National Library of Energy BETA

Sample records for uo uo uo

  1. Spectroscopic Studies of the Several Isomers of UO3

    SciTech Connect (OSTI)

    Sweet, Lucas E.; Reilly, Dallas D.; Abrecht, David G.; Buck, Edgar C.; Meier, David E.; Su, Yin-Fong; Brauer, Carolyn S.; Schwantes, Jon M.; Tonkyn, Russell G.; Szecsody, James E.; Blake, Thomas A.; Johnson, Timothy J.

    2013-09-26

    Uranium trioxide is known to adopt seven different structural forms. While these structural forms have been well characterized using x-ray or neutron diffraction techniques, little work has been done to characterize their spectroscopic properties, particularly of the pure phases. Since the structural isomers of UO3 all have similar thermodynamic stabilities and most tend to hydrolyze under open atmospheric conditions, mixtures of UO3 phases and the hydrolysis products are common. Much effort went into isolating pure phases of UO3. Utilizing x-ray diffraction as a sample identification check, UV/Vis/NIR spectroscopic signatures of α-UO3, β-UO3, γ-UO3 and UO2(OH)2 products were obtained. The spectra of the pure phases can now be used to characterize typical samples of UO3, which are often mixtures of isomers.

  2. METHOD FOR PREPARATION OF UO$sub 2$ PARTICLES

    DOE Patents [OSTI]

    Johnson, J.R.; Taylor, A.J.

    1959-09-22

    A method is described for the preparation of highdensity UO/sub 2/ particles within the size range of 40 to 100 microns. In accordance with the invention UO/sub 2/ particles are autoclaved with an aqueous solution of uranyl ions. The resulting crystals are reduced to UO/sub 2/ and the UO/sub 2/ is heated to at least 1000 deg C to effect densification. The resulting UO/sub 2/ particles are screened, and oversize particles are crushed and screened to recover the particles within the desired size range.

  3. Final Report: Manganese Redox Mediation of UO2 Stability and...

    Office of Scientific and Technical Information (OSTI)

    Meter Scale Dynamics Citation Details In-Document Search Title: Final Report: Manganese Redox Mediation of UO2 Stability and Uranium Fate in the Subsurface: Molecular and Meter ...

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

    Office of Scientific and Technical Information (OSTI)

    Journal Article: Oxidative Dissolution of UO2 in a Simulated ... Citation Details In-Document Search Title: Oxidative ... Publication Date: 2014-03-18 OSTI Identifier: 1124154 Report ...

  5. RECOMMENDATIONS FOR UO3 PLANT BIOASSAY

    SciTech Connect (OSTI)

    Carbaugh, Eugene H.

    2010-07-12

    Alternative urine bioassay programs are described for application with decontamination and decommissioning activities at the Hanford UO3 Plant. The alternatives are based on quarterly or monthly urine bioassay for recycled uranium, assuming multiple acute inhalation intakes of recycled uranium occurring over a year. The inhalations are assumed to be 5µm AMAD particles of 80% absorption type F and 20% absorption type M. Screening levels, expressed as daily uranium mass excretion rates in urine, and the actions associated with these levels are provided for both quarterly and monthly sampling frequencies.

  6. PREPARATION OF HIGH DENSITY UO$sub 2$

    DOE Patents [OSTI]

    Googin, J.M.

    1959-09-29

    A method is presented for the preparation of highdensity UO/sub 2/ from UF/sub 6/. In accordance with the invention, UF/sub 6/ is reacted with water and concentrated ammonium hydroxide is added to the resulting aqueous solution of UO/ sub 2/F/sub 2/. The resulting precipitate is calcined to U/sub 3/O/sub 8/ an d the U/sub 3/O/sub 8/ is reduced to UO/sub 2/ with a gaseous mixture comprised of carbon monoxide and carbon dioxide at a temperature of from 1600 to 1900 deg C.

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

    SciTech Connect (OSTI)

    Burgett, Eric; Deo, Chaitanya; Phillpot, Simon

    2015-05-08

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

  8. Simple but Stronger UO, Double but Weaker UNMe Bonds: The Tale Told by Cp2UO and Cp2UNR

    SciTech Connect (OSTI)

    LPCNO, CNRS-UPS-INSA, INSA Toulouse; Institut Charles Gerhardt, CNRS, Universite Montpellier; Laboratoire de Chimie et Physique Quantiques, CNRS, IRSAMC, Universite Paul Sabatier; Andersen, Richard; Barros, Noemi; Maynau, Daniel; Maron, Laurent; Eisenstein, Odile; Zi, Guofu; Andersen, Richard

    2007-06-27

    The free energies of reaction and the activation energies are calculated, with DFT (B3PW91) and small RECP (relativistic core potential) for uranium, for the reaction of Cp2UNMe and Cp2UO with MeCCMe and H3Si-Cl that yields the corresponding addition products. CAS(2,7) and DFT calculations on Cp2UO and Cp2UNMe give similar results, which validates the use of DFT calculations in these cases. The calculated results mirror the experimental reaction of [1,2,4-(CMe3)3C5H2]2UNMe with dimethylacetylene and [1,2,4-(CMe3)3C5H2]2UO with Me3SiCl. The net reactions are controlled by the change in free energy between the products and reactants, not by the activation energies, and therefore by the nature of the UO and UNMe bonds in the initial and final states. A NBO analysis indicates that the U-O interaction in Cp2UO is composed of a single U-O bond with three lone pairs of electrons localized on oxygen, leading to a polarized U-O fragment. In contrast, the U-NMe interaction in Cp2UNMe is composed of a and component and a lone pairof electrons localized on the nitrogen, resulting in a less polarized UNMe fragment, in accord with the lower electronegativity of NMe relative to O. The strongly polarized U(+)-O(-) bond is calculated to be about 70 kcal mol-1 stronger than the less polarized U=NMe bond.

  9. U(v) in metal uranates: A combined experimental and theoretical study of MgUO4, CrUO4, and FeUO4

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

    Guo, Xiaofeng; Tiferet, Eitan; Qi, Liang; Solomon, Jonathan M.; Lanzirotti, Antonio; Newville, Matthew; Engelhard, Mark H.; Kukkadapu, Ravi K.; Wu, Di; Ilton, Eugene S.; et al

    2016-02-02

    Although pentavalent uranium can exist in aqueous solution, its presence in the solid state is uncommon. Metal monouranates, MgUO4, CrUO4 and FeUO4 were synthesized for detailed structural and energetic investigations. Structural characteristics of these uranates used powder X-ray diffraction, synchrotron X-ray absorption spectroscopy, X-ray photoelectron spectroscopy, and 57Fe-Mossbauer spectroscopy. Enthalpies of formation were measured by high temperature oxide melt solution calorimetry. Density functional theory (DFT) calculations provided both structural and energetic information. The measured structural and thermodynamic properties show good consistency with those predicted from DFT. The presence of U5+ has been solidly confirmed in CrUO4 and FeUO4, which aremore » thermodynamically stable compounds, and the origin and stability of U5+ in the system was elaborated by DFT. Lastly, the structural and thermodynamic behaviour of U5+ elucidated in this work is relevant to fundamental actinide redox chemistry and to applications in the nuclear industry and radioactive waste disposal.« less

  10. Unveiling the Behavior of UO2 Under Extreme Physical Conditions...

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

    Such understanding, as the interplay between UO2 and its zirconium cladding at the atomic scale, will allow researchers to design, test and deploy "accident-tolerant fuels" that do ...

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

    SciTech Connect (OSTI)

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

    1994-02-01

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

  12. Synthesis and sintering of UN-UO2 fuel composites

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

    Jaques, Brian J.; Watkins, Jennifer; Croteau, Joseph R.; Alanko, Gordon A.; Tyburska-Puschel, Beata; Meyer, Mitch; Xu, Peng; Lahoda, Edward J.; Butt, Darryl P.

    2015-06-17

    In this study, the design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO2, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO2 to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uraniummore » using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO2 in a planetary ball mill. UN and UN – UO2 composite pellets were sintered in Ar – (0–1 at%) N2 to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO2 composite pellets were also sintered in Ar – 100 ppm N2 to assess the effects of temperature (1700–2000 °C) on the final grain morphology and phase concentration.« less

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

    SciTech Connect (OSTI)

    Washenfelder, D.J.

    1993-12-01

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

  14. Thermodynamic studies of studtite thermal decomposition pathways via amorphous intermediates UO3, U2O7, and UO4

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

    Guo, Xiaofeng; Wu, Di; Xu, Hongwu; Burns, Peter C.; Navrotsky, Alexandra

    2016-06-08

    The thermal decomposition of studtite (UO2)O2(H2O)2·2H2O results in a series of intermediate X-ray amorphous materials with general composition UO3+x (x = 0, 0.5, 1). As an extension of a structural study on U2O7, this work provides detailed calorimetric data on these amorphous oxygen-rich materials since their energetics and thermal stability are unknown. These were characterized in situ by thermogravimetry, and mass spectrometry. Ex situ X-ray diffraction and infrared spectroscopy characterized their chemical bonding and local structures. This detailed characterization formed the basis for obtaining formation enthalpies by high temperature oxide melt solution calorimetry. The thermodynamic data demonstrate the metastability ofmore » the amorphous UO3+x materials, and explain their irreversible and spontaneous reactions to generate oxygen and form metaschoepite. Thus, formation of studtite in the nuclear fuel cycle, followed by heat treatment, can produce metastable amorphous UO3+x materials that pose the risk of significant O2 gas. Quantitative knowledge of the energy landscape of amorphous UO3+x was provided for stability analysis and assessment of conditions for decomposition.« less

  15. Modelling oxygen self-diffusion in UO2 under pressure

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

    Cooper, Michael William D.; Grimes, R. W.; Fitzpatrick, M. E.; Chroneos, A.

    2015-10-22

    Access to values for oxygen self-diffusion over a range of temperatures and pressures in UO2 is important to nuclear fuel applications. Here, elastic and expansivity data are used in the framework of a thermodynamic model, the cBΩ model, to derive the oxygen self-diffusion coefficient in UO2 over a range of pressures (0–10 GPa) and temperatures (300–1900 K). Furthermore, the significant reduction in oxygen self-diffusion as a function of increasing hydrostatic pressure, and the associated increase in activation energy, is identified.

  16. METHOD FOR PREPARATION OF SPHERICAL UO$sub 4$

    DOE Patents [OSTI]

    Gregory, J.F. Jr.; Levey, R.P. Jr.

    1962-06-01

    A method is given for continuously precipitating ura nium peroxide in the form of spherical particles. Seed crystals are formed in a first reaction zone by introducing an acidified aqueous uranyl nitrate solution and an aqueous hydrogen peroxide solution at a ratio of 5 to 20 per cent of the stoichiometric amount required for complete precipitation. After a mean residence time of 2 to 5 minutes in the first reaction zone, the resulting mixture is introduced into a second reaction zone, together with a large excess of hydrogen peroxide solution. The resulting UO4 is rapidly separated from the mother liquor after an over-all residence time of 5 to 11 minutes. The first reaction is maintained at a temperature of 85 to 90 deg C and the second zone above 50 deg C. Additional reaction zones may be employed for further crystal growth. The UO/sub 4/ is converted to U/sub 3/O/sub 8/ or UO/sub 2/ by heating in air or hydrogen atmosphere. This method is particularly useful for the preparation of spherical UO/sub 2/ particles 10 to 25 microns in diameter. (AEC)

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

    SciTech Connect (OSTI)

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

    2009-11-01

    he objective of this paper is to demonstrate that NF3 fluorinates uranium metal, UO2, UF4, UO3, U3O8, and UO2F22H2O 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.

  18. Density Functional Theory Calculations of Mass Transport in UO2

    SciTech Connect (OSTI)

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

    2012-06-26

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

  19. Tandem dissolution of UO 3 in amide-based acidic ionic liquid and in situ electrodeposition of UO 2 with regeneration of the ionic liquid: a closed cycle

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

    Wanigasekara, Eranda; Freiderich, John W.; Sun, Xiao-Guang; Meisner, Roberta A.; Luo, Huimin; Delmau, Lætitia H.; Dai, Sheng; Moyer, Bruce A.

    2016-05-19

    A closed cycle is demonstrated for the tandem dissolution and electroreduction of UO3 to UO2 with regeneration of the acidic ionic liquid. The dissolution is achieved by use of the acidic ionic liquid N,N-dimethylacetimidium bis(trifluoromethanesulfonimide) in 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonimide) serving as the diluent. Bulk electrolysis performed at 1.0 V vs. Ag reference yields a dark brown-black uranium deposit (UO2) on the cathode. Anodic oxidation of water in the presence of dimethylacetamide regenerates the acidic ionic liquid. We have demonstrated the individual steps in the cycle together with a sequential dissolution, electroreduction, and regeneration cycle.

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

    SciTech Connect (OSTI)

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

    2015-01-01

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

  1. PROCESS FOR THE PRODUCTION OF AN ACTIVATED FORM OF UO$sub 2$

    DOE Patents [OSTI]

    Polissar, M.J.

    1957-09-24

    A process for producing a highly active form of UO/sub 2/ characterized both by rapid oxidation in air and by rapid chlorination with CCl/sub 4/ vapor at an elevated temperature is reported. In accordance with the process, commercial UO/sub 2/, is subjected to a series of oxidation-reduction operations to produce a form of UC/sub 2/ of enhanced reactivity. By treatimg commercial UO/sub 2/ at a temperature between 335 and 485 deg C with methane, then briefly with an oxygen containing gas and followimg this by a second treatment with a methane containing gas, the original relatively stable charge of UO/sub 2/ will be transformed into an active form of UO/sub 2/.

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

    SciTech Connect (OSTI)

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

    1995-01-01

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

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

    SciTech Connect (OSTI)

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

    2015-10-08

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

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

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

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

    2015-10-08

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

  5. Dissolution process for ZrO.sub.2 -UO.sub.2 -CaO fuels

    DOE Patents [OSTI]

    Paige, Bernice E.

    1976-06-22

    The present invention provides an improved dissolution process for ZrO.sub.2 -UO.sub.2 -CaO-type pressurized water reactor fuels. The zirconium cladding is dissolved with hydrofluoric acid, immersing the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers in the resulting zirconium-dissolver-product in the dissolver vessel, and nitric acid is added to the dissolver vessel to facilitate dissolution of the uranium from the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers.

  6. Simulations and Experimental Measurements of UO2 Thermal Conductivity

    SciTech Connect (OSTI)

    Stanek, Christopher Richard; Gofryk, Krzysztof; Tonks, Michael; Andersson, Anders David Ragnar; Liu, Xiang-Yang; Lashley, Jason Charles; Uberuaga, Blas P.; Mcclellan, Kenneth James

    2015-04-10

    Spin-phonon interactions lead to low κ of UO2 (and behave like a defect), and this has implications for nuclear fuel performace. The inability to capture spin-phonon scattering leads to inherent errors. The interplay between magnetism and structural asymmetry in UO2 displays rich physics. Grain boundary structure plays a role which must be taken into account.

  7. Final Report: Manganese Redox Mediation of UO2 Stability and Uranium Fate

    Office of Scientific and Technical Information (OSTI)

    in the Subsurface: Molecular and Meter Scale Dynamics (Technical Report) | SciTech Connect Report: Manganese Redox Mediation of UO2 Stability and Uranium Fate in the Subsurface: Molecular and Meter Scale Dynamics Citation Details In-Document Search Title: Final Report: Manganese Redox Mediation of UO2 Stability and Uranium Fate in the Subsurface: Molecular and Meter Scale Dynamics One strategy to remediate U contamination in the subsurface is the immobilization of U via injection of an

  8. DISPERSION ELEMENT CONSISTING OF CHROMIUM COATED UO$sup 2$ PARTICLES UNIFORMLY DISTRIBUTED IN A ZIRCALOY MATRIX

    DOE Patents [OSTI]

    Cain, F.M. Jr.; Eck, J.E.

    1963-05-01

    A nuclear fuel element consisting of metal coated UO/sub 2/ particles dispersed in a matrix of Zircalloy and having a cladding of Zircalloy is presented. (AEC)

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

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

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

    2015-08-01

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

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

    SciTech Connect (OSTI)

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

    2015-08-01

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

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

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

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

    2015-02-11

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

  12. Synthesis and sintering of UN-UO2 fuel composites

    SciTech Connect (OSTI)

    Jaques, Brian J.; Watkins, Jennifer; Croteau, Joseph R.; Alanko, Gordon A.; Tyburska-Puschel, Beata; Meyer, Mitch; Xu, Peng; Lahoda, Edward J.; Butt, Darryl P.

    2015-06-17

    In this study, the design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO2, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO2 to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO2 in a planetary ball mill. UN and UN – UO2 composite pellets were sintered in Ar – (0–1 at%) N2 to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO2 composite pellets were also sintered in Ar – 100 ppm N2 to assess the effects of temperature (1700–2000 °C) on the final grain morphology and phase concentration.

  13. PREPARATION OF UO$sub 2$ FOR NUCLEAR REACTOR FUEL PELLETS

    DOE Patents [OSTI]

    Googin, J.M.

    1962-06-01

    A method is given for preparing high-density UO/sub 2/ compacts. An aqueous uranyl fluoride solution is contacted with an aqueous ammonium hydroxide solution at an ammonium to-uranium ratio of 25: 1 to 30:1 to form a precipitate. The precipitate is separated from the- mother liquor, dried, and contacted with steam at a uniform temperature within the range of 400 to 650 deg C to produce U/ sub 3/O/sub 8/. The U/sub 3/O/sub 8/ is red uced to UO/sub 2/ with hydrogen at a uniform temperature within the range of 550 to 600 deg C. The UO/sub 2/ is then compressed into compacts and sintered. High-density compacts are fabricated to close tolerances without use of a binder and without machining or grinding. (AEC)

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

    SciTech Connect (OSTI)

    Skim, Y. S.

    1999-11-12

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

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

    SciTech Connect (OSTI)

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

    2014-07-01

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

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

    DOE Patents [OSTI]

    Moore, J.E.

    1960-07-12

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

  17. PUREX/UO3 Facilities deactivation lessons learned history

    SciTech Connect (OSTI)

    Gerber, M.S.

    1996-09-19

    Disconnecting the criticality alarm permanently in June 1996 signified that the hazards in the PUREX (plutonium-uranium extraction) plant had been so removed and reduced that criticality was no longer a credible event. Turning off the PUREX criticality alarm also marked a salient point in a historic deactivation project, 1 year before its anticipated conclusion. The PUREX/UO3 Deactivation Project began in October 1993 as a 5-year, $222.5- million project. As a result of innovations implemented during 1994 and 1995, the project schedule was shortened by over a year, with concomitant savings. In 1994, the innovations included arranging to send contaminated nitric acid from the PUREX Plant to British Nuclear Fuels, Limited (BNFL) for reuse and sending metal solutions containing plutonium and uranium from PUREX to the Hanford Site tank farms. These two steps saved the project $36.9- million. In 1995, reductions in overhead rate, work scope, and budget, along with curtailed capital equipment expenditures, reduced the cost another $25.6 million. These savings were achieved by using activity-based cost estimating and applying technical schedule enhancements. In 1996, a series of changes brought about under the general concept of ``reengineering`` reduced the cost approximately another $15 million, and moved the completion date to May 1997. With the total savings projected at about $75 million, or 33.7 percent of the originally projected cost, understanding how the changes came about, what decisions were made, and why they were made becomes important. At the same time sweeping changes in the cultural of the Hanford Site were taking place. These changes included shifting employee relations and work structures, introducing new philosophies and methods in maintaining safety and complying with regulations, using electronic technology to manage information, and, adopting new methods and bases for evaluating progress. Because these changes helped generate cost savings and were

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

    SciTech Connect (OSTI)

    Thein, S.M.

    2000-07-26

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

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

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

    SciTech Connect (OSTI)

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

    2012-05-02

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

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

    SciTech Connect (OSTI)

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

    2015-10-14

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

  2. Fabrication of ThO/sub 2/ and ThO/sub 2/-UO/sub 2/ fuel pellets

    SciTech Connect (OSTI)

    Davis, N.C.; Matthews, R.B.; White, G.D.; Hart, P.E.

    1980-06-01

    In this presentation, ThO/sub 2/ and ThO/sub 2/-UO/sub 2/ pellet fuel development activities leading to the production of kilogram quantities of fuel are described. Conventional dry ball milling was used to produce ThO/sub 2/ and ThO/sub 2/-UO/sub 2/ mixtures that were pressed and sintered to 95% TD with a homogeneous distribution of the components.

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

    SciTech Connect (OSTI)

    Wittman, Richard S.; Buck, Edgar C.

    2012-09-01

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

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

    DOE Patents [OSTI]

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

    1975-12-01

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

  5. Development of Innovative Accident Tolerant High Thermal Conductivity UO2-Diamond Composite Fuel Pellets

    SciTech Connect (OSTI)

    Tulenko, James; Subhash, Ghatu

    2016-01-01

    The University of Florida (UF) evaluated a composite fuel consisting of UO2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO2 – SiC and UO2 – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO2 may greatly enhanced the thermal conductivity of the UO2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

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

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1994-10-18

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

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

    SciTech Connect (OSTI)

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

    2015-07-21

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

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

    DOE Patents [OSTI]

    Tomczuk, Zygmunt; Miller, William E.

    1994-01-01

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

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

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

    SciTech Connect (OSTI)

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

    2015-02-11

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

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

    SciTech Connect (OSTI)

    Thompson, Alexander E.; Wolverton, C.

    2011-10-01

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

  12. Pellet fabrication development using thermally denitrated UO{sub 2} powder

    SciTech Connect (OSTI)

    Davis, N.C.; Griffin, C.W.

    1992-05-01

    Pacific Northwest Laboratory (PNL) has evaluted, on a laboratory scale, the characteristics and pellet fabrication properties of UO{sub 3} powder prepared by the thermal denitration process. Excellent quality, 96% TD (percent of theoretical density) pellets were produced from development lots of this powder. Apparently, the key to making this highly sinterable powder from uranyl nitrate is the addition of ammonium nitrate (NH{sub 4}NO{sub 3}) to the feed solution prior to thermal denitration. Powder lots were processed with and without the NH{sub 4}NO{sub 3} addition in the feed solution. The lots included samples from the ORNL laboratory rotary kiln and from a larger scale rotary kiln at National Lead of Ohio (NLO). In the PNL evaluation, samples of UO{sub 3} were calculated and reduced to UO{sub 2}, followed by conventional process procedures to compare the sinterability of the powder lots. The high density pellets made from the powder lots, which included the NH{sub 4}NO{sub 3} addition, were reduced to Fast Breeder Reactor (FBR) density range of 88 to 92% TD by the use of poreformers. The NH{sub 4}NO{sub 3} addition also improved the sinterability properties of uranium oxide powders that contain thorium and cerium. Thorium and cerium were used as ``stand-in`` for plutonium used in urania-plutonia FBR fuel pellets. A very preliminary examination of a single lot of thermally denitrated uranium-plutonium oxide powder was made. This powder lot was made with the NH{sub 3}NO{sub 3} addition and produced pellets just above the FBR density range.

  13. 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)

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

    SciTech Connect (OSTI)

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

    2007-06-01

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

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

    SciTech Connect (OSTI)

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

    2015-09-30

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

  16. Synchrotron characterization of nanograined UO2 grain growth

    SciTech Connect (OSTI)

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

    2015-09-30

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

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

    SciTech Connect (OSTI)

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

    1980-11-01

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

  18. Direct electrolytic reduction of UO{sub 2} vs. U{sub 3}O{sub 8}

    SciTech Connect (OSTI)

    Barnes, L.A.; Willit, J.L.

    2007-07-01

    UO{sub 2} and U{sub 3}O{sub 8} have been electrochemically reduced to uranium metal by direct electrolytic reduction at 650 deg. C in molten LiCl-Li{sub 2}O. Differences in electrolyte concentrations and efficiency have been observed in the reduction process as a function of using UO{sub 2} vs. U{sub 3}O{sub 8} as the feed material. Numerous tests with UO{sub 2} as the feed material have been conducted to optimize the electrochemical cell design and process chemistry. These studies have reproducibly demonstrated greater than 99% reduction with current efficiencies as high as 75% and only a slight decrease in the Li{sub 2}O concentration. Initial scoping experiments using U{sub 3}O{sub 8} have achieved greater than 93% reduction to uranium metal. However, in the U{sub 3}O{sub 8} experiments, the Li{sub 2}O concentration dropped significantly upon introducing the feed basket into the cell electrolyte due to formation of Li{sub 2}UO{sub 4}., but under appropriate conditions could be fully reduced to metal. (authors)

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

    SciTech Connect (OSTI)

    Michael R Tonks; Yongfeng Zhang; Xianming Bai

    2014-06-01

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

  20. Multi-scale modeling of inter-granular fracture in UO2

    SciTech Connect (OSTI)

    Chakraborty, Pritam; Zhang, Yongfeng; Tonks, Michael R.; Biner, S. Bulent

    2015-03-01

    A hierarchical multi-scale approach is pursued in this work to investigate the influence of porosity, pore and grain size on the intergranular brittle fracture in UO2. In this approach, molecular dynamics simulations are performed to obtain the fracture properties for different grain boundary types. A phase-field model is then utilized to perform intergranular fracture simulations of representative microstructures with different porosities, pore and grain sizes. In these simulations the grain boundary fracture properties obtained from molecular dynamics simulations are used. The responses from the phase-field fracture simulations are then fitted with a stress-based brittle fracture model usable at the engineering scale. This approach encapsulates three different length and time scales, and allows the development of microstructurally informed engineering scale model from properties evaluated at the atomistic scale.

  1. New insight into UO2F2 particulate structure by micro-Raman spectroscopy

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

    Stefaniak, Elzbieta A.; Darchuk, Larysa; Sapundjiev, Danislav; Kips, Ruth E.; Aregbe, Yetunde; Grieken, Rene Van

    2013-02-19

    Uranyl fluoride particles produced via hydrolysis of uranium hexafluoride have been deposited on different substrates: polished graphite disks, silver foil, stainless steel and gold-coated silicon wafer, and measured with micro-Raman spectroscopy (MRS). All three metallic substrates enhanced the Raman signal delivered by UO2F2 in comparison to graphite. The fundamental stretching of the U–O band appeared at 867 cm–1 in case of the graphite substrate, while in case of the others it was shifted to lower frequencies (down to 839 cm–1). All applied metallic substrates showed the expected effect of Raman signal enhancement; however the gold layer appeared to be mostmore » effective. Lastly, application of new substrates provides more information on the molecular structure of uranyl fluoride precipitation, which is interesting for nuclear safeguards and nuclear environmental analysis.« less

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

    SciTech Connect (OSTI)

    Kosaka, Yuji; Watanabe, Seiichi; Arakawa, Yasushi

    2007-07-01

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

  3. Direct Electrodeposition of UO2 from Uranyl Bis(trifluoromethanesulfonyl)imide Dissolved in 1-Ethyl-3-methylimidazolium Bis(trifluoromethanesulfonyl)imide Room Temperature Ionic Liquid System

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

    Freiderich, John W.; Wanigasekara, Eranda P.; Sun, Xiao-Guang; Meisner, Roberta Ann; Meyer, III, Harry M.; Luo, Huimin; Delmau, Laetitia Helene; Dai, Sheng; Moyer, Bruce A

    2013-11-11

    Our study demonstrates a direct electrodeposition of UO2 at a Pt cathode from a solution of uranyl bis(trifluoromethanesulfonyl)imide [UO2(NTf2)2)] in a bulk room-temperature ionic liquid (RTIL), 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide (EMIM+NTf2–). Cyclic voltammetry (CV) studies revealed two reduction waves corresponding to the conversion of uranium(VI) to uranium(IV), and a mechanism for the overall electroreduction is proposed. A controlled-potential experiment was performed, holding the reduction potential at–1.0 V for 24 h to obtain a brown-black deposit of UO2 on the Pt cathode. The Faradaic efficiency of the reduction process was determined to be >80%. The UO2deposit was characterized by powder X-ray diffraction (XRD)more » and X-ray photoelectron spectroscopy (XPS).« less

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

    SciTech Connect (OSTI)

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

    2014-10-31

    In light water reactor fuel, gaseous fission products segregate to grain boundaries, resulting in the nucleation and growth of large intergranular fission gas bubbles. Based on the mechanisms established from density functional theory (DFT) and empirical potential calculations 1, continuum models for diffusion of xenon (Xe), uranium (U) vacancies and U interstitials in UO2 have been derived for both intrinsic conditions and under irradiation. Segregation of Xe to grain boundaries is described by combining the bulk diffusion model with a model for the interaction between Xe atoms and three different grain boundaries in UO2 ( Σ5 tilt, Σ5 twist and a high angle random boundary),as derived from atomistic calculations. All models are implemented in the MARMOT phase field code, which is used to calculate effective Xe and U diffusivities as well as redistribution for a few simple microstructures.

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

    SciTech Connect (OSTI)

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

    2011-04-15

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

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

    SciTech Connect (OSTI)

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

    2014-01-01

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

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

    SciTech Connect (OSTI)

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

    2002-06-15

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

  8. Chemical reactions during ThO{sub 2} and ThO{sub 2}-UO{sub 2} fuel fabrication

    SciTech Connect (OSTI)

    Clayton, J.C.

    1994-08-01

    The chemical reactions that occur during the fabrication of fuel pellets for the Shipping port Light Water Breeder Reactor are discussed. These include (1) precipitation and pyrolysis of thorium oxalate, (2) precipitation, calcination, and hydrogen reduction of ammonium diuranate, (3) comminution, granulation with an organic binder, and cold compaction of ThO{sub 2} and ThO{sub 2}-UO{sub 2} powders, (4) decarburization of the organic binder in CO{sub 2} at temperature up to 925 C, and (5) sintering in moist hydrogen at temperature up to 1790 C. Thorium oxalate precipitation and pyrolysis temperatures were the primary process variables for controlling the resulting thoria powder properties. Coprecipitated metal sulfates were converted to transition metal sulfides during calcining - the conversion of thorium oxalate to thorium oxide. The critical variable for controlling the urania powder properties was the hydrogen reduction temperature. Thermodynamic analyses showed that the efficiency of carbon removal from cold-compacted ThO{sub 2} and ThO{sub 2}-UO{sub 2} pellets by CO{sub 2} oxidation increases with temperature and, at temperatures around 900 C, substantially complete oxidation of carbon to carbon monoxide gas should occur. The carbon content of the ThO{sub 2} and ThO{sub 2}-UO{sub 2} fuels was further reduced during the initial heating in the hydrogen sintering cycle through the formation of methane gas. Additions of water vapor to the hydrogen sintering atmosphere also aided in carbon removal.

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

    SciTech Connect (OSTI)

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

    2014-03-10

    Uranium dioxide (UO2) is the most common fuel material in commercial nuclear power reactors. Despite its numerous advantages such as high melting point, good high-temperature stability, good chemical compatibility with cladding and coolant, and resistance to radiation, it suffers from low thermal conductivity that can result in large temperature gradients within the UO2 fuel pellet, causing it to crack and release fission gases. Thermal swelling of the pellets also limits the lifetime of UO2 fuel in the reactor. To mitigate these problems, we propose to develop novel UO2 fuel with uniformly distributed carbon nanotubes (CNTs) that can provide high-conductivity thermal pathways and can eliminate fuel cracking and fission gas release due to high temperatures. CNTs have been investigated extensively for the past decade to explore their unique physical properties and many potential applications. CNTs have high thermal conductivity (6600 W/mK for an individual single- walled CNT and >3000 W/mK for an individual multi-walled CNT) and high temperature stability up to 2800°C in vacuum and about 750°C in air. These properties make them attractive candidates in preparing nano-composites with new functional properties. The objective of the proposed research is to develop high thermal conductivity of UO2–CNT composites without affecting the neutronic property of UO2 significantly. The concept of this goal is to utilize a rapid sintering method (5–15 min) called spark plasma sintering (SPS) in which a mixture of CNTs and UO2 powder are used to make composites with different volume fractions of CNTs. Incorporation of these nanoscale materials plays a fundamentally critical role in controlling the performance and stability of UO2 fuel. We will use a novel in situ growth process to grow CNTs on UO2 particles for rapid sintering and develop UO2-CNT composites. This method is expected to provide a uniform distribution of CNTs at various volume fractions so that a high

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

    SciTech Connect (OSTI)

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

    2010-12-01

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

  11. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    SciTech Connect (OSTI)

    Bai, Xian -Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2015-12-22

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctly predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. Furthermore, an analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.

  12. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

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

    Bai, Xian -Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2015-12-22

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctlymore » predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. Furthermore, an analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.« less

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

    SciTech Connect (OSTI)

    Van Brutzel, Laurent

    2008-07-01

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

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

    SciTech Connect (OSTI)

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

    2013-10-01

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

  15. Thermionic emission and work function of U and UO/sub 2/

    SciTech Connect (OSTI)

    McLean, W.; Chen, H.L.

    1985-02-01

    Thermionic emission measurements have been used to determine the work function (phi) of pure and oxidized uranium samples between 1100 and 1300/sup 0/K; Auger electron spectroscopy (AES) was used to verify the cleanliness and compositions of the samples. It was found that impurities present in ppM amounts in the bulk U segregated to the surface upon heating and had an appreciable effect on the zero-field emission currents as well as the slopes of the Schottkey curves obtained at various temperatures. A combination of ion-sputtering and ultra-high vacuum (UHV) annealing at high temperatures was successful in reducing the total impurity level on the hot surfaces to approx.5%. At this low concentration of impurities, well-behaved Richardson line plots were obtained with A = 135 A cm/sup -2/ K/sup -2/ and phi = 3.54 eV for pure U, and A = 128 A cm/sup -2/ K/sup -2/ and phi = 3.19 eV for UO/sub 2/. The Schottkey coefficients for clean U approached their ideal values at fields > 400 V/cm.

  16. Probing the Oxygen Environment in UO22+ by Solid-State O-17 Nuclear Magnetic Resonance Spectroscopy and Relativistic Density Functional Calculations

    SciTech Connect (OSTI)

    Cho, Herman M.; De Jong, Wibe A.; Soderquist, Chuck Z.

    2010-02-28

    A combined theoretical and solid-state O-17 NMR study of the electronic structure of the uranyl ion UO22+ in (NH4)4UO2(CO3)3 and rutherfordine UO2CO3 is presented, the former representing a system with a hydrogen-bonding environment around the uranyl oxygens, and the latter exemplifying a uranyl environment without hydrogens. A fully relativistic ab initio treatment reveals unique features of the U-O covalent bond, including the finding of O-17 chemical shift anisotropies that are among the largest ever reported (>1200 ppm). Computational results for the oxygen electric field gradient tensor are found to be consistently larger in magnitude than experimental solid-state O-17 NMR measurements in a 7.05 T magnetic field indicate. A modified version of the Solomon theory of the two-spin echo amplitude for a spin-5/2 nucleus is developed and applied to the analysis of the O-17 echo signal of UO22+. The William R. Wiley environmental Molecular Sciences Laboratory is a US Department of Energy national scientific user facility located at Pacific Northwest National Laboratory (PNNL) in Richland, Washington. PNNL is operated by Battelle for the US Department of Energy.

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

    SciTech Connect (OSTI)

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

    2012-05-30

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

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

    SciTech Connect (OSTI)

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

    2012-10-30

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

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

    Office of Legacy Management (LM)

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

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

    SciTech Connect (OSTI)

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

    1982-01-01

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

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

    SciTech Connect (OSTI)

    Uberuaga, Blas P.

    2012-05-02

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

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

    SciTech Connect (OSTI)

    Collins, J.L.

    2004-12-02

    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.

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

    SciTech Connect (OSTI)

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

    2015-08-28

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

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

    SciTech Connect (OSTI)

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

    2015-09-30

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

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

    SciTech Connect (OSTI)

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

    2014-11-01

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

  6. Theoretical investigation of thermodynamic stability and mobility of the oxygen vacancy in ThO2UO2 solid solutions

    SciTech Connect (OSTI)

    Liu, B.; Aidhy, D. S.; Zhang, Y.; Weber, W. J.

    2014-10-16

    The thermodynamic stability and the migration energy barriers of oxygen vacancies in ThO2UO2 solid solutions are investigated by density functional theory calculations. In pure ThO2, the formation energy of oxygen vacancy is 7.58 eV and 1.46 eV under O rich and O poor conditions, respectively, while its migration energy barrier is 1.97 eV. The addition of UO2 into ThO2 significantly decreases the energetics of formation and migration of the oxygen vacancy. Among the range of UO2-ThO2 solid solutions studied in this work, UO2 exhibits the lowest formation energy (5.99 eV and -0.13 eV under O rich and O poor conditions, respectively) and Th0.25U0.75O2 exhibits the lowest migration energy barrier (~ 1 eV). Moreover, by considering chemical potential, the phase diagram of oxygen vacancy as a function of both temperature and oxygen partial pressure is shown, which could help to gain experimental control over oxygen vacancy concentration.

  7. Theoretical investigation of thermodynamic stability and mobility of the oxygen vacancy in ThO2 –UO2 solid solutions

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

    Liu, B.; Aidhy, D. S.; Zhang, Y.; Weber, W. J.

    2014-10-16

    The thermodynamic stability and the migration energy barriers of oxygen vacancies in ThO2 –UO2 solid solutions are investigated by density functional theory calculations. In pure ThO2, the formation energy of oxygen vacancy is 7.58 eV and 1.46 eV under O rich and O poor conditions, respectively, while its migration energy barrier is 1.97 eV. The addition of UO2 into ThO2 significantly decreases the energetics of formation and migration of the oxygen vacancy. Among the range of UO2-ThO2 solid solutions studied in this work, UO2 exhibits the lowest formation energy (5.99 eV and -0.13 eV under O rich and O poormore » conditions, respectively) and Th0.25U0.75O2 exhibits the lowest migration energy barrier (~ 1 eV). Moreover, by considering chemical potential, the phase diagram of oxygen vacancy as a function of both temperature and oxygen partial pressure is shown, which could help to gain experimental control over oxygen vacancy concentration.« less

  8. Y H-S I-H HATIOHAL LEAth~~Y~~OF' OtUO ' Industrial Hygiene No...

    Office of Legacy Management (LM)

    H-S I-H HATIOHAL LEAthYOF' OtUO ' Industrial Hygiene No. P.O. Box 158)At.' He&bykation Sample Nos. ? Sk. 0 qq Cinchnail 31;Obio Type of SampleCI" lz -- HEALTH AND SAFETY ...

  9. Direct Electrodeposition of UO2 from Uranyl Bis(trifluoromethanesulfonyl)imide Dissolved in 1-Ethyl-3-methylimidazolium Bis(trifluoromethanesulfonyl)imide Room Temperature Ionic Liquid System

    SciTech Connect (OSTI)

    Freiderich, John W.; Wanigasekara, Eranda P.; Sun, Xiao-Guang; Meisner, Roberta Ann; Meyer, III, Harry M.; Luo, Huimin; Delmau, Laetitia Helene; Dai, Sheng; Moyer, Bruce A

    2013-11-11

    Our study demonstrates a direct electrodeposition of UO2 at a Pt cathode from a solution of uranyl bis(trifluoromethanesulfonyl)imide [UO2(NTf2)2)] in a bulk room-temperature ionic liquid (RTIL), 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide (EMIM+NTf2). Cyclic voltammetry (CV) studies revealed two reduction waves corresponding to the conversion of uranium(VI) to uranium(IV), and a mechanism for the overall electroreduction is proposed. A controlled-potential experiment was performed, holding the reduction potential at–1.0 V for 24 h to obtain a brown-black deposit of UO2 on the Pt cathode. The Faradaic efficiency of the reduction process was determined to be >80%. The UO2deposit was characterized by powder X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS).

  10. Synthesis and structure of R{sub 2}[UO{sub 2}(NO{sub 3}){sub 2}(NCS){sub 2}] (R = Rb or Cs)

    SciTech Connect (OSTI)

    Serezhkin, V. N.; Peresypkina, E. V.; Grigor’eva, V. A.; Virovets, A. V.; Serezhkina, L. B.

    2015-01-15

    Crystals Rb{sub 2}[UO{sub 2}(NO{sub 3}){sub 2}(NCS){sub 2}] (I) and Cs{sub 2}[UO{sub 2}(NO{sub 3}){sub 2}(NCS){sub 2}] (II) have been synthesized and studied by IR spectroscopy and X-ray diffraction. Crystals I are monoclinic, with the following parameters: a = 12.2118(5) Å, b = 10.2545(3) Å, c = 11.8754(4) Å, β = 110.287(1)°, sp. gr. C2/c, Z = 4, and R = 0.0523. Crystals II are orthorhombic, with a = 13.7309(3) Å, b = 10.5749(2) Å, c = 10.1891(2) Å, sp. gr. Pnma, Z = 4, and R = 0.0411. The basic structural units of crystals I and II are one-core complexes [UO{sub 2}(NO{sub 3}){sub 2}(NCS){sub 2}]{sup 2−}, which belong to the crystallochemical group cis-AB{sub 2}{sup 01}M{sub 2}{sup 1} (A = UO{sub 2}{sup 2+}, B{sup 01} = NO{sub 3}{sup −}, M{sup 1} = NCS{sup −}), which are combined into a framework via electrostatic interactions with ions of alkaline metals R (R = Rb or Cs). The structural features of crystals I and II, which condition the formation of [UO{sub 2}(NO{sub 3}){sub 2}(NCS){sub 2}]{sup 2−} complexes with a cis rather than a trans position of isothiocyanate ions in the coordination sphere of uranyl ions, are discussed.

  11. New insight into UO2F2 particulate structure by micro-Raman spectroscopy

    SciTech Connect (OSTI)

    Stefaniak, Elzbieta A.; Darchuk, Larysa; Sapundjiev, Danislav; Kips, Ruth E.; Aregbe, Yetunde; Grieken, Rene Van

    2013-02-19

    Uranyl fluoride particles produced via hydrolysis of uranium hexafluoride have been deposited on different substrates: polished graphite disks, silver foil, stainless steel and gold-coated silicon wafer, and measured with micro-Raman spectroscopy (MRS). All three metallic substrates enhanced the Raman signal delivered by UO2F2 in comparison to graphite. The fundamental stretching of the U–O band appeared at 867 cm–1 in case of the graphite substrate, while in case of the others it was shifted to lower frequencies (down to 839 cm–1). All applied metallic substrates showed the expected effect of Raman signal enhancement; however the gold layer appeared to be most effective. Lastly, application of new substrates provides more information on the molecular structure of uranyl fluoride precipitation, which is interesting for nuclear safeguards and nuclear environmental analysis.

  12. Phase-field simulations of intragranular fission gas bubble evolution in UO2 under post-irradiation thermal annealing

    SciTech Connect (OSTI)

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

    2013-05-15

    Fission gas bubble is one of evolving microstructures, which affect thermal mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking, in operating nuclear fuels. Therefore, fundamental understanding of gas bubble evolution kinetics is essential to predict the thermodynamic property and performance changes of fuels. In this work, a generic phasefield model was developed to describe the evolution kinetics of intra-granular fission gas bubbles in UO2 fuels under post-irradiation thermal annealing conditions. Free energy functional and model parameters are evaluated from atomistic simulations and experiments. Critical nuclei size of the gas bubble and gas bubble evolution were simulated. A linear relationship between logarithmic bubble number density and logarithmic mean bubble diameter is predicted which is in a good agreement with experimental data.

  13. Modeling the Distribution of Acidity within Nuclear Fuel (UO{sub 2}) Corrosion Product Deposits and Porous Sites

    SciTech Connect (OSTI)

    Cheong, W.J.; Keech, P.G.; Wren, J.C.; Shoesmith, D.W.; Qin, Z.

    2007-07-01

    A model for acidity within pores within corrosion products on anodically-dissolving UO{sub 2} was developed using Comsol Multiphysics 3.2 to complement ongoing electrochemical measurements. It was determined that a depression of pH within pores can be maintained if: electrochemically measured dissolution currents used in the calculations are attenuated to reflect very localized pores; corrosion potentials exceed -250 mV (vs. SCE); and pore depths are >1 {mu}m for 300 mV or >100 {mu}m for -50 mV (vs. SCE). Mixed diffusional-chemical equilibria control is suggested through deviations in the shapes between pH-potential and pH-pore depth plots. (authors)

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

    SciTech Connect (OSTI)

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

    2010-01-15

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

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

    SciTech Connect (OSTI)

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

    2012-01-15

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

  16. High temperature synthesis of two open-framework uranyl silicates with ten-ring channels: Cs{sub 2}(UO{sub 2}){sub 2}Si{sub 8}O{sub 19} and Rb{sub 2}(UO{sub 2}){sub 2}Si{sub 5}O{sub 13}

    SciTech Connect (OSTI)

    Babo, Jean-Marie; Albrecht-Schmitt, Thomas E.

    2013-01-15

    The uranyl silicates Cs{sub 2}(UO{sub 2}){sub 2}Si{sub 8}O{sub 19} and Rb{sub 2}(UO{sub 2}){sub 2}Si{sub 5}O{sub 13} were obtained by mixing stoichiometric amounts of uranium metal, tellurium dioxide, silicon dioxide, and an excess of correspondent alkali metal halide flux. These compounds crystallize in the orthorhombic space groups Pnma and C222 with eight and two units per unit cell, respectively. Their crystal structures are dominated by zippered pentagonal bipyramidal chains of UO{sub 7} and silicates layer that are further connected into 3D frameworks. The cesium compound has silicate double layers while rubidium has a single layer. Six-ring voids and ten-ring channels are found in both compounds. - Graphical abstract: A view of the three-dimensional network structure of Cs{sub 2}(UO{sub 2}){sub 2}Si{sub 8}O{sub 19}. Highlights: Black-Right-Pointing-Pointer Three-dimensional uranium silicates. Black-Right-Pointing-Pointer Analogs of natural uranyl silicate minerals. Black-Right-Pointing-Pointer Complexity and symmetry ambiguity of uranyl silicates.

  17. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

    SciTech Connect (OSTI)

    Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

    2007-08-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

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

    SciTech Connect (OSTI)

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

    2014-03-01

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

  20. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

    SciTech Connect (OSTI)

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-07-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  1. The gas-phase bis-uranyl nitrate complex [(UO2)2(NO3)5]-: infrared spectrum and structure

    SciTech Connect (OSTI)

    Groenewold, G. S.; van Stipdonk, Michael J.; Oomens, Jos; De Jong, Wibe A.; McIIwain, Michael E.

    2011-12-01

    The infrared spectrum of the bis-uranyl nitrate complex [(UO2)2(NO3)5]- was measured in the gas phase using multiple photon dissociation (IRMPD). Intense absorptions corresponding to the nitrate symmetric and asymmetric vibrations, and the uranyl asymmetric vibration were observed. The nitrate v3 vibrations indicate the presence of nitrate in a bridging configuration bound to both uranyl cations, and probably two distinct pendant nitrates in the complex. The coordination environment of the nitrate ligands and the uranyl cations were compared to those in the mono-uranyl complex. Overall, the uranyl cation is more loosely coordinated in the bis-uranyl complex [(UO2)2(NO3)5]- compared to the mono-complex [UO2(NO3)3]-, as indicated by a higher O-U-O asymmetric stretching (v3) frequency. However, the pendant nitrate ligands are more strongly bound in the bis-complex than they are in the mono-uranyl complex, as indicated by the v3 frequencies of the pendant nitrate, which are split into nitrosyl and O-N-O vibrations as a result of bidentate coordination. These phenomena are consistent with lower electron density donation per uranyl by the nitrate bridging two uranyl centers compared to that of a pendant nitrate in the mono-uranyl complex. The lowest energy structure predicted by density functional theory (B3LYP functional) calculations was one in which the two uranyl molecules bridged by a single nitrate coordinated in a bis-bidentate fashion. Each uranyl molecule was coordinated by two pendant nitrate ligands. The corresponding vibrational spectrum was in excellent agreement with the IRMPD measurement, confirming the structural assignment.

  2. Topologically identical, but geometrically isomeric layers in hydrous α-, β-Rb[UO{sub 2}(AsO{sub 3}OH)(AsO{sub 2}(OH){sub 2})]·H{sub 2}O and anhydrous Rb[UO{sub 2}(AsO{sub 3}OH)(AsO{sub 2}(OH){sub 2})

    SciTech Connect (OSTI)

    Yu, Na; Klepov, Vladislav V.; Villa, Eric M.; Bosbach, Dirk; Suleimanov, Evgeny V.; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.; Alekseev, Evgeny V.

    2014-07-01

    The hydrothermal reaction of uranyl nitrate with rubidium nitrate and arsenic (III) oxide results in the formation of polymorphic α- and β-Rb[UO{sub 2}(AsO{sub 3}OH)(AsO{sub 2}(OH){sub 2})]·H{sub 2}O (α-, β-RbUAs) and the anhydrous phase Rb[UO{sub 2}(AsO{sub 3}OH)(AsO{sub 2}(OH){sub 2})] (RbUAs). These phases were structurally, chemically and spectroscopically characterized. The structures of all three compounds are based upon topologically identical, but geometrically isomeric layers. The layers are linked with each other by means of the Rb cations and hydrogen bonding. Dehydration experiments demonstrate that water deintercalation from hydrous α- and β-RbUAs yields anhydrous RbUAs via topotactic reactions. - Graphical abstract: Three different layer geometries observed in the structures of Rb[UO{sub 2}(AsO{sub 3}OH)(AsO{sub 2}(OH){sub 2})] and α- and β- Rb[UO{sub 2}(AsO{sub 3}OH)(AsO{sub 2}(OH){sub 2})]·H{sub 2}O. Two different coordination environments of uranium polyhedra (types I and II) are shown schematically on the top of the figure. - Highlights: • Three new uranyl arsenates were synthesized from the hydrothermal reactions. • The phases consist of the topologically identical but geometrically different layers. • Topotactic transitions were observed in the processes of mono-hyrates dehydration.

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

    SciTech Connect (OSTI)

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

    1996-12-30

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

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

    SciTech Connect (OSTI)

    Elam, K.R.

    2003-10-07

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

  5. 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)

    Babo, Jean-Marie; Albrecht-Schmitt, Thomas E.

    2013-10-15

    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) Å, b=11.052(2) Å, c=10.666(2) Å and β=93.897(3)°), P1{sup ¯} (a=7.051(2) Å, b=7.198(2) Å, c=8.314(2) Å, α=107.897(3)°, β=102.687(3)° and γ=100.564(3)°) and C2/c (a=17.862(4) Å, b=6.931(1) Å, c=20.133(4) Å and β=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.

  6. Molten salt flux synthesis and structure of the new layered uranyl tellurite, K{sub 4}[(UO{sub 2}){sub 5}(TeO{sub 3}){sub 2}O{sub 5}

    SciTech Connect (OSTI)

    Woodward, Jonathan D.; Albrecht-Schmitt, Thomas E. . E-mail: albreth@auburn.edu

    2005-09-15

    The reaction of UO{sub 3} and TeO{sub 3} with a KCl flux at 800 deg. C for 3 days yields single crystals of K{sub 4}[(UO{sub 2}){sub 5}(TeO{sub 3}){sub 2}O{sub 5}]. The structure of the title compound consists of layered, two-dimensional {sub {infinity}}{sup 2}[(UO{sub 2}){sub 5}(TeO{sub 3}){sub 2}O{sub 5}]{sup 4-} sheets arranged in a stair-like topology separated by potassium cations. Contained within these sheets are one-dimensional uranium oxide ribbons consisting of UO{sub 7} pentagonal bipyramids and UO{sub 6} tetragonal bipyramids. The ribbons are in turn linked by corner-sharing with trigonal pyramidal TeO{sub 3} units to form sheets. The lone-pair of electrons from the TeO{sub 3} groups are oriented in opposite directions with respect to one another on each side of the sheets rendering each individual sheet nonpolar. The potassium cations form contacts with nearby tellurite units and axial uranyl oxygen atoms. Crystallographic data (193K, MoK{alpha}, {lambda}=0.71073A): triclinic, space group P1-bar , a=6.8514(5)A, b=7.1064(5)A, c=11.3135(8)A, {alpha}=99.642(1){sup o}, {beta}=93.591(1){sup o}, {gamma}=100.506(1){sup o}, V=531.48(7)A{sup 3}, Z=1,R(F)=4.19% for 149 parameters and 2583 reflections with I>2{sigma}(I)

  7. LWR fuel assembly designs for the transmutation of LWR Spent Fuel TRU with FCM and UO{sub 2}-ThO{sub 2} Fuels

    SciTech Connect (OSTI)

    Bae, G.; Hong, S. G.

    2013-07-01

    In this paper, transmutation of transuranic (TRU) nuclides from LWR spent fuels is studied by using LWR fuel assemblies which consist of UO{sub 2}-ThO{sub 2} fuel pins and FCM (Fully Ceramic Microencapsulated) fuel pins. TRU from LWR spent fuel is loaded in the kernels of the TRISO particle fuels of FCM fuel pins. In the FCM fuel pins, the TRISO particle fuels are distributed in SiC matrix having high thermal conductivity. The loading patterns of fuel pins and the fuel compositions are searched to have high transmutation rate and feasible neutronic parameters including pin power peaking, temperature reactivity coefficients, and cycle length. All studies are done only in fuel assembly calculation level. The results show that our fuel assembly designs have good transmutation performances without multi-recycling and without degradation of the safety-related neutronic parameters. (authors)

  8. RELAP5/MOD3.2 Analysis of a VVER-1000 Reactor with UO{sub 2} Fuel and Mixed-Oxide Fuel

    SciTech Connect (OSTI)

    Hassan, Yassin A.; Fu Chun

    2004-12-15

    A RELAP5/MOD3.2 model of the VVER-1000/MODEL V320 nuclear power plant was modified and a large-break loss-of-coolant accident (LBLOCA) in the cold leg was simulated. In this analysis, the core consisted of 162 UO{sub 2} assemblies and 1 mixed-oxide assembly. The results from the simulation were compared with the results from a similar study performed with the Russian computer program TECH-M. An uncertainty analysis was performed on the peak cladding temperature following a similar methodology called code scaling, applicability, and uncertainty. Monte Carlo calculations were performed using the response surface inferred from 15 runs of RELAP5 calculations. The result of this study shows that the emergency core coolant system would be sufficient to keep the cladding temperature during the LBLOCA scenario well below the required maximum limit.

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

    SciTech Connect (OSTI)

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

    2015-09-01

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

  10. Fission gas and iodine release measured in IFA-430 up to 15 GWd/t UO/sub 2/ burnup. [PWR; BWR

    SciTech Connect (OSTI)

    Appelhans, A.D.; Turnbull, J.A.; White, R.J.

    1983-01-01

    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper presents a summary of the results up to December, 1982. The data cover fuel centerline temperatures ranging from 700 to 1500/sup 0/C for average linear heat ratings of 16 to 35 kW/m. The measurements have been performed for the period between 4.2 and 14.8 GWd/t UO/sub 2/ of burnup of the Instrumented Fuel Assembly 430 (IFA-430). The measurement program has been directed toward quantifying the release of the short-lived radioactive noble gases and iodines.

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

    SciTech Connect (OSTI)

    Doucet, M.; Durant Terrasson, L.; Mouton, J.

    2006-07-01

    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

  12. Possible Demonstration of a Polaronic Bose-Einstein(-Mott) Condensate in UO2(+x) by Ultrafast THz Spectroscopy and Microwave Dissipation

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

    Conradson, Steven D.; Gilbertson, Steven M.; Daifuku, Stephanie L.; Kehl, Jeffrey A.; Durakiewicz, Tomasz; Andersson, David A.; Bishop, Alan R.; Byler, Darrin D.; Maldonado, Pablo; Oppeneer, Peter M.; et al

    2015-10-16

    Bose-Einstein condensates (BECs) composed of polarons would be an advance because they would combine coherently charge, spin, and a crystal lattice. Following our earlier report of unique structural and spectroscopic properties, we now identify potentially definitive evidence for polaronic BECs in photo- and chemically doped UO2(+x) on the basis of exceptional coherence in the ultrafast time dependent terahertz absorption and microwave spectroscopy results that show collective behavior including dissipation patterns whose precedents are condensate vortex and defect disorder and condensate excitations. Furthermore, that some of these signatures of coherence in an atom-based system extend to ambient temperature suggests a novelmore »mechanism that could be a synchronized, dynamical, disproportionation excitation, possibly via the solid state analog of a Feshbach resonance that promotes the coherence. Such a mechanism would demonstrate that the use of ultra-low temperatures to establish the BEC energy distribution is a convenience rather than a necessity, with the actual requirement for the particles being in the same state that is not necessarily the ground state attainable by other means. Interestingly, a macroscopic quantum object created by chemical doping that can persist to ambient temperature and resides in a bulk solid would be revolutionary in a number of scientific and technological fields.« less

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

    SciTech Connect (OSTI)

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

    2007-11-01

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

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

    SciTech Connect (OSTI)

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

    2014-11-07

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

  15. Possible Demonstration of a Polaronic Bose-Einstein(-Mott) Condensate in UO2(+x) by Ultrafast THz Spectroscopy and Microwave Dissipation

    SciTech Connect (OSTI)

    Conradson, Steven D.; Gilbertson, Steven M.; Daifuku, Stephanie L.; Kehl, Jeffrey A.; Durakiewicz, Tomasz; Andersson, David A.; Bishop, Alan R.; Byler, Darrin D.; Maldonado, Pablo; Oppeneer, Peter M.; Valdez, James A.; Neidig, Michael L.; Rodriguez, George

    2015-10-16

    Bose-Einstein condensates (BECs) composed of polarons would be an advance because they would combine coherently charge, spin, and a crystal lattice. Following our earlier report of unique structural and spectroscopic properties, we now identify potentially definitive evidence for polaronic BECs in photo- and chemically doped UO2(+x) on the basis of exceptional coherence in the ultrafast time dependent terahertz absorption and microwave spectroscopy results that show collective behavior including dissipation patterns whose precedents are condensate vortex and defect disorder and condensate excitations. Furthermore, that some of these signatures of coherence in an atom-based system extend to ambient temperature suggests a novel mechanism that could be a synchronized, dynamical, disproportionation excitation, possibly via the solid state analog of a Feshbach resonance that promotes the coherence. Such a mechanism would demonstrate that the use of ultra-low temperatures to establish the BEC energy distribution is a convenience rather than a necessity, with the actual requirement for the particles being in the same state that is not necessarily the ground state attainable by other means. Interestingly, a macroscopic quantum object created by chemical doping that can persist to ambient temperature and resides in a bulk solid would be revolutionary in a number of scientific and technological fields.

  16. Possible Demonstration of a Polaronic Bose-Einstein(-Mott) Condensate in UO2(+x) by Ultrafast THz Spectroscopy and Microwave Dissipation

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

    Conradson, Steven D.; Gilbertson, Steven M.; Daifuku, Stephanie L.; Kehl, Jeffrey A.; Durakiewicz, Tomasz; Andersson, David A.; Bishop, Alan R.; Byler, Darrin D.; Maldonado, Pablo; Oppeneer, Peter M.; et al

    2015-10-16

    Bose-Einstein condensates (BECs) composed of polarons would be an advance because they would combine coherently charge, spin, and a crystal lattice. Following our earlier report of unique structural and spectroscopic properties, we now identify potentially definitive evidence for polaronic BECs in photo- and chemically doped UO2(+x) on the basis of exceptional coherence in the ultrafast time dependent terahertz absorption and microwave spectroscopy results that show collective behavior including dissipation patterns whose precedents are condensate vortex and defect disorder and condensate excitations. Furthermore, that some of these signatures of coherence in an atom-based system extend to ambient temperature suggests a novelmore » mechanism that could be a synchronized, dynamical, disproportionation excitation, possibly via the solid state analog of a Feshbach resonance that promotes the coherence. Such a mechanism would demonstrate that the use of ultra-low temperatures to establish the BEC energy distribution is a convenience rather than a necessity, with the actual requirement for the particles being in the same state that is not necessarily the ground state attainable by other means. Interestingly, a macroscopic quantum object created by chemical doping that can persist to ambient temperature and resides in a bulk solid would be revolutionary in a number of scientific and technological fields.« less

  17. Mesoscale Benchmark Demonstration Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing

    SciTech Connect (OSTI)

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David

    2012-04-11

    A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

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

    SciTech Connect (OSTI)

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

    2009-04-15

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

  20. Pipe diffusion at dislocations in UO2

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

    TEM images of high burn-up fuel pellets show that the dislocations are heavily decorated by fission product precipitates, including gas bubbles and metallic particles 5. This is ...

  1. Structure and dynamics of complexes of the uranyl ion with nonamethylimidodiphosphoramide (NIPA). 2. NMR studies of complexes (UO/sub 2/(NIPA)/sub 2/X)(CIO/sub 4/)/sub 2/ with X = H/sub 2/O, MeOH, EtOH, or Me/sub 2/CO

    SciTech Connect (OSTI)

    Rodehueser, L.; Rubini, P.R.; Bokolo, K.; Delpuech, J.J.

    1982-03-01

    The /sup 31/P and /sup 1/H spectra at -90/sup 0/C of the title uranyl complex ions (prepared as solutions of the solid perchlorates in inert anhydrous organic solvents (CH/sub 3/NO/sub 2/, CH/sub 2/Cl/sub 2/)) reveal a pentacoordinated arrangement of two symmetrically doubly bonded NIPA molecules and one solvent molecule about the uranyl group. In the case of (UO/sub 2/(NIPA)/sub 2/(EtOH))(ClO/sub 4/)/sub 2/, an intermolecular exchange between bound and free ethanol molecules is observed above -75/sup 0/C upon addition of ethanol to a solution of the complex. The observed rate law, k/sub inter/ = kK(EtOH)/(1 + K(EtOH) is accounted for by the existence of an outer-sphere complex (UO/sub 2//sup 2 +/(NIPA)/sub 2/(EtOH))EtOH in fast equilibrium (K) with the initial complex and free ethanol. The rate-determining step (k) consists of an outer-sphere to inner-sphere interchange of ethanol molecules. The thermodynamic and kinetic parameters are K(25/sup 0/C) = 15.8 dm/sup 3/ mol/sup -1/, k(25/sup 0/C) = 1.0 x 10/sup 4/s/sup -1/, ..delta..H and ..delta..H/sub inter//sup + +/ = -4.8 and 7.6 kcal mol/sup -1/, and ..delta..S and ..delta..S/sub inter//sup + +/ = 10.7 and -14.7 eu. A second exchange takes place at higher temperatures (above -30/sup 0/C) yielding full dynamic equivalence of the phosphorus nuclei of the coordinated NIPA molecules. The observed rate law k/sub intra/ = k/sub ex/(1 + K(EtOH)) reveals that the internal rearrangement of NIPA molecules occurs on the complex ion (UO/sub 2/(NIPA)/sub 2/(EtOH))/sup 2 +/ but not on the outer-sphere complex: k/sub ex/(25/sup 0/C) = 0.91 x 10/sup 3/s/sup -1/, ..delta..H/sub intra//sup + +/ = 10.6 kcal mol/sup -1/ and ..delta..S/sub intra//sup + +/ = -9.4 eu. Possible mechanisms for this exchange are discussed. 5 figures, 2 tables.

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

    Office of Scientific and Technical Information (OSTI)

    Authors: Xian-Ming Bai ; Anter El-Azab ; Jianguo Yu ; Todd R. Allen Publication Date: 2013-01-01 OSTI Identifier: 1057705 Report Number(s): INLJOU-12-27046 DOE Contract Number: ...

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

    SciTech Connect (OSTI)

    Gerber, M.S.

    1997-11-25

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

  4. Materials Data on UO2 (SG:225) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2015-01-27

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  5. Modeling of Fission Gas Release in UO2

    SciTech Connect (OSTI)

    MH Krohn

    2006-01-23

    A two-stage gas release model was examined to determine if it could provide a physically realistic and accurate model for fission gas release under Prometheus conditions. The single-stage Booth model [1], which is often used to calculate fission gas release, is considered to be oversimplified and not representative of the mechanisms that occur during fission gas release. Two-stage gas release models require saturation at the grain boundaries before gas is release, leading to a time delay in release of gases generated in the fuel. Two versions of a two-stage model developed by Forsberg and Massih [2] were implemented using Mathcad [3]. The original Forsbers and Massih model [2] and a modified version of the Forsberg and Massih model that is used in a commercially available fuel performance code (FRAPCON-3) [4] were examined. After an examination of these models, it is apparent that without further development and validation neither of these models should be used to calculate fission gas release under Prometheus-type conditions. There is too much uncertainty in the input parameters used in the models. In addition. the data used to tune the modified Forsberg and Massih model (FRAPCON-3) was collected under commercial reactor conditions, which will have higher fission rates relative to Prometheus conditions [4].

  6. METHOD OF MAKING UO$sub 2$-Bi SLURRIES

    DOE Patents [OSTI]

    Hahn, H.T.

    1960-05-24

    A process is given of preparing an easily dispersible slurry of uranium dioxide in bismuth. A mixture of bismuth oxide, uranium, and bismuth are heated in a capsule to a temperature over the melting point of bismuth oxide. The amount of bismuth oxide used is less than that stoichiometrically required because the oxygen in the capsule also enters into the reaction.

  7. Materials Data on UOs2 (SG:227) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2015-02-09

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  8. OSMOSE program : statistical review of oscillation measurements in the MINERVE reactor R1-UO2 configuration.

    SciTech Connect (OSTI)

    Stoven, G.; Klann, R.; Zhong, Z.; Nuclear Engineering Division

    2007-08-28

    The OSMOSE program is a collaboration on reactor physics experiments between the United States Department of Energy and the France Commissariat Energie Atomique. At the working level, it is a collaborative effort between the Argonne National Laboratory and the CEA Cadarache Research Center. The objective of this program is to measure very accurate integral reaction rates in representative spectra for the actinides important to future nuclear system designs, and to provide the experimental data for improving the basic nuclear data files. The main outcome of the OSMOSE measurement program will be an experimental database of reactivity-worth measurements in different neutron spectra for the heavy nuclides. This database can then be used as a benchmark to verify and validate reactor analysis codes. The OSMOSE program (Oscillation in Minerve of isotopes in Eupraxic Spectra) aims at improving neutronic predictions of advanced nuclear fuels through oscillation measurements in the MINERVE facility on samples containing the following separated actinides: {sup 232}Th, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm, and {sup 245}Cm. The first part of this report provides an overview of the experimental protocol and the typical processing of a series of experimental results which is currently performed at CEA-Cadarache. In the second part of the report, improvements to this technique are presented, as well as the program that was created to process oscillation measurement results from the MINERVE facility in the future.

  9. REACTOR HAVING NaK-UO$sub 2$ SLURRY HELICALLY POSITIONED IN A GRAPHITE MODERATOR

    DOE Patents [OSTI]

    Rodin, M.B.; Carter, J.C.

    1962-05-15

    A reactor utilizing 20% enriched uranium consists of a central graphite island in cylindrical form, with a spiral coil of tubing fitting against the central island. An external graphite moderator is placed around the central island and coil. A slurry of uranium dioxide dispersed in alkali metal passes through the coil to transfer heat externally to the reactor. There are also conventional controls for regulating the nuclear reaction. (AEC)

  10. A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide...

    Office of Scientific and Technical Information (OSTI)

    the CONFU assembly exhibits negative reactivity feedback coefficients comparable in ... NUCLEAR FUELS; PWR TYPE REACTORS; REACTIVITY COEFFICIENTS; REPROCESSING; SIMULATION; ...

  11. Project report to STB/UO, Northern New Mexico Community College two- year college initiative: Biotechnology

    SciTech Connect (OSTI)

    1996-03-01

    This report summarizes the experiences gained in a project involving faculty direct undergraduate research focused on biotechnology and its applications. The biology program at Northern New Mexico Community College has been involved in screening for mutations in human DNA and has developed the ability to perform many standard and advanced molecular biology techniques. Most of these are based around the polymerase chain reaction (PCR) and include the use of single strand conformation polymorphism analysis (SSCP), denaturing gradient gel electrophoresis (DGGE) in the screening for mutant DNA molecules, and the capability to sequence PCR generated fragments of DNA using non-isotopic imaging. At Northern, these activities have a two-fold objective: (1) to bring current molecular biology techniques to the teaching laboratory, and (2) to support the training of minority undergraduates in research areas that stimulate them to pursue advanced degrees in the sciences.

  12. Materials Data on K2UO4 (SG:139) by Materials Project

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

    Kristin Persson

    2014-11-02

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  13. Materials Data on Na3UO4 (SG:65) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2014-11-02

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

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

    SciTech Connect (OSTI)

    Gerber, M.S.

    1993-11-01

    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.

  15. Random-Walk Monte Carlo Simulation of Intergranular Gas Bubble Nucleation in UO2 Fuel

    SciTech Connect (OSTI)

    Yongfeng Zhang; Michael R. Tonks; S. B. Biner; D.A. Andersson

    2012-11-01

    Using a random-walk particle algorithm, we investigate the clustering of fission gas atoms on grain bound- aries in oxide fuels. The computational algorithm implemented in this work considers a planar surface representing a grain boundary on which particles appear at a rate dictated by the Booth flux, migrate two dimensionally according to their grain boundary diffusivity, and coalesce by random encounters. Specifically, the intergranular bubble nucleation density is the key variable we investigate using a parametric study in which the temperature, grain boundary gas diffusivity, and grain boundary segregation energy are varied. The results reveal that the grain boundary bubble nucleation density can vary widely due to these three parameters, which may be an important factor in the observed variability in intergranular bubble percolation among grain boundaries in oxide fuel during fission gas release.

  16. Recovery of UO{sub 2}/PuO{sub 2} in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1992-01-01

    This invention is comprised of a process for converting PuO{sub 2} and U0{sub 2} present in an electrorefiner to the chlorides, by contacting the PuO{sub 2} and U0{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 U0{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 0{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 0{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}.

  17. Materials Data on Ca3UO6 (SG:148) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2014-11-02

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  18. Strong electron correlation in UO{sub 2}{sup -}: A photoelectron...

    Office of Scientific and Technical Information (OSTI)

    ... Molecular Engineering of Ministry of Education, Tsinghua University, Beijing 100084 ... Resource Type: Journal Article Resource Relation: Journal Name: Journal of Chemical Physics; ...

  19. Materials Data on BaUO3 (SG:62) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2014-11-02

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  20. Materials Data on SrUO4 (SG:166) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2014-11-02

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

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

    Office of Legacy Management (LM)

    ... Several colleges and universities, including the University of California, the University ... NAME California Inst. of Technology University of California Yale Heavy Ion Linear ...

  2. Safety testing of AGR-2 UO2 compacts 3-3-2 and 3-4-2

    SciTech Connect (OSTI)

    Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.; Montgomery, Fred C.

    2015-09-01

    Post-irradiation examination (PIE) is in progress on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2) [Collin 2014]. The AGR-2 PIE will build upon new information and understanding acquired throughout the recently-concluded six-year AGR-1 PIE campaign [Demkowicz et al. 2015] and establish a database for the different AGR-2 fuel designs.

  3. Fully-coupled engineering and mesoscale simulations of thermal conductivity in UO2 fuel using an implicit multiscale approach

    SciTech Connect (OSTI)

    Michael Tonks; Derek Gaston; Cody Permann; Paul Millett; Glen Hansen; Chris Newman

    2009-08-01

    Reactor fuel performance is sensitive to microstructure changes during irradiation (such as fission gas and pore formation). This study proposes an approach to capture microstructural changes in the fuel by a two-way coupling of a mesoscale phase field irradiation model to an engineering scale, finite element calculation. This work solves the multiphysics equation system at the engineering-scale in a parallel, fully-coupled, fully-implicit manner using a preconditioned Jacobian-free Newton Krylov method (JFNK). A sampling of the temperature at the Gauss points of the coarse scale is passed to a parallel sequence of mesoscale calculations within the JFNK function evaluation phase of the calculation. The mesoscale thermal conductivity is calculated in parallel, and the result is passed back to the engineering-scale calculation. As this algorithm is fully contained within the JFNK function evaluation, the mesoscale calculation is nonlinearly consistent with the engineering-scale calculation. Further, the action of the Jacobian is also consistent, so the composite algorithm provides the strong nonlinear convergence properties of Newton's method. The coupled model using INL's \\bison\\ code demonstrates quadratic nonlinear convergence and good parallel scalability. Initial results predict the formation of large pores in the hotter center of the pellet, but few pores on the outer circumference. Thus, the thermal conductivity is is reduced in the center of the pellet, leading to a higher internal temperature than that in an unirradiated pellet.

  4. Fonsi.Leo.DOC

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

    ... empty and 147 full); uranium dioxide (UO 2 ) (UO 2 inventory on the Hanford Site consists of depleted and normal uranium pellets, powder, and fuel pins containing UO 2 pellets). ...

  5. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    SciTech Connect (OSTI)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  6. Recalculation of the Critical Size and Multiplication Constant of a Homogeneous UO{sub 2} - D{sub 2}O Mixtures

    DOE R&D Accomplishments [OSTI]

    Wigner, E. P.; Weinberg, A. M.; Stephenson, J.

    1944-02-11

    The multiplication constant and optimal concentration of a slurry pile is recalculated on the basis of Mitchell`s experiments on resonance absorption. The smallest chain reacting unit contains 45 to 55 m{sup 3}of D{sub 2}O. (auth)

  7. Recalculation of the Critical Size and Multiplication Constant of a Homogeneous UO{sub 2}-D{sub 2}O Mixtures

    DOE R&D Accomplishments [OSTI]

    Wigner, E. P.; Weinberg, A. M.; Stephenson, J.

    1944-02-11

    The multiplication constant and optimal concentration of a slurry pile is recalculated on the basis of Mitchell's experiments on resonance absorption. The smallest chain reacting unit contains 45 to 55 m{sup 3}of d{sub 2}O. (auth).

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

    SciTech Connect (OSTI)

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

    2015-05-13

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

  9. VERA Core Physics Benchmark Progression Problems Specifications

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

    ... It is a very thin ZrB 2 coating on selected UO 2 fuel pellets in an assembly. Because the ... Gadolinia is mixed homogeneously within the UO 2 fuel pellets for a few select rods in the ...

  10. DOE-HDBK-1113-98, CN 1, Reaffirm; Radiological Safety Training...

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

    ... The UO 2 powder is compacted into cylindrical pellets that are loaded into thin walled ... The UO 2 powder is compacted into cylindrical pellets that are loaded into thinwalled ...

  11. Advanced Nuclear Final Solicitation Fact Sheet_Dec-2014

    Office of Environmental Management (EM)

    ... is "reconverted" from enriched UF6 gas from enrichment plants; (2) formation of UO2 pellets from UO2 powder through compaction and sintering; and(3) fuel assembly (i.e. insertion ...

  12. METHOD OF PREPARING A CERAMIC FUEL ELEMENT

    DOE Patents [OSTI]

    Ross, W.T.; Bloomster, C.H.; Bardsley, R.E.

    1963-09-01

    A method is described for preparing a fuel element from -325 mesh PuO/ sub 2/ and -20 mesh UO/sub 2/, and the steps of screening --325 mesh UO/sub 2/ from the -20 mesh UO/sub 2/, mixing PuO/sub 2/ with the --325 mesh UO/sub 2/, blending this mixture with sufficient --20 mesh UO/sub 2/ to obtain the desired composition, introducing the blend into a metal tube, repeating the procedure until the tube is full, and vibrating the tube to compact the powder are included. (AEC)

  13. Investigation of the Polymorphs and Hydrolysis of Uranium Trioxide

    SciTech Connect (OSTI)

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

    2013-04-01

    This work focuses on progress in gaining a better understanding of the polymorphic nature of the UO3-water system, one of several important materials associated with the nuclear fuel cycle. The UO3-water system is complex and has not been fully characterized, even though these species are common throughout the fuel cycle. Powder x-ray diffraction, Raman and fluorescence characterization was performed on polymorphic forms of UO3 and UO3 hydrolysis products for the purpose of developing some predictive capability of estimating process history and utility, e.g. for polymorphic phases of unknown origin. Specifically, we have investigated three industrially relevant production pathways of UO3 and discovered a previously unknown low temperature route to β-UO3. Pure phases of UO3, hydrolysis products and starting materials were used to establish optical spectroscopic signatures for these compounds.

  14. Molten uranium dioxide structure and dynamics

    SciTech Connect (OSTI)

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

    2014-11-21

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

  15. Molten uranium dioxide structure and dynamics

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

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

    2014-11-21

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

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

    SciTech Connect (OSTI)

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

    2014-12-01

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

  17. SEPARATION OF URANIUM AND PLUTONIUM OXIDES

    DOE Patents [OSTI]

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

    1961-12-01

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

  18. Microsoft Word - DOE-ID-14-057 University of Florida EC B3-6.doc

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

    7 SECTION A. Project Title: Experimental Validation of UO2 Microstructural Evolution for NEAMS tool MARMOT - University of Florida SECTION B. Project Description The University of Florida will experimentally validate the NEAMS tool for the computer model MARMOT. Surrogate oxide (CeO2) and depleted uranium oxide (UO2) will be used for this study and will be formed into pellets of various types then characterized. SECTION C. Environmental Aspects / Potential Sources of Impact The depleted UO 2

  19. ENG-RCAL-028-r1.PDF

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

    for the Shipment of Unirradiated Uranium Preparer: J. L. Boles Date 53100 ... of the transportation of unirradiated uranium in the form of metallic billets, UO 3 ...

  20. Possible Demonstration of a Polaronic Bose-Einstein(-Mott) Condensate...

    Office of Scientific and Technical Information (OSTI)

    Demonstration of a Polaronic Bose-Einstein(-Mott) Condensate in UO2(+x) by Ultrafast THz Spectroscopy and Microwave Dissipation Citation Details In-Document Search Title: Possible ...

  1. OPTICAL PROPERTIES OF A MECHANICALLY POLISHED AND AIR-EQUILIBRATED...

    Office of Scientific and Technical Information (OSTI)

    Conference: OPTICAL PROPERTIES OF A MECHANICALLY POLISHED AND AIR-EQUILIBRATED 111 UO2 SURFACE BY RAMAN AND ELLIPSOMETRIC SPECTROSCOPY Citation Details In-Document Search Title: ...

  2. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    S. ; Gu, Genda ; Gilbertson, Steve M. ; Taylor, Antoinette ; Rodriguez, George A ... 5f electrons inUO2 Au, Yongqiang Q ; Taylor, Antoinette J ; Durakiewicz, Tomasz ; ...

  3. FUEL FOR NEUTRONIC REACTORS AND PROCESS OF MAKING

    DOE Patents [OSTI]

    Abraham, B.M.; Flotow, H.E.

    1961-05-01

    A fuel material is offered for nuclear reactors consisting of UO/sub 2// sub .//sub 0//sub 0/ suspended in a sodium-containing liquid metal.

  4. final ERI-2142 18-1501 Analysis of Potential Effects on Domestic...

    Office of Environmental Management (EM)

    ... may contain natural uranium pellets, such that some uranium does not require enrichment; the feed material for these natural uranium pellets may also be made from UO 2 , ...

  5. ARQ07-4.indd

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

    ... uranium, low-enriched uranium fuel pellets, highly enriched uranium, plutonium, and contaminated scrap metal. First case: uranium pellets Uranium dioxide (UO 2 ) pellets are used ...

  6. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... parameters are investigated based on available experimental and atomic scale data. ... Lattice expansion of UO2 was observed in x-ray diffraction after ion irradiation up to ...

  7. Microsoft Word - 1aDOE-ID-12-047 Westinghouse EC B3-6 NRC.doc

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

    A. Project Title: Development of LWR Fuels Enhanced Accident Tolerance - Westinghouse Electric ... the currently used Zr + UO2 fuel system with and enhanced accident tolerant fuel. ...

  8. Ordinary Isotropic Peridynamic Models Position Aware Viscoelastic...

    Office of Scientific and Technical Information (OSTI)

    ... state t(Y) pcox + 2u"o + 2 a,:cr( - el) i rfl Sandia National laboratories John ... state t(Y) ymx + 2u"o + 2 g,cr( - el) i Governing equation for l . 1 . l + ...

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

    SciTech Connect (OSTI)

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

    2014-12-01

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

  10. METHOD AND APPARATUS FOR CALCINING SALT SOLUTIONS

    DOE Patents [OSTI]

    Lawroski, S.; Jonke, A.A.; Taecker, R.G.

    1961-10-31

    A method is given for converting uranyl nitrate solution into solid UO/ sub 3/, The solution is sprayed horizontally into a fluidized bed of UO/sub 3/ particles at 310 to 350 deg C by a nozzle of the coaxial air jet type at about 26 psig, Under these conditions the desired conversion takes place, and caking in the bed is avoided.

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

    SciTech Connect (OSTI)

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D.; Stogov, Yu. V.

    2014-12-15

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

  12. Accelerator breeder with uranium, thorium target

    SciTech Connect (OSTI)

    Takahashi, H.; Powell, J.; Kouts, H.

    1983-01-01

    An accelerator breeder, that uses a low-enriched fuel as the target material, can produce substantial amounts of fissile material and electric power. A study of H/sub 2/O- and D/sub 2/O-cooled, UO/sub 2/, U, (depleted U), or thorium indicates that U-metal fuel produces a good fissile production rate and electrical power of about 60% higher than UO/sub 2/ fuel. Thorium fuel has the same order of magnitude as UO/sub 2/ fuel for fissile-fuel production, but the generating electric power is substantially lower than in a UO/sub 2/ reactor. Enriched UO/sub 2/ fuel increases the generating electric power but not the fissile-material production rate. The Na-cooled breeder target has many advantages over the H/sub 2/O-cooled breeder target.

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

    SciTech Connect (OSTI)

    Carnahan, C.L.

    1987-12-01

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

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

    SciTech Connect (OSTI)

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

    1991-11-01

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

  15. Complexation of Gluconate with Uranium(VI) in Acidic Solutions: Thermodynamic Study with Structural Analysis

    SciTech Connect (OSTI)

    Zhang, Zhicheng; Helms, G.; Clark, S. B.; Tian, Guoxin; Zanonato, PierLuigi; Rao, Linfeng

    2009-01-05

    Within the pC{sub H} range of 2.5 to 4.2, gluconate forms three uranyl complexes UO{sub 2}(GH{sub 4}){sup +}, UO{sub 2}(GH{sub 3})(aq), and UO{sub 2}(GH{sub 3})(GH{sub 4}){sup -}, through the following reactions: (1) UO{sub 2}{sup 2+} + GH{sub 4}{sup -} = UO{sub 2}(GH{sub 4}){sup +}, (2) UO{sub 2}{sup 2+} + GH{sub 4}{sup -} = UO{sub 2}(GH{sub 3})(aq) + H{sup +}, and (3) UO{sub 2}{sup 2+} + 2GH{sub 4}{sup -} = UO{sub 2}(GH{sub 3})(GH{sub 4}){sup -} + H{sup +}. Complexes were inferred from potentiometric, calorimetric, NMR, and EXAFS studies. Correspondingly, the stability constants and enthalpies were determined to be log {Beta}{sub 1} = 2.2 {+-} 0.3 and {Delta}H{sub 1} = 7.5 {+-} 1.3 kJ mol{sup -1} for reaction (1), log {Beta}{sub 2} = -(0.38 {+-} 0.05) and {Delta}H{sub 2} = 15.4 {+-} 0.3 kJ mol{sup -1} for reaction (2), and log {Beta}{sub 3} = 1.3 {+-} 0.2 and {Delta}H{sub 3} = 14.6 {+-} 0.3 kJ mol{sup -1} for reaction (3), at I = 1.0 M NaClO{sub 4} and t = 25 C. The UO{sub 2}(GH{sub 4}){sup +} complex forms through the bidentate carboxylate binding to U(VI). In the UO{sub 2}(GH{sub 3})(aq) complex, hydroxyl-deprotonated gluconate (GH{sub 3}{sup 2-}) coordinates to U(VI) through the five-membered ring chelation. For the UO{sub 2}(GH{sub 3})(GH{sub 4}){sup -} complex, multiple coordination modes are suggested. These results are discussed in the context of trivalent and pentavalent actinide complexation by gluconate.

  16. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A.

    1990-01-01

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

  17. ALTERATION OF U(VI)-PHASES UNDER OXIDIZING CONDITIONS

    SciTech Connect (OSTI)

    A.P. Deditius; S. Utsunomiya; R.C. Ewing

    2006-02-21

    Uranium-(VI) phases are the primary alteration products of the UO{sub 2} in spent nuclear fuel and the UO{sub 2+x}, in natural uranium deposits. The U(VI)-phases generally form sheet structures of edge-sharing UO{sub 2}{sup 2+} polyhedra. The complexity of these structures offers numerous possibilities for coupled-substitutions of trace metals and radionuclides. The incorporation of radionuclides into U(VI)-structures provides a potential barrier to their release and transport in a geologic repository that experiences oxidizing conditions. In this study, we have used natural samples of UO{sub 2+x}, to study the U(VI)-phases that form during alteration and to determine the fate of the associated trace elements.

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

    SciTech Connect (OSTI)

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

    2013-08-14

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

  19. MEMORANDUM TO: FILE - Oh'

    Office of Legacy Management (LM)

    ... of. Law NCJW York, New York 01 MATERIALS FRO3 DOB L ....-- --. -' - Norsrbt IO, 10x1 "uos.es Ca.e YIIe . -.. . . . . . . . Env. Prot. Bureau N.Y.S. Department of Law Nev ...

  20. Effect of temperature on the complexation of Uranium(VI) with fluoride in aqueous solutions

    SciTech Connect (OSTI)

    Tian, Guoxin; Rao, Linfeng

    2009-05-18

    Complexation of U(VI) with fluoride at elevated temperatures in aqueous solutions was studied by spectrophotometry. Four successive complexes, UO{sub 2}F{sup +}, UO{sub 2}F{sub 2}(aq), UO{sub 2}F{sub 3}{sup -}, and UO{sub 2}F{sub 4}{sup 2-}, were identified, and the stability constants at 25, 40, 55, and 70 C were calculated. The stability of the complexes increased as the temperature was elevated. The enthalpies of complexation at 25 C were determined by microcalorimetry. Thermodynamic parameters indicate that the complexation of U(VI) with fluoride in aqueous solutions at 25 to 70 C is slightly endothermic and entropy-driven. The Specific Ion Interaction (SIT) approach was used to obtain the thermodynamic parameters of complexation at infinite dilution. Structural information on the U(VI)/fluoride complexes was obtained by extended X-ray absorption fine structure spectroscopy.

  1. MOX Fuel Presentation to Duke Board of Directors

    National Nuclear Security Administration (NNSA)

    PuO 2 with 95% depleted UO 2 - Like LEU fuel pellets, MOX fuel pellets are primarily uranium * Fission power comes primarily from plutonium (Pu 239 ) instead of uranium (U 235 )...

  2. E P GPT collectively denotes new ...

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

    ... The model includes 16 cylindrical fuel rods (8 MOX and 8 UO 2 ) moderated by water. For this model, the flux has 256 dimensions representing 128 spa- tial nodes with equal spatial ...

  3. OPTICAL PROPERTIES OF A MECHANICALLY POLISHED AND AIR-EQUILIBRATED...

    Office of Scientific and Technical Information (OSTI)

    AND AIR-EQUILIBRATED 111 UO2 SURFACE BY RAMAN AND ELLIPSOMETRIC SPECTROSCOPY Citation Details In-Document Search Title: OPTICAL PROPERTIES OF A MECHANICALLY POLISHED AND AIR-EQUI...

  4. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. ... Laboratory Low Activity BetaGamma Sources Waste Stream at the Area 5 Radioactive ...

  5. ContaCt

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

    Time-dependent experiments have been successfully performed using a broad-spectrum neutron beam. EnErgy-rESoLvEd nEUtron iMaging Tungsten (red) diffused into UO 2 rod. Fuel pellets ...

  6. DOE-HDBK-1017/2-93; DOE Fundamentals Handbook Material Science...

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

    ... The UO 2 is formed into pellets and clad with zircaloy (water-cooled reactors) or ... Also, pelletized fuel requires a tubular container to hold the pellets in the required ...

  7. CX-012689: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Experimental Validation of UO2 Microstructural Evolution for NEAMS tool MARMOT – University of Florida CX(s) Applied: B3.6Date: 41869 Location(s): FloridaOffices(s): Nuclear Energy

  8. Uranium Mining, Conversion, and Enrichment Industries

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

    ... At the fabrication facility, the enriched UF 6 is converted into UO 2 powder, and then formed into small ceramic pellets. These pellets are then loaded into metal tubes and ...

  9. Microsoft PowerPoint - MOX Adventure_Reactor Subcommittee_Tamara...

    National Nuclear Security Administration (NNSA)

    3 MOX Fuel - General MOX fuel pellets from former weapons plutonium Blend of 5% PuO 2 with 95% depleted UO 2 Like LEU fuel pellets, MOX fuel pellets are primarily ...

  10. NEAMS Update

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

    ... PWR Pressurized water reactor RIA Reactivity-initiated accident RPL Reactors Product Line SFR Sodium fast reactor UO Uranium oxide VHTR Very high-temperature gas-cooled reactor ...

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

    SciTech Connect (OSTI)

    Broczkowski, Michael E.; Noel, Jamie J.; Shoesmith, David W.

    2007-07-01

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

  12. Emergent Properties of the Bose-Einstein-Hubbard Condensate in...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Emergent Properties of the Bose-Einstein-Hubbard Condensate in UO2(+x) Citation Details In-Document Search Title: Emergent Properties of the Bose-Einstein-Hubbard ...

  13. Stanford Synchrotron Radiation Lightsource

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

    Nanoparticulate FeS as an Effective Redox Buffer to Prevent Uraninite (UO2) Oxidation August 2013 SSRL Science Summary by Manuel Gnida Figure A major concern in the nuclear age is...

  14. Office of Nuclear Energy Teachers' Edition Doe...

    Energy Savers [EERE]

    ... 2 Water Graphite Fast Neutron Breeder Reactor (FBR) Japan, France, Russia 2 1 PuO2 and UO 2 Liquid sodium None needed ... Source: IAEA June 2012, Nuclear Power Reactors in the World ...

  15. Directory

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

    Mechanical Behavior of UO2 at Sub-grain Length Scales: Quantification of Elastic, Plastic and Creep ... (Properties) 91813 9:46 AM 91813 9:46 AM Send Document Link...

  16. L

    Office of Legacy Management (LM)

    ... contract to the Manhattan &gineer District (MED) and the Atomic Energy Commission @EC). ... wnverting UO, to green salt (UFJ, operated during World War II and tbe following 2 years. ...

  17. DOE - Office of Legacy Management -- University of California...

    Office of Legacy Management (LM)

    Subject: List of California Sites; May 17, 1989 CA.05-3 - AEC Memorandum; Ball to Smith; Subject: 500 Pounds UO3 - SR-1952; July 10, 1951 CA.05-4 - AEC Memorandum; Blatzs to ...

  18. Uranium Downblending and Disposition Project Technology Readiness...

    Office of Environmental Management (EM)

    ... transfer high specific activity UO 3 as a dry, thermally hot solid, which is expected to ... for an electrostatic precipitator or a scrubber to allow the removal of these daughter ...

  19. 2013 Publications Resulting from the Use of NERSC Resources

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

    ... the Uranyl-SO2 Complex, UO2(CH3SO2)(SO2)-," J. Phys. Chem. ... Nuclear Instruments and Methods A683 (2012) 78-90, 11 August ... Cyclops Tensor Framework: reducing communication and ...

  20. Quasielastic neutron scattering with in situ humidity control: Water dynamics in uranyl fluoride

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

    Miskowiec, A.; Kirkegaard, M. C.; Herwig, K. W.; Trowbridge, L.; Mamontov, E.; Anderson, B.

    2016-03-04

    The authors confirm that water vapor pressure is the driving thermodynamic force for the conversion of the anhydrous structure to [(UO2F2)(H2O)]7 ? (H2O)4, and they demonstrate the feasibility of extending this approach to aqueous forms of UO2F2+ xH2O. This method has general applicability to systems in which water content itself is a driving variable for structural or dynamical phase transitions.

  1. PREPARATION OF SPHERICAL URANIUM DIOXIDE PARTICLES

    DOE Patents [OSTI]

    Levey, R.P. Jr.; Smith, A.E.

    1963-04-30

    This patent relates to the preparation of high-density, spherical UO/sub 2/ particles 80 to 150 microns in diameter. Sinterable UO/sub 2/ powder is wetted with 3 to 5 weight per cent water and tumbled for at least 48 hours. The resulting spherical particles are then sintered. The sintered particles are useful in dispersion-type fuel elements for nuclear reactors. (AEC)

  2. On the mechanical stability of uranyl peroxide hydrates: Implications for nuclear fuel degradation

    SciTech Connect (OSTI)

    Weck, Philippe F.; Kim, Eunja; Buck, Edgar C.

    2015-09-11

    The mechanical properties and stability of studtite, (UO2)(O2)(H2O)2·2H2O, and metastudtite, (UO2)(O2)(H2O)2, two important corrosion phases observed on spent nuclear fuel exposed to water, have been investigated using density functional perturbation theory. While (UO2)(O2)(H2O)2 satisfies the necessary and sufficient Born criteria for mechanical stability, (UO2)(O2)(H2O)2·2H2O is found to be mechanically metastable, which might be the underlying cause of the irreversibility of the studtite to metastudtite transformation. According to Pugh’s and Poisson’s ratios and the Cauchy pressure, both phases are considered ductile and shear modulus is the parameter limiting their mechanical stability. Debye temperatures of 294 and 271 K are predicted for polycrystalline (UO2)(O2)(H2O)2·2H2O and (UO2)(O2)(H2O)2, suggesting a lower micro-hardness of metastudtite.

  3. On the mechanical stability of uranyl peroxide hydrates: Implications for nuclear fuel degradation

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

    Weck, Philippe F.; Kim, Eunja; Buck, Edgar C.

    2015-09-11

    The mechanical properties and stability of studtite, (UO2)(O2)(H2O)2·2H2O, and metastudtite, (UO2)(O2)(H2O)2, two important corrosion phases observed on spent nuclear fuel exposed to water, have been investigated using density functional perturbation theory. While (UO2)(O2)(H2O)2 satisfies the necessary and sufficient Born criteria for mechanical stability, (UO2)(O2)(H2O)2·2H2O is found to be mechanically metastable, which might be the underlying cause of the irreversibility of the studtite to metastudtite transformation. According to Pugh's and Poisson's ratios and the Cauchy pressure, both phases are considered ductile and shear modulus is the parameter limiting their mechanical stability. Furthermore, debye temperatures of 294 and 271 K are predictedmore » for polycrystalline (UO2)(O2)(H2O)2·2H2O and (UO2)(O2)(H2O)2, suggesting a lower micro-hardness of metastudtite.« less

  4. Barium uranyl diphosphonates

    SciTech Connect (OSTI)

    Nelson, Anna-Gay D.; Alekseev, Evgeny V.; Ewing, Rodney C.; Albrecht-Schmitt, Thomas E.

    2012-08-15

    Three Ba{sup 2+}/UO{sub 2}{sup 2+} methylenediphosphonates have been prepared from mild hydrothermal treatment of uranium trioxide, methylendiphosphonic acid (C1P2) with barium hydroxide octahydrate, barium iodate monohydrate, and small aliquots of HF at 200 Degree-Sign C. These compounds, Ba[UO{sub 2}[CH{sub 2}(PO{sub 3}){sub 2}]{center_dot}1.4H{sub 2}O (Ba-1), Ba{sub 3}[(UO{sub 2}){sub 4}(CH{sub 2}(PO{sub 3}){sub 2}){sub 2}F{sub 6}]{center_dot}6H{sub 2}O (Ba-2), and Ba{sub 2}[(UO{sub 2}){sub 2}(CH{sub 2}(PO{sub 3}){sub 2})F{sub 4}]{center_dot}5.75H{sub 2}O (Ba-3) all adopt layered structures based upon linear uranyl groups and disphosphonate molecules. Ba-2 and Ba-3 are similar in that they both have UO{sub 5}F{sub 2} pentagonal bipyramids that are bridged and chelated by the diphosphonate moiety into a two-dimensional zigzag anionic sheet (Ba-2) and a one-dimensional ribbon anionic chain (Ba-3). Ba-1, has a single crystallographically unique uranium metal center where the C1P2 ligand solely bridges to form [UO{sub 2}[CH{sub 2}(PO{sub 3}){sub 2}]{sup 2-} sheets. The interlayer space of the structures is occupied by Ba{sup 2+}, which, along with the fluoride ion, mediates the structure formed and maintains overall charge balance. - Graphical abstract: Illustration of the stacking of the layers in Ba{sub 3}[(UO{sub 2}){sub 4}(CH{sub 2}(PO{sub 3}){sub 2}){sub 2})F{sub 6}]{center_dot}6H{sub 2}O viewed along the c-axis. The structure is constructed from UO{sub 7} pentagonal bipyramidal units, U(1)O{sub 7}=gray, U(2)O{sub 7}=yellow, barium=blue, phosphorus=magenta, fluorine=green, oxygen=red, carbon=black, and hydrogen=light peach. Highlights: Black-Right-Pointing-Pointer The polymerization of the UO{sub 2}{sup 2+} sites to form uranyl dimers leads to structural variations in compounds. Black-Right-Pointing-Pointer Barium cations stitch uranyl diphosphonate anionic layers together, and help mediate structure formation. Black-Right-Pointing-Pointer HF acts as both a

  5. The burnup dependence of light water reactor spent fuel oxidation

    SciTech Connect (OSTI)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies

  6. EFRC CMSNF Major Accomplishments

    SciTech Connect (OSTI)

    D. Hurley; Todd R. Allen

    2014-09-01

    The mission of the Center for Material Science of Nuclear Fuels (CMSNF) has been to develop a first-principles-based understanding of thermal transport in the most widely used nuclear fuel, UO2, in the presence of defect microstructure associated with radiation environments. The overarching goal within this mission was to develop an experimentally validated multiscale modeling capability directed toward a predictive understanding of the impact of radiation and fission-product induced defects and microstructure on thermal transport in nuclear fuel. Implementation of the mission was accomplished by integrating the physics of thermal transport in crystalline solids with microstructure science under irradiation through multi institutional experimental and computational materials theory teams from Idaho National Laboratory, Oak Ridge National Laboratory, Purdue University, the University of Florida, the University of Wisconsin, and the Colorado School of Mines. The Centers research focused on five major areas: (i) The fundamental aspects of anharmonicity in UO2 crystals and its impact on thermal transport; (ii) The effects of radiation microstructure on thermal transport in UO2; (iii) The mechanisms of defect clustering in UO2 under irradiation; (iv) The effect of temperature and oxygen environment on the stoichiometry of UO2; and (v) The mechanisms of growth of dislocation loops and voids under irradiation. The Center has made important progress in each of these areas, as summarized below.

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

    SciTech Connect (OSTI)

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

    1997-02-01

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

  8. Characterization of urania vaporization with transpiration coupled thermogravimetry

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

    McMurray, J. W.

    2015-12-05

    Determining equilibrium vapor pressures of materials is made easier by transpiration measurements. However, the traditional technique involves condensing the volatiles entrained in a carrier gas outside of the hot measurement zone. One potential problem is deposition en route to a cooled collector. Thermogravimetric analysis (TGA) can be used to measure in situ mass loss due to vaporization and therefore obviate the need to analyze the entire gas train due to premature plating of vapor species. Therefore, a transpiration coupled TGA technique was used to determine equilibrium pressures of UO3 gas over fluorite structure UO2+x and U3O8 at T = (1573more » and 1773) K. Moreover, we compared to calculations from models and databases in the open literature. Our study gives clarity to the thermochemical data for UO3 gas and validates the mass loss transpiration method using thermogravimetry for determining equilibrium vapor pressures of non-stoichiometric oxides.« less

  9. Characterization of urania vaporization with transpiration coupled thermogravimetry

    SciTech Connect (OSTI)

    McMurray, J. W.

    2015-12-05

    Determining equilibrium vapor pressures of materials is made easier by transpiration measurements. However, the traditional technique involves condensing the volatiles entrained in a carrier gas outside of the hot measurement zone. One potential problem is deposition en route to a cooled collector. Thermogravimetric analysis (TGA) can be used to measure in situ mass loss due to vaporization and therefore obviate the need to analyze the entire gas train due to premature plating of vapor species. Therefore, a transpiration coupled TGA technique was used to determine equilibrium pressures of UO3 gas over fluorite structure UO2+x and U3O8 at T = (1573 and 1773) K. Moreover, we compared to calculations from models and databases in the open literature. Our study gives clarity to the thermochemical data for UO3 gas and validates the mass loss transpiration method using thermogravimetry for determining equilibrium vapor pressures of non-stoichiometric oxides.

  10. Calculation of the thermodynamic properties of fuel-vapor species from spectroscopic data

    SciTech Connect (OSTI)

    Green, D.W.

    1980-09-01

    Measured spectroscopic data, estimated molecular parameters, and a densty-of-states model for electronic structure have been used to calculate thermodynamic functions for gaseous ThO, ThO/sub 2/, UO, UO/sub 2/, UO/sub 3/, PuO, and PuO/sub 2/. Various methods for estimating parameters have been considered and numerically evaluated. The sensitivity of the calculated thermodynamic functions to molecular parameters has been examined quantitatively. New values of the standard enthalpies of formation at 298.15/sup 0/K have been derived from the best available ..delta..G/sup 0//sub f/ equations and the calculated thermodynamic functions. Estimates of the uncertainties have been made for measured and estimated data as well as for various mathematical and physical approximations. Tables of the thermodynamic functions to 6000/sup 0/K are recommended for gaseous thorium, uranium, and plutonium oxides.

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

    SciTech Connect (OSTI)

    Tsai, C.

    1981-11-01

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

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

    SciTech Connect (OSTI)

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

    2014-07-28

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

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

    SciTech Connect (OSTI)

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

    2014-01-01

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

  14. On the mechanical stability of uranyl peroxide hydrates: Implications for nuclear fuel degradation

    SciTech Connect (OSTI)

    Weck, Philippe F.; Kim, Eunja; Buck, Edgar C.

    2015-09-11

    The mechanical properties and stability of studtite, (UO2)(O2)(H2O)2·2H2O, and metastudtite, (UO2)(O2)(H2O)2, two important corrosion phases observed on spent nuclear fuel exposed to water, have been investigated using density functional perturbation theory. While (UO2)(O2)(H2O)2 satisfies the necessary and sufficient Born criteria for mechanical stability, (UO2)(O2)(H2O)2·2H2O is found to be mechanically metastable, which might be the underlying cause of the irreversibility of the studtite to metastudtite transformation. According to Pugh's and Poisson's ratios and the Cauchy pressure, both phases are considered ductile and shear modulus is the parameter limiting their mechanical stability. Furthermore, debye temperatures of 294 and 271 K are predicted for polycrystalline (UO2)(O2)(H2O)2·2H2O and (UO2)(O2)(H2O)2, suggesting a lower micro-hardness of metastudtite.

  15. Diffusion and Adsorption of Uranyl Carbonate Species in Nanosized Mineral Fractures

    SciTech Connect (OSTI)

    Kerisit, Sebastien N.; Liu, Chongxuan

    2012-02-07

    Atomistic simulations were performed to study the diffusion and adsorption of Ca{sub 2}UO{sub 2}(CO{sub 3}){sub 3} and of some of its constituent species, i.e., UO{sub 2}{sup 2+}, CO{sub 3}{sup 2-}, and UO{sub 2}CO{sub 3}, in feldspar nano-sized fractures. Feldspar is important to uranium remediation efforts at the U.S. Department of Energy Hanford site as it has been found in recent studies to host contaminants within its intragrain fractures. In addition, uranyl carbonate species are known to dominate U(VI) speciation in conditions relevant to the Hanford site. Molecular dynamics (MD) simulations showed that the presence of the feldspar surface diminishes the diffusion coefficients of all the species considered in this work and that the diffusion coefficients do not reach their bulk aqueous solution values in the center of a 2.5 nm fracture. Moreover, the MD simulations showed that the rate of decrease in the diffusion coefficients with decreasing distance from the surface is greater for larger adsorbing species. Free energy profiles of the same species adsorbing on the feldspar surface revealed a large exothermic free energy of adsorption for UO{sub 2}{sup 2+} and UO{sub 2}CO{sub 3}, which are able to adsorb to the surface with their uranium atom directly bonded to a surface hydroxyl oxygen, whereas adsorption of CO{sub 3}{sup 2-} and Ca{sub 2}UO{sub 2}(CO{sub 3}){sub 3}, which attach to the surface via hydrogen bonding from a surface hydroxyl group to a carbonate oxygen, was calculated to be either only slightly exothermic or endothermic.

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

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth

    2002-09-01

    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.

  17. Oxidation and crystal field effects in uranium

    SciTech Connect (OSTI)

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

    2015-07-06

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

  18. Effects of Time, Heat, and Oxygen on K Basin Sludge Agglomeration, Strength, and Solids Volume

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Sinkov, Sergey I.; Schmidt, Andrew J.; Daniel, Richard C.; Burns, Carolyn A.

    2011-01-04

    Sludge disposition will be managed in two phases under the K Basin Sludge Treatment Project. The first phase is to retrieve the sludge that currently resides in engineered containers in the K West (KW) Basin pool at ~10 to 18°C. The second phase is to retrieve the sludge from interim storage in the sludge transport and storage containers (STSCs) and treat and package it in preparation for eventual shipment to the Waste Isolation Pilot Plant. The work described in this report was conducted to gain insight into how sludge may change during long-term containerized storage in the STSCs. To accelerate potential physical and chemical changes, the tests were performed at temperatures and oxygen partial pressures significantly greater than those expected in the T Plant canyon cells where the STSCs will be stored. Tests were conducted to determine the effects of 50°C oxygenated water exposure on settled quiescent uraninite (UO2) slurry and a full simulant of KW containerized sludge to determine the effects of oxygen and heat on the composition and mechanical properties of sludge. Shear-strength measurements by vane rheometry also were conducted for UO2 slurry, mixtures of UO2 and metaschoepite (UO3•2H2O), and for simulated KW containerized sludge. The results from these tests and related previous tests are compared to determine whether the settled solids in the K Basin sludge materials change in volume because of oxidation of UO2 by dissolved atmospheric oxygen to form metaschoepite. The test results also are compared to determine if heating or other factors alter sludge volumes and to determine the effects of sludge composition and settling times on sludge shear strength. It has been estimated that the sludge volume will increase with time because of a uranium metal → uraninite → metaschoepite oxidation sequence. This increase could increase the number of containers required for storage and increase overall costs of sludge management activities. However, the volume

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

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

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

    2016-04-09

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

  20. Measuring the Noble Metal and Iodine Composition of Extracted Noble Metal Phase from Spent Nuclear Fuel Using Instrumental Neutron Activation Analysis

    SciTech Connect (OSTI)

    Palomares, R. I.; Dayman, Kenneth J.; Landsberger, Sheldon; Biegalski, Steven R.; Soderquist, Chuck Z.; Casella, Amanda J.; Brady Raap, Michaele C.; Schwantes, Jon M.

    2015-04-01

    Mass quantities of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis (NAA). Nuclide presence is predicted using fission yield analysis, and mass quantification is derived from standard gamma spectroscopy and radionuclide decay analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO2 fuel dissolved in nitric acid and UO2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared. Lastly, the implications of the rapid analytic speed of instrumental NAA are discussed in relation to potential nuclear forensics applications.

  1. Analysis of the thorium axial blanket experiments in the PROTEUS reactor

    SciTech Connect (OSTI)

    White, J. R.; Ingersoll, D. T.; Schmocker, U.

    1980-01-01

    An extensive program of reactor physics experiments in GCFR fuel pin lattices has been completed recently at the PROTEUS critical facility located at EIR laboratory in Switzerland. The PROTEUS reactor consists of a central test zone surrounded by a uranium buffer and thermal driver region. The test lattices included a PuO/sub 2//UO/sub 2/ fuel region with internal and axial blankets of UO/sub 2/, ThO/sub 2/, and thorium metal. Detailed analysis of the thorium-bearing lattices has been performed at EIR and at ORNL in order to validate nuclear data and methods used for reactor physics analysis of advanced GCFR designs.

  2. Time-Resolved Infrared Reflectance Studies of the Dehydration-Induced Transformation of Uranyl Nitrate Hexahydrate to the Trihydrate Form

    SciTech Connect (OSTI)

    Johnson, Timothy J.; Sweet, Lucas E.; Meier, David E.; Mausolf, Edward J.; Kim, Eunja; Weck, Philippe F.; Buck, Edgar C.; McNamara, Bruce K.

    2015-10-01

    Uranyl nitrate is a key species in the nuclear fuel cycle. However, this species is known to exist in different states of hydration, including the hexahydrate ([UO2(NO3)2(H2O)6] often called UNH), the trihydrate [UO2(NO3)2(H2O)3 or UNT], and in very dry environments the dihydrate form [UO2(NO3)2(H2O)2]. Their relative stabilities depend on both water vapor pressure and temperature. In the 1950s and 1960s the different phases were studied by infrared transmission spectroscopy, but were limited both by instrumental resolution and by the ability to prepare the samples for transmission. We have revisited this problem using time-resolved reflectance spectroscopy, which requires no sample preparation and allows dynamic analysis while the sample is exposed to a flow of N2 gas. Samples of known hydration state were prepared and confirmed via X-ray diffraction patterns of known species. In reflectance mode the hexahydrate UO2(NO3)2(H2O)6 has a distinct uranyl asymmetric stretch band at 949.0 cm-1 that shifts to shorter wavelengths and broadens as the sample desiccates and recrystallizes to the trihydrate, first as a shoulder growing in on the blue edge but ultimately results in a doublet band with reflectance peaks at 966 and 957 cm-1. The data are consistent with transformation from UNH to UNT as UNT has two inequivalent UO22+ sites. The dehydration of UO2(NO3)2(H2O)6 to UO2(NO3)2(H2O)3 is both a structural and morphological change that has the lustrous lime green UO2(NO3)2(H2O)6 crystals changing to the matte greenish yellow of the trihydrate solid. The phase transformation and crystal structures were confirmed by density functional theory calculations and optical microscopy methods, both of which showed a transformation with two distinct sites for the uranyl cation in the trihydrate, with but one in the hexahydrate.

  3. Time-resolved infrared reflectance studies of the dehydration-induced transformation of uranyl nitrate hexahydrate to the trihydrate form

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

    Johnson, Timothy J.; Sweet, Lucas E.; Meier, David E.; Edward J. Mausolf; Kim, Eunja; Weck, Philippe F.; Buck, Edgar C.; Bruce K. McNamara

    2015-09-08

    Uranyl nitrate is a key species in the nuclear fuel cycle. However, this species is known to exist in different states of hydration, including the hexahydrate ([UO2(NO3)2(H2O)6] often called UNH), the trihydrate [UO2(NO3)2(H2O)3 or UNT], and in very dry environments the dihydrate form [UO2(NO3)2(H2O)2]. Their relative stabilities depend on both water vapor pressure and temperature. In the 1950s and 1960s, the different phases were studied by infrared transmission spectroscopy but were limited both by instrumental resolution and by the ability to prepare the samples for transmission. We have revisited this problem using time-resolved reflectance spectroscopy, which requires no sample preparationmore » and allows dynamic analysis while the sample is exposed to a flow of N2 gas. Samples of known hydration state were prepared and confirmed via X-ray diffraction patterns of known species. In reflectance mode the hexahydrate UO2(NO3)2(H2O)6 has a distinct uranyl asymmetric stretch band at 949.0 cm–1 that shifts to shorter wavelengths and broadens as the sample desiccates and recrystallizes to the trihydrate, first as a shoulder growing in on the blue edge but ultimately results in a doublet band with reflectance peaks at 966 and 957 cm–1. The data are consistent with transformation from UNH to UNT as UNT has two inequivalent UO22+ sites. The dehydration of UO2(NO3)2(H2O)6 to UO2(NO3)2(H2O)3 is both a structural and morphological change that has the lustrous lime green UO2(NO3)2(H2O)6 crystals changing to the matte greenish yellow of the trihydrate solid. As a result, the phase transformation and crystal structures were confirmed by density functional theory calculations and optical microscopy methods, both of which showed a transformation with two distinct sites for the uranyl cation in the trihydrate, with only one in the hexahydrate.« less

  4. Variation of stability constants of thorium and uranium oxalate complexes with ionic strength

    SciTech Connect (OSTI)

    Erten, H.N; Mohammed, A.K.; Choppin, G.R.

    1993-12-31

    Extraction of Th(IV) and UO{sub 2}{sup 2+} by a solution of TTA and HDEHP, respectively, in toluene was used to obtain stability constants of their oxalate complexes in 1, 3, 5, 7 and 9 M ionic strength (NaClO{sub 2}) solutions. The complexes formed were the MOx, MHOx, MOx{sub 2} and M(HOx){sub 2} (M = Th, UO{sub 2}) species. The values were analyzed by the Specific Interaction Theory and agreed to I {le} 3 M but required an additional term for fitting at I > 3 M.

  5. Confirmation of shutdown cooling effects

    SciTech Connect (OSTI)

    Sato, Kotaro Tabuchi, Masato; Sugimura, Naoki; Tatsumi, Masahiro

    2015-12-31

    After the Fukushima accidents, all nuclear power plants in Japan have gradually stopped their operations and have long periods of shutdown. During those periods, reactivity of fuels continues to change significantly especially for high-burnup UO{sub 2} fuels and MOX fuels due to radioactive decays. It is necessary to consider these isotopic changes precisely, to predict neutronics characteristics accurately. In this paper, shutdown cooling (SDC) effects of UO{sub 2} and MOX fuels that have unusual operation histories are confirmed by the advanced lattice code, AEGIS. The calculation results show that the effects need to be considered even after nuclear power plants come back to normal operation.

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

    SciTech Connect (OSTI)

    Gofryk, K.; Jaime, M.

    2014-12-01

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

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

    SciTech Connect (OSTI)

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

    2015-03-05

    In this study, oxygen transport is central to many properties of oxides such as stoichiometric changes, phase transformation and ionic conductivity. In this paper, we report a mechanism for oxygen transport in uranium dioxide (UO2) in which the kinetics is mediated by defect clustering dynamics. In particular, the kinetic Monte Carlo (KMC) method has been used to investigate the kinetics of oxygen transport in UO2 under the condition of creation and annihilation of oxygen vacancies and interstitials as well as oxygen interstitial clustering, with variable offstoichiometry and temperature conditions. It is found that in hypo-stoichiometric UO2-x, oxygen transport is well described by the vacancy diffusion mechanism while in hyper-stoichiometric UO2+x, oxygen interstitial cluster diffusion contributes significantly to oxygen transport kinetics, particularly at high temperatures and high off-stoichiometry levels. It is also found that diinterstitial clusters and single interstitials play dominant roles in oxygen diffusion while other larger clusters have negligible contributions. However, the formation, coalescence and dissociation of these larger clusters indirectly affects the overall oxygen diffusion due to their interactions with mono and di-interstitials, thus providing a explanation of the experimental observation of saturation or even drop of oxygen diffusivity at high off-stoichiometry.

  8. URANIUM SEPARATION PROCESS

    DOE Patents [OSTI]

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

    1959-03-10

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

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

    SciTech Connect (OSTI)

    Lewis, K.D.

    1994-09-01

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

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

    SciTech Connect (OSTI)

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

    2015-03-07

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

  11. Thermodynamics of fission products in dispersion fuel designs - first principles modeling of defect behavior in bulk and at interfaces

    SciTech Connect (OSTI)

    Liu, Xiang-yand; Uberuaga, Blas P; Nerikar, Pankaj; Sickafus, Kurt E; Stanek, Chris R

    2009-01-01

    Density functional theory (DFT) calculations of fission product (Xe, Sr, and Cs) incorporation and segregation in alkaline earth metal oxides, HfO{sub 2} and UO{sub 2} oxides, and the MgO/(U, Hf, Ce)O{sub 2} interfaces have been carried out. In the case of UO{sub 2}, the calculations were performed using spin polarization and with a Hubbard U term characterizing the on-sit Coulomb repulsion between the localized 5f electrons. The fission product solution energies in bulk UO{sub 2{+-}x} have been calculated as a function of non-stoichiometry x, and were compared to that in MgO. These calculations demonstrate that the fission product incorporation energies in MgO are higher than in HfO{sub 2}. However, this trend is reversed or reduced for alkaline earth oxides with larger cation sizes. The solution energies of fission products in MgO are substantially higher than in UO{sub 2{+-}x}, except for the case of Sr in the hypostoichiometric case. Due to size effects, the thermodynamic driving force of segregation for Xe and Cs from bulk MgO to the MgO/fluorite interface is strong. However, this driving force is relatively weak for Sr.

  12. Developing a High Thermal Conductivity Fuel with Silicon Carbide Additives

    SciTech Connect (OSTI)

    baney, Ronald; Tulenko, James

    2012-11-20

    The objective of this research is to increase the thermal conductivity of uranium oxide (UO{sub 2}) without significantly impacting its neutronic properties. The concept is to incorporate another high thermal conductivity material, silicon carbide (SiC), in the form of whiskers or from nanoparticles of SiC and a SiC polymeric precursor into UO{sub 2}. This is expected to form a percolation pathway lattice for conductive heat transfer out of the fuel pellet. The thermal conductivity of SiC would control the overall fuel pellet thermal conductivity. The challenge is to show the effectiveness of a low temperature sintering process, because of a UO{sub 2}-SiC reaction at 1,377°C, a temperature far below the normal sintering temperature. Researchers will study three strategies to overcome the processing difficulties associated with pore clogging and the chemical reaction of SiC and UO{sub 2} at temperatures above 1,300°C:

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

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

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

    2015-03-05

    In this study, oxygen transport is central to many properties of oxides such as stoichiometric changes, phase transformation and ionic conductivity. In this paper, we report a mechanism for oxygen transport in uranium dioxide (UO2) in which the kinetics is mediated by defect clustering dynamics. In particular, the kinetic Monte Carlo (KMC) method has been used to investigate the kinetics of oxygen transport in UO2 under the condition of creation and annihilation of oxygen vacancies and interstitials as well as oxygen interstitial clustering, with variable offstoichiometry and temperature conditions. It is found that in hypo-stoichiometric UO2-x, oxygen transport is wellmore » described by the vacancy diffusion mechanism while in hyper-stoichiometric UO2+x, oxygen interstitial cluster diffusion contributes significantly to oxygen transport kinetics, particularly at high temperatures and high off-stoichiometry levels. It is also found that diinterstitial clusters and single interstitials play dominant roles in oxygen diffusion while other larger clusters have negligible contributions. However, the formation, coalescence and dissociation of these larger clusters indirectly affects the overall oxygen diffusion due to their interactions with mono and di-interstitials, thus providing a explanation of the experimental observation of saturation or even drop of oxygen diffusivity at high off-stoichiometry.« less

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

    SciTech Connect (OSTI)

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

    2008-07-08

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

  15. Pyrolytic carbon-coated nuclear fuel

    DOE Patents [OSTI]

    Lindemer, Terrence B.; Long, Jr., Ernest L.; Beatty, Ronald L.

    1978-01-01

    An improved nuclear fuel kernel having at least one pyrolytic carbon coating and a silicon carbon layer is provided in which extensive interaction of fission product lanthanides with the silicon carbon layer is avoided by providing sufficient UO.sub.2 to maintain the lanthanides as oxides during in-reactor use of said fuel.

  16. Hydrofluoric Acid Corrosion Testing on Unplated and Electroless Gold-Plated Samples

    SciTech Connect (OSTI)

    Osborne, P.E.; Icenhour, A.S.; Del Cul, G.D.

    2000-08-01

    The Molten Salt Reactor Experiment (MSRE) remediation requires that almost 40 kg of uranium hexafluoride (UF6) be converted to uranium oxide (UO). In the process of this conversion, six moles of hydrofluoric acid (HP) are produced for each mole of UF6 converted.

  17. CX-011566: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Mechanical Behavior of Uranium Oxide (UO2) at Sub-grain Length Scales: Quantification of Elastic, Plastic and Creep Properties via Microscale Testing CX(s) Applied: B3.6 Date: 11/18/2013 Location(s): Arizona Offices(s): Idaho Operations Office

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

    SciTech Connect (OSTI)

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

    2014-11-01

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

  19. The Development of Models to Optimize Selection of Nuclear Fuels through Atomic-Level Simulation

    SciTech Connect (OSTI)

    Prof. Simon Phillpot; Prof. Susan B. Sinnott; Prof. Hans Seifert; Prog. James Tulenko

    2009-01-26

    Demonstrated that FRAPCON can be modified to accept data generated from first principles studies, and that the result obtained from the modified FRAPCON make sense in terms of the inputs. Determined the temperature dependence of the thermal conductivity of single crystal UO2 from atomistic simulation.

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

    SciTech Connect (OSTI)

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

    2008-07-15

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

  1. Water-Moderated and -Reflected Slabs of Uranium Oxyfluoride

    SciTech Connect (OSTI)

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

    2010-09-01

    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.

  2. I Mr. J. C. Delaney

    Office of Legacy Management (LM)

    ... F?UO Product is &k;e.i,j is &,,pl&' fof te"&d:,jth&r charged to the blender. A dry ... The same fume scrubber system also is connected to a hood over the U02,powder dissolver to ...

  3. Synthesis and structures of new uranyl malonate complexes with carbamide derivatives

    SciTech Connect (OSTI)

    Serezhkina, L. B.; Grigor’ev, M. S.; Medvedkov, Ya. A.; Serezhkin, V. N.

    2015-09-15

    Crystals of new malonate-containing uranyl complexes [UO{sub 2}(C{sub 3}H{sub 2}O{sub 4})(Imon)(H{sub 2}O)] (I) and [UO{sub 2}(C{sub 3}H{sub 2}O4)(Meur){sub 3}] (II), where Imon is imidazolidin-2-one (ethylenecarbamide) and Meur is methyl-carbamide, have been synthesized and studied by X-ray diffraction. Both compounds crystallize in the monoclinic system with the following unit-cell parameters (at 100 K): a = 11.1147(10) Å, b = 6.9900(6) Å, c = 14.4934(12) Å, β = 92.042(2)°, V = 1125.30(17) Å{sup 3}, sp. gr. P2{sub 1}/n, Z = 4, R{sub 1} = 0.0398 (I); a = 16.6613(5) Å, b = 9.5635(3) Å, c = 22.9773(6) Å, β = 103.669(2)°, V = 3557.51(18) Å{sup 3}, sp. gr. C2/c, Z = 8, R{sub 1} = 0.0207 (II). The crystals are composed of electroneutral chains [UO{sub 2}(C{sub 3}H{sub 2}O{sub 4})(Imon)(H{sub 2}O)] and mononuclear groups [UO{sub 2}(C{sub 3}H{sub 2}O{sub 4})(Meur){sub 3}] as the structural units belonging to the crystal-chemical groups AT{sup 11}M{sub 2}{sup 1} and AB{sup 01}M{sub 3}{sup 1} (A =UO{sub 2}{sup 2+}, T{sup 11} and B{sup 01} = C{sub 3}H{sub 2}, M{sup 1} = Imon, H{sub 2}O, or Meur), respectively, of uranyl complexes. The packing modes of the uranyl-containing complexes were analyzed by the method of molecular Voronoi—Dirichlet polyhedra.

  4. New three-dimensional inorganic frameworks based on the uranophane-type sheet in monoamine templated uranyl-vanadates

    SciTech Connect (OSTI)

    Jouffret, Laurent; Shao Zhenmian

    2010-10-15

    Seven new uranyl vanadates with mono-protonated amine or tetramethylammonium used as structure directing cations, (C{sub 2}NH{sub 8}){sub 2{l_brace}}[(UO{sub 2})(H{sub 2}O)][(UO{sub 2})(VO{sub 4})]{sub 4{r_brace}}.H{sub 2}O (DMetU5V4) (C{sub 2}NH{sub 8}){l_brace}[(UO{sub 2})(H{sub 2}O){sub 2}][(UO{sub 2})(VO{sub 4})]{sub 3{r_brace}}.H{sub 2}O (DMetU4V3), (C{sub 5}NH{sub 6}){sub 2{l_brace}}[(UO{sub 2})(H{sub 2}O)][(UO{sub 2})(VO{sub 4})]{sub 4{r_brace}}.H{sub 2}O (PyrU5V4), (C{sub 3}NH{sub 10}){l_brace}[(UO{sub 2})(H{sub 2}O){sub 2}][(UO{sub 2})(VO{sub 4})]{sub 3{r_brace}}.H{sub 2}O (isoPrU4V3), (N(CH{sub 3}){sub 4}){l_brace}[(UO{sub 2})(H{sub 2}O){sub 2}][(UO{sub 2})(VO{sub 4})]{sub 3{r_brace}}.H{sub 2}O (TMetU4V3), (C{sub 6}NH{sub 14}){l_brace}[(UO{sub 2})(H{sub 2}O){sub 2}][(UO{sub 2})(VO{sub 4})]{sub 3{r_brace}}.H{sub 2}O (CHexU4V3), and (C{sub 4}NH{sub 12}){l_brace}[(UO{sub 2})(H{sub 2}O)][(UO{sub 2})(VO{sub 4})]{sub 3{r_brace}} (TButU4V3) were prepared from mild-hydrothermal reactions using dimethylamine, pyridine, isopropylamine, tetramethylammonium hydroxide, cyclohexylamine and tertiobutylamine, respectively, with uranyl nitrate and vanadium oxide in acidic medium. The structures were solved using single-crystal X-ray diffraction data. The compounds exhibit three-dimensional uranyl-vanadate inorganic frameworks built from uranophane-type uranyl-vanadate layers pillared by uranyl polyhedra with cavities in between occupied by protonated organic moieties. In the uranyl-vanadate layers the orientations of the vanadate tetrahedra give new geometrical isomers leading to unprecedented pillared systems and new inorganic frameworks with U/V=4/3. Crystallographic data: (DMetU5V4) orthorhombic, Cmc2{sub 1} space group, a=15.6276(4), b=14.1341(4), c=13.6040(4) A; (DMetU4V3) monoclinic, P2{sub 1}/n space group, a=10.2312(4), b=13.5661(7), c=17.5291(7) A, {beta}=96.966(2); (PyrU5V4), triclinic, P1 space group, a=9.6981(3), b=9.9966(2), c=10.5523(2) A, {alpha}=117

  5. Analysis of key safety metrics of thorium utilization in LWRs

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

    Ade, Brian J.; Bowman, Stephen M.; Worrall, Andrew; Powers, Jeffrey

    2016-04-08

    Here, thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel (233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on conceptsmore » that mix thorium with uranium (UO2 + ThO2) or that add fertile thorium (ThO2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO2 + ThO2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO2 + UO2 and ThO2 + PuO2 against those of UO2 and typical UO2 + PuO2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of

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

    SciTech Connect (OSTI)

    Zickafoose, M.S.

    1997-11-01

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

  7. Fuels for research and test reactors, status review: July 1982

    SciTech Connect (OSTI)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO/sub 2/ rod fuels. Among new fuels, those given major emphasis include H/sub 3/Si-Al dispersion and UO/sub 2/ caramel plate fuels.

  8. The behavior of measured SEU at low altitude during periods of high solar activity

    SciTech Connect (OSTI)

    Harboe-Sorensen, R.; Daly, E.J.; Adams, L. ); Underwood, C.I.; Ward, J. )

    1990-12-01

    The UoSAT-2 spacecraft, launched in 1984 into a polar orbit of altitude 700 km has a number of systems which have been observed to experience single-event upsets at significant rates. Geographically, the upsets are strongly concentrated in the South-Atlantic Anomaly region from which it has been deduced that in this region they are due to the products of proton-induced nuclear reactions in the devices. During the year 1989, several solar flare events occurred which elevated the upset rates at high latitudes. The October 19 event, in particular, resulted in very high high-latitude upset rates. The authors separate and analyze these data, deriving upset rates for the various memory devices under quiet cosmic-ray, South Atlantic anomaly and solar flare conditions. The authors report on the results of the heavy ion and proton testing of UoSAT memories which were undertaken in order to compare predictions and observations.

  9. Feasibility Study of MOX Fuel Online Burnup Analysis

    SciTech Connect (OSTI)

    Dennis, M.L.; Usman, S.

    2006-07-01

    This research is an extension of well established Non-Destructive Analysis of UO fuel using gamma spectroscopy of Cs-137 and other related isotopes. Given the performance similarities between UO fuel and MOX fuel, investigations are underway to develop similar correlation for MOX. MOX fuel burnup and decay simulations are being performed using ORIGEN-ARP (Oak Ridge Isotope Generation and Depletion Code - Automatic Rapid Processing). Simulation results are being analyzed and will be used to determine performance specifications of a detection system for field applications. Analysis of isotopic activity from irradiated fuel will be used to develop correlations to determine burn-up and Plutonium content of MOX fuel. These results will be particularly useful in view of the recent interest in MOX fuel. (authors)

  10. A comment on the thermal conductivity of (U,Pu)O2 and (U,Th)O2 by molecular dynamics with adjustment for phonon-spin scattering

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

    Cooper, Michael William D.; Liu, Xiang -Yang; Stanek, Christopher Richard; Andersson, David Anders

    2016-07-15

    In this study, a new approach for adjusting molecular dynamics results on UO2 thermal conductivity to include phonon-spin scattering has been used to improve calculations on Ux Pu1–x O2 and UxTh1xO2. We demonstrate that by including spin scattering a strong asymmetry as a function of uranium actinide fraction, x, is obtained. Greater degradation is shown for UxTh1–xO2 than UxPu1-xO2. Minimum thermal conductivities are predicted at U0.97Pu0.03O2 and U0.58Th0.42O2, although the degradation in UxPu1–xO2 is negligible relative to pure UO2.