1960-01-26

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

  10. METHOD FOR PURIFYING URANIUM

    DOE Patents [OSTI]

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

    1960-04-26

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

  11. Uranium Marketing Annual Report -

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

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

  12. Uranium Marketing Annual Report -

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

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

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

    SciTech Connect (OSTI)

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

    1993-08-01

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

  14. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  15. Future Sulfur Dioxide Emissions

    SciTech Connect (OSTI)

    Smith, Steven J.; Pitcher, Hugh M.; Wigley, Tom M.

    2005-12-01

    The importance of sulfur dioxide emissions for climate change is now established, although substantial uncertainties remain. This paper presents projections for future sulfur dioxide emissions using the MiniCAM integrated assessment model. A new income-based parameterization for future sulfur dioxide emissions controls is developed based on purchasing power parity (PPP) income estimates and historical trends related to the implementation of sulfur emissions limitations. This parameterization is then used to produce sulfur dioxide emissions trajectories for the set of scenarios developed for the Special Report on Emission Scenarios (SRES). We use the SRES methodology to produce harmonized SRES scenarios using the latest version of the MiniCAM model. The implications, and requirements, for IA modeling of sulfur dioxide emissions are discussed. We find that sulfur emissions eventually decline over the next century under a wide set of assumptions. These emission reductions result from a combination of emission controls, the adoption of advanced electric technologies, and a shift away from the direct end use of coal with increasing income levels. Only under a scenario where incomes in developing regions increase slowly do global emission levels remain at close to present levels over the next century. Under a climate policy that limits emissions of carbon dioxide, sulfur dioxide emissions fall in a relatively narrow range. In all cases, the relative climatic effect of sulfur dioxide emissions decreases dramatically to a point where sulfur dioxide is only a minor component of climate forcing by the end of the century. Ecological effects of sulfur dioxide, however, could be significant in some developing regions for many decades to come.

  16. Validation of MCNP with X6.XS cross-section set on the SUN Sparc Station 1+ computer for nominally 5 weight percent {sup 235}U enriched uranium systems

    SciTech Connect (OSTI)

    Lewis, K.D.

    1994-09-01

    The national Atomic Vapor Laser Isotope Separation (AVLIS) project has conducted extensive nuclear criticality safety analyses both in the design of Uranium Demonstration System (UDS) equipment and in AVLIS plant design/plant deployment activities. Currently, the design limit of an AVLIS plant calls for uranium product enriched in {sup 235}U to 5 wt %. Since an objective of an AVLIS plant is to deliver its product in a form readily usable by customers, uranium enriched in {sup 235}U will appear in a variety of forms, including metallic; as oxides, e.g., UO{sub 2}, UO{sub 3}; as fluorides, e.g., UF{sub 6}, UF{sub 4}, UO{sub 2}F{sub 2}; as nitrates or nitrides, e.g., UO{sub 2} (NO{sub 3}){sub 2}; and perhaps as uranium salts mixed with hydrocarbons such as oil. A wide range of neutron moderation levels, ranging from zero to optimal, and beyond can also be anticipated in an AVLIS plant, because of decontamination and cleaning activities and other wet chemistry processes that may be required.

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

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

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

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

    SciTech Connect (OSTI)

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

    2009-04-01

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

  19. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Yeager, J.H.

    1958-08-12

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

  20. A Uranium Bioremediation Reactive Transport Benchmark

    SciTech Connect (OSTI)

    Yabusaki, Steven B.; Sengor, Sevinc; Fang, Yilin

    2015-06-01

    A reactive transport benchmark problem set has been developed based on in situ uranium bio-immobilization experiments that have been performed at a former uranium mill tailings site in Rifle, Colorado, USA. Acetate-amended groundwater stimulates indigenous microorganisms to catalyze the reduction of U(VI) to a sparingly soluble U(IV) mineral. The interplay between the flow, acetate loading periods and rates, microbially-mediated and geochemical reactions leads to dynamic behavior in metal- and sulfate-reducing bacteria, pH, alkalinity, and reactive mineral surfaces. The benchmark is based on an 8.5 m long one-dimensional model domain with constant saturated flow and uniform porosity. The 159-day simulation introduces acetate and bromide through the upgradient boundary in 14-day and 85-day pulses separated by a 10 day interruption. Acetate loading is tripled during the second pulse, which is followed by a 50 day recovery period. Terminal electron accepting processes for goethite, phyllosilicate Fe(III), U(VI), and sulfate are modeled using Monod-type rate laws. Major ion geochemistry modeled includes mineral reactions, as well as aqueous and surface complexation reactions for UO2++, Fe++, and H+. In addition to the dynamics imparted by the transport of the acetate pulses, U(VI) behavior involves the interplay between bioreduction, which is dependent on acetate availability, and speciation-controlled surface complexation, which is dependent on pH, alkalinity and available surface complexation sites. The general difficulty of this benchmark is the large number of reactions (74), multiple rate law formulations, a multisite uranium surface complexation model, and the strong interdependency and sensitivity of the reaction processes. Results are presented for three simulators: HYDROGEOCHEM, PHT3D, and PHREEQC.

  1. Uranium Oxide Aerosol Transport in Porous Graphite

    SciTech Connect (OSTI)

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

    2012-01-23

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

  2. Method for converting uranium oxides to uranium metal

    DOE Patents [OSTI]

    Duerksen, Walter K. (Norris, TN)

    1988-01-01

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

  3. Innovative Elution Processes for Recovering Uranium from Seawater

    SciTech Connect (OSTI)

    Wai, Chien; Tian, Guoxin; Janke, Christopher

    2014-05-29

    Utilizing amidoxime-based polymer sorbents for extraction of uranium from seawater has attracted considerable interest in recent years. Uranium collected in the sorbent is recovered typically by elution with an acid. One drawback of acid elution is deterioration of the sorbent which is a significant factor that limits the economic competitiveness of the amidoxime-based sorbent systems for sequestering uranium from seawater. Developing innovative elution processes to improve efficiency and to minimize loss of sorbent capacity become essential in order to make this technology economically feasible for large-scale industrial applications. This project has evaluated several elution processes including acid elution, carbonate elution, and supercritical fluid elution for recovering uranium from amidoxime-based polymer sorbents. The elution efficiency, durability and sorbent regeneration for repeated uranium adsorption- desorption cycles in simulated seawater have been studied. Spectroscopic techniques are used to evaluate chemical nature of the sorbent before and after elution. A sodium carbonate-hydrogen peroxide elution process for effective removal of uranium from amidoxime-based sorbent is developed. The cause of this sodium carbonate and hydrogen peroxide synergistic leaching of uranium from amidoxime-based sorbent is attributed to the formation of an extremely stable uranyl peroxo-carbonato complex. The efficiency of uranium elution by the carbonate-hydrogen peroxide method is comparable to that of the hydrochloric acid elution but damage to the sorbent material is much less for the former. The carbonate- hydrogen peroxide elution also does not need any elaborate step to regenerate the sorbent as those required for hydrochloric acid leaching. Several CO2-soluble ligands have been tested for extraction of uranium from the sorbent in supercritical fluid carbon dioxide. A mixture of hexafluoroacetylacetone and tri-n-butylphosphate shows the best result but uranium removal from the sorbent reaches only 80% after 10 hours of leaching. Some information regarding coordination of vanadium with amidoxime molecules and elution of vanadium from amidoxime- based sorbents is also given in the report.

  4. Dispersion of UO{sub 2}F{sub 2} aerosol and HF vapor in the operating floor during winter ventilation at the Paducah Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Kim, S.H.; Chen, N.C.J.; Taleyarkhan, R.P.; Keith, K.D.; Schmidt, R.W.; Carter, J.C.

    1996-12-30

    The gaseous diffusion process is currently employed at two plants in the US: the Paducah Gaseous Diffusion Plant and the Portsmouth Gaseous Diffusion Plant. As part of a facility-wide safety evaluation, a postulated design basis accident involving large line-rupture induced releases of uranium hexafluoride (UF{sub 6}) into the process building of a gaseous diffusion plant (GDP) is evaluated. When UF{sub 6} is released into the atmosphere, it undergoes an exothermic chemical reaction with moisture (H{sub 2}O) in the air to form vaporized hydrogen fluoride (HF) and aerosolized uranyl fluoride (UO{sub 2}F{sub 2}). These reactants disperse in the process building and transport through the building ventilation system. The ventilation system draws outside air into the process building, distributes it evenly throughout the building, and discharges it to the atmosphere at an elevated temperature. Since air is recirculated from the cell floor area to the operating floor, issues concerning in-building worker safety and evacuation need to be addressed. Therefore, the objective of this study is to evaluate the transport of HF vapor and UO{sub 2}F{sub 2} aerosols throughout the operating floor area following B-line break accident in the cell floor area.

  5. Carbon dioxide removal process

    DOE Patents [OSTI]

    Baker, Richard W.; Da Costa, Andre R.; Lokhandwala, Kaaeid A.

    2003-11-18

    A process and apparatus for separating carbon dioxide from gas, especially natural gas, that also contains C.sub.3+ hydrocarbons. The invention uses two or three membrane separation steps, optionally in conjunction with cooling/condensation under pressure, to yield a lighter, sweeter product natural gas stream, and/or a carbon dioxide stream of reinjection quality and/or a natural gas liquids (NGL) stream.

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

    SciTech Connect (OSTI)

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

    2013-02-01

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

  7. Influence of the Electronic Structure and Optical Properties of CeO2 and UO2 for Characterization with UV-Laser Assisted Atom Probe Tomography

    SciTech Connect (OSTI)

    Billy Valderrama; H.B. Henderson; C. Yablinsky; J. Gan; T.R. Allen; M.V. Manuel

    2015-09-01

    Oxide materials are used in numerous applications such as thermal barrier coatings, nuclear fuels, and electrical conductors and sensors, all applications where nanometer-scale stoichiometric changes can affect functional properties. Atom probe tomography can be used to characterize the precise chemical distribution of individual species and spatially quantify the oxygen to metal ratio at the nanometer scale. However, atom probe analysis of oxides can be accompanied by measurement artifacts caused by laser-material interactions. In this investigation, two technologically relevant oxide materials with the same crystal structure and an anion to cation ratio of 2.00, pure cerium oxide (CeO2) and uranium oxide (UO2) are studied. It was determined that electronic structure, optical properties, heat transfer properties, and oxide stability strongly affect their evaporation behavior, thus altering their measured stoichiometry, with thermal conductance and thermodynamic stability being strong factors.

  8. About the Uranium Mine Team | Department of Energy

    Energy Savers [EERE]

    Uranium Mine Team About the Uranium Mine Team Text coming

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

    SciTech Connect (OSTI)

    Lloyd, R.

    1980-10-01

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

  10. Preparation of uranium compounds

    DOE Patents [OSTI]

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

    2013-02-19

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

  11. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    S2. Uranium feed deliveries, enrichment services, and uranium loaded by owners and operators of U.S. civilian nuclear power reactors, 1994-2014 million pounds U3O8 equivalent million separative work units (SWU) Year Feed deliveries by owners and operators of U.S. civilian nuclear power reactors Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors U.S.-origin enrichment services purchased Foreign-origin enrichment services purchased Total purchased enrichment services

  12. 2014 Uranium Marketing Annual Report

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

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

  13. 2014 Uranium Marketing Annual Report

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

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

  14. METHOD OF ROLLING URANIUM

    DOE Patents [OSTI]

    Smith, C.S.

    1959-08-01

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

  15. Uranium Purchases Report

    Reports and Publications (EIA)

    1996-01-01

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

  16. highly enriched uranium

    National Nuclear Security Administration (NNSA)

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

  17. Domestic Uranium Production Report

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

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

  18. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  19. Uranium Marketing Annual Report -

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

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

  20. Uranium Marketing Annual Report -

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

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

  1. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  2. Uranium Marketing Annual Report -

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

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

  3. Uranium Marketing Annual Report -

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

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

  4. Uranium Marketing Annual Report -

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

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

  5. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  6. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  7. U.S.Uranium Reserves

    Gasoline and Diesel Fuel Update (EIA)

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

  8. 2014 Uranium Marketing Annual Report

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

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

  9. 2014 Domestic Uranium Production Report

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

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

  10. 2014 Uranium Marketing Annual Report

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

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

  11. 2014 Uranium Marketing Annual Report

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

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

  12. 2014 Uranium Marketing Annual Report

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

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

  13. 2014 Uranium Marketing Annual Report

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

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

  14. Carbon dioxide sensor

    DOE Patents [OSTI]

    Dutta, Prabir K. (Worthington, OH); Lee, Inhee (Columbus, OH); Akbar, Sheikh A. (Hilliard, OH)

    2011-11-15

    The present invention generally relates to carbon dioxide (CO.sub.2) sensors. In one embodiment, the present invention relates to a carbon dioxide (CO.sub.2) sensor that incorporates lithium phosphate (Li.sub.3PO.sub.4) as an electrolyte and sensing electrode comprising a combination of lithium carbonate (Li.sub.2CO.sub.3) and barium carbonate (BaCO.sub.3). In another embodiment, the present invention relates to a carbon dioxide (CO.sub.2) sensor has a reduced sensitivity to humidity due to a sensing electrode with a layered structure of lithium carbonate and barium carbonate. In still another embodiment, the present invention relates to a method of producing carbon dioxide (CO.sub.2) sensors having lithium phosphate (Li.sub.3PO.sub.4) as an electrolyte and sensing electrode comprising a combination of lithium carbonate (Li.sub.2CO.sub.3) and barium carbonate (BaCO.sub.3).

  15. PUREX/UO{sub 3} facilities deactivation lessons learned: History

    SciTech Connect (OSTI)

    Gerber, M.S.

    1997-11-25

    In May 1997, a historic deactivation project at the PUREX (Plutonium URanium EXtraction) facility at the Hanford Site in south-central Washington State concluded its activities (Figure ES-1). The project work was finished at $78 million under its original budget of $222.5 million, and 16 months ahead of schedule. Closely watched throughout the US Department of Energy (DOE) complex and by the US Department of Defense for the value of its lessons learned, the PUREX Deactivation Project has become the national model for the safe transition of contaminated facilities to shut down status.

  16. Derivation of effective fission gas diffusivities in UO2 from lower length scale simulations and implementation of fission gas diffusion models in BISON

    SciTech Connect (OSTI)

    Andersson, Anders David Ragnar; Pastore, Giovanni; Liu, Xiang-Yang; Perriot, Romain Thibault; Tonks, Michael; Stanek, Christopher Richard

    2014-11-07

    This report summarizes the development of new fission gas diffusion models from lower length scale simulations and assessment of these models in terms of annealing experiments and fission gas release simulations using the BISON fuel performance code. Based on the mechanisms established from density functional theory (DFT) and empirical potential calculations, continuum models for diffusion of xenon (Xe) in UO2 were derived for both intrinsic conditions and under irradiation. The importance of the large XeU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and stability. These models were implemented in the MARMOT phase field code, which is used to calculate effective Xe diffusivities for various irradiation conditions. The effective diffusivities were used in BISON to calculate fission gas release for a number of test cases. The results are assessed against experimental data and future directions for research are outlined based on the conclusions.

  17. India's Worsening Uranium Shortage

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-01-15

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

  18. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  19. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  20. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  1. Depleted uranium management alternatives

    SciTech Connect (OSTI)

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

    1994-08-01

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

  2. Near surface stoichiometry in UO2: A density functional theory study

    SciTech Connect (OSTI)

    Yu, Jianguo; Valderrama, Billy; Henderson, Hunter B.; Manuel, Michele V.; Allen, Todd

    2015-08-01

    The mechanisms of oxygen stoichiometry variation in UO2 at different temperature and oxygen partial pressure are important for understanding the dynamics of microstructure in these crystals. However, very limited experimental studies have been performed to understand the atomic structure of UO2 near surface and defect effects of near surface on stoichiometry in which the system can exchange atoms with the external reservoir. In this study, the near (110) surface relaxation and stoichiometry in UO2 have been studied with density functional theory (DFT) calculations. On the basis of the point-defect model (PDM), a general expression for the near surface stoichiometric variation is derived by using DFT total-energy calculations and atomistic thermodynamics, in an attempt to pin down the mechanisms of oxygen exchange between the gas environment and defected UO2. By using the derived expression, it is observed that, under poor oxygen conditions, the stoichiometry of near surface is switched from hyperstoichiometric at 300 K with a depth around 3 nm to near-stoichiometric at 1000 K and hypostoichiometric at 2000 K. Furthermore, at very poor oxygen concentrations and high temperatures, our results also suggest that the bulk of the UO2 prefers to be hypostoichiometric, although the surface is near-stoichiometric.

  3. Near surface stoichiometry in UO2: A density functional theory study

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

    Yu, Jianguo; Valderrama, Billy; Henderson, Hunter B.; Manuel, Michele V.; Allen, Todd

    2015-08-01

    The mechanisms of oxygen stoichiometry variation in UO2 at different temperature and oxygen partial pressure are important for understanding the dynamics of microstructure in these crystals. However, very limited experimental studies have been performed to understand the atomic structure of UO2 near surface and defect effects of near surface on stoichiometry in which the system can exchange atoms with the external reservoir. In this study, the near (110) surface relaxation and stoichiometry in UO2 have been studied with density functional theory (DFT) calculations. On the basis of the point-defect model (PDM), a general expression for the near surface stoichiometric variationmore » is derived by using DFT total-energy calculations and atomistic thermodynamics, in an attempt to pin down the mechanisms of oxygen exchange between the gas environment and defected UO2. By using the derived expression, it is observed that, under poor oxygen conditions, the stoichiometry of near surface is switched from hyperstoichiometric at 300 K with a depth around 3 nm to near-stoichiometric at 1000 K and hypostoichiometric at 2000 K. Furthermore, at very poor oxygen concentrations and high temperatures, our results also suggest that the bulk of the UO2 prefers to be hypostoichiometric, although the surface is near-stoichiometric.« less

  4. Domestic Uranium Production Report

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

    9. Summary production statistics of the U.S. uranium industry, 1993-2014 Year Exploration and development surface drilling (million feet) Exploration and development drilling expenditures 1 (million dollars) Mine production of uranium (million pounds U3O8) Uranium concentrate production (million pounds U3O8) Uranium concentrate shipments (million pounds U3O8) Employment (person-years) 1993 1.1 5.7 2.1 3.1 3.4 871 1994 0.7 1.1 2.5 3.4 6.3 980 1995 1.3 2.6 3.5 6.0 5.5 1,107 1996 3.0 7.2 4.7 6.3

  5. Uranium Marketing Annual Report -

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

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

  6. Uranium Marketing Annual Report -

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

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

  7. Uranium Marketing Annual Report -

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

    9. Contracted purchases of uranium by owners and operators of U.S. civilian nuclear power reactors, signed in 2014, by delivery year, 2015-24 thousand pounds U3O8 equivalent Year...

  8. Uranium Marketing Annual Report -

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

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

  9. Uranium Marketing Annual Report

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

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

  10. Uranium Marketing Annual Report -

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

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

  11. Uranium Marketing Annual Report -

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

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

  12. Uranium Marketing Annual Report -

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

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

  13. supercritical carbon dioxide

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

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

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

    DOE Patents [OSTI]

    Kreuzmann, A.B.

    1982-10-27

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

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

    DOE Patents [OSTI]

    Kreuzmann, Alvin B. (Cincinnati, OH)

    1983-01-01

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

  16. Fission gas release from UO{sub 2+x} in defective light water reactor fuel rods

    SciTech Connect (OSTI)

    Skim, Y. S.

    1999-11-12

    A simplified semi-empirical model predicting fission gas release form UO{sub 2+x} fuel to the fuel rod plenum as a function of stoichiometry excess (x) is developed to apply to the fuel of a defective LWR fuel rod in operation. The effect of fuel oxidation in enhancing gas diffusion is included as a parabolic dependence of the stoichiometry excess. The increase of fission gas release in a defective BWR fuel rod is at the most 3 times higher than in an intact fuel rod because of small extent of UO{sub 2} oxidation. The major enhancement contributor in fission gas release of UO{sub 2+x} fuel is the increased diffusivity due to stoichiometry excess rather than the higher temperature caused by degraded fuel thermal conductivity.

  17. Criticality Safety Study of UF6and UO2F2in 8-in. Inner Diameter Piping

    SciTech Connect (OSTI)

    Elam, K.R.

    2003-10-07

    The purpose of this report is to provide an evaluation of the criticality safety aspects of using up to 8-in.-inner-diameter (ID) piping as part of a system to monitor the {sup 235}U enrichment in uranium hexafluoride (UF{sub 6}) gas both before and after an enrichment down-blending operation. The evaluated operation does not include the blending stage but includes only the monitors and the piping directly associated with the monitors, which are in a separate room from the blending operation. There are active controls in place to limit the enrichment of the blended UF{sub 6} to a maximum of 5 weight percent (wt%) {sup 235}U. Under normal operating conditions of temperature and pressure, the UF{sub 6} will stay in the gas phase and criticality will not be credible. The two accidents of concern are solidification of the UF{sub 6} along with some hydrofluoric acid (HF) and water or moisture ingress, which would cause the UF{sub 6} gas to react and form a hydrated uranyl fluoride (UO{sub 2}F{sub 2}) solid or solution. Of these two types of accidents, the addition of water and formation of UO{sub 2}F{sub 2} is the most reactive scenario and thus limits related to UO{sub 2}F{sub 2} will bound the limits related to UF{sub 6}. Two types of systems are included in the monitoring process. The first measures the enrichment of the approximately 90 wt% enriched UF{sub 6} before it is blended. This system uses a maximum 4-in.-(10.16-cm-) ID pipe, which is smaller than the 13.7-cm-cylinder-diameter subcritical limit for UO{sub 2}F{sub 2} solution of any enrichment as given in Table 1 of American National Standard ANSI/ANS-8.1.1 Therefore, this system poses no criticality concerns for either accident scenario. The second type of system includes two enrichment monitors for lower-enriched UF{sub 6}. One monitors the approximately 1.5 wt% enriched UF{sub 6} entering the blending process, and the second monitors the approximately 5 wt% enriched UF{sub 6} coming out of the blending process. Both use a maximum 8-in.-(20.32-cm-) ID piping, where the length of the larger ID piping is approximately 9.5 m. This diameter of piping is below the 26.6-cm-cylinder-diameter subcritical limit for 5 wt% enriched UO{sub 2}F{sub 2} solutions as given in Table 6 of ANSI/ANS-8.1. Therefore, for up to 5 wt% enriched UF{sub 6}, this piping does not present a criticality concern for either accident scenario. Calculations were performed to determine the enrichment level at which criticality could become a concern in these 8-in.-ID piping sections. Both unreflected and fully water-reflected conditions were considered.

  18. Emergent Properties of the Bose-Einstein-Hubbard Condensate in UO2(+x)

    Office of Scientific and Technical Information (OSTI)

    (Technical Report) | SciTech Connect Technical Report: Emergent Properties of the Bose-Einstein-Hubbard Condensate in UO2(+x) Citation Details In-Document Search Title: Emergent Properties of the Bose-Einstein-Hubbard Condensate in UO2(+x) Authors: Conradson, Steven D. [1] ; Durakiewicz, Tomasz [1] + Show Author Affiliations Los Alamos National Laboratory Publication Date: 2013-04-10 OSTI Identifier: 1073727 Report Number(s): LA-UR-13-22555 DOE Contract Number: AC52-06NA25396 Resource Type:

  19. Microstructure evolution in Xe-irradiated UO2 at room temperature

    SciTech Connect (OSTI)

    L.F. He; J. Pakarinen; M.A. Kirk; J. Gan; A.T. Nelson; X.-M. Bai; A. El-Azab; T.R. Allen

    2014-07-01

    In situ Transmission Electron Microscopy was conducted for single crystal UO2 to understand the microstructure evolution during 300 keV Xe irradiation at room temperature. The dislocation microstructure evolution was shown to occur as nucleation and growth of dislocation loops at low irradiation doses, followed by transformation to extended dislocation segments and tangles at higher doses. Xe bubbles with dimensions of 1-2 nm were observed after room-temperature irradiation. Electron Energy Loss Spectroscopy indicated that UO2 remained stoichiometric under room temperature Xe irradiation.

  20. Nuclear Fuel Facts: Uranium | Department of Energy

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

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

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

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

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

  2. Influence of uranium hydride oxidation on uranium metal behaviour

    SciTech Connect (OSTI)

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

    2013-07-01

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

  3. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1989-01-01

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

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

    SciTech Connect (OSTI)

    Stewart, B.D.; Amos, R.T.; Nico, P.S.; Fendorf, S.

    2010-03-15

    Uranium is a pollutant of concern to both human and ecosystem health. Uranium's redox state often dictates its partitioning between the aqueous- and solid-phases, and thus controls its dissolved concentration and, coupled with groundwater flow, its migration within the environment. In anaerobic environments, the more oxidized and mobile form of uranium (UO{sub 2}{sup 2+} and associated species) may be reduced, directly or indirectly, by microorganisms to U(IV) with subsequent precipitation of UO{sub 2}. However, various factors within soils and sediments may limit biological reduction of U(VI), inclusive of alterations in U(VI) speciation and competitive electron acceptors. Here we elucidate the impact of U(VI) speciation on the extent and rate of reduction with specific emphasis on speciation changes induced by dissolved Ca, and we examine the impact of Fe(III) (hydr)oxides (ferrihydrite, goethite and hematite) varying in free energies of formation on U reduction. The amount of uranium removed from solution during 100 h of incubation with S. putrefaciens was 77% with no Ca or ferrihydrite present but only 24% (with ferrihydrite) and 14% (no ferrihydrite) were removed for systems with 0.8 mM Ca. Imparting an important criterion on uranium reduction, goethite and hematite decrease the dissolved concentration of calcium through adsorption and thus tend to diminish the effect of calcium on uranium reduction. Dissimilatory reduction of Fe(III) and U(VI) can proceed through different enzyme pathways, even within a single organism, thus providing a potential second means by which Fe(III) bearing minerals may impact U(VI) reduction. We quantify rate coefficients for simultaneous dissimilatory reduction of Fe(III) and U(VI) in systems varying in Ca concentration (0 to 0.8 mM), and using a mathematical construct implemented with the reactive transport code MIN3P, we reveal the predominant influence of uranyl speciation, specifically the formation of uranyl-calcium-carbonato complexes, and ferrihydrite on the rate and extent of uranium reduction in complex geochemical systems.

  5. Uranium hexafluoride handling. Proceedings

    SciTech Connect (OSTI)

    Not Available

    1991-12-31

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

  6. Criticality Benchmark Analysis of Water-Reflected Uranium Oxyfluoride Slabs

    SciTech Connect (OSTI)

    Margaret A. Marshall; John D. Bess

    2009-11-01

    A series of twelve experiments were conducted in the mid 1950's at the Oak Ridge National Laboratory Critical Experiments Facility to determine the critical conditions of a semi-infinite water-reflected slab of aqueous uranium oxyfluoride (UO2F2). A different slab thickness was used for each experiment. Results from the twelve experiment recorded in the laboratory notebook were published in Reference 1. Seven of the twelve experiments were determined to be acceptable benchmark experiments for the inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. This evaluation will not only be available to handbook users for the validation of computer codes and integral cross-section data, but also for the reevaluation of experimental data used in the ANSI/ANS-8.1 standard. This evaluation is important as part of the technical basis of the subcritical slab limits in ANSI/ANS-8.1. The original publication of the experimental results was used for the determination of bias and bias uncertainties for subcritical slab limits, as documented by Hugh Clark's paper 'Subcritical Limits for Uranium-235 Systems'.

  7. Uranium-titanium-niobium alloy

    DOE Patents [OSTI]

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

    1990-01-01

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

  8. 2014 Uranium Marketing Annual Report

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

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

  9. 2014 Uranium Marketing Annual Report

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

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

  10. 2014 Uranium Marketing Annual Report

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

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

  11. 2014 Uranium Marketing Annual Report

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

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

  12. Process for sequestering carbon dioxide and sulfur dioxide

    DOE Patents [OSTI]

    Maroto-Valer, M. Mercedes (State College, PA); Zhang, Yinzhi (State College, PA); Kuchta, Matthew E. (State College, PA); Andresen, John M. (State College, PA); Fauth, Dan J. (Pittsburgh, PA)

    2009-10-20

    A process for sequestering carbon dioxide, which includes reacting a silicate based material with an acid to form a suspension, and combining the suspension with carbon dioxide to create active carbonation of the silicate-based material, and thereafter producing a metal salt, silica and regenerating the acid in the liquid phase of the suspension.

  13. 2014 Domestic Uranium Production Report

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

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

  14. Domestic Uranium Production Report

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

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

  15. PROCESS OF PREPARING URANIUM CARBIDE

    DOE Patents [OSTI]

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

    1964-03-24

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

  16. Uranium immobilization and nuclear waste

    SciTech Connect (OSTI)

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

    1982-02-01

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

  17. Carbon dioxide and climate

    SciTech Connect (OSTI)

    Not Available

    1990-10-01

    Scientific and public interest in greenhouse gases, climate warming, and global change virtually exploded in 1988. The Department's focused research on atmospheric CO{sub 2} contributed sound and timely scientific information to the many questions produced by the groundswell of interest and concern. Research projects summarized in this document provided the data base that made timely responses possible, and the contributions from participating scientists are genuinely appreciated. In the past year, the core CO{sub 2} research has continued to improve the scientific knowledge needed to project future atmospheric CO{sub 2} concentrations, to estimate climate sensitivity, and to assess the responses of vegetation to rising concentrations of CO{sub 2} and to climate change. The Carbon Dioxide Research Program's goal is to develop sound scientific information for policy formulation and governmental action in response to changes of atmospheric CO{sub 2}. The Program Summary describes projects funded by the Carbon Dioxide Research Program during FY 1990 and gives a brief overview of objectives, organization, and accomplishments.

  18. Thermal Behavior of Advanced UO{sub 2} Fuel at High Burnup

    SciTech Connect (OSTI)

    Muller, E.; Lambert, T.; Silberstein, K.; Therache, B.

    2007-07-01

    To improve the fuel performance, advanced UO{sub 2} products are developed to reduce significantly Pellet-Cladding Interaction and Fission Gas Release to increase high burnup safety margins on Light Water Reactors. To achieve the expected improvements, doping elements are currently used, to produce large grain viscoplastic UO{sub 2} fuel microstructures. In that scope, AREVA NP is conducting the qualification of a new UO{sub 2} fuel pellet obtained by optimum chromium oxide doping. To assess the fuel thermal performance, especially the fuel conductivity degradation with increasing burnup and also the kinetics of fission gas release under transient operating conditions, an instrumented in-pile experiment, called REMORA, has been developed by the CEA. One segment base irradiated for five cycles in a French EDF commercial PWR ({approx} 62 GWd/tM) was consequently re-instrumented with a fuel centerline thermocouple and an advanced pressure sensor. The design of this specific sensor is based on the counter-pressure principle and avoids any drift phenomenon due to nuclear irradiation. This rodlet was then irradiated in the GRIFFONOS rig of the Osiris experimental reactor at CEA Saclay. This device, located in the periphery of the core, is designed to perform test under conditions close to those prevailing in French PWR reactor. Power variations are carried out by translating the device relatively to the core. Self - powered neutron detectors are positioned in the loop in order to monitor the power the whole time of the irradiation. The re-irradiation of the REMORA experiment consisted of a stepped ramp to power in order to point out a potential degradation of the fuel thermal conductivity with increasing burnup. During the first part of the irradiation, most of the measurements were performed at low power in order to take into account the irradiation effects on UO{sub 2} thermal conductivity at high burnup in low range of temperature. The second part of the irradiation consisted in power cycling with one steady-state at several powers (290 W/cm and 360 W/cm) to assess both the thermal conductivity at higher temperature (until 1600 deg. C) and the fission gas release kinetic. This paper summarizes and discusses the main results assessed for this advanced UO{sub 2} fuel: on the one hand, the thermal performances indicate that the fuel thermal conductivity is similar to the one of the standard UO{sub 2} fuel type (the thermal conductivity damage under irradiation can be modelling alike) and, on the other hand, the test results show low fission gas release in comparison with UO{sub 2} standard fuel. (authors)

  19. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  20. Corrosion-resistant uranium

    DOE Patents [OSTI]

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

    1983-01-01

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

  1. Corrosion-resistant uranium

    DOE Patents [OSTI]

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

    1981-10-21

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

  2. Domestic Uranium Production Report

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

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

  3. Theoretical investigation of the impact of grain boundaries and fission gases on UO2 thermal conductivity

    SciTech Connect (OSTI)

    Du, Shiyu; Andersson, Anders D.; Germann, Timothy C.; Stanek, Christopher R.

    2012-05-02

    Thermal conductivity is one of the most important metrics of nuclear fuel performance. Therefore, it is crucial to understand the impact of microstructure features on thermal conductivity, especially since the microstructure evolves with burn-up or time in the reactor. For example, UO{sub 2} fuels are polycrystalline and for high-burnup fuels the outer parts of the pellet experience grain sub-division leading to a very fine grain structure. This is known to impact important physical properties such as thermal conductivity as fission gas release. In a previous study, we calculated the effect of different types of {Sigma}5 grain boundaries on UO{sub 2} thermal conductivity and predicted the corresponding Kapitza resistances, i.e. the resistance of the grain boundary in relation to the bulk thermal resistance. There have been reports of pseudoanisotropic effects for the thermal conductivity in cubic polycrystalline materials, as obtained from molecular dynamics simulations, which means that the conductivity appears to be a function of the crystallographic direction of the temperature gradient. However, materials with cubic symmetry should have isotropic thermal conductivity. For this reason it is necessary to determine the cause of this apparent anisotropy and in this report we investigate this effect in context of our earlier simulations of UO{sub 2} Kapitza resistances. Another source of thermal resistance comes from fission products and fission gases. Xe is the main fission gas and when generated in sufficient quantity it dissolves from the lattice and forms gas bubbles inside the crystalline structure. We have performed studies of how Xe atoms dissolved in the UO{sub 2} matrix or precipitated as bubbles impact thermal conductivity, both in bulk UO{sub 2} and in the presence of grain boundaries.

  4. Supercritical Carbon Dioxide / Reservoir Rock Chemical Interactions...

    Open Energy Info (EERE)

    Supercritical Carbon Dioxide Reservoir Rock Chemical Interactions Jump to: navigation, search Geothermal Lab Call Projects for Supercritical Carbon Dioxide Reservoir Rock...

  5. Optimize carbon dioxide sequestration, enhance oil recovery

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

    Optimize carbon dioxide sequestration, enhance oil recovery Optimize carbon dioxide sequestration, enhance oil recovery The simulation provides an important approach to estimate...

  6. High loading uranium fuel plate

    DOE Patents [OSTI]

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

    1990-01-01

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

  7. Uranium Reduction by Clostridia

    SciTech Connect (OSTI)

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

    2006-04-05

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

  8. Method of preparation of uranium nitride

    DOE Patents [OSTI]

    Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

    2013-07-09

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

  9. Removing oxygen from a solvent extractant in an uranium recovery process

    DOE Patents [OSTI]

    Hurst, Fred J. (Oak Ridge, TN); Brown, Gilbert M. (Knoxville, TN); Posey, Franz A. (Concord, TN)

    1984-01-01

    An improvement in effecting uranium recovery from phosphoric acid solutions is provided by sparging dissolved oxygen contained in solutions and solvents used in a reductive stripping stage with an effective volume of a nonoxidizing gas before the introduction of the solutions and solvents into the stage. Effective volumes of nonoxidizing gases, selected from the group consisting of argon, carbon dioxide, carbon monoxide, helium, hydrogen, nitrogen, sulfur dioxide, and mixtures thereof, displace oxygen from the solutions and solvents thereby reduce deleterious effects of oxygen such as excessive consumption of elemental or ferrous and accumulation of complex iron phosphates or cruds.

  10. Method for oxygen reduction in a uranium-recovery process. [US DOE patent application

    DOE Patents [OSTI]

    Hurst, F.J.; Brown, G.M.; Posey, F.A.

    1981-11-04

    An improvement in effecting uranium recovery from phosphoric acid solutions is provided by sparging dissolved oxygen contained in solutions and solvents used in a reductive stripping stage with an effective volume of a nonoxidizing gas before the introduction of the solutions and solvents into the stage. Effective volumes of nonoxidizing gases, selected from the group consisting of argon, carbon dioxide, carbon monoxide, helium, hydrogen, nitrogen, sulfur dioxide, and mixtures thereof, displace oxygen from the solutions and solvents thereby reduce deleterious effects of oxygen such as excessive consumption of elemental or ferrous iron and accumulation of complex iron phosphates or cruds.

  11. Thermal transport in UO2 with defects and fission products by molecular dynamics simulations

    SciTech Connect (OSTI)

    Liu, Xiang-Yang; Cooper, Michael William Donald; Mcclellan, Kenneth James; Lashley, Jason Charles; Byler, Darrin David; Stanek, Christopher Richard; Andersson, Anders David Ragnar

    2015-10-14

    The importance of the thermal transport in nuclear fuel has motivated a wide range of experimental and modelling studies. In this report, the reduction of thermal transport in UO2 due to defects and fission products has been investigated using non-equilibrium MD simulations, with two sets of empirical potentials for studying the degregation of UO2 thermal conductivity including a Buckingham type interatomic potential and a recently developed EAM type interatomic potential. Additional parameters for U5+ and Zr4+ in UO2 have been developed for the EAM potential. The thermal conductivity results from MD simulations are then corrected for the spin-phonon scattering through Callaway model formulations. To validate the modelling results, comparison was made with experimental measurements on single crystal hyper-stoichiometric UO2+x samples.

  12. Reducing carbon dioxide to products

    DOE Patents [OSTI]

    Cole, Emily Barton; Sivasankar, Narayanappa; Parajuli, Rishi; Keets, Kate A

    2014-09-30

    A method reducing carbon dioxide to one or more products may include steps (A) to (C). Step (A) may bubble said carbon dioxide into a solution of an electrolyte and a catalyst in a divided electrochemical cell. The divided electrochemical cell may include an anode in a first cell compartment and a cathode in a second cell compartment. The cathode may reduce said carbon dioxide into said products. Step (B) may adjust one or more of (a) a cathode material, (b) a surface morphology of said cathode, (c) said electrolyte, (d) a manner in which said carbon dioxide is bubbled, (e), a pH level of said solution, and (f) an electrical potential of said divided electrochemical cell, to vary at least one of (i) which of said products is produced and (ii) a faradaic yield of said products. Step (C) may separate said products from said solution.

  13. Method for fabricating uranium foils and uranium alloy foils

    DOE Patents [OSTI]

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

    2006-09-05

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

  14. Recuperative supercritical carbon dioxide cycle

    DOE Patents [OSTI]

    Sonwane, Chandrashekhar; Sprouse, Kenneth M; Subbaraman, Ganesan; O'Connor, George M; Johnson, Gregory A

    2014-11-18

    A power plant includes a closed loop, supercritical carbon dioxide system (CLS-CO.sub.2 system). The CLS-CO.sub.2 system includes a turbine-generator and a high temperature recuperator (HTR) that is arranged to receive expanded carbon dioxide from the turbine-generator. The HTR includes a plurality of heat exchangers that define respective heat exchange areas. At least two of the heat exchangers have different heat exchange areas.

  15. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  16. recycled_uranium.cdr

    Office of Legacy Management (LM)

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

  17. ELECTROLYSIS OF THORIUM AND URANIUM

    DOE Patents [OSTI]

    Hansen, W.N.

    1960-09-01

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

  18. $sup 18$O enrichment process in UO$sub 2$F$sub 2$ utilizing laser light

    DOE Patents [OSTI]

    DePoorter, G.L.; Rofer-DePoorter, C.K.

    1975-12-01

    Photochemical reaction induced by laser light is employed to separate oxygen isotopes. A solution containing UO$sub 2$F$sub 2$, HF, H$sub 2$O and a large excess of CH$sub 3$OH is irradiated with laser light of appropriate wavelength to differentially excite the UO$sub 2$$sup 2+$ ions containing $sup 16$O atoms and cause a reaction to proceed in accordance with the reaction 2 UO$sub 2$F$sub 2$ + CH$sub 3$OH + 4 HF $Yields$ 2 UF$sub 4$ down arrow + HCOOH + 3 H$sub 2$O. Irradiation is discontinued when about 10 percent of the UO$sub 2$F$sub 2$ has reacted, the UF$sub 4$ is filtered from the reaction mixture and the residual CH$sub 3$OH and HF plus the product HCOOH and H$sub 2$O are distilled away from the UO$sub 2$F$sub 2$ which is thereby enriched in the $sup 18$O isotope, or the solution containing the UO$sub 2$F$sub 2$ may be photochemically processed again to provide further enrichment in the $sup 18$O isotope.

  19. Recovery of UO[sub 2]/PuO[sub 2] in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1994-10-18

    A process is described for converting PuO[sub 2] and UO[sub 2] present in an electrorefiner to the chlorides, by contacting the PuO[sub 2] and UO[sub 2] with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the UO[sub 2] and PuO[sub 2] to metals while converting Li metal to Li[sub 2]O. Li[sub 2]O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting O[sub 2] out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li[sub 2]O to disassociate to O[sub 2] and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl[sub 2].

  20. Recovery of UO.sub.2 /Pu O.sub.2 in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Zygmunt (Lockport, IL); Miller, William E. (Naperville, IL)

    1994-01-01

    A process for converting PuO.sub.2 and UO.sub.2 present in an electrorefiner to the chlorides, by contacting the PuO.sub.2 and UO.sub.2 with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the UO.sub.2 and PuO.sub.2 to metals while converting Li metal to Li.sub.2 O. Li.sub.2 O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting O.sub.2 out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li.sub.2 O to disassociate to O.sub.2 and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl.sub.2.

  1. VANE Uranium One JV | Open Energy Information

    Open Energy Info (EERE)

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

  2. SEPARATION OF THORIUM FROM URANIUM

    DOE Patents [OSTI]

    Bane, R.W.

    1959-09-01

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

  3. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  4. 2014 Domestic Uranium Production Report

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

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

  5. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

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

  6. 2014 Domestic Uranium Production Report

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

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

  7. 2014 Domestic Uranium Production Report

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

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

  8. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

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

  9. 2014 Uranium Marketing Annual Survey

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

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

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

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-01-20

    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.

  11. Determination of the Relative Amount of Fluorine in Uranium Oxyfluoride Particles using Secondary Ion Mass Spectrometry and Optical Spectroscopy

    SciTech Connect (OSTI)

    Kips, R; Kristo, M J; Hutcheon, I D; Amonette, J; Wang, Z; Johnson, T; Gerlach, D; Olsen, K B

    2009-05-29

    Both nuclear forensics and environmental sampling depend upon laboratory analysis of nuclear material that has often been exposed to the environment after it has been produced. It is therefore important to understand how those environmental conditions might have changed the chemical composition of the material over time, particularly for chemically sensitive compounds. In the specific case of uranium enrichment facilities, uranium-bearing particles stem from small releases of uranium hexafluoride, a highly reactive gas that hydrolyzes upon contact with moisture from the air to form uranium oxyfluoride (UO{sub 2}F{sub 2}) particles. The uranium isotopic composition of those particles is used by the International Atomic Energy Agency (IAEA) to verify whether a facility is compliant with its declarations. The present study, however, aims to demonstrate how knowledge of time-dependent changes in chemical composition, particle morphology and molecular structure can contribute to an even more reliable interpretation of the analytical results. We prepared a set of uranium oxyfluoride particles at the Institute for Reference Materials and Measurements (IRMM, European Commission, Belgium) and followed changes in their composition, morphology and structure with time to see if we could use these properties to place boundaries on the particle exposure time in the environment. Because the rate of change is affected by exposure to UV-light, humidity levels and elevated temperatures, the samples were subjected to varying conditions of those three parameters. The NanoSIMS at LLNL was found to be the optimal tool to measure the relative amount of fluorine in individual uranium oxyfluoride particles. At PNNL, cryogenic laser-induced time-resolved U(VI) fluorescence microspectroscopy (CLIFS) was used to monitor changes in the molecular structure.

  12. Calculating Atomic Number Densities for Uranium

    Energy Science and Technology Software Center (OSTI)

    1993-01-01

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

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

    Office of Scientific and Technical Information (OSTI)

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

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

    Office of Scientific and Technical Information (OSTI)

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

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

    Office of Scientific and Technical Information (OSTI)

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

  16. Uranium Resources Inc URI | Open Energy Information

    Open Energy Info (EERE)

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

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

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

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

  18. Fluorination behavior of UO{sub 2}F{sub 2} in the presence of F{sub 2} and O{sub 2}

    SciTech Connect (OSTI)

    Matsuda, Minoru; Sato, Nobuaki; Kirishima, Akira; Tochiyama, Osamu

    2007-07-01

    To apply the fluoride volatility process to the spent nuclear fuel, fluorination of UO{sub 2} by fluorine has been studied. In this reaction, it is possible that the U-O-F compounds, such as UO{sub 2}F{sub 2}, are produced. Therefore, study of such compounds is useful in order to know the fluorination behavior of UO{sub 2}. This paper presents the fluorination behavior of UO{sub 2}F{sub 2} in the presence of F{sub 2} and O{sub 2}, analyzed by thermogravimetry and differential thermal analysis (TG-DTA) method using anti-corrosion type differential thermo-balance. In fluorine gas, exothermic peaks appeared and volatilization of UF{sub 6}. In oxygen gas, only slowly pace decomposition was measured from UO{sub 22} to UF{sub 6} and UO{sub 3}. (authors)

  19. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  20. Domestic Uranium Production Report

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

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

  1. Domestic Uranium Production Report

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

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

  2. Supercritical Fluid Extraction and Separation of Uranium from Other Actinides

    SciTech Connect (OSTI)

    Donna L. Quach; Bruce J. Mincher; Chien M. Wai

    2014-06-01

    This paper investigates the feasibility of separating uranium from other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of an extraction and counter current stripping technique, which would be a more efficient and environmentally benign technology for used nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U(VI), Np(VI), Pu(IV), and Am(III)) were extracted in sc-CO2 modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, the separation of uranium from plutonium in sc-CO2 modified with TBP was successful at nitric acid concentrations of less than 3 M in the presence of acetohydroxamic acid or oxalic acid, and the separation of uranium from neptunium was successful at nitric acid concentrations of less than 1 M in the presence of acetohydroxamic acid, oxalic acid, or sodium nitrite.

  3. Prospects for the recovery of uranium from seawater

    SciTech Connect (OSTI)

    Best, F.R.; Driscoll, M.

    1986-04-01

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

  4. Aspects of uranium chemistry pertaining to UF{sub 6} cylinder handling

    SciTech Connect (OSTI)

    Ritter, R.L.; Barber, E.J.

    1991-12-31

    Under normal conditions, the bulk of UF{sub 6} in storage cylinders will be in the solid state with an overpressure of gaseous UF{sub 6} well below one atmosphere. Corrosion of the interior of the cylinder will be very slow, with formation of a small amount of reduced fluoride, probably U{sub 2}F{sub 9}. The UO{sub 3}-HF-H{sub 2}O phase diagram indicates that reaction of any inleaking water vapor with the solid UF{sub 6} will generate the solid material [H{sub 3}O]{sub 2}(U(OH){sub 4}F{sub 4}) in equilibrium with an aqueous HF solution containing only small amounts of uranium. The corrosion of the steel cylinder by these materials may be enhanced over that observed with gaseous anhydrous UF{sub 6}.

  5. A roadmap to uranium ionic liquids: Anti-crystal engineering

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

    Yaprak, Damla; Spielberg, Eike T.; Bäcker, Tobias; Richter, Mark; Mallick, Bert; Klein, Axel; Mudring, Anja -Verena

    2014-04-15

    In the search for uranium-based ionic liquids, tris(N,N-dialkyldithiocarbamato)uranylates have been synthesized as salts of the 1-butyl-3-methylimidazolium (C4mim) cation. As dithiocarbamate ligands binding to the UO22+ unit, tetra-, penta-, hexa-, and heptamethylenedithiocarbamates, N,N-diethyldithiocarbamate, N-methyl-N-propyldithiocarbamate, N-ethyl-N-propyldithiocarbamate, and N-methyl-N-butyldithiocarbamate have been explored. X-ray single-crystal diffraction allowed unambiguous structural characterization of all compounds except N-methyl-N-butyldithiocarbamate, which is obtained as a glassy material only. In addition, powder X-ray diffraction as well as vibrational and UV/Vis spectroscopy, supported by computational methods, were used to characterize the products. Differential scanning calorimetry was employed to investigate the phase-transition behavior depending on the N,N-dialkyldithiocarbamato ligand with the aim tomore » establish structure–property relationships regarding the ionic liquid formation capability. Compounds with the least symmetric N,N-dialkyldithiocarbamato ligand and hence the least symmetric anions, tris(N-methyl-N-propyldithiocarbamato)uranylate, tris(N-ethyl-N-propyldithiocarbamato)uranylate, and tris(N-methyl-N-butyldithiocarbamato)uranylate, lead to the formation of (room-temperature) ionic liquids, which confirms that low-symmetry ions are indeed suitable to suppress crystallization. As a result, these materials combine low melting points, stable complex formation, and hydrophobicity and are therefore excellent candidates for nuclear fuel purification and recovery.« less

  6. Uranium Downblending and Disposition Project Technology Readiness

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

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

  7. PREPARATION OF URANIUM-ALUMINUM ALLOYS

    DOE Patents [OSTI]

    Moore, R.H.

    1962-09-01

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

  8. First-principles study of noble gas impurities and defects in UO{sub 2}

    SciTech Connect (OSTI)

    Thompson, Alexander E.; Wolverton, C.

    2011-10-01

    We performed a series of density functional theory + U (DFT + U) calculations to explore the energetics of various defects in UO{sub 2}, i.e., noble gases (He, Ne, Ar, Kr, Xe), Schottky defects, and the interaction between these defects. We found the following: (1) collinear antiferromagnetic UO{sub 2} has an energy-lowering distortion of the oxygen sublattice from ideal fluorite positions; (2) DFT + U qualitatively affects the formation volume of Schottky defect clusters in UO{sub 2} (without U the formation volume is negative, but including U the formation volume is positive); (3) the configuration of the Schottky defect cluster is dictated by a competition between electrostatic and surface energy effects; (4) the incorporation energy of inserting noble gas atoms into an interstitial site has a strong dependence on the volume of the noble gas atom, corresponding to the strain it causes in the interstitial site, from He (0.98 eV) to Xe (9.73 eV); (5) the energetics of each of the noble gas atoms incorporated in Schottky defects show strong favorable binding, due to strain relief associated with moving the noble gas atom from the highly strained interstitial position into the vacant space of the Schottky defect; and (6) for argon, krypton, and xenon, the binding energy of a noble gas impurity with the Schottky defect is larger than the formation energy of a Schottky defect, thereby making the formation of Schottky defects thermodynamically favorable in the presence of these large impurities.

  9. Solar and Photovoltaic Data from the University of Oregon Solar Radiation Monitoring Laboratory (UO SRML)

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    The UO SRML is a regional solar radiation data center whose goal is to provide sound solar resource data for planning, design, deployment, and operation of solar electric facilities in the Pacific Northwest. The laboratory has been in operation since 1975. Solar data includes solar resource maps, cumulative summary data, daily totals, monthly averages, single element profile data, parsed TMY2 data, and select multifilter radiometer data. A data plotting program and other software tools are also provided. Shade analysis information and contour plots showing the effect of tilt and orientation on annual solar electric system perfomance make up a large part of the photovoltaics data.(Specialized Interface)

  10. An Overview of Current and Past W-UO[2] CERMET Fuel Fabrication Technology

    SciTech Connect (OSTI)

    Douglas E. Burkes; Daniel M. Wachs; James E. Werner; Steven D. Howe

    2007-06-01

    Studies dating back to the late 1940s performed by a number of different organizations and laboratories have established the major advantages of Nuclear Thermal Propulsion (NTP) systems, particularly for manned missions. A number of NTP projects have been initiated since this time; none have had any sustained fuel development work that appreciably contributed to fuel fabrication or performance data from this era. As interest in these missions returns and previous space nuclear power researchers begin to retire, fuel fabrication technologies must be revisited, so that established technologies can be transferred to young researchers seamlessly and updated, more advanced processes can be employed to develop successful NTP fuels. CERMET fuels, specifically W-UO2, are of particular interest to the next generation NTP plans since these fuels have shown significant advantages over other fuel types, such as relatively high burnup, no significant failures under severe transient conditions, capability of accommodating a large fission product inventory during irradiation and compatibility with flowing hot hydrogen. Examples of previous fabrication routes involved with CERMET fuels include hot isostatic pressing (HIPing) and press and sinter, whereas newer technologies, such as spark plasma sintering, combustion synthesis and microsphere fabrication might be well suited to produce high quality, effective fuel elements. These advanced technologies may address common issues with CERMET fuels, such as grain growth, ductile to brittle transition temperature and UO2 stoichiometry, more effectively than the commonly accepted ‘traditional’ fabrication routes. Bonding of fuel elements, especially if the fabrication process demands production of smaller element segments, must be investigated. Advanced brazing techniques and compounds are now available that could produce a higher quality bond segment with increased ease in joining. This paper will briefly address the history of CERMET fuel fabrication technology as related to the GE 710 and ANL Nuclear Rocket Programs, in addition to discussing future plans, viable alternatives and preliminary investigations for W-UO2 CERMET fuel fabrication. The intention of the talk is to provide the brief history and tie in an overview of current programs and investigations as related to NTP based W-UO2 CERMET fuel fabrication, and hopefully peak interest in advanced fuel fabrication technologies.

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

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

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

  12. Uranium hexafluoride bibliography

    SciTech Connect (OSTI)

    Burnham, S.L.

    1988-01-01

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

  13. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  14. Domestic Uranium Production Report

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

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

  15. Domestic Uranium Production Report

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

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

  16. Domestic Uranium Production Report

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

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

  17. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  18. Y-12 and uranium history

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

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

  19. 2014 Uranium Marketing Annual Report

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

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

  20. 2014 Uranium Marketing Annual Report

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

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

  1. 2014 Uranium Marketing Annual Report

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

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

  2. 2014 Uranium Marketing Annual Report

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

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

  3. Synchrotron characterization of nanograined UO2 grain growth

    SciTech Connect (OSTI)

    Mo, Kun; Miao, Yinbin; Yun, Di; Jamison, Laura M.; Lian, Jie; Yao, Tiankei

    2015-09-30

    This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL) and aims at providing experimental data for the validation of the mesoscale simulation code MARMOT. MARMOT is a mesoscale multiphysics code that predicts the coevolution of microstructure and properties within reactor fuel during its lifetime in the reactor. It is an important component of the Moose-Bison-Marmot (MBM) code suite that has been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. In order to ensure the accuracy of the microstructure based materials models being developed within the MARMOT code, extensive validation efforts must be carried out. In this report, we summarize our preliminary synchrotron radiation experiments at APS to determine the grain size of nanograin UO2. The methodology and experimental setup developed in this experiment can directly apply to the proposed in-situ grain growth measurements. The investigation of the grain growth kinetics was conducted based on isothermal annealing and grain growth characterization as functions of duration and temperature. The kinetic parameters such as activation energy for grain growth for UO2 with different stoichiometry are obtained and compared with molecular dynamics (MD) simulations.

  4. Supplying materials needed for grain growth characterizations of nano-grained UO2

    SciTech Connect (OSTI)

    Mo, Kun; Miao, Yinbin; Yun, Di; Jamison, Laura M.; Lian, Jie; Yao, Tiankei

    2015-09-30

    This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL) and aims at providing experimental data for the validation of the mesoscale simulation code MARMOT. MARMOT is a mesoscale multiphysics code that predicts the coevolution of microstructure and properties within reactor fuel during its lifetime in the reactor. It is an important component of the Moose-Bison-Marmot (MBM) code suite that has been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. In order to ensure the accuracy of the microstructure based materials models being developed within the MARMOT code, extensive validation efforts must be carried out. In this report, we summarize our preliminary synchrotron radiation experiments at APS to determine the grain size of nanograin UO2. The methodology and experimental setup developed in this experiment can directly apply to the proposed in-situ grain growth measurements. The investigation of the grain growth kinetics was conducted based on isothermal annealing and grain growth characterization as functions of duration and temperature. The kinetic parameters such as activation energy for grain growth for UO2 with different stoichiometry are obtained and compared with molecular dynamics (MD) simulations.

  5. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, B.A.

    1983-06-10

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

  6. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  7. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, Bruce A. (Kennewick, WA)

    1986-01-01

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

  8. Beneficial Uses of Depleted Uranium

    SciTech Connect (OSTI)

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

    1997-08-01

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

  9. AREVA NP next generation fresh UO{sub 2} fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    SciTech Connect (OSTI)

    Doucet, M.; Landrieu, M.; Montgomery, R.; O' Donnell, B.

    2007-07-01

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO{sub 2} shipping casks being agreed within a lot of countries including USA, France, Germany, Belgium, Sweden, China, and South Africa - and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up-to-date shipping cask (so called MAP project) gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: - Preferential flooding; - Fuel rod lattice pitch expansion for full length of fuel assemblies; - Neutron absorber penalty; -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - Boral{sup TM} as neutron absorber; - High density polyethylene (HDPE) or Nylon as neutron moderator; - Foam as thermal and mechanical protection. The cask is designed to handle the complete AREVA NP fuel assembly types from the 14x14 to the 18x18 design with a {sup 235}U enrichment up to 5.0% enriched natural uranium (ENU) and enriched reprocessed uranium (ERU). After a brief presentation of the computer codes and the description of the shipping cask, calculation results and comparisons between SCALE and CRISTAL will be discussed. (authors)

  10. Method for carbon dioxide sequestration

    DOE Patents [OSTI]

    Wang, Yifeng; Bryan, Charles R.; Dewers, Thomas; Heath, Jason E.

    2015-09-22

    A method for geo-sequestration of a carbon dioxide includes selection of a target water-laden geological formation with low-permeability interbeds, providing an injection well into the formation and injecting supercritical carbon dioxide (SC--CO.sub.2) into the injection well under conditions of temperature, pressure and density selected to cause the fluid to enter the formation and splinter and/or form immobilized ganglia within the formation. This process allows for the immobilization of the injected SC--CO.sub.2 for very long times. The dispersal of scCO2 into small ganglia is accomplished by alternating injection of SC--CO.sub.2 and water. The injection rate is required to be high enough to ensure the SC--CO.sub.2 at the advancing front to be broken into pieces and small enough for immobilization through viscous instability.

  11. High capacity carbon dioxide sorbent

    DOE Patents [OSTI]

    Dietz, Steven Dean; Alptekin, Gokhan; Jayaraman, Ambalavanan

    2015-09-01

    The present invention provides a sorbent for the removal of carbon dioxide from gas streams, comprising: a CO.sub.2 capacity of at least 9 weight percent when measured at 22.degree. C. and 1 atmosphere; an H.sub.2O capacity of at most 15 weight percent when measured at 25.degree. C. and 1 atmosphere; and an isosteric heat of adsorption of from 5 to 8.5 kilocalories per mole of CO.sub.2. The invention also provides a carbon sorbent in a powder, a granular or a pellet form for the removal of carbon dioxide from gas streams, comprising: a carbon content of at least 90 weight percent; a nitrogen content of at least 1 weight percent; an oxygen content of at most 3 weight percent; a BET surface area from 50 to 2600 m.sup.2/g; and a DFT micropore volume from 0.04 to 0.8 cc/g.

  12. CARBON DIOXIDE AS A FEEDSTOCK.

    SciTech Connect (OSTI)

    CREUTZ,C.; FUJITA,E.

    2000-12-09

    This report is an overview on the subject of carbon dioxide as a starting material for organic syntheses of potential commercial interest and the utilization of carbon dioxide as a substrate for fuel production. It draws extensively on literature sources, particularly on the report of a 1999 Workshop on the subject of catalysis in carbon dioxide utilization, but with emphasis on systems of most interest to us. Atmospheric carbon dioxide is an abundant (750 billion tons in atmosphere), but dilute source of carbon (only 0.036 % by volume), so technologies for utilization at the production source are crucial for both sequestration and utilization. Sequestration--such as pumping CO{sub 2} into sea or the earth--is beyond the scope of this report, except where it overlaps utilization, for example in converting CO{sub 2} to polymers. But sequestration dominates current thinking on short term solutions to global warming, as should be clear from reports from this and other workshops. The 3500 million tons estimated to be added to the atmosphere annually at present can be compared to the 110 million tons used to produce chemicals, chiefly urea (75 million tons), salicylic acid, cyclic carbonates and polycarbonates. Increased utilization of CO{sub 2} as a starting material is, however, highly desirable, because it is an inexpensive, non-toxic starting material. There are ongoing efforts to replace phosgene as a starting material. Creation of new materials and markets for them will increase this utilization, producing an increasingly positive, albeit small impact on global CO{sub 2} levels. The other uses of interest are utilization as a solvent and for fuel production and these will be discussed in turn.

  13. Domestic Uranium Production Report

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

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

  14. NEAMS FPL M2 Milestone Report: Development of a UO? Grain Size Model using Multicale Modeling and Simulation

    SciTech Connect (OSTI)

    Michael R Tonks; Yongfeng Zhang; Xianming Bai

    2014-06-01

    This report summarizes development work funded by the Nuclear Energy Advanced Modeling Simulation program's Fuels Product Line (FPL) to develop a mechanistic model for the average grain size in UO? fuel. The model is developed using a multiscale modeling and simulation approach involving atomistic simulations, as well as mesoscale simulations using INL's MARMOT code.

  15. Inherently safe in situ uranium recovery

    DOE Patents [OSTI]

    Krumhansl, James L; Brady, Patrick V

    2014-04-29

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

  16. Uranium vacancy mobility at the Σ5 symmetric tilt and Σ5 twist grain boundaries in UO₂

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

    Uberuaga, Blas Pedro; Andersson, David A.

    2015-10-01

    Ionic transport at grain boundaries in oxides dictates a number of important phenomena, from ionic conductivity to sintering to creep. For nuclear fuels, it also influences fission gas bubble nucleation and growth. Here, using a combination of atomistic calculations and object kinetic Monte Carlo (okMC) simulations, we examine the kinetic pathways associated with uranium vacancies at two model grain boundaries in UO2. The barriers for vacancy motion were calculated using the nudged elastic band method at all uranium sites at each grain boundary and were used as the basis of the okMC simulations. For both boundaries considered – a simplemore » tilt and a simple twist boundary – the mobility of uranium vacancies is significantly higher than in the bulk. For the tilt boundary, there is clearly preferred migration along the tilt axis as opposed to in the perpendicular direction while, for the twist boundary, migration is essentially isotropic within the boundary plane. These results show that cation defect mobility in fluorite-structured materials is enhanced at certain types of grain boundaries and is dependent on the boundary structure with the tilt boundary exhibiting higher rates of migration than the twist boundary.« less

  17. Optimize carbon dioxide sequestration, enhance oil recovery

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

    Optimize carbon dioxide sequestration, enhance oil recovery Optimize carbon dioxide sequestration, enhance oil recovery The simulation provides an important approach to estimate the potential of storing carbon dioxide in depleted oil fields while simultaneously maximizing oil production. January 8, 2014 Schematic of a water-alternating-with-gas flood for CO2 sequestration and enhanced oil recovery. Schematic of a water-alternating-with-gas flood for CO2 sequestration and enhanced oil recovery.

  18. Optimize carbon dioxide sequestration, enhance oil recovery

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

    Optimize carbon dioxide sequestration, enhance oil recovery Optimize carbon dioxide sequestration, enhance oil recovery The simulation provides an important approach to estimate the potential of storing carbon dioxide in depleted oil fields while simultaneously maximizing oil production. January 8, 2014 Schematic of a water-alternating-with-gas flood for CO2 sequestration and enhanced oil recovery. Schematic of a water-alternating-with-gas flood for CO2 sequestration and enhanced oil recovery.

  19. How Atomic Vibrations Transform Vanadium Dioxide

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

    How Atomic Vibrations Transform Vanadium Dioxide How Atomic Vibrations Transform Vanadium Dioxide Calculations Confirm Material's Potential for Next-Generation Electronics, Energy November 10, 2014 Contact: Dawn Levy, levyd@ornl.gov, 865.576.6448 Budaivibe Vanadium atoms (blue) have unusually large thermal vibrations that stabilize the metallic state of a vanadium dioxide crystal. Red depicts oxygen atoms. Image credit: Oak Ridge National Laboratory For more than 50 years, scientists have

  20. Irradiation behaviour of the large grained UO{sub 2} fuel pellet in the transient conditions

    SciTech Connect (OSTI)

    Kosaka, Yuji; Watanabe, Seiichi; Arakawa, Yasushi

    2007-07-01

    In order to achieve a high duty fuel rod design, it is the key issue to suppress the fission gas release from the view point of the fuel rod inner pressure design. The large grain UO{sub 2} pellet is one of the candidates to meet such a requirement by reducing the fission gas release especially at high power and/or high burnup. We have demonstrated the fuel performance of the large grain pellet in the PWR irradiation conditions, which was fabricated with no additive but with active UO{sub 2} powder through the conventional pelletizing process for the normal grain size pellet. According to the mechanism of the fission gas retention, there may be a concern about the larger gas bubble swelling of the large grain pellet at the power transient conditions which may increase the potential of the PCMI failure. In this paper, we focus on the differences of the dimensional change in comparison among the pellets with the different grain sizes at the power transient conditions. The power ramp tests were carried out on the high burnup fuel rods of normal and large grain pellet with no additive, which had been irradiated in the PWR conditions up to around 60 GWd/t at peak position. The detailed PIE results revealed that the volume increment due to the power ramp clearly showed the dependence on the grain size as well as the fission gas release and suggested that the larger grain with no additive may suppress the gas bubble swelling at the power transient conditions. According to the experimental results, it is concluded that the large grain pellet with no additive does not deteriorate the irradiation performance during the power transient conditions from the view point of the gas bubble swelling. (authors)

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

    Office of Scientific and Technical Information (OSTI)

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

  2. Uranium Management and Policy | Department of Energy

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

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

  3. Project Profile: Direct Supercritical Carbon Dioxide Receiver...

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

    characterize, and experimentally demonstrate a novel high-temperature receiver technology using supercritical carbon dioxide (s-CO2) directly as the heat transfer fluid (HTF). ...

  4. ARM - Measurement - Carbon dioxide (CO2) flux

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

    carbon dioxide, a heavy, colorless greenhouse gas. Categories Atmospheric Carbon, Surface Properties Instruments The above measurement is considered scientifically relevant for the...

  5. Case Study: Transcritical Carbon Dioxide Supermarket Refrigeration...

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

    one year of operating experience with a transcritical carbon dioxide (TC CO2) booster refrigeration system at Delhaize America's Hannaford supermarket location in Turner, Maine. ...

  6. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  7. Domestic Uranium Production Report - Quarterly

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

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

  8. Continuous reduction of uranium tetrafluoride

    SciTech Connect (OSTI)

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

    1993-10-21

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

  9. High strength uranium-tungsten alloys

    DOE Patents [OSTI]

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

    1991-01-01

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

  10. High strength uranium-tungsten alloy process

    DOE Patents [OSTI]

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

    1990-01-01

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

  11. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  12. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  13. 2014 Uranium Market Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  14. METHOD OF PROTECTIVELY COATING URANIUM

    DOE Patents [OSTI]

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

    1959-02-01

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

  15. 2014 Domestic Uranium Production Report

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

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

  16. Domestic Uranium Production Report - Quarterly

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

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

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

    SciTech Connect (OSTI)

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

    2012-07-25

    Electron backscatter diffraction (EBSD) was used to examine the microstructures of unalloyed uranium, U-6Nb, U-10Mo, and U-0.75Ti. For unalloyed uranium, we used EBSD to examine the effects of various processes on microstructures including casting, rolling and forming, recrystallization, welding, and quasi-static and shock deformation. For U-6Nb we used EBSD to examine the microstructural evolution during shape memory loading. EBSD was used to study chemical homogenization in U-10Mo, and for U-0.75Ti, we used EBSD to study the microstructure and texture evolution during thermal cycling and deformation. The studied uranium alloys have significant microstructural and chemical differences and each of these alloys presents unique preparation challenges. Each of the alloys is prepared by a sequence of mechanical grinding and polishing followed by electropolishing with subtle differences between the alloys. U-6Nb and U-0.75Ti both have martensitic microstructures and both require special care in order to avoid mechanical polishing artifacts. Unalloyed uranium has a tendency to rapidly oxidize when exposed to air and a two-step electropolish is employed, the first step to remove the damaged surface layer resulting from the mechanical preparation and the second step to passivate the surface. All of the alloying additions provide a level of surface passivation and different one and two step electropolishes are employed to create good EBSD surfaces. Because of its low symmetry crystal structure, uranium exhibits complex deformation behavior including operation of multiple deformation twinning modes. EBSD was used to observe and quantify twinning contributions to deformation and to examine the fracture behavior. Figure 1 shows a cross section of two mating fracture surfaces in cast uranium showing the propensity of deformation twinning and intergranular fracture largely between dissimilarly oriented grains. Deformation of U-6Nb in the shape memory regime occurs by the motion of twin boundaries formed during the martensitic transformation. Deformation actually results in a coarsening of the microstructure making EBSD more practical following a limited amount of strain. Figure 2 shows the microstructure resulting from 6% compression. Casting of U-10Mo results in considerable chemical segregation as is apparent in Figure 2a. The segregation subsists through rolling and heat treatment processes as shown in Figure 2b. EBSD was used to study the effects of homogenization time and temperature on chemical heterogeneity. It was found that times and temperatures that result in a chemically homogeneous microstructure also result in a significant increase in grain size. U-0.75Ti forms an acicular martinsite as shown in Figure 4. This microstructure prevails through cycling into the higher temperature solid uranium phases.

  18. Gas Phase Uranyl Activation: Formation of a Uranium Nitrosyl Complex from Uranyl Azide

    SciTech Connect (OSTI)

    Gong, Yu; De Jong, Wibe A.; Gibson, John K.

    2015-05-13

    Activation of the oxo bond of uranyl, UO22+, was achieved by collision induced dissociation (CID) of UO2(N3)Cl2– in a quadrupole ion trap mass spectrometer. The gas phase complex UO2(N3)Cl2– was produced by electrospray ionization of solutions of UO2Cl2 and NaN3. CID of UO2(N3)Cl2– resulted in the loss of N2 to form UO(NO)Cl2–, in which the “inert” uranyl oxo bond has been activated. Formation of UO2Cl2– via N3 loss was also observed. Density functional theory computations predict that the UO(NO)Cl2– complex has nonplanar Cs symmetry and a singlet ground state. Analysis of the bonding of the UO(NO)Cl2– complex shows that the side-on bonded NO moiety can be considered as NO3–, suggesting a formal oxidation state of U(VI). Activation of the uranyl oxo bond in UO2(N3)Cl2– to form UO(NO)Cl2– and N2 was computed to be endothermic by 169 kJ/mol, which is energetically more favorable than formation of NUOCl2– and UO2Cl2–. The observation of UO2Cl2– during CID is most likely due to the absence of an energy barrier for neutral ligand loss.

  19. Reactions of aluminum with uranium fluorides and oxyfluorides

    SciTech Connect (OSTI)

    Leitnaker, J.M.; Nichols, R.W.; Lankford, B.S.

    1991-12-31

    Every 30 to 40 million operating hours a destructive reaction is observed in one of the {approximately}4000 large compressors that move UF{sub 6} through the gaseous diffusion plants. Despite its infrequency, such a reaction can be costly in terms of equipment and time. Laboratory experiments reveal that the presence of moderate pressures of UF{sub 6} actually cools heated aluminum, although thermodynamic calculations indicate the potential for a 3000-4000{degrees}C temperature rise. Within a narrow and rather low (<100 torr; 1 torr = 133.322 Pa) pressure range, however, the aluminum is seen to react with sufficient heat release to soften an alumina boat. Three things must occur in order for aluminum to react vigorously with either UF{sub 6} or UO{sub 2}F{sub 2}. 1. An initiating source of heat must be provided. In the compressors, this source can be friction, permitted by disruption of the balance of the large rotating part or by creep of the aluminum during a high-temperature treatment. In the absence of this heat source, compressors have operated for 40 years in UF{sub 6} without significant reaction. 2. The film protecting the aluminum must be breached. Melting (of UF{sub 5} at 620 K or aluminum at 930 K) can cause such a breach in laboratory experiments. In contrast, holding Al samples in UF{sub 6} at 870 K for several hours produces only moderate reaction. Rubbing in the cascade can undoubtedly breach the protective film. 3. Reaction products must not build up and smother the reaction. While uranium products tend to dissolve or dissipate in molten aluminum, AIF{sub 3} shows a remarkable tendency to surround and hence protect even molten aluminum. Hence the initial temperature rise must be rapid and sufficient to move reactants into a temperature region in which products are removed from the reaction site.

  20. Development of pulsed neutron uranium logging instrument

    SciTech Connect (OSTI)

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

    2015-03-15

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

  1. Process for alloying uranium and niobium

    DOE Patents [OSTI]

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

    1991-01-01

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

  2. Domestic Uranium Production Report - Energy Information Administration

    Gasoline and Diesel Fuel Update (EIA)

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

  3. In-Situ Measurements of Low Enrichment Uranium Holdup Process Gas Piping at K-25 - Paper for Waste Management Symposia 2010 East Tennessee Technology Park Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    Rasmussen B.

    2010-01-01

    This document is the final version of a paper submitted to the Waste Management Symposia, Phoenix, 2010, abstract BJC/OR-3280. The primary document from which this paper was condensed is In-Situ Measurement of Low Enrichment Uranium Holdup in Process Gas Piping at K-25 Using NaI/HMS4 Gamma Detection Systems, BJC/OR-3355. This work explores the sufficiency and limitations of the Holdup Measurement System 4 (HJVIS4) software algorithms applied to measurements of low enriched uranium holdup in gaseous diffusion process gas piping. HMS4 has been used extensively during the decommissioning and demolition project of the K-25 building for U-235 holdup quantification. The HMS4 software is an integral part of one of the primary nondestructive assay (NDA) systems which was successfully tested and qualified for holdup deposit quantification in the process gas piping of the K-25 building. The initial qualification focused on the measurement of highly enriched UO{sub 2}F{sub 2} deposits. The purpose of this work was to determine if that qualification could be extended to include the quantification of holdup in UO{sub 2}F{sub 2} deposits of lower enrichment. Sample field data are presented to provide evidence in support of the theoretical foundation. The HMS4 algorithms were investigated in detail and found to sufficiently compensate for UO{sub 2}F{sub 2} source self-attenuation effects, over the range of expected enrichment (4-40%), in the North and East Wings of the K-25 building. The limitations of the HMS4 algorithms were explored for a described set of conditions with respect to area source measurements of low enriched UO{sub 2}F{sub 2} deposits when used in conjunction with a 1 inch by 1/2 inch sodium iodide (NaI) scintillation detector. The theoretical limitations of HMS4, based on the expected conditions in the process gas system of the K-25 building, are related back to the required data quality objectives (DQO) for the NBA measurement system established for the K-25 demolition project. The combined review of the HMS software algorithms and supporting field measurements lead to the conclusion that the majority of process gas pipe measurements are adequately corrected for source self-attenuation using HMS4. While there will be instances where the UO{sub 2}F{sub 2} holdup mass presents an infinitely thick deposit to the NaI-HMS4 system these situations are expected to be infrequent. This work confirms that the HMS4 system can quantify UO{sub 2}F{sub 2} holdup, in its current configuration (deposition, enrichment, and geometry), below the DQO levels for the K-25 building decommissioning and demolition project. For an area measurement of process gas pipe in the K-25 building, if an infinitely thick UO{sub 2}F{sub 2} deposit is identified in the range of enrichment of {approx}4-40%, the holdup quantity exceeds the corresponding DQO established for the K-25 building demolition project.

  4. Electrobiocommodities from Carbon Dioxide: Enhancing Microbial

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

    Electrosynthesis with Synthetic Electromicrobiology and System Design | Department of Energy Electrobiocommodities from Carbon Dioxide: Enhancing Microbial Electrosynthesis with Synthetic Electromicrobiology and System Design Electrobiocommodities from Carbon Dioxide: Enhancing Microbial Electrosynthesis with Synthetic Electromicrobiology and System Design Presentation by Derek Lovley, UMass Amherst, during the "Targeting High-Value Challenges" panel at the Hydrogen, Hydrocarbons,

  5. 2014 Uranium Marketing Annual Report

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

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

  6. Uranium deposition study on aluminum: results of early tests

    SciTech Connect (OSTI)

    Hughes, M.R.; Nolan, T.A.

    1984-06-19

    Laboratory experiments to quantify uranium compound deposition on Aluminum 3003 test coupons have been initiated. These experiments consist of exposing the coupons to normal assay UF/sub 6/ (0.7% /sup 235/U) in nickel reaction vessels under various conditions of UF/sub 6/ pressure, temperature, and time. To-date, runs from 5 minutes to 2000 hr have been completed at a UF/sub 6/ pressure of 100 torr and at a temperature of 60/sup 0/C. Longer exposure times are in progress. Initial results indicated that a surface film of uranium, primarily as uranyl fluoride (UO/sub 2/F/sub 2/), is deposited very soon after exposure to UF/sub 6/. In a five minute UF/sub 6/ exposure at a temperature of 60/sup 0/C, an average of 2.9 ..mu..g U/cm/sup 2/ was deposited; after 24 hr the deposit typically increased to 5.0 ..mu..g/cm/sup 2/ and then increased to 10.4 ..mu..g/cm/sup 2/ after 2000 hr. This amount of deposit (at 2000 hr exposure) would contribute roughly 10 to 20% to the total 186 keV gamma signal obtained from a GCEP product header pipe being operated at UF/sub 6/ pressures of 2 to 5 torr. The amount of isotopic exchange which would occur in the deposit in the event that HEU and LEU productions were alternated is considered. It is felt that isotopic exchange would not occur to any significant amount within the fixed deposit during relatively short HEU production periods since the HEU would be present primarily as adsorbed UF/sub 6/ molecules on the surface of the deposit. The adsorbed HEU molecules would be removed by evacuation and diluted by LEU production. Major increases in the deposit count would be observed if a leak occurred or moisture was introduced into the system while HEU was being produced.

  7. Uranium isotopes fingerprint biotic reduction

    SciTech Connect (OSTI)

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

    2015-04-20

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

  8. Uranium isotopes fingerprint biotic reduction

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

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

    2015-04-20

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

  9. Layered solid sorbents for carbon dioxide capture (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Layered solid sorbents for carbon dioxide capture Citation Details In-Document Search Title: Layered solid sorbents for carbon dioxide capture You are accessing a document from...

  10. Method for carbon dioxide sequestration (Patent) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Method for carbon dioxide sequestration Citation Details In-Document Search Title: Method for carbon dioxide sequestration You are accessing a document from the Department of...

  11. Electrocatalysts for carbon dioxide conversion

    DOE Patents [OSTI]

    Masel, Richard I; Salehi-Khojin, Amin

    2015-04-21

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

  12. Mathematical simulation of the amplification of 1790-nm laser radiation in a nuclear-excited He – Ar plasma containing nanoclusters of uranium compounds

    SciTech Connect (OSTI)

    Kosarev, V A; Kuznetsova, E E

    2014-02-28

    The possibility of applying dusty active media in nuclearpumped lasers has been considered. The amplification of 1790-nm radiation in a nuclear-excited dusty He – Ar plasma is studied by mathematical simulation. The influence of nanoclusters on the component composition of the medium and the kinetics of the processes occurring in it is analysed using a specially developed kinetic model, including 72 components and more than 400 reactions. An analysis of the results indicates that amplification can in principle be implemented in an active laser He – Ar medium containing 10-nm nanoclusters of metallic uranium and uranium dioxide. (lasers)

  13. A Fission Gas Release Model for High-Burnup LWR ThO{sub 2}-UO{sub 2} Fuel

    SciTech Connect (OSTI)

    Long, Yun; Yi Yuan; Kazimi, Mujid S.; Ballinger, Ronald G.; Pilat, Edward E.

    2002-06-15

    Fission gas release in thoria-urania fuel has been investigated by creating a specially modified FRAPCON-3 code. Because of the reduced buildup of {sup 239}Pu and a flatter distribution of {sup 233}U, the new model THUPS (Thoria-Urania Power Shape) was developed to calculate the radial power distribution, including the effects of both plutonium and {sup 233}U. Additionally, a new porosity model for the rim region was introduced at high burnup. The mechanisms of fission gas release in ThO{sub 2}-UO{sub 2} fuel are expected to be essentially similar to those of UO{sub 2} fuel; therefore, the general formulations of the existing fission gas release models in FRAPCON-3 were retained. However, the gas diffusion coefficient was adjusted to a lower level to account for the smaller observed release fraction in the thoria-based fuel. To model the accelerated fission gas release at high burnup properly, a new athermal fission gas release model was introduced. The modified version of FRAPCON-3 was calibrated using the measured fission gas release data from the light water breeder reactor. Using the new model to calculate the gas release in typical pressurized water reactor hot pins gives data that indicate that the ThO{sub 2}-UO{sub 2} fuel will have considerably lower fission gas release above a burnup of 50 MWd/kg HM.

  14. Inherently safe in situ uranium recovery.

    SciTech Connect (OSTI)

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

    2009-05-01

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

  15. Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance

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

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

  16. High-resolution mineralogical characterization and biogeochemical modeling of uranium reaction pathways at the FRC

    SciTech Connect (OSTI)

    Chen Zhu

    2006-06-15

    High-Resolution Mineralogical Characterization and Biogeochemical Modeling of Uranium Reduction Pathways at the Oak Ridge Field-Research Center (FRC) Chen Zhu, Indiana University, David R. Veblen, Johns Hopkins University We have successfully completed a proof-of-concept, one-year grant on a three-year proposal from the former NABIR program, and here we seek additional two-year funding to complete and publish the research. Using a state-of-the-art 300-kV, atomic resolution, Field Emission Gun Transmission Electron Microscope (TEM), we have successfully identified three categories of mineral hosts for uranium in contaminated soils: (1) iron oxides; (2) mixed manganese-iron oxides; and (3) uranium phosphates. Method development using parallel electron energy loss spectroscopy (EELS) associated with the TEM shows great promise for characterizing the valence states of immobilized U during bioremediation. We have also collected 27 groundwater samples from two push-pull field biostimulation tests, which form two time series from zero to approximately 600 hours. The temporal evolution in major cations, anions, trace elements, and the stable isotopes 34S, 18O in sulfate, 15N in nitrate, and 13C in dissolved inorganic carbon (DIC) clearly show that biostimulation resulted in reduction of nitrate, Mn(IV), Fe(III), U(VI), sulfate, and Tc(VII), and these reduction reactions were intimately coupled with a complex network of inorganic reactions evident from alkalinity, pH, Na, K, Mg, and Ca concentrations. From these temporal trends, apparent zero order rates were regressed. However, our extensive suite of chemical and isotopic data sets, perhaps the first and only comprehensive data set available at the FRC, show that the derived rates from these field biostimulation experiments are composite and lump-sum rates. There were several reactions that were occurring at the same time but were masked by these pseudo-zero order rates. A reaction-path model comprising a total of nine redox couples (NO3–/NH4+, MnO2(s)/Mn2+, Fe(OH)3(s) /Fe2+, TcO4–/TcO2(s), UO22+/UO2(s), SO42–/HS–, CO2/CH4, ethanol/acetate, and H+/H2.) is used to simulate the temporal biogeochemical evolution observed in the field tests. Preliminary results show that the models based on thermodynamics and more complex rate laws can generate the apparent zero order rates when several concurrent or competing reactions occur. Professor Alex Halliday of Oxford University, UK, and his postdoctoral associates are measuring the uranium isotopes in our groundwater samples. Newly developed state-of-the-art analytical techniques in measuring variability in 235U/238U offer the potential to distinguish biotic and abiotic uranium reductive mechanisms.

  17. Low temperature synthesis and sintering of d-UO2 nanoparticles.

    SciTech Connect (OSTI)

    Nenoff, Tina Maria; Ferreira, Summer Rhodes; Robinson, David B.; Jacobs, Benjamin W.; Provencio, Paula Polyak; Huang, Jian Yu

    2010-12-01

    We report on the novel room temperature method of synthesizing advanced nuclear fuels; a method that virtually eliminates any volatility of components. This process uses radiolysis to form stable nanoparticle (NP) nuclear transuranic (TRU) fuel surrogates and in-situ heated stage TEM to sinter the NPs. The radiolysis is performed at Sandia's Gamma Irradiation Facility (GIF) 60Co source (3 x 10{sup 6} rad/hr). Using this method, sufficient quantities of fuels for research purposes can be produced for accelerated advanced nuclear fuel development. We are focused on both metallic and oxide alloy nanoparticles of varying compositions, in particular d-U, d-U/La alloys and d-UO2 NPs. We present detailed descriptions of the synthesis procedures, the characterization of the NPs, the sintering of the NPs, and their stability with temperature. We have employed UV-vis, HRTEM, HAADF-STEM imaging, single particle EDX and EFTEM mapping characterization techniques to confirm the composition and alloying of these NPs.

  18. Intergranular fracture in UO2: derivation of traction-separation law from atomistic simulations

    SciTech Connect (OSTI)

    Yongfeng Zhang; Paul C Millett; Michael R Tonks; Xian-Ming Bai; S Bulent Biner

    2013-10-01

    In this study, the intergranular fracture behavior of UO2 was studied by molecular dynamics simulations using the Basak potential. In addition, the constitutive traction-separation law was derived from atomistic data using the cohesive-zone model. In the simulations a bicrystal model with the (100) symmetric tilt E5 grain boundaries was utilized. Uniaxial tension along the grain boundary normal was applied to simulate Mode-I fracture. The fracture was observed to propagate along the grain boundary by micro-pore nucleation and coalescence, giving an overall intergranular fracture behavior. Phase transformations from the Fluorite to the Rutile and Scrutinyite phases were identified at the propagating crack tips. These new phases are metastable and they transformed back to the Fluorite phase at the wake of crack tips as the local stress concentration was relieved by complete cracking. Such transient behavior observed at atomistic scale was found to substantially increase the energy release rate for fracture. Insertion of Xe gas into the initial notch showed minor effect on the overall fracture behavior.

  19. Atomic Scale Modelling of the Primary Damage State of Irradiated UO{sub 2} Matrix

    SciTech Connect (OSTI)

    Van Brutzel, Laurent

    2008-07-01

    Large scale classical molecular dynamics simulations have been carried out to study the primary damage state due to a-decay self irradiation in UO{sub 2} matrix. Simulations of energetic displacement cascades up to the realistic energy of the recoil nucleus at 80 keV provide new informations on defect production, their spatial distribution and their clustering. The discrepancy with the classical linear theory NRT (Norton-Robinson-Torrens) law on the creation of the number of point defects is discussed. Study of cascade overlap sequence shows a saturation of the number of point defects created as the dose increases. Toward the end of the overlap sequence, large stable clusters of vacancies are observed. The values of athermal diffusion coefficients coming from the ballistic collisions and the additional point defects created during the cascades are estimated from these simulations to be, in all the cases, less than 10-26 m{sup 2}/s. Finally, the influence of a grain boundary of type Sigma 5 is analysed. It has been found that the energy of the cascades are dissipated along the interface and that most of the point defects are created at the grain boundary. (authors)

  20. EIA - Greenhouse Gas Emissions - Carbon Dioxide Emissions

    Gasoline and Diesel Fuel Update (EIA)

    2. Carbon Dioxide Emissions 2.1. Total carbon dioxide emissions Annual U.S. carbon dioxide emissions fell by 419 million metric tons in 2009 (7.1 percent), to 5,447 million metric tons (Figure 9 and Table 6). The annual decrease-the largest over the 19-year period beginning with the 1990 baseline-puts 2009 emissions 608 million metric tons below the 2005 level, which is the Obama Administration's benchmark year for its goal of reducing U.S. emissions by 17 percent by 2020. The key factors

  1. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  2. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  3. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  4. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  5. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  6. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  7. Depleted uranium disposal options evaluation

    SciTech Connect (OSTI)

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

    1994-05-01

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

  8. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  9. uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

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

  10. :- : DRILLING URANIUM BILLETS ON A

    Office of Legacy Management (LM)

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

  11. Uranium Metal Analysis via Selective Dissolution

    SciTech Connect (OSTI)

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

    2008-09-10

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

  12. Uranium Leasing Program | Department of Energy

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

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

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

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

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

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

    SciTech Connect (OSTI)

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

    2013-07-01

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

  15. Carbon Dioxide Emission Factors for Coal

    Reports and Publications (EIA)

    1994-01-01

    The Energy Information Administration (EIA) has developed factors for estimating the amount of carbon dioxide emitted, accounting for differences among coals, to reflect the changing "mix" of coal in U.S. coal consumption.

  16. Recycling Carbon Dioxide to Make Plastics

    Broader source: Energy.gov [DOE]

    The world’s first successful large-scale production of a polypropylene carbonate polymer using waste carbon dioxide as a key raw material has resulted from a projected funded in part by the U.S. Department of Energy.

  17. Potentiometric determination of uranium in organic extracts

    SciTech Connect (OSTI)

    Bodnar, L.Z.

    1980-05-01

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

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

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

    Processing Facility ... Uranium Processing Facility Team Signs Partnering Agreement Posted: July 18, 2014 - 4:39pm Front row, left to right: Bill Priest, Consolidated Nuclear...

  19. Colorimetric detection of uranium in water

    DOE Patents [OSTI]

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

    2012-03-13

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

  20. Radiological Safety Training for Uranium Facilities

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

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

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

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

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

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

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

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

  3. Carbon dioxide-soluble polymers and swellable polymers for carbon dioxide applications

    DOE Patents [OSTI]

    DeSimone, Joseph M.; Birnbaum, Eva; Carbonell, Ruben G.; Crette, Stephanie; McClain, James B.; McCleskey, T. Mark; Powell, Kimberly R.; Romack, Timothy J.; Tumas, William

    2004-06-08

    A method for carrying out a catalysis reaction in carbon dioxide comprising contacting a fluid mixture with a catalyst bound to a polymer, the fluid mixture comprising at least one reactant and carbon dioxide, wherein the reactant interacts with the catalyst to form a reaction product. A composition of matter comprises carbon dioxide and a polymer and a reactant present in the carbon dioxide. The polymer has bound thereto a catalyst at a plurality of chains along the length of the polymer, and wherein the reactant interacts with the catalyst to form a reaction product.

  4. ARM - Measurement - Carbon dioxide (CO2) concentration

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

    concentration ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Carbon dioxide (CO2) concentration The amount of carbon dioxide, a heavy, colorless greenhouse gas, per unit of volume. Categories Atmospheric Carbon Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all

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

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

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

  6. High strength and density tungsten-uranium alloys

    DOE Patents [OSTI]

    Sheinberg, Haskell (Los Alamos, NM)

    1993-01-01

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

  7. Promoting Uranium Immobilization by the Activities of Microbial Phosphatases

    SciTech Connect (OSTI)

    Robert J. Martinez; Melanie J. Beazley; Samuel M. Webb; Martial Taillefert; and Patricia A. Sobecky

    2007-04-19

    The overall objective of this project is to examine the activity of nonspecific phosphohydrolases present in naturally occurring subsurface microorganisms for the purpose of promoting the immobilization of radionuclides through the production of uranium [U(VI)] phosphate precipitates. Specifically, we hypothesize that the precipitation of U(VI) phosphate minerals may be promoted through the microbial release and/or accumulation of PO4 3- as a means to detoxify radionuclides and heavy metals. An experimental approach was designed to determine the extent of phosphatase activity in bacteria previously isolated from contaminated subsurface soils collected at the ERSP Field Research Center (FRC) in Oak Ridge, TN. Screening of 135 metal resistant isolates for phosphatase activity indicated the majority (75 of 135) exhibited a phosphatase-positive phenotype. During this phase of the project, a PCR based approach has also been designed to assay FRC isolates for the presence of one or more classes of the characterized non-specific acid phophastase (NSAP) genes likely to be involved in promoting U(VI) precipitation. Testing of a subset of Pb resistant (Pbr) Arthrobacter, Bacillus and Rahnella strains indicated 4 of the 9 Pbr isolates exhibited phosphatase phenotypes suggestive of the ability to bioprecipitate U(VI). Two FRC strains, a Rahnella sp. strain Y9602 and a Bacillus sp. strain Y9-2, were further characterized. The Rahnella sp. exhibited enhanced phosphatase activity relative to the Bacillus sp. Whole-cell enzyme assays identified a pH optimum of 5.5, and inorganic phosphate accumulated in pH 5.5 synthetic groundwater (designed to mimic FRC conditions) incubations of both strains in the presence of a model organophosphorus substrate provided as the sole C and P source. Kinetic experiments showed that these two organisms can grow in the presence of 200 μM dissolved uranium and that Rahnella is much more efficient in precipitating U(VI) than Bacillus sp. The precipitation of U(VI) must be mediated by biological activity as less than 3% soluble U(VI) was removed either from the abiotic or the heat-killed cell controls. Interestingly, the pH has a strong effect on growth and U(VI) biomineralization rates by Rahnella. Thermodynamic modeling identifies autunite-type minerals [Ca(UO2)2(PO4)2] as the precipitate likely formed in the synthetic FRC groundwater conditions at all pH investigated. Extended X-ray absorption fine structure measurements have recently confirmed that the precipitate found in these incubations is an autunite and meta-autunite-type mineral. A kinetic model of U biomineralization at the different pH indicates that hydrolysis of organophosphate can be described using simple Monod kinetics and that uranium precipitation is accelerated when monohydrogen phosphate is the main orthophosphate species in solution. Overall, these experiments and ongoing soil slurry incubations demonstrate that the biomineralization of U(VI) through the activity of phosphatase enzymes can be expressed in a wide range of geochemical conditions pertaining to the FRC site.

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

    Energy Savers [EERE]

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

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

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

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

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

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

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

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

    National Nuclear Security Administration (NNSA)

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

  12. Decommissioning of U.S. Uranium Production Facilities

    Reports and Publications (EIA)

    1995-01-01

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

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

    Office of Environmental Management (EM)

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

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

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

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

  15. 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...

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

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

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

  17. 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...

  18. 3rd Quarter 2015 Domestic Uranium Production Report

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

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

  19. Domestic Uranium Production Report 4th Quarter 2015

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

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

  20. Competing retention pathways of uranium upon reaction with Fe(II)

    SciTech Connect (OSTI)

    Massey, Michael S.; Lezama Pacheco, Juan S.; Jones, Morris; Ilton, Eugene S.; Cerrato, Jose M.; Bargar, John R.; Fendorf, Scott

    2014-10-01

    Biogeochemical retention processes, including adsorption, reductive precipitation, and incorporation into host minerals, are important in contaminant transport, remediation, and geologic deposition of uranium. Recent work has shown that U can become incorporated into iron (hydr)oxide minerals, with a key pathway arising from Fe(II)-induced transformation of ferrihydrite, (Fe(OH)3•nH2O) to goethite (?-FeO(OH)); this is a possible U retention mechanism in soils and sediments. Several key questions, however, remain unanswered regarding U incorporation into iron (hydr)oxides and this pathway’s contribution to U retention, including: (i) the competitiveness of U incorporation versus reduction to U(IV) and subsequent precipitation of UO2; (ii) the oxidation state of incorporated U; (iii) the effects of uranyl aqueous speciation on U incorporation; and, (iv) the mechanism of U incorporation. Here we use a series of batch reactions conducted at pH ~7, [U(VI)] from 1 to 170 ?M, [Fe(II)] from 0 to 3 mM, and [Ca] at 0 or 4 mM) coupled with spectroscopic examination of reaction products of Fe(II)-induced ferrihydrite transformation to address these outstanding questions. Uranium retention pathways were identified and quantified using extended x-ray absorption fine structure (EXAFS) spectroscopy, x-ray powder diffraction, x-ray photoelectron spectroscopy, and transmission electron microscopy. Analysis of EXAFS spectra showed that 14 to 89% of total U was incorporated into goethite, upon reaction with Fe(II) and ferrihydrite. Uranium incorporation was a particularly dominant retention pathway at U concentrations ? 50 ?M when either uranyl-carbonato or calcium-uranyl-carbonato complexes were dominant, accounting for 64 to 89% of total U. With increasing U(VI) and Fe(II) concentrations, U(VI) reduction to U(IV) became more prevalent, but U incorporation remained a functioning retention pathway. These findings highlight the potential importance of U(V) incorporation within iron oxides as a retention process of U across a wide range of biogeochemical environments and the sensitivity of uranium retention processes to operative (bio)geochemical conditions.

  1. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  2. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  3. 2014 Domestic Uranium Production Report

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

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

  4. 2014 Domestic Uranium Production Report

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

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

  5. 2014 Domestic Uranium Production Report

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

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

  6. 2014 Uranium Marketing Annual Report

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

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

  7. 2014 Uranium Marketing Annual Report

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

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

  8. 2014 Uranium Marketing Annual Report

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

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

  9. 2014 Uranium Marketing Annual Survey

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

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

  10. Enhanced Generic Phase-field Model of Irradiation Materials: Fission Gas Bubble Growth Kinetics in Polycrystalline UO2

    SciTech Connect (OSTI)

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert O.; Gao, Fei; Sun, Xin

    2012-05-30

    Experiments show that inter-granular and intra-granular gas bubbles have different growth kinetics which results in heterogeneous gas bubble microstructures in irradiated nuclear fuels. A science-based model predicting the heterogeneous microstructure evolution kinetics is desired, which enables one to study the effect of thermodynamic and kinetic properties of the system on gas bubble microstructure evolution kinetics and morphology, improve the understanding of the formation mechanisms of heterogeneous gas bubble microstructure, and provide the microstructure to macroscale approaches to study their impact on thermo-mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking. In our previous report 'Mesoscale Benchmark Demonstration, Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing', we developed a phase-field model to simulate the intra-granular gas bubble evolution in a single crystal during post-irradiation thermal annealing. In this work, we enhanced the model by incorporating thermodynamic and kinetic properties at grain boundaries, which can be obtained from atomistic simulations, to simulate fission gas bubble growth kinetics in polycrystalline UO2 fuels. The model takes into account of gas atom and vacancy diffusion, vacancy trapping and emission at defects, gas atom absorption and resolution at gas bubbles, internal pressure in gas bubbles, elastic interaction between defects and gas bubbles, and the difference of thermodynamic and kinetic properties in matrix and grain boundaries. We applied the model to simulate gas atom segregation at grain boundaries and the effect of interfacial energy and gas mobility on gas bubble morphology and growth kinetics in a bi-crystal UO2 during post-irradiation thermal annealing. The preliminary results demonstrate that the model can produce the equilibrium thermodynamic properties and the morphology of gas bubbles at grain boundaries for given grain boundary properties. More validation of the model capability in polycrystalline is underway.

  11. Method for fabricating laminated uranium composites

    DOE Patents [OSTI]

    Chapman, L.R.

    1983-08-03

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

  12. Scrap uranium recycling via electron beam melting

    SciTech Connect (OSTI)

    McKoon, R.

    1993-11-01

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

  13. Atomistic Calculations of the Effect of Minor Actinides on Thermodynamic and Kinetic Properties of UO{sub 2{+-}x}

    SciTech Connect (OSTI)

    Deo, Chaitanya; Adnersson, Davis; Battaile, Corbett; uberuaga, Blas

    2012-10-30

    The team will examine how the incorporation of actinide species important for mixed oxide (MOX) and other advanced fuel designs impacts thermodynamic quantities of the host UO{sub 2} nuclear fuel and how Pu, Np, Cm and Am influence oxygen mobility. In many cases, the experimental data is either insufficient or missing. For example, in the case of pure NpO2, there is essentially no experimental data on the hyperstoichiometric form it is not even known if hyperstoichiometry NpO{sub 2{+-}x} is stable. The team will employ atomistic modeling tools to calculate these quantities

  14. E&nr Ph. S. W.. Wahhgt~n. D.C. 200242174, TIkpbnc (202) 48a60uo

    Office of Legacy Management (LM)

    75' 00.955 L' E&nr Ph. S. W.. Wahhgt~n. D.C. 200242174, TIkpbnc (202) 48a60uo 7117-03.87.cdy.43 23 September 1987 CR CA.d M r. Andrew Wallo, III, NE-23 Division of Facility & Site Decoaunissioning Projects U.S. Department of Energy Germantown, Maryland 20545 Dear M r. Wallo: ELIMINATION RECOMMENDATION -- COLLEGES AND UNIVERSITIES M /4.0-03 kl 77.0% I - The attached elimination reconunendation was prepared in accordance rlL.0~ with your suggestion during our meeting on 22 September. The

  15. Carbon dioxide capture process with regenerable sorbents

    DOE Patents [OSTI]

    Pennline, Henry W.; Hoffman, James S.

    2002-05-14

    A process to remove carbon dioxide from a gas stream using a cross-flow, or a moving-bed reactor. In the reactor the gas contacts an active material that is an alkali-metal compound, such as an alkali-metal carbonate, alkali-metal oxide, or alkali-metal hydroxide; or in the alternative, an alkaline-earth metal compound, such as an alkaline-earth metal carbonate, alkaline-earth metal oxide, or alkaline-earth metal hydroxide. The active material can be used by itself or supported on a substrate of carbon, alumina, silica, titania or aluminosilicate. When the active material is an alkali-metal compound, the carbon-dioxide reacts with the metal compound to generate bicarbonate. When the active material is an alkaline-earth metal, the carbon dioxide reacts with the metal compound to generate carbonate. Spent sorbent containing the bicarbonate or carbonate is moved to a second reactor where it is heated or treated with a reducing agent such as, natural gas, methane, carbon monoxide hydrogen, or a synthesis gas comprising of a combination of carbon monoxide and hydrogen. The heat or reducing agent releases carbon dioxide gas and regenerates the active material for use as the sorbent material in the first reactor. New sorbent may be added to the regenerated sorbent prior to subsequent passes in the carbon dioxide removal reactor.

  16. Uranium Pyrophoricity Phenomena and Prediction

    SciTech Connect (OSTI)

    DUNCAN, D.R.

    2000-04-20

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

  17. Experimental investigations of long-term interactions of molten UO/sub 2/ with MgO and concrete at Argonne National Laboratory. [LMFBR

    SciTech Connect (OSTI)

    Stein, R.P.; Farhadieh, R.; Pedersen, D.R.; Gunther, W.H.; Purviance, R.T.

    1982-01-01

    Experimental work at Argonne is being performed to investigate the long-term molten-core-debris retention capability of the ex-vessel cavity following a postulated meltdown accident. The eventual objective of the work is to determine if normal structural material (concrete) or a specifically selected sacrificial material (MgO) located in the ex-vessel cavity region can effectively contain molten core debris. The materials under investigation at ANL are various types of concrete (limestone, basalt and magnetite) and commercially-available MgO brick. Results are presented of the status of real material experimental investigation at ANL into (1) molten UO/sub 2/ pool heat transfer, (2) long-term molten UO/sub 2/ penetration into concrete and (3) long-term molten UO/sub 2/ penetration into refractory substrates. The decay heating in the fuel has been simulated by direct electrical heating permitting the study of the long-term interaction.

  18. SEQUESTERING CARBON DIOXIDE IN COALBEDS

    SciTech Connect (OSTI)

    K.A.M. Gasem; R.L. Robinson, Jr.; J.E. Fitzgerald; Z. Pan; M. Sudibandriyo

    2003-04-30

    The authors' long-term goal is to develop accurate prediction methods for describing the adsorption behavior of gas mixtures on solid adsorbents over complete ranges of temperature, pressure, and adsorbent types. The originally-stated, major objectives of the current project are to: (1) measure the adsorption behavior of pure CO{sub 2}, methane, nitrogen, and their binary and ternary mixtures on several selected coals having different properties at temperatures and pressures applicable to the particular coals being studied, (2) generalize the adsorption results in terms of appropriate properties of the coals to facilitate estimation of adsorption behavior for coals other than those studied experimentally, (3) delineate the sensitivity of the competitive adsorption of CO{sub 2}, methane, and nitrogen to the specific characteristics of the coal on which they are adsorbed; establish the major differences (if any) in the nature of this competitive adsorption on different coals, and (4) test and/or develop theoretically-based mathematical models to represent accurately the adsorption behavior of mixtures of the type for which measurements are made. As this project developed, an important additional objective was added to the above original list. Namely, we were encouraged to interact with industry and/or governmental agencies to utilize our expertise to advance the state of the art in coalbed adsorption science and technology. As a result of this additional objective, we participated with the Department of Energy and industry in the measurement and analysis of adsorption behavior as part of two distinct investigations. These include (a) Advanced Resources International (ARI) DOE Project DE-FC26-00NT40924, ''Adsorption of Pure Methane, Nitrogen, and Carbon Dioxide and Their Mixtures on Wet Tiffany Coal'', and (b) the DOE-NETL Project, ''Round Robin: CO{sub 2} Adsorption on Selected Coals''. These activities, contributing directly to the DOE projects listed above, also provided direct synergism with the original goals of our work. Specific accomplishments of this project are summarized below in three broad categories: experimentation, model development, and coal characterization.

  19. Modified biokinetic model for uranium from analysis of acute exposure to UF6

    SciTech Connect (OSTI)

    Fisher, D.R.; Kathren, R.L.; Swint, M.J. )

    1991-03-01

    Urinalysis measurements from 31 workers acutely exposed to uranium hexafluoride (UF6) and its hydrolysis product UO2F2 (during the 1986 Gore, Oklahoma UF6-release accident) were used to develop a modified recycling biokinetic model for soluble U compounds. The model is expressed as a five-compartment exponential equation: yu(t) = 0.086e-2.77t + 0.0048e-0.116t + 0.00069e-0.0267t + 0.00017 e-0.00231t + 2.5 x 10(-6) e-0.000187t, where yu(t) is the fractional daily urinary excretion and t is the time after intake, in days. The excretion constants of the five exponential compartments correspond to residence half-times of 0.25, 6, 26, 300, and 3,700 d in the lungs, kidneys, other soft tissues, and in two bone volume compartments, respectively. The modified recycling model was used to estimate intake amounts, the resulting committed effective dose equivalent, maximum kidney concentrations, and dose equivalent to bone surfaces, kidneys, and lungs.

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

    SciTech Connect (OSTI)

    Ladd-Lively, Jennifer L

    2014-01-01

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

  1. Uranium Leasing Program Environmental Documents | Department of Energy

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

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

  2. Uranium Lease Tracts Location Map | Department of Energy

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

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

  3. Polymers for metal extractions in carbon dioxide

    DOE Patents [OSTI]

    DeSimone, Joseph M. (7315 Crescent Ridge Dr., Chapel Hill, NC 27516); Tumas, William (1130 Big Rock Loop, Los Alamos, NM 87544); Powell, Kimberly R. (103 Timber Hollow Ct. Apartment 323, Chapel Hill, NC 27514); McCleskey, T. Mark (1930 Camino Mora, Los Alamos, NM 87544); Romack, Timothy J. (5810 Forest Ridge Dr., Durham, NC 27713); McClain, James B. (8530 Sommersweet La., Raleigh, NC 27612); Birnbaum, Eva R. (1930 Camino Mora, Los Alamos, NM 87544)

    2001-01-01

    A composition useful for the extraction of metals and metalloids comprises (a) carbon dioxide fluid (preferably liquid or supercritical carbon dioxide); and (b) a polymer in the carbon dioxide, the polymer having bound thereto a ligand that binds the metal or metalloid; with the ligand bound to the polymer at a plurality of locations along the chain length thereof (i.e., a plurality of ligands are bound at a plurality of locations along the chain length of the polymer). The polymer is preferably a copolymer, and the polymer is preferably a fluoropolymer such as a fluoroacrylate polymer. The extraction method comprises the steps of contacting a first composition containing a metal or metalloid to be extracted with a second composition, the second composition being as described above; and then extracting the metal or metalloid from the first composition into the second composition.

  4. Microstructure changes and thermal conductivity reduction in UO2 following 3.9 MeV He2+ ion irradiation

    SciTech Connect (OSTI)

    Janne Pakrinen; Marat Khafizov; Lingfeng He; Chris Wetland; Jian Gan; Andrew T. Nelson; David H Hurley; Anter El-Azab; Todd R Allen

    2014-11-01

    The microstructural changes and associated effects on thermal conductivity were examined in UO2 after irradiation using 3.9 MeV He2+ ions. Lattice expansion of UO2 was observed in x-ray diffraction after ion irradiation up to 5×1016 He2+/cm2 at low-temperature (< 200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 µm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 µm) in the sample subjected to 5×1016 He2+/cm2, the highest fluence reached, while similar features were not detected at 9×1015 He2+/cm2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55 % for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.

  5. FASTGRASS implementation in BISON and Fission gas behavior characterization in UO2 and connection to validating MARMOT

    SciTech Connect (OSTI)

    Yun, Di; Mo, Kun; Ye, Bei; Jamison, Laura M.; Miao, Yinbin; Lian, Jie; Yao, Tiankei

    2015-09-30

    This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL). Two major accomplishments in FY 15 are summarized in this report: (1) implementation of the FASTGRASS module in the BISON code; and (2) a Xe implantation experiment for large-grained UO2. Both BISON AND MARMOT codes have been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. To contribute to the development of the Moose-Bison-Marmot (MBM) code suite, we have implemented the FASTGRASS fission gas model as a module in the BISON code. Based on rate theory formulations, the coupled FASTGRASS module in BISON is capable of modeling LWR oxide fuel fission gas behavior and fission gas release. In addition, we conducted a Xe implantation experiment at the Argonne Tandem Linac Accelerator System (ATLAS) in order to produce the needed UO2 samples with desired bubble morphology. With these samples, further experiments to study the fission gas diffusivity are planned to provide validation data for the Fission Gas Release Model in MARMOT codes.

  6. Enhanced carbon dioxide capture upon incorporation of

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

    N,N'-dimethylethylenediamine in the metal-organic framework CuBTTri | Center for Gas SeparationsRelevant to Clean Energy Technologies | Blandine Jerome Enhanced carbon dioxide capture upon incorporation of N,N'-dimethylethylenediamine in the metal-organic framework CuBTTri Previous Next List Thomas M. McDonald, Deanna M. D'Alessandro, Rajamani Krishna and Jeffrey R. Long, Chem. Sci., 2011,2, 2022-2028 DOI: 10.1039/C1SC00354B Graphical abstract: Enhanced carbon dioxide capture upon

  7. uranium

    National Nuclear Security Administration (NNSA)

    a>

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

  8. Technical Basis for Assessing Uranium Bioremediation Performance

    SciTech Connect (OSTI)

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

    2008-04-01

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

  9. Genome-Based Models to Optimize In Situ Bioremediation of Uranium and Harvesting Electrical Energy from Waste Organic Matter

    SciTech Connect (OSTI)

    Lovley, Derek R

    2012-12-28

    The goal of this research was to provide computational tools to predictively model the behavior of two microbial communities of direct relevance to Department of Energy interests: 1) the microbial community responsible for in situ bioremediation of uranium in contaminated subsurface environments; and 2) the microbial community capable of harvesting electricity from waste organic matter and renewable biomass. During this project the concept of microbial electrosynthesis, a novel form of artificial photosynthesis for the direct production of fuels and other organic commodities from carbon dioxide and water was also developed and research was expanded into this area as well.

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

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

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

  11. Record of Decision for the Uranium Leasing Program Programmatic

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

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

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

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

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

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

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

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

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

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

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

  15. Final Uranium Leasing Program Programmatic Environmental Impact Statement

    Office of Environmental Management (EM)

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

  16. RECOVERY OF URANIUM FROM CARBONATE LEACH LIQUORS

    DOE Patents [OSTI]

    Wilson, H.F.

    1958-07-01

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

  17. Uranium Mining, Conversion, and Enrichment Industries

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

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

  18. Project Profile: 10-Megawatt Supercritical Carbon Dioxide Turbine...

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

    aim to demonstrate a multi-megawatt power cycle using supercritical carbon dioxide (s-CO2) as the working fluid. The use of carbon dioxide instead of steam allows higher...

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

    Office of Scientific and Technical Information (OSTI)

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

  20. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  1. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  2. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  3. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  4. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    1 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 thousand pounds U 3 O 8 equivalent 2010 2011 2012 2013 P2014 Owners and operators of U.S. civilian nuclear power reactors 86,527 89,835 97,647 113,007 116,047 U.S. brokers and traders 11,125 6,841 5,677 7,926 5,798 U.S. converter, enrichers, fabricators, and producers 13,608 15,428 17,611 13,416 12,766 Total commercial inventories 111,259 112,104 120,936 134,418 134,611 thousand pounds U 3 O 8

  5. 2014 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    3 2014 Uranium Marketing Annual Report Release Date: May 13, 2015 Next Release Date: May 2016 Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price Quantity with reported price Weighted- average price First 7,119 38.24 7,175 34.34 6,665 30.26 Second 7,119 48.64 7,175 41.29 6,665 35.11 Third 7,119 51.16 7,175 45.89 6,665 39.29 Fourth 7,119 54.15 7,175 49.84 6,665 43.36 Fifth 7,119 56.93 7,175 53.17 6,665 46.74 Sixth 7,119 59.98 7,175 57.24 6,665

  6. Domestic Uranium Production Report - Quarterly

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

    1. Total production of uranium concentrate in the United States, 1996 - 4th quarter 2015 pounds U3O8 Calendar-year quarter 1st quarter 2nd quarter 3rd quarter 4th quarter Calendar-year total 1996 1,734,427 1,460,058 1,691,796 1,434,425 6,320,706 1997 1,149,050 1,321,079 1,631,384 1,541,052 5,642,565 1998 1,151,587 1,143,942 1,203,042 1,206,003 4,704,574 1999 1,196,225 1,132,566 1,204,984 1,076,897 4,610,672 2000 1,018,683 983,330 981,948 973,585 3,975,545 2001 709,177 748,298 628,720 553,060

  7. 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.

    1997-07-01

    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.

  8. Geothermal Startup Will Put Carbon Dioxide to Good Use

    Broader source: Energy.gov [DOE]

    Geothermal power holds enormous opportunities to provide affordable, clean energy that avoids greenhouse gases like carbon dioxide (CO2).

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

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

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

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

  11. Array of titanium dioxide nanostructures for solar energy utilization

    DOE Patents [OSTI]

    Qiu, Xiaofeng; Parans Paranthaman, Mariappan; Chi, Miaofang; Ivanov, Ilia N; Zhang, Zhenyu

    2014-12-30

    An array of titanium dioxide nanostructures for solar energy utilization includes a plurality of nanotubes, each nanotube including an outer layer coaxial with an inner layer, where the inner layer comprises p-type titanium dioxide and the outer layer comprises n-type titanium dioxide. An interface between the inner layer and the outer layer defines a p-n junction.

  12. Acid sorption regeneration process using carbon dioxide

    DOE Patents [OSTI]

    King, C. Judson (Kensington, CA); Husson, Scott M. (Anderson, SC)

    2001-01-01

    Carboxylic acids are sorbed from aqueous feedstocks onto a solid adsorbent in the presence of carbon dioxide under pressure. The acids are freed from the sorbent phase by a suitable regeneration method, one of which is treating them with an organic alkylamine solution thus forming an alkylamine-carboxylic acid complex which thermally decomposes to the desired carboxylic acid and the alkylamine.

  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)

    1986-01-01

    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. Highly Enriched Uranium Disposition | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    needs primarily by down-blending, or converting, it into low enriched uranium (LEU). Once down-blended, the material can no longer be used for nuclear weapons. To the extent...

  15. The Uranium Resource: A Comparative Analysis

    SciTech Connect (OSTI)

    Schneider, Erich A.; Sailor, William C.

    2007-07-01

    An analogy was drawn between uranium and thirty five minerals for which the USGS maintains extensive records. The USGS mineral price data, which extends from 1900 to the present, was used to create a simple model describing long term price evolution. Making the assumption that the price of uranium, a geologically unexceptional mineral, will evolve in a manner similar to that of the USGS minerals, the model was used to project its price trend for this century. Based upon the precedent set by the USGS data, there is an 80% likelihood that the price of uranium will decline. Moreover, the most likely scenario would see the equilibrium price of uranium decline by about 40% by mid-century. (authors)

  16. Ex Parte Communications- Uranium Producers of America

    Broader source: Energy.gov [DOE]

    On Thursday, February 12, 2015, representatives from the Uranium  Producers  of America (UPA) met with the Department of Energy (DOE) officials to discuss the management of the federal excess...

  17. PROCESSES OF RECLAIMING URANIUM FROM SOLUTIONS

    DOE Patents [OSTI]

    Zumwalt, L.R.

    1959-02-10

    A process is described for reclaiming residual enriched uranium from calutron wash solutions containing Fe, Cr, Cu, Ni, and Mn as impurities. The solution is adjusted to a pH of between 2 and 4 and is contacted with a metallic reducing agent, such as iron or zinc, in order to reduce the copper to metal and thereby remove it from the solution. At the same time the uranium present is reduced to the uranous state The solution is then contacted with a precipitate of zinc hydroxide or barium carbonate in order to precipitate and carry uranium, iron, and chromium away from the nickel and manganese ions in the solution. The uranium is then recovered fronm this precipitate.

  18. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    1. U.S. Forward-Cost Uranium Reserves by State, Year-End 2008 State 50lb 100lb Ore (million tons) Gradea (%) U3O8 (million lbs) Ore (million tons) Gradea (%) U3O8 (million lbs)...

  19. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    Methodology The U.S. uranium ore reserves reported by EIA for specific MFC categories represent the sums of quantities estimated to occur in known deposits on properties where data...

  20. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    2. U.S. Forward-Cost Uranium Reserves by Mining Method, Year-End 2008 Mining Method 50 per pound 100 per pound Ore (million tons) Gradea (percent U3O8) U3O8 (million pounds) Ore...

  1. Process for reducing beta activity in uranium

    DOE Patents [OSTI]

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

    1985-04-11

    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.

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

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

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

  3. Federal Actions to Address Impacts of Uranium

    Office of Legacy Management (LM)

    Federal Actions to Address Impacts of Uranium Contamination in the Navajo Nation 2014 Page | i TABLE OF CONTENTS Executive Summary ....................................................................................................................... 1 Introduction .................................................................................................................................... 2 Summary of Work Completed 2008-2012

  4. The ultimate disposition of depleted uranium

    SciTech Connect (OSTI)

    Lemons, T.R.

    1991-12-31

    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.

  5. Depleted uranium: A DOE management guide

    SciTech Connect (OSTI)

    1995-10-01

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

  6. Uranium Leasing Program Documents | Department of Energy

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

    Documents Uranium Leasing Program Documents U.S. District Court's Order of October 18, 2011, in Colorado Environmental Coalition v. Office of Legacy Management, Civil Action No. 08-cv-01624 (D. Colo.). The Court has issued the injunctive relief described on pages 51-52 of the Order. U.S. District Court's Order of February 27, 2012, in Colorado Environmental Coalition v. Office of Legacy Management, Civil Action No. 08-cv-01624 (D. Colo.). Uranium Lease Tracts Location Map

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

    SciTech Connect (OSTI)

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

    2008-07-01

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

  8. Electrochemical method of producing eutectic uranium alloy and apparatus

    DOE Patents [OSTI]

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

    1995-01-01

    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.

  9. Phase-field simulations of intragranular fission gas bubble evolution in UO2 under post-irradiation thermal annealing

    SciTech Connect (OSTI)

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert O.; Gao, Fei; Sun, Xin

    2013-05-15

    Fission gas bubble is one of evolving microstructures, which affect thermal mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking, in operating nuclear fuels. Therefore, fundamental understanding of gas bubble evolution kinetics is essential to predict the thermodynamic property and performance changes of fuels. In this work, a generic phasefield model was developed to describe the evolution kinetics of intra-granular fission gas bubbles in UO2 fuels under post-irradiation thermal annealing conditions. Free energy functional and model parameters are evaluated from atomistic simulations and experiments. Critical nuclei size of the gas bubble and gas bubble evolution were simulated. A linear relationship between logarithmic bubble number density and logarithmic mean bubble diameter is predicted which is in a good agreement with experimental data.

  10. Y H-S I-H HATIOHAL LEAth~~Y~~OF' OtUO ' Industrial Hygiene No.

    Office of Legacy Management (LM)

    H-S I-H HATIOHAL LEAth~~Y~~OF' OtUO ' Industrial Hygiene No. P.O. Box 158)At.' He&by~kation Sample Nos. ? Sk. 0 qq! Cinchnail 31;Obio Type of SampleCI" lz -- HEALTH AND SAFETY DIVISIQN Analyze for u Method &An.ly,i, ;i __ = - . -- Industrial Hygiene No. P.O. Box 158 Mt.~He&lly Qq99 Q' ' - Ciacian& 31. 01 Sample Nos. 992' I HtAL I H ANU SAFt H-S 1-H J MATI~ LEID-WAIJY. OF OIUO station rio Type of S&h d' r dQsr CYS ..- . . -.. ..- -. -- :TY DIVISICJN Analyze for d t I I I

  11. 2014 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    11 2014 Domestic Uranium Production Report Release Date: April 30, 2015 Next Release Date: May 2016 Total Land and Other 2003 W W 31.3 NA NA NA W 2004 10.6 27.8 48.4 NA NA NA 86.9 2005 18.1 58.2 59.7 NA NA NA 136.0 2006 40.1 65.9 115.2 41.0 23.3 50.9 221.2 2007 67.5 90.4 178.2 77.7 50.3 50.2 336.2 2008 81.9 221.2 164.4 65.2 50.2 49.1 467.6 2009 35.4 141.0 104.0 17.3 24.2 62.4 280.5 2010 44.6 133.3 99.5 20.2 34.5 44.7 277.3 2011 53.6 168.8 96.8 19.6 43.5 33.7 319.2 2012 66.6 186.9 99.4 16.8 33.3

  12. Uranium mill ore dust characterization

    SciTech Connect (OSTI)

    Knuth, R.H.; George, A.C.

    1980-11-01

    Cascade impactor and general air ore dust measurements were taken in a uranium processing mill in order to characterize the airborne activity, the degree of equilibrium, the particle size distribution and the respirable fraction for the /sup 238/U chain nuclides. The sampling locations were selected to limit the possibility of cross contamination by airborne dusts originating in different process areas of the mill. The reliability of the modified impactor and measurement techniques was ascertained by duplicate sampling. The results reveal no significant deviation from secular equilibrium in both airborne and bulk ore samples for the /sup 234/U and /sup 230/Th nuclides. In total airborne dust measurements, the /sup 226/Ra and /sup 210/Pb nuclides were found to be depleted by 20 and 25%, respectively. Bulk ore samples showed depletions of 10% for the /sup 226/Ra and /sup 210/Pb nuclides. Impactor samples show disequilibrium of /sup 226/Ra as high as +-50% for different size fractions. In these samples the /sup 226/Ra ratio was generally found to increase as particle size decreased. Activity median aerodynamic diameters of the airborne dusts ranged from 5 to 30 ..mu..m with a median diameter of 11 ..mu..m. The maximum respirable fraction for the ore dusts, based on the proposed International Commission on Radiological Protection's (ICRP) definition of pulmonary deposition, was < 15% of the total airborne concentration. Ore dust parameters calculated for impactor duplicate samples were found to be in excellent agreement.

  13. Extraction of furfural with carbon dioxide

    SciTech Connect (OSTI)

    Gamse, T.; Marr, R.; Froeschl, F.; Siebenhofer, M.

    1997-01-01

    A new approach to separate furfural from aqueous waste has been investigated. Recovery of furfural and acetic acid from aqueous effluents of a paper mill has successfully been applied on an industrial scale since 1981. The process is based on the extraction of furfural and acetic acid by the solvent trooctylphosphineoxide (TOPO). Common extraction of both substances may cause the formation of resin residues. Improvement was expected by selective extraction of furfural with chlorinated hydrocarbons, but ecological reasons stopped further development of this project. The current investigation is centered in the evaluation of extraction of furfural by supercritical carbon dioxide. The influence of temperature and pressure on the extraction properties has been worked out. The investigation has considered the multi-component system furfural-acetic acid-water-carbon dioxide. Solubility of furfural in liquid and supercritical carbon dioxide has been measured, and equilibrium data for the ternary system furfural-water-CO{sub 2} as well as for the quaternary system furfural-acetic acid-water-CO{sub 2} have been determined. A high-pressure extraction column has been used for evaluation of mass transfer rates.

  14. RESOLUTION OF URANIUM ISOTOPES WITH KINETIC PHOSPHORESCENCE ANALYSIS

    SciTech Connect (OSTI)

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

    2013-04-01

    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.

  15. LM Progressing with Uranium Mines Report to Congress | Department...

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

    Progressing with Uranium Mines Report to Congress LM Progressing with Uranium Mines Report to Congress July 12, 2013 - 10:50am Addthis As reported in an earlier Program Update...

  16. DOE - Office of Legacy Management -- Abandoned Uranium Mines

    Office of Legacy Management (LM)

    Uranium Mines Report to Congress The U.S. Department of Energy (DOE) Office of Legacy Management completed a report on defense-related uranium mines in consultation with...

  17. Uranium at Y-12: Inspection | Y-12 National Security Complex

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

    Inspection Uranium at Y-12: Inspection Posted: July 22, 2013 - 3:36pm | Y-12 Report | Volume 10, Issue 1 | 2013 Inspection of enriched uranium is performed by dimensional checks...

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

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

    Recovery Uranium at Y-12: Recovery Posted: July 22, 2013 - 3:44pm | Y-12 Report | Volume 10, Issue 1 | 2013 Recovery involves reclaiming uranium from numerous sources and...

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

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

    ... Uranium at Y-12: Accountability Posted: July 22, 2013 - 3:37pm | Y-12 Report | Volume 10, Issue 1 | 2013 Accountability of enriched uranium is facilitated by the ability to put...

  20. Think Uranium. Think Y-12 | Y-12 National Security Complex

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

    | Y-12 Report | Volume 10, Issue 1 | 2013 Uranium fever: Much like the California gold rush of 1849, the uranium flurry of 1949 led Geiger counter-toting prospectors to scour...

  1. Y-12 Bulletin Uranium Articles | Y-12 National Security Complex

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

    Bulletin Uranium ... Y-12 Bulletin Uranium Articles Posted: July 22, 2013 - 3:13pm | Y-12 Report | Volume 10, Issue 1 | 2013 These and other articles can be found in archived...

  2. 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...

  3. Uranium distribution in relation to sedimentary facies, Kern Lake, California

    SciTech Connect (OSTI)

    Merifield, P.M.; Carlisle, D.; Idiz, E.; Anderhalt, R.; Reed, W.E.; Lamar, D.L.

    1980-04-01

    Kern Lake has served as a sink for drainage from the southern Sierra Nevada and, in lesser amounts, from the southern Temblor Range. Both areas contain significant uranium source rocks. The uranium content in Holocene Kern Lake sediments correlates best with the mud (silt and clay) fraction. It correlates less well with organic carbon. Biotite grains could account for much of the uranium in the sand fraction, and perhaps the silt fraction as well. The data suggest that fixation of uranium by adsorption on mineral grains is a dominant process in this lake system. Further work is required to determine the importance of cation-exchange of uranium on clays and micas and of organically complexed uranium adsorbed to mineral surfaces. These findings also raise the question of whether uranium transport down the Kern River occurs largely as uranium adsorbed to mineral surfaces.

  4. The Uranium Processing Facility (UPF) Finite Element Meshing Discussion |

    Office of Environmental Management (EM)

    Department of Energy The Uranium Processing Facility (UPF) Finite Element Meshing Discussion The Uranium Processing Facility (UPF) Finite Element Meshing Discussion The Uranium Processing Facility (UPF) Finite Element Meshing Discussion Loring Wyllie Arne Halterman Degenkolb Engineers, San Francisco PDF icon The Uranium Processing Facility (UPF) Finite Element Meshing Discussion More Documents & Publications SASSI Subtraction Method Effects at Various DOE projects October 2009 Seismic

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

    SciTech Connect (OSTI)

    Marsh, Terence L.

    2013-07-30

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

  6. Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl

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

    Complexes with Glutarimidedioxime Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl Complexes with Glutarimidedioxime Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl Complexes with Glutarimidedioxime Print Sunday, 14 October 2012 00:00 The ocean is an important source of uranium if it can be extracted economically. Extraction of uranium from seawater is very challenging, not only because it is in an extremely low concentration, but also because

  7. Manhattan Project: Early Uranium Research, 1939-1941

    Office of Scientific and Technical Information (OSTI)

    Ernest Lawrence, Arthur Compton, Vannevar Bush, and James Conant discuss uranium research, Berkeley, March 29, 1940. EARLY URANIUM RESEARCH (1939-1941) Events > Early Government Support, 1939-1942 Einstein's Letter, 1939 Early Uranium Research, 1939-1941 Piles and Plutonium, 1939-1941 Reorganization and Acceleration, 1940-1941 The MAUD Report, 1941 A Tentative Decision to Build the Bomb, 1941-1942 President Franklin D. Roosevelt responded to the call for government support of uranium research

  8. DOE Releases Excess Uranium Inventory Plan | Department of Energy

    Energy Savers [EERE]

    Excess Uranium Inventory Plan DOE Releases Excess Uranium Inventory Plan December 16, 2008 - 8:51am Addthis WASHINGTON, D.C. - The United States Department of Energy (DOE) today issued its Excess Uranium Inventory Management Plan (the Plan), which outlines the Department's strategy for the management and disposition of its excess uranium inventories. The Plan highlights DOE's ongoing efforts to enhance national security and promote a healthy domestic nuclear infrastructure through the efficient

  9. Testing for Uranium Deuteride Initiation in Liquid Deuterium

    SciTech Connect (OSTI)

    Siekhaus, W. J.; Teslich, N. E.; Kucheyev, S. O.; Go, J.

    2015-10-29

    This report offers a description of the testing related to Uranium foil and its interaction with liquid deuterium.

  10. Excess Uranium Inventory Management Plan | Department of Energy

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

    Excess Uranium Inventory Management Plan Excess Uranium Inventory Management Plan The 2013 Excess Uranium Inventory Management Plan describes a framework for the effective management of the Energy Department's surplus uranium inventory in support of meeting its critical environmental cleanup and national security missions. The Plan is not a commitment to specific activities beyond those that have already been contracted nor is it a restriction on actions that the Department may undertake in the

  11. Uranium Leasing Program Draft Programmatic EIS Issued for Public Comment |

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

    Department of Energy Uranium Leasing Program Draft Programmatic EIS Issued for Public Comment Uranium Leasing Program Draft Programmatic EIS Issued for Public Comment March 15, 2013 - 11:08am Addthis Uranium Leasing Program Draft Programmatic EIS Issued for Public Comment DOE has issued the Draft Uranium Leasing Program Programmatic Environmental Impact Statement (ULP PEIS)(DOE/EIS-0472D) for public review and comment. The document is available here and on the ULP PEIS website. Under the

  12. Retrieval of buried depleted uranium from the T-1 trench

    SciTech Connect (OSTI)

    Burmeister, M.; Castaneda, N.; Greengard, T. |; Hull, C.; Barbour, D.; Quapp, W.J.

    1998-07-01

    The Trench 1 remediation project will be conducted this year to retrieve depleted uranium and other associated materials from a trench at Rocky Flats Environmental Technology Site. The excavated materials will be segregated and stabilized for shipment. The depleted uranium will be treated at an offsite facility which utilizes a novel approach for waste minimization and disposal through utilization of a combination of uranium recycling and volume efficient uranium stabilization.

  13. Borehole Logging Methods for Exploration and Evaluation of Uranium Deposits

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

    (1967) | Department of Energy Borehole Logging Methods for Exploration and Evaluation of Uranium Deposits (1967) Borehole Logging Methods for Exploration and Evaluation of Uranium Deposits (1967) Borehole Logging Methods for Exploration and Evaluation of Uranium Deposits (1967) PDF icon Borehole Logging Methods for Exploration and Evaluation of Uranium Deposits (1967) More Documents & Publications Gamma-Ray Logging Workshop (February 1981) Grade Assignments for Models Used for

  14. Uranium Processing Facility Site Readiness Subproject Completed on Time and

    National Nuclear Security Administration (NNSA)

    Under Budget | National Nuclear Security Administration Library / Press Releases / Uranium Processing Facility Site Readiness Subproject Completed ... Uranium Processing Facility Site Readiness Subproject Completed on Time and Under Budget Press Release Mar 13, 2015 Washington D.C.--The Uranium Processing Facility (UPF) project celebrates its first major milestone with the completion of site readiness work, delivered on time and under budget. "UPF is essential to our Nation's uranium

  15. Reimbursements to Licensees of Active Uranium and Thorium Processing Sites,

    Energy Savers [EERE]

    Fiscal Year 2009 and 2010 Status Report | Department of Energy Reimbursements to Licensees of Active Uranium and Thorium Processing Sites, Fiscal Year 2009 and 2010 Status Report Reimbursements to Licensees of Active Uranium and Thorium Processing Sites, Fiscal Year 2009 and 2010 Status Report Reimbursements to Licensees of Active Uranium and Thorium Processing Sites, Fiscal Year 2009 and 2010 Status Report (March 2010) PDF icon Reimbursements to Licensees of Active Uranium and Thorium

  16. TRACE ELEMENT ANALYSES OF URANIUM MATERIALS

    SciTech Connect (OSTI)

    Beals, D; Charles Shick, C

    2008-06-09

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

  17. Chapter 20 - Uranium Enrichment Decontamination & Decommissioning Fund

    Energy Savers [EERE]

    0. Uranium Enrichment Decontamination and Decommissioning Fund 20-1 CHAPTER 20 URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND 1. INTRODUCTION. a. Purpose. To establish policies and procedures for the financial management, accounting, budget preparation, cash management of the Uranium Enrichment Decontamination and Decommissioning Fund, referred to hereafter as the Fund. b. Applicability. This chapter applies to all Departmental elements, including the National Nuclear Security

  18. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, John P.

    1992-01-01

    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.

  19. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, J.P.

    1992-03-17

    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.

  20. Uranium Processing Facility Site Readiness Subproject Completed on Time and

    National Nuclear Security Administration (NNSA)

    Under Budget | National Nuclear Security Administration Field Offices / Welcome to the NNSA Production Office / NPO News Releases / Uranium Processing Facility Site Readiness Subproject Completed ... Uranium Processing Facility Site Readiness Subproject Completed on Time and Under Budget The Uranium Processing Facility (UPF) project celebrates its first major milestone with the completion of site readiness work, delivered on time and under budget.

  1. The CNG process: Acid gas removal with liquid carbon dioxide

    SciTech Connect (OSTI)

    Liu, Y.C.; Auyang, L.; Brown, W.R.

    1987-01-01

    The CNG acid gas removal process has two unique features: the absorption of sulfur-containing compounds and other trace contaminants with liquid carbon dioxide, and the regeneration of pure liquid carbon dioxide by triple-point crystallization. The process is especially suitable for treating gases which contain large amounts of carbon dioxide and much smaller amounts (relative to carbon dioxide) of hydrogen sulfide. Capital and energy costs are lower than conventional solvent processes. Further, products of the CNG process meet stringent purity specifications without undue cost penalties. A process demonstration unit has been constructed and operated to demonstrate the two key steps of the CNG process. Hydrogen sulfide and carbonyl sulfide removal from gas streams with liquid carbon dioxide absorbent to sub-ppm concentrations has been demonstrated. The production of highly purified liquid carbon dioxide (less than 0.1 ppm total contaminant) by triple-point crystallization also has been demonstrated.

  2. Evaluation of an automatic uranium titration system

    SciTech Connect (OSTI)

    Lewis, K.

    1980-01-01

    The titration system utilizes the constant current coulometric titration of Goldbeck and Lerner. U(VI) is reduced to U(IV) by Fe(II). V(V) is generated to titrate the U(IV), and the titration is followed potentiometrically. The evaluation shows that the recovery of uranium is 100% at the 40-mg level. The accuracy is generally +-0.10% or better. The smallest sample weight at which reliable results were obtained was 40 mg of uranium. Time for one analysis is 15 minutes. Advantages and disadvantages of the automated titrator are listed. (DLC)

  3. Aseismic design criteria for uranium enrichment plants

    SciTech Connect (OSTI)

    Beavers, J.E.

    1980-01-01

    In this paper technological, economical, and safety issues of aseismic design of uranium enrichment plants are presented. The role of management in the decision making process surrounding these issues is also discussed. The resolution of the issues and the decisions made by management are controlling factors in developing aseismic design criteria for any facility. Based on past experience in developing aseismic design criteria for the GCEP various recommendations are made for future enrichment facilities, and since uranium enrichment plants are members of the nuclear fuel cycle the discussion and recommendations presented herein are applicable to other nonreactor nuclear facilities.

  4. Uranium Marketing Annual Report - Energy Information Administration

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Marketing Annual Report With Data for 2014 | Release Date: May 13, 2015 | Next Release Date: May 2016 | full report Previous reports Year: 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 2003 2002 2001 2000 1999 1998 1997 1996 1995 1994 1993 1992 Go Uranium purchases and prices Owners and operators of U.S. civilian nuclear power reactors ("civilian owner/operators" or "COOs") purchased a total of 53 million pounds U3O8e (equivalent1) of deliveries from U.S.

  5. Uranium in the Savannah River Site environment

    SciTech Connect (OSTI)

    Evans, A.G.; Bauer, L.R.; Haselow, J.S.; Hayes, D.W.; Martin, H.L.; McDowell, W.L.; Pickett, J.B.

    1992-12-09

    The purpose of this report is to consolidate the history of environmental uranium studies conducted by SRS and to describe the status of uranium in the environment. The report is intended to be a living document'' that will be updated periodically. This draft issue, February 1992, documents studies that occurred from 1954 to 1989. Data in this report are taken primarily from annual and semiannual environmental reports for SRS. Semiannual reports were published from 1954 through 1962. Annual reports have been published since 1963. Occasionally unpublished data are included in this report for completeness.

  6. Uranium in the Savannah River Site environment

    SciTech Connect (OSTI)

    Evans, A.G.; Bauer, L.R.; Haselow, J.S.; Hayes, D.W.; Martin, H.L.; McDowell, W.L.; Pickett, J.B.

    1992-12-09

    The purpose of this report is to consolidate the history of environmental uranium studies conducted by SRS and to describe the status of uranium in the environment. The report is intended to be a ``living document`` that will be updated periodically. This draft issue, February 1992, documents studies that occurred from 1954 to 1989. Data in this report are taken primarily from annual and semiannual environmental reports for SRS. Semiannual reports were published from 1954 through 1962. Annual reports have been published since 1963. Occasionally unpublished data are included in this report for completeness.

  7. METHOD OF HOT ROLLING URANIUM METAL

    DOE Patents [OSTI]

    Kaufmann, A.R.

    1959-03-10

    A method is given for quickly and efficiently hot rolling uranium metal in the upper part of the alpha phase temperature region to obtain sound bars and sheets possessing a good surface finish. The uranium metal billet is heated to a temperature in the range of 1000 deg F to 1220 deg F by immersion iii a molten lead bath. The heated billet is then passed through the rolls. The temperature is restored to the desired range between successive passes through the rolls, and the rolls are turned down approximately 0.050 inch between successive passes.

  8. Electrochemical Membrane for Carbon Dioxide Separation and Power Generation

    Office of Scientific and Technical Information (OSTI)

    (Conference) | SciTech Connect Conference: Electrochemical Membrane for Carbon Dioxide Separation and Power Generation Citation Details In-Document Search Title: Electrochemical Membrane for Carbon Dioxide Separation and Power Generation uelCell Energy, Inc. (FCE) has developed a novel system concept for separation of carbon dioxide (CO2) from greenhouse gas (GHG) emission sources using an electrochemical membrane (ECM). The salient feature of the ECM is its capability to produce electric

  9. A Novel System for Carbon Dioxide Capture Utilizing Electrochemical

    Office of Scientific and Technical Information (OSTI)

    Membrane Technology (Journal Article) | SciTech Connect Journal Article: A Novel System for Carbon Dioxide Capture Utilizing Electrochemical Membrane Technology Citation Details In-Document Search Title: A Novel System for Carbon Dioxide Capture Utilizing Electrochemical Membrane Technology FuelCell Energy, Inc. (FCE), in collaboration with Pacific Northwest National Laboratory (PNNL) and URS Corporation, is developing a novel Combined Electric Power and Carbon-Dioxide Separation (CEPACS)

  10. Carbon Dioxide Capture at a Reduced Cost - Energy Innovation Portal

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

    Industrial Technologies Industrial Technologies Find More Like This Return to Search Carbon Dioxide Capture at a Reduced Cost Lawrence Berkeley National Laboratory Contact LBL About This Technology Technology Marketing Summary Scientists at Berkeley Lab have developed a method that reduces the expense of capturing carbon dioxide generated by the combustion of fossil fuels. This technology would allow power plants and the chemical and cement industries to better sequester carbon dioxide and

  11. Haverford College Researchers Create Carbon Dioxide-Separating Polymer

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

    Haverford College Researchers Create Carbon Dioxide-Separating Polymer Haverford College Researchers Create Carbon Dioxide-Separating Polymer August 1, 2012 Rebecca Raber, rraber@haverford.edu, +1 610 896 1038 gtoc.jpg Carbon dioxide gas separation is important for many environmental and energy applications. Molecular dynamics simulations are used to characterize a two-dimensional hydrocarbon polymer, PG-ES1, that uses a combination of surface adsorption and narrow pores to separate carbon

  12. Project Profile: High-Efficiency Receivers for Supercritical Carbon Dioxide

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

    Cycles | Department of Energy Receivers for Supercritical Carbon Dioxide Cycles Project Profile: High-Efficiency Receivers for Supercritical Carbon Dioxide Cycles Brayton logo Brayton Energy, under the 2012 SunShot Concentrating Solar Power (CSP) R&D FOA, is building and testing a new solar receiver that uses supercritical carbon dioxide (s-CO2) as the heat-transfer fluid. The research team is designing the receiver to withstand higher operating temperatures and pressures than

  13. ARM - Lesson Plans: Plant Growth and Carbon Dioxide

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

    Plant Growth and Carbon Dioxide Outreach Home Room News Publications Traditional Knowledge Kiosks Barrow, Alaska Tropical Western Pacific Site Tours Contacts Students Study Hall About ARM Global Warming FAQ Just for Fun Meet our Friends Cool Sites Teachers Teachers' Toolbox Lesson Plans Lesson Plans: Plant Growth and Carbon Dioxide Objective The objective is to show how carbon dioxide in the air affects plant growth. Materials Each group of students will need the following: Graph paper Pencil

  14. Case Study: Transcritical Carbon Dioxide Supermarket Refrigeration Systems

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

    | Department of Energy Transcritical Carbon Dioxide Supermarket Refrigeration Systems Case Study: Transcritical Carbon Dioxide Supermarket Refrigeration Systems This case study documents one year of operating experience with a transcritical carbon dioxide (TC CO2) booster refrigeration system at Delhaize America's Hannaford supermarket location in Turner, Maine. This supermarket, which began operation in June 2013, is the first supermarket installation in the U.S. of a TC CO2 booster

  15. 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.; Stogov, Yu. V.

    2014-12-15

    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.

  16. Selection and Characterization of Carbon Black and Surfactants for Development of Small Scale Uranium Oxicarbide Kernels

    SciTech Connect (OSTI)

    Contescu, Cristian I

    2006-01-01

    This report supports the effort for development of small scale fabrication of UCO (a mixture of UO{sub 2} and UC{sub 2}) fuel kernels for the generation IV high temperature gas reactor program. In particular, it is focused on optimization of dispersion conditions of carbon black in the broths from which carbon-containing (UO{sub 2} {center_dot} H{sub 2}O + C) gel spheres are prepared by internal gelation. The broth results from mixing a hexamethylenetetramine (HMTA) and urea solution with an acid-deficient uranyl nitrate (ADUN) solution. Carbon black, which is previously added to one or other of the components, must stay dispersed during gelation. The report provides a detailed description of characterization efforts and results, aimed at identification and testing carbon black and surfactant combinations that would produce stable dispersions, with carbon particle sizes below 1 {micro}m, in aqueous HMTA/urea and ADUN solutions. A battery of characterization methods was used to identify the properties affecting the water dispersability of carbon blacks, such as surface area, aggregate morphology, volatile content, and, most importantly, surface chemistry. The report introduces the basic principles for each physical or chemical method of carbon black characterization, lists the results obtained, and underlines cross-correlations between methods. Particular attention is given to a newly developed method for characterization of surface chemical groups on carbons in terms of their acid-base properties (pK{sub a} spectra) based on potentiometric titration. Fourier-transform infrared (FTIR) spectroscopy was used to confirm the identity of surfactants, both ionic and non-ionic. In addition, background information on carbon black properties and the mechanism by which surfactants disperse carbon black in water is also provided. A list of main physical and chemical properties characterized, samples analyzed, and results obtained, as well as information on the desired trend or range of values generally associated with better dispersability, is provided in the Appendix. Special attention was given to characterization of several surface-modified carbon blacks produced by Cabot Corporation through proprietary diazonium salts chemistry. As demonstrated in the report, these advanced carbons offer many advantages over traditional dispersions. They disperse very easily, do not require intensive mechanical shearing or sonication, and the particle size of the dispersed carbon black aggregates is in the target range of 0.15-0.20 {micro}m. The dispersions in water and HMTA/urea solutions are stable for at least 30 days; in conditions of simulated broth, the dispersions are stable for at least 6 hours. It is proposed that the optimization of the carbon black dispersing process is possible by replacing traditional carbon blacks and surfactants with surface-modified carbon blacks having suitable chemical groups attached on their surface. It is recognized that the method advanced in this report for optimizing the carbon black dispersion process is based on a limited number of tests made in aqueous and simulated broth conditions. The findings were corroborated by a limited number of tests carried out with ADUN solutions by the Nuclear Science and Technology Division at Oak Ridge National Laboratory (ORNL). More work is necessary, however, to confirm the overall recommendation based on the findings discussed in this report: namely, that the use of surface-modified carbon blacks in the uranium-containing broth will not adversely impact the chemistry of the gelation process, and that high quality uranium oxicarbide (UCO) kernels will be produced after calcination.

  17. Haverford College Researchers Create Carbon Dioxide-Separating...

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

    dioxide gas separation is important for many environmental and energy applications. Molecular dynamics simulations are used to characterize a two-dimensional hydrocarbon...

  18. Molecular Simulation of Carbon Dioxide Brine and Clay Mineral...

    Office of Scientific and Technical Information (OSTI)

    Title: Molecular Simulation of Carbon Dioxide Brine and Clay Mineral Interactions and Determination of Contact Angles. Abstract not provided. Authors: Tenney, Craig M ; Cygan, ...

  19. Molecular Simulation of Carbon Dioxide Nanodroplets on Clay in...

    Office of Scientific and Technical Information (OSTI)

    Title: Molecular Simulation of Carbon Dioxide Nanodroplets on Clay in Deep Saline Aquifers. Authors: Tenney, Craig M. Publication Date: 2012-06-01 OSTI Identifier: 1073284 Report ...

  20. Molecular Simulation of Carbon Dioxide Nanodroplets on Clay Surfaces...

    Office of Scientific and Technical Information (OSTI)

    Title: Molecular Simulation of Carbon Dioxide Nanodroplets on Clay Surfaces in Deep Saline Aquifers. Authors: Tenney, Craig M. Publication Date: 2013-01-01 OSTI Identifier: 1063603 ...

  1. Comparison of methods for geologic storage of carbon dioxide...

    Office of Scientific and Technical Information (OSTI)

    Journal Article: Comparison of methods for geologic storage of carbon dioxide in saline formations Citation Details In-Document Search Title: Comparison of methods for geologic...

  2. Carbon Dioxide Information Analysis Center (CDIAC)-Fossil Fuel...

    Open Energy Info (EERE)

    Fuel CO2 Emissions Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Carbon Dioxide Information Analysis Center (CDIAC)-Fossil Fuel CO2 Emissions AgencyCompany...

  3. Carbon Dioxide Emissions Associated with Bioenergy and Other...

    Open Energy Info (EERE)

    and Other Biogenic Sources Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Carbon Dioxide Emissions Associated with Bioenergy and Other Biogenic Sources AgencyCompany...

  4. CarBen Version 3: Multisector Carbon Dioxide Emissions Accounting...

    Open Energy Info (EERE)

    Name: CarBen Version 3: Multisector Carbon Dioxide Emissions Accounting Tool Focus Area: Geothermal Power Topics: Policy, Deployment, & Program Impact Website: www.netl.doe.gov...

  5. High-Efficiency Receivers for Supercritical Carbon Dioxide Cycles

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

    Brayton Energy's supercritical carbon dioxide (s-CO 2 ) solar receiver has the potential to significantly improve reliability, increase efficiency, and reduce costs of CSP systems. ...

  6. Beneficial Use of Carbon Dioxide in Precast Concrete Production...

    Office of Scientific and Technical Information (OSTI)

    in binding matrix. Two typical precast products are examined for their capacity to store carbon dioxide during the production. They are concrete blocks and fiber-cement panels. ...

  7. Carbon dioxide absorbent and method of using the same

    SciTech Connect (OSTI)

    Perry, Robert James; O'Brien, Michael Joseph

    2014-06-10

    In accordance with one aspect, the present invention provides a composition which contains the amino-siloxane structures I, or III, as described herein. The composition is useful for the capture of carbon dioxide from process streams. In addition, the present invention provides methods of preparing the amino-siloxane composition. Another aspect of the present invention provides methods for reducing the amount of carbon dioxide in a process stream employing the amino-siloxane compositions of the invention, as species which react with carbon dioxide to form an adduct with carbon dioxide.

  8. Carbon Dioxide Geological Sequestration in Fractured Porous Rocks

    Office of Scientific and Technical Information (OSTI)

    Training and Research on Probabilistic Hydro-Thermo-Mechanical Modeling of Carbon Dioxide Geological Sequestration in Fractured Porous Rocks Gutierrez, Marte 54 ENVIRONMENTAL...

  9. Carbon Dioxide Capture and Storage Demonstration in Developing...

    Open Energy Info (EERE)

    Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Carbon Dioxide Capture and Storage Demonstration in Developing Countries: Analysis of Key Policy Issues and Barriers...

  10. Method for carbon dioxide sequestration (Patent) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    interbeds, providing an injection well into the formation and injecting supercritical carbon dioxide (SC--CO.sub.2) into the injection well under conditions of ...

  11. Beneficial Use of Carbon Dioxide in Precast Concrete Production...

    Office of Scientific and Technical Information (OSTI)

    of Carbon Dioxide in Precast Concrete Production Shao, Yixin 36 MATERIALS SCIENCE Clean Coal Technology Coal - Environmental Processes Clean Coal Technology Coal - Environmental...

  12. High Performance Composite Membranes for Separation of Carbon Dioxide from

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

    Methane | Center for Gas SeparationsRelevant to Clean Energy Technologies | Blandine Jerome High Performance Composite Membranes for Separation of Carbon Dioxide from Methane

  13. Innovative Concepts for Beneficial Reuse of Carbon Dioxide | Department of

    Office of Environmental Management (EM)

    Energy Innovative Concepts for Beneficial Reuse of Carbon Dioxide Innovative Concepts for Beneficial Reuse of Carbon Dioxide Funding for 12 projects to test innovative concepts for the beneficial use of carbon dioxide (CO2) was announced by the U.S. Department of Energy. The awards are part of $1.4 billion in funding from the American Recovery and Reinvestment Act (ARRA) for projects that will capture carbon dioxide from industrial sources. These 12 projects will engage in a first phase

  14. Carbon dioxide absorbent and method of using the same

    DOE Patents [OSTI]

    Perry, Robert James; O'Brien, Michael Joseph

    2015-12-29

    In accordance with one aspect, the present invention provides a composition which contains the amino-siloxane structures I, or III, as described herein. The composition is useful for the capture of carbon dioxide from process streams. In addition, the present invention provides methods of preparing the amino-siloxane composition. Another aspect of the present invention provides methods for reducing the amount of carbon dioxide in a process stream employing the amino-siloxane compositions of the invention, as species which react with carbon dioxide to form an adduct with carbon dioxide.

  15. Possibility of nuclear pumped laser experiment using low enriched uranium

    SciTech Connect (OSTI)

    Obara, Toru; Takezawa, Hiroki [Center for Research into Innovative Nuclear Energy Systems Tokyo Institute of Technology 2-12-1-N1-19, Ookayama Meguro-ku, Tokyo 152-8550 (Japan)

    2012-06-06

    Possibility to perform experiments for nuclear pumped laser oscillation by using low enriched uranium is investigated. Kinetic analyses are performed for two types of reactor design, one is using highly enriched uranium and the other is using low enriched uranium. The reactor design is based on the experiment reactor in IPPE. The results show the oscillation of nuclear pumped laser in the case of low enriched uranium reactor is also possible. The use of low enriched uranium in the experiment will make experiment easier.

  16. Method for extracting and sequestering carbon dioxide

    DOE Patents [OSTI]

    Rau, Gregory H. (Castro Valley, CA); Caldeira, Kenneth G. (Livermore, CA)

    2005-05-10

    A method and apparatus to extract and sequester carbon dioxide (CO.sub.2) from a stream or volume of gas wherein said method and apparatus hydrates CO.sub.2, and reacts the resulting carbonic acid with carbonate. Suitable carbonates include, but are not limited to, carbonates of alkali metals and alkaline earth metals, preferably carbonates of calcium and magnesium. Waste products are metal cations and bicarbonate in solution or dehydrated metal salts, which when disposed of in a large body of water provide an effective way of sequestering CO.sub.2 from a gaseous environment.

  17. Apparatus for extracting and sequestering carbon dioxide

    DOE Patents [OSTI]

    Rau, Gregory H. (Castro Valley, CA); Caldeira, Kenneth G. (Livermore, CA)

    2010-02-02

    An apparatus and method associated therewith to extract and sequester carbon dioxide (CO.sub.2) from a stream or volume of gas wherein said apparatus hydrates CO.sub.2 and reacts the resulting carbonic acid with carbonate. Suitable carbonates include, but are not limited to, carbonates of alkali metals and alkaline earth metals, preferably carbonates of calcium and magnesium. Waste products are metal cations and bicarbonate in solution or dehydrated metal salts, which when disposed of in a large body of water provide an effective way of sequestering CO.sub.2 from a gaseous environment.

  18. Capture of carbon dioxide by hybrid sorption

    DOE Patents [OSTI]

    Srinivasachar, Srivats

    2014-09-23

    A composition, process and system for capturing carbon dioxide from a combustion gas stream. The composition has a particulate porous support medium that has a high volume of pores, an alkaline component distributed within the pores and on the surface of the support medium, and water adsorbed on the alkaline component, wherein the proportion of water in the composition is between about 5% and about 35% by weight of the composition. The process and system contemplates contacting the sorbent and the flowing gas stream together at a temperature and for a time such that some water remains adsorbed in the alkaline component when the contact of the sorbent with the flowing gas ceases.

  19. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

    SciTech Connect (OSTI)

    Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

    2007-08-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

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

    SciTech Connect (OSTI)

    1995-07-05

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