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Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

Uranium dioxide electrolysis  

DOE Patents [OSTI]

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.

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

2009-12-29T23:59:59.000Z

2

Molten uranium dioxide structure and dynamics  

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

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.

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

2014-11-20T23:59:59.000Z

3

Standard specification for sintered gadolinium oxide-uranium dioxide pellets  

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

4

THE ENERGY SPECTRA OF URANIUM ATOMS SPUTTERED FROM URANIUM METAL AND URANIUM DIOXIDE TARGETS  

E-Print Network [OSTI]

THE ENERGY SPECTRA OF URANIUM ATOMS SPUTTERED FROM URANIUM METAL AND URANIUM DIOXIDE TARGETS Thesis. I have benefitted from conversations with many persons w~ile engaged in this project. I would like

Winfree, Erik

5

Dry process fluorination of uranium dioxide using ammonium bifluoride  

E-Print Network [OSTI]

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

Yeamans, Charles Burnett, 1978-

2003-01-01T23:59:59.000Z

6

Molecular Dynamics Simulation of Thermodynamic Properties in Uranium Dioxide  

SciTech Connect (OSTI)

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

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

2014-03-01T23:59:59.000Z

7

Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide  

SciTech Connect (OSTI)

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.

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

2012-07-31T23:59:59.000Z

8

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

SciTech Connect (OSTI)

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.

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

2014-01-01T23:59:59.000Z

9

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

SciTech Connect (OSTI)

Conclusions 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.

K. Gofryk; M. Jaime

2014-12-01T23:59:59.000Z

10

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

SciTech Connect (OSTI)

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.

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

2009-11-01T23:59:59.000Z

11

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2007-01-01T23:59:59.000Z

12

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

1999-01-01T23:59:59.000Z

13

An analysis of the impact of having uranium dioxide mixed in with plutonium dioxide  

SciTech Connect (OSTI)

An assessment was performed to show the impact on airborne release fraction, respirable fraction, dose conversion factor and dose consequences of postulated accidents at the Plutonium Finishing Plant involving uranium dioxide rather than plutonium dioxide.

MARUSICH, R.M.

1998-10-21T23:59:59.000Z

14

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2000-01-01T23:59:59.000Z

15

Radiative heat transfer in porous uranium dioxide  

SciTech Connect (OSTI)

Due to low thermal conductivity and high emissivity of UO{sub 2}, it has been suggested that radiative heat transfer may play a significant role in heat transfer through pores of UO{sub 2} fuel. This possibility was computationally investigated and contribution of radiative heat transfer within pores to overall heat transport in porous UO{sub 2} quantified. A repeating unit cell was developed to model approximately a porous UO{sub 2} fuel system, and the heat transfer through unit cells representing a wide variety of fuel conditions was calculated using a finite element computer program. Conduction through solid fuel matrix as wekk as pore gas, and radiative exchange at pore surface was incorporated. A variety of pore compositions were investigated: porosity, pore size, shape and orientation, temperature, and temperature gradient. Calculations were made in which pore surface radiation was both modeled and neglected. The difference between yielding the integral contribution of radiative heat transfer mechanism to overall heat transport. Results indicate that radiative component of heat transfer within pores is small for conditions representative of light water reactor fuel, typically less than 1% of total heat transport. It is much larger, however, for conditions present in liquid metal fast breeder reactor fuel; during restructuring of this fuel type early in life, the radiative heat transfer mode was shown to contribute as much as 10-20% of total heat transport in hottest regions of fuel.

Hayes, S.L. [Texas A and M Univ., College Station, TX (United States)] [Texas A and M Univ., College Station, TX (United States)

1992-12-01T23:59:59.000Z

16

Surface blistering and flaking of sintered uranium dioxide samples under high dose gas implantation and annealing  

E-Print Network [OSTI]

Surface blistering and flaking of sintered uranium dioxide samples under high dose gas implantation-sur-Yvette, France. a guillaume.martin@cea.fr Keywords: uranium dioxide, helium, hydrogen, implantation, blistering, flaking Abstract. High helium contents will be generated within minor actinide doped uranium dioxide

Boyer, Edmond

17

Fire hazards analysis for the uranium oxide (UO{sub 3}) facility  

SciTech Connect (OSTI)

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

Wyatt, D.M.

1994-12-06T23:59:59.000Z

18

Process for continuous production of metallic uranium and uranium alloys  

DOE Patents [OSTI]

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.

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

1995-06-06T23:59:59.000Z

19

Process for continuous production of metallic uranium and uranium alloys  

DOE Patents [OSTI]

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.

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

1995-01-01T23:59:59.000Z

20

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

SciTech Connect (OSTI)

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

Mac Donald, Philip Elsworth

2002-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Standard specification for sintered (Uranium-Plutonium) dioxide pellets  

E-Print Network [OSTI]

1.1 This specification covers finished sintered and ground (uranium-plutonium) dioxide pellets for use in thermal reactors. It applies to uranium-plutonium dioxide pellets containing plutonium additions up to 15 % weight. This specification may not completely cover the requirements for pellets fabricated from weapons-derived plutonium. 1.2 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all applicable international, federal, state, and local regulations pertaining to possessing, processing, shipping, or using source or special nuclear material. Examples of U.S. government documents are Code of Federal Regulations Title 10, Part 50Domestic Licensing of Production and Utilization Facilities; Code of Federal Regulations Title 10, Part 71Packaging and Transportation of Radioactive Material; and Code of Federal Regulations Tit...

American Society for Testing and Materials. Philadelphia

2001-01-01T23:59:59.000Z

22

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

DOE Patents [OSTI]

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.

Stinton, David P. (Knoxville, TN)

1983-01-01T23:59:59.000Z

23

Heavy Ion Beam in Resolution of the Critical Point Problem for Uranium and Uranium Dioxide  

E-Print Network [OSTI]

Important advantages of heavy ion beam (HIB) irradiation of matter are discussed in comparison with traditional sources - laser heating, electron beam, electrical discharge etc. High penetration length (~ 10 mm) is of primary importance for investigation of dense matter properties. This gives an extraordinary chance to reach the uniform heating regime when HIB irradiation is being used for thermophysical property measurements. Advantages of HIB heating of highly-dispersive samples are claimed for providing free and relatively slow quasi-isobaric heating without fast hydrodynamic expansion of heated sample. Perspective of such HIB application are revised for resolution of long-time thermophysical problems for uranium and uranium-bearing compounds (UO2). The priorities in such HIB development are stressed: preferable energy levels, beam-time duration, beam focusing, deposition of the sample etc.

Igor Iosilevskiy; Victor Gryaznov

2010-05-23T23:59:59.000Z

24

THERMODYNAMIC MODEL FOR URANIUM DIOXIDE BASED NUCLEAR FUEL  

SciTech Connect (OSTI)

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

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

2007-01-01T23:59:59.000Z

25

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2006-01-01T23:59:59.000Z

26

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

SciTech Connect (OSTI)

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.

Nelson, Andrew T. [Los Alamos National Laboratory

2012-08-30T23:59:59.000Z

27

Spark Plasma Sintering of W-UO2 Cermets  

SciTech Connect (OSTI)

About 50 vol.% 3 um depleted uranium dioxide (UO2) powder was encapsulated within a tungsten super alloy matrix produced from sub-micron tungsten powders using the Spark Plasma Sintering (SPS) process. An additive of 25 atom-percent (at.%) rhenium was included within the tungsten matrix to improve the ductility and fracture toughness of the ceramic–metallic (cermet) matrix. Cermet fabrication to 97.9% of the theoretical cermet density was achieved by sintering at 1500 degrees C with 40 MPa of applied pressure for 20 min. The results presented are from the first known trials of W–UO2 and nuclear cermet production via SPS.

R. C. O'Brien; N. D. Jerred

2013-02-01T23:59:59.000Z

28

Extraction of Uranium from Aqueous Solutions Using Ionic Liquid and Supercritical Carbon Dioxide in Conjunction  

SciTech Connect (OSTI)

Uranyl ions (UO2)2+ in aqueous nitric acid solutions can be extracted into supercritical CO2 (sc-CO2) via an imidazolium-based ionic liquid using tri-n-butylphosphate (TBP) as a complexing agent. The transfer of uranium from the ionic liquid to the supercritical fluid phase was monitored by UV/Vis spectroscopy using a high-pressure fiberoptic cell. The form of the uranyl complex extracted into the supercritical CO2 phase was found to be UO2(NO3)2(TBP)2. The extraction results were confirmed by UV/Vis spectroscopy and by neutron activation analysis. This technique could potentially be used to extract other actinides for applications in the field of nuclear waste management.

Wang, Joanna S.; Sheaff, Chrystal N.; Yoon, Byunghoon; Addleman, Raymond S.; Wai, Chien M.

2009-01-01T23:59:59.000Z

29

High Thermal Conductivity UO2-BeO Nulcear Fuel: Neutronic Performance Assessments and Overview of Fabrication  

E-Print Network [OSTI]

The objective of this work was to evaluate a new high conductivity nuclear fuel form. Uranium dioxide (UO2) is a very effective nuclear fuel, but it’s performance is limited by its low thermal conductivity. The fuel concept considered here is a...

Naramore, Michael J

2010-08-03T23:59:59.000Z

30

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2005-01-01T23:59:59.000Z

31

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

SciTech Connect (OSTI)

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.

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

2012-05-15T23:59:59.000Z

32

Dissolution of metal oxides and separation of uranium from lanthanides and actinides in supercritical carbon dioxide  

SciTech Connect (OSTI)

This paper investigates the feasibility of extracting and separating uranium from lanthanides and other actinides by using supercritical fluid carbon dioxide (sc-CO{sub 2}) as a solvent modified with tri-n-butylphosphate (TBP) for the development of a counter current stripping technique, which would be a more efficient and environmentally benign technology for spent nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U, Pu, and Np) and europium were extracted in sc-CO{sub 2} 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, uranium/europium and uranium/plutonium extraction and separation in sc-CO{sub 2} modified with TBP is successful at nitric acid concentrations of less than 6 M and at nitric acid concentrations of less than 3 M with acetohydroxamic acid or oxalic acid, respectively. A scheme for recycling uranium from spent nuclear fuel by using sc-CO{sub 2} and counter current stripping columns is presented. (authors)

Quach, D.L.; Wai, C.M. [Department of Chemistry, University of Idaho, Moscow, Idaho 83844 (United States); Mincher, B.J. [Idaho National Lab, Idaho Falls, Idaho (United States)

2013-07-01T23:59:59.000Z

33

DISSOLUTION OF METAL OXIDES AND SEPARATION OF URANIUM FROM LANTHANIDES AND ACTINIDES IN SUPERCRITICAL CARBON DIOXIDE  

SciTech Connect (OSTI)

This paper investigates the feasibility of extracting and separating uranium from lanthanides and other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of a counter current stripping technique, which would be a more efficient and environmentally benign technology for spent nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U, Pu, and Np) and europium 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, uranium/europium and uranium/plutonium extraction and separation in sc-CO2 modified with TBP is successful at nitric acid concentrations of less than 6 M and at nitric acid concentrations of less than 3 M with acetohydroxamic acid or oxalic acid, respectively. A scheme for recycling uranium from spent nuclear fuel by using sc-CO2 and counter current stripping columns is presented.

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

2013-10-01T23:59:59.000Z

34

Final Report: Manganese Redox Mediation of UO2 Stability and Uranium Fate in the Subsurface: Molecular and Meter Scale Dynamics  

SciTech Connect (OSTI)

One strategy to remediate U contamination in the subsurface is the immobilization of U via injection of an electron donor, e.g., acetate, which leads to stimulation of the bioreduction of U(VI), the more mobile form of U, to U(IV), the less mobile form. This process is inevitably accompanied by the sequential reductive dissolution of Mn and Fe oxides and often continuing into sulfate-reducing conditions. When these reducing zones, which accumulate U(IV), experience oxidizing conditions, reduced Fe and Mn can be reoxidized forming Fe and Mn oxides that, along with O2, can impact the stability of U(IV). The focus of our project has been to investigate (i) the effects of Mn(II) on the dissolution of UO2 under both reducing and oxidizing conditions, (ii) the oxidative dissolution of UO2 by soluble Mn(III), (iii) the fate of U once it is oxidized by MnO2 in both laboratory and field settings, and (iv) the effects of groundwater constituents on the coupled Mn(II)/U(IV) oxidation process. Additionally, studies of the interaction of Se, found at the DOE site at Rifle, CO, and Mn cycling were initiated to understand if observed seasonal fluctuations of Se and Mn are directly linked and whether any such linkages can affect the stability of U(IV).

Tebo, Bradley M. [OSHU; Tebo, Bradley M.

2014-09-02T23:59:59.000Z

35

Method for fluorination of uranium oxide  

DOE Patents [OSTI]

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.

Petit, George S. (Oak Ridge, TN)

1987-01-01T23:59:59.000Z

36

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

SciTech Connect (OSTI)

Conclusions: 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.

K. Gofryk; N. Harrison; M. Jaime

2014-12-01T23:59:59.000Z

37

Etching of UO{sub 2} in NF{sub 3} RF Plasma Glow Discharge  

SciTech Connect (OSTI)

A series of room temperature, low pressure (10.8 to 40 Pa), low power (25 to 210 W) RF plasma glow discharge experiments with UO{sub 2} were conducted to demonstrate that plasma treatment is a viable method for decontaminating UO{sub 2} from stainless steel substrates. Experiments were conducted using NF{sub 3} gas to decontaminate depleted uranium dioxide from stainless-steel substrates. Depleted UO{sub 2} samples each containing 129.4 Bq were prepared from 100 microliter solutions of uranyl nitrate hexahydrate solution. The amorphous UO{sub 2} in the samples had a relatively low density of 4.8 gm/cm{sub 3}. Counting of the depleted UO{sub 2} on the substrate following plasma immersion was performed using liquid scintillation counting with alpha/beta discrimination due to the presence of confounding beta emitting daughter products, {sup 234}Th and {sup 234}Pa. The alpha emission peak from each sample was integrated using a gaussian and first order polynomial fit to improve quantification. The uncertainties in the experimental measurement of the etched material were estimated at about {+-} 2%. Results demonstrated that UO{sub 2} can be completely removed from stainless-steel substrates after several minutes processing at under 200 W. At 180 W and 32.7 Pa gas pressure, over 99% of all UO{sub 2} in the samples was removed in just 17 minutes. The initial etch rate in the experiments ranged from 0.2 to 7.4 {micro}m/min. Etching increased with the plasma absorbed power and feed gas pressure in the range of 10.8 to 40 Pa. A different pressure effect on UO{sub 2} etching was also noted below 50 W in which etching increased up to a maximum pressure, {approximately}23 Pa, then decreased with further increases in pressure.

John M. Veilleux

1999-08-01T23:59:59.000Z

38

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

SciTech Connect (OSTI)

Conclusions 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.

K. Gofryk; V. Zapf; M. Jaime

2014-12-01T23:59:59.000Z

39

Final Version: Orbital Specificity in the Unoccupied States of UO2 from Resonant Inverse Photoelectron Spectroscopy  

SciTech Connect (OSTI)

One of the crucial questions of all actinide electronic structure determinations is the issue of 5f versus 6d character and the distribution of these components across the density of states. Here, a break-though experiment is discussed, which has allowed the direct determination of the U5f and U6d contributions to the unoccupied density of states (UDOS) in Uranium Dioxide. A novel Resonant Inverse Photoelectron (RIPES) and X-ray Emission Spectroscopy (XES) investigation of UO{sub 2} is presented. It is shown that the U5f and U6d components are isolated and identified unambiguously.

Tobin, J G; Yu, S W

2012-03-12T23:59:59.000Z

40

Refinement in the ultrasonic velocity data and estimation of the critical parameters for molten uranium dioxide  

E-Print Network [OSTI]

accurate exper- imental measurements on the density, and heat capacity of liquid UO2 up to $8000 K density and isobaric heat capacity, much more easily than other conventional methods [3,4]. Many of state for liquid urania has also been developed which predicts a critical temperature (Tc) % 10500 K

Azad, Abdul-Majeed

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

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

SciTech Connect (OSTI)

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 1000, 1300, 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 samples.

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

2015-01-01T23:59:59.000Z

42

Cameco UO3 Materials Analysis  

SciTech Connect (OSTI)

Uranium trioxide (UO{sub 3}) was characterized using a variety of techniques to better understand its physical properties. Scanning electron microscope (SEM) images were collected to examine particle morphology, which consisted of semi-spherical particles that tended to agglomerate before sonication. Particle size analysis revealed a singular mode distribution with a mean particle size of 43.0 {micro}m. After sonication a bimodal distribution was produced with peak particle sizes at 0.226 {micro}m and 9.43 {micro}m. The O/U ratio was measured to be 3.09 by Cameco in 2009, by gravimetric analysis. X-ray diffraction (XRD) showed that the sample was mostly {gamma}-UO{sub 3} (87.1%) with a small amount of UO{sub 3} {center_dot} 0.80 H{sub 2}O (12.9%). Bulk and tap densities were determined to be 3.678 {+-} 0.2 and 4.81 {+-} 0.2 g/cm{sup 3}, respectively (crystalline density is 7.3 g/cm{sup 3}). The stoichiometry was measured to be 2.99 in 2012.

Hill, Mary Ann [Los Alamos National Laboratory; Nolen, Blake Penfield [Los Alamos National Laboratory; Wermer, Joseph R. [Los Alamos National Laboratory; Wilkerson, Marianne P. [Los Alamos National Laboratory; Fredenburg, David A. [Los Alamos National Laboratory; Wagner, Gregory L. [Los Alamos National Laboratory; Papin, Pallas A. [Los Alamos National Laboratory; Scott, Brian L. [Los Alamos National Laboratory; Guidry, Dennis Ray [Los Alamos National Laboratory

2012-07-12T23:59:59.000Z

43

Possible effects of UO/sub 2/ oxidation on light water reactor spent fuel performance in long-term geologic disposal  

SciTech Connect (OSTI)

Disposal of spent nuclear fuel in a conventionally mined geologic formation is the nearest-term option for permanently isolating radionuclides from the biosphere. Because irradiated uranium dioxide (UO/sub 2/) fuel pellets retain 95 to 99% of the radionuclides generated during normal light water reactor operation, they may represent a significant barrier to radionuclide release. This document presents a technical assessment of published literature representing the current level of understanding of spent fuel characteristics and conditions that may degrade pellet integrity during a geologic disposal sequence. A significant deterioration mechanism is spent UO/sub 2/ oxidation with possible consequences identified as fission gas release, rod diameter increases, cladding breach extension, and release of solid fuel particles containing radionuclides. Areas requiring further study to support development of a comprehensive spent fuel performance prediction model are highlighted. A program and preliminary schedule to obtain the information needed to develop model correlations are also presented.

Almassy, M.Y.; Woodley, R.E.

1982-08-01T23:59:59.000Z

44

Raman Investigation of The Uranium Compounds U3O8, UF4, UH3 and UO3 under Pressure at Room Temperature  

SciTech Connect (OSTI)

Our current state-of-the-art X-ray diffraction experiments are primarily sensitive to the position of the uranium atom. While the uranium - low-Z element bond (such as U-H or U-F) changes under pressure and temperature the X-ray diffraction investigations do not reveal information about the bonding or the stoichiometry. Questions that can be answered by Raman spectroscopy are (i) whether the bonding strength changes under pressure, as observed by either blue- or red-shifted peaks of the Raman active bands in the spectrum and (ii) whether the low-Z element will eventually be liberated and leave the host lattice, i.e. do the fluorine, oxygen, or hydrogen atoms form dimers after breaking the bond to the uranium atom. Therefore Raman spectra were also collected in the range where those decomposition products would appear. Raman is particularly well suited to these types of investigations due to its sensitivity to trace amounts of materials. One challenge for Raman investigations of the uranium compounds is that they are opaque to visible light. They absorb the incoming radiation and quickly heat up to the point of decomposition. This has been dealt with in the past by keeping the incoming laser power to very low levels on the tens of milliWatt range consequently affecting signal to noise. Recent modern investigations also used very small laser spot sizes (micrometer range) but ran again into the problem of heating and chemical sensitivity to the environment. In the studies presented here (in contrast to all other studies that were performed at ambient conditions only) we employ micro-Raman spectroscopy of samples situated in a diamond anvil cell. This increases the trustworthiness of the obtained data in several key-aspects: (a) We surrounded the samples in the DAC with neon as a pressure transmitting medium, a noble gas that is absolutely chemically inert. (b) Through the medium the sample is thermally heat sunk to the diamond anvils, diamond of course possessing the very best heat conductivity of any material. Therefore local heating and decomposition are avoided, a big challenge with other approaches casting doubts on their results. (c) This in turn benefits the signal/noise ratio tremendously since the Raman features of uranium-compounds are very small. The placement of the samples in DACs allows for higher laser powers to impinge on the sample spot while keeping the spot-size larger than in previous studies and keep the samples from heating up. Raman spectroscopy is a very sensitive non-invasive technique and we will show that it is even possible to distinguish the materials by their origin / manufacturer as we have studied samples from Cameco (Canada) and IBI-Labs (US-Florida) and can compare with ambient literature data for samples from Strem (US-MA) and Areva (Pierrelatte, France).

Lipp, M J; Jenei, Z; Park-Klepeis, J; Evans, W J

2011-12-15T23:59:59.000Z

45

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

SciTech Connect (OSTI)

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.

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

2014-01-01T23:59:59.000Z

46

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

SciTech Connect (OSTI)

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.

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

2014-03-10T23:59:59.000Z

47

PUREX/UO{sub 3} deactivation project management plan  

SciTech Connect (OSTI)

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.

Washenfelder, D.J.

1993-12-01T23:59:59.000Z

48

Electrolytic process for preparing uranium metal  

DOE Patents [OSTI]

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.

Haas, Paul A. (Knoxville, TN)

1990-01-01T23:59:59.000Z

49

UoS Motor Accident Report Form COMPANY DETAILS  

E-Print Network [OSTI]

UNIV01FL02 UoS Motor Accident Report Form COMPANY DETAILS INSURED: University of Sussex ADDRESS: LOCATION: DESCRIPTION OF HOW ACCIDENT HAPPENED: PLEASE DRAW A SKETCH OF THE ACCIDENT: #12;DRIVER DETAILS: PREVIOUS ACCIDENTS: ADDRESS: VEHICLE DETAILS DATE VEHICLE PURCHASED: MAKE/MODEL: REGISTRATION: MILEAGE

Sussex, University of

50

Aquifer restoration at in-situ leach uranium mines: evidence for natural restoration processes  

SciTech Connect (OSTI)

Pacific Northwest Laboratory conducted experiments with aquifer sediments and leaching solution (lixiviant) from an in-situ leach uranium mine. The data from these laboratory experiments and information on the normal distribution of elements associated with roll-front uranium deposits provide evidence that natural processes can enhance restoration of aquifers affected by leach mining. Our experiments show that the concentration of uranium (U) in solution can decrease at least an order of magnitude (from 50 to less than 5 ppM U) due to reactions between the lixiviant and sediment, and that a uranium solid, possibly amorphous uranium dioxide, (UO/sub 2/), can limit the concentration of uranium in a solution in contact with reduced sediment. The concentrations of As, Se, and Mo in an oxidizing lixiviant should also decrease as a result of redox and precipitation reactions between the solution and sediment. The lixiviant concentrations of major anions (chloride and sulfate) other than carbonate were not affected by short-term (less than one week) contact with the aquifer sediments. This is also true of the total dissolved solids level of the solution. Consequently, we recommend that these solution parameters be used as indicators of an excursion of leaching solution from the leach field. Our experiments have shown that natural aquifer processes can affect the solution concentration of certain constituents. This effect should be considered when guidelines for aquifer restoration are established.

Deutsch, W.J.; Serne, R.J.; Bell, N.E.; Martin, W.J.

1983-04-01T23:59:59.000Z

51

High temperature redox reactions with uranium: Synthesis and characterization of Cs(UO{sub 2})Cl(SeO{sub 3}), Rb{sub 2}(UO{sub 2}){sub 3}O{sub 2}(SeO{sub 3}){sub 2}, and RbNa{sub 5}U{sub 2}(SO{sub 4}){sub 7}  

SciTech Connect (OSTI)

Cs(UO{sub 2})Cl(SeO{sub 3}) (1), Rb{sub 2}(UO{sub 2}){sub 3}O{sub 2}(SeO{sub 3}){sub 3} (2), and RbNa{sub 5}U{sub 2}(SO{sub 4}){sub 7} (3) single crystals were synthesized using CsCl, RbCl, and a CuCl/NaCl eutectic mixture as fluxes, respectively. Their lattice parameters and space groups are as follows: P2{sub 1}/n (a=6.548(1) Ĺ, b=11.052(2) Ĺ, c=10.666(2) Ĺ and ?=93.897(3)°), P1{sup Ż} (a=7.051(2) Ĺ, b=7.198(2) Ĺ, c=8.314(2) Ĺ, ?=107.897(3)°, ?=102.687(3)° and ?=100.564(3)°) and C2/c (a=17.862(4) Ĺ, b=6.931(1) Ĺ, c=20.133(4) Ĺ and ?=109.737(6)°. The small anionic building units found in these compounds are SeO{sub 3}{sup 2?} and SO{sub 4}{sup 2?} tetrahedra, oxide, and chloride. The crystal structure of the first compound is composed of [(UO{sub 2}){sub 2}Cl{sub 2}(SeO{sub 3}){sub 2}]{sup 2?} chains separated by Cs{sup +} cations. The structure of (2) is constructed from [(UO{sub 2}){sub 3}O{sub 11}]{sup 16?} chains further connected through selenite units into layers stacked perpendicularly to the [0 1 0] direction, with Rb{sup +} cations intercalating between them. The structure of compound (3) is made of uranyl sulfate layers formed by edge and vertex connections between dimeric [U{sub 2}O{sub 16}] and [SO{sub 4}] polyhedra. These layers contain unusual sulfate–metal connectivity as well as large voids. - Graphical abstract: A new family of uranyl selenites and sulfates has been prepared by high-temperature redox reactions. This compounds display new bonding motifs. Display Omitted - Highlights: • Low-dimensional Uranyl Oxoanion compounds. • Conversion of U(IV) to U(VI) at high temperatures. • Dimensional reduction by both halides and stereochemically active lone-pairs.

Babo, Jean-Marie [Department of Chemistry and Biochemistry, Florida State University, 95 Chieftan Way, Tallahassee, FL 32306-4390 (United States); Department of Civil and Environmental Engineering and Earth Sciences and Department of Chemistry and Biochemistry, University of Notre Dame, Notre Dame, IN 46556 (United States); Albrecht-Schmitt, Thomas E., E-mail: talbrechtschmitt@gmail.com [Department of Chemistry and Biochemistry, Florida State University, 95 Chieftan Way, Tallahassee, FL 32306-4390 (United States)

2013-10-15T23:59:59.000Z

52

Structural evolution of the double perovskites Sr{sub 2}B'UO{sub 6} (B' = Mn, Fe, Co, Ni, Zn) upon reduction: Magnetic behavior of the uranium cations  

SciTech Connect (OSTI)

Highlights: {yields} Evolution of the double perovskites Sr{sub 2}B'UO{sub 6} upon reduction were studied by XRPD. {yields} Orthorhombic (Pnma) disordered perovskites SrB'{sub 0.5-x}U{sub 0.5+x}O{sub 3} were obtained at 900 {sup o}C. {yields} U{sup 5+/4+} and Zn{sup 2+} cations are distributed at random over the octahedral positions. {yields} AFM ordering for the perovskite with B' = Zn appears below 30 K. -- Abstract: We describe the preparation of five perovskite oxides obtained upon reduction of Sr{sub 2}B'UO{sub 6} (B' = Mn, Fe, Co, Ni, Zn) with H{sub 2}/N{sub 2} (5%/95%) at 900 {sup o}C during 8 h, and their structural characterization by X-ray powder diffraction (XRPD). During the reduction process there is a partial segregation of the elemental metal when B' = Co, Ni, Fe, and the corresponding B'O oxide when B' = Mn, Zn. Whereas the parent, oxygen stoichiometric double perovskites Sr{sub 2}B'UO{sub 6} are long-range ordered concerning B' and U cations. The crystal structures of the reduced phases, SrB'{sub 0.5-x}U{sub 0.5+x}O{sub 3} with 0.37 < x < 0.27, correspond to simple, disordered perovskites; they are orthorhombic, space group Pnma (No. 62), with a full cationic disorder at the B site. Magnetic measurements performed on the phase with B' = Zn, indicate uncompensated antiferromagnetic ordering of the U{sup 5+}/U{sup 4+} sublattice below 30 K.

Pinacca, R.M., E-mail: rmp@unsl.edu.ar [Area de Quimica General e Inorganica 'Dr. Gabino F. Puelles', Departamento de Quimica, Facultad de Quimica, Bioquimica y Farmacia, Universidad Nacional de San Luis, Chacabuco y Pedernera, 5700 San Luis (Argentina); Viola, M.C.; Pedregosa, J.C. [Area de Quimica General e Inorganica 'Dr. Gabino F. Puelles', Departamento de Quimica, Facultad de Quimica, Bioquimica y Farmacia, Universidad Nacional de San Luis, Chacabuco y Pedernera, 5700 San Luis (Argentina)] [Area de Quimica General e Inorganica 'Dr. Gabino F. Puelles', Departamento de Quimica, Facultad de Quimica, Bioquimica y Farmacia, Universidad Nacional de San Luis, Chacabuco y Pedernera, 5700 San Luis (Argentina); Carbonio, R.E. [INFIQC (CONICET), Departamento de Fisicoquimica, Facultad de Ciencias Quimicas, Universidad Nacional de Cordoba, Ciudad Universitaria, X5000HUA Cordoba (Argentina)] [INFIQC (CONICET), Departamento de Fisicoquimica, Facultad de Ciencias Quimicas, Universidad Nacional de Cordoba, Ciudad Universitaria, X5000HUA Cordoba (Argentina); Lope, M.J. Martinez; Alonso, J.A. [Instituto de Ciencia de Materiales de Madrid, C.S.I.C., Cantoblanco, 28049 Madrid (Spain)] [Instituto de Ciencia de Materiales de Madrid, C.S.I.C., Cantoblanco, 28049 Madrid (Spain)

2011-11-15T23:59:59.000Z

53

Bioremediation of Uranium Plumes with Nano-scale  

E-Print Network [OSTI]

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

Fay, Noah

54

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

SciTech Connect (OSTI)

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.

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

2010-01-29T23:59:59.000Z

55

Fundamental study on recovery uranium oxide from HEPA filters  

SciTech Connect (OSTI)

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

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

1993-12-31T23:59:59.000Z

56

Modeling of UO{sub 2} oxidation in steam atmosphere  

SciTech Connect (OSTI)

Nuclear fuel oxidation is an important phenomenon affecting fission product behavior. As indicated by a number of studies, uranium dioxide shows a very wide range of nonstoichiometric states. In steam, fuel oxidation produces a hyperstoichiometric composition, changing the transport properties. Variation of stoichiometry changes diffusion coefficients for oxygen, noble gases, and fission products substantially, affecting the release of fission products.

Dobrov, B.V.; Likhanskii, V.V. [Triniti Research Center, Triniti, Moscow (Russian Federation); Ozrin, V.D. [Nuclear Safety Institute IBREA, Moscow (Russian Federation)] [and others

1997-12-01T23:59:59.000Z

57

Standard test method for determination of impurities in nuclear grade uranium compounds by inductively coupled plasma mass spectrometry  

E-Print Network [OSTI]

1.1 This test method covers the determination of 67 elements in uranium dioxide samples and nuclear grade uranium compounds and solutions without matrix separation by inductively coupled plasma mass spectrometry (ICP-MS). The elements are listed in Table 1. These elements can also be determined in uranyl nitrate hexahydrate (UNH), uranium hexafluoride (UF6), triuranium octoxide (U3O8) and uranium trioxide (UO3) if these compounds are treated and converted to the same uranium concentration solution. 1.2 The elements boron, sodium, silicon, phosphorus, potassium, calcium and iron can be determined using different techniques. The analyst's instrumentation will determine which procedure is chosen for the analysis. 1.3 The test method for technetium-99 is given in Annex A1. 1.4 The values stated in SI units are to be regarded as standard. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish ...

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

58

TRIMOLECULAR REACTIONS OF URANIUM HEXAFLUORIDE WITH WATER  

SciTech Connect (OSTI)

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.

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

2010-02-25T23:59:59.000Z

59

UO Policy Library Resource for  

E-Print Network [OSTI]

UO Policy Library Resource for Policy Owners To ensure University- wide consistency in the formulation, review, approval, and implementation of policies, the Policy Library has provided a resource section for policy owners. It helps answer questions such as: Is this a policy or procedure? What

Oregon, University of

60

Conversion of depleted uranium hexafluoride to a solid uranium compound  

DOE Patents [OSTI]

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.

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

2001-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

UO  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched Ferromagnetism in Layeredof EnergyLeaseEnergyUNCLASSIFIED 2 1IsotopeFigure 1.

62

Identification of secondary phases formed during unsaturated reaction of UO{sub 2} with EJ-13 water  

SciTech Connect (OSTI)

A set of experiments, wherein UO{sub 2} has been contacted by dripping water, has been conducted over a period of 182.5 weeks. The experiments are being conducted to develop procedures to study spent fuel reaction under unsaturated conditions that are expected to exist over the lifetime of the proposed Yucca Mountain repository site. One half of the experiments have been terminated, while one half are ongoing. Analyses of solutions that have dripped from the reacted UO{sub 2} have been performed for all experiments, while the reacted UO{sub 2} surfaces have been examined for the terminated experiments. A pulse of uranium release from the UO{sub 2} solid, combined with the formation of schoepite on the surface of the UO{sub 2}, was observed between 39 and 96 weeks of reaction. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporated cations from the EJ-13 water and included boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are continuing to monitor whether additional changes in solution chemistry or secondary phase formation occurs. 6 refs., 2 figs., 2 tabs.

Bates, J.K.; Tani, B.S.; Veleckis, E.

1989-11-01T23:59:59.000Z

63

Identification of secondary phases formed during unsaturated reaction of UO{sub 2} with EJ-13 water  

SciTech Connect (OSTI)

A set of experiments, wherein UO{sub 2} has been contacted by dripping water, has been conducted over a period of 182.5 weeks. The experiments are being conducted to develop procedures to study spent fuel reaction under unsaturated conditions that are expected to exist over the lifetime of the proposed Yucca Mountain repository site. One half of the experiments have been terminated, while one half are ongoing. Analyses of solutions that have dripped from the reacted UO{sub 2} have been performed for all experiments, while reacted UO{sub 2} surfaces have been examined for the terminated experiments. A pulse of uranium release from the UO{sub 2} solid, combined with the formation of schoepite on the surface of the UO{sub 2}, was observed between 39 and 96 weeks of reaction. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporated cations from the EJ-13 water and include boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are continuing to monitor whether additional changes in solution chemistry or secondary phase formation occurs.

Bates, J.K.; Tani, B.S.; Veleckis, E.; Wronkiewicz, D.J. [Argonne National Lab., IL (USA)

1990-12-31T23:59:59.000Z

64

Depleted uranium hexafluoride: Waste or resource?  

SciTech Connect (OSTI)

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

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

1995-07-01T23:59:59.000Z

65

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

SciTech Connect (OSTI)

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.

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

2011-10-01T23:59:59.000Z

66

Extraction of uranium from spent fuels using liquefied gases  

SciTech Connect (OSTI)

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)

Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi [EcoTopia Science Institute, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, 464-8603 (Japan)

2007-07-01T23:59:59.000Z

67

Dissolution characteristics of mixed UO{sub 2} powders in J-13 water under saturated conditions  

SciTech Connect (OSTI)

The Yucca Mountain Project/Spent Fuel program at Argonne National Laboratory is designed to determine radionuclide release rates by exposing high-level waste to repository-relevant groundwater. To gain experience for the tests with spent fuel, a scoping experiment was conducted at room temperature to determine the uranium release rate from an unirradiated UO{sub 2} powder mixture (14.3 wt % enrichment in {sup 235}U) to J-13 water under saturated conditions. Another goal set for the experiment was to develop a method for utilizing isotope dilution techniques to determine whether the dissolution rate of UO{sub 2} matrix is in accordance with an existing kinetic model. Results of these analyses revealed unequal uranium dissolution rates from the enriched and depleted portions of the powder mixture because of undisclosed differences between them. Although the presence of this inhomogeneity has precluded the application of the kinetic model, it also provided an opportunity to elaborate on the utilization of isotope dilution data in recognizing and quantifying such conditions. Detailed listings of uranium release and solution chemistry data are presented. Other problems commonly associated with spent fuel, such as the effectiveness of filtering media, the existence of uranium concentration peaks during early stages of the leach tests, the need for concentration corrections due to water replenishments of sample volumes, and experience derived from isotope dilution data are discussed in the context of the present results. 10 refs., 5 figs., 7 tabs.

Veleckis, E.; Hoh, J.C.

1991-03-01T23:59:59.000Z

68

OXYGEN DIFFUSION IN UO2-x  

E-Print Network [OSTI]

~ K.C. K:i.m, "Oxygen Diffusion in Hypostoichiometricsystem for enriching uo 2 in oxygen-18 or for stoichiometry+nal of Nuclear Materials OXYGEN DIFFUSION IN U0 2 _:x K.C.

Kim, K.C.

2013-01-01T23:59:59.000Z

69

Bisphosphine dioxides  

DOE Patents [OSTI]

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.

Moloy, Kenneth G. (Charleston, WV)

1990-01-01T23:59:59.000Z

70

Bisphosphine dioxides  

DOE Patents [OSTI]

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.

Moloy, K.G.

1990-02-20T23:59:59.000Z

71

Crystal fields in UO2 - revisited  

SciTech Connect (OSTI)

We performed inelastic neutron scattering (INS) in order to re-investigate the crystal-field ground state and the level splitting in UO{sub 2}. Previous INS studies on UO{sub 2} by Amorelli et al. [Physical Review B 15, 1989, 1856] uncovered four excitations at low temperatures in the 150-180 meV range. Considering the dipole-allowed transitions, only three of these transitions could be explained by the published crystal-field model. Our INS results on a different UO{sub 2} sample revealed that the unaccounted peak at about 180 meV is a spurious one, and thus not intrinsic to UO{sub 2}. In good agreement with Amoretti's results, we corroborated that the ground-state of UO{sub 2} is the {Lambda}{sub 5} triplet, and we computed that the fourth- and six-order crystal field parameters are V{sub 4} = -116 meV and V{sub 6} = 26 meV, respectively. We also studied the INS response of the non-magnetic U{sub 0.4}Th{sub 0.6}O{sub 2}. The splitting for this thorium-doped compound is similar to the one of UO{sub 2}, which orders antiferromagnetically at low temperatures. Therefore, we can conclude that magnetic interactions only weakly perturb the energy level splitting, which is dominated by strong crystal fields.

Nakotte, Heinz [Los Alamos National Laboratory; Rajatram, R [NMSU/UNIV OF N.C.; Kern, S [COLORADO STATE UNIV; Mcqueeney, R J [AMES LAB; Lander, G H [EUROPEAN COMMISIONS, JRC; Robinson, R A [BRAGG INSTITUTE

2009-01-01T23:59:59.000Z

72

Uranium Ore Uranium is extracted  

E-Print Network [OSTI]

Milling of Uranium Ore Uranium is extracted from ore with strong acids or bases. The uranium is concentrated in a solid substance called"yellowcake." Chemical Conversion Plants convert the uranium in yellowcake to uranium hexafluoride (UF6 ), a compound that can be made into nuclear fuel. Enrichment

73

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

SciTech Connect (OSTI)

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.

Carnahan, C.L.

1987-12-01T23:59:59.000Z

74

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

SciTech Connect (OSTI)

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.

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

2008-07-08T23:59:59.000Z

75

Surface Decontamination of System Components in Uranium Conversion Plant at KAERI  

SciTech Connect (OSTI)

A chemical decontamination process using nitric acid solution was selected as in-situ technology for recycle or release with authorization of a large amount of metallic waste including process system components such as tanks, piping, etc., which is generated by dismantling a retired uranium conversion plant at Korea Atomic Energy Research Institute (KAERI). The applicability of nitric acid solution for surface decontamination of metallic wastes contaminated with uranium compounds was evaluated through the basic research on the dissolution of UO2 and ammonium uranyl carbonate (AUC) powder. Decontamination performance was verified by using the specimens contaminated with such uranium compounds as UO2 and AUC taken from the uranium conversion plant. Dissolution rate of UO2 powder is notably enhanced by the addition of H2O2 as an oxidant even in the condition of a low concentration of nitric acid and low temperature compared with those in a nitric acid solution without H2O2. AUC powders dissolve easily in nitric acid solutions until the solution pH attains about 2.5 {approx} 3. Above that solution pH, however, the uranium concentration in the solution is lowered drastically by precipitation as a form of U3(NH3)4O9 . 5H2O. Decontamination performance tests for the specimens contaminated with UO2 and AUC were quite successful with the application of decontamination conditions obtained through the basic studies on the dissolution of UO2 and AUC powders.

Choi, W. K.; Kim, K. N.; Won, H. J.; Jung, C. H.; Oh, W. Z.

2003-02-25T23:59:59.000Z

76

Thorium dioxide: properties and nuclear applications  

SciTech Connect (OSTI)

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.

Belle, J.; Berman, R.M. (eds.)

1984-01-01T23:59:59.000Z

77

Leaching action of EJ-13 water on unirradiated UO{sub 2} surfaces under unsaturated conditions at 90{degree}C: Interim report  

SciTech Connect (OSTI)

A set of experiments, based on the application of the Unsaturated Test method to the reaction of UO{sub 2} with EJ-13 water, has been conducted over a period of 182.5 weeks. One half of the experiments have been terminated, while one half are still ongoing. Solutions that have dripped from UO{sub 2} specimens have been analyzed for all experiments, while the reacted UO{sub 2} surfaces have been examined for only the terminated experiments. A pulse of uranium release from the UO{sub 2} solid, in conjunction with the formation of dehydrated schoepite on the surface of the UO{sub 2}, was observed during the 39- to 96-week period. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporate cations from the EJ-13 water and include boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are being continued to monitor for additional changes in solution composition and secondary phase formation, and have now reached the 319-week period. 9 refs., 17 figs., 25 tabs.

Wronkiewicz, D.J.; Bates, J.K.; Gerding, T.J.; Veleckis, E.; Tani, B.S.

1991-07-01T23:59:59.000Z

78

Oxidative Dissolution of UO2 in a Simulated Groundwater Containing Synthetic Nanocrystalline Mackinawite  

SciTech Connect (OSTI)

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 conditions. Field and laboratory studies have implicated iron sulfide minerals as redox buffers or oxidant scavengers that may slow oxidation of reduced U(VI) solid phases by oxygen and Fe(III). Yet, the inhibition mechanism(s) and reaction rates of uraninite (UO2) oxidative dissolution by oxic species such as oxygen in FeS-bearing systems remain largely unresolved. To address this knowledge gap, abiotic batch experiments were conducted with synthetic UO2 in the presence and absence of synthetic mackinawite (FeS) under simulated groundwater conditions of pH = 7, PO2 = 0.02 atm, and PCO2 = 0.05 atm (equivalent to total dissolved carbonate of 0.01 M). The kinetic profiles of dissolved uranium indicate that FeS inhibited UO2 dissolution for 51 hr by effectively scavenging oxygen and keeping dissolved oxygen (DO) low. During this time period, oxidation of structural Fe(II) and S(-II) of FeS were found to control the DO levels, leading to the formation of iron oxyhydroxides and elemental sulfur, respectively, as verified by X-ray diffraction (XRD), Mössbauer and X-ray absorption spectroscopy (XAS). After FeS was depleted due to oxidation, DO levels increased and UO2 oxidative dissolution occurred at an initial rate of rm = 1.2 ± 0.4 ×10-8 mol•g-1•s-1, higher than rm = 5.4 ± 0.3 ×10-9 mol•g-1•s-1 in the control experiment where FeS was absent. Soluble U(VI) products were adsorbed by iron oxyhydroxides (i.e. nanogoethite and ferrihydrite) formed from FeS oxidation, which facilitated the detachment of U(VI) surface complexes and more rapid dissolution of UO2. XAS analysis confirmed the adsorption of U(VI) species, and also showed that U(VI) was not significantly incorporated into iron oxyhydroxide structure. This work reveals that both the oxygen scavenging by FeS and the adsorption of U(VI) to FeS oxidation products may be important in U reductive immobilization systems subject to redox cycling events.

Bi, Yuqiang; Hyun, Sung Pil; Kukkadapu, Ravi K.; Hayes, Kim F.

2013-02-01T23:59:59.000Z

79

Vibrational Spectroscopy of Mass Selected [UO2(ligand)n]2+ Complexes in the Gas Phase  

SciTech Connect (OSTI)

The gas-phase infrared spectra of discrete uranyl ([UO2]2+) complexes ligated with acetone and/or acetonitrile were used to evaluate systematic trends of ligation on the position of the O=U=O stretch, and to enable rigorous comparison with the results of computational studies. Ionic uranyl complexes isolated in a Fourier transform ion cyclotron resonance mass spectrometer were fragmented via infrared multiphoton dissociation using a free electron laser scanned over the mid-IR wavelengths. The asymmetric O=U=O stretching frequency was measured at 1017 cm-1 for [UO2(CH3COCH3)2]2+, and was systematically red shifted to 1000 and 988 cm-1 by the addition of a third and fourth acetone ligands, respectively, which was consistent with more donation of electron density to the uranium center in complexes with higher coordination number. The experimental measurements were in good agreement with values generated computationally using LDA, B3LYP, and ZORA-PW91 approaches. In contrast to the uranyl frequency shifts, the carbonyl frequencies of the acetone ligands were progressively blue shifted as the number of ligands increased from 2 to 4, and approached that of free acetone. This observation was consistent with the formation of weaker noncovalent bonds between uranium and the carbonyl oxygen as the extent of ligation increases. Similar trends were observed for [UO2(CH3CN)n]2+ complexes although the magnitude of the red shift in the uranyl frequency upon addition more acetonitrile ligands was smaller than for acetone, consistent with the more modest nucleophilic nature of acetonitrile. This conclusion was amplified by the uranyl stretching frequencies measured for mixed acetone/acetonitrile complexes, which showed that substitution of one acetone for one acetonitrile produced a modest red shift of 3 to 6 cm-1.

Gary S. Groenewold; Anita Gianotto; Michael Vanstipdonk; Kevin C. Cossel; David T. Moore,; Nick Polfer; Jos Oomens

2006-03-01T23:59:59.000Z

80

OXYGEN DIFFUSION IN HYPOSTOICHIOMETRIC URANIUM DIOXIDE  

E-Print Network [OSTI]

Research Division OXYGEN DIFFUSION IN HYPOSTOICHIOMETRIC11905 -DISCLAIMER - OXYGEN DIFFUSION IN HYPOSTOICHIOMETRICc o n e e n i g woroxygen self-diffusion coefficient

Kim, Kee Chul

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

NUCLEAR ISOTOPIC DILUTION OF HIGHLY ENRICHED URANIUM BY DRY BLENDING VIA THE RM-2 MILL TECHNOLOGY  

SciTech Connect (OSTI)

DOE has initiated numerous activities to focus on identifying material management strategies to disposition various excess fissile materials. In particular the INEEL has stored 1,700 Kg of offspec HEU at INTEC in CPP-651 vault facility. Currently, the proposed strategies for dispositioning are (a) aqueous dissolution and down blending to LEU via facilities at SRS followed by shipment of the liquid LEU to NFS for fabrication into LWR fuel for the TVA reactors and (b) dilution of the HEU to 0.9% for discard as a waste stream that would no longer have a criticality or proliferation risk without being processed through some type of enrichment system. Dispositioning this inventory as a waste stream via aqueous processing at SRS has been determined to be too costly. Thus, dry blending is the only proposed disposal process for the uranium oxide materials in the CPP-651 vault. Isotopic dilution of HEU to typically less than 20% by dry blending is the key to solving the dispositioning issue (i.e., proliferation) posed by HEU stored at INEEL. RM-2 mill is a technology developed and successfully tested for producing ultra-fine particles by dry grinding. Grinding action in RM-2 mill produces a two million-fold increase in the number of particles being blended in a centrifugal field. In a previous study, the concept of achieving complete and adequate blending and mixing (i.e., no methods were identified to easily separate and concentrate one titanium compound from the other) in remarkably short processing times was successfully tested with surrogate materials (titanium dioxide and titanium mono-oxide) with different particle sizes, hardness and densities. In the current project, the RM-2 milling technology was thoroughly tested with mixtures of natural uranium oxide (NU) and depleted uranium oxide (DU) stock to prove its performance. The effects of mill operating and design variables on the blending of NU/DU oxides were evaluated. First, NU and DU both made of the same oxide, UO{sub 3}, was used in the testing. Next, NU made up of UO{sub 3} and DU made up of UO{sub 2} was used in the test work. In every test, the blend achieved was characterized by spatial sampling of the ground product and analyzing for {sup 235}U concentration. The test work proved that these uranium oxide materials can be blended successfully. The spatial concentration was found to be uniform. Next, sintered thorium oxide pellets were used as surrogate for light water breeder reactor pellets (LWBR). To simulate LWBR pellet dispositioning, the thorium oxide pellets were first ground to a powder form and then the powder was blended with NU. In these tests also the concentration of {sup 235}U and {sup 232}Th in blended products fell within established limits proving the success of RM-2 milling technology. RM-2 milling technology is applicable to any dry radioactive waste, especially brittle solids that can be ground up and mixed with the non-radioactive stock.

Raj K. Rajamani; Sanjeeva Latchireddi; Vikas Devrani; Harappan Sethi; Roger Henry; Nate Chipman

2003-08-01T23:59:59.000Z

82

advanced doped uo2: Topics by E-print Network  

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

N. Creton1,a Physics Websites Summary: layer during the anionic oxidation of UO2 pellets induced very important mechanical stresses due to the crystallographic lattice...

83

PUREX/UO3 Facilities deactivation lessons learned history  

SciTech Connect (OSTI)

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

Gerber, M.S.

1996-09-19T23:59:59.000Z

84

Dissolution rates of uranium compounds in simulated lung fluid  

SciTech Connect (OSTI)

Maximum dissolution rates of uranium into simulated lung fluid from a variety of materials were measured at 37/sup 0/in the where f/sub i/ is in order to estimate clearance rates from the deep lung. A batch procedure was utilized in which samples containing as little as 10 ..mu..g of natural uranium could be tested. The materials included: products of uranium mining, milling and refining operations, coal fly ash, an environmental sample from a site exposed to multiple uranium sources, and purified samples of (NH/sub 4/)/sub 2/U/sub 2/O/sub 7/ U/sub 3/O/sub 8/, UO/sub 2/, and UF/sub 4/. Dissolution of uranium from several materials indicated the presence of more than one type of uranium compound; but in all cases, the fraction F of uranium remaining undissolved at any time t could be represented by the sum of up to three terms in the series: F = ..sigma../sub i/f/sub i/ exp (-0.693t/UPSILON/sub i/), where f/sub i/ is the initial fraction of component i with dissolution half-time epsilon/sub i/. Values of epsilon/sub i/ varied from 0.01 day to several thousand days depending on the physical and chemical form of the uranium. Dissolution occurred predominantly by formation of the (UO/sub 2/(CO/sub 3/)/sub 3/)/sup 4 -/ ion; and as a result, tetravalent uranium compounds dissolved slowly. Dissolution rates of size-separated yellow-cake aerosols were found to be more closely correlated with specific surface area than with aerodynamic diameter.

Kalkwarf, D.R.

1981-01-01T23:59:59.000Z

85

Uranium industry annual 1997  

SciTech Connect (OSTI)

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

NONE

1998-04-01T23:59:59.000Z

86

URANIUM IN ALKALINE ROCKS  

E-Print Network [OSTI]

Greenland," in Uranium Exploration Geology, Int. AtomicOklahoma," 1977 Nure Geology Uranium Symposium, Igneous HostMcNeil, M. , 1977. "Geology of Brazil's Uranium and Thorium

Murphy, M.

2011-01-01T23:59:59.000Z

87

Uranium-contaminated soils: Ultramicrotomy and electron beam analysis  

SciTech Connect (OSTI)

Uranium-contaminated soils from the U.S. Department of Energy (DOE) Fernald Site, Ohio, have been examined by a combination of scanning electron microscopy with backscattered electron imaging (SEM/BSE) and analytical electron microscopy (AEM). The inhomogeneous distribution of particulate uranium phases in the soil required the development of a method for using ultramicrotomy to prepare transmission electron microscopy (TEM) thin sections of the SEM mounts. A water-miscible resin was selected that allowed comparison between SEM and TEM images, permitting representative sampling of the soil. Uranium was found in iron oxides, silicates (soddyite), phosphates (autunites), and fluorite (UO{sub 2}). No uranium was detected in association with phyllosilicates in the soil.

Buck, E.C.; Dietz, N.L.; Bates, J.K.; Cunnane, J.C.

1994-02-01T23:59:59.000Z

88

Investigation of Uranium Polymorphs  

SciTech Connect (OSTI)

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

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

2011-08-01T23:59:59.000Z

89

Production of Depleted UO2Kernels for the Advanced Gas-Cooled Reactor Program for Use in TRISO Coating Development  

SciTech Connect (OSTI)

The main objective of the Depleted UO{sub 2} Kernels Production Task at Oak Ridge National Laboratory (ORNL) was to conduct two small-scale production campaigns to produce 2 kg of UO{sub 2} kernels with diameters of 500 {+-} 20 {micro}m and 3.5 kg of UO{sub 2} kernels with diameters of 350 {+-} 10 {micro}m for the U.S. Department of Energy Advanced Fuel Cycle Initiative Program. The final acceptance requirements for the UO{sub 2} kernels are provided in the first section of this report. The kernels were prepared for use by the ORNL Metals and Ceramics Division in a development study to perfect the triisotropic (TRISO) coating process. It was important that the kernels be strong and near theoretical density, with excellent sphericity, minimal surface roughness, and no cracking. This report gives a detailed description of the production efforts and results as well as an in-depth description of the internal gelation process and its chemistry. It describes the laboratory-scale gel-forming apparatus, optimum broth formulation and operating conditions, preparation of the acid-deficient uranyl nitrate stock solution, the system used to provide uniform broth droplet formation and control, and the process of calcining and sintering UO{sub 3} {center_dot} 2H{sub 2}O microspheres to form dense UO{sub 2} kernels. The report also describes improvements and best past practices for uranium kernel formation via the internal gelation process, which utilizes hexamethylenetetramine and urea. Improvements were made in broth formulation and broth droplet formation and control that made it possible in many of the runs in the campaign to produce the desired 350 {+-} 10-{micro}m-diameter kernels, and to obtain very high yields.

Collins, J.L.

2004-12-02T23:59:59.000Z

90

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

SciTech Connect (OSTI)

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.

Not Available

2011-06-22T23:59:59.000Z

91

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents [OSTI]

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.

Friedman, Horace A. (Oak Ridge, TN)

1985-01-01T23:59:59.000Z

92

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents [OSTI]

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.

Friedman, H.A.

1984-06-13T23:59:59.000Z

93

Vendor Control UoW 1730 (Rev. 10/07)  

E-Print Network [OSTI]

Vendor Control Use Only UoW 1730 (Rev. 10/07) ACCOUNTING DETAIL U.S. Taxpayer ID Number 1. Vendor. VENDOR'S CERTIFICATE: I hereby certify that the items and totals listed herein are proper charges

Borenstein, Elhanan

94

advanced uo2 fuel: Topics by E-print Network  

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

Last Page Topic Index 1 Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets CERN Preprints Summary: A transversal mapping of the Gd concentration was measured in...

95

Vibrational Spectroscopy of Mass-Selected [UO2(ligand)n]2+ Complexes in the Gas Phase: Comparison with Theory  

SciTech Connect (OSTI)

The gas-phase infrared spectra of discrete uranyl ([UO2]2+) complexes ligated with acetone and/or acetonitrile were used to evaluate systematic trends of ligation on the position of the OdUdO stretch and to enable rigorous comparison with the results of computational studies. Ionic uranyl complexes isolated in a Fourier transform ion cyclotron resonance mass spectrometer were fragmented via infrared multiphoton dissociation using a free electron laser scanned over the mid-IR wavelengths. The asymmetric OdUdO stretching frequency was measured at 1017 cm-1 for [UO2(CH3COCH3)2]2+ and was systematically red shifted to 1000 and 988 cm-1 by the addition of a third and fourth acetone ligand, respectively, which was consistent with increased donation of electron density to the uranium center in complexes with higher coordination number. The values generated computationally using LDA, B3LYP, and ZORA-PW91 were in good agreement with experimental measurements. In contrast to the uranyl frequency shifts, the carbonyl frequencies of the acetone ligands were progressively blue shifted as the number of ligands increased from two to four and approached that of free acetone. This observation was consistent with the formation of weaker noncovalent bonds between uranium and the carbonyl oxygen as the extent of ligation increases. Similar trends were observed for [UO2(CH3CN)n]2+ complexes, although the uranyl asymmetric stretching frequencies were greater than those measured for acetone complexes having equivalent coordination, which is consistent with the fact that acetonitrile is a weaker nucleophile than is acetone. This conclusion was confirmed by the uranyl stretching frequencies measured for mixed acetone/acetonitrile complexes, which showed that substitution of one acetone for one acetonitrile produced a modest red shift of 3-6 cm-1.

Gary S. Groenewold; Anita K. Gianotto

2006-03-01T23:59:59.000Z

96

Vibrational Spectroscopy of Mass-Selected [UO?(ligand)n]˛? Complexes in the Gas Phase: Comparison with Theory  

SciTech Connect (OSTI)

The gas-phase infrared spectra of discrete uranyl ([UO?]˛?) complexes ligated with acetone and/or acetonitrile were used to evaluate systematic trends of ligation on the position of the O=U=O stretch, and to enable rigorous comparison with the results of computational studies. Ionic uranyl complexes isolated in a Fourier transform ion cyclotron resonance mass spectrometer were fragmented via infrared multiphoton dissociation using a free electron laser scanned over the mid-IR wavelengths. The asymmetric O=U=O stretching frequency was measured at 1017 cm?ą for [UO?(CH?COCH?)?]˛? and was systematically red shifted to 1000 and 988 cm?ą by the addition of a third and fourth acetone ligand, respectively, which was consistent with increased donation of electron density to the uranium center in complexes with higher coordination number. The values generated computationally using LDA, B3LYP, and ZORA-PW91 were in good agreement with experimental measurements. In contrast to the uranyl frequency shifts, the carbonyl frequencies of the acetone ligands were progressively blue shifted as the number of ligands increased from 2 to 4, and approached that of free acetone. This observation was consistent with the formation of weaker noncovalent bonds between uranium and the carbonyl oxygen as the extent of ligation increases. Similar trends were observed for [UO?(CH?CN)n]˛? complexes, although the magnitude of the red shift in the uranyl frequency upon addition of more acetonitrile ligands was smaller than for acetone, consistent with the more modest nucleophilic nature of acetonitrile. This conclusion was confirmed by the uranyl stretching frequencies measured for mixed acetone/acetonitrile complexes, which showed that substitution of one acetone for one acetonitrile produced a modest red shift of 3 to 6 cm?ą.

Groenewold, G. S.; Gianotto, Anita K.; Cossel, Kevin C.; Van Stipdonk, Michael J.; Moore, David T.; Polfer, Nick; Oomens, Jos; De Jong, Wibe A.; Visscher, Lucas

2006-03-18T23:59:59.000Z

97

Electronic structure and ionicity of actinide oxides from first principles L. Petit,1,2,* A. Svane,1 Z. Szotek,2 W. M. Temmerman,2 and G. M. Stocks3  

E-Print Network [OSTI]

. A mixture of UO2 and PuO2, where Pu is blended with either natural or depleted uranium, constitutes. INTRODUCTION Actinide oxides play a dominant role in the nuclear fuel cycle.1 For many years, uranium dioxide

Svane, Axel Torstein

98

Water-Moderated and -Reflected Slabs of Uranium Oxyfluoride  

SciTech Connect (OSTI)

A series of ten experiments were conducted at the Oak Ridge National Laboratory Critical Experiment Facility in December 1955, and January 1956, in an attempt to determine critical conditions for a slab of aqueous uranium oxyfluoride (UO2F2). These experiments were recorded in an Oak Ridge Critical Experiments Logbook and results were published in a journal of the American Nuclear Society, Nuclear Science and Engineering, by J. K. Fox, L. W. Gilley, and J. H. Marable (Reference 1). The purpose of these experiments was to obtain the minimum critical thickness of an effectively infinite slab of UO2F2 solution by extrapolation of experimental data. To do this the slab thickness was varied and critical solution and water-reflector heights were measured using two different fuel solutions. Of the ten conducted experiments eight of the experiments reached critical conditions but the results of only six of the experiments were published in Reference 1. All ten experiments were evaluated from which five critical configurations were judged as acceptable criticality safety benchmarks. The total uncertainty in the acceptable benchmarks is between 0.25 and 0.33 % ?k/keff. UO2F2 fuel is also evaluated in HEU-SOL-THERM-043, HEU-SOL-THERM-011, and HEU-SOL-THERM-012, but these those evaluation reports are for large reflected and unreflected spheres. Aluminum cylinders of UO2F2 are evaluated in HEU-SOL-THERM-050.

Margaret A. Marshall; John D. Bess; J. Blair Briggs; Clinton Gross

2010-09-01T23:59:59.000Z

99

Uranium industry annual 1996  

SciTech Connect (OSTI)

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.

NONE

1997-04-01T23:59:59.000Z

100

Nitrogen dioxide detection  

DOE Patents [OSTI]

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.

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

1993-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Potential incorporation of transuranics into uranium phases  

SciTech Connect (OSTI)

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

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

1999-12-07T23:59:59.000Z

102

New insights into uranium (VI) sol-gel processing  

SciTech Connect (OSTI)

Nuclear Magnetic Resonance (NMR) investigations on the Oak Ridge National Laboratory process for sol-gel synthesis of microspherical nuclear fuel (UO{sub 2}), has been extremely useful in sorting out the chemical mechanism in the sol-gel steps. {sup 13}C, {sup 15}N, and {sup 1}H NMR studies on the HMTA gelation agent (Hexamethylene tetramine, C{sub 6}H{sub 12}N{sub 4}) has revealed near quantitative stability of this adamantane-like compound in the sol-gel process, contrary to its historical role as an ammonia source for gelation from the worldwide technical literature. {sup 17}O NMR of uranyl (UO{sub 2}{sup ++}) hydrolysis fragments produced in colloidal sols has revealed the selective formation of a uranyl trimer, ((UO{sub 2}){sub 3}({mu}{sub 3}-O)({mu}{sub 2}-OH){sub 3}){sup +}, induced by basic hydrolysis with the HMTA gelation agent. Spectroscopic results will be presented to illustrate that trimer condensation occurs during sol-gel processing leading to layered polyanionic hydrous uranium oxides in which HMTAH{sup +} is occluded as an intercalation'' cation. Subsequent sol-gel processing of microspheres by ammonia washing results in in-situ exchange and formation of a layered hydrous ammonium uranate with a proposed structural formula of (NH{sub 4}){sub 2} ((UO{sub 2}){sub 8} O{sub 4} (OH){sub 10}) {center dot} 8H{sub 2}O. This compound is the precursor to sintered UO{sub 2} ceramic fuel. 23 refs., 10 figs.

King, C.M.; Thompson, M.C.; Buchanan, B.R. (Westinghouse Savannah River Co., Aiken, SC (USA)); King, R.B. (Georgia Univ., Athens, GA (USA). Dept. of Chemistry); Garber, A.R. (South Carolina Univ., Columbia, SC (USA). Dept. of Chemistry)

1990-01-01T23:59:59.000Z

103

Uranium Industry Annual, 1992  

SciTech Connect (OSTI)

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

Not Available

1993-10-28T23:59:59.000Z

104

UO{sub 3} plant turnover - facility description document  

SciTech Connect (OSTI)

This document was developed to provide a facility description for those portions of the UO{sub 3} Facility being transferred to Bechtel Hanford Company, Inc. (BHI) following completion of facility deactivation. The facility and deactivated state condition description is intended only to serve as an overview of the plant as it is being transferred to BHI.

Clapp, D.A.

1995-01-01T23:59:59.000Z

105

FEASIBILITY STUDY OF DUPOLY TO RECYCLE DEPLETED URANIUM.  

SciTech Connect (OSTI)

DUPoly, depleted uranium (DU) powder microencapsulated in a low-density polyethylene binder, has been demonstrated as an innovative and efficient recycle product, a very durable high density material with significant commercial appeal. DUPoly was successfully prepared using uranium tetrafluoride (UF{sub 4}) ''green salt'' obtained from Fluor Daniel-Fernald, a U.S. Department of Energy reprocessing facility near Cincinnati, Ohio. Samples containing up to 90 wt% UF{sub 4} were produced using a single screw plastics extruder, with sample densities of up to 3.97 {+-} 0.08 g/cm{sup 3} measured. Compressive strength of as-prepared samples (50-90 wt% UF4 ) ranged from 1682 {+-} 116 psi (11.6 {+-} 0.8 MPa) to 3145 {+-} 57 psi (21.7 {+-} 0.4 MPa). Water immersion testing for a period of 90 days produced no visible degradation of the samples. Leach rates were low, ranging from 0.02 % (2.74 x 10{sup {minus}6} gm/gm/d) for 50 wt% UF{sub 4} samples to 0.72 % (7.98 x 10{sup {minus}5} gm/gm/d) for 90 wt% samples. Sample strength was not compromised by water immersion. DUPoly samples containing uranium trioxide (UO{sub 3}), a DU reprocessing byproduct material stockpiled at the Savannah River Site, were gamma irradiated to 1 x 10{sup 9} rad with no visible deterioration. Compressive strength increased significantly, however: up to 200% for samples with 90 wt% UO{sub 3}. Correspondingly, percent deformation (strain) at failure was decreased for all samples. Gamma attenuation data on UO{sub 3} DUPoly samples yielded mass attenuation coefficients greater than those for lead. Neutron removal coefficients were calculated and shown to correlate well with wt% of DU. Unlike gamma attenuation, both hydrogenous and nonhydrogenous materials interact to attenuate neutrons.

ADAMS,J.W.; LAGERAAEN,P.R.; KALB,P.D.; RUTENKROGER,S.P.

1998-02-01T23:59:59.000Z

106

Oxidative Dissolution of UO2 in a Simulated Groundwater Containing...  

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

Groundwater Containing Synthetic Nanocrystalline Mackinawite. Abstract: The long-term success of in situ reductive immobilization of uranium (U) depends on the stability of...

107

Density Functional Theory Calculations of Mass Transport in UO2  

SciTech Connect (OSTI)

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.

Andersson, Anders D. [Los Alamos National Laboratory; Dorado, Boris [CEA; Uberuaga, Blas P. [Los Alamos National Laboratory; Stanek, Christopher R. [Los Alamos National Laboratory

2012-06-26T23:59:59.000Z

108

Uranium industry annual 1994  

SciTech Connect (OSTI)

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

NONE

1995-07-05T23:59:59.000Z

109

What's Next for 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...

110

Lattice thermal conductivity of UO{sub 2} using ab-initio and classical molecular dynamics  

SciTech Connect (OSTI)

We applied the non-equilibrium ab-initio molecular dynamics and predict the lattice thermal conductivity of the pristine uranium dioxide for up to 2000?K. We also use the equilibrium classical molecular dynamics and heat-current autocorrelation decay theory to decompose the lattice thermal conductivity into acoustic and optical components. The predicted optical phonon transport is temperature independent and small, while the acoustic component follows the Slack relation and is in good agreement with the limited single-crystal experimental results. Considering the phonon grain-boundary and pore scatterings, the effective lattice thermal conductivity is reduced, and we show it is in general agreement with the sintered-powder experimental results. The charge and photon thermal conductivities are also addressed, and we find small roles for electron, surface polaron, and photon in the defect-free structures and for temperatures below 1500?K.

Kim, Hyoungchul [Department of Mechanical Engineering, University of Michigan, Ann Arbor, Michigan 48109 (United States); High-Temperature Energy Materials Research Center, Korea Institute of Science and Technology, Seoul 136–791 (Korea, Republic of); Kim, Moo Hwan [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang 790-784 (Korea, Republic of); Kaviany, Massoud, E-mail: kaviany@umich.edu [Department of Mechanical Engineering, University of Michigan, Ann Arbor, Michigan 48109 (United States); Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang 790-784 (Korea, Republic of)

2014-03-28T23:59:59.000Z

111

Standard test methods for arsenic in uranium hexafluoride  

E-Print Network [OSTI]

1.1 These test methods are applicable to the determination of total arsenic in uranium hexafluoride (UF6) by atomic absorption spectrometry. Two test methods are given: Test Method A—Arsine Generation-Atomic Absorption (Sections 5-10), and Test Method B—Graphite Furnace Atomic Absorption (Appendix X1). 1.2 The test methods are equivalent. The limit of detection for each test method is 0.1 ?g As/g U when using a sample containing 0.5 to 1.0 g U. Test Method B does not have the complete collection details for precision and bias data thus the method appears as an appendix. 1.3 Test Method A covers the measurement of arsenic in uranyl fluoride (UO2F2) solutions by converting arsenic to arsine and measuring the arsine vapor by flame atomic absorption spectrometry. 1.4 Test Method B utilizes a solvent extraction to remove the uranium from the UO2F2 solution prior to measurement of the arsenic by graphite furnace atomic absorption spectrometry. 1.5 Both insoluble and soluble arsenic are measured when UF6 is...

American Society for Testing and Materials. Philadelphia

2005-01-01T23:59:59.000Z

112

Final Uranium Leasing Program Programmatic Environmental Impact...  

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

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

113

Depleted Uranium Technical Brief  

E-Print Network [OSTI]

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

114

Uranium Oxide Aerosol Transport in Porous Graphite  

SciTech Connect (OSTI)

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.

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

2012-01-23T23:59:59.000Z

115

Uranium(VI) extraction by TBP in the presence of HDBP  

SciTech Connect (OSTI)

The influence of di-n-butyl phosphoric acid (HDBP) upon extraction of uranium(VI) by tri-n-butyl phosphate (TBP) from 0.5--3.0 M nitric acid solutions has been studied. It has been shown that the uranium(VI) distribution coefficient D{sub U} for extraction by 1.1 M TBP in tri-decane or xylene is increased when HDBP is present in the organic phase. For iso-molar solutions of (TBP + HDBP) with a total concentration of 0.36 M, and Uranium(VI) aqueous concentration up to 10--20 g/l, a maximum value of D{sub U} is observed when TBP/HDBP = 1; for higher U(VI) concentration the maximum gradually disappears, with D{sub U} growing monotonically with the HDBP content in the organic phase. Uranium(VI) absorption spectra for 1.1 M TBP in tri-decane or xylene, containing HDBP, provide evidence for the formation of compounds, of which composition is intermediate between uranyl nitrate--TBP disolvate and the U(VI)--HDBP complex. It is proposed that these intermediate compounds are UO{sub 2}(NO{sub 3}){sub 2}HDBP.TBP and UO{sub 2}(NO{sub 3}){sub 2}(HDBP){sub 2}.

Fedorov, Yu.S.; Zilberman, B.Ya.; Kulikov, S.M.; Blazheva, I.V.; Mishin, E.N. [V.G. Khlopin Radium Inst., Saint-Petersburg (Russian Federation); Wallwork, A.L.; Denniss, I.S.; May, I. [British Nuclear Fuels plc, Sellafield (United Kingdom); Hill, N.J. [British Nuclear Fuels plc, Risley (United Kingdom)

1999-03-01T23:59:59.000Z

116

Innovative Elution Processes for Recovering Uranium from Seawater  

SciTech Connect (OSTI)

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.

Wai, Chien; Tian, Guoxin; Janke, Christopher

2014-05-29T23:59:59.000Z

117

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

SciTech Connect (OSTI)

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.

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

2013-02-01T23:59:59.000Z

118

Method for converting uranium oxides to uranium metal  

DOE Patents [OSTI]

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.

Duerksen, Walter K. (Norris, TN)

1988-01-01T23:59:59.000Z

119

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

E-Print Network [OSTI]

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

Hart, Gus

120

Sulfur Dioxide Regulations (Ohio)  

Broader source: Energy.gov [DOE]

This chapter of the law establishes that the Ohio Environmental Protection Agency provides sulfur dioxide emission limits for every county, as well as regulations for the emission, monitoring and...

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Carbon dioxide removal process  

DOE Patents [OSTI]

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.

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

2003-11-18T23:59:59.000Z

122

Welding of uranium and uranium alloys  

SciTech Connect (OSTI)

The major reported work on joining uranium comes from the USA, Great Britain, France and the USSR. The driving force for producing this technology base stems from the uses of uranium as a nuclear fuel for energy production, compact structures requiring high density, projectiles, radiation shielding, and nuclear weapons. This review examines the state-of-the-art of this technology and presents current welding process and parameter information. The welding metallurgy of uranium and the influence of microstructure on mechanical properties is developed for a number of the more commonly used welding processes.

Mara, G.L.; Murphy, J.L.

1982-03-26T23:59:59.000Z

123

EPA Update: NESHAP Uranium Activities  

E-Print Network [OSTI]

for underground uranium mining operations (Subpart B) EPA regulatory requirements for operating uranium mill for Underground Uranium Mining Operations (Subpart B) #12;5 EPA Regulatory Requirements for Underground Uranium uranium mines include: · Applies to 10,000 tons/yr ore production, or 100,000 tons/mine lifetime · Ambient

124

Uranium hexafluoride public risk  

SciTech Connect (OSTI)

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

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

1994-08-01T23:59:59.000Z

125

Los Alamos probes mysteries of uranium dioxide's thermal conductivity  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas Conchas recovery challenge fund Las Conchas recoveryNuclearPhysicist honored

126

Uranium Mill Tailings Management  

SciTech Connect (OSTI)

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

Nelson, J.D.

1982-01-01T23:59:59.000Z

127

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

SciTech Connect (OSTI)

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.

Lloyd, R.

1980-10-01T23:59:59.000Z

128

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

SciTech Connect (OSTI)

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

Choi, W.K.; Song, P.S.; Oh, W.Z.; Jung, C.H. [Korea Atomic Energy Research Institute (Korea, Republic of); Min, B.Y. [Chungnam National University, 220 Gung-Dong, Yusung-Gu Taejon 305-764 (Korea, Republic of)

2007-07-01T23:59:59.000Z

129

Preparation of uranium compounds  

DOE Patents [OSTI]

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.

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

2013-02-19T23:59:59.000Z

130

Sampling, characterization, and remote sensing of aerosols formed in the atmospheric hydrolysis of uranium hexafluoride  

SciTech Connect (OSTI)

When gaseous uranium hexafluoride (UF/sub 6/) is released into the atmosphere, it rapidly reacts with ambient moisture to form an aerosol of uranyl fluoride (UO/sub 2/F/sub 2/) and hydrogen fluoride (HF). As part of our Safety Analysis program, we have performed several experimental releases of HF/sub 6/ in contained volumes in order to investigate techniques for sampling and characterizing the aerosol materials. The aggregate particle morphology and size distribution have been found to be dependent upon several conditions, including the temperature of the UF/sub 6/ at the time of its release, the relative humidity of the air into which it is released, and the elapsed time after the release. Aerosol composition and settling rate have been investigated using stationary samplers for the separate collection of UO/sub 2/F/sub 2/ and HF and via laser spectroscopic remote sensing (Mie scatter and infrared spectroscopy). 25 refs., 16 figs., 5 tabs.

Bostick, W.D.; McCulla, W.H.; Pickrell, P.W.

1984-05-01T23:59:59.000Z

131

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)

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.

Kim, S.H.; Chen, N.C.J.; Taleyarkhan, R.P.; Keith, K.D.; Schmidt, R.W. [Oak Ridge National Lab., TN (United States); Carter, J.C. [J.C. Carter Associates, Inc., Knoxville, TN (United States)

1996-12-30T23:59:59.000Z

132

Uranium industry annual 1993  

SciTech Connect (OSTI)

Uranium production in the United States has declined dramatically from a peak of 43.7 million pounds U{sub 3}O{sub 8} (16.8 thousand metric tons uranium (U)) in 1980 to 3.1 million pounds U{sub 3}O{sub 8} (1.2 thousand metric tons U) in 1993. This decline is attributed to the world uranium market experiencing oversupply and intense competition. Large inventories of uranium accumulated when optimistic forecasts for growth in nuclear power generation were not realized. The other factor which is affecting U.S. uranium production is that some other countries, notably Australia and Canada, possess higher quality uranium reserves that can be mined at lower costs than those of the United States. Realizing its competitive advantage, Canada was the world`s largest producer in 1993 with an output of 23.9 million pounds U{sub 3}O{sub 8} (9.2 thousand metric tons U). The U.S. uranium industry, responding to over a decade of declining market prices, has downsized and adopted less costly and more efficient production methods. The main result has been a suspension of production from conventional mines and mills. Since mid-1992, only nonconventional production facilities, chiefly in situ leach (ISL) mining and byproduct recovery, have operated in the United States. In contrast, nonconventional sources provided only 13 percent of the uranium produced in 1980. ISL mining has developed into the most cost efficient and environmentally acceptable method for producing uranium in the United States. The process, also known as solution mining, differs from conventional mining in that solutions are used to recover uranium from the ground without excavating the ore and generating associated solid waste. This article describes the current ISL Yang technology and its regulatory approval process, and provides an analysis of the factors favoring ISL mining over conventional methods in a declining uranium market.

Not Available

1994-09-01T23:59:59.000Z

133

Project Profile: Direct Supercritical Carbon Dioxide Receiver...  

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

Carbon Dioxide Receiver Development Project Profile: Direct Supercritical Carbon Dioxide Receiver Development National Renewable Energy Laboratory logo The National...

134

CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS  

E-Print Network [OSTI]

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

Hart, Gus

135

Carbon dioxide sensor  

DOE Patents [OSTI]

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).

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

2011-11-15T23:59:59.000Z

136

CARBON DIOXIDE EMISSION REDUCTION  

E-Print Network [OSTI]

.5 Primary Energy Use and Carbon Dioxide Emissions for Selected US Chemical Subsectors in 1994 ...............................................................................................................16 Table 2.7 1999 Energy Consumption and Specific Energy Consumption (SEC) in the U.S. Cement Efficiency Technologies and Measures in Cement Industry.................22 Table 2.9 Energy Consumption

Delaware, University of

137

WISE Uranium Project - Fact Sheet  

E-Print Network [OSTI]

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

Hazards From Depleted

138

India's Worsening Uranium Shortage  

SciTech Connect (OSTI)

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.

Curtis, Michael M.

2007-01-15T23:59:59.000Z

139

Depleted uranium management alternatives  

SciTech Connect (OSTI)

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.

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

1994-08-01T23:59:59.000Z

140

CARBON DIOXIDE FIXATION.  

SciTech Connect (OSTI)

Solar carbon dioxide fixation offers the possibility of a renewable source of chemicals and fuels in the future. Its realization rests on future advances in the efficiency of solar energy collection and development of suitable catalysts for CO{sub 2} conversion. Recent achievements in the efficiency of solar energy conversion and in catalysis suggest that this approach holds a great deal of promise for contributing to future needs for fuels and chemicals.

FUJITA,E.

2000-01-12T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Method for the recovery of uranium values from uranium tetrafluoride  

DOE Patents [OSTI]

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.

Kreuzmann, Alvin B. (Cincinnati, OH)

1983-01-01T23:59:59.000Z

142

Method for the recovery of uranium values from uranium tetrafluoride  

DOE Patents [OSTI]

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.

Kreuzmann, A.B.

1982-10-27T23:59:59.000Z

143

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

SciTech Connect (OSTI)

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

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

1995-11-30T23:59:59.000Z

144

Benchmarking of Graphite Reflected Critical Assemblies of UO2  

SciTech Connect (OSTI)

A series of experiments were carried out in 1963 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 253 tightly-packed fuel rods (1.27 cm triangular pitch) with graphite reflectors [1], the second part used 253 graphite-reflected fuel rods organized in a 1.506 cm triangular pitch [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods with a 1.506 cm triangular pitch. [3] Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. The first part of this experimental series has been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5] and is discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power systems. [6

Margaret A. Marshall; John D. Bess

2011-11-01T23:59:59.000Z

145

Uranium Enrichment Decontamination and Decommissioning Fund's...  

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

Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2008 and 2007 Financial Statement Audit Uranium Enrichment Decontamination and Decommissioning Fund's...

146

Process for electrolytically preparing uranium metal  

DOE Patents [OSTI]

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.

Haas, Paul A. (Knoxville, TN)

1989-01-01T23:59:59.000Z

147

Controlling uranium reactivity March 18, 2008  

E-Print Network [OSTI]

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

Meyer, Karsten

148

Influence of uranium hydride oxidation on uranium metal behaviour  

SciTech Connect (OSTI)

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)

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

2013-07-01T23:59:59.000Z

149

Process for sequestering carbon dioxide and sulfur dioxide  

DOE Patents [OSTI]

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.

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)

2009-10-20T23:59:59.000Z

150

Carbon Dioxide Reduction Through Urban Forestry  

E-Print Network [OSTI]

. Retrieval Terms: urban forestry, carbon dioxide, sequestration, avoided energy The Authors E. Gregory McCarbon Dioxide Reduction Through Urban Forestry: Guidelines for Professional and Volunteer Tree; Simpson, James R. 1999. Carbon dioxide reduction through urban forestry

Standiford, Richard B.

151

A DualDisk File System: ext4 Mihai Budiu  

E-Print Network [OSTI]

uranium oxide, UO2 [8,4] and hexavalent uranium based fluorides, UF6 [5], oxides, CaUO4 [9] and CdUO4 [10

Budiu, Mihai

152

Uranium-titanium-niobium alloy  

DOE Patents [OSTI]

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.

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

1990-01-01T23:59:59.000Z

153

Uranium deposits of Brazil  

SciTech Connect (OSTI)

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

NONE

1991-09-01T23:59:59.000Z

154

Uranium hexafluoride handling. Proceedings  

SciTech Connect (OSTI)

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.

Not Available

1991-12-31T23:59:59.000Z

155

DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL  

SciTech Connect (OSTI)

The primary goal of this investigation was to evaluate the effectiveness of the chop-leach process, with nitric acid solvent, to produce a nominally 300 g/L [U] and 1 M [H{sup +}] product solution. The results of this study show that this processing technique is appropriate for applications in which a low free acid and moderately high U content are desired. The 7.75 L of product solution, which was over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H{sup +}] in the range 0.8-1.2 M. A secondary goal was to test the effectiveness of this treatment for the removal of actinides from Zircaloy cladding to produce a low-level radioactive waste (LLW) cladding product. Analysis of the cladding shows that actinides are present in the cladding at a concentration of about 5000 {eta}Ci/g, which is about 50 times greater than the acceptable transuranium element limit in low level radioactive waste. It appears that the concentration of nitric acid used for this dissolution study (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate to completely digest the UO{sub 2} present in the spent fuel. The mass of insoluble material collected from the initial treatments with nitric acid, 340 g, was much higher than expected, and analysis of this insoluble residue showed that it contained at least 200 g U.

Kessinger, G.; Thompson, M.

2009-08-07T23:59:59.000Z

156

Magnetic resonance as a structural probe of a uranium (VI) sol-gel process  

SciTech Connect (OSTI)

NMR investigations on the ORNL process for sol-gel synthesis of microspherical nuclear fuel (UO{sub 2}), has been useful in sorting out the chemical mechanism in the sol-gel steps. {sup 13}C, {sup 15}N, and {sup 1}H NMR studies on the HMTA gelation agent (Hexamethylene tetramine, C{sub 6}H{sub l2}N{sub 4}) has revealed near quantitative stability of this adamantane-like compound in the sol-Gel process, contrary to its historical role as an ammonia source for gelation from the worldwide technical literature. {sub 17}0 NMR of uranyl (UO{sub 2}{sup ++}) hydrolysis fragments produced in colloidal sols has revealed the selective formation of a uranyl trimer, ((UO{sub 2}){sub 3}({mu}{sub 3}-O)({mu}{sub 2}-OH){sub 3}){sup +}, induced by basic hydrolysis with the HMTA gelation agent. Spectroscopic results show that trimer condensation occurs during sol-gel processing leading to layered polyanionic hydrous uranium oxides in which HMTAH{sup +} is occluded as an intercalation'' cation. Subsequent sol-gel processing of microspheres by ammonia washing results in in-situ ion exchange and formation of a layered hydrous ammonium uranate with a proposed structural formula of (NH{sub 4}){sub 2}((UO{sub 2}){sub 8}O{sub 4}(OH){sub 10}) {center dot} 8H{sub 2}0. This compound is the precursor to sintered U0{sub 2} ceramic fuel.

King, C.M.; Thompson, M.C.; Buchanan, B.R. (Westinghouse Savannah River Co., Aiken, SC (United States)); King, R.B. (Georgia Univ., Athens, GA (United States). Dept. of Chemistry); Garber, A.R. (South Carolina Univ., Columbia, SC (United States). Dept. of Chemistry)

1989-01-01T23:59:59.000Z

157

Magnetic resonance as a structural probe of a uranium (VI) sol-gel process  

SciTech Connect (OSTI)

NMR investigations on the ORNL process for sol-gel synthesis of microspherical nuclear fuel (UO{sub 2}), has been useful in sorting out the chemical mechanism in the sol-gel steps. {sup 13}C, {sup 15}N, and {sup 1}H NMR studies on the HMTA gelation agent (Hexamethylene tetramine, C{sub 6}H{sub l2}N{sub 4}) has revealed near quantitative stability of this adamantane-like compound in the sol-Gel process, contrary to its historical role as an ammonia source for gelation from the worldwide technical literature. {sub 17}0 NMR of uranyl (UO{sub 2}{sup ++}) hydrolysis fragments produced in colloidal sols has revealed the selective formation of a uranyl trimer, [(UO{sub 2}){sub 3}({mu}{sub 3}-O)({mu}{sub 2}-OH){sub 3}]{sup +}, induced by basic hydrolysis with the HMTA gelation agent. Spectroscopic results show that trimer condensation occurs during sol-gel processing leading to layered polyanionic hydrous uranium oxides in which HMTAH{sup +} is occluded as an ``intercalation`` cation. Subsequent sol-gel processing of microspheres by ammonia washing results in in-situ ion exchange and formation of a layered hydrous ammonium uranate with a proposed structural formula of (NH{sub 4}){sub 2}[(UO{sub 2}){sub 8}O{sub 4}(OH){sub 10}] {center_dot} 8H{sub 2}0. This compound is the precursor to sintered U0{sub 2} ceramic fuel.

King, C.M.; Thompson, M.C.; Buchanan, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); King, R.B. [Georgia Univ., Athens, GA (United States). Dept. of Chemistry; Garber, A.R. [South Carolina Univ., Columbia, SC (United States). Dept. of Chemistry

1989-12-31T23:59:59.000Z

158

Case Study: Transcritical Carbon Dioxide Supermarket Refrigeration...  

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

Case Study: Transcritical Carbon Dioxide Supermarket Refrigeration Systems Case Study: Transcritical Carbon Dioxide Supermarket Refrigeration Systems This case study documents one...

159

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...

160

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 The simulation provides an important approach to estimate the potential of storing carbon dioxide in depleted oil fields...

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Uranium Extraction From Laboratory Synthesized, Uranium-Doped Hydrous Ferric Oxides  

SciTech Connect (OSTI)

The extractability of uranium (U) from synthetic hydrous ferric oxides has been shown to decrease as a function of mineral ripening, consistent with the hypothesis that the ripening process decrease contaminant lability. To evaluate this process, three hydrous ferric oxide (HFO) suspensions were co-precipitated with uranyl (UO22+) and maintained at pH 7.0 ± 0.1. Uranyl was added to the HFO post-precipitation in a fourth suspension. Two suspensions also contained either co-precipitated silicate (Si-U-HFO) or phosphate (P-U-HFO). After precipitation of the HFOs, at time intervals of one week, one month, six months, one year, and 2 years, aliquots of the suspensions were contacted with five solutions for a range of time. The extracts were analyzed for U and iron (Fe). The results are consistent with the hypothesis that U and Fe extractability will decrease as the mineral phase ripens. All extracting solutions exhibited some degree of selectivity for U, as the proportional extraction of U exceeded that for congruent dissolution. Micro X-ray diffraction analysis indicates the transformation from an amorphous phase to a material containing substantial proportions of crystalline goethite and hematite, except the P-U-HFO which remained primarily amorphous. Further analysis of the co-precipitates by the Mössbauer technique and scanning electron microscopy (SEM) provides further evidence of mineralogic ripening

Smith, Steven C.; Douglas, Matthew; Moore, Dean A.; Kukkadapu, Ravi K.; Arey, Bruce W.

2009-03-01T23:59:59.000Z

162

Carbon dioxide and climate  

SciTech Connect (OSTI)

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.

Not Available

1990-10-01T23:59:59.000Z

163

Criticality Benchmark Analysis of Water-Reflected Uranium Oxyfluoride Slabs  

SciTech Connect (OSTI)

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'.

Margaret A. Marshall; John D. Bess

2009-11-01T23:59:59.000Z

164

Uranium immobilization and nuclear waste  

SciTech Connect (OSTI)

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.

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

1982-02-01T23:59:59.000Z

165

Corrosion-resistant uranium  

DOE Patents [OSTI]

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.

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

1981-10-21T23:59:59.000Z

166

Corrosion-resistant uranium  

DOE Patents [OSTI]

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.

Hovis, Jr., Victor M. (Kingston, TN); Pullen, William C. (Knoxville, TN); Kollie, Thomas G. (Oak Ridge, TN); Bell, Richard T. (Knoxville, TN)

1983-01-01T23:59:59.000Z

167

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uranium

168

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

SciTech Connect (OSTI)

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.

Elam, K.R.

2003-10-07T23:59:59.000Z

169

Ultraslow Wave Nuclear Burning of Uranium-Plutonium Fissile Medium on Epithermal Neutrons  

E-Print Network [OSTI]

For a fissile medium, originally consisting of uranium-238, the investigation of fulfillment of the wave burning criterion in a wide range of neutron energies is conducted for the first time, and a possibility of wave nuclear burning not only in the region of fast neutrons, but also for cold, epithermal and resonance ones is discovered for the first time. For the first time the results of the investigation of the Feoktistov criterion fulfillment for a fissile medium, originally consisting of uranium-238 dioxide with enrichments 4.38%, 2.00%, 1.00%, 0.71% and 0.50% with respect to uranium-235, in the region of neutron energies 0.015-10.0eV are presented. These results indicate a possibility of ultraslow wave neutron-nuclear burning mode realization in the uranium-plutonium media, originally (before the wave initiation by external neutron source) having enrichments with respect to uranium-235, corresponding to the subcritical state, in the regions of cold, thermal, epithermal and resonance neutrons. In order to validate the conclusions, based on the slow wave neutron-nuclear burning criterion fulfillment depending on the neutron energy, the numerical modeling of ultraslow wave neutron-nuclear burning of a natural uranium in the epithermal region of neutron energies (0.1-7.0eV) was conducted for the first time. The presented simulated results indicate the realization of the ultraslow wave neutron-nuclear burning of the natural uranium for the epithermal neutrons.

V. D. Rusov; V. A. Tarasov; M. V. Eingorn; S. A. Chernezhenko; A. A. Kakaev; V. M. Vashchenko; M. E. Beglaryan

2014-09-29T23:59:59.000Z

170

Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets  

E-Print Network [OSTI]

A transversal mapping of the Gd concentration was measured in UO2-Gd2O3 nuclear fuel pellets by electron paramagnetic resonance spectroscopy (EPR). The quantification was made from the comparison with a Gd2O3 reference sample. The nominal concentration in the pellets is UO2: 7.5 % Gd2O3. A concentration gradient was found, which indicates that the Gd2O3 amount diminishes towards the edges of the pellets. The concentration varies from (9.3 +/- 0.5)% in the center to (5.8 +/- 0.3)% in one of the edges. The method was found to be particularly suitable for the precise mapping of the distribution of Gd3+ ions in the UO2 matrix.

D. Tobia; E. L. Winkler; J. Milano; A. Butera; R. Kempf; L. Bianchi; F. Kaufmann

2014-02-28T23:59:59.000Z

171

Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets  

E-Print Network [OSTI]

A transversal mapping of the Gd concentration was measured in UO2-Gd2O3 nuclear fuel pellets by electron paramagnetic resonance spectroscopy (EPR). The quantification was made from the comparison with a Gd2O3 reference sample. The nominal concentration in the pellets is UO2: 7.5 % Gd2O3. A concentration gradient was found, which indicates that the Gd2O3 amount diminishes towards the edges of the pellets. The concentration varies from (9.3 +/- 0.5)% in the center to (5.8 +/- 0.3)% in one of the edges. The method was found to be particularly suitable for the precise mapping of the distribution of Gd3+ ions in the UO2 matrix.

Tobia, D; Milano, J; Butera, A; Kempf, R; Bianchi, L; Kaufmann, F

2014-01-01T23:59:59.000Z

172

High loading uranium fuel plate  

DOE Patents [OSTI]

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.

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

1990-01-01T23:59:59.000Z

173

Microscale Controls on the Fate of Contaminant Uranium in the Vadose Zone, Hanford Site, Washington  

SciTech Connect (OSTI)

An alkaline brine containing uranyl (UO22+) leaked to the thick unsaturated zone at the Hanford Site. X-ray and electron microprobe imaging showed that the uranium was associated with a minority of clasts, specifically granitic clasts occupying less than four percent of the sediment volume. XANES analysis at micron resolution showed the uranium to be hexavalent. The uranium was precipitated in microfractures as radiating clusters of uranyl silicates, and sorbed uranium was not observed on other surfaces. Compositional determinations of the 1-3 µm precipitates were difficult, but indicated a sodium potassium uranyl silicate, likely sodium boltwoodite. Observations suggested that uranyl was removed from pore waters by diffusion and precipitation in microfractures, where dissolved silica within the granite-equilibrated solution would cause supersaturation with respect to sodium boltwoodite. This hypothesis was tested using a diffusion reaction model operating at microscale. Conditions favoring precipitation were simulated to be transient, and driven by the compositional contrast between pore and fracture space. Pore-space conditions, including alkaline pH, were eventually imposed on the microfracture environment. However, conditions favoring precipitation were prolonged within the microfracture by reaction at the silicate mineral surface to buffer pH in a solubility limiting acidic state, and to replenish dissolved silica. During this time, uranyl was additionally removed to the fracture space by diffusion from pore space. Uranyl is effectively immobilized within the microfracture environment within the presently unsaturated vadose zone.

McKinley, James P.; Zachara, John M.; Liu, Chongxuan; Heald, Steve M.; Prenitzer, Brenda I.; Kempshall, Brian

2006-04-15T23:59:59.000Z

174

Removing oxygen from a solvent extractant in an uranium recovery process  

DOE Patents [OSTI]

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.

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

1984-01-01T23:59:59.000Z

175

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

DOE Patents [OSTI]

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.

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

1981-11-04T23:59:59.000Z

176

Thorium and uranium diphosphonates: Syntheses, structures, and spectroscopic properties  

SciTech Connect (OSTI)

Four new thorium and uranium diphosphonate compounds, [H{sub 3}O]{l_brace}Th{sub 2}[C{sub 6}H{sub 4}(PO{sub 3}){sub 2}]{sub 2}F{r_brace} (Thbbp-1), An{sub 2}{l_brace}(O{sub 3}PC{sub 6}H{sub 4}PO{sub 3}H){sub 2}[C{sub 6}H{sub 4}(PO{sub 3}H){sub 2}]{r_brace} [An=Th(IV), U(IV)] (Thbbp-2)/(U4bbp), and [(C{sub 2}H{sub 5})(CH{sub 3}){sub 3}N][(UO{sub 2}){sub 3}(O{sub 3}PC{sub 6}H{sub 4}PO{sub 3}H){sub 2}F(H{sub 2}O)] (U6bbp) have been synthesized hydrothermally using 1,4-benzenebisphosphonic acid as ligand. The crystal structures of these compounds were determined by single crystal X-ray diffraction. Thbbp-1 and Thbbp-2 contain seven-coordinate Th(IV) within ThO{sub 6}F and ThO{sub 7} units with capped trigonal prismatic and capped octahedral geometries, respectively. U4bbp is isotypic with Thbbp-2. The structure of U6bbp contains U(VI) is the common seven-coordinate pentagonal bipyramid. - Graphical abstract: Coordination polyhedra and luminescence properties in thorium and uranium compounds. Highlights: Black-Right-Pointing-Pointer Three-dimensional thorium and uranium complexes. Black-Right-Pointing-Pointer Conversion of U(VI) to U(IV) under hydrothermal condition. Black-Right-Pointing-Pointer Unusual seven-coordinate thorium complexes exhibiting capped octahedral and capped trigonal prismatic geometries.

Adelani, Pius O. [Department of Civil Engineering and Geological Sciences, and Department of Chemistry and Biochemistry, University of Notre Dame, Notre Dame, IN 46556 (United States); Albrecht-Schmitt, Thomas E., E-mail: talbrec1@nd.edu [156 Fitzpatrick Hall, University of Notre Dame, Notre Dame, IN 46556 (United States)

2012-08-15T23:59:59.000Z

177

Uranium from seawater  

SciTech Connect (OSTI)

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

Gregg, D.; Folkendt, M.

1982-09-21T23:59:59.000Z

178

CARBON DIOXIDE AND OUR OCEAN LEGACY  

E-Print Network [OSTI]

is a biologist at the California State Univer- sity San Marcos, with expertise in the effects of carbon dioxideCARBON DIOXIDE AND OUR OCEAN LEGACY G Carbon Dioxide: Our Role The United States is the single. Every day the average American adds about 118 pounds of carbon dioxide to the atmos- phere, due largely

179

Carbon Dioxide Sequestration Industrial-scale processes are available for separating carbon dioxide from the post-  

E-Print Network [OSTI]

Carbon Dioxide Sequestration Industrial-scale processes are available for separating carbon dioxide dioxide separation and sequestration because the lower cost of carbon dioxide separation from for injection of carbon dioxide into oil or gas-bearing formations. An advantage of sequestration involving

180

Safety Criticality Standards Using the French CRISTAL Code Package: Application to the AREVA NP UO{sub 2} Fuel Fabrication Plant  

SciTech Connect (OSTI)

Criticality safety evaluations implement requirements to proof of sufficient sub critical margins outside of the reactor environment for example in fuel fabrication plants. Basic criticality data (i.e., criticality standards) are used in the determination of sub critical margins for all processes involving plutonium or enriched uranium. There are several criticality international standards, e.g., ARH-600, which is one the US nuclear industry relies on. The French Nuclear Safety Authority (DGSNR and its advising body IRSN) has requested AREVA NP to review the criticality standards used for the evaluation of its Low Enriched Uranium fuel fabrication plants with CRISTAL V0, the recently updated French criticality evaluation package. Criticality safety is a concern for every phase of the fabrication process including UF{sub 6} cylinder storage, UF{sub 6}-UO{sub 2} conversion, powder storage, pelletizing, rod loading, assembly fabrication, and assembly transportation. Until 2003, the accepted criticality standards were based on the French CEA work performed in the late seventies with the APOLLO1 cell/assembly computer code. APOLLO1 is a spectral code, used for evaluating the basic characteristics of fuel assemblies for reactor physics applications, which has been enhanced to perform criticality safety calculations. Throughout the years, CRISTAL, starting with APOLLO1 and MORET 3 (a 3D Monte Carlo code), has been improved to account for the growth of its qualification database and for increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating a modern basic microscopic cross section set based on JEF2.2 and the comprehensive APOLLO2 and MORET 4 codes. APOLLO2 is well suited for criticality standards calculations as it includes a sophisticated self shielding approach, a P{sub ij} flux determination, and a 1D transport (S{sub n}) process. CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and the implementation of user-friendly graphical interfaces. Due to its comprehensive physical simulation and thanks to its broad qualification database with more than a thousand benchmark/calculation comparisons, CRISTAL V0 provides outstanding and reliable accuracy for criticality evaluations for configurations covering the entire fuel cycle (i.e. from enrichment, pellet/assembly fabrication, transportation, to fuel reprocessing). After a brief description of the calculation scheme and the physics algorithms used in this code package, results for the various fissile media encountered in a UO{sub 2} fuel fabrication plant will be detailed and discussed. (authors)

Doucet, M.; Durant Terrasson, L.; Mouton, J. [AREVA-NP (France)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Method of preparation of uranium nitride  

DOE Patents [OSTI]

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.

Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

2013-07-09T23:59:59.000Z

182

URANIUM MILLING ACTIVITIES AT SEQUOYAH FUELS CORPORATION  

E-Print Network [OSTI]

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

unknown authors

183

Carbon Dioxide: Threat or Opportunity?  

E-Print Network [OSTI]

catastrophic long term effects on world climate. An alternative to discharging carbon dioxide into the atmosphere is to find new uses. One possible use is in 'Biofactories'. Biofactories may be achieved by exploiting two new developing technologies: Solar...

McKinney, A. R.

1982-01-01T23:59:59.000Z

184

Reducing carbon dioxide to products  

DOE Patents [OSTI]

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.

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

2014-09-30T23:59:59.000Z

185

Recuperative supercritical carbon dioxide cycle  

DOE Patents [OSTI]

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.

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

2014-11-18T23:59:59.000Z

186

UNIVERSITY OF OREGON SOLAR MONITORING LABORATORY The University of Oregon (UO) Solar Moni-  

E-Print Network [OSTI]

i UNIVERSITY OF OREGON SOLAR MONITORING LABORATORY The University of Oregon (UO) Solar Moni- toring Laboratory has been measuring incident solar radiation since 1975. Current support for this work comes from the Regional Solar Radiation Monitoring Project (RSRMP), a utility consortium project including the Bon

Oregon, University of

187

f Fk66iCOP-] NBSIR 86-3422 uoL_ i 1  

E-Print Network [OSTI]

f Fk66iCOP-] NBSIR 86-3422 uoL_ i 1 The Performance of A Conventional Residential Sized Heat Pump RESIDENTIAL SIZED HEAT PUMP OPERATING WITH A NONAZEOTROPIC BINARY REFRIGERANT MIXTURE William Mulroy David unmodified residential heat pump designed for R22 when charged with a nonazeotropic refrigerant mixture (NARM

Oak Ridge National Laboratory

188

Radiation-Induced Decomposition of U(VI) Phase to Nanocrystals of UO2  

SciTech Connect (OSTI)

U{sup 6+}-phases are common alteration products, under oxidizing conditions, of uraninite and the UO{sub 2} in spent nuclear fuel. These U{sup 6+}-phases are subjected to a radiation field caused by the {alpha}-decay of U, or in the case of spent nuclear fuel, incorporated actinides, such as {sup 239}Pu and {sup 237}Np. In order to evaluate the effects of {alpha}-decay events on the stability of the U{sup 6+}-phases, we report, for the first time, the results of ion beam irradiations (1.0 MeV Kr{sup 2+}) of U{sup 6+}-phases. The heavy-particle irradiations are used to simulate the ballistic interactions of the recoil-nucleus of an {alpha}-decay event with the surrounding structure. The Kr{sup 2+}-irradiation decomposed the U{sup 6+}-phases to UO{sub 2} nanocrystals at doses as low as 0.006 displacements per atom (dpa). U{sup 6+}-phases accumulate substantial radiation doses ({approx}1.0 displacement per atom) within 100,000 years if the concentration of incorporated {sup 239}Pu is as high as 1 wt%. Similar nanocrystals of UO{sub 2} were observed in samples from the natural fission reactors at Oklo, Gabon. Multiple cycles of radiation-induced decomposition to UO{sub 2} followed by alteration to U{sup 6+}-phases provide a mechanism for the remobilization of incorporated radionuclides.

S. Utsunomiya; R.C. Ewing; L. Wang

2005-06-13T23:59:59.000Z

189

additives doped-uo2 pellets: Topics by E-print Network  

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

additives doped-uo2 pellets First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Journal of Nuclear...

190

UO Organizational Development and Training HUMAN RESOURCES, ORGANIZATIONAL DEVELOPMENT AND TRAINING  

E-Print Network [OSTI]

UO Organizational Development and Training HUMAN RESOURCES, ORGANIZATIONAL DEVELOPMENT AND TRAINING Organizational Development and Training HUMAN RESOURCES, ORGANIZATIONAL DEVELOPMENT AND TRAINING 5210 University with your field of choice. Explore what they have to offer members and consider learning from, and creating

Oregon, University of

191

Method for fabricating uranium foils and uranium alloy foils  

DOE Patents [OSTI]

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.

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

2006-09-05T23:59:59.000Z

192

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

SciTech Connect (OSTI)

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.

Du, Shiyu [Los Alamos National Laboratory; Andersson, Anders D. [Los Alamos National Laboratory; Germann, Timothy C. [Los Alamos National Laboratory; Stanek, Christopher R. [Los Alamos National Laboratory

2012-05-02T23:59:59.000Z

193

Unexpected, Stable Form of Uranium Detected | EMSL  

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

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

194

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

SciTech Connect (OSTI)

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.

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

2009-05-29T23:59:59.000Z

195

Determination of kinetic coefficients for the simultaneous reduction of sulfate and uranium by Desulfovibrio desulfuricans bacteria  

SciTech Connect (OSTI)

Uranium contamination of groundwaters and surface waters near abandoned mill tailings piles is a serious concern in many areas of the western United States. Uranium usually exists in either the U(IV) or the U(VI) oxidation state. U(VI) is soluble in water and, as a result, is very mobile in the environment. U(IV), however, is generally insoluble in water and, therefore, is not subject to aqueous transport. In recent years, researchers have discovered that certain anaerobic microorganisms, such as the sulfate-reducing bacteria Desulfovibrio desulfuricans, can mediate the reduction of U(VI) to U(IV). Although the ability of this microorganism to reduce U(VI) has been studied in some detail by previous researchers, the kinetics of the reactions have not been characterized. The purpose of this research was to perform kinetic studies on Desulfovibrio desulficans bacteria during simultaneous reduction of sulfate and uranium and to determine the phase in which uranium exists after it has been reduced and precipitated from solution. The studies were conducted in a laboratory-scale chemostat under substrate-limited growth conditions with pyruvate as the substrate. Kinetic coefficients for substrate utilization and cell growth were calculated using the Monod equation. The maximum rate of substrate utilization (k) was determined to be 4.70 days{sup {minus}1} while the half-velocity constant (K{sub s}) was 140 mg/l COD. The yield coefficient (Y) was determined to be 0.17 mg cells/mg COD while the endogenous decay coefficient (k{sub d}) was calculated as 0.072 days{sup {minus}1}. After reduction, U(IV) Precipitated from solution in the uraninite (UO{sub 2}) phase. Uranium removal efficiency as high as 90% was achieved in the chemostat.

Tucker, M.D.

1995-05-01T23:59:59.000Z

196

Supercritical Fluid Extraction and Separation of Uranium from Other Actinides  

SciTech Connect (OSTI)

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.

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

2014-06-01T23:59:59.000Z

197

Criticality experiments with low enriched UO/sub 2/ fuel rods in water containing dissolved gadolinium  

SciTech Connect (OSTI)

The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO/sub 2/ and PuO/sub 2/-UO/sub 2/ fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO/sub 2/ rods at two enrichments (2.35 wt % and 4.31 wt % /sup 235/U) and on mixed fuel-water assemblies of UO/sub 2/ and PuO/sub 2/-UO/sub 2/ rods containing 4.31 wt % /sup 235/U and 2 wt % PuO/sub 2/ in natural UO/sub 2/ respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in /sup 235/U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel.

Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

1984-02-01T23:59:59.000Z

198

2013 Domestic Uranium Production Report  

E-Print Network [OSTI]

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

199

Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems  

SciTech Connect (OSTI)

A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO[sub 2]F[sub 2] and H[sub 2]O) and hydrofluoric-acid-moderated uranium hexaflouride (UF[sub 6] and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % [sup 235]U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

Jordan, W.C.; Turner, J.C.

1992-12-01T23:59:59.000Z

200

Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems  

SciTech Connect (OSTI)

A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO{sub 2}F{sub 2} and H{sub 2}O) and hydrofluoric-acid-moderated uranium hexaflouride (UF{sub 6} and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % {sup 235}U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

Jordan, W.C.; Turner, J.C.

1992-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Synthesis and X-ray structural investigation of K{sub 8}[(UO{sub 2}){sub 2}(C{sub 2}O{sub 4}){sub 2}(SeO{sub 4}){sub 4}] . 2H{sub 2}O  

SciTech Connect (OSTI)

Single crystals of the compound K{sub 8}[(UO{sub 2}){sub 2}(C{sub 2}O{sub 4}){sub 2}(SeO{sub 4})4] . 2H{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 = 14.9290(4) A, b = 7.2800(2) A, c = 15.3165(4) A, {beta} = 109.188(1){sup o}, V = 1572.17(7) A{sup 3}, space group P2{sub 1}/n, Z = 2, and R = 0.0297. The uranium-containing structural units of crystals I are dimers of the composition [(UO {sub 2}){sub 2}(C{sub 2}O{sub 4}){sub 2}(SeO{sub 4}){sub 4}]{sup 8-}, which belong to the crystal-chemical group AB{sup 01}B{sup 2}M{sup 1} (A = UO{sub 2}{sup 2+}, B{sup 01} = C{sub 2}O{sub 4}{sup 2-}, B{sup 2} = SeO{sub 4}{sup 2-}, M{sup 1} = SeO{sub 4}{sup 2-}) of the uranyl complexes. The [(UO{sub 2}){sub 2}(C{sub 2}O{sub 4}){sub 2}(SeO{sub 4}){sub 4}]{sup 8-} dimers are joined into a three-dimensional framework through electrostatic interactions with the outer-sphere potassium cations.

Serezhkina, L. B., E-mail: lserezh@ssu.samara.ru [Samara State University (Russian Federation); Peresypkina, E. V.; Virovets, A. V. [Russian Academy of Sciences, Nikolaev Institute of Inorganic Chemistry, Siberian Branch (Russian Federation); Verevkin, A. G.; Pushkin, D. V. [Samara State University (Russian Federation)

2009-01-15T23:59:59.000Z

202

Domestic Uranium Production Report  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines AboutDecember 2005 (Thousand9, 2015Year109 AppendixCostsDistributedSep-1410. Uranium

203

Domestic Uranium Production Report  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines AboutDecember 2005 (Thousand9, 2015Year109 AppendixCostsDistributedSep-1410. Uranium9.

204

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.

205

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.

206

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.

207

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.5.

208

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.5.3.

209

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from

210

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uraniumb.

211

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uraniumb.7.

212

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma.

213

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma.9.

214

Fingerprinting Uranium | EMSL  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsing ZirconiaPolicyFeasibilityField Office FinalFinancingFingerprinting Uranium

215

Gas-phase energies of actinide oxides -- an assessment of neutral and cationic monoxides and dioxides from thorium to curium  

SciTech Connect (OSTI)

An assessment of the gas-phase energetics of neutral and singly and doubly charged cationic actinide monoxides and dioxides of thorium, protactinium, uranium, neptunium, plutonium, americium, and curium is presented. A consistent set of metal-oxygen bond dissociation enthalpies, ionization energies, and enthalpies of formation, including new or revised values, is proposed, mainly based on recent experimental data and on correlations with the electronic energetics of the atoms or cations and with condensed-phase thermochemistry.

Marcalo, Joaquim; Gibson, John K.

2009-08-10T23:59:59.000Z

216

UO{sub 2} corrosion in high surface-area-to-volume batch experiments.  

SciTech Connect (OSTI)

Unsaturated drip tests have been used to investigate the alteration of unirradiated UO{sub 2} and spent UO{sub 2} fuel in an unsaturated environment such as may be expected in the proposed repository at Yucca Mountain. In these tests, simulated groundwater is periodically injected onto a sample at 90 C in a steel vessel. The solids react with the dripping groundwater and water condensed on surfaces to form a suite of U(VI) alteration phases. Solution chemistry is determined from leachate at the bottom of each vessel after the leachate stops interacting with the solids. A more detailed knowledge of the compositional evolution of the leachate is desirable. By providing just enough water to maintain a thin film of water on a small quantity of fuel in batch experiments, we can more closely monitor the compositional changes to the water as it reacts to form alteration phases.

Bates, J. K.; Finch, R. J.; Hanchar, J. M.; Wolf, S. F.

1997-12-08T23:59:59.000Z

217

UO2 CORROSION IN HIGH SURFACE-AREA-TO-VOLUME BATCH EXPERIMENTS  

SciTech Connect (OSTI)

Unsaturated drip tests have been used to investigate the alteration of unirradiated UO{sub 2} and spent UO{sub 2} fuel in an unsaturated environment, such as may be expected in the proposed repository at Yucca Mountain. In these tests, simulated groundwater is periodically injected onto a sample at 90 C in a steel vessel. The solids react with the dripping groundwater and water condensed on surfaces to form a suite of U(VI) alteration phases. Solution chemistry is determined from leachate at the bottom of each vessel after the leachate stops interacting with the solids. A more detailed knowledge of the compositional evolution of the leachate is desirable. By providing just enough water to maintain a thin film of water on a small quantity of fuel in batch experiments, we can more closely monitor the compositional changes to the water as it reacts to form alteration phases.

Finch, Robert J.; Wolf, Stephen F.; Hanchar, John M.; Bates, John K.

1998-05-11T23:59:59.000Z

218

Bubble formation and Kr distribution in Kr-irradiated UO2  

SciTech Connect (OSTI)

In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO2 were conducted to understand the inert gas bubble behavior in oxide nuclear fuel. The bubble size and volume swelling are shown as a weak function of ion dose but strongly depend on the temperature. The Kr bubble formation at room temperature was observed for the first time. The depth profiles of implanted Kr determined by atom probe tomography are in good agreement with the calculated profiles by SRIM, but the measured concentration of Kr is about 1/3 of calculated one. This difference is mainly due to low solubility of Kr in UO2 matrix, which has been confirmed by both density-functional theory calculations and chemical equilibrium analysis.

L.F. He; B. Valderrama; A.-R. Hassan; J. Yu; M. Gupta; J. Pakarinen; H.B. Henderson; J. Gan; M.A. Kirk; A.T. Nelson; M.V. Manuel; A. El-Azab; T.R. Allen

2015-01-01T23:59:59.000Z

219

SEISMIC MONITORING OF CARBON DIOXIDE FLUID FLOW  

E-Print Network [OSTI]

SEISMIC MONITORING OF CARBON DIOXIDE FLUID FLOW J. E. Santos1, G. B. Savioli2, J. M. Carcione3, D´e, Argentina SEISMIC MONITORING OF CARBON DIOXIDE FLUID FLOW ­ p. #12;Introduction. I Storage of CO2). SEISMIC MONITORING OF CARBON DIOXIDE FLUID FLOW ­ p. #12;Introduction. II CO2 is separated from natural

Santos, Juan

220

APPENDIX J Partition Coefficients For Uranium  

E-Print Network [OSTI]

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

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Final Report - Gas Generation Testing of Uranium Metal in Simulated K Basin Sludge and in Grouted Sludge Waste Forms  

SciTech Connect (OSTI)

The Waste Isolation Pilot Plant (WIPP) is being considered for the disposal of K Basin sludge as RH-TRU. Because the hydrogen gas concentration in the 55-gallon RH-TRU sealed drums to be transported to WIPP is limited by flammability safety, the number of containers and shipments likely will be driven by the rate of hydrogen generated by the uranium metal-water reaction (U + 2 H{sub 2}O {yields} UO{sub 2} + 2 H{sub 2}) in combination with the hydrogen generated from water and organic radiolysis. Gas generation testing was conducted with uranium metal particles of known surface area, in simulated K West (KW) Basin canister sludge and immobilized in candidate grout solidification matrices. This study evaluated potential for Portland cement and magnesium phosphate grouts to inhibit the reaction of water with uranium metal in the sludge and thereby permit higher sludge loading to the disposed waste form. The best of the grouted waste forms decreased the uranium metal-water reaction by a factor of four.

Delegard, Calvin H.; Schmidt, Andrew J.; Sell, Rachel L.; Sinkov, Sergei I.; Bryan, Samuel A.; Gano, Sue; Thornton, Brenda M.

2004-08-19T23:59:59.000Z

222

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

DOE Patents [OSTI]

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].

Tomczuk, Z.; Miller, W.E.

1994-10-18T23:59:59.000Z

223

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

SciTech Connect (OSTI)

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}.

Ritter, R.L.; Barber, E.J. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)

1991-12-31T23:59:59.000Z

224

The End of Cheap Uranium  

E-Print Network [OSTI]

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

Michael Dittmar

2011-06-21T23:59:59.000Z

225

Safe Operating Procedure SAFETY PROTOCOL: URANIUM  

E-Print Network [OSTI]

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

Farritor, Shane

226

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

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

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

227

Leaching patterns and secondary phase formation during unsaturated leaching of UO{sub 2} at 90{degrees}C  

SciTech Connect (OSTI)

Experiments are being conducted that examine the reaction of UO{sub 2} with dripping oxygenated ground water at 90{degrees}C. The experiments are designed to identify secondary phases formed during UO{sub 2} alteration, evaluate parameters controlling U release, and act as scoping tests for studies with spent fuel. This study is the first of its kind that examines the alteration of UO{sub 2} under unsaturated conditions expected to exist at the proposed Yucca Mountain repository site. Results suggest the UO{sub 2} matrix will readily react within a few months after being exposed to simulated Yucca Mountain conditions. A pulse of rapid U release, combined with the formation of dehydrated schoepite on the UO{sub 2} surface, characterizes the reaction between one to two years. Rapid dissolution of intergrain boundaries and spallation of UO{sub 2} granules appears to be responsible for much of the U released. Differential release of the UO{sub 2} granules may be responsible for much of the variation observed between duplicate experiments. Less than 5 wt % of the released U remains in solution or in a suspended form, while the remaining settles out of solution as fine particles or is reprecipitated as secondary phases. Subsequent to the pulse period, U release rates decline and a more stable assemblage of uranyl silicate phases are formed by incorporating cations from the ground water leachant. Uranophane, boltwoodite, and sklodowskite appear as the final solubility limiting phases that form in these tests. This observed paragenetic sequence (from uraninite to schoepite-type phases to uranyl silicates) is identical to those observed in weathered zones of natural uraninite occurrences. The combined results indicate that the release of radionuclides from spent fuel may not be limited by U solubility constraints, but that spallation of particulate matter may be an important, if not the dominant release mechanism affecting release.

Wronkiewicz, D.J.; Bates, J.K.; Gerding, T.J.; Veleckis, E.; Tani, B.S.

1991-11-01T23:59:59.000Z

228

Feasibility of Partial ZrO[subscript 2] Coatings on Outer Surface of Annular UO[subscript 2] Pellets to Control Gap Conductance  

E-Print Network [OSTI]

The viability of depositing a thin porous coating of zirconia on the outer surface of an annular UO[subscript 2] pellet

Feinroth, H.

229

Laser induced phosphorescence uranium analysis  

DOE Patents [OSTI]

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.

Bushaw, Bruce A. (Kennewick, WA)

1986-01-01T23:59:59.000Z

230

Laser induced phosphorescence uranium analysis  

DOE Patents [OSTI]

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.

Bushaw, B.A.

1983-06-10T23:59:59.000Z

231

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

SciTech Connect (OSTI)

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

Chodak, P. III

1996-05-01T23:59:59.000Z

232

Excited States and Luminescent Properties of UO2F2 and Its Solvated Complexes in Aqueous Solution  

SciTech Connect (OSTI)

The electronic absorption and emission spectra of free UO2F2 and its water solvated complexes below 32,000 cm?1 are investigated at the levels of ab initio CASPT2 and CCSD(T) with inclusion of scalar relativistic and spin-orbit coupling effects. The influence of the water coordination on the electronic spectra of UO2F2 is explored by investigating the excited states of solvated complexes (H2O)nUO2F2 (n = 1?3). In these uranyl-complexes, water coordination is found to have appreciable influence on the 3? (? = 1g) character of the luminescent state and on the electronic spectral shape. The simulated luminescence spectral curves based on the calculated spectral parameters of (H2O)nUO2F2 from CCSD(T) approach agree well with experimental spectra in aqueous solution at both near liquid helium temperature and room temperature. The possible luminescence spectra of free UO2F2 in gas phase are predicted based on CASPT2 and CCSD(T) results, respectively, by considering three symmetric vibration modes. The effect of competition between spin-orbital coupling and ligand field repulsion on the luminescent state properties is discussed.

Su, Jing; Wang, Zheming; Pan, Duoqiang; Li, Jun

2014-08-20T23:59:59.000Z

233

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

E-Print Network [OSTI]

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

Helmreich, Grant

2012-02-14T23:59:59.000Z

234

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)

235

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

SciTech Connect (OSTI)

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.

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

2007-06-01T23:59:59.000Z

236

SEISMIC MONITORING OF CARBON DIOXIDE FLUID FLOW  

E-Print Network [OSTI]

SEISMIC MONITORING OF. CARBON DIOXIDE FLUID FLOW. J. E. Santos. 1. , G. B. Savioli. 2. , J. M. Carcione. 3. , D. Gei. 3. 1. CONICET, IGPUBA, Fac.

santos

237

Evidence of uranium biomineralization in sandstone-hosted roll-front uranium deposits, northwestern China  

E-Print Network [OSTI]

Evidence of uranium biomineralization in sandstone-hosted roll-front uranium deposits, northwestern Available online 25 January 2005 Abstract We show evidence that the primary uranium minerals, uraninite-front uranium deposits, Xinjiang, northwestern China were biogenically precipitated and psuedomorphically

Fayek, Mostafa

238

Grain boundary corrosion and alteration phase formation during the oxidative dissolution of UO{sub 2} pellets  

SciTech Connect (OSTI)

Alteration behavior of UO{sub 2} pellets following reaction under unsaturated drip-test conditions at 90 C for up to 10 years was examined by solid phase and leachate analyses. Sample reactions were characterized by preferential dissolution of grain boundaries between the original press-sintered UO{sub 2} granules comprising the samples, development of a polygonal network of open channels along the intergrain boundaries, and spallation of surface granules that had undergone severe grain boundary corrosion. The development of a dense mat of alteration phases after 2 years of reaction trapped loose granules, resulting in reduced rates of particulate U release. The paragenetic sequence of alteration phases that formed on the present samples was similar to that observed in surficial weathering zones of natural uraninite (UO{sub 2}) deposits, with alkali and alkaline earth uranyl silicates representing the long-term solubility-limiting phases for U in both systems.

Wronkiewicz, D.J.; Buck, E.C.; Bates, J.K.

1996-12-31T23:59:59.000Z

239

VAPOR + LIQUID EQUILIBRIUM OF WATER, CARBON DIOXIDE, AND THE BINARY SYSTEM WATER + CARBON DIOXIDE FROM  

E-Print Network [OSTI]

(for water: the SPC-, SPC/E-, and TIP4P-potential models; for carbon dioxide: the EPM2 potential model dioxide are calculated. For water, the SPC- and TIP4P-models give superior results for the vapor pressure when compared to the SPC/E-model. The vapor liquid equilibrium of the binary mixture carbon dioxide

240

Inherently safe in situ uranium recovery  

DOE Patents [OSTI]

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.

Krumhansl, James L; Brady, Patrick V

2014-04-29T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Uranium Acquisition | Y-12 National Security Complex  

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

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

242

The End of Cheap Uranium  

E-Print Network [OSTI]

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

Dittmar, Michael

2011-01-01T23:59:59.000Z

243

Determination of kinetic coefficients for the reduction and removal of uranium from water by the Desulfovibrio desulfuricans bacteria  

SciTech Connect (OSTI)

Uranium contamination of groundwater and surface water from abandoned uranium mill tailings piles is a serious concern in many areas of the western United States. U(VI) is soluble in water and, as a result, is relatively mobile in the environment. U(IV), however, is generally insoluble in water and, therefore, is not subject to aqueous transport. In recent years, researchers have discovered that certain microorganisms, such as the sulfate-reducing bacteria Desuffiovibrio desulfricans, can mediate the reduction of U(VI) to U(IV) by anaerobic respiration. Although the ability of this microorganism to reduce U(VI) has been studied in some detail by previous researchers, the kinetics of the reaction have not been characterized. The purpose of this research was to perform kinetic studies on Desuffiovibrio desulfricans during simultaneous reduction of sulfate and uranium and to determine the mineral phase of uranium after it has been reduced. The studies were conducted in a laboratory-scale chemostat under substrate-limited growth conditions with pyruvate as the substrate. The maximum rate of substrate utilization (k) was determined to be 4.70 days{sup -1} while the half-velocity constant (Ks) was 140 mg CODA. The yield coefficient (Y) was determined to be 0. 17 mg cells/mg COD while the endogenous decay coefficient (kd) was found to be 0.072 days{sup -1}. After reduction, U(IV) precipitated from solution in the uraninite (UO{sub 2}) phase as predicted by thermodynamics. Uranium removal efficiency as high as 90% was achieved in the chemostat.

Tucker, M.D.; Barton, L.L.; Thomson, B.M. [Sandia National Labs., Albuquerque, NM (United States)

1996-12-31T23:59:59.000Z

244

Evaluation of Heterogeneous Options: Effects of MgO versus UO2 Matrix Selection for Minor Actinide Targets in a Sodium Fast Reactor  

SciTech Connect (OSTI)

The primary focus of this work was to compare MgO with UO2 as target matrix material options for burning minor actinides in a transmutation target within a sodium fast reactor. This analysis compared the transmutation performance of target assemblies having UO2 matrix to those having specifically MgO inert matrix.

M. Pope; S. Bays; R. Ferrer

2008-03-01T23:59:59.000Z

245

High strength uranium-tungsten alloys  

DOE Patents [OSTI]

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.

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

1991-01-01T23:59:59.000Z

246

High strength uranium-tungsten alloy process  

DOE Patents [OSTI]

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.

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

1990-01-01T23:59:59.000Z

247

Clean Air Act Requirements: Uranium Mill Tailings  

E-Print Network [OSTI]

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

248

URANIUM MILL TAILINGS RADON FLUX CALCULATIONS  

E-Print Network [OSTI]

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

249

Remediation and Recovery of Uranium from Contaminated  

E-Print Network [OSTI]

Remediation and Recovery of Uranium from Contaminated Subsurface Environments with Electrodes K E L that Geobacter species can effectively remove uranium from contaminated groundwater by reducing soluble U was stably precipitated until reoxidized in the presence of oxygen. When an electrode was placed in uranium

Lovley, Derek

250

Uranium Watch REGULATORY CONFUSION: FEDERALAND STATE  

E-Print Network [OSTI]

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

251

D Riso-R-429 Automated Uranium  

E-Print Network [OSTI]

routinely used analytical techniques for uranium determina- tions in geological samples, fissionCM i D Riso-R-429 Automated Uranium Analysis by Delayed-Neutron Counting H. Kunzendorf, L. Løvborg AUTOMATED URANIUM ANALYSIS BY DELAYED-NEUTRON COUNTING H. Kunzendorf, L. Løvborg and E.M. Christiansen

252

Y-12 Uranium Exposure Study  

SciTech Connect (OSTI)

Following the recent restart of operations at the Y-12 Plant, the Radiological Control Organization (RCO) observed that the enriched uranium exposures appeared to involve insoluble rather than soluble uranium that presumably characterized most earlier Y-12 operations. These observations necessitated changes in the bioassay program, particularly the need for routine fecal sampling. In addition, it was not reasonable to interpret the bioassay data using metabolic parameter values established during earlier Y-12 operations. Thus, the recent urinary and fecal bioassay data were interpreted using the default guidance in Publication 54 of the International Commission on Radiological Protection (ICRP); that is, inhalation of Class Y uranium with an activity median aerodynamic diameter (AMAD) of 1 {micro}m. Faced with apparently new workplace conditions, these actions were appropriate and ensured a cautionary approach to worker protection. As additional bioassay data were accumulated, it became apparent that the data were not consistent with Publication 54. Therefore, this study was undertaken to examine the situation.

Eckerman, K.F.; Kerr, G.D.

1999-08-05T23:59:59.000Z

253

Quantifying differences in the impact of variable chemistry on equilibrium uranium(VI) adsorption properties of aquifer sediments  

SciTech Connect (OSTI)

Uranium adsorption-desorption on sediment samples collected from the Hanford 300-Area, Richland, WA varied extensively over a range of field-relevant chemical conditions, complicating assessment of possible differences in equilibrium adsorption properties. Adsorption equilibrium was achieved in 500-1000 hours although dissolved uranium concentrations increased over thousands of hours owing to changes in aqueous chemical composition driven by sediment-water reactions. A non-electrostatic surface complexation reaction, >SOH + UO22+ + 2CO32- = >SOUO2(CO3HCO3)2-, provided the best fit to experimental data for each sediment sample resulting in a range of conditional equilibrium constants (logKc) from 21.49 to 21.76. Potential differences in uranium adsorption properties could be assessed in plots based on the generalized mass-action expressions yielding linear trends displaced vertically by differences in logKc values. Using this approach, logKc values for seven sediment samples were not significantly different. However, a significant difference in adsorption properties between one sediment sample and the fines (<0.063 mm) of another could be demonstrated despite the fines requiring a different reaction stoichiometry. Estimates of logKc uncertainty were improved by capturing all data points within experimental errors. The mass-action expression plots demonstrate that applying models outside the range of conditions used in model calibration greatly increases potential errors.

Stoliker, Deborah L.; Kent, Douglas B.; Zachara, John M.

2011-09-16T23:59:59.000Z

254

Impact of uranyl-calcium-carbonato complexes on uranium(VI) adsorption to synthetic and natural sediments  

SciTech Connect (OSTI)

Adsorption on soil and sediment solids may decrease aqueous uranium concentrations and limit its propensity for migration in natural and contaminated settings. Uranium adsorption will be controlled in large part by its aqueous speciation, with a particular dependence on the presence of dissolved calcium and carbonate. Here we quantify the impact of uranyl speciation on adsorption to both goethite and sediments from the Hanford Clastic Dike and Oak Ridge Melton Branch Ridgetop formations. Hanford sediments were preconditioned with sodium acetate and acetic acid to remove carbonate grains, and Ca and carbonate were reintroduced at defined levels to provide a range of aqueous uranyl species. U(VI) adsorption is directly linked to UO{sub 2}{sup 2+} speciation, with the extent of retention decreasing with formation of ternary uranyl-calcium-carbonato species. Adsorption isotherms under the conditions studied are linear, and K{sub d} values decrease from 48 to 17 L kg{sup -1} for goethite, from 64 to 29 L kg{sup -1} for Hanford sediments, and from 95 to 51 L kg{sup -1} for Melton Branch sediments as the Ca concentration increases from 0 to 1 mM at pH 7. Our observations reveal that, in carbonate-bearing waters, neutral to slightly acidic pH values ({approx}5) and limited dissolved calcium are optimal for uranium adsorption.

Stewart, B.D. [Stanford University; Mayes, Melanie [ORNL; Fendorf, Scott [ORNL

2010-01-01T23:59:59.000Z

255

Preparation of thorium-uranium gel spheres  

SciTech Connect (OSTI)

Ceramic oxide spheres with diameters of 15 to 1500 ..mu..m are being evaluated for fabrication of power reactor fuel rods. (Th,U)O/sub 2/ spheres can be prepared by internal or external chemical gelation of nitrate solutions or oxide sols. Two established external gelation techniques were tested but proved to be unsatisfactory for the intended application. Established internal gelation techniques for UO/sub 2/ spheres were applied with minor modifications to make 75% ThO/sub 2/-25% UO/sub 2/ spheres that sinter to diameters of 200 to 1400 ..mu..m (99% T.D.).

Spence, R.D.; Haas, P.A.

1980-01-01T23:59:59.000Z

256

Acute and chronic toxicity of uranium compounds to Ceriodaphnia-Daphnia dubia  

SciTech Connect (OSTI)

A study to determine the acute and chronic toxicity of uranyl nitrate, hydrogen uranyl phosphate, and uranium dioxide to the organism Ceriodaphnia dubia was conducted. The toxicity tests were conducted by two independent environmental consulting laboratories. Part of the emphasis for this determination was based on concerns expressed by SCDHEC, which was concerned that a safety factor of 100 must be applied to the previous 1986 acute toxicity result of 0.22 mg/L for Daphnia pulex, This would have resulted in the LETF release limits being based on an instream concentration of 0.0022 mg/L uranium. The NPDES Permit renewal application to SCDHEC utilized the results of this study and recommended that the LETF release limit for uranium be based an instream concentration of 0.004 mg/L uranium. This is based on the fact that the uranium releases from the M-Area LETF will be in the hydrogen uranyl phosphate form, or a uranyl phosphate complex at the pH (6--10) of the Liquid Effluent Treatment Facility effluent stream, and at the pH of the receiving stream (5.5 to 7.0). Based on the chronic toxicity of hydrogen uranyl phosphate, a lower uranium concentration limit for the Liquid Effluent Treatment Facility outfall vs. the existing NPDES permit was recommended: The current NPDES permit ``Guideline`` for uranium at outfall M-004 is 0.500 mg/L average and 1.0 mg/L maximum, at a design flowrate of 60 gpm. It was recommended that the uranium concentration at the M-004 outfall be reduced to 0.28 mg/L average, and 0.56 mg/L, maximum, and to reduce the design flowrate to 30 gpm. The 0.28 mg/L concentration will provide an instream concentration of 0.004 mg/L uranium. The 0.28 mg/L concentration at M-004 is based on the combined flows from A-014, A-015, and A-011 outfalls (since 1985) of 1840 gpm (2.65 MGD) and was the flow rate which was utilized in the 1988 NPDES permit renewal application.

Pickett, J.B.; Specht, W.L.; Keyes, J.L.

1993-03-31T23:59:59.000Z

257

atmospheric sulphur dioxide: Topics by E-print Network  

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

carbon dioxide CERN Preprints Summary: The primary ingredient of Anthropogenic Global Warming hypothesis is the assumption that atmospheric carbon dioxide variations are the cause...

258

Carbon dioxide-assisted fabrication of highly uniform submicron...  

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

dioxide-assisted fabrication of highly uniform submicron-sized colloidal carbon spheres via hydrothermal carbonization Carbon dioxide-assisted fabrication of highly uniform...

259

Optimize carbon dioxide sequestration, enhance oil recovery  

E-Print Network [OSTI]

- 1 - Optimize carbon dioxide sequestration, enhance oil recovery January 8, 2014 Los Alamos simulation to optimize carbon dioxide (CO2) sequestration and enhance oil recovery (CO2-EOR) based on known production. Due to carbon capture and storage technology advances, prolonged high oil prices

260

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

SciTech Connect (OSTI)

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.

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

2012-07-25T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Process for alloying uranium and niobium  

DOE Patents [OSTI]

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.

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

1991-01-01T23:59:59.000Z

262

Uranium 2014 resources, production and demand  

E-Print Network [OSTI]

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

Organisation for Economic Cooperation and Development. Paris

2014-01-01T23:59:59.000Z

263

Uranium 2005 resources, production and demand  

E-Print Network [OSTI]

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

Organisation for Economic Cooperation and Development. Paris

2006-01-01T23:59:59.000Z

264

RELAP5/MOD3.2 analysis of a VVER-1000 reactor with UO[2] fuel and MOX fuel  

E-Print Network [OSTI]

.2 results showed a good agreement with calculations obtained with TECH-M computer program. The cladding temperatures of the MOX assembly have been compared with that of the hot UO? assembly. The peak cladding temperature of MOX assembly is about 55 K higher...

Fu, Chun

2012-06-07T23:59:59.000Z

265

Evaluation of sintering effects on SiC incorporated UO2 kernels under Ar and Ar-4%H2 environments  

SciTech Connect (OSTI)

Silicon carbide (SiC) is suggested as an oxygen getter in UO2 kernels used for TRISO particle fuels to lower oxygen potential and prevent kernel migration during irradiation. Scanning electron microscopy and X-ray diffractometry analyses performed on sintered kernels verified that internal gelation process can be used to incorporate SiC in urania fuel kernels. Sintering in either Ar or Ar-4%H2 at 1500 C lowered the SiC content in the UO2 kernels to some extent. Formation of UC was observed as the major chemical phase in the process, while other minor phases such as U3Si2C2, USi2, U3Si2, and UC2 were also identified. UC formation was presumed to be occurred by two reactions. The first was the SiC reaction with its protective SiO2 oxide layer on SiC grains to produce volatile SiO and free carbon that subsequently reacted with UO2 to form UC. The second process was direct UO2 reaction with SiC grains to form SiO, CO, and UC, especially in Ar-4%H2. A slightly higher density and UC content was observed in the sample sintered in Ar-4%H2, but the use of both atmospheres produced kernels with ~95% of theoretical density. It is suggested that incorporating CO in the sintering gas would prevent UC formation and preserve the initial SiC content.

Silva, Chinthaka M [ORNL] [ORNL; Lindemer, Terrence [Harbach Engineering and Solutions] [Harbach Engineering and Solutions; Hunt, Rodney Dale [ORNL] [ORNL; Collins, Jack Lee [ORNL] [ORNL; Terrani, Kurt A [ORNL] [ORNL; Snead, Lance Lewis [ORNL] [ORNL

2013-01-01T23:59:59.000Z

266

14UO TANK,OPENING REPORT NO.5. October 20th -November 26th (37 days total; 27 working days).  

E-Print Network [OSTI]

14UO TANK,OPENING REPORT NO.5. October 20th - November 26th (37 days total; 27 working days). Since the tank was last closed the accelerator ran for 97 days.until this opening which was scheduled to replace was done during the tank-open period. We believe that there would be value in gIvIng our assessments

Chen, Ying

267

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

SciTech Connect (OSTI)

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.

Tonks, Michael R. [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

2014-06-01T23:59:59.000Z

268

Reactions of aluminum with uranium fluorides and oxyfluorides  

SciTech Connect (OSTI)

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.

Leitnaker, J.M.; Nichols, R.W.; Lankford, B.S. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)

1991-12-31T23:59:59.000Z

269

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

270

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)

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.

Rasmussen B.

2010-01-01T23:59:59.000Z

271

D9 experiment: heat removal from stratified UO/sub 2/ debris  

SciTech Connect (OSTI)

The D9 experiment investigated the coolability of a shallow (77 mm), stratified urania bed in sodium. The bed was fission heated in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories to simulate the effects of radioactive decay heating. It was the first stratified debris bed experiment to use an extended UO/sub 2/ particle size distribution (0.038 to 4.0 mm). Dryout occurred at powers ranging from 0.10 to 0.58 W/g, which was close to the incipient boiling power and before channels penetrated the subcooled zone in the bed, even with subcoolings as low as 80/sup 0/C. Channel penetration was observed after dryout began, but the bed became only moderately more coolable. All these observations agree with current models.

Ottinger, C A; Mitchell, G W; Lipinski, R J; Kelly, J E

1985-04-01T23:59:59.000Z

272

Displacement of crude oil by carbon dioxide  

E-Print Network [OSTI]

by Carbon Dioxide (December 1980) Olusegun Omole, B. S. , University of Ibadan, Nigeria Chairman of Advisory Committee: Dr. J. S. Osoba It has long been recognized that carbon dioxide could be used as an oil recovery agent. Both laboratory and field...- tion. Crude oil from the Foster Field in West Texas, of 7 cp and 34 API, 0 was used as the oil in place. Oil displacements were conducted at pres- sures between 750 psig and 1800 ps1g, and at a temperature of 110 F. 0 Carbon dioxide was injected...

Omole, Olusegun

1980-01-01T23:59:59.000Z

273

Polyacrylamide-hydroxyapatite composite: Preparation, characterization and adsorptive features for uranium and thorium  

SciTech Connect (OSTI)

The composite of synthetically produced hydroxyapatite (HAP) and polyacrylamide was prepared (PAAm-HAP) and characterized by BET, FT-IR, TGA, XRD, SEM and PZC analysis. The adsorptive features of HAP and PAAm-HAP were compared for UO{sub 2}{sup 2+} and Th{sup 4+}. The entrapment of HAP into PAAm-HAP did not change the structure of HAP. Both structures had high affinity to the studied ions. The adsorption capacity of PAAm-HAP was than that of HAP. The adsorption dependence on pH and ionic intensity provided supportive evidences for the effect of complex formation on adsorption process. The adsorption kinetics was well compatible to pseudo second order model. The values of enthalpy and entropy changes were positive. Th{sup 4+} adsorption from the leachate obtained from a regional fluorite rock confirmed the selectivity of PAAm-HAP for this ion. In consequence, PAAm-HAP should be considered amongst favorite adsorbents for especially deposition of nuclear waste containing U and Th, and radionuclide at secular equilibrium with these elements. - Graphical abstract: SEM images of hydroxyapatite (HAP) and polyacrylamide-hydroxyapatite (PAAm-HAP), and the adsorption isotherms for Uranium and Thorium. Highlights: Black-Right-Pointing-Pointer Composite of PAAm-HAP was synthesized from hydroxyapatite and polyacrylamide. Black-Right-Pointing-Pointer The materials were characterized by BET, FT-IR, XRD, SEM, TGA and PZC analysis. Black-Right-Pointing-Pointer HAP and PAAm-HAP had high sorption capacity and very rapid uptake for UO{sub 2}{sup 2+} and Th{sup 4+}. Black-Right-Pointing-Pointer Super porous PAAm was obtained from PAAm-HAP after its removal of HAP content. Black-Right-Pointing-Pointer The composite is potential for deposition of U, Th and its associate radionuclides.

Baybas, Demet, E-mail: dbaybas@cumhuriyet.edu.tr [Cumhuriyet University, Faculty of Science, Department of Chemistry, Kayseri, Sivas 58140 (Turkey)] [Cumhuriyet University, Faculty of Science, Department of Chemistry, Kayseri, Sivas 58140 (Turkey); Ulusoy, Ulvi, E-mail: ulusoy@cumhuriyet.edu.tr [Cumhuriyet University, Faculty of Science, Department of Chemistry, Kayseri, Sivas 58140 (Turkey)] [Cumhuriyet University, Faculty of Science, Department of Chemistry, Kayseri, Sivas 58140 (Turkey)

2012-10-15T23:59:59.000Z

274

Identifying and Developing New, Carbon Dioxide Consuming Processes , Sudheer Indalaa  

E-Print Network [OSTI]

of propane, styrene from ethyl benzene and carbon dioxide, and methanol from hydrogenation of carbon dioxide408b Identifying and Developing New, Carbon Dioxide Consuming Processes Aimin Xua , Sudheer Indalaa@hal.lamar.edu, yawscl@hal.lamar.edu Key words; Carbon Dioxide Processes, Greenhouse Gases, Chemical Complex, Sustainable

Pike, Ralph W.

275

Reports on investigations of uranium anomalies. National Uranium Resource Evaluation  

SciTech Connect (OSTI)

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

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

1982-10-01T23:59:59.000Z

276

The New Generation of Uranium In Situ Recovery Facilities: Design Improvements Should Reduce Radiological Impacts Relative to First Generation Uranium Solution Mining Plants  

SciTech Connect (OSTI)

In the last few years, there has been a significant increase in the demand for Uranium as historical inventories have been consumed and new reactor orders are being placed. Numerous mineralized properties around the world are being evaluated for Uranium recovery and new mining / milling projects are being evaluated and developed. Ore bodies which are considered uneconomical to mine by conventional methods such as tunneling or open pits, can be candidates for non-conventional recovery techniques, involving considerably less capital expenditure. Technologies such as Uranium In Situ Leaching / In Situ Recovery (ISL / ISR - also referred to as 'solution mining'), have enabled commercial scale mining and milling of relatively small ore pockets of lower grade, and are expected to make a significant contribution to overall world wide uranium supplies over the next ten years. Commercial size solution mining production facilities have operated in the US since the mid 1970's. However, current designs are expected to result in less radiological wastes and emissions relative to these 'first' generation plants (which were designed, constructed and operated through the 1980's). These early designs typically used alkaline leach chemistries in situ including use of ammonium carbonate which resulted in groundwater restoration challenges, open to air recovery vessels and high temperature calcining systems for final product drying vs the 'zero emissions' vacuum dryers as typically used today. Improved containment, automation and instrumentation control and use of vacuum dryers in the design of current generation plants are expected to reduce production of secondary waste byproduct material, reduce Radon emissions and reduce potential for employee exposure to uranium concentrate aerosols at the back end of the milling process. In Situ Recovery in the U.S. typically involves the circulation of groundwater, fortified with oxidizing (gaseous oxygen e.g) and complexing agents (carbon dioxide, e.g) into an ore body, solubilizing the uranium in situ, and then pumping the solutions to the surface where they are fed to a processing plant ( mill). Processing involves ion exchange and may also include precipitation, drying or calcining and packaging operations depending on facility specifics. This paper presents an overview of the ISR process and the health physics monitoring programs developed at a number of commercial scale ISL / ISR Uranium recovery and production facilities as a result of the radiological character of these processes. Although many radiological aspects of the process are similar to that of conventional mills, conventional-type tailings as such are not generated. However, liquid and solid byproduct materials may be generated and impounded. The quantity and radiological character of these by products are related to facility specifics. Some special monitoring considerations are presented which are required due to the manner in which radon gas is evolved in the process and the unique aspects of controlling solution flow patterns underground. The radiological character of these processes are described using empirical data collected from many operating facilities. Additionally, the major aspects of the health physics and radiation protection programs that were developed at these first generation facilities are discussed and contrasted to circumstances of the current generation and state of the art of uranium ISR technologies and facilities. In summary: This paper has presented an overview of in situ Uranium recovery processes and associated major radiological aspects and monitoring considerations. Admittedly, the purpose was to present an overview of those special health physics considerations dictated by the in situ Uranium recovery technology, to point out similarities and differences to conventional mill programs and to contrast these alkaline leach facilities to modern day ISR designs. As evidenced by the large number of ISR projects currently under development in the U.S. and worldwide, non conventional Uranium recovery techniques

Brown, S.H. [CHP, SHB INC., Centennial, Colorado (United States)

2008-07-01T23:59:59.000Z

277

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

E-Print Network [OSTI]

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

Matthews, Isaac A

2010-01-01T23:59:59.000Z

278

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

E-Print Network [OSTI]

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

Hwang, Chiachi

2009-01-01T23:59:59.000Z

279

Regulating carbon dioxide capture and storage  

E-Print Network [OSTI]

This essay examines several legal, regulatory and organizational issues that need to be addressed to create an effective regulatory regime for carbon dioxide capture and storage ("CCS"). Legal, regulatory, and organizational ...

De Figueiredo, Mark A.

2007-01-01T23:59:59.000Z

280

Carbon Dioxide Emission Factors for Coal  

Reports and Publications (EIA)

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.

1994-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Uranium 2009 resources, production and demand  

E-Print Network [OSTI]

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

Organisation for Economic Cooperation and Development. Paris

2010-01-01T23:59:59.000Z

282

Uranium Metal Analysis via Selective Dissolution  

SciTech Connect (OSTI)

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.

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

2008-09-10T23:59:59.000Z

283

Depleted uranium disposal options evaluation  

SciTech Connect (OSTI)

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.

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

1994-05-01T23:59:59.000Z

284

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

E-Print Network [OSTI]

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

Boyer, Edmond

285

SHEEP MOUNTAIN URANIUM PROJECT CROOKS GAP, WYOMING  

E-Print Network [OSTI]

;PROJECT OVERVIEW ·Site Location·Site Location ·Fremont , Wyoming ·Existing Uranium Mine Permit 381C·Existing Uranium Mine Permit 381C ·Historical Operation ·Western Nuclear Crooks Gap Project ·Mined 1956 ­ 1988 and Open Pit Mining ·Current Mine Permit (381C) ·Updating POO, Reclamation Plan & Bond ·Uranium Recovery

286

Review of uranium bioassay techniques  

SciTech Connect (OSTI)

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

Bogard, J.S.

1996-04-01T23:59:59.000Z

287

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

E-Print Network [OSTI]

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

Meyer, Karsten

288

Statistical data of the uranium industry  

SciTech Connect (OSTI)

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

none,

1982-01-01T23:59:59.000Z

289

Adsorptive Stripping Voltammetric Measurements of Trace Uranium...  

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

Measurements of Trace Uranium at the Bismuth Film Electrode. Abstract: Bismuth-coated carbon-fiber electrodes have been successfully applied for adsorptive-stripping...

290

Biogeochemical Processes In Ethanol Stimulated Uranium Contaminated...  

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

A laboratory incubation experiment was conducted with uranium contaminated subsurface sediment to assess the geochemical and microbial community response to ethanol amendment. A...

291

Colorimetric detection of uranium in water  

DOE Patents [OSTI]

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.

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

2012-03-13T23:59:59.000Z

292

Uranium Weapons Components Successfully Dismantled | National...  

National Nuclear Security Administration (NNSA)

Successfully Dismantled March 20, 2007 Uranium Weapons Components Successfully Dismantled Oak Ridge, TN Continuing its efforts to reduce the size of the U.S. nuclear weapons...

293

Review The Toxicity of Depleted Uranium  

E-Print Network [OSTI]

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

Wayne Briner

294

High strength and density tungsten-uranium alloys  

DOE Patents [OSTI]

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.

Sheinberg, Haskell (Los Alamos, NM)

1993-01-01T23:59:59.000Z

295

Distribution of uranium-bearing phases in soils from Fernald  

SciTech Connect (OSTI)

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

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

1993-12-31T23:59:59.000Z

296

THE KINETICS OF LASER PULSE VAPORIZATION OF URANIUM DIOXIDE BY MASS SPECTROMETRY  

E-Print Network [OSTI]

B. Nicholson, "VENUS-II: An LMFBR Disassembly Program," ANL-metal fast breeder reactor (LMFBR) safety analysis. Most

Tsai, Chuen-horng

2012-01-01T23:59:59.000Z

297

Secondary Uranium-Phase Paragenesis and Incorporation of Radionuclides into Secondary Phase  

SciTech Connect (OSTI)

The purpose of this analysis/model report (AMR) is to assess the potential for uranium (U) (VI) compounds, formed during the oxidative corrosion of spent uranium-oxide (UO{sub 2}) fuels, to sequester certain radionuclides and, thereby, limit their release. The ''unsaturated drip tests'' being conducted at Argonne National Laboratory (ANL) provide the basis of this AMR (Table 1). The ANL drip tests on spent fuel are the only experiments on fuel corrosion from which solids have been analyzed for trace levels of radionuclides. Brief summaries are provided of the results from other selected corrosion and dissolution experiments on spent UO{sub 2} fuels, specifically those conducted under nominally oxidizing conditions. Discussions of the current understanding of thermodynamic and kinetic properties of U(VI) compounds is provided in order to outline the scientific basis for modeling precipitation and dissolution of potential radionuclide-bearing phases under repository-relevant conditions. Attachment I provides additional information on corrosion mechanisms and behaviors of radionuclides in the tests at ANL. Attachment II reviews occurrence, formation, and alteration (collectively known as paragenesis) of naturally occurring U(VI) minerals because natural mineral occurrences can be used to assess the possible long-term behaviors of U(VI) compounds formed in short-term laboratory experiments and to extrapolate experimental results to repository-relevant time scales. This AMR develops a model for calculating dissolved concentrations of radionuclides that are incorporated into U(VI) compounds, which is an alternative to models currently used in TSPA to calculate dissolved concentration limits for certain radionuclides. In particular, the model developed in this AMR applies to Np (neptunium) concentrations being controlled by solid uranyl oxyhydroxides that are known to contain trace levels of Np. The results of this AMR and the conceptual model developed from it and presented in Section 6.7.2.3 are primarily intended to support sensitivity evaluations in performance assessment. This AMR was developed in accordance with the ''Technical Work Plan for Waste Form Degradation Process Model Report for SR'' (CRWMS M&O 2000a). The scope of this AMR is outlined in the section ''Mixed Phase Dissolved Radionuclide Concentration Limits'' of the technical work plan.

R. Finch

2001-06-05T23:59:59.000Z

298

President Truman Increases Production of Uranium and Plutonium...  

National Nuclear Security Administration (NNSA)

Increases Production of Uranium and Plutonium October 09, 1950 President Truman Increases Production of Uranium and Plutonium Washington, DC President Truman approves a 1.4...

299

Atomistic Simulations of Uranium Incorporation into Iron (Hydr...  

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

of Uranium Incorporation into Iron (Hydr)Oxides. Atomistic Simulations of Uranium Incorporation into Iron (Hydr)Oxides. Abstract: Atomistic simulations were carried out to...

300

Toxic Substances Control Act Uranium Enrichment Federal Facility...  

Office of Environmental Management (EM)

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

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Geochemical Controls on Contaminant Uranium in Vadose Hanford...  

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

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

302

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

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

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

303

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

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

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

304

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

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

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

305

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...

306

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

Energy Savers [EERE]

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

307

DOE Extends Public Comment Period for the Draft Uranium Leasing...  

Office of Environmental Management (EM)

Extends Public Comment Period for the Draft Uranium Leasing Program Programmatic Environmental Impact Statement DOE Extends Public Comment Period for the Draft Uranium Leasing...

308

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...

309

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

Office of Environmental Management (EM)

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

310

A brief history of the PUREX and UO{sub 3} facilities  

SciTech Connect (OSTI)

The Plutonium-Uranium Extraction (PUREX) Plant, conceived during the early Cold War years, was a vehicle to increase significantly US nuclear weapons production capacity. The original PUREX Plant was a concrete rectangle 1,005 feet long and 61.5 feet wide. The shielding capacity of the concrete was designed so that personnel in non-regulated service areas would not receive radiation in excess of 0.1 millirem per hour. This report discusses the design of the PUREX Plant, the production chronology, projects and equipment changes, equipment decontamination and reuse, waste management, and contamination events that have occurred during the operation of the plant. Additionally, the development and history of the Uranium Trioxide Plant are also covered.

Gerber, M.S.

1993-11-01T23:59:59.000Z

311

Dissolution of Irradiated Commercial UO2 Fuels in Ammonium Carbonate and Hydrogen Peroxide  

SciTech Connect (OSTI)

We propose and test a disposition path for irradiated nuclear fuel using ammonium carbonate and hydrogen peroxide media. We demonstrate on a 13 g scale that >98% of the irradiated fuel dissolves. Subsequent expulsion of carbonate from the dissolver solution precipitates >95% of the plutonium, americium, curium, and substantial amounts of fission products, effectively partitioning the fuel at the dissolution step. Uranium can be easily recovered from solution by any of several means, such as ion exchange, solvent extraction, or direct precipitation. Ammonium carbonate can be evaporated from solution and recovered for re-use, leaving an extremely compact volume of fission products, transactinides, and uranium. Stack emissions are predicted to be less toxic, less radioactive, chemically simpler, and simpler to treat than those from the conventional PUREX process.

Soderquist, Chuck Z.; Johnsen, Amanda M.; McNamara, Bruce K.; Hanson, Brady D.; Chenault, Jeffrey W.; Carson, Katharine J.; Peper, Shane M.

2011-01-18T23:59:59.000Z

312

Increasing carbon dioxideIncreasing carbon dioxide & its effect on forest& its effect on forest  

E-Print Network [OSTI]

ecosystem's natural capacity toA forest ecosystem's natural capacity to capture energy, capture energy's natural capacity toA forest ecosystem's natural capacity to capture energy, capture energy, sustain life10/13/2010 1 Increasing carbon dioxideIncreasing carbon dioxide & its effect on forest& its effect

Gray, Matthew

313

RADIATION-INDUCED DECOMPOSITION OF U(VI) ALTERATION PHASES OF UO2  

SciTech Connect (OSTI)

U{sup 6+}-phases are common alteration products of spent nuclear fuel under oxidizing conditions, and they may potentially incorporate actinides, such as long-lived {sup 239}Pu and {sup 237}Np, delaying their transport to the biosphere. In order to evaluate the ballistic effects of {alpha}-decay events on the stability of the U{sup 6+}-phases, we report, for the first time, the results of ion beam irradiations (1.0 MeV Kr{sup 2+}) for six different structures of U{sup 6+}-phases: uranophane, kasolite, boltwoodite, saleeite, carnotite, and liebigite. The target uranyl-minerals were characterized by powder X-ray diffraction and identification confirmed by SAED (selected area electron diffraction) in TEM (transmission electron microscopy). The TEM observation revealed no initial contamination of uraninite in these U{sup 6+} phases. All of the samples were irradiated with in situ TEM observation using 1.0 MeV Kr{sup 2+} in the IVEM (intermediate-voltage electron microscope) at the IVEM-Tandem Facility of Argonne National Laboratory. The ion flux was 6.3 x 10{sup 11} ions/cm{sup 2}/sec. The specimen temperatures during irradiation were 298 and 673 K, respectively. The Kr{sup 2+}-irradiation decomposed the U{sup 6+}-phases to nanocrystals of UO{sub 2} at doses as low as 0.006 dpa. The cumulative doses for the pure U{sup 6+}-phases, e.g., uranophane, at 0.1 and 1 million years (m.y.) are calculated to be 0.009 and 0.09 dpa using SRIM2003. However, with the incorporation of 1 wt.% {sup 239}Pu, the calculated doses reach 0.27 and {approx}1.00 dpa in ten thousand and one hundred thousand years, respectively. Under oxidizing conditions, multiple cycles of radiation-induced decomposition to UO{sub 2} followed by alteration to U{sup 6+}-phases should be further investigated to determine the fate of trace elements that may have been incorporated in the U{sup 6+}-phases.

S. Utsunomiya; R.C. Ewing

2005-09-08T23:59:59.000Z

314

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

SciTech Connect (OSTI)

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.

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

2014-10-01T23:59:59.000Z

315

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

SciTech Connect (OSTI)

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

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

2012-01-01T23:59:59.000Z

316

Bioremediation of uranium contaminated soils and wastes  

SciTech Connect (OSTI)

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

Francis, A.J.

1998-12-31T23:59:59.000Z

317

Uranium Management - Preservation of a National Asset  

SciTech Connect (OSTI)

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

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

2002-02-27T23:59:59.000Z

318

Use of a permeable biological reaction barrier for groundwater remediation at a uranium mill tailings remedial action (UMTRA) site  

SciTech Connect (OSTI)

Previous work at the University of New Mexico and elsewhere has shown that sulfate reducing bacteria are capable of reducing uranium from the soluble +6 oxidation state to the insoluble +4 oxidation state. This chemistry forms the basis of a proposed groundwater remediation strategy in which microbial reduction would be used to immobilize soluble uranium. One such system would consist of a subsurface permeable barrier which would stimulate microbial growth resulting in the reduction of sulfate and nitrate and immobilization of metals while permitting the unhindered flow of ground water through it. This research investigated some of the engineering considerations associated with a microbial reducing barrier such as identifying an appropriate biological substrate, estimating the rate of substrate utilization, and identifying the final fate of the contaminants concentrated in the barrier matrix. The performance of batch reactors and column systems that treated simulated plume water was evaluated using cellulose, wheat straw, alfalfa hay, sawdust, and soluble starch as substrates. The concentrations of sulfate, nitrate, and U(VI) were monitored over time. Precipitates from each system were collected and the precipitated U(IV) was determined to be crystalline UO{sub 2}(s) by X-ray Diffraction. The results of this study support the proposed use of cellulosic substrates as candidate barrier materials.

Thombre, M.S.; Thomson, B.M.; Barton, L.L. [Univ. of New Mexico, Albuquerque, NM (United States)

1997-12-31T23:59:59.000Z

319

IPNS enriched uranium booster target  

SciTech Connect (OSTI)

Since startup in 1981, IPNS has operated on a fully depleted /sup 238/U target. With the booster as in the present system, high energy protons accelerated to 450 MeV by the Rapid Cycling Synchrotron are directed at the target and by mechanisms of spallation and fission of the uranium, produce fast neutrons. The neutrons from the target pass into adjacent moderator where they slow down to energies useful for spectroscopy. The target cooling systems and monitoring systems have operated very reliably and safely during this period. To provide higher neutron intensity, we have developed plans for an enriched uranium (booster) target. HETC-VIM calculations indicate that the target will produce approx.90 kW of heat, with a nominal x5 gain (k/sub eff/ = 0.80). The neutron beam intensity gain will be a factor of approx.3. Thermal-hydraulic and heat transport calculations indicate that approx.1/2 in. thick /sup 235/U discs are subject to about the same temperatures as the present /sup 238/U 1 in. thick discs. The coolant will be light demineralized water (H/sub 2/O) and the coolant flow rate must be doubled. The broadening of the fast neutron pulse width should not seriously affect the neutron scattering experiments. Delayed neutrons will appear at a level about 3% of the total (currently approx.0.5%). This may affect backgrounds in some experiments, so that we are assessing measures to control and correct for this (e.g., beam tube choppers). Safety analyses and neutronic calculations are nearing completion. Construction of the /sup 235/U discs at the ORNL Y-12 facility is scheduled to begin late 1985. The completion of the booster target and operation are scheduled for late 1986. No enriched uranium target assembly operating at the projected power level now exists in the world. This effort thus represents an important technological experiment as well as being a ''flux enhancer''.

Schulke, A.W. Jr.

1985-01-01T23:59:59.000Z

320

Carbon dioxide capture process with regenerable sorbents  

DOE Patents [OSTI]

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.

Pennline, Henry W. (Bethel Park, PA); Hoffman, James S. (Library, PA)

2002-05-14T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Molten-Salt Depleted-Uranium Reactor  

E-Print Network [OSTI]

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

Dong, Bao-Guo; Gu, Ji-Yuan

2015-01-01T23:59:59.000Z

322

Method for fabricating laminated uranium composites  

DOE Patents [OSTI]

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.

Chapman, L.R.

1983-08-03T23:59:59.000Z

323

Scrap uranium recycling via electron beam melting  

SciTech Connect (OSTI)

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.

McKoon, R.

1993-11-01T23:59:59.000Z

324

National Uranium Resource Evaluation, Tonopah quadrangle, Nevada  

SciTech Connect (OSTI)

The Tonopah Quadrangle, Nevada, was evaluated using National Uranium Resource Evaluation criteria to identify and delineate areas favorable for uranium deposits. Investigations included reconnaissance and detailed surface geologic and radiometric studies, geochemical sampling and evaluation, analysis and ground-truth followup of aerial radiometric and hydrogeochemical and stream-sediment reconnaissance data, and subsurface data evaluation. The results of these investigations indicate environments favorable for hydroallogenic uranium deposits in Miocene lacustrine sediments of the Big Smoky Valley west of Tonopah. The northern portion of the Toquima granitic pluton is favorable for authigenic uranium deposits. Environments considered unfavorable for uranium deposits include Quaternary sediments; intermediate and mafic volcanic and metavolcanic rocks; Mesozoic, Paleozoic, and Precambrian sedimentary and metasedimentary rocks; those plutonic rocks not included within favorable areas; and those felsic volcanic rocks not within the Northumberland and Mount Jefferson calderas.

Hurley, B W; Parker, D P

1982-04-01T23:59:59.000Z

325

Uranium in prehistoric Indian pottery  

E-Print Network [OSTI]

present in the sample, and the cross l section of the process (the measure of the probability of a neutron interacting with an uranium atom), In general, a daughter product 235 of U fission is analyzed on a detector which counts either gamma rays... for quantitative analysis of various elements on archaeological artifacts, Manganese has been determined in Mesoamerican pot sherds (Bennyhoff and Heizer 1965). A Pu-Be radioisotope neutron source with a flux of 4 x 10 4 -2 -1 neutrons cm sec was used...

Filberth, Ernest William

2012-06-07T23:59:59.000Z

326

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of Tables3 Uranium

327

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of Tables3 Uranium11

328

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of6,2009Uranium

329

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of6,2009UraniumNext

330

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of6,2009UraniumNext

331

U.S.Uranium Reserves  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProved ReservesFeet)per Thousand28 198 18BiomassThree-Dimensional SeismicUranium

332

2013 Uranium Marketing Annual Report  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:SeadovCooperativeA2. World liquids consumption by region,Purchases2 U.S.Feed6a. Uranium

333

2013 Uranium Marketing Annual Report  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:SeadovCooperativeA2. World liquids consumption by region,Purchases2 U.S.Feed6a.4. Uranium

334

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

SciTech Connect (OSTI)

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.

Ladd-Lively, Jennifer L [ORNL] [ORNL

2014-01-01T23:59:59.000Z

335

Intergranular fracture in UO{sub 2}: derivation of traction-separation law from atomistic simulations  

SciTech Connect (OSTI)

In this study, the intergranular fracture behavior of UO{sub 2} 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 ?5 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. (authors)

Zhang, Yongfeng; Millett, P.C.; Tonks, M.R.; Bai, Xian-Ming; Biner, S.B. [Fuels Modeling and Simulation Department, Idaho National Laboratory - INL, Idaho Falls, ID 83415 (United States)

2013-07-01T23:59:59.000Z

336

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

SciTech Connect (OSTI)

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.

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

2013-10-01T23:59:59.000Z

337

Interface control document between PUREX/UO{sub 3} Plant Transition and Solid Waste Disposal Division  

SciTech Connect (OSTI)

This interface control document (ICD) between PUREX/UO{sub 3} Plant Transition (PPT) and Solid Waste Disposal Division (SWD) establishes at a top level the functional responsibilities of each division where interfaces exist between the two divisions. Since the PUREX Transition and Solid Waste Disposal divisions operate autonomously, it is important that each division has a clear understanding of the other division`s expectations regarding these interfaces. This ICD primarily deals with solid wastes generated by the PPT. In addition to delineating functional responsibilities, the ICD includes a baseline description of those wastes that will require management as part of the interface between the divisions. The baseline description of wastes includes waste volumes and timing for use in planning the proper waste management capabilities: the primary purpose of this ICD is to ensure defensibility of expected waste stream volumes and Characteristics for future waste management facilities. Waste descriptions must be as complete as-possible to ensure adequate treatment, storage, and disposal capability will exist. The ICD also facilitates integration of existing or planned waste management capabilities of the PUREX. Transition and Solid Waste Disposal divisions. The ICD does not impact or affect the existing processes or procedures for shipping, packaging, or approval for shipping wastes by generators to the Solid Waste Division.

Duncan, D.R.

1994-06-30T23:59:59.000Z

338

Report of clean out and flushing of UO{sub 3} Plant processing equipment: Revision 1  

SciTech Connect (OSTI)

The UO{sub 3} Plant went through a clean out leading to the deactivation of the facility. This clean out consisted of three phases. Phase 1 consisted of the removal of residual process material and the deactivation of most process equipment and instrumentation. Phase 2 consisted of the fixing or removal of contamination so storm water processing would be no longer required. Phase 3 consisted of the remaining activities that had to be completed before the facility was turned over to the Surplus Facility Program. Since the activities of Phase 2 and 3 were closely related, these two phases were worked simultaneously. The first part of this document summarizes the Phase 1 clean out procedures and their results. Phase 1 was completed on February 28, 1994. The second part summarizes the Phase 2/3 clean out procedures and their results. Phase 2/3 was completed before December 31, 1994. Because tanks and equipment were flushed simultaneously or in a specific sequence, the clean out processes are discussed per workplan.

Gonsalves, E.

1994-12-02T23:59:59.000Z

339

Graphite and Beryllium Reflector Critical Assemblies of UO2 (Benchmark Experiments 2 and 3)  

SciTech Connect (OSTI)

INTRODUCTION A series of experiments was carried out in 1962-65 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2 wt% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 252 tightly-packed fuel rods (1.27-cm triangular pitch) with graphite reflectors [1], the second part used 252 graphite-reflected fuel rods organized in a 1.506-cm triangular-pitch array [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods in a 1.506-cm-triangular-pitch configuration and in a 7-tube-cluster configuration [3]. Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. All three experiments in the series have been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5]. The evaluation of the first experiment in the series was discussed at the 2011 ANS Winter meeting [6]. The evaluations of the second and third experiments are discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power systems [7].

Margaret A. Marshall; John D. Bess

2012-11-01T23:59:59.000Z

340

Reaction of titanium polonides with carbon dioxide  

SciTech Connect (OSTI)

It has been ascertained that heating titanium and tantalum in carbon dioxide to temperatures of 500 or 800/sup 0/C alters the composition of the gas phase, causing the advent of carbon monoxide and lowering the oxygen content. Investigation of the thermal stability of titanium polonides in a carbon dioxide medium has shown that titanium mono- and hemipolonides are decomposed at temperatures below 350/sup 0/C. The temperature dependence of the vapor pressure of polonium produced in the decomposition of these polonides in a carbon dioxide medium have been determined by a radiotensimetric method. The enthalpy of the process, calculated from this relationship, is close to the enthalpy of vaporization of elementary polonium in vacuo.

Abakumov, A.S.; Malyshev, M.L.; Reznikova, N.F.

1987-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

SEQUESTERING CARBON DIOXIDE IN COALBEDS  

SciTech Connect (OSTI)

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.

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

2003-04-30T23:59:59.000Z

342

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

SciTech Connect (OSTI)

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

Francis, C. W.

1993-09-01T23:59:59.000Z

343

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

SciTech Connect (OSTI)

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.

Lovley, Derek R

2012-12-28T23:59:59.000Z

344

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

SciTech Connect (OSTI)

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

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

1995-12-31T23:59:59.000Z

345

Improving Natural Uranium Utilization By Using Thorium in Low Moderation PWRs - A Preliminary Neutronic Scoping Study  

SciTech Connect (OSTI)

The Th-U fuel cycle is not quite self-sustainable when used in water-cooled reactors and with fuel burnups higher than a few thousand of MWd/t characteristic of CANDU reactors operating with a continuous refueling. For the other industrially mature water-cooled reactors (i.e. PWRs and BWRs) it is economically necessary that the fuel has enough reactivity to reach fuel burnups of the order of a few tens of thousand of MWd/t. In this particular case, an additional input of fissile material is necessary to complement the bred fissile U-233. This additional fissile material could be included in the form of Highly Enriched Uranium (HEU) at the fabrication of the Th-U fuel. The objective of this preliminary neutronic scoping study is to determine (1) how much HEU and, consequently, how much natural uranium is necessary in such Th-U fuel cycle with U recycling and (2) how much TRansUranics (TRU=Pu, Np, Am and Cm) are produced. These numbers are then compared with those of a standard UO2 PWR. The thorium reactors considered have a homogeneous hexagonal lattice made up of the same (Th-U)O2 pins. Furthermore, at this point, we are not considering the use of blankets inside or outside the core. The lattice pitch has been varied to estimate the effect of the water-to-fuel volume ratio, and light water as well as heavy water have been considered. For most cases, an average burnup at discharge of 45,000 MWd/t has been considered.

Gilles Youinou; Ignacio Somoza

2010-10-01T23:59:59.000Z

346

Uranium Cluster Chemistry DOI: 10.1002/anie.200906605  

E-Print Network [OSTI]

Uranium Cluster Chemistry DOI: 10.1002/anie.200906605 Tetranuclear Uranium Clusters by Reductive in the coordination chemistry and small-molecule reactivity of uranium. Among the intriguing reactivity patterns of tetravalent uranium with 3,5-dimethylpyrazolate (Me2PzĂ? ) led to forma- tion of an unprecedented homoleptic

347

Breath is a mixture of nitrogen, oxygen, carbon dioxide, water  

E-Print Network [OSTI]

12 SCIENCE Breath is a mixture of nitrogen, oxygen, carbon dioxide, water vapour, inert gases. On the basis of proton affinity, the major constituents of air and breath (nitrogen, oxygen, carbon dioxide

348

A methodology for forecasting carbon dioxide flooding performance  

E-Print Network [OSTI]

A methodology was developed for forecasting carbon dioxide (CO2) flooding performance quickly and reliably. The feasibility of carbon dioxide flooding in the Dollarhide Clearfork "AB" Unit was evaluated using the methodology. This technique is very...

Marroquin Cabrera, Juan Carlos

1998-01-01T23:59:59.000Z

349

Carbon Dioxide Capture/Sequestration Tax Deduction (Kansas)  

Broader source: Energy.gov [DOE]

Carbon Dioxide Capture/Sequestration Tax Deduction allows a taxpayer a deduction to adjusted gross income with respect to the amortization of the amortizable costs of carbon dioxide capture,...

350

Louisiana Geologic Sequestration of Carbon Dioxide Act (Louisiana)  

Broader source: Energy.gov [DOE]

This law establishes that carbon dioxide and sequestration is a valuable commodity to the citizens of the state. Geologic storage of carbon dioxide may allow for the orderly withdrawal as...

351

The Greenness of Cities: Carbon Dioxide Emissions and Urban Development  

E-Print Network [OSTI]

carbon dioxide emissions per 1,000 cubic feet of natural gas. In this case, there is much less energy

Glaeser, Edward L.; Kahn, Matthew E.

2008-01-01T23:59:59.000Z

352

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

SciTech Connect (OSTI)

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.

Fisher, D.R.; Kathren, R.L.; Swint, M.J. (Pacific Northwest Laboratory, Richland, WA (USA))

1991-03-01T23:59:59.000Z

353

Technical Basis for Assessing Uranium Bioremediation Performance  

SciTech Connect (OSTI)

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.

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

2008-04-01T23:59:59.000Z

354

Thermochemical Modeling of the Uranium-Cerium-Oxygen System  

SciTech Connect (OSTI)

The objective of the Fuel Cycle R&D Program, Advanced Fuels campaign is to provide the research and development necessary to develop low loss, high quality nuclear fuels for ultra-high burnup reactor operation. Primary work in this area will be focused on the ceramic and metallic fuel systems. The goal of the current work is to enhance the understanding of ceramic nuclear fuel thermochemistry to support fuel research and development efforts. The thermochemical behavior of oxide nuclear fuel under irradiation is dependent on the oxygen to metal ratio (O:M). In fluorite-structured fuel, the actinide metal cation is bonded with {approx}2 oxygen atoms on a crystal lattice and as the metal atoms fission, fission fragments and free oxygen are created. The resulting fission fragments will contain some oxide forming elements, however these are insufficient to bind to all the liberated oxygen and therefore, there is an average increase in O:M with fuel burnup. Some of the fission products also form species that will migrate to and react with the cladding surface in a phenomenon known as Fuel Clad Chemical Interaction (FCCI). Cladding corrosion is life-limiting so it is desirable to understand influencing factors, such as oxide thermochemistry, which can be used to guide the design and fabrication of higher burn up fuel. A phased oxide fuel thermochemical model development effort is underway within the Advanced Fuels Campaign. First models of binary oxide systems are developed. For nuclear fuel system this means U-O and transuranic systems such as Pu-O, Np-O and Am-O. Next, the binary systems will be combined to form pseudobinary systems such as U-Pu-O, etc. The model development effort requires the use of data to allow optimization based on known thermochemical parameters as a function of composition and temperature. Available data is mined from the literature and supplemented by experimental work as needed. Due to the difficulty of performing fuel fabrication development with actinide materials, fundamental studies with uranium are performed using surrogate materials as stand-ins for transuranic elements. In most cases, cerium can be used as a suitable substitute for plutonium when performing O:M and sintering kinetics studies because of identical valence states. Differences exist between the magnitude of reported thermodynamic data of (U,Pu)O{sub x} and (U,Ce)O{sub x}, however the change in oxygen potential versus O:M follows the same trend for both systems. Cerium is also a major fission product element, and thus understanding its behavior in fuel is an important issue as well.

Voit, Stewart L [ORNL; Besmann, Theodore M [ORNL

2010-10-01T23:59:59.000Z

355

Air Pollution XVI 247 Emissions of Nitrogen Dioxide from Modern  

E-Print Network [OSTI]

Air Pollution XVI 247 Emissions of Nitrogen Dioxide from Modern Diesel Vehicles G.A. Bishop and D negative implications for local photochemical ozone production. Keywords: Nitrogen dioxide, automobile strategies, Lemaire [1] suggests that nitrogen dioxide (NO2) was forgotten as a separate component of the NOx

Denver, University of

356

Thermal Infrared Radiation and Carbon Dioxide in the Atmosphere  

E-Print Network [OSTI]

dioxide Water vapor #12;Atmospheric composition (parts per million by volume) · Nitrogen (N2) 780Thermal Infrared Radiation and Carbon Dioxide in the Atmosphere Bill Satzer 3M Company #12;Outline,840 · Oxygen (O2) 209,460 · Argon (Ar) 9340 · Carbon dioxide (CO2) 394 · Methane (CH4) 1.79 · Ozone (O3) 0

Olver, Peter

357

Nanostructured Tin Dioxide Materials for Gas Sensor Applications  

E-Print Network [OSTI]

CHAPTER 30 Nanostructured Tin Dioxide Materials for Gas Sensor Applications T. A. Miller, S. D) levels for some species. Tin dioxide (also called stannic oxide or tin oxide) semi- conductor gas sensors undergone extensive research and development. Tin dioxide (SnO2) is the most important material for use

Wooldridge, Margaret S.

358

Designed amyloid fibers as materials for selective carbon dioxide capture  

E-Print Network [OSTI]

Designed amyloid fibers as materials for selective carbon dioxide capture Dan Lia,b,c,1 , Hiroyasu demonstrate that amyloids, self-assembling protein fibers, are effective for selective carbon dioxide capture. Solid-state NMR proves that amyloid fibers containing alkylamine groups reversibly bind carbon dioxide

359

Array of titanium dioxide nanostructures for solar energy utilization  

DOE Patents [OSTI]

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.

Qiu, Xiaofeng; Parans Paranthaman, Mariappan; Chi, Miaofang; Ivanov, Ilia N; Zhang, Zhenyu

2014-12-30T23:59:59.000Z

360

Glutamate Surface Speciation on Amorphous Titanium Dioxide and  

E-Print Network [OSTI]

Glutamate Surface Speciation on Amorphous Titanium Dioxide and Hydrous Ferric Oxide D I M I T R I (HFO) and titanium dioxide exhibit similar strong attachment of many adsorbates including biomolecules on amorphous titanium dioxide. The results indicate that glutamate adsorbs on HFO as a deprotonated divalent

Sverjensky, Dimitri A.

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Chukwuemeka I. Okoye Carbon Dioxide Solubility and Absorption Rate in  

E-Print Network [OSTI]

Copyright by Chukwuemeka I. Okoye 2005 #12;Carbon Dioxide Solubility and Absorption Rate _______________________ Nicholas A. Peppas #12;Carbon Dioxide Solubility and Absorption Rate in Monoethanolamine/Piperazine/H2O for. #12;iii Carbon Dioxide Solubility and Absorption Rate in Monoethanolamine/Piperazine/H2O

Rochelle, Gary T.

362

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

E-Print Network [OSTI]

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

Duhovic, Selma

2012-01-01T23:59:59.000Z

363

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

E-Print Network [OSTI]

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

Rao, Linfeng

2011-01-01T23:59:59.000Z

364

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

E-Print Network [OSTI]

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

Hwang, Chiachi

2009-01-01T23:59:59.000Z

365

Capstone Depleted Uranium Aerosols: Generation and Characterization  

SciTech Connect (OSTI)

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

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

2004-10-19T23:59:59.000Z

366

Introduction Air Quality and Nitrogen Dioxide  

E-Print Network [OSTI]

- Global update 2005. Primary sources of air pollutants include combustion products from power generationIntroduction Air Quality and Nitrogen Dioxide Air pollution can be defined as "the presence effects to man and/or the environment". (DEFRA) "Clean air is considered to be a basic requirement

367

Carbon Dioxide Corrosion: Modelling and Experimental Work  

E-Print Network [OSTI]

Carbon Dioxide Corrosion: Modelling and Experimental Work Applied to Natural Gas Pipelines Philip in the corrosion related research institutions at IFE and the Ohio University or any other scientific research;#12;Introduction - v - Summary CO2 corrosion is a general problem in the industry and it is expensive. The focus

368

Atmospheric Lifetime of Fossil Fuel Carbon Dioxide  

E-Print Network [OSTI]

Atmospheric Lifetime of Fossil Fuel Carbon Dioxide David Archer,1 Michael Eby,2 Victor Brovkin,3 released from combustion of fossil fuels equilibrates among the various carbon reservoirs of the atmosphere literature on the atmospheric lifetime of fossil fuel CO2 and its impact on climate, and we present initial

Scherer, Norbert F.

369

Hydroelectric Reservoirs -the Carbon Dioxide and Methane  

E-Print Network [OSTI]

Hydroelectric Reservoirs - the Carbon Dioxide and Methane Emissions of a "Carbon Free" Energy an overview on the greenhouse gas production of hydroelectric reservoirs. The goals are to point out the main how big the greenhouse gas emissions from hydroelectric reservoirs are compared to thermo-power plants

Fischlin, Andreas

370

Acid sorption regeneration process using carbon dioxide  

DOE Patents [OSTI]

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.

King, C. Judson (Kensington, CA); Husson, Scott M. (Anderson, SC)

2001-01-01T23:59:59.000Z

371

Carbon dioxide storage professor Martin Blunt  

E-Print Network [OSTI]

of CCS storage there are over a hundred sites worldwide where Co2 is injected under- ground as partCarbon dioxide storage professor Martin Blunt executive summary Carbon Capture and Storage (CCS and those for injection and storage in deep geological formations. all the individual elements operate today

372

Carbon Dioxide Capture from Coal-Fired  

E-Print Network [OSTI]

. LFEE 2005-002 Report #12;#12;i ABSTRACT Investments in three coal-fired power generation technologiesCarbon Dioxide Capture from Coal-Fired Power Plants: A Real Options Analysis May 2005 MIT LFEE 2005 environment. The technologies evaluated are pulverized coal (PC), integrated coal gasification combined cycle

373

Carbon Dioxide Corrosion and Inhibition Studies  

E-Print Network [OSTI]

· Corrosion inhibition very important in the oil industry · Film forming inhibitors containing nitrogenCarbon Dioxide Corrosion and Inhibition Studies Kristin Gilida #12;Outline · Background = Zreal + Zim Rp 1/Corr Rate #12;Tafel · Measures corrosion rate directly · Measures iCORR from A and C

Petta, Jason

374

Crystal Chemistry of Early Actinides (Thorium, Uranium, and Neptunium) and Uranium Mesoporous Materials.  

E-Print Network [OSTI]

??Despite their considerable global importance, the structural chemistry of actinides remains understudied. Thorium and uranium fuel cycles are used in commercial nuclear reactors in India… (more)

Sigmon, Ginger E.

2010-01-01T23:59:59.000Z

375

Prokaryotic microorganisms in uranium mining waste piles and their interactions with uranium and other heavy metals.  

E-Print Network [OSTI]

??The influence of uranyl and sodium nitrate under aerobic and anaerobic conditions on the microbial community structure of a soil sample from the uranium mining… (more)

Geißler, Andrea

2007-01-01T23:59:59.000Z

376

Depleted uranium disposition study -- Supplement, Revision 1  

SciTech Connect (OSTI)

The Department of Energy Office of Weapons and Materials Planning has requested a supplemental study to update the recent Depleted Uranium Disposition report. This supplemental study addresses new disposition alternatives and changes in status.

Becker, G.W.

1993-11-01T23:59:59.000Z

377

In situ remediation of uranium contaminated groundwater  

SciTech Connect (OSTI)

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

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

1997-12-31T23:59:59.000Z

378

In situ remediation of uranium contaminated groundwater  

SciTech Connect (OSTI)

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

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

1997-02-01T23:59:59.000Z

379

Process for reducing beta activity in uranium  

DOE Patents [OSTI]

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.

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

1986-01-01T23:59:59.000Z

380

Method of recovering uranium from aqueous solution  

SciTech Connect (OSTI)

Anion exchange resin derived from insoluble crosslinked polymers of vinyl benzyl chloride which are prepared by polymerizing vinyl benzyl chloride and a crosslinking monomer are particularly suitable in the treatment of uranium bearing leach liquors.

Albright, R.L.

1980-01-22T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Innovative design of uranium startup fast reactors  

E-Print Network [OSTI]

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

Fei, Tingzhou

2012-01-01T23:59:59.000Z

382

Process for reducing beta activity in uranium  

DOE Patents [OSTI]

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.

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

1985-04-11T23:59:59.000Z

383

Depleted uranium: A DOE management guide  

SciTech Connect (OSTI)

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.

NONE

1995-10-01T23:59:59.000Z

384

The ultimate disposition of depleted uranium  

SciTech Connect (OSTI)

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.

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

1991-12-31T23:59:59.000Z

385

BIOREMEDIATION OF URANIUM CONTAMINATED SOILS AND WASTES.  

SciTech Connect (OSTI)

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

FRANCIS,A.J.

1998-09-17T23:59:59.000Z

386

Material property correlations for uranium mononitride  

E-Print Network [OSTI]

MATERIAL PROPERTY CORRELATIONS FOR URANIUM MONONITRIDE A Thesis by STEVEN LOWE HAYES Submitted to the Office of Graduate Studies of Texas ARM University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE August... 1989 Major Subject: Nuclear Engineering MATERIAL PROPERTY CORRELATIONS FOR URANIUM MONONITRIDE A Thesis by STEVEN LOWE HAYES Approved as to style and content by: K. L. Peddicord (Chair of Committee) R. R. Hart (Member) C. P. Burger (Member...

Hayes, Steven Lowe

2012-06-07T23:59:59.000Z

387

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

SciTech Connect (OSTI)

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

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

2012-10-30T23:59:59.000Z

388

Electrochemical method of producing eutectic uranium alloy and apparatus  

DOE Patents [OSTI]

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.

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

1995-01-01T23:59:59.000Z

389

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

SciTech Connect (OSTI)

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

NONE

1995-07-05T23:59:59.000Z

390

Instrument Development and Measurements of the Atmospheric Pollutants Sulfur Dioxide, Nitrate Radical, and Nitrous Acid by Cavity Ring-down Spectroscopy and Cavity Enhanced Absorption Spectroscopy  

E-Print Network [OSTI]

A. , A method of nitrogen dioxide and sulphur dioxidedetermination of nitrogen dioxide and sulfur dioxide in theDOAS) have measured nitrogen dioxide (NO 2 ), nitrate

Medina, David Salvador

2011-01-01T23:59:59.000Z

391

Monte Carlo analysis of burnup-dependent plutonium concentration profiles in UO{sub 2} and MOX fuel pins  

SciTech Connect (OSTI)

The ability to accurately predict fuel performance is an essential requirement for fuel design studies. Prediction of plutonium concentration profiles in an irradiated fuel pin is important for fuel performance analysis and spent-fuel storage. The MCNP coupling with ORIGEN2 (MCWO) burnup calculation code as demonstrated in this paper can analyze the rim effect in UO{sub 2} and mixed-oxide (MOX) fuel pins. Acceptance of a code such as MCWO depends very strongly on its validation. Validation involves the benchmark of the code predictions to the in-pile experimental data and results of post-irradiation examinations (PIEs). In this paper, a validation was made by comparing the MCWO calculated results with the VIM-BURN code, which has been validated against PIE data. The validated MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. In this paper, Pu concentration (wt%) and fission power profiles versus burnup of UO{sub 2} and reactor-grade (RG)-MOX fuel pins were calculated with MCWO, and results are discussed.

Chang, G.S. [Lockheed Martin Idaho Technologies, Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

1998-09-01T23:59:59.000Z

392

D10 experiment: coolability of UO/sub 2/ debris in sodium with downward heat removal. [LMFBR  

SciTech Connect (OSTI)

The LMFBR Debris Coolability Program at Sandia National Laboratories investigates the coolability of particle beds which may form following a severe accident involving core disassembly in a nuclear reactor. The D series experiments utilize fission heating of fully enriched UO/sub 2/ particles submerged in sodium to realistically simulate decay heating. The D10 experiment is the first in the series to study the effects of bottom cooling of the debris that could be provided in an actual accident condition by structural materials onto which the debris might settle. Additionally, the D10 experiment was designed to achieve maximum temperatures in the debris approaching the melting point of UO/sub 2/. The experiment was successfully operated for over 50 hours and investigated downward heat removal in a packed bed at specific powers of 0.16 to 0.58 W/g. Dryout in the debris was achieved at powers from 0.42 to 0.58 W/g. Channels were induced in the bed and channeled bed dryout was achieved at powers of 1.06 to 1.77 W/g. Maximum temperatures in excess of 2500/sup 0/C were attained.

Mitchell, G.W.; Ottinger, C.A.; Meister, H.

1984-12-01T23:59:59.000Z

393

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

SciTech Connect (OSTI)

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.

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

2014-11-01T23:59:59.000Z

394

Coolability of stratified UO/sub 2/ debris in sodium with downward heat removal: The D13 experiment  

SciTech Connect (OSTI)

The LMFBR Debris Coolability Program at Sandia National Laboratories investigates the coolability of particle beds that may form following a severe accident involving core disassembly in a nuclear reactor. The D series experiments utilize fission heating of fully enriched UO/sub 2/ particles submerged in sodium to realistically simulate decay heating. The D13 experiment is the first in the series to study the effects of bottom cooling of stratified debris, which could be provided in an actual accident condition by structural materials onto which the debris might settle. Additionally, the D13 experiment was designed to achieve maximum temperatures in the debris approaching the melting point of UO/sub 2/. The experiment was operated for over 40 hours and investigated downward heat removal at specific powers of 0.22 to 2.58 W/g. Channeled dryout in the debris was achieved at powers from 0.94 to 2.58 W/g. Maximum temperatures approaching 2700/sup 0/C were attained. Bottom heat removal was up to 750 kW/m/sup 2/ as compared to 450 kW/m/sup 2/ in the D10 experiment.

Ottinger, C.A.; Mitchell, G.W.; Reed, A.W.; Meister, H.

1987-03-01T23:59:59.000Z

395

Anisotropic reactive ion etching of vanadium dioxide  

E-Print Network [OSTI]

. Weichold Vanadium dioxide (V02) was anisotropically reactive ion etched using carbon tetrafluoride (CF4) . CF4, as an etch gas, provided the chemistry along with the control needed to achieve an anisotropic etch. This chemistry was practically inert... with vanadium quite easily. This leads to interest in using a fluorine- based chemistry. The goal of this research is to produce a selective anisotropic reactive ion etch for VO2 /photoresist using only carbon tetrafluoride (CFq) . Reactive ion etching...

Radle, Byron K

1990-01-01T23:59:59.000Z

396

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...

397

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

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

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

398

Assessments of long-term uranium supply availability  

E-Print Network [OSTI]

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

Zaterman, Daniel R

2009-01-01T23:59:59.000Z

399

Prospects for the recovery of uranium from seawater  

E-Print Network [OSTI]

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

Best, F. R.

1980-01-01T23:59:59.000Z

400

Monitoring Uranium Transformations Determined by the Evolution of Biogeochemical Processes  

SciTech Connect (OSTI)

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

Marsh, Terence L.

2013-07-30T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Depleted Uranium in Kosovo Post-Conflict Environmental Assessment  

E-Print Network [OSTI]

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

Unep Scientific; Mission Kosovo

402

Uranium Mill Tailings Remedial Action Project surface project management plan  

SciTech Connect (OSTI)

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

Not Available

1994-09-01T23:59:59.000Z

403

Microbial Janitors: Enabling natural microbes to clean up uranium contamination  

E-Print Network [OSTI]

Microbial Janitors: Enabling natural microbes to clean up uranium contamination Oak Ridge to the development of the atomic bomb. Uranium enrichment activities on the Oak Ridge Reservation in the 1940s until then the uranium and nitrate contamination has spread through the ground and now covers an area of about 7 km

404

Plutonium recovery from spent reactor fuel by uranium displacement  

DOE Patents [OSTI]

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.

Ackerman, John P. (Downers Grove, IL)

1992-01-01T23:59:59.000Z

405

Standard Review Plan for In Situ Leach Uranium  

E-Print Network [OSTI]

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

406

EPA Uranium Program Update Loren W. Setlow and  

E-Print Network [OSTI]

30, 2008 #12;2 Overview EPA Radiation protection program Uranium reports and abandoned mine lands and Liability Act #12;4 Uranium Reports and Abandoned Mine Lands Program ·Technologically Enhanced Naturally Occurring Radioactive Materials from Uranium Mining, Volume I: Mining and Reclamation Background (Revised

407

Soil to plant transfer of 238 Th on a uranium  

E-Print Network [OSTI]

Soil to plant transfer of 238 U, 226 Ra and 232 Th on a uranium mining-impacted soil from species grown in soils from southeastern China contaminated with uranium mine tailings were analyzed The radioactive waste (e.g. tailings) produced by uranium mining activities contains a series of long

Hu, Qinhong "Max"

408

Plutonium recovery from spent reactor fuel by uranium displacement  

DOE Patents [OSTI]

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.

Ackerman, J.P.

1992-03-17T23:59:59.000Z

409

Composition of the U.S. DOE Depleted Uranium Inventory  

E-Print Network [OSTI]

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

Concentration Of Less

410

Modeling Uranium-Proton Ion Exchange in Biosorption  

E-Print Network [OSTI]

threatening heavy metals because of its high toxicity and some radioactivity. Excessive amounts of uranium seaweed biomass was used to remove the heavy metal uranium from the aqueous solution. Uranium biosorption the heavy metal uptake performance of different biosorbents.LangmuirandFreundlichmodelsoftengenerally fit

Volesky, Bohumil

411

Estimating terrestrial uranium and thorium by antineutrino flux measurements  

E-Print Network [OSTI]

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

Mcdonough, William F.

412

A Geostatistical Study of the Uranium Deposit at Kvanefjeld,  

E-Print Network [OSTI]

with the geology. It is also shown that, although anisotropy exists, the uranium variation has a secondRisa-R-468 A Geostatistical Study of the Uranium Deposit at Kvanefjeld, The Ilimaussaq Intrusion A GEOSTATISTICAL STUDY OF THE URANIUM DEPOSIT AT KVANEFJELD, THE ILIMAUSSAQ INTRUSION, SOUTH GREENLAND Flemming

413

Depleted uranium plasma reduction system study  

SciTech Connect (OSTI)

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

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

1994-12-01T23:59:59.000Z

414

Gel and process for preventing carbon dioxide break through  

SciTech Connect (OSTI)

A process is described for retarding the flow of carbon dioxide in carbon dioxide break-through fingers in a subterranean formation, the process comprising: (a) introducing a gas selected from the group consisting of carbon dioxide and gases containing carbon dioxide into a subterranean deposit containing carbon dioxide break-through fingers; (b) after the carbon dioxide break-through fingers have sorbed a predetermined amount of the gas, stopping the flow of the gas into the subterranean formation, (c) after stopping the flow of the gas into the subterranean formation, introducing an effective amount of a gel-forming composition into the subterranean formation and into the carbon dioxide break-through fingers, the gel-forming composition being operable, when contacting carbon dioxide break-through fingers containing the brine which has absorbed substantial amounts of carbon dioxide to form a gel in the fingers which is operable for retarding the flow of the gas in the finger. The gel-forming composition comprises: i. an aqueous solution comprising a first substance selected from the group consisting of polyvinyl alcohols, polyvinyl alcohol copolymers, and mixtures thereof, and ii. an amount of a second substance selected from the group consisting of aldehydes, aldehyde generating substances, acetals, acetal generating substances, and mixtures thereof.

Sandiford, B.B.; Zillmer, R.C.

1987-06-16T23:59:59.000Z

415

The Greenness of Cities: Carbon Dioxide Emissions and Urban Development  

E-Print Network [OSTI]

dioxide impact of electricity consumption in different majorand residential electricity consumption. Car usage and homefor fuel oil and electricity consumption. We then use

Glaeser, Edward L.; Kahn, Matthew E.

2008-01-01T23:59:59.000Z

416

Carbon dioxide absorbent and method of using the same  

SciTech Connect (OSTI)

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.

Perry, Robert James; O'Brien, Michael Joseph

2014-06-10T23:59:59.000Z

417

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...

418

assisted silicon dioxide: Topics by E-print Network  

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

dioxide substrates is described. The approach consists of solid such as displays and thin-film polycrystalline solar cells. Particularly important for low- cost thin-film solar...

419

Synthesis and structure of Cs[UO{sub 2}(SeO{sub 4})(OH)] . nH{sub 2}O (n = 1.5 or 1)  

SciTech Connect (OSTI)

The synthesis and single-crystal X-ray diffraction study of Cs[UO{sub 2}(SeO{sub 4})(OH)] . 1.5H{sub 2}O (I) and Cs[UO{sub 2}(SeO{sub 4})(OH)] . H{sub 2}O (II) are performed. Compound I crystallizes in the monoclinic crystal system, a = 7.2142(2) A, b = 14.4942(4) A, c = 8.9270(3) A, {beta} = 112.706(1){sup o}, space group P2{sub 1}/m, Z = 4, and R = 0.0222. Compound II is monoclinic, a = 8.4549(2) A, b = 11.5358(3) A, c = 9.5565(2) A, {beta} = 113.273(1){sup o}, space group P2{sub 1}/c, Z = 4, and R = 0.0219. The main structural units of crystals I and II are [UO{sub 2}(SeO{sub 4})(OH)]{sup -} layers which belong to the AT{sup 3}M{sup 2} crystal chemical group of uranyl complexes (A = UO{sub 2}{sup 2+}, T{sup 3} = SeO{sub 4}{sup 2-}, and M{sup 2} = OH{sup -}). In structure I, johannite-like layers are found. Structure II is a topological isomer of I. The two structures differ in the number of U(VI) atoms bound to the central atom by all bridging ligands.

Serezhkina, L. B., E-mail: lserezh@ssu.samara.ru [Samara State University (Russian Federation); Peresypkina, E. V.; Virovets, A. V. [Russian Academy of Sciences, Institute of Inorganic Chemistry, Siberian Branch (Russian Federation); Pushkin, D. V.; Verevkin, A. G. [Samara State University (Russian Federation)

2010-05-15T23:59:59.000Z

420

Possible Bose-condensated Behavior in a Quantum Phase Originating in a Collective Excitation in the Chemically and Optically Doped Mott-Hubbard System UO2+x  

SciTech Connect (OSTI)

The pinned charge defects in U4O9, and U3O7 that are the single phase fluoritestructured derivatives of UO2 have been characterized by U L3 EXAFS at 30, 100, and 200 K, xray and neutron pair distribution function analysis, O K edge XAS and non-resonant inelastic xray scattering, and Raman spectroscopy, while mobile charge defects were investigated by femtosecond time-resolved pump-probe laser spectroscopy on single crystal UO2 between 7 and 300 K. The results from all of these measurements show highly complex and anomalous behaviors, which we attribute to a charge-lattice instability in UO2 that most likely originates in the intersection of the ground U(IV) and a proximate uranyl-like excited state in a conic section, causing a breakdown of the Born-Oppenheimer approximation. Furthermore, the photoinduced quasiparticles undergo a gap-opening condensation between 50 and 60 K. Doped UO2 may therefore exhibit novel correlated electron physics that extends beyond that of the cuprate-manganite-pnictide family of compounds.

Conradson, Steven D.; Durakiewicz, Tomasz; Espinosa-Faller, Francisco J.; An, Yong Q.; Andersson , David; Bishop, Alan R.; Boland, Kevin S.; Bradley, Joseph A.; Byler, Darrin D.; Clark, David L.; Conradson, Dylan R.; Conradson, Leilani L.; Costello, Alison E.; Hess, Nancy J.; Lander, Gerard H.; Llobet, Anna; Martucci, Mary B.; de Leon, Jose M.; Nordlund, Dennis; Lezama-Pacheco, Juan S.; Proffen, Thomas E.; Rodriguez, George; Schwarz, Daniel E.; Seidler, Gerald T.; Taylor, Antoinette; Trugman, Stuart A.; Tyson, Trevor A.; Valdez, James A.

2013-09-23T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

UO Purchasing & Contracting Services rev: August 13, 2012 Text Requirements for Level 1 and Level 2 Signature Authority Purchase Orders and Invoices.  

E-Print Network [OSTI]

UO Purchasing & Contracting Services rev: August 13, 2012 Text Requirements for Level 1 and Level 2 the payment is due and payable (may be same person exercising L1SA)] · Either: 1. List contract # or purchase number); of 2. If you do not have a written contract or purchase order, generally describe what goods and

422

UO Purchasing & Contracting Services rev: August 19, 2014 Text Requirements for Level 1 and Level 2 Contracting Authority Purchase Orders and Invoices.  

E-Print Network [OSTI]

UO Purchasing & Contracting Services rev: August 19, 2014 Text Requirements for Level 1 and Level 2 Contracting Authority Purchase Orders and Invoices. Purchase Orders - Level 1 Contracting Authority: · L1CA [Insert on First Line of Document Text] · [name of individual exercising Level 1 Contracting Authority

423

UO Department of Chemistry -Faculty Research Interests Berglund, Andy -The primary goals of the Berglund lab are to understand the molecular basis of the human disease myotonic  

E-Print Network [OSTI]

applications in solar energy harvesting and electrochemical energy storage. Chartoff, Richard - The UO Polymer and thermodynamics of quantum states of molecules embedded in a quantum environment. Lonergan, Mark C. - Research interesting electrical and electrochemical phenomena in solid-state systems. Marcus, Andrew - The Marcus group

Cina, Jeff

424

UoE Employees How to gain access to internal vacancies As a current University employee, you will be eligible to access (via the jobs site) vacancies  

E-Print Network [OSTI]

UoE Employees ­ How to gain access to internal vacancies As a current University employee, you will be eligible to access (via the jobs site) vacancies advertised internally, in addition to those advertised gain access to all vacancies (including those advertised to internal applicants only) whenever you log

Edinburgh, University of

425

Evaporation of Enriched Uranium Solutions Containing Organophosphates  

SciTech Connect (OSTI)

The Savannah River Site has enriched uranium (EU) solution which has been stored for almost 10 years since being purified in the second uranium cycle of the H area solvent extraction process. The preliminary SRTC data, in conjunction with information in the literature, is promising. However, very few experiments have been run, and none of the results have been confirmed with repeat tests. As a result, it is believed that insufficient data exists at this time to warrant Separations making any process or program changes based on the information contained in this report. When this data is confirmed in future testing, recommendations will be presented.

Pierce, R.A.

1999-03-18T23:59:59.000Z

426

Decarburization of uranium via electron beam processing  

SciTech Connect (OSTI)

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

McKoon, R H

1998-10-23T23:59:59.000Z

427

Progress toward uranium scrap recycling via EBCHR  

SciTech Connect (OSTI)

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

McKoon, R.H.

1994-11-01T23:59:59.000Z

428

Laboratory Directed Research and Development (LDRD) on Mono-uranium Nitride Fuel Development for SSTAR and Space Applications  

SciTech Connect (OSTI)

The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy & Environment (E&E) and Chemistry & Material Sciences (C&MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E&E and C&MS Directorates co-sponsored this Laboratory Directed Research & Development (LDRD) Project on Mono-Uranium Nitride Fuel Development for SSTAR and Space Applications. In fact, three out of the six GEN-IV reactor concepts consider using the nitride-based fuel, as shown in Table 1. SSTAR is a liquid-metal cooled, fast reactor. It uses nitride fuel in a sealed reactor vessel that could be shipped to the user and returned to the supplier having never been opened in its long operating lifetime. This sealed reactor concept envisions no fuel refueling nor on-site storage of spent fuel, and as a result, can greatly enhance proliferation resistance. However, the requirement for a sealed, long-life core imposes great challenges to research and development of the nitride fuel and its cladding. Cladding is an important interface between the fuel and coolant and a barrier to prevent fission gas release during normal and accidental conditions. In fabricating the nitride fuel rods and assemblies, the cladding material should be selected based on its the coolant-side corrosion properties, the chemical/physical interaction with the nitride fuel, as well as their thermal and neutronic properties. The US NASA space reactor, the SP-100 was designed to use mono-uranium nitride fuel. Although the SP-100 reactor was not commissioned, tens of thousand of nitride fuel pellets were manufactured and lots of them, cladded in Nb-1-Zr had been irradiated in fast test reactors (FFTF and EBR-II) with good irradiation results. The Russian Naval submarines also use nitride fuel with stainless steel cladding (HT-9) in Pb-Bi coolant. Although the operating experience of the Russian submarine is not readily available, such combination of fuel, cladding and coolant has been proposed for a commercial-size liquid-metal cooled fast reactor (BREST-300). Uranium mono-nitride fuel is studied in this LDRD Project due to its favorable properties such as its high actinide density and high thermal conductivity. The thermal conductivity of mono-nitride is 10 times higher than that of oxide (23 W/m-K for UN vs. 2.3 W/m-K for UO{sub 2} at 1000 K) and its melting temperature is much higher than that of metal fuel (2630 C for UN vs. 1132 C for U metal). It also has relatively high actinide density, (13.51 gU/cm{sup 3} in UN vs. 9.66 gU/cm{sup 3} in UO{sub 2}) which is essential for a compact reactor core design. The objective of this LDRD Project is to: (1) Establish a manufacturing capability for uranium-based ceramic nuclear fuel, (2) Develop a computational capability to analyze nuclear fuel performance, (3) Develop a modified UN-based fuel that can support a compact long-life reactor core, and (4) Collaborate with the Nuclear Engineering Department of UC Berkeley on nitride fuel reprocessing and disposal in a geologic repository.

Choi, J; Ebbinghaus, B; Meiers, T; Ahn, J

2006-02-09T23:59:59.000Z

429

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

E-Print Network [OSTI]

Simplifying strong electronic correlations in uranium: Localized uranium heavy-fermion UM2Zn20 (M AtĂłmica, 8400 Bariloche, Argentina 6 Department of Chemistry and Biochemistry, University of Delaware-field effects corroborate an ionic-like uranium electronic configura- tion in UM2Zn20. DOI: 10.1103/PhysRevB.78

Lawrence, Jon

430

Sulfur dioxide removal by enhanced electrostatics  

SciTech Connect (OSTI)

The economic removal of sulfur dioxide (SO{sub 2}) still represents a significant technical challenge which could determine the use of certain types of fossil fuels for energy production. This paper will present the preliminary results of an innovative research project utilizing a low-cost wet electrostatic precipitator to remove sulfur dioxide. There are many aspects for gas removal in an electrostatic precipitator which are not currently being used. This project utilizes electron attachment of free electrons onto gas molecules and ozone generation to remove sulfur dioxide which is a typical flue gas pollutant. This research was conducted on a bench-scale, wet electrostatic precipitator. A direct-current negative discharge corona is used to generate the ozone in-situ. This ozone will be used to oxidize SO{sub 2} to form sulfuric acid, which is very soluble in water. However, it is believed that the primary removal mechanism is electron attachment of the free electrons from the corona which force the SO{sub 2} to go to equilibrium with the water and be removed from the gas stream. Forcing the equilibrium has been shown to achieve removal efficiencies of up to 70%. The bench scale unit has been designed to operate wet or dry, positive and negative for comparison purposes. The applied dc voltage is variable from 0 to 100 kV, the flow rate is a nominal 7 m{sup 3}/hr and the collecting electrode area is 0.20 m{sup 2}. Tests are conducted on a simulated flue gas stream with SO{sub 2} ranging from 0 to 4,000 ppmv. This paper presents the results of tests conducted to determine the effect of operating conditions on removal efficiency. The removal efficiency was found to vary with gas residence time, water flow rate, inlet concentration, applied power, and the use of corona pulsing.

Larkin, K.; Tseng, C.; Keener, T.C.; Khang, S.J. [Univ. of Cincinnati, OH (United States)

1997-12-31T23:59:59.000Z

431

Nitrogen dioxide, sulfur dioxide, and ammonia detector for remote sensing of vehicle emissions  

E-Print Network [OSTI]

with sulfuric and nitric acids formed from at- mospheric oxidations of sulfur dioxide SO2 and nitrogen oxides mobile sources comes from the combustion of sulfur compounds in fuel. The U.S. is in the process of reducing sulfur in fuel for all mobile sources. This process begins with ultralow sulfur on-road diesel

Denver, University of

432

Retrieval of ozone and nitrogen dioxide concentrations from Stratospheric Aerosol and Gas Experiment III (SAGE III)  

E-Print Network [OSTI]

Retrieval of ozone and nitrogen dioxide concentrations from Stratospheric Aerosol and Gas extinction. We retrieve ozone and nitrogen dioxide number densities and aerosol extinction from transmission), Retrieval of ozone and nitrogen dioxide concentrations from Stratospheric Aerosol and Gas Experiment III

433

6/4/2013 Page 1 of 12 Nitrogen Dioxide SOP Standard Operating Procedures  

E-Print Network [OSTI]

6/4/2013 Page 1 of 12 Nitrogen Dioxide SOP Standard Operating Procedures Nitrogen Dioxide and Nitric Oxide Print a copy and insert into your laboratory the precautions and safe handling procedures for the use of Nitrogen Dioxide

Cohen, Ronald C.

434

Satellite observations of ozone and nitrogen dioxide: from retrievals to emission estimates  

E-Print Network [OSTI]

Satellite observations of ozone and nitrogen dioxide: from retrievals to emission estimates #12 Satellite observations of ozone and nitrogen dioxide: from retrievals to emission es- timates / by Bas Subject headings: satellite retrieval / nitrogen dioxide / ozone / air pollution / emis- sion estimates

Haak, Hein

435

Apparatus for extracting and sequestering carbon dioxide  

DOE Patents [OSTI]

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.

Rau, Gregory H. (Castro Valley, CA); Caldeira, Kenneth G. (Livermore, CA)

2010-02-02T23:59:59.000Z

436

Method for extracting and sequestering carbon dioxide  

DOE Patents [OSTI]

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.

Rau, Gregory H. (Castro Valley, CA); Caldeira, Kenneth G. (Livermore, CA)

2005-05-10T23:59:59.000Z

437

Capture of carbon dioxide by hybrid sorption  

DOE Patents [OSTI]

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.

Srinivasachar, Srivats

2014-09-23T23:59:59.000Z

438

A Vortex Contactor for Carbon Dioxide Separations  

SciTech Connect (OSTI)

Many analysts identify carbon dioxide (CO2) capture and separation as a major roadblock in efforts to cost effectively mitigate greenhouse gas emissions via sequestration. An assessment 4 conducted by the International Energy Agency (IEA) Greenhouse Gas Research and Development Programme cited separation costs from $35 to $264 per tonne of CO2 avoided for a conventional coal fired power plant utilizing existing capture technologies. Because these costs equate to a greater than 40% increase in current power generation rates, it appears obvious that a significant improvement in CO2 separation technology is required if a negative impact on the world economy is to be avoided.

Raterman, Kevin Thomas; Mc Kellar, Michael George; Turner, Terry Donald; Podgorney, Anna Kristine; Stacey, Douglas Edwin; Stokes, B.; Vranicar, J.

2001-05-01T23:59:59.000Z

439

Capture of Carbon Dioxide Archived Projects  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation Proposed New Substation Sites Proposed Route BTRICGEGR-N-Capture of Carbon Dioxide Archived

440

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

SciTech Connect (OSTI)

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.

NONE

1995-07-05T23:59:59.000Z

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

Selection and Characterization of Carbon Black and Surfactants for Development of Small Scale Uranium Oxicarbide Kernels  

SciTech Connect (OSTI)

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.

Contescu, Cristian I [ORNL

2006-01-01T23:59:59.000Z

442

Geodatabase of the South Texas Uranium District  

E-Print Network [OSTI]

Uranium and its associated trace elements and radionuclides are ubiquitous in the South Texas Tertiary environment. Surface mining of this resource from the 1960s through the early 1980s at over sixty locations has left an extensive anthropological footprint (Fig. 1) in the lower Nueces and San Antonio river basins. Reclamation of mining initiated after 1975 has been under the regulatory authority of the Railroad Commission of Texas (RCT). However, mines that were active before the Texas Surface Mining Act of 1975 was enacted, and never reclaimed, are now considered abandoned. The Abandoned Mine Land Section of the RCT is currently reclaiming these pre-regulation uranium mines with funding from the federal government. The RCT monitors the overall effectiveness of this process through post-reclamation radiation and vegetative cover surveys, water quality testing, slope stability and erosion control monitoring. Presently a number of graduate and postgraduate students are completing research on the watershed and reservoir distribution of trace elements and radionuclides downstream of the South Texas Uranium District. The question remains as to whether the elevated levels of uranium, its associated trace elements and radiation levels in the South Texas environment are due to mining

Mark Beaman; William Wade Mcgee

443

The Quest for the Heaviest Uranium Isotope  

E-Print Network [OSTI]

We study Uranium isotopes and surrounding elements at very large neutron number excess. Relativistic mean field and Skyrme-type approaches with different parametrizations are used in the study. Most models show clear indications for isotopes that are stable with respect to neutron emission far beyond N=184 up to the range of around N=258.

S. Schramm; D. Gridnev; D. V. Tarasov; V. N. Tarasov; W. Greiner

2012-01-17T23:59:59.000Z

444

The multiphoton ionization of uranium hexafluoride  

SciTech Connect (OSTI)

Multiphoton ionization (MPI) time-of-flight mass spectroscopy and photoelectron spectroscopy studies of UF{sub 6} have been conducted using focused light from the Nd:YAG laser fundamental ({lambda}=1064 nm) and its harmonics ({lambda}=532, 355, or 266 nm), as well as other wavelengths provided by a tunable dye laser. The MPI mass spectra are dominated by the singly and multiply charged uranium ions rather than by the UF{sub x}{sup +} fragment ions even at the lowest laser power densities at which signal could be detected. The laser power dependence of U{sup n+} ions signals indicates that saturation can occur for many of the steps required for their ionization. In general, the doubly-charged uranium ion (U{sup 2+}) intensity is much greater than that of the singly-charged uranium ion (U{sup +}). For the case of the tunable dye laser experiments, the U{sup n+} (n = 1- 4) wavelength dependence is relatively unstructured and does not show observable resonance enhancement at known atomic uranium excitation wavelengths. The dominance of the U{sup 2+} ion and the absence or very small intensities of UF{sub x}{sup +} fragments, along with the unsaturated wavelength dependence, indicate that mechanisms may exist other than ionization of bare U atoms after the stepwise photodissociation of F atoms from the parent molecule.

Armstrong, D.P. (Oak Ridge K-25 Site, TN (United States). UEO Enrichment Technical Operations Div.)

1992-05-01T23:59:59.000Z

445

Radiological health aspects of uranium milling  

SciTech Connect (OSTI)

This report describes the operation of conventional and unconventional uranium milling processes, the potential for occupational exposure to ionizing radiation at the mill, methods for radiological safety, methods of evaluating occupational radiation exposures, and current government regulations for protecting workers and ensuring that standards for radiation protection are adhered to. In addition, a survey of current radiological health practices is summarized.

Fisher, D.R.; Stoetzel, G.A.

1983-05-01T23:59:59.000Z

446

Investigation of Trace Uranium in Biological Matrices  

E-Print Network [OSTI]

complex. As a result, the data varies in its breadth and quality due to the variety of sources.[41-44] Additional studies have been undertaken to understand the effects of using depleted uranium munitions in war and the accompanying exposures.[45...

Miller, James Christopher

2013-05-31T23:59:59.000Z

447

Faraday rotation spectroscopy of nitrogen dioxide based on a widely tunable external cavity quantum cascade laser  

E-Print Network [OSTI]

Faraday rotation spectroscopy of nitrogen dioxide based on a widely tunable external cavity quantum: Faraday Rotation Spectroscopy, external-cavity quantum cascade laser, nitrogen dioxide, trace

448

ORNL/CDIAC-143 CARBON DIOXIDE, HYDROGRAPHIC, AND CHEMICAL DATA OBTAINED DURING THE  

E-Print Network [OSTI]

Kozyr Carbon Dioxide Information Analysis Center Oak Ridge National Laboratory Oak Ridge, Tennessee, U Prepared by the Carbon Dioxide Information Analysis Center OAK RIDGE NATIONAL LABORATORY Oak Ridge

449

E-Print Network 3.0 - applied carbon dioxide Sample Search Results  

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

7 By-Products Utilization Summary: Center for By-Products Utilization DRAFT REPORT CARBON DIOXIDE SEQUESTRATION IN CEMENTITIOUS... -MILWAUKEE 12;CARBON DIOXIDE...

450

E-Print Network 3.0 - american carbon dioxide Sample Search Results  

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

7 By-Products Utilization Summary: Center for By-Products Utilization DRAFT REPORT CARBON DIOXIDE SEQUESTRATION IN CEMENTITIOUS... -MILWAUKEE 12;CARBON DIOXIDE...

451

E-Print Network 3.0 - ammonia-water-carbon dioxide mixtures Sample...  

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

Summary: . The possibility of using carbonation process as a direct means for carbon dioxide sequestration is yet... . Carbon dioxide gas is the principal greenhouse...

452

E-Print Network 3.0 - air carbon dioxide Sample Search Results  

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

Summary: . The possibility of using carbonation process as a direct means for carbon dioxide sequestration is yet... . Carbon dioxide gas is the principal greenhouse...

453

ORNL/CDIAC-34 Carbon Dioxide Information Analysis Center and  

E-Print Network [OSTI]

Research U.S. Department of Energy Budget Activity Number KP 12 04 01 0 Prepared by the Carbon Dioxide. Burtis Carbon Dioxide Information Analysis Center Environmental Sciences Division Publication No. 4777's (DOE) Environmental Sciences Division, Office of Biological and Environmental Research (OBER

454

World Energy Consumption and Carbon Dioxide Emissions: 1950 2050  

E-Print Network [OSTI]

-U" relation with a within- sample peak between carbon dioxide emissions (and energy use) per capita and perWorld Energy Consumption and Carbon Dioxide Emissions: 1950 Ń 2050 Richard Schmalensee, Thomas M capita income. Using the income and population growth assumptions of the Intergovernmental Panel

455

Method for synthesis of titanium dioxide nanotubes using ionic liquids  

SciTech Connect (OSTI)

The invention is directed to a method for producing titanium dioxide nanotubes, the method comprising anodizing titanium metal in contact with an electrolytic medium containing an ionic liquid. The invention is also directed to the resulting titanium dioxide nanotubes, as well as devices incorporating the nanotubes, such as photovoltaic devices, hydrogen generation devices, and hydrogen detection devices.

Qu, Jun; Luo, Huimin; Dai, Sheng

2013-11-19T23:59:59.000Z

456

Pilot Plant Study of Carbon Dioxide Capture by Aqueous Monoethanolamine  

E-Print Network [OSTI]

i Pilot Plant Study of Carbon Dioxide Capture by Aqueous Monoethanolamine Topical Report Prepared Pilot Plant Study of Carbon Dioxide Capture by Aqueous Monoethanolamine Ross Edward Dugas, M capture using monoethanolamine (MEA). MEA is an appropriate choice for a baseline study since

Rochelle, Gary T.

457

Carbon Dioxide Capture by Chemical Absorption: A Solvent Comparison Study  

E-Print Network [OSTI]

1 Carbon Dioxide Capture by Chemical Absorption: A Solvent Comparison Study by Anusha Kothandaraman Students #12;2 #12;3 Carbon Dioxide Capture by Chemical Absorption: A Solvent Comparison Study by Anusha with electricity generation accounting for 40% of the total1 . Carbon capture and sequestration (CCS) is one

458

Carbon Dioxide Capture DOI: 10.1002/anie.200902836  

E-Print Network [OSTI]

Carbon Dioxide Capture DOI: 10.1002/anie.200902836 Highly Selective CO2 Capture in Flexible 3D Coordination Polymer Networks** Hye-Sun Choi and Myunghyun Paik Suh* Carbon dioxide capture has been warming, and the development of efficient methods for capturing CO2 from industrial flue gas has become

Paik Suh, Myunghyun

459

The surface science of titanium dioxide Ulrike Diebold*  

E-Print Network [OSTI]

The surface science of titanium dioxide Ulrike Diebold* Department of Physics, Tulane University, New Orleans, LA 70118, USA Manuscript received in final form 7 October 2002 Abstract Titanium dioxide is reviewed on the adsorption and reaction of a wide variety of inorganic molecules (H2, O2, H2O, CO, CO2, N2

Diebold, Ulrike

460

Carbon Dioxide, Global Warming, and Michael Crichton's "State of Fear"  

E-Print Network [OSTI]

Carbon Dioxide, Global Warming, and Michael Crichton's "State of Fear" Bert W. Rust Mathematical- tioned the connection between global warming and increasing atmospheric carbon dioxide by pointing out of these plots to global warming have spilled over to the real world, inviting both praise [4, 17] and scorn [15

Rust, Bert W.

Note: This page contains sample records for the topic "uranium dioxide uo" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


461

Exhaust Gas Sensor Based On Tin Dioxide For Automotive Application  

E-Print Network [OSTI]

Exhaust Gas Sensor Based On Tin Dioxide For Automotive Application Arthur VALLERON a,b , Christophe, Engineering Materials Department The aim of this paper is to investigate the potentialities of gas sensor based on semi-conductor for exhaust gas automotive application. The sensing element is a tin dioxide

Paris-Sud XI, Université de

462

Carbon dioxide sequestration in concrete in different curing environments  

E-Print Network [OSTI]

Carbon dioxide sequestration in concrete in different curing environments Y.-m. Chun, T.R. Naik, USA ABSTRACT: This paper summarizes the results of an investigation on carbon dioxide (CO2) sequestration in concrete. Concrete mixtures were not air entrained. Concrete mixtures were made containing

Wisconsin-Milwaukee, University of

463

Absorption of Carbon Dioxide in Aqueous Piperazine/Methyldiethanolamine  

E-Print Network [OSTI]

Absorption of Carbon Dioxide in Aqueous Piperazine/Methyldiethanolamine Sanjay Bishnoi and Gary T dioxide absorption in 0.6 M piperazine PZ r4 M methyldiethanolamine ( )MDEA was measured in a wetted wall loading. The absorption rate did not follow pseudo first-order beha®ior except at ®ery low loading. All

Rochelle, Gary T.

464

Development of a Carbon Dioxide Monitoring Rotorcraft Unmanned Aerial Vehicle  

E-Print Network [OSTI]

stage to prevent potential danger to workforce and material, and carbon capture and sequestration (CCSDevelopment of a Carbon Dioxide Monitoring Rotorcraft Unmanned Aerial Vehicle Florian Poppa and Uwe the development of a carbon dioxide (CO2) sensing rotorcraft unmanned aerial vehicle (RUAV) and the experiences

Zimmer, Uwe

465