National Library of Energy BETA

Sample records for ms uo uo

  1. 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.

  2. Microstructure changes and thermal conductivity reduction in UO2 following

    Office of Scientific and Technical Information (OSTI)

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

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

    Office of Scientific and Technical Information (OSTI)

    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 ...

  4. 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.

  5. 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

  6. Oxidative Dissolution of UO2 in a Simulated Groundwater Containing

    Office of Scientific and Technical Information (OSTI)

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

  7. Oxidative Dissolution of UO2 in a Simulated Groundwater Containing

    Office of Scientific and Technical Information (OSTI)

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

  8. 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.

  9. 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.

  10. 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)

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

    SciTech Connect (OSTI)

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

    2009-11-01

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

  12. Density Functional Theory Calculations of Mass Transport in UO2

    SciTech Connect (OSTI)

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

    2012-06-26

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

  13. 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/.

  14. 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.

  15. 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).

  16. 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

  17. 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.

  18. 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.

  19. 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)

  20. PUREX/UO3 Facilities deactivation lessons learned history

    SciTech Connect (OSTI)

    Gerber, M.S.

    1996-09-19

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

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

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

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

    2015-01-01

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

  2. 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

  3. 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.

  4. 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)

  5. 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.

  6. 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.

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

    Office of Scientific and Technical Information (OSTI)

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

  8. 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.

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

    Office of Scientific and Technical Information (OSTI)

    Minimization (Journal Article) | SciTech Connect A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide Waste Minimization Citation Details In-Document Search Title: A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide Waste Minimization A new COmbined NonFertile and Uranium (CONFU) fuel assembly is proposed to limit the actinides that need long-term high-level waste storage from the pressurized water reactor (PWR) fuel cycle. In the CONFU assembly concept,

  10. 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.

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

  12. 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.

  13. 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.

  14. 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.

  15. $sup 18$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.

  16. 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.

  17. 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].

  18. 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].

  19. 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.

  20. 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)

  1. 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.

  2. 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)

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

    SciTech Connect (OSTI)

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

    2007-06-01

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

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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)

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

    SciTech Connect (OSTI)

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

    2014-10-31

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

  11. 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}.

  12. 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.

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

    SciTech Connect (OSTI)

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

    2002-06-15

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

  14. 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 2800C in vacuum and about 750C 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 UO2CNT composites without affecting the neutronic property of UO2 significantly. The concept of this goal is to utilize a rapid sintering method (515 min) called spark plasma sintering (SPS) in which a mixture of CNTs and UO2 powder are used to make composites with different volume fractions of CNTs. Incorporation of these nanoscale materials plays a fundamentally critical role in controlling the performance and stability of UO2 fuel. We will use a novel in situ growth process to grow CNTs on UO2 particles for rapid sintering and develop UO2-CNT composites. This method is expected to provide a uniform distribution of CNTs at various volume fractions so that a high thermally conductive UO2-CNT composite is obtained with a minimal volume fraction of CNTs. The mixtures are sintered in the SPS facility at a range of temperatures, pressures, and time durations so as to identify the optimal processing conditions to obtain the desired microstructure of sintered UO2-CNT pellets. The second objective of the proposed work is to identify the optimal volume fraction of CNTs in the microstructure of the composites that provides the desired high thermal conductivity yet retaining the mechanical strength required for efficient function as a reactor fuel. We will systematically study the resulting microstructure (grain size, porosity, distribution of CNTs, etc.) obtained at various SPS processing conditions using optical microscopy, scanning electron microscopy (SEM), and transmission electron microscope (TEM). We will conduct indentation hardness measurements and uniaxial strength measurements as a function of volume fraction of CNTs to determine the mechanical strength and compare them to the properties of UO2. The fracture surfaces will be studied to determine the fracture characteristics that may relate to the observed cracking during service. Finally, we will perform thermal conductivity measurements on all the composites up to 1000 C. This study will relate the microstructure, mechanical properties, and thermal properties at various volume fractions of CNTs. The overall intent is to identify optimal processing conditions that will provide a well-consolidated compact with optimal microstructure and thermo-mechanical properties. The deliverables include: (1) fully characterized UO2-CNT composite with optimal CNT volume fraction and high thermal conductivity and (2) processing conditions for production of UO2-CNT composite pellets using SPS method.

  15. 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.

  16. 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)

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

    SciTech Connect (OSTI)

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

    2010-12-01

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

  18. 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.

  19. 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

  20. 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.

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

    SciTech Connect (OSTI)

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

    2012-05-30

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

  2. 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

  3. 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

  4. 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.

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

    SciTech Connect (OSTI)

    Uberuaga, Blas P.

    2012-05-02

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

  6. 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.

  7. 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 51016 He2+/cm2 at low-temperature (< 200 C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 m 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 m) in the sample subjected to 51016 He2+/cm2, the highest fluence reached, while similar features were not detected at 91015 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.

  8. 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.

  9. 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 ...

  10. 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.

  11. 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).

  12. [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 +}.

  13. 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.

  14. {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)

  15. 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

  16. 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.

  17. 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.

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

    SciTech Connect (OSTI)

    Elam, K.R.

    2003-10-07

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

  19. 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.

  20. 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)

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. 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

  7. 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 methods used in this study.

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

  9. [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.

  10. Pipe diffusion at dislocations in UO2

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

    of pipe diffusion to the overall O 2 and U 4+ diffusion is also discussed. 2014 Elsevier B.V. All rights reserved. 1. Introduction During its lifetime in-pile nuclear fuel...

  11. 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.

  12. 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].

  13. 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.

  14. 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

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

    Office of Scientific and Technical Information (OSTI)

    APA Chicago Bibtex Export Metadata Endnote Excel CSV XML Save to My Library Send to Email Send to Email Email address: Content: Close Send Cite: MLA Format Close Cite: APA ...

  16. 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.

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

    Office of Scientific and Technical Information (OSTI)

    Understanding the migration kinetics of radiation-induced point defects and defect clusters is a key to predicting the microstructural evolution and mass transport in nuclear ...

  18. 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

  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. 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

  2. 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.

  3. 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.

  4. 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}.

  5. 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.

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

    Office of Scientific and Technical Information (OSTI)

    Assembly for Actinide Waste Minimization Citation Details In-Document Search Title: A Combined Nonfertile and UOsub 2 PWR Fuel Assembly for Actinide Waste Minimization A new ...

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

    Office of Legacy Management (LM)

    District (MED) and Atomic Energy Commission (AEC) activities ... also identified and included on the FUSRAP site list. ... Louis University Washington University. North Carolina ...

  8. 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.

  9. 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.

  10. MS/MS Automated Selected Ion Chromatograms

    Energy Science and Technology Software Center (OSTI)

    2005-12-12

    This program can be used to read a LC-MS/MS data file from either a Finnigan ion trap mass spectrometer (.Raw file) or an Agilent Ion Trap mass spectrometer (.MGF and .CDF files) and create a selected ion chromatogram (SIC) for each of the parent ion masses chosen for fragmentation. The largest peak in each SIC is also identified, with reported statistics including peak elution time, height, area, and signal to noise ratio. It creates severalmore » output files, including a base peak intensity (BPI) chromatogram for the survey scan, a BPI for the fragmentation scans, an XML file containing the SIC data for each parent ion, and a "flat file" (ready for import into a database) containing summaries of the SIC data statistics.« less

  11. 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).

  12. 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.

  13. LA-983%MS

    Office of Legacy Management (LM)

    LA-983%MS p 1);: 3 -: ,,i .' . , , -- .. >: 1.. . bob l,,,o* atlonal t&,oratov is operated by the University of California for the United States Department of Energy under ...

  14. MS Based Metabonomics

    SciTech Connect (OSTI)

    Want, Elizabeth J.; Metz, Thomas O.

    2010-03-01

    Metabonomics is the latest and least mature of the systems biology triad, which also includes genomics and proteomics, and has its origins in the early orthomolecular medicine work pioneered by Linus Pauling and Arthur Robinson. It was defined by Nicholson and colleagues in 1999 as the quantitative measurement of perturbations in the metabolite complement of an integrated biological system in response to internal or external stimuli, and is often used today to describe many non-global types of metabolite analyses. Applications of metabonomics are extensive and include toxicology, nutrition, pharmaceutical research and development, physiological monitoring and disease diagnosis. For example, blood samples from millions of neonates are tested routinely by mass spectrometry (MS) as a diagnostic tool for inborn errors of metabolism. The metabonome encompasses a wide range of structurally diverse metabolites; therefore, no single analytical platform will be sufficient. Specialized sample preparation and detection techniques are required, and advances in NMR and MS technologies have led to enhanced metabonome coverage, which in turn demands improved data analysis approaches. The role of MS in metabonomics is still evolving as instrumentation and software becomes more sophisticated and as researchers realize the strengths and limitations of current technology. MS offers a wide dynamic range, high sensitivity, and reproducible, quantitative analysis. These attributes are essential for addressing the challenges of metabonomics, as the range of metabolite concentrations easily exceeds nine orders of magnitude in biofluids, and the diversity of molecular species ranges from simple amino and organic acids to lipids and complex carbohydrates. Additional challenges arise in generating a comprehensive metabolite profile, downstream data processing and analysis, and structural characterization of important metabolites. A typical workflow of MS-based metabonomics is shown in Figure 1. Gas chromatography-(GC)-MS was the most commonly used MS-based method for small molecule analysis in the 1970s and 1980s. It is still used today for the detection of many metabolic disorders and plays a strong role in plant metabonomics. Liquid chromatography (LC)-MS approaches have grown in popularity for metabolite studies, due to simpler sample preparation, reduced analysis times through the introduction of ultra-high performance liquid chromatography (UPLC)-MS and the ability to observe a wider range of metabolites. This chapter will discuss the role of MS in metabonomics, the techniques involved in this exciting area, and the current and future applications of the field. The various bioinformatics tools and multivariate analysis techniques used to maximize information recovery and to aid in the interpretation of the very large data sets typically obtained in metabonomics studies will also be discussed. While there are many different MS-based approaches utilized in metabonomics studies, emphasis will be placed on more established methods.

  15. IMS - MS Data Extractor

    Energy Science and Technology Software Center (OSTI)

    2015-10-20

    An automated drift time extraction and computed associated collision cross section software tool for small molecule analysis with ion mobility spectrometry-mass spectrometry (IMS-MS). The software automatically extracts drift times and computes associated collision cross sections for small molecules analyzed using ion mobility spectrometry-mass spectrometry (IMS-MS) based on a target list of expected ions provided by the user.

  16. Ms. Maggie Owen, Chair

    Office of Environmental Management (EM)

    Todd Mullins Department of Energy Portsmouth/Paducah Project Office 1017 Majestic Drive, Suite 200 Lexington, Kentucky 40513 (859) 219-4000 NOV 1 4 1014 Federal Facility Agreement Manager Division of Waste Management Kentucky Department for Environmental Protection 200 Fair Oaks Lane, 2 nd Floor Frankfort, Kentucky 40601 Ms. Jennifer Tufts Federal Facility Agreement Manager U.S. Environmental Protection Agency, Region 4 61 Forsyth Street Atlanta, Georgia 30303 Dear Mr. Mullins and Ms. Tufts:

  17. ICP-MS Workshop

    SciTech Connect (OSTI)

    Carman, April J.; Eiden, Gregory C.

    2014-11-01

    This is a short document that explains the materials that will be transmitted to LLNL and DNN HQ regarding the ICP-MS Workshop held at PNNL June 17-19th. The goal of the information is to pass on to LLNL information regarding the planning and preparations for the Workshop at PNNL in preparation of the SIMS workshop at LLNL.

  18. LA-13859-MS

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

    0733-MS UC-15 Issued: June 1986 LA-10733-HS DE86 013070 Equations for Plutonium and Americium-241 Decay Corrections T. E. Sampson J. L Parker DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsi- bility for the accuracy, completeness, or usefulness of any

  19. MS, II-J

    Office of Legacy Management (LM)

    I' ; ,' Departm&th of Energy 1 MS, II-J Washington. DC 20585 ' . I I The Honorable John Gallagher ,)fl', /',' ' 103 E. Michigan Avenue .i., ,.' Battle Creek, Michigan 49016 _. Dear Mayor Gallagheri d,---, " '/ approachto openness i.n: with the: public. In (FUSRAP)i.is responsible agencies, determining ~author~ity, performing remedial action to cleanup sites to meet current radiological protection requirements.. A conservative set of technical evaluation guidelines is used in these

  20. Ms. Maria Galanti

    Energy Savers [EERE]

    HAR 2 4 lDII Ohio Environmental Protection Agency Southeast District Office 2195 Front Street Logan, Ohio 43138 Dear Ms. Galanti: PPPO-03-1158259-11 CONSTRUCTION COMPLETION REPORT FOR REMOVAL OF THE X-533 SWITCHYARD COMPLEX AT THE PORTSMOUTH GASEOUS DIFFUSION PLANT, PIKETON, OHIO The Department of Energy is submitting the enclosed Construction Completion Report for Removal of the X-533 Switchyard Complex at the Portsmouth Gaseous Diffusion Plant, Piketon, Ohio (DOEIPPPO/03-0174&D1) to the

  1. Ms. Susan Leckband, Chair W

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

    MAR 1 9 2009 Ms. Susan Leckband, Chair W o r d Advisory Board 713 Jadwin, Suite 4 achland, WA 99352 Dear Ms. Leckband: Thank you for your letter dated December 5,2008, providing...

  2. Is LA-12152-MS

    Office of Scientific and Technical Information (OSTI)

    LA-12152-MS DE91 016813 A Weibull Brittle Material Failure Model for the ABAQUS Computer Program Joel Bennett L ('r^^r5' /A\ n^rnr?i/'7^(^ '-°s Alamos National Laboratory l y j ^ /AAUCSILI LI U i y j ^ LOS Alamos.New Mexico 87545 ^ _ . * i - DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes

  3. Ms. Margaret Owen, Chair

    Office of Environmental Management (EM)

    November 23,2011 Oak Ridge Site Specific Advisory Board P.O. Box 2001 Oak Ridge, Tennessee 3 7831 Dear Ms. Owen: Thank you for your September 19, 2011, letter recommending that we identify waste that could be transported for disposal by rail instead of highway, and improve communication with local communities impacted by the loading and unloading of the waste from one conveyance to the other. Within the Office of Environmental Management (EM), we routinely look at our transport options when we

  4. METLIN: MS/MS metabolite data from the MAGGIE Project

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

    METLIN is a metabolite database for metabolomics containing over 50,000 structures, it also represents a data management system designed to assist in a broad array of metabolite research and metabolite identification by providing public access to its repository of current and comprehensive MS/MS metabolite data. An annotated list of known metabolites and their mass, chemical formula, and structure are available on the METLIN website. Each metabolite is conveniently linked to outside resources such as the the Kyoto Encyclopedia of Genes and Genomes (KEGG) for further reference and inquiry. MS/MS data is also available on many of the metabolites. The list is expanding continuously as more metabolite information is being deposited and discovered. [from http://metlin.scripps.edu/] Metlin is a component of the MAGGIE Project. MAGGIE is funded by the DOE Genomics: GTL and is an acronym for "Molecular Assemblies, Genes, and Genomics Integrated Efficiently."

  5. 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.

  6. LA-5097-MS INFORMAL REPORT

    Office of Legacy Management (LM)

    5097-MS INFORMAL REPORT lamos lamos scientific laboratory scientific laboratory of the University of California of the University of California LOS ALAMOS. NEW MEXICO 87544 LOS ALAMOS. NEW MEXICO 87544 Los AIamos Land Areas Environmental Radiation Survey 1972 . In the interest of prompt distribution, this LAMS re port was not edited by the Technical Information staff. Printed in the United States of America. Available from National Technical Information Service U. S. Department of Commerce 5285

  7. Ms. Maria Galanti Site Coordinator

    Energy Savers [EERE]

    DEC 23 ZDto PPPO-03-1088949-11 Ohio Environmental Protection Agency 2195 Front Street Logan, Ohio 43138 Dear Ms. Galanti: REVISED CONSTRUCTION COMPLETION REPORT FOR PHASE I OF THE REMOVAL OF THE X-633 RECIRCULATING COOLING WATER COMPLEX AT THE PORTSMOUTH GASEOUS DIFFUSION PLANT, PIKETON, OHIO AND RESPONSES TO COMMENTS The Department of Energy is SUbmitting the enclosed revised Construction Completion Report for Phase I of the Removal of the X-633 ReCirculating Cooling Water Complex at the

  8. Ms. Maria Galanti Site Coordinator

    Energy Savers [EERE]

    3 0 2015 Ohio Environmental Protection Agency Southeast District Office 2195 Front Street Logan, Ohio 43138 Dear Ms. Galanti: PPP0-03-3065331-15 FINAL RECORD OF DECISION FOR THE PROCESS BUILDINGS AND COMPLEX FACILITIES DECONTAMINATION AND DECOMMISSIONING EVALUATION PROJECT AT THE PORTSMOUTH GASEOUS DIFFUSION PLANT, PIKETON, OHIO (DOE/PPP0/03-0425&Dl) References: 1. Letter from W. Murphie to M. Galanti, "Record of Decision for the Process Buildings and Complex Facilities Decontamination

  9. Category:Jackson, MS | Open Energy Information

    Open Energy Info (EERE)

    Jackson, MS Jump to: navigation, search Go Back to PV Economics By Location Media in category "Jackson, MS" The following 16 files are in this category, out of 16 total....

  10. Ms. Maria Galanti Site Coordinator

    Energy Savers [EERE]

    ? 5 2011 PPPO-03-1251788-11 Ohio Environmental Protection Agency Southeast District Office 2195 Front Street Logan, Ohio 43138 Dear Ms. Galanti: TRANSMITTAL OF Dl CONSTRUCTION COMPLETION REPORT FOR PHASES I AND II OF THE REMOVAL OF THE X-760 CHEMICAL ENGINEERING BUILDING AT THE PORTSMOUTH GASEOUS DIFFUSION PLANT, PIKETON, OHIO (DOE/PPPO/03-0196&Dl) Reference: Letter from M. Galanti to J. Bradbume, "Construction Completion Report for Phases I and II ofthe Removal of the X-760 Chemical

  11. TAU Portable Performance Profiling Tools Sameer Shende

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

    TAU Portable Performance Profiling Tools Sameer Shende Department of CIS, University of Oregon, Advanced Computing Laboratory, Los Alamos National Laboratory sameer@cs.uoregon.edu Tuning and Analysis Utilities http://www.acl.lanl.gov/tau TAU Profiling Team Members (In alphabetical order) Peter Beckman (LANL) Prof. Janice Cuny (UO) Steve Karmesin (LANL) Kathleen Lindlan (UO) Prof. Allen D. Malony (UO) Sameer Shende (UO, LANL) Tuning and Analysis Utilities http://www.acl.lanl.gov/tau TAU: Tuning

  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. ARM - Campaign Instrument - pyran-eko-ms-801

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

    eko-ms-801 Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Campaign Instrument : Pyranometer-eko-ms-801 (PYRAN-EKO-MS-801

  14. LC-IMS-MS Feature Finder

    SciTech Connect (OSTI)

    2013-03-07

    LC-IMS-MS Feature Finder is a command line software application which searches for possible molecular ion signatures in multidimensional liquid chromatography, ion mobility spectrometry, and mass spectrometry data by clustering deisotoped peaks with similar monoisotopic mass values, charge states, elution times, and drift times. The software application includes an algorithm for detecting multiple conformations and co-eluting species in the ion mobility dimension. LC-IMS-MS Feature Finder is designed to create an output file with detected features that includes associated information about the detected features.

  15. ARM - Campaign Instrument - ptr-ms

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

    govInstrumentsptr-ms Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Campaign Instrument : Proton Transfer Reaction Mass Spectrometer (PTR-MS) Instrument Categories Aerosols Campaigns 2006 MAX-Mex-Megacity Aerosol eXperiment - Mexico City [ Download Data ] Off Site Campaign : various, including non-ARM sites, 2006.03.03 - 2006.03.28 Carbonaceous Aerosol and Radiation Effects Study (CARES) - Surface Meteorological Sounding [ Download Data ] Off

  16. LC-IMS-MS Feature Finder

    Energy Science and Technology Software Center (OSTI)

    2013-03-07

    LC-IMS-MS Feature Finder is a command line software application which searches for possible molecular ion signatures in multidimensional liquid chromatography, ion mobility spectrometry, and mass spectrometry data by clustering deisotoped peaks with similar monoisotopic mass values, charge states, elution times, and drift times. The software application includes an algorithm for detecting multiple conformations and co-eluting species in the ion mobility dimension. LC-IMS-MS Feature Finder is designed to create an output file with detected features thatmore » includes associated information about the detected features.« less

  17. LA-11873-MS The Forster, Dexter, and

    Office of Scientific and Technical Information (OSTI)

    LA-11873-MS The Forster, Dexter, and Inokuti-Hirayama Models of the Time Dependence of Fluorescence Amplitude: An Annotated Bibliography \)\) DISTRIBUTION OF THIS DOCUMENT IS UNLIM ITED Los Alamos National Laboratory is operated by the University of California for the United States Department of Energy under contract W-7405-ENG-36. DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency

  18. Ms. Judy Clayton, Chair Paducah Citizens Advisory Board EHI Consultant...

    Office of Environmental Management (EM)

    0 Ms. Judy Clayton, Chair Paducah Citizens Advisory Board EHI Consultants, Inc. 1 1 1 Memorial Drive Paducah, Kentucky 4200 1 Dear Ms. Clayton: Thank you for your recent letter...

  19. Bioenergia Brasil S A MS | Open Energy Information

    Open Energy Info (EERE)

    Brasil S A MS Jump to: navigation, search Name: Bioenergia Brasil SA (MS) Place: Mato Grosso do Sul, Brazil Product: Company developing a 100m litre per year ethanol plant in Mato...

  20. LA--12O48-MS DE91

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

    -12O48-MS DE91 010299 I ~ I i Tm Thou.mwl Yews of Solitude? 0)1 llzfuhwh!lll Illfrlwiml i)ffo fhc Wmtefsolfltiwl Pilot Project Rqwsitory GrL~gor!/B L')/fo)"(i* Craig W. Kirhood** HmwjOfwf7y Marfit~/. Pmquak!!i+ ~~~~n~~~~L..Al...s.Me. M.xico 87541 L A N L % D I O T D I U E Table of Contents Preface . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. .. .. .. .. .. D O . . . . . . . . . . . ...ooOO.OO..OOOOO"OO".OO """"."" 'ti 1. Introduction . .

  1. 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.

  2. 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

  3. Neutron and Nuclear Science To/MS: Distribution From/MS: Stephen Wender/H855

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

    memorandum Neutron and Nuclear Science To/MS: Distribution From/MS: Stephen Wender/H855 Phone/Fax: 7-1344/5-3705 E-mail: wender@lanl.gov Symbol: LANSCE-NS-14-02 Date: February 4, 2014 Subject: AUTHORIZATIONS AND ASSIGNMENTS I. LANSCE-NS Additional Duty Assignments ALARA Coordinator Ron Nelson Crane Coordinator Gregg Chaparro Facility Coordinator Steve Wender Electrical Safety Officer William Waganaar ES&H Officer Steve Wender Forklift Coordinator Tim Medina Lockout/Tagout Coordinator Ron

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

    Office of Scientific and Technical Information (OSTI)

    UO2(+x) by Ultrafast THz Spectroscopy and Microwave Dissipation (Journal Article) | SciTech Connect 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 Demonstration of a Polaronic Bose-Einstein(-Mott) Condensate in UO2(+x) by Ultrafast THz Spectroscopy and Microwave Dissipation Bose-Einstein condensates (BECs) composed of polarons would be an advance because they

  5. 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)

  6. 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.

  7. ICP-MS Data Analysis Software

    Energy Science and Technology Software Center (OSTI)

    1999-01-14

    VG2Xl - this program reads binary data files generated by VG instrumentals inductively coupled plasma-mass spectrometers using PlasmaQuad Software Version 4.2.1 and 4.2.2 running under IBM OS/2. ICPCalc - this module is a macro for Microsoft Excel written in VBA (Virtual Basic for Applications) that performs data analysis for ICP-MS data required for nuclear materials that cannot readily be done with the vendor''s software. VG2GRAMS - This program reads binary data files generated by VGmore » instruments inductively coupled plasma mass spectrometers using PlasmaQuad software versions 4.2.1 and 4.2.2 running under IBM OS/2.« less

  8. 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)

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

  9. 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.

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

    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...

  11. January 2013 Most Viewed Documents for Fission And Nuclear Technologie...

    Office of Scientific and Technical Information (OSTI)

    E DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL Kessinger, G.; Thompson, M. Safeguardability of advanced spent fuel conditioning process Li, T. K. (Tien K.) ...

  12. OPTICAL PROPERTIES OF A MECHANICALLY POLISHED AND AIR-EQUILIBRATED [111]

    Office of Scientific and Technical Information (OSTI)

    UO2 SURFACE BY RAMAN AND ELLIPSOMETRIC SPECTROSCOPY (Conference) | SciTech Connect Conference: OPTICAL PROPERTIES OF A MECHANICALLY POLISHED 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-EQUILIBRATED [111] UO2 SURFACE BY RAMAN AND ELLIPSOMETRIC SPECTROSCOPY Optical constants of a [111] UO{sub 2} surface, aged in air, were measured in the range from .8 and 5 eV

  13. Theoretical investigation of thermodynamic stability and mobility of the

    Office of Scientific and Technical Information (OSTI)

    oxygen vacancy in ThO2 -UO2 solid solutions (Journal Article) | DOE PAGES Theoretical investigation of thermodynamic stability and mobility of the oxygen vacancy in ThO2 -UO2 solid solutions « Prev Next » Title: Theoretical investigation of thermodynamic stability and mobility of the oxygen vacancy in ThO2 -UO2 solid solutions The thermodynamic stability and the migration energy barriers of oxygen vacancies in ThO2 -UO2 solid solutions are investigated by density functional theory

  14. Municipal Energy Agency of MS | Open Energy Information

    Open Energy Info (EERE)

    Name: Municipal Energy Agency of MS Place: Mississippi Phone Number: (601) 362-2252 Facebook: https:www.facebook.compagesMunicipal-Energy-Agency-of-Mississippi Outage...

  15. Commander, Naval Base ATTN: Ms. Cheryl Barnett Building N-26

    Office of Legacy Management (LM)

    Building N-26 Code N 9 E Norfolk, Virginia 23511-6002 Dear Ms. Barnett: I enjoyed speaking with you on the phone. The Department of Energy (DOE) has established its Formerly...

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. This is a paper model of the MS2 virus

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

    a paper model of the MS2 virus. MS2 is a nanoscale virus that lives in the human gut. It is benign and doesn't interfere with us. A virus, it attaches to a host cell that it identifies using protein markers on its exterior. Once attached, it injects its RNA genetic material into the cell. This material then co-opts the cell's metabolism to produce copies of the virus, which the cell releases back into the organism's tissues. Researchers have found that this mechanism has potential as a treatment

  1. CONCURRENC RTG. SYMBOL GC-34 Ms. Mary Beth Brado

    Office of Legacy Management (LM)

    MAY 2 9 1980 CONCURRENC RTG. SYMBOL GC-34 Ms. Mary Beth Brado "*N'W Town of Lewiston * i..,! 1375 Ridge Road ^r'8 Lewiston, New York 14092 RTG.SYuBOL Dear Ms. Brado: .- ,l13. INirIA Lss iQ. W'Mott This is in response to your letter of January 29, 1980, and subsequent ..... ,. telephone discussions with irr. Brazley of my office, concerning land use 5/ /8 restrictions on the 1,511 acres declared surplus in the Towns of Lewiston RGSYMOL. and Porter, New York. In regard to your question of

  2. When worlds collide - Mac to MS-DOS. [Data transfer to and from Apple Macintosh computers and MS-DOS based personal computers

    SciTech Connect (OSTI)

    Busbey, A.B.

    1989-04-01

    A number of methods and products, both hardware and software, to allow data exchange between Apple Macintosh computers and MS-DOS based systems. These included serial null modem connections, MS-DOS hardware and/or software emulation, MS-DOS disk-reading hardware and networking.

  3. PNM Resources 2401 Aztec NE, MS-Z100

    Energy Savers [EERE]

    PNM Resources 2401 Aztec NE, MS-Z100 Albuquerque, NM 87107 505-241-2025 Fax 505 241-2384 PNMResources.com October 29, 2013 Mr. Christopher Lawrence Office of Electricity Delivery and Energy Reliability (OE-20) U.S. Department of Energy 1000 Independence Avenue, SW Washington, DC 20585 Submitted electronically via email to: Christopher.Lawrence@hq.doe.gov Dear Mr. Lawrence: Subject: Department of Energy (DOE)- Improving Performance of Federal Permitting and Review of Infrastructure Projects,

  4. Ms. Julie Smith Office of Electricity Delivery and Energy Reliability

    Energy Savers [EERE]

    5, 2013 Ms. Julie Smith Office of Electricity Delivery and Energy Reliability Mail Code OE-20 U.S. Department of Energy 1000 Independence Avenue Washington, D.C. 20585 Juliea.smith@hq.doe.gov; Christopher.lawrence@hq.doe.gov Re: DOE RFI "Improving Performance of Federal Permitting and Review of Infrastructure Projects The American people support increased production and consumption of renewable energy according to credible public opinion polls. Too often the most appropriate sites for wind,

  5. Method for factor analysis of GC/MS data

    DOE Patents [OSTI]

    Van Benthem, Mark H; Kotula, Paul G; Keenan, Michael R

    2012-09-11

    The method of the present invention provides a fast, robust, and automated multivariate statistical analysis of gas chromatography/mass spectroscopy (GC/MS) data sets. The method can involve systematic elimination of undesired, saturated peak masses to yield data that follow a linear, additive model. The cleaned data can then be subjected to a combination of PCA and orthogonal factor rotation followed by refinement with MCR-ALS to yield highly interpretable results.

  6. 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.

  7. 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.

  8. An Anonymous Referee Report (Technical Report) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Technical Report: An Anonymous Referee Report Citation Details In-Document Search Title: An Anonymous Referee Report This is a combined experimental and theoretical study of the compound UO2, UO2(NO3)2(H20)6 and UO0.75Pu0.25O2, using resonant inelastic x-ray scattering (RXIS), high resolution x-ray absorption (XAS) and LDA and LDA-U calculations. Authors: Tobin, J. G. [1] + Show Author Affiliations Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States) Publication Date:

  9. Possible Bose-condensate Behavior in a Quantum Phase Originating in a

    Office of Scientific and Technical Information (OSTI)

    Collective Excitation in the Chemically and Optically Doped Mott-Hubbard System UO2+x (Journal Article) | SciTech Connect Possible Bose-condensate Behavior in a Quantum Phase Originating in a Collective Excitation in the Chemically and Optically Doped Mott-Hubbard System UO2+x Citation Details In-Document Search Title: Possible Bose-condensate Behavior in a Quantum Phase Originating in a Collective Excitation in the Chemically and Optically Doped Mott-Hubbard System UO2+x Authors: Conradson,

  10. Plains and Eastern Clean Line Transmission Line: Comment from Ms. Callahan

    Energy Savers [EERE]

    | Department of Energy Ms. Callahan Plains and Eastern Clean Line Transmission Line: Comment from Ms. Callahan Comment submitted on updated Part 2 application. PDF icon Comment from Ms. Callahan 07-13-15.pdf More Documents & Publications Plains and Eastern Clean Line Transmission Line: Comment from Cynthia Blansett (COE) Plains and Eastern Clean Line Transmission Line: Comment from Dr. Contreras Plains and Eastern Clean Line Transmission Line: Comment from Simon Mahan (SWEA)

  11. Microsoft Word - DOE-ID-14-080 South Florida EC B3-6.doc

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

    uranium "nano-traps" as well as the investigation of their performance on UO 2 2+ adsorption in the lab scale. Collaborators will engage in ambient seawater uranium adsorption...

  12. 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.

  13. 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 )...

  14. 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...

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

    Office of Scientific and Technical Information (OSTI)

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

  16. POST-DEACTIVATION SURVEILLANCE AND MAINTENANCE PLANNING

    Office of Environmental Management (EM)

    ... main incoming steam to the UO3 facility has been shut off and the main isolation valve locktagged closed. KEH Utilities will remove the valve during the scheduled spring outage. ...

  17. 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.

  18. 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

  19. NEAMS Update

    Energy Savers [EERE]

    ... 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 ...

  20. Publications and Presentations at Scientific Meetings | Stanford...

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

    Giammar, and B.M. Tebo (2008) Indirect UO2 oxidation by Mn(II)-oxidizing spores of Bacillus sp. strain SG-1 and the effect of U and Mn concentrations. Environ. Sci. Technol....

  1. 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 ...

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

    SciTech Connect (OSTI)

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

    2013-08-14

    Subsurface bacteria including sulfate reducing bacteria (SRB) reduce soluble U(VI) to insoluble U(IV) with subsequent precipitation of UO2. We have shown that SRB reduce U(VI) to nanometer-sized UO2 particles (1-5 nm) which are both intra- and extracellular, with UO2 inside the cell likely physically shielded from subsequent oxidation processes. We evaluated the UO2 nanoparticles produced by Desulfovibrio desulfuricans G20 under growth and non-growth conditions in the presence of lactate or pyruvate and sulfate, thiosulfate, or fumarate, using ultrafiltration and HR-TEM. Results showed that a significant mass fraction of bioreduced U (35-60%) existed as a mobile phase when the initial concentration of U(VI) was 160 M. Further experiments with different initial U(VI) concentrations (25 - 900 ?M) in MTM with PIPES or bicarbonate buffers indicated that aggregation of uraninite depended on the initial concentrations of U(VI) and type of buffer. It is known that under some conditions SRB-mediated UO2 nanocrystals can be reoxidized (and thus remobilized) by Fe(III)-(hydr)oxides, common constituents of soils and sediments. To elucidate the mechanism of UO2 reoxidation by Fe(III) (hydr)oxides, we studied the impact of Fe and U chelating compounds (citrate, NTA, and EDTA) on reoxidation rates. Experiments were conducted in anaerobic batch systems in PIPES buffer. Results showed EDTA significantly accelerated UO2 reoxidation with an initial rate of 9.5?M day-1 for ferrihydrite. In all cases, bicarbonate increased the rate and extent of UO2 reoxidation with ferrihydrite. The highest rate of UO2 reoxidation occurred when the chelator promoted UO2 and Fe(III) (hydr)oxide dissolution as demonstrated with EDTA. When UO2 dissolution did not occur, UO2 reoxidation likely proceeded through an aqueous Fe(III) intermediate as observed for both NTA and citrate. To complement to these laboratory studies, we collected U-bearing samples from a surface seep at the Rifle field site and have measured elevated U concentrations in oxic iron-rich sediments. To translate experimental results into numerical analysis of U fate and transport, a reaction network was developed based on Sani et al. (2004) to simulate U(VI) bioreduction with concomitant UO2 reoxidation in the presence of hematite or ferrihydrite. The reduction phase considers SRB reduction (using lactate) with the reductive dissolution of Fe(III) solids, which is set to be microbially mediated as well as abiotically driven by sulfide. Model results show the oxidation of HS by Fe(III) directly competes with UO2 reoxidation as Fe(III) oxidizes HS preferentially over UO2. The majority of Fe reduction is predicted to be abiotic, with ferrihydrite becoming fully consumed by reaction with sulfide. Predicted total dissolved carbonate concentrations from the degradation of lactate are elevated (log(pCO2) ~ 1) and, in the hematite system, yield close to two orders-of-magnitude higher U(VI) concentrations than under initial carbonate concentrations of 3 mM. Modeling of U(VI) bioreduction with concomitant reoxidation of UO2 in the presence of ferrihydrite was also extended to a two-dimensional field-scale groundwater flow and biogeochemically reactive transport model for the South Oyster site in eastern Virginia. This model was developed to simulate the field-scale immobilization and subsequent reoxidation of U by a biologically mediated reaction network.

  3. 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...

  4. 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...

  5. Characterization of Fatty Acids in Crenarchaeota by GC-MS and...

    Office of Scientific and Technical Information (OSTI)

    Title: Characterization of Fatty Acids in Crenarchaeota by GC-MS and NMR Lipids composed ... Because lipids are energy currency and cell signaling molecules, their presence in Archaea ...

  6. ST. LCUIS ST. LallS JOKIN KFms Clrv S!. LrMS ST. Lcm

    Office of Legacy Management (LM)

    Ho. 30 !2121/87 smt6w cm ST. LCUIS ST. LallS JOKIN KFms Clrv S!. LrMS ST. Lcm

  7. Ms Van T Nguyen | U.S. DOE Office of Science (SC)

    Office of Science (SC) Website

    firm as a Chemical Engineer. Education: M.S., Environmental Science and Engineering, Virginia Polytechnic Institute and State University, 1993 B.S., Chemical Engineering, ...

  8. 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.

  9. Gas chromatograph-mass spectrometer (GC/MS) system for quantitative analysis of reactive chemical compounds

    DOE Patents [OSTI]

    Grindstaff, Quirinus G.

    1992-01-01

    Described is a new gas chromatograph-mass spectrometer (GC/MS) system and method for quantitative analysis of reactive chemical compounds. All components of such a GC/MS system external to the oven of the gas chromatograph are programmably temperature controlled to operate at a volatilization temperature specific to the compound(s) sought to be separated and measured.

  10. 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.

  11. 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

  12. File:USDA-CE-Production-GIFmaps-MS.pdf | Open Energy Information

    Open Energy Info (EERE)

    MS.pdf Jump to: navigation, search File File history File usage Mississippi Ethanol Plant Locations Size of this preview: 776 600 pixels. Full resolution (1,650 1,275...

  13. Ms Linda Cerrone | U.S. DOE Office of Science (SC)

    Office of Science (SC) Website

    Prior to her employment with DOE, Ms. Cerrone served 13 years with Mason & Hanger-Silas ... manager with Scurlock-Permian Corporation (formerly the Permian Corporation) for 11 years. ...

  14. WSRC-MS-94-0605 Wall Thinning Acceptance Criteria for Degraded...

    Office of Scientific and Technical Information (OSTI)

    WSRC-MS-94-0605 Wall Thinning Acceptance Criteria for Degraded Carbon Steel Piping Systems Using FAD Methodology (U) by P. S. Lam Westinghouse Savannah River Company Savannah River ...

  15. Training is sponsoring two MS PowerPoint Courses. The course...

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

    Training is sponsoring two MS PowerPoint Courses. The course options feature level 1 and level 2. This email is to gauge interest. The associated cost per class is 200.00. If...

  16. Ms. Judy Clayton, Chair Paducah Citizens Advisory Board EHI Consultants, Inc.

    Office of Environmental Management (EM)

    0 Ms. Judy Clayton, Chair Paducah Citizens Advisory Board EHI Consultants, Inc. 1 1 1 Memorial Drive Paducah, Kentucky 4200 1 Dear Ms. Clayton: Thank you for your recent letter providing a recommendation to include option periods in Department of Energy Requests for Proposals for prime contracts. The Federal Acquisition Regulation (FAR) provides the total of the basic and option period of a contract for services shall not exceed 5 years, unless approved in accordance with agency procedures. In

  17. Mr. Christopher Lawrence and Ms. Julie Smith Office of Electricity Delivery and Energy Reliability

    Energy Savers [EERE]

    Mr. Christopher Lawrence and Ms. Julie Smith Office of Electricity Delivery and Energy Reliability Mail Code: OE-20 U.S. Department of Energy 1000 Independence Avenue, SW Washington, D.C. 20585 Dear Mr. Lawrence and Ms. Smith, The Western Governors' Association (WGA) is submitting these comments in response to the Department of Energy's (DOE) Request for Information (RFI), dated August 29, 2013 1 . The RFI outlines a proposed process to establish a coordinated series of meetings and other

  18. Ms. Brenda Edwards U.S. Department of Energy, Buildings Technology Program

    Energy Savers [EERE]

    8, 2013 Ms. Brenda Edwards U.S. Department of Energy, Buildings Technology Program Mail Stop EE-2J 1000 Independence Ave. SW Washington DC 20585-0121 RE: Revisions to Energy Efficiency Enforcement Regulations EERE-2011-BT-TD- 0005 Dear Ms. Edwards: The National Marine Manufacturers Association (NMMA) appreciates the opportunity to respond to the Department of Energy (DOE), Office of Energy Efficiency and Renewable Energy's request for information published in the Federal Register (78 Fed. Reg.

  19. Ms. Julie A. Smith Mr. Christopher Lawrence Office of Electricity Delivery and Energy Reliability

    Energy Savers [EERE]

    October 30, 2013 Ms. Julie A. Smith Mr. Christopher Lawrence Office of Electricity Delivery and Energy Reliability Mail Code: OE-20 U.S. Department of Energy 1000 Independence Avenue, SW Washington, DC 20585 Dear Ms. Smith and Mr. Lawrence: The Association of Fish and Wildlife Agencies (AFWA) would like to provide comments on the Federal Register Notice Request for Information (RFI) on Improving Performance of Federal Permitting and Review of Infrastructure Projects, Federal Register Document

  20. Modeling and simulation of Red Teaming. Part 1, Why Red Team M&S?

    SciTech Connect (OSTI)

    Skroch, Michael J.

    2009-11-01

    Red teams that address complex systems have rarely taken advantage of Modeling and Simulation (M&S) in a way that reproduces most or all of a red-blue team exchange within a computer. Chess programs, starting with IBM's Deep Blue, outperform humans in that red-blue interaction, so why shouldn't we think computers can outperform traditional red teams now or in the future? This and future position papers will explore possible ways to use M&S to augment or replace traditional red teams in some situations, the features Red Team M&S should possess, how one might connect live and simulated red teams, and existing tools in this domain.

  1. M186

    National Nuclear Security Administration (NNSA)

    . S . Department of E n m National NucIear Security A d ~ t i o n P.O. Box M50 Oak Ridge, TN 37831 PAGE ! of 3 PAGES Ahf~NDMENT OF SOLICITATION/MODWJCATZON OF CONTaACT --- I - I 8. NAMEAND ADDRESS OF CONTMCn,R (Uo., me#, &my, Zm We) I 9 k AMENDMENT OF SOLKITATION NO. 1. CONTRACT I D CODE A C 2. kMENDMENTMODFICATION NO. MI86 B a h d & W ~ X T a d Y - 1 2 , LLC P.0, Box 2009 MS 8014 Oak Ridge, 'JX 37?33143014 3. EFlBIWE DATE Sct Block l k . Offasmust~~h&edgtroaeiptdthis t prior to Ihc

  2. Improving Alpha Spectrometry Energy Resolution by Ion Implantation with ICP-MS

    SciTech Connect (OSTI)

    Dion, Michael P.; Liezers, Martin; Farmer, Orville T.; Miller, Brian W.; Morley, Shannon M.; Barinaga, Charles J.; Eiden, Gregory C.

    2015-01-01

    We report results of a novel technique using an Inductively Coupled Plasma Mass Spectrometer (ICP-MS) as a method of source preparation for alpha spectrometry. This method produced thin, contaminant free 241Am samples which yielded extraordinary energy resolution which appear to be at the lower limit of the detection technology used in this research.

  3. 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 associated with spent fuel oxidation (Section 5).

  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 mineralizing agent and a ligand.

  5. 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.

  6. 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}.

  7. Response Model for Kodak Biomax-MS Film to X Rays

    SciTech Connect (OSTI)

    Knauer, J.P.; Marshall, F.J.; Yaakobi, B.; Anderson, D.; Schmitt, B.A.; Chandler, K.M.; Pikuz, S.A.; Shelkovenko, T.A.; Mitchell, M.D.; Hammer, D.A.

    2007-01-24

    X-raysensitive film is used for a variety of imaging and spectroscopic diagnostics for high-temperature plasmas. New film becomes available as older films are phased out of production. Biomax-MS is a T-grain class of film that is proposed as a replacement for Kodak DEF film. A model of its response to x rays is presented. Data from dimensional measurements of the film, x-ray transmission measurements, SEM micrograph images, and x-ray calibration are used to develop this sensitivity model of Biomax-MS film as a function of x-ray energy and angle of incidence. Relative response data provide a check of the applicability of this model to determine the x-ray flux from spectrum data. This detailed film characterization starts with simple mathematical models and extends them to T-grain type film.

  8. Response model for Kodak Biomax-MS film to x rays

    SciTech Connect (OSTI)

    Knauer, J. P.; Marshall, F. J.; Yaakobi, B.; Anderson, D.; Schmitt, B. A.; Chandler, K. M.; Pikuz, S. A.; Shelkovenko, T. A.; Mitchell, M. D.; Hammer, D. A.

    2006-10-15

    X-ray-sensitive film is used for a variety of imaging and spectroscopic diagnostics for high-temperature plasmas. Replacement film must be found as older films are phased out of production. Biomax-MS is a 'T-grain' class of film that is proposed as a replacement for Kodak DEF and a model of its response to x rays is presented. Data from dimensional measurements of the film, x-ray transmission measurements, scanning electron microscopy micrograph images, and x-ray calibration are used to develop this sensitivity model of Biomax-MS film as a function of x-ray energy and angle of incidence. Relative response data provide a check of the applicability of this model to determine the x-ray flux from spectrum data. This detailed film characterization starts with simple mathematical models and extends them to T-grain-type film.

  9. PIK M.S. Onegin Petersburg Nuclear Physics Institute 2015 Super Heavy Elements Symposium

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

    possibilities of heavy actinide isotopes production in high-flux reactor PIK M.S. Onegin Petersburg Nuclear Physics Institute 2015 Super Heavy Elements Symposium Reactor PIK 2011 - Criticality reached 2013 - Construction finished 2014 - 2017 Licensing and neutron stations and installation construction 2018 - Power operation scheduled The maximum heat output 100 MW Heat transfer agent water Reflector heavy water Number of horizontal experimental channels 10 Number inclined experimental channels 6

  10. Paul W. King, Ph.D., M.S. | Bioenergy | NREL

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

    Paul W. King Paul W. King, Ph.D., M.S. Scientist VI - Supervisor, Photobiology Group Paul.King@nrel.gov | 303-384-6277 Research Interests Paul King's research interests are broadly in the study of photobiological and artificial photosynthetic systems for converting solar energy into biofuels. Specific areas of interest include using molecular, biochemical and biophysical techniques to investigate the catalytic mechanisms and structure-function of hydrogenases; bioenergetics of hydrogen

  11. To: Ms. Brenda Edwards U.S. Department of Energy, Building Technologies Program

    Energy Savers [EERE]

    0, 2012 To: Ms. Brenda Edwards U.S. Department of Energy, Building Technologies Program Mailstop EE-2J, 1000 Independence Avenue Washington, DC 20585-0121 Phone: (202) 586-2945 cc: Michelle Blaise (VP, ComEd Engineering & Project Management) Joseph Watson (Director, Federal Government Affairs) Martin Rave (Prin Engineer, ComEd Distribution Standards) From: Peter Tyschenko (Manager, ComEd Distribution Standards) Two Lincoln Centre Oakbrook Terrace, IL 60181-4260 Phone: (630) 576-6998 Subject:

  12. To: Ms. Brenda Edwards U.S. Department of Energy, Building Technologies Program

    Energy Savers [EERE]

    8, 2011 To: Ms. Brenda Edwards U.S. Department of Energy, Building Technologies Program Mailstop EE-2J, 1000 Independence Avenue Washington, DC 20585-0121 Phone: (202) 586-2945 cc: Michelle Blaise (VP, ComEd Engineering & Project Management) Joseph Watson (Director, Federal Government Affairs) Martin Rave (Prin Engineer, ComEd Distribution Standards) From: Peter Tyschenko (Manager, ComEd Distribution Standards) Two Lincoln Centre Oakbrook Terrace, IL 60181-4260 Phone: (630) 576-6998 Subject:

  13. Development of chiral LC-MS methods for small molecules and their applications i

    Office of Scientific and Technical Information (OSTI)

    chiral LC-MS methods for small molecules and their applications i n the analysis of enantiomeric composition and pharmacokinetic studies Meera Jay Desai A dissertation submitted to the graduate faculty in partial fulfillment of the requirements for the degree of DOCTOR OF PHILOSOPHY Major: Analytical Chemistry Program of Study Committee: Daniel W. Armstrong, Major Professor Edward S. Yeung Robsrt S. Houk Victor S.-Y. Lin Gregory Phillips Iowa State University Ames, Iowa 2004 .. 11 Graduate

  14. Evaluation of Ultra-Low Background Materials for Uranium and Thorium Using ICP-MS

    SciTech Connect (OSTI)

    Hoppe, Eric W.; Overman, Nicole R.; LaFerriere, Brian D.

    2013-08-08

    An increasing number of physics experiments require low background materials for their construction. The presence of Uranium and Thorium and their progeny in these materials present a variety of unwanted background sources for these experiments. The sensitivity of the experiments continues to drive the necessary levels of detection ever lower as well. This requirement for greater sensitivity has rendered direct radioassay impractical in many cases requiring large quantities of material, frequently many kilograms, and prolonged counting times, often months. Other assay techniques have been employed such as Neutron Activation Analysis but this requires access to expensive facilities and instrumentation and can be further complicated and delayed by the formation of unwanted radionuclides. Inductively Coupled Plasma Mass Spectrometry (ICP-MS) is a useful tool and recent advancements have increased the sensitivity particularly in the elemental high mass range of U and Th. Unlike direct radioassay, ICP-MS is a destructive technique since it requires the sample to be in liquid form which is aspirated into a high temperature plasma. But it benefits in that it usually requires a very small sample, typically about a gram. Here we will discuss how a variety of low background materials such as copper, polymers, and fused silica are made amenable to ICP-MS assay and how the arduous task of maintaining low backgrounds of U and Th is achieved.

  15. Evaluation of ultra-low background materials for uranium and thorium using ICP-MS

    SciTech Connect (OSTI)

    Hoppe, E. W.; Overman, N. R.; LaFerriere, B. D.

    2013-08-08

    An increasing number of physics experiments require low background materials for their construction. The presence of Uranium and Thorium and their progeny in these materials present a variety of unwanted background sources for these experiments. The sensitivity of the experiments continues to drive the necessary levels of detection ever lower as well. This requirement for greater sensitivity has rendered direct radioassay impractical in many cases requiring large quantities of material, frequently many kilograms, and prolonged counting times, often months. Other assay techniques have been employed such as Neutron Activation Analysis but this requires access to expensive facilities and instrumentation and can be further complicated and delayed by the formation of unwanted radionuclides. Inductively Coupled Plasma Mass Spectrometry (ICP-MS) is a useful tool and recent advancements have increased the sensitivity particularly in the elemental high mass range of U and Th. Unlike direct radioassay, ICP-MS is a destructive technique since it requires the sample to be in liquid form which is aspirated into a high temperature plasma. But it benefits in that it usually requires a very small sample, typically about a gram. This paper discusses how a variety of low background materials such as copper, polymers, and fused silica are made amenable to ICP-MS assay and how the arduous task of maintaining low backgrounds of U and Th is achieved.

  16. Sources of Technical Variability in Quantitative LC-MS Proteomics: Human Brain Tissue Sample Analysis.

    SciTech Connect (OSTI)

    Piehowski, Paul D.; Petyuk, Vladislav A.; Orton, Daniel J.; Xie, Fang; Moore, Ronald J.; Ramirez Restrepo, Manuel; Engel, Anzhelika; Lieberman, Andrew P.; Albin, Roger L.; Camp, David G.; Smith, Richard D.; Myers, Amanda J.

    2013-05-03

    To design a robust quantitative proteomics study, an understanding of both the inherent heterogeneity of the biological samples being studied as well as the technical variability of the proteomics methods and platform is needed. Additionally, accurately identifying the technical steps associated with the largest variability would provide valuable information for the improvement and design of future processing pipelines. We present an experimental strategy that allows for a detailed examination of the variability of the quantitative LC-MS proteomics measurements. By replicating analyses at different stages of processing, various technical components can be estimated and their individual contribution to technical variability can be dissected. This design can be easily adapted to other quantitative proteomics pipelines. Herein, we applied this methodology to our label-free workflow for the processing of human brain tissue. For this application, the pipeline was divided into four critical components: Tissue dissection and homogenization (extraction), protein denaturation followed by trypsin digestion and SPE clean-up (digestion), short-term run-to-run instrumental response fluctuation (instrumental variance), and long-term drift of the quantitative response of the LC-MS/MS platform over the 2 week period of continuous analysis (instrumental stability). From this analysis, we found the following contributions to variability: extraction (72%) >> instrumental variance (16%) > instrumental stability (8.4%) > digestion (3.1%). Furthermore, the stability of the platform and its suitability for discovery proteomics studies is demonstrated.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. Toward Joint Hypothesis-Tests Seismic Event Screening Analysis: Ms|mb and Event Depth

    SciTech Connect (OSTI)

    Anderson, Dale; Selby, Neil

    2012-08-14

    Well established theory can be used to combine single-phenomenology hypothesis tests into a multi-phenomenology event screening hypothesis test (Fisher's and Tippett's tests). Commonly used standard error in Ms:mb event screening hypothesis test is not fully consistent with physical basis. Improved standard error - Better agreement with physical basis, and correctly partitions error to include Model Error as a component of variance, correctly reduces station noise variance through network averaging. For 2009 DPRK test - Commonly used standard error 'rejects' H0 even with better scaling slope ({beta} = 1, Selby et al.), improved standard error 'fails to rejects' H0.

  2. Market Research Survey of Commercial Off-The-Shelf (COTS) Portable MS Systems for IAEA Safeguards Applications

    SciTech Connect (OSTI)

    Hart, Garret L.; Hager, George J.; Barinaga, Charles J.; Duckworth, Douglas C.

    2013-02-01

    This report summarizes the results for the market research survey of mass spectrometers that are deemed pertinent to International Atomic Energy Agency (IAEA) needs and strategic objectives. The focus of the report is on MS instruments that represent currently available (or soon to be) commercial off-the shelf (COTS) technology and weigh less than 400 pounds. A compilation of all available MS instruments (36 COTS and 2 R&D) is presented, along with pertinent information regarding each instrument.

  3. 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.

  4. 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)22H2O, 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)22H2O 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)22H2O and (UO2)(O2)(H2O)2, suggesting a lower micro-hardness of metastudtite.

  5. Infrared Multiphoton Dissociation Spectroscopy of a Gas-Phase Complex of Uranyl and 3-Oxa-Glutaramide: An Extreme Red-Shift of the [O=U=O]? Asymmetric Stretch

    SciTech Connect (OSTI)

    Gibson, John K.; Hu, Hanshi; Van Stipdonk, Michael J.; Berden, Giel; Oomens, Jos; Li, Jun

    2015-04-09

    The gas-phase complex UO?(TMOGA)?? (TMOGA = tetramethyl-3-oxa-glutaramide) prepared by electrospray ionization was characterized by infrared multiphoton dissociation (IRMPD) spectroscopy. The IRMPD spectrum from 7001800 cm? was interpreted using a computational study based on density functional theory. The predicted vibrational frequencies are in good agreement with the measured values, with an average deviation of only 8 cm? (<1%) and a maximum deviation of 21 cm? (<2%). The only IR peak assigned to the linear uranyl moiety was the asymmetric ?? mode, which appeared at 965 cm? and was predicted by DFT as 953 cm?. This ?? frequency is red-shifted relative to bare uranyl, UO??, by ca. 150 cm? due to electron donation from the TMOGA ligands. Based on the degree of red-shifting, it is inferred that two TMOGA oxygen-donor ligands have a greater effective gas basicity than the four monodentate acetone ligands in UO?(acetone)??. The uranyl ?? frequency was also computed for uranyl coordinated by two TMGA ligands, in which the central Oether of TMOGA has been replaced by CH?. The computed ?? for UO?(TMGA)??, 950 cm?, is essentially the same as that for UO?(TMOGA)??, suggesting that electron donation to uranyl from the Oether of TMOGA is minor. The computed ?? asymmetric stretching frequencies for the three actinyl complexes, UO?(TMOGA)??, NpO?(TMOGA)?? and PuO?(TMOGA)??, are comparable. This similarity is discussed in the context of the relationship between ?? and intrinsic actinide-oxygen bond energies in actinyl complexes.

  6. Characterization of CMPO and its radiolysis products by Direct Infusion ESI-MS

    SciTech Connect (OSTI)

    G. S. Groenewold; G. Elias; B. J. Mincher; S. P. Mezyk

    2012-09-01

    Direct infusion electrospray ionization-mass spectrometry (ESI-MS) approaches were developed for rapid identification of octyl,phenyl,(N,N-(diisobutyl)carbamoylmethyl) phosphine oxide (CMPO) and impurity compounds formed during alpha and gamma irradiation experiments. CMPO is an aggressive Lewis base, and produces extremely abundant metal complex ions in the ESI-MS analysis that make identification of low abundance compounds that are less nucleophilic challenging. Radiolysis products were identified using several approaches including restricting ion trapping so as to exclude the abundant natiated CMPO ions, extraction of acidic products using aqueous NaOH, and extraction of basic products using HNO3. These approaches generated protonated, natiated and deprotonated species derived from CMPO degradation products formed via radiolytic cleavages of several different bonds. Cleavages of the amide and methylene-phosphoryl bonds appear to be favored by both forms of irradiation, while alpha irradiation also appears to induce cleavage of the methylene-carbonyl bond. The degradation products observed are formed from recombination of the initially formed radicals with hydrogen, methyl, isopropyl and hydroxyl radicals that are derived either from CMPO, or the dodecane solvent.

  7. Reduction of Solvent Effect in Reverse Phase Gradient Elution LC-ICP-MS

    SciTech Connect (OSTI)

    Patrick Allen Sullivan

    2005-12-17

    Quantification in liquid chromatography (LC) is becoming very important as more researchers are using LC, not as an analytical tool itself, but as a sample introduction system for other analytical instruments. The ability of LC instrumentation to quickly separate a wide variety of compounds makes it ideal for analysis of complex mixtures. For elemental speciation, LC is joined with inductively coupled plasma mass spectrometry (ICP-MS) to separate and detect metal-containing, organic compounds in complex mixtures, such as biological samples. Often, the solvent gradients required to perform complex separations will cause matrix effects within the plasma. This limits the sensitivity of the ICP-MS and the quantification methods available for use in such analyses. Traditionally, isotope dilution has been the method of choice for LC-ICP-MS quantification. The use of naturally abundant isotopes of a single element in quantification corrects for most of the effects that LC solvent gradients produce within the plasma. However, not all elements of interest in speciation studies have multiple naturally occurring isotopes; and polyatomic interferences for a given isotope can develop within the plasma, depending on the solvent matrix. This is the case for reverse phase LC separations, where increasing amounts of organic solvent are required. For such separations, an alternative to isotope dilution for quantification would be is needed. To this end, a new method was developed using the Apex-Q desolvation system (ESI, Omaha, NE) to couple LC instrumentation with an ICP-MS device. The desolvation power of the system allowed greater concentrations of methanol to be introduced to the plasma prior to destabilization than with direct methanol injection into the plasma. Studies were performed, using simulated and actual linear methanol gradients, to find analyte-internal standard (AIS) pairs whose ratio remains consistent (deviations {+-} 10%) over methanol concentration ranges of 5%-35% (simulated) and 8%-32% (actual). Quadrupole (low resolution) and sector field (high resolution) ICP-MS instrumentation were utilized in these studies. Once an AIS pair is determined, quantification studies can be performed. First, an analysis is performed by adding both elements of the AIS pair post-column while performing the gradient elution without sample injection. A comparison of the ratio of the measured intensities to the atomic ratio of the two standards is used to determine a correction factor that can be used to account for the matrix effects caused by the mobile phase. Then, organic and/or biological molecules containing one of the two elements in the AIS pair are injected into the LC column. A gradient method is used to vary the methanol-water mixture in the mobile phase and to separate out the compounds in a given sample. A standard solution of the second ion in the AIS pair is added continuously post-column. By comparing the ratio of the measured intensities to the atomic ratio of the eluting compound and internal standard, the concentration of the injected compound can be determined.

  8. 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.

  9. 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 might decrease because of decreases in the water-volume fraction caused by sludge solid reactions, compaction, or intergrowth and recrystallization of metaschoepite. In that case, fewer STSCs may be needed, but the shear strength would increase, and this could challenge recovery by water jet erosion and require more aggressive retrieval methods. Overall, the tests described herein indicate that the settled solids volume remains the same or decreases with time. The only case for which the sludge solids volumes increase with time is for the expansion factor attendant upon the anoxic corrosion of uranium metal to produce UO2 and subsequent reaction with oxygen to form equimolar UO2.25 and UO3•2H2O.

  10. Possible Bose-condensate behavior in a quantum phase originating in a

    Office of Scientific and Technical Information (OSTI)

    collective excitation in the chemically and optically doped Mott-Hubbard system UO[subscript 2+x] (Journal Article) | SciTech Connect Possible Bose-condensate behavior in a quantum phase originating in a collective excitation in the chemically and optically doped Mott-Hubbard system UO[subscript 2+x] Citation Details In-Document Search Title: Possible Bose-condensate behavior in a quantum phase originating in a collective excitation in the chemically and optically doped Mott-Hubbard system

  11. 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

  12. 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.

  13. 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.

  14. Pentaerythritol Tetranitrate (PETN) Surveillance by HPLC-MS: Instrumental Parameters Development

    SciTech Connect (OSTI)

    Harvey, C A; Meissner, R

    2005-11-04

    Surveillance of PETN Homologs in the stockpile here at LLNL is currently carried out by high performance liquid chromatography (HPLC) with ultra violet (UV) detection. Identification of unknown chromatographic peaks with this detection scheme is severely limited. The design agency is aware of the limitations of this methodology and ordered this study to develop instrumental parameters for the use of a currently owned mass spectrometer (MS) as the detection system. The resulting procedure would be a ''drop-in'' replacement for the current surveillance method (ERD04-524). The addition of quadrupole mass spectrometry provides qualitative identification of PETN and its homologs (Petrin, DiPEHN, TriPEON, and TetraPEDN) using a LLNL generated database, while providing mass clues to the identity of unknown chromatographic peaks.

  15. Mass Spectrometry Data from the Biological MS Data and Software Distribution Center

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

    Anderson, Gordon

    The mass spectrometry capabilities at Pacific Northwest National Laboratory (PNNL) are primarily applied to biological research, with an emphasis on proteomics and metabolomics. Many of these cutting-edge mass spectrometry capabilities and bioinformatics methods are housed in the Department of Energy's Environmental Molecular Sciences Laboratory (EMSL), a national scientific user facility operated by PNNL. These capabilities have been developed and acquired through cooperation between the EMSL national scientific user program and PNNL programmatic research. At the website of the Biological MS Data and Software Distribution Center, the following resources are made available: PNNL-developed software tools and source code, PNNL-generated raw data and processed results, links to publications that used the data and results available on this site, and tutorials and user manuals. [taken from http://omics.pnl.gov/

  16. Elemental speciation in biomolecules by LC-ICP-MS with magnetic sector and collision cell instruments

    SciTech Connect (OSTI)

    Wang, Jin

    1999-11-08

    A methodology that can monitor and identify inorganic elements in biological and environmental systems was developed. Size exclusion chromatography (SEC) separates biomolecules, which are then nebulized by a microconcentric nebulizer. The resulting aerosol is desolved and introduced into either a high resolution ICP-MS device or a quadrupole device with a collision cell. Because of the high sensitivity and spectral resolution and high sample introduction efficiency, many unusual or difficult elements, such as Cr, Se, Cd and U, can be observed at ambient levels bound to proteins in human serum. These measurements are made in only a few minutes without preliminary isolation and preconcentration steps. Serum samples can be titrated with spikes of various elements to determine which proteins bind a given metal and oxidation state. Experiments concerning the effects of breaking disulfide linkages and denaturation on metal binding in proteins were also investigated. Elemental distribution in liver extract was also obtained.

  17. Application of Printed Circuit Board Technology to FT-ICR MS Analyzer Cell Construction and Prototyping

    SciTech Connect (OSTI)

    Leach, Franklin E.; Norheim, Randolph V.; Anderson, Gordon A.; Pasa-Tolic, Ljiljana

    2014-12-01

    Although Fourier transform ion cyclotron resonance mass spectrometry (FT-ICRMS) remains themass spectrometry platform that provides the highest levels of performance for mass accuracy and resolving power, there is room for improvement in analyzer cell design as the ideal quadrupolar trapping potential has yet to be generated for a broadband MS experiment. To this end, analyzer cell designs have improved since the field’s inception, yet few research groups participate in this area because of the high cost of instrumentation efforts. As a step towards reducing this barrier to participation and allowing for more designs to be physically tested, we introduce a method of FT-ICR analyzer cell prototyping utilizing printed circuit boards at modest vacuum conditions. This method allows for inexpensive devices to be readily fabricated and tested over short intervals and should open the field to laboratories lacking or unable to access high performance machine shop facilities because of the required financial investment.

  18. Studies on the content of heavy metals in Aries River using ICP-MS

    SciTech Connect (OSTI)

    Voica, Cezara Kovacs, Melinda Feher, Ioana

    2013-11-13

    Among the industrial branches, the mining industry has always been an important source of environmental pollution, both aesthetically and chemically. Through this paper results of ICP-MS characterization of Aries River Basin are reported. Mining activities from this area has resulted in contamination of environment and its surrounding biota. This is clearly evidenced in analyzed water samples, especially from Baia de Aries site where increased amount of trace elements as Cr, Zn, As, Se, Cd, Pb and U were founded. Also in this site greater amount of rare earth elements was evidenced also. Through monitoring of Aries River from other non-mining area it was observed that the quantitative content of heavy metals was below the maximum permissible levels which made us to conclude that the water table wasn't seriously affected (which possibly might be attributed to the cessation of mining activities in this area from a few years ago)

  19. Profiling of Adrenocorticotropic Hormone and Arginine Vasopressin in Human Pituitary Gland and Tumor Thin Tissue Sections using Droplet-Based Liquid Microjunction Surface Sampling-HPLC-ESI-MS/MS

    SciTech Connect (OSTI)

    Kertesz, Vilmos; Van Berkel, Gary J; Calligaris, David; Feldman, Daniel R; Changelian, Armen; Laws, Edward R; Santagata, Sandro; Agar, Nathalie YR

    2015-01-01

    Described here are the results from the profiling of the proteins arginine vasopressin (AVP) and adrenocorticotropic hormone (ACTH) from normal human pituitary gland and pituitary adenoma tissue sections using a fully automated droplet-based liquid microjunction surface sampling-HPLC-ESI-MS/MS system for spatially resolved sampling, HPLC separation, and mass spectral detection. Excellent correlation was found between the protein distribution data obtained with this droplet-based liquid microjunction surface sampling-HPLC-ESI-MS/MS system and those data obtained with matrix assisted laser desorption ionization (MALDI) chemical imaging analyses of serial sections of the same tissue. The protein distributions correlated with the visible anatomic pattern of the pituitary gland. AVP was most abundant in the posterior pituitary gland region (neurohypophysis) and ATCH was dominant in the anterior pituitary gland region (adenohypophysis). The relative amounts of AVP and ACTH sampled from a series of ACTH secreting and non-secreting pituitary adenomas correlated with histopathological evaluation. ACTH was readily detected at significantly higher levels in regions of ACTH secreting adenomas and in normal anterior adenohypophysis compared to non-secreting adenoma and neurohypophysis. AVP was mostly detected in normal neurohypophysis as anticipated. This work demonstrates that a fully automated droplet-based liquid microjunction surface sampling system coupled to HPLC-ESI-MS/MS can be readily used for spatially resolved sampling, separation, detection, and semi-quantitation of physiologically-relevant peptide and protein hormones, such as AVP and ACTH, directly from human tissue. In addition, the relative simplicity, rapidity and specificity of the current methodology support the potential of this basic technology with further advancement for assisting surgical decision-making.

  20. A comparison of the y-Radiolysis of TODGA and T(EH)DGA using UHPLC-ESI-MS analysis

    SciTech Connect (OSTI)

    Zarzana, Christopher A.; Groenewold, Gary S.; Mincher, Bruce J.; Mezyk, Stephen P.; Wilden, Andreas; Schmidt, Holger; Modolo, Giuseppe; Wishart, James F.; Cook, Andrew R.

    2015-04-27

    Solutions of the diglycolamide extractants TODGA and T(EH)DGA in n-dodecane were subjected to γ- irradiation in the presence and absence of an acidic aqueous phase. These solutions were then analyzed using UHPLC-ESI-MS to determine the rates of radiolytic decay of the two extractants neat and in contact with respect to the acidity of the contacted aqueous phase, as well as to identify radiolysis products. The presence or absence of an acidic aqueous phase was shown to have no influence on the measured decay rates, nor did the side-chain have an influence. A number of radiolysis products were identified, consistent with those previously identified for these two compounds using GC-MS. The identity of these radiolysis products suggests that the bonds most vulnerable to radiolytic attack are those in the dyglycolamide center of these molecules, and not on the side-chains.  The agreement of these results with previous work using GC-MS indicates supports the further use of UHPLC-ESI-MS as a tool for studying diglycolamide extractant systems.

  1. A comparison of the y-Radiolysis of TODGA and T(EH)DGA using UHPLC-ESI-MS analysis

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

    Zarzana, Christopher A.; Groenewold, Gary S.; Mincher, Bruce J.; Mezyk, Stephen P.; Wilden, Andreas; Schmidt, Holger; Modolo, Giuseppe; Wishart, James F.; Cook, Andrew R.

    2015-04-27

    Solutions of the diglycolamide extractants TODGA and T(EH)DGA in n-dodecane were subjected to γ- irradiation in the presence and absence of an acidic aqueous phase. These solutions were then analyzed using UHPLC-ESI-MS to determine the rates of radiolytic decay of the two extractants neat and in contact with respect to the acidity of the contacted aqueous phase, as well as to identify radiolysis products. The presence or absence of an acidic aqueous phase was shown to have no influence on the measured decay rates, nor did the side-chain have an influence. A number of radiolysis products were identified, consistent with thosemore » previously identified for these two compounds using GC-MS. The identity of these radiolysis products suggests that the bonds most vulnerable to radiolytic attack are those in the dyglycolamide center of these molecules, and not on the side-chains.  The agreement of these results with previous work using GC-MS indicates supports the further use of UHPLC-ESI-MS as a tool for studying diglycolamide extractant systems.« less

  2. A comparison of the y-Radiolysis of TODGA and T(EH)DGA using UHPLC-ESI-MS analysis

    SciTech Connect (OSTI)

    Zarzana, Christopher A.; Groenewold, Gary S.; Mincher, Bruce J.; Mezyk, Stephen P.; Wilden, Andreas; Schmidt, Holger; Modolo, Giuseppe; Wishart, James F.; Cook, Andrew R.

    2015-04-27

    Solutions of the diglycolamide extractants TODGA and T(EH)DGA in n-dodecane were subjected to?- irradiation in the presence and absence of an acidic aqueous phase. These solutions were then analyzed using UHPLC-ESI-MS to determine the rates of radiolytic decay of the two extractants neat and in contact with respect to the acidity of the contacted aqueous phase, as well as to identify radiolysis products. The presence or absence of an acidic aqueous phase was shown to have no influence on the measured decay rates, nor did the side-chain have an influence. A number of radiolysis products were identified, consistent with those previously identified for these two compounds using GC-MS. The identity of these radiolysis products suggests that the bonds most vulnerable to radiolytic attack are those in the dyglycolamide center of these molecules, and not on the side-chains.The agreement of these results with previous work using GC-MS indicates supports the further use of UHPLC-ESI-MS as a tool for studying diglycolamide extractant systems.

  3. Characterization of Fatty Acids in Crenarchaeota by GC-MS and NMR

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

    Hamerly, Timothy; Tripet, Brian; Wurch, Louie; Hettich, Robert L.; Podar, Mircea; Bothner, Brian; Copié, Valérie

    2015-01-01

    Lipids composed of condensed isoprenyl units connected to glycerol backbones by ether linkages are a distinguishing feature of Archaea. Data suggesting that fatty acids with linear hydrocarbon chains are present in some Archaea have been available for decades. However, lack of genomic and biochemical evidence for the metabolic machinery required to synthesize and degrade fatty acids has left the field unclear on this potentially significant biochemical aspect. Because lipids are energy currency and cell signaling molecules, their presence in Archaea is significant for understanding archaeal biology. A recent large-scale bioinformatics analysis reignited the debate as to the importance ofmore » fatty acids in Archaea by presenting genetic evidence for the presence of enzymes required for anabolic and catabolic fatty acid metabolism across the archaeal domain. Here, we present direct biochemical evidence from gas chromatography-mass spectrometry (GC-MS) and nuclear magnetic resonance (NMR) spectroscopy for the presence of fatty acids in two members of the Crenarchaeota, Sulfolobus solfataricus and Ignicoccus hospitalis . This is the first report providing biochemical data for the existence of fatty acids in these Crenarchaeota, opening new discussions on energy balance and the potential for the discovery of new thermostable enzymes for industry.« less

  4. Surface Cleaning Techniques: Ultra-Trace ICP-MS Sample Preparation and Assay of HDPE

    SciTech Connect (OSTI)

    Overman, Nicole R.; Hoppe, Eric W.; Addleman, Raymond S.

    2013-06-01

    The world’s most sensitive radiation detection and assay systems depend upon ultra-low background (ULB) materials to reduce unwanted radiological backgrounds. Herein, we evaluate methods to clean HDPE, a material of interest to ULB systems and the means to provide rapid assay of surface and bulk contamination. ULB level material and ultra-trace level detection of actinide elements is difficult to attain, due to the introduction of contamination from sample preparation equipment such as pipette tips, sample vials, forceps, etc. and airborne particulate. To date, literature available on the cleaning of such polymeric materials and equipment for ULB applications and ultra-trace analyses is limited. For these reasons, a study has been performed to identify an effective way to remove surface contamination from polymers in an effort to provide improved instrumental detection limits. Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) was utilized to assess the effectiveness of a variety of leachate solutions for removal of inorganic uranium and thorium surface contamination from polymers, specifically high density polyethylene (HDPE). HDPE leaching procedures were tested to optimize contaminant removal of thorium and uranium. Calibration curves for thorium and uranium ranged from 15 ppq (fg/mL) to 1 ppt (pg/mL). Detection limits were calculated at 6 ppq for uranium and 7 ppq for thorium. Results showed the most effective leaching reagent to be clean 6 M nitric acid for 72 hour exposures. Contamination levels for uranium and thorium found in the leachate solutions were significant for ultralow level radiation detection applications.

  5. I

    Office of Legacy Management (LM)

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  6. 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/.

  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.

    2015-03-07

    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.

  8. 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.

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

    SciTech Connect (OSTI)

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

    2008-07-08

    Uraninite (UO2) and metaschoepite (UO32H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21C and 50C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.0040.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21C than the particles prepared at 50C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Winter 2013 Working Groups

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

    1 :00---2:00pm; M SE C onference R oom ( 3062 H H D ow) January 1 7 Jimmy Chen ( Phillips g roup) February 7 Michael K uo ( Ku g roup) February 2 8 Vladimir S toica (note: l...

  15. 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.

  16. 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

  17. 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

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

    SciTech Connect (OSTI)

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

    2014-01-22

    Uranyl phosphate solids are often found with uranium ores, and their low solubility makes them promising target phases for in situ remediation of uranium-contaminated subsurface environments. The products and solubility of uranium(VI) precipitated with phosphate can be affected by the pH, dissolved inorganic carbon (DIC) concentration, and co-solute composition (e.g. Na+/Ca2+) of the groundwater. Batch experiments were performed to study the effect of these parameters on the products and extent of uranium precipitation induced by phosphate addition. In the absence of co-solute cations, chernikovite [H3O(UO2)(PO4)3H2O] precipitated despite uranyl orthophosphate [(UO2)3(PO4)24H2O] being thermodynamically more favorable under certain conditions. As determined using X-ray diffraction, electron microscopy, and laser induced fluorescence spectroscopy, the presence of Na+ or Ca2+ as a co-solute led to the precipitation of sodium autunite ([Na2(UO2)2(PO4)2] and autunite [Ca(UO2)2(PO4)2]), which are structurally similar to chernikovite. In the presence of sodium, the dissolved U(VI) concentrations were generally in agreement with equilibrium predictions of sodium autunite solubility. However, in the calcium-containing systems, the observed concentrations were below the predicted solubility of autunite, suggesting the possibility of uranium adsorption to or incorporation in a calcium phosphate precipitate in addition to the precipitation of autunite.

  19. 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.

  20. 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.

  1. 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.

  2. Project Final Report: Ubiquitous Computing and Monitoring System (UCoMS) for Discovery and Management of Energy Resources

    SciTech Connect (OSTI)

    Tzeng, Nian-Feng; White, Christopher D.; Moreman, Douglas

    2012-07-14

    The UCoMS research cluster has spearheaded three research areas since August 2004, including wireless and sensor networks, Grid computing, and petroleum applications. The primary goals of UCoMS research are three-fold: (1) creating new knowledge to push forward the technology forefronts on pertinent research on the computing and monitoring aspects of energy resource management, (2) developing and disseminating software codes and toolkits for the research community and the public, and (3) establishing system prototypes and testbeds for evaluating innovative techniques and methods. Substantial progress and diverse accomplishment have been made by research investigators in their respective areas of expertise cooperatively on such topics as sensors and sensor networks, wireless communication and systems, computational Grids, particularly relevant to petroleum applications.

  3. Ms. Elizabeth Withers, EIS Document Manager U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration (NNSA)

    August 14, 2002 Ms. Elizabeth Withers, EIS Document Manager U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA) Office of Los Alamos Site Operations (OLASO) 528 35 th Street Los Alamos, NM 87544 Re: Comments regarding the scope of the Environmental Impact Statement (EIS) for the proposed Chemistry and Metallurgy Research (CMR) Building Replacement (CMRR) Project (CMRRP) at Los Alamos National Laboratory (LANL) (the "CMRRP EIS") Dear Elizabeth - I won't be

  4. Dehydration of Uranyl Nitrate Hexahydrate to Uranyl Nitrate Trihydrate under Ambient Conditions as Observed via Dynamic Infrared Reflectance Spectroscopy

    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.

    2015-05-22

    the hexahydrate [UO2(NO3)2(H2O)6] (UNH) and the trihydrate [UO2(NO3)2(H2O)3] (UNT) forms. Their stabilities depend on both relative humidity and temperature. Both phases have previously been studied by infrared transmission spectroscopy, but the data were limited by both instrumental resolution and the ability to prepare the samples as pellets without desiccating it. We report time-resolved infrared (IR) measurements using an integrating sphere that allow us to observe the transformation from the hexahydrate to the trihydrate simply by flowing dry nitrogen gas over the sample. Hexahydrate samples were prepared and confirmed via known XRD patterns, then measured in reflectance mode. The hexahydrate has a distinct uranyl asymmetric stretch band at 949.0 cm-1 that shifts to shorter wavelengths and broadens as the sample dehydrates and recrystallizes to the trihydrate, first as a blue edge shoulder but ultimately resulting in a doublet band with reflectance peaks at 966 and 957 cm-1. The data are consistent with transformation from UNH to UNT since UNT has two non-equivalent UO22+ sites. The dehydration of UO2(NO3)2(H2O)6 to UO2(NO3)2(H2O)3 is both a morphological and structural change that has the lustrous lime green crystals changing to the dull greenish yellow of the trihydrate. Crystal structures and phase transformation were confirmed theoretically using DFT calculations and experimentally via microscopy methods. Both methods showed a transformation with two distinct sites for the uranyl cation in the trihydrate, as opposed to a single crystallographic site in the hexahydrate.

  5. Benchmarking the New RESRAD-OFFSITE Source Term Model with DUST-MS and GoldSim - 13377

    SciTech Connect (OSTI)

    Cheng, J.J.; Kamboj, S.; Gnanapragasam, E.; Yu, C.

    2013-07-01

    RESRAD-OFFSITE is a computer code developed by Argonne National Laboratory under the sponsorship of U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC). It is designed on the basis of RESRAD (onsite) code, a computer code designated by DOE and NRC for evaluating soil-contaminated sites for compliance with human health protection requirements pertaining to license termination or environmental remediation. RESRAD-OFFSITE has enhanced capabilities of modeling radionuclide transport to offsite locations and calculating potential radiation exposure to offsite receptors. Recently, a new source term model was incorporated into RESRAD-OFFSITE to enhance its capability further. This new source term model allows simulation of radionuclide releases from different waste forms, in addition to the soil sources originally considered in RESRAD (onsite) and RESRAD-OFFSITE codes. With this new source term model, a variety of applications can be achieved by using RESRAD-OFFSITE, including but not limited to, assessing the performance of radioactive waste disposal facilities. This paper presents the comparison of radionuclide release rates calculated by the new source term model of RESRAD-OFFSITE versus those calculated by DUST-MS and GoldSim, respectively. The focus of comparison is on the release rates of radionuclides from the bottom of the contaminated zone that was assumed to contain radioactive source materials buried in soil. The transport of released contaminants outside of the primary contaminated zone is beyond the scope of this paper. Overall, the agreement between the RESRAD-OFFSITE results and the DUST-MS and GoldSim results is fairly good, with all three codes predicting identical or similar radionuclide release profiles over time. Numerical dispersion in the DUST-MS and GoldSim results was identified as potentially contributing to the disagreement in the release rates. In general, greater discrepancy in the release rates was found for short-lived, fast-moving radionuclides than for long-lived, slow-moving radionuclides. (authors)

  6. A WSRC-MS-g8-00318 Heat Transfer Model of Above and Underground Insulated Piping

    Office of Scientific and Technical Information (OSTI)

    WSRC-MS-g8-00318 Heat Transfer Model of Above and Underground Insulated Piping Systems by K. C. Kwon Westinghouse Savannah River Company Savannah River Site Aiken, South Carolina 29808 A document prepared for ASME CONFERENCE - HEAT EXCHANGER COMMITTEE MEETING 8 , INTERNATIONAL JOINT POWER GENERATION CONFERENCE 1998 at Baltimore, MA, USA from 8/23/98 - 8/26/98. DOE Contract No. DE-AC09-96SR18500 This paper was prepared in connection with work done under the above contract number with the U. S.

  7. NanoLC-FT-ICR MS improves proteome coverage attainable for ~3000 laser microdissected breast carcinoma cells

    SciTech Connect (OSTI)

    Umar, Arzu; Luider, Theo N.; Foekens, J. A.; Pasa-Tolic, Liljiana

    2007-01-29

    Genomics and proteomics assays hold great promise for unrevealing molecular events that underlie human disease. Essential to this quest is the ability to effectively analyze clinical samples, but this task is considerably complicated by tissue heterogeneity. Laser capture microdissection (LCM) can be used to selectively isolate targeted cell populations (such as tumor cells) from their native tissue environment. However, the small number of cells that are typically procured by LCM severely limits the proteome coverage and biomarker discovery potential achievable by conventional proteomics platforms. Herein, we report on the use of a nano liquid chromatography-Fourier transform ion clyclotron resonance mass spectrometry (nLC-FTICR MS) platform for analyzing protein digests of approximately 3,000 LCM-derived tumor cells from breast carcinoma tissue, which corresponds to approximately 300 ng of total protein. A total of 2,836 peptides were identified by matching LC-MS data to accurate mass and time (AMT) tag databases that were previously established for the human mammary epithelium and several breast cancer cell lines. The peptide identifications correspond to 1,139 unique proteins confidently identified with 2 or more peptides. Based on categorization by Gene Ontology, identified proteins appear to cover a wide variety of biological functions and cellular compartments. This work demonstrates that a substantial number of proteins can be identified from a limited number of cells using the AMT tag approach and opens a door for high throughput in-depth proteomics analysis of clinical samples.

  8. A Comparison of the ?-Radiolysis of TODGA and T(EH)DGA Using UHPLC-ESI-MS Analysis

    SciTech Connect (OSTI)

    Chris A. Zarzana; Gary S. Groenewold; Bruce J. Mincher; Stephen P. Mezyk; Giuseppe Modolo; Andreas Wildens; Holgar Schmidt; James F. Wishart; Andrew R. Cook

    2015-03-01

    Solutions of TODGA and T(EH)DGA in n-dodecane were subjected to ?-irradiation in the presence and absence of an aqueous nitric acid phase and analyzed using UHPLC-ESI-MS to determine the rates of radiolytic decay of the two extractants, as well as to identify radiolysis products. The DGA concentrations decreased exponentially with increasing dose, and the measured degradation rate constants were uninfluenced by the presence or absence of an acidic aqueous phase, or by chemical variations in the alkyl side-chains. The DGA degradation was attributed to reactions of the dodecane radical cation, whose kinetics were measured for TODGA using picosecond electron pulse radiolysis to be k2 = (9.72 1.10) 109 M-1 s-1. The identified radiolysis products suggest that the bonds most vulnerable to radiolytic attack are those in the diglycolamide center of these molecules and not on the side-chains.

  9. Study of organic compounds evolved during the co-firing of coal and refuse derived fuel using TG/MS

    SciTech Connect (OSTI)

    Puroshothama, Shobha; Lu, R.; Yang, Xiaodong

    1996-10-01

    The evolution of organic compounds during the combustion of carbonaceous fuel coupled with solid waste disposal and limited landfill space has been a cause for concern. Co-firing high sulfur coal with refuse derived fuel seems an attractive alternative technique to tackle the dual problem of controlling SO{sub x} emissions as well as those of the chlorinated organic toxins. The TG serves to emulate the conditions of the fluidized bed combustor and the MS serves as the detector for evolved gases. This versatile combination is used to study the decomposition pathway as well as predict the conditions at which various compounds are formed and may serve as a means of reducing the formation of these chlorinated organic compounds.

  10. Identification and decay of the 0.48 ms 13/2{sup +} isomer in {sup 181}Hg

    SciTech Connect (OSTI)

    Andreyev, A. N.; Antalic, S.; Saro, S.; Ackermann, D.; Comas, V. F.; Heinz, S.; Heredia, J. A.; Hessberger, F. P.; Khuyagbaatar, J.; Kojouharov, I.; Kindler, B.; Lommel, B.; Mann, R.; Cocolios, T. E.; Elseviers, J.; Huyse, M.; Duppen, P. Van; Venhart, M.; Franchoo, S.; Hofmann, S.

    2009-10-15

    A new isomer with a half-life of 0.48(2) ms was identified in the nuclide {sup 181}Hg, which was produced in the complete fusion reaction {sup 40}Ca+{sup 144}Sm{yields}{sup 184}Pb* at the velocity filter SHIP (GSI, Darmstadt). The isomeric state was tentatively assigned a spin-parity of 13/2{sup +}. We propose that this isomer de-excites by a yet unobserved low-energy, strongly converted {gamma}-ray transition, followed by a newly identified cascade composed of a 90.3 keV M1 and a 71.4 keV E2 {gamma}-ray transition.

  11. 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}=117.194(1), {beta}=113.551(1), {gamma}=92.216(1){sup o}; (isoPrU4V3) monoclinic, P2{sub 1}/n space group, a=10.3507(1), b=13.6500(2), c=17.3035(2) A, {beta}=97.551(1){sup o}; (TMetU4V3) orthorhombic, Pbca space group, a=17.1819(2), b=13.6931(1), c=21.4826(2) A; (CHexU4V3), triclinic P-1 space group, a=9.8273(6), b=11.0294(7), c=12.7506(8) A, {alpha}=98.461(3), {beta}=96.437(3), {gamma}=105.955(3){sup o}; (TButU4V3), monoclinic, P2{sub 1}/m space group, a=9.8048(4), b=17.4567(8), c=15.4820(6) A, {beta}=106.103(2). - Graphical abstract: The various type of PBP pillars P2, P3, P4, and P4' in the three-dimensional inorganic frameworks based on the uranophane-type sheet in monoamine templated uranyl-vanadates.

  12. Absolute calibration of Kodak Biomax-MS film response to x rays in the 1.5- to 8-keV energy range

    SciTech Connect (OSTI)

    Marshall, F. J.; Knauer, J. P.; Anderson, D.; Schmitt, B. L

    2006-10-15

    The absolute response of Kodak Biomax-MS film to x rays in the range from 1.5- to 8-keV has been measured using a laboratory electron-beam generated x-ray source. The measurements were taken at specific line energies by using Bragg diffraction to produce monochromatic beams of x rays. Multiple exposures were taken on Biomax MS film up to levels exceeding optical densities of 2 as measured by a microdensitometer. The absolute beam intensity for each exposure was measured with a Si(Li) detector. Additional response measurements were taken with Kodak direct exposure film (DEF) so as to compare the results of this technique to previously published calibrations. The Biomax-MS results have been fitted to a semiempirical mathematical model (Knauer et al., these proceedings). Users of the model can infer absolute fluences from observed exposure levels at either interpolated or extrapolated energies. To summarize the results: Biomax MS has comparable sensitivity to DEF film below 3 keV but has reduced sensitivity above 3 keV ({approx}50%). The lower exposure results from thinner emulsion layers, designed for use with phosphor screens. The ease with which Biomax-MS can be used in place of DEF (same format film, same developing process, and comparable sensitivity) makes it a good replacement.

  13. A comparison of the γ-radiolysis of TODGA and T(EH)DGA using UHPLC-ESI-MS analysis

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

    Chris A. Zarzana; Cook, Andrew R.; Groenewold, Gary S.; Mincher, Bruce J.; Mezyk, Stephen P.; Wilden, Andreas; Schmidt, Holger; Modolo, Giuseppe; Wishart, James F.

    2015-03-11

    Solutions of TODGA and T(EH)DGA in n-dodecane were subjected to γ-irradiation in the presence and absence of an aqueous nitric acid phase and analyzed using UHPLC-ESI-MS to determine the rates of radiolytic decay of the two extractants, as well as to identify radiolysis products. The DGA concentrations decreased exponentially with increasing dose, and the measured degradation rate constants were uninfluenced by the presence or absence of an acidic aqueous phase, or by chemical variations in the alkyl side-chains. The DGA degradation was attributed to reactions of the dodecane radical cation, whose kinetics were measured for TODGA using picosecond electron pulsemore » radiolysis to be k2 = (9.72 ± 1.10) × 109 M–1 s–1. Furthermore, the identified radiolysis products suggest that the bonds most vulnerable to radiolytic attack are those in the diglycolamide center of these molecules and not on the side-chains.« less

  14. Distributed computing strategies for processing of FT-ICR MS imaging datasets for continuous mode data visualization

    SciTech Connect (OSTI)

    Smith, Donald F.; Schulz, Carl; Konijnenburg, Marco; Kilic, Mehmet; Heeren, Ronald M.

    2015-03-01

    High-resolution Fourier transform ion cyclotron resonance (FT-ICR) mass spectrometry imaging enables the spatial mapping and identification of biomolecules from complex surfaces. The need for long time-domain transients, and thus large raw file sizes, results in a large amount of raw data (“big data”) that must be processed efficiently and rapidly. This can be compounded by largearea imaging and/or high spatial resolution imaging. For FT-ICR, data processing and data reduction must not compromise the high mass resolution afforded by the mass spectrometer. The continuous mode “Mosaic Datacube” approach allows high mass resolution visualization (0.001 Da) of mass spectrometry imaging data, but requires additional processing as compared to featurebased processing. We describe the use of distributed computing for processing of FT-ICR MS imaging datasets with generation of continuous mode Mosaic Datacubes for high mass resolution visualization. An eight-fold improvement in processing time is demonstrated using a Dutch nationally available cloud service.

  15. LA-23336-MS

    Office of Scientific and Technical Information (OSTI)

    Materials for Liquid-Lead and Lead-Bismuth Eutectic Spallation Neutron Source John ... Oxide scales were formed on the surface of the sample. Figure 2 shows the weight changes ...

  16. Ms. Maggie Owen, Chair

    Office of Environmental Management (EM)

    Myron Iwanski, Anderson County Mayor Mark Watson, Oak Ridge city Mayor Ron Woody, Roan ... FORS David Adler, EM-92 Melyssa Noe, EM-95 Joy Sager, EM-91 Susan M.Cange, Acting Manager ...

  17. LA-23336-MS

    Office of Scientific and Technical Information (OSTI)

    ... Hours to Severe Corrosion Temperature Range ("C)

  18. LA-5052-MS

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

    2271 CHEMISTRY-GENERAL TID-4500, 14th Ed. LOS ALAMOS SCIENTIFIC LABORATORY OF THE UNIVERSITY OF CALIFORNIA LOS ALAMOS NEW MEXICO REPORT WRITTEN: August 1958 REPORT DISTRIBUTED: March 17, 1959 COMPRESSIBILITY FACTORS AND FUGACITY COEFFICIENTS CALCULATED FROM THE BEATTIE-BRIDGEMAN EQUATION OF STATE FOR HYDROGEN, NITROGEN, OXYGEN, CARBON DIOXIDE, AMMONIA, METHANE, AND HELIUM by C. E. Holley, J r . W. J. Worlton R. K. Zeigler » * This report expresses the opinions of the author or authors and does

  19. MS FORTRAN Extended Libraries

    Energy Science and Technology Software Center (OSTI)

    1986-09-01

    DISPPAK is a set of routines for use with Microsoft FORTRAN programs that allows the flexible display of information on the screen of an IBM PC in both text and graphics modes. The text mode routines allow the cursor to be placed at an arbitrary point on the screen and text to be displayed at the cursor location, making it possible to create menus and other structured displays. A routine to set the color ofmore » the characters that these routines display is also provided. A set of line drawing routines is included for use with IBM''s Color Graphics Adapter or an equivalent board (such as the Enhanced Graphics Adapter in CGA emulation mode). These routines support both pixel coordinates and a user-specified set of real number coordinates. SUBPAK is a function library which allows Microsoft FORTRAN programs to calculate random numbers, issue calls to the operating system, read individual characters from the keyboard, perform Boolean and shift operations, and communicate with the I/O ports of the IBM PC. In addition, peek and poke routines, a routine that returns the address of any variable, and routines that can access the system time and date are included.« less

  20. Is LA-12152-MS

    Office of Scientific and Technical Information (OSTI)

    ... Treaty Organization Advisory Group for Aerospace Research and Development (December ... B. Hildebrand, Advanced Calculus for Applications (Prentice-Hall, Inc., Englewood ...

  1. LA-11224-MS

    Office of Scientific and Technical Information (OSTI)

    FEHM: FINITE ELEMENT HEAT AND MASS TRANSFER CODE George l y v o l o s k i , Zora Dash, and Sharad K e l k a r ABSTRACT The f i n i t e element heat and mass (FEHM) t r a n s f e r ...

  2. LA-11224-MS

    Office of Scientific and Technical Information (OSTI)

    i m u l a t e geothermal and h o t d r y r o c k r e s e r v o i r s . I t i s a l s o a p p l i c a b l e t o n a t u r a l - s t a t e s t u d i e s o f geothermal systems and ...

  3. LA-10256-MS

    Office of Legacy Management (LM)

    ... CHUPADERA MESA IN SITU 37Cs DATA FOR SOILS BY LAND FORM ... IB-27B (1976). Water Above Natural Background," DOE Order ... -1.5 to 2 miles east of a gas pump house 2052 71477; ...

  4. Ms. Maria Galanti

    Energy Savers [EERE]

    ... Silt fencing was installed to control sediment runoff, and sediment traps were installed at potentially-impacted catch basins. An X-533 Stormwater Pollution Prevention Plan was ...

  5. Ms. Margaret Owen, Chair

    Office of Environmental Management (EM)

    that applauds the Department of Energy's (DOE) Environmental Management (EM) cleanup work and recommends the identification of "unique assets" for retention with community input into those decisions. Although no specific assets were mentioned in the recommendation, I hope that you have made your concerns known to your local EM Site-Specific Advisory Board (EM SSAB) contacts and site managers. EM does evaluate potential reuse of assets as it plans its cleanup efforts. Decisions

  6. Ms. Margaret Owen, Chair

    Office of Environmental Management (EM)

    recommending that the Department of Energy (DOE) consider using Federal transport and/or disposition funds to relocate designated cultural/historic property items to outside organizations, when the organizations are unable to fund the relocation themselves. As you summarized in your letter, it is the current policy of DOE and Office of Environmental Management (EM) to donate property with cultural or historic value when that property no longer has reuse value to the Government, has been deemed

  7. 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.; 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 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. 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.

  8. Electrochemical behavior of simulated debris from a severe accident using a molten salt system

    SciTech Connect (OSTI)

    Takahashi, Yuya; Nakamura, Hitoshi; Yamada, Akira; Mizuguchi, Koji; Fujita, Reiko

    2013-07-01

    In a severe nuclear accident, the fuel in the reactor may melt, forming debris, which contains a UO{sub 2}-ZrO{sub 2} stable oxide mixture and parts of the reactor, such as Zircaloy and iron components. Proper handling of the debris is a critically important issue. The debris does not have the same composition as spent fuel, and so it is impossible to apply conventional reprocessing technology directly. In this study, we successfully separated Zr and Fe from simulated debris using NaCl-KCl molten salt electrolysis, and we selectively recovered the Zr and Fe. The simulated debris was made from Zr, Fe, and CeO{sub 2}. The CeO{sub 2} was used for simulating stable UO{sub 2}-ZrO{sub 2}. With this approach, it should be possible to reduce the volume of the debris by recovering metals, which can then be treated as low level radioactive wastes.

  9. AGR-2 Irradiation Test Final As-Run Report, Rev 2

    SciTech Connect (OSTI)

    Blaise Collin

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO2 fuel, while fast fluence values ranged from 1.94 to 3.47´1025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53´1025 n/m2 (E >0.18 MeV) for UO2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10-6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2´10-6. In the UO2 capsule (Capsule 3), the R/B values during the first three cycles were below 10-7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

  10. AGR-2 irradiation test final as-run report, Rev. 1

    SciTech Connect (OSTI)

    Collin, Blaise

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities; (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing; and, (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO2 fuel, while fast fluence values ranged from 1.94 to 3.47´1025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53´1025 n/m2 (E >0.18 MeV) for UO2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10-6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2´10-6. In the UO2 capsule (Capsule 3), the R/B values during the first three cycles were below 10-7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

  11. AGR-2 IRRADIATION TEST FINAL AS-RUN REPORT

    SciTech Connect (OSTI)

    Blaise, Collin

    2014-07-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO2 fuel, while fast fluence values ranged from 1.94 to 3.47´1025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53´1025 n/m2 (E >0.18 MeV) for UO2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10-6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2´10-6. In the UO2 capsule (Capsule 3), the R/B values during the first three cycles were below 10-7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

  12. 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)

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

  14. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    SciTech Connect (OSTI)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K.; Oomori, T.

    2013-07-01

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

  15. Method for fluorination of uranium oxide

    DOE Patents [OSTI]

    Petit, George S. (Oak Ridge, TN)

    1987-01-01

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

  16. Fuel System Compatibility Issues for Prometheus-1

    SciTech Connect (OSTI)

    DC Noe; KB Gibbard; MH Krohn

    2006-01-20

    Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO{sub 2} as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO{sub 2}-based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined.

  17. Biogeochemical Mechanisms Controlling Reduced Radionuclide Particle Properties and Stability

    SciTech Connect (OSTI)

    Jim K. Fredrickson; John M. Zachara; Matthew J. Marshall; Alex S. Beliaev

    2006-06-01

    Uranium and Technetium are the major risk-driving contaminants at Hanford and other DOE sites. These radionuclides have been shown to be reduced by dissimilatory metal reducing bacteria (DMRB) under anoxic conditions. Laboratory studies have demonstrated that reduction results in the formation of poorly soluble hydrous oxides, UO2(s) and TcO2n?H2O(s), that are believed to limit mobility in the environment. The mechanisms of microbial reduction of U and Tc have been the focus of considerable research in the Environmental Remediation Sciences Program (ERSP). In spite of equal or greater importance in terms of controlling the environmental fate of the contaminants relatively little is known regarding the precipitation mechanism(s), reactivity, persistence, and transport of biogenic UO2(s) and TcO2(s).

  18. Relocation and freezing of liquefied fuel-rod material. [PWR

    SciTech Connect (OSTI)

    Moore, R.L.; Broughton, J.M.

    1982-01-01

    Severe degraded core cooling accidents, such as occurred at TMI-2 can potentially reach temperatures in excess of cladding melting. When the molten cladding is in contact with UO/sub 2/ fuel, the UO/sub 2/ will be dissolved contributing significantly to the total amount of liquefied material flowing down the rod and eventually freezing in a lower, cooler region of the core. The primary objectives of this paper are to evaluate the relocation and freezing characteristics of liquefied fuel rod material over a wide range of system conditions, physical characteristics of the fuel rod and liquefied material, and material thermo-physical properties to determine the relative influence of the controlling parameters. First the analytical model used in the analysis is briefly reviewed. The results of the analyses are then presented and discussed, and this is followed by the conclusions.

  19. METHOD FOR DECONTAMINATION OF REACTOR SOLUTIONS

    DOE Patents [OSTI]

    Maraman, W.J.; Baxman, H.R.; Baker, R.D.

    1959-05-01

    A process for U recovery from phosphate fuel solutions is described. To fuel solution drawn from the reactor is added Fe(NO/sub 3/)/sub 3/ which destroys the U complex and forms ferric phosphate complex. The UO/sub 2/(NO/sub 3/)/sub 2/ formed is extracted into TBP-kerosene in a countercurrent column. The TBP contalning UO/sub 2/(NO/sub 3/)/sub 2/ is further purified by an aqueous Al(NO/ sub 3/)/sub 3/ scrub solution. The pregnant solution then goes to an H/sub 3/PO/ sub 4/ stripping and kerosene washing column. The H/sub 3/PO/sub 4/--uranyl phosphate solution is separated at the bottom and boiled to remove HNO/sub 3/ then diluted to fuel solution make-up strength. (T.R.H.)

  20. Selectivity in ligand binding to uranyl compounds: A synthetic, structural, thermodynamic and computational study

    SciTech Connect (OSTI)

    Arnold, John

    2015-01-21

    The uranyl cation (UO??) is the most abundant form of uranium on the planet. It is estimated that 4.5 billion tons of uranium in this form exist in sea water. The ability to bind and extract the uranyl cation from aqueous solution while separating it from other elements would provide a limitless source of nuclear fuel. A large body of research concerns the selective recognition and extraction of uranyl. A stable molecule, the cation has a linear O=U=O geometry. The short U-O bonds (1.78 ) arise from the combination of uranium 5f/6d and oxygen 2p orbitals. Due to the oxygen moieties being multiply bonded, these sites were not thought to be basic enough for Lewis acidic coordination to be a viable approach to sequestration.

  1. Aerosols released during large-scale integral MCCI tests in the ACE Program

    SciTech Connect (OSTI)

    Fink, J.K.; Thompson, D.H.; Spencer, B.W.; Sehgal, B.R.

    1992-04-01

    As part of the internationally sponsored Advanced Containment Experiments (ACE) program, seven large-scale experiments on molten core concrete interactions (MCCIs) have been performed at Argonne National Laboratory. One of the objectives of these experiments is to collect and characterize all the aerosols released from the MCCIs. Aerosols released from experiments using four types of concrete (siliceous, limestone/common sand, serpentine, and limestone/limestone) and a range of metal oxidation for both BWR and PWR reactor core material have been collected and characterized. Release fractions were determined for UO{sup 2}, Zr, the fission-products: BaO, SrO, La{sub 2}O{sub 3}, CeO{sub 2}, MoO{sub 2}, Te, Ru, and control materials: Ag, In, and B{sub 4}C. Release fractions of UO{sub 2} and the fission products other than Te were small in all tests. However, release of control materials was significant.

  2. Aerosols released during large-scale integral MCCI tests in the ACE Program

    SciTech Connect (OSTI)

    Fink, J.K.; Thompson, D.H.; Spencer, B.W. ); Sehgal, B.R. )

    1992-01-01

    As part of the internationally sponsored Advanced Containment Experiments (ACE) program, seven large-scale experiments on molten core concrete interactions (MCCIs) have been performed at Argonne National Laboratory. One of the objectives of these experiments is to collect and characterize all the aerosols released from the MCCIs. Aerosols released from experiments using four types of concrete (siliceous, limestone/common sand, serpentine, and limestone/limestone) and a range of metal oxidation for both BWR and PWR reactor core material have been collected and characterized. Release fractions were determined for UO{sup 2}, Zr, the fission-products: BaO, SrO, La{sub 2}O{sub 3}, CeO{sub 2}, MoO{sub 2}, Te, Ru, and control materials: Ag, In, and B{sub 4}C. Release fractions of UO{sub 2} and the fission products other than Te were small in all tests. However, release of control materials was significant.

  3. Enhancing Biological Analyses with Three Dimensional Field Asymmetric Ion Mobility, Low Field Drift Time Ion Mobility and Mass Spectrometry (FAIMS/IMS-MS) Separations

    SciTech Connect (OSTI)

    Zhang, Xing; Ibrahim, Yehia M.; Chen, Tsung-Chi; Kyle, Jennifer E.; Norheim, Randolph V.; Monroe, Matthew E.; Smith, Richard D.; Baker, Erin Shammel

    2015-06-30

    We report the first evaluation of a platform coupling a high speed field asymmetric ion mobility spectrometry microchip (FAIMS) with drift tube ion mobility and mass spectrometry (IMS-MS). The FAIMS/IMS-MS platform was used to analyze biological samples and simultaneously acquire multidimensional information of detected features from the measured FAIMS compensation fields and IMS drift times, while also obtaining accurate ion masses. These separations thereby increase the overall separation power, resulting increased information content, and provide more complete characterization of more complex samples. The separation conditions were optimized for sensitivity and resolving power by the selection of gas compositions and pressures in the FAIMS and IMS separation stages. The resulting performance provided three dimensional separations, benefitting both broad complex mixture studies and targeted analyses by e.g. improving isomeric separations and allowing detection of species obscured by chemical noise and other interfering peaks.

  4. Absolute Calibration of Kodak Biomax-MS Film Response to X Rays in the 1.5- to 8-keV Energy Range

    SciTech Connect (OSTI)

    Marshall, F.J.; Knauer, J.P.; Anderson, D.; Schmitt, B.L.

    2006-09-28

    The absolute response of Kodak Biomax-MS film to x rays in the range from 1.5- to 8-keV has been measured using a laboratory e-beam generated x-ray source. The measurements were taken at specific line energies by using Bragg diffraction to produce monochromatic beams of x rays. Multiple exposures were taken on Biomax MS film up to levels exceeding optical densities of 2 as measured by a microdensitometer. The absolute beam intensity for each exposure was measured with a Si(Li) detector. Additional response measurements were taken with Kodak direct exposure film (DEF) so as to compare the results of this technique to previously published calibrations.

  5. Fluorescence spectra and biological activity of aerosolized bacillus spores and MS2 bacteriophage exposed to ozone at different relative humidities in a rotating drum

    SciTech Connect (OSTI)

    Ratnesar-Shumate, Shanna; Pan, Yong-Le; Hill, Steven C.; Kinahan, Sean; Corson, Elizabeth; Eshbaugh, Jonathan; Santarpia, Joshua L.

    2015-10-14

    Biological aerosols (bioaerosols) released into the environment may undergo physical and chemical transformations when exposed to atmospheric constituents such as solar irradiation, reactive oxygenated species, ozone, free radicals, water vapor and pollutants. Aging experiments were performed in a rotating drum chamber subjecting bioaerosols, Bacillus thuringiensis Al Hakam (BtAH) spores and MS2 bacteriophages to ozone at 0 and 150 ppb, and relative humidities (RH) at 10%, 50%, and 80+%. Fluorescence spectra and intensities of the aerosols as a function of time in the reaction chamber were measured with a single particle fluorescence spectrometer (SPFS) and an Ultra-Violet Aerodynamic Particle Sizer® Spectrometer (UV-APS). Losses in biological activity were measured by culture and quantitative polymerase chain reaction (q-PCR) assay. For both types of aerosols the largest change in fluorescence emission was between 280 and 400 nm when excited at 263 nm followed by fluorescence emission between 380 and 700 nm when excited at 351 nm. The fluorescence for both BtAH and MS2 were observed to decrease significantly at high ozone concentration and high RH when excited at 263 nm excitation. The decreases in 263 nm excited fluorescence are indicative of hydrolysis and oxidation of tryptophan in the aerosols. Fluorescence measured with the UV-APS (355-nm excitation) increased with time for both BtAH and MS2 aerosols. A two log loss of MS2 bacteriophage infectivity was observed in the presence of ozone at ~50% and 80% RH when measured by culture and normalized for physical losses by q-PCR. Viability of BtAH spores after exposure could not be measured due to the loss of genomic material during experiments, suggesting degradation of extracelluar DNA attributable to oxidation. The results of these studies indicate that the physical and biological properties of bioaerosols change significantly after exposure to ozone and water vapor.

  6. Fluorescence spectra and biological activity of aerosolized bacillus spores and MS2 bacteriophage exposed to ozone at different relative humidities in a rotating drum

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

    Ratnesar-Shumate, Shanna; Pan, Yong-Le; Hill, Steven C.; Kinahan, Sean; Corson, Elizabeth; Eshbaugh, Jonathan; Santarpia, Joshua L.

    2015-10-14

    Biological aerosols (bioaerosols) released into the environment may undergo physical and chemical transformations when exposed to atmospheric constituents such as solar irradiation, reactive oxygenated species, ozone, free radicals, water vapor and pollutants. Aging experiments were performed in a rotating drum chamber subjecting bioaerosols, Bacillus thuringiensis Al Hakam (BtAH) spores and MS2 bacteriophages to ozone at 0 and 150 ppb, and relative humidities (RH) at 10%, 50%, and 80+%. Fluorescence spectra and intensities of the aerosols as a function of time in the reaction chamber were measured with a single particle fluorescence spectrometer (SPFS) and an Ultra-Violet Aerodynamic Particle Sizer® Spectrometermore » (UV-APS). Losses in biological activity were measured by culture and quantitative polymerase chain reaction (q-PCR) assay. For both types of aerosols the largest change in fluorescence emission was between 280 and 400 nm when excited at 263 nm followed by fluorescence emission between 380 and 700 nm when excited at 351 nm. The fluorescence for both BtAH and MS2 were observed to decrease significantly at high ozone concentration and high RH when excited at 263 nm excitation. The decreases in 263 nm excited fluorescence are indicative of hydrolysis and oxidation of tryptophan in the aerosols. Fluorescence measured with the UV-APS (355-nm excitation) increased with time for both BtAH and MS2 aerosols. A two log loss of MS2 bacteriophage infectivity was observed in the presence of ozone at ~50% and 80% RH when measured by culture and normalized for physical losses by q-PCR. Viability of BtAH spores after exposure could not be measured due to the loss of genomic material during experiments, suggesting degradation of extracelluar DNA attributable to oxidation. The results of these studies indicate that the physical and biological properties of bioaerosols change significantly after exposure to ozone and water vapor.« less

  7. Energy Frontier Research Center Center for Materials Science of Nuclear

    Office of Scientific and Technical Information (OSTI)

    Fuels (Technical Report) | SciTech Connect Frontier Research Center Center for Materials Science of Nuclear Fuels Citation Details In-Document Search Title: Energy Frontier Research Center Center for Materials Science of Nuclear Fuels Scientific Successes * The first phonon density of states (PDOS) measurements for UO2 to include anharmonicity were obtained using time-of-flight inelastic neutron scattering at the Spallation Neutron Source (SNS), and an innovative, experimental-based

  8. Universal fuel basket for use with an improved oxide reduction vessel and electrorefiner vessel

    DOE Patents [OSTI]

    Herrmann, Steven D.; Mariani, Robert D.

    2002-01-01

    A basket, for use in the reduction of UO.sub.2 to uranium metal and in the electrorefining of uranium metal, having a continuous annulus between inner and outer perforated cylindrical walls, with a screen adjacent to each wall. A substantially solid bottom and top plate enclose the continuous annulus defining a fuel bed. A plurality of scrapers are mounted adjacent to the outer wall extending longitudinally thereof, and there is a mechanism enabling the basket to be transported remotely.

  9. Energy Frontier Research Centers

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

    Fuels (Technical Report) | SciTech Connect Technical Report: Energy Frontier Research Center Center for Materials Science of Nuclear Fuels Citation Details In-Document Search Title: Energy Frontier Research Center Center for Materials Science of Nuclear Fuels Scientific Successes * The first phonon density of states (PDOS) measurements for UO2 to include anharmonicity were obtained using time-of-flight inelastic neutron scattering at the Spallation Neutron Source (SNS), and an innovative,

  10. AFV CoverSheet

    Office of Scientific and Technical Information (OSTI)

    7701 (Accepted Manuscript) Possible Demonstration of a Polaronic Bose-Einstein(-Mott) Condensate in UO2(+x) by Ultrafast THz Spectroscopy and Microwave Dissipation Conradson, Steven D; Gilbertson, Steve Michael; Daifuku, Stephanie L; Kehl, Jeffrey A; Durakiewicz, Tomasz; Andersson, Anders David Ragnar; Bishop, Alan; Byler, Darrin David; Maldonado, Pablo; Oppeneer, Peter; Valdez, James Anthony; Neidig, Michael L; Rodriguez, George Provided by the author(s) and the Los Alamos National Laboratory

  11. Ultrasound enhanced process for extracting metal species in supercritical fluids

    DOE Patents [OSTI]

    Wai, Chien M.; Enokida, Youichi

    2006-10-31

    Improved methods for the extraction or dissolution of metals, metalloids or their oxides, especially lanthanides, actinides, uranium or their oxides, into supercritical solvents containing an extractant are disclosed. The disclosed embodiments specifically include enhancing the extraction or dissolution efficiency with ultrasound. The present methods allow the direct, efficient dissolution of UO2 or other uranium oxides without generating any waste stream or by-products.

  12. LANSCE | Lujan Center | Sample and Equipment Shipping Instructions

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

    Sample Shipping Hazardous Nanoparticles Radioactive, Nuclear, Special Nuclear Materials Contacts Lujan Center Leader Aaron Couture (acting) 505.667.1730 Deputy Leader Fredrik Tovesson 505.665.9652 Deputy Leader & Experimental Area Manager Charles Kelsey 505.665.5579 Experiment Coordinator Charles Kelsey (acting) 505.667.8755 User Program Administration lujan-uo@lanl.gov Administrative Assistant Julie Quintana-Valdez 505.665.5390 Department of Energy, National Nuclear Security Administration

  13. LANSCE | Lujan Center | Tips for Writing Beamtime Proposals

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

    Aaron Couture (acting) 505.667.1730 Deputy Leader Fredrik Tovesson 505.665.9652 Deputy Leader & Experimental Area Manager Charles Kelsey 505.665.5579 Experiment Coordinator Charles Kelsey (acting) 505.667.8755 User Program Administration lujan-uo@lanl.gov Administrative Assistant Julie Quintana-Valdez 505.665.5390 Department of Energy, National Nuclear Security Administration nnsa.energy.gov Tips for Writing Beamtime Proposals LANSCE User Resources Tips for a Successful Proposal Proposals

  14. LCLS Users' Organization Executive Committee | Linac Coherent Light Source

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

    Users' Organization Executive Committee SAVE THE DATE: SSRL/LCLS Users' Conference and Workshops, October 5-7, 2016 Read summary of 2015 users' conference. During the annual meeting, users also have the opportunity to vote for their Users Executive Committee Representatives. The LCLS Users' Organization (LCLS UO) provides an organized framework and independent vehicle for interaction between the scientists who are interested in using the Linac Coherent Light Source (the users) and LCLS/SLAC

  15. Assessment of Homogeneous Thorium/Uranium Fuel for Pressurized Water

    Office of Scientific and Technical Information (OSTI)

    Reactors (Journal Article) | SciTech Connect Assessment of Homogeneous Thorium/Uranium Fuel for Pressurized Water Reactors Citation Details In-Document Search Title: Assessment of Homogeneous Thorium/Uranium Fuel for Pressurized Water Reactors The homogeneous ThO{sub 2}-UO{sub 2} fuel cycle option for a pressurized water reactor (PWR) of current technology is investigated. The fuel cycle assessment was carried out by calculating the main performance parameters: natural uranium and separative

  16. Criticality safety evaluation report for the 100 KE Basin sandfilter backwash pit

    SciTech Connect (OSTI)

    Erickson, D.G.

    1995-01-01

    This analysis presents the technical basis for establishing a safe mass limit for continued operations of the KE Basin sandfilter backwash pit. The main analysis is based on a very conservative UO{sub 2}-PuO{sub 2}-H{sub 2}O system using the measured isotopic concentrations of uranium and plutonium in the sludge. Appendix C contains analyses of the sandfilter backwash pit utilizing all verified materials presently in the pit, and gives new limits based on assumptions made.

  17. Energy Frontier Research Center Center for Materials Science of Nuclear

    Office of Scientific and Technical Information (OSTI)

    Fuels (Technical Report) | SciTech Connect Technical Report: Energy Frontier Research Center Center for Materials Science of Nuclear Fuels Citation Details In-Document Search Title: Energy Frontier Research Center Center for Materials Science of Nuclear Fuels Scientific Successes * The first phonon density of states (PDOS) measurements for UO2 to include anharmonicity were obtained using time-of-flight inelastic neutron scattering at the Spallation Neutron Source (SNS), and an innovative,

  18. Vegetation

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

    Vegetation 250 o 250 N A Community _ Loblolly Pine D Bottomland Hardwood I!!!I Carolina Bay Wetland _ Bottomland HardwodlPine W Streams ~ Roads A/; Rails [2] SRS Bays Will Hydric Soils 500 Meters Soils Soil Series and Phase D DoA D DoB DRm rn Uo Figure 24-1. Plant COll/llll/lzities and soils associated with the Cypress Bay Set-Aside Area. sc 24-5 Set-Aside 24: Cypress Bay

  19. Draft report on melt point as a function of composition for urania-based systems

    SciTech Connect (OSTI)

    Valdez, James A; Byler, Darrin D

    2012-06-08

    This report documents the testing of a urania (UO{sub 2.00}) sample as a baseline and the attempt to determine the melt point associated with 4 compositions of urania-ceria and urania-neodymia pseudo binaries provided by ORNL, with compositions of 95/5, and 80/20 and of (U/Ce)O{sub 2.00} and (U/Nd)O{sub 2.00} in the newly developed ceramic melt point determination system. A redesign of the system using parts fabricated from tungsten was undertaken in order to help prevent contamination and tungsten carbide formation in the crucibles. The previously developed system employed mostly graphite parts that were shown to react with the sample containment black-body crucible leading to unstable temperature readings and crucible failure, thus the redesign. Measured melt point values of UO{sub 2.00} and U{sub 0.95}Ce{sub 0.05}O{sub 2.00}, U{sub 0.80}Ce{sub 0.20}O{sub 2.00}, U{sub 0.95}Nd{sub 0.05}O{sub 2.00} and U{sub 0.80}Nd{sub 0.20}O{sub 2.00} were measured using a 2-color pyrometer. The value measured for UO{sub 2.00} was consistent with the published accepted value 2845 C {+-} 25 C, although a wide range of values has been published by researchers and will be discussed later in the text. For comparison, values obtained from a published binary phase diagram of UO{sub 2}-Nd{sub 2}O{sub 3} were used for comparison with our measure values. No literature melt point values for comparison with the measurements performed in this study were found for (U/Ce)O{sub 2.00} in our stoichiometry range.

  20. Electrochemistry of LiCl-Li2O-H2O Molten Salt Systems

    SciTech Connect (OSTI)

    Natalie J. Gese; Batric Pesic

    2013-03-01

    Uranium can be recovered from uranium oxide (UO2) spent fuel through the combination of the oxide reduction and electrorefining processes. During oxide reduction, the spent fuel is introduced to molten LiCl-Li2O salt at 650 degrees C and the UO2 is reduced to uranium metal via two routes: (1) electrochemically, and (2) chemically by lithium metal (Li0) that is produced electrochemically. However, the hygroscopic nature of both LiCl and Li2O leads to the formation of LiOH, contributing hydroxyl anions (OH-), the reduction of which interferes with the Li0 generation required for the chemical reduction of UO2. In order for the oxide reduction process to be an effective method for the treatment of uranium oxide fuel, the role of moisture in the LiCl-Li2O system must be understood. The behavior of moisture in the LiCl-Li2O molten salt system was studied using cyclic voltammetry, chronopotentiometry and chronoamperometry, while reduction to hydrogen was confirmed with gas chromatography.

  1. PRODUCTION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-08-27

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

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

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-09-10

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

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

    SciTech Connect (OSTI)

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

    2015-04-01

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

  4. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    SciTech Connect (OSTI)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  5. Radionuclide release from spent fuel under geologic disposal conditions: An overview of experimental and theoretical work through 1985

    SciTech Connect (OSTI)

    Reimus, P.W.; Simonson, S.A.

    1988-04-01

    This report presents an overview of experimental and theoretical work on radionuclide release from spent fuel and uranium dioxide (UO/sub 2/) under geologic disposal conditions. The purpose of the report is to provide a source book of information that can be used to develop models that describe radionuclide release from spent fuel waste packages. Modeling activities of this nature will be conducted within the Waste Package Program (WPP) of the Department of Energy's Salt Repository Project (SRP). The topics discussed include experimental methods for investigating radionuclide release, how results have been reported from radionuclide release experiments, theoretical studies of UO/sub 2/ and actinide solubility, results of experimental studies of radionuclide release from spent fuel and UO/sub 2/ (i.e., the effects of different variables on radionuclide release), characteristics of spent fuel pertinent to radionuclide release, and status of modeling of radionuclide release from spent fuel. Appendix A presents tables of data from spent fuel radionuclide release experiments. These data have been digitized from graphs that appear in the literature. An annotated bibliography of literature on spent fuel characterization is provided in Appendix B.

  6. The thermal conductivity of mixed fuel UxPu1-xO2: molecular dynamics simulations

    SciTech Connect (OSTI)

    Liu, Xiang-Yang; Cooper, Michael William Donald; Stanek, Christopher Richard; Andersson, Anders David Ragnar

    2015-10-16

    Mixed oxides (MOX), in the context of nuclear fuels, are a mixture of the oxides of heavy actinide elements such as uranium, plutonium and thorium. The interest in the UO2-PuO2 system arises from the fact that these oxides are used both in fast breeder reactors (FBRs) as well as in pressurized water reactors (PWRs). The thermal conductivity of UO2 fuel is an important material property that affects fuel performance since it is the key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. For this reason it is important to understand the thermal conductivity of MOX fuel and how it differs from UO2. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of mixing on the thermal conductivity of UxPu1-xO2, as a function of PuO2 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.

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

    SciTech Connect (OSTI)

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

    1994-08-01

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

  8. Molecular Dynamics Simulation of Thermodynamic Properties in Uranium Dioxide

    SciTech Connect (OSTI)

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

    2014-03-01

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

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

    SciTech Connect (OSTI)

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

    2004-05-12

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

  10. Multiscale Simulation of Thermo-mechancial Processes in Irradiated Fission-reactor Materials.

    SciTech Connect (OSTI)

    Simon R. Phillpot

    2012-06-08

    The work funded from this project has been published in six papers, with two more in draft form, with submission planned for the near future. The papers are: (1) Kinetically-Evolving Irradiation-Induced Point-Defect Clusters in UO{sub 2} by Molecular-Dynamics Simulation; (2) Kinetically driven point-defect clustering in irradiated MgO by molecular-dynamics simulation; (3) Grain-Boundary Source/Sink Behavior for Point Defect: An Atomistic Simulation Study; (4) Energetics of intrinsic point defects in uranium dioxide from electronic structure calculations; (5) Thermodynamics of fission products in UO{sub 2{+-}x}; and (6) Atomistic study of grain boundary sink strength under prolonged electron irradiation. The other two pieces of work that are currently being written-up for publication are: (1) Effect of Pores and He Bubbles on the Thermal Transport Properties of UO2 by Molecular Dynamics Simulation; and (2) Segregation of Ruthenium to Edge Dislocations in Uranium Dioxide.

  11. Simulation of the effects of grain boundary fission gas during thermal transients

    SciTech Connect (OSTI)

    Fenske, G.R.; Emerson, J.E.; Beiersdorf, B.A.

    1984-11-01

    This report presents the results of an initial set of out-of-cell transient heating experiments performed on unirradiated UO/sub 2/ pellets fabricated to simulate the effect of grain boundary fission gas on fuel swelling and cladding failure. The fabrication involved trapping high-pressure argon on internal pores by sintering annular UO/sub 2/ pellets in a hot isostatic press (HIP). The pellet stack was subjected to two separate transients (DGF83-03A and -03B). Figures show photomicrographs of HIPped and non-HIPped UO/sub 2/, respectively, and the adjacent cladding after DGF83-03B. Fuel melting occurred at the center of both the HIPped and non-HIPped pellets; however, a dark ring is present near the center in the HIPped fuel but not in the non-HIPped fuel. This dark band is a high-porosity region due to increased grain boundary/edge swelling in that pellet. In contrast, grain boundary/edge swelling did not occur in the non-HIPped pellets. Thus, the presence of the high-pressure argon trapped on internal pores during sintering in the HIP altered the microstructural behavior. Results of these preliminary tests indicate that the microstructural behavior of HIPped fuel during thermal transients is different from the behavior of conventionally fabricated fuel.

  12. Conceptual Design of a CERMET NTR Fission Core Using Multiphysics Modeling Techniques

    SciTech Connect (OSTI)

    Jonathan A. Webb; Brian J. Gross; William T. Taitano

    2011-08-01

    An initial pre-conceptual CERMET Nuclear Thermal Propulsion reactor system is investigated within this paper. Reactor configurations are investigated where the fuel consists of 60 vol.% UO2 and 40 vol.% W where the UO2 consists of Gd2O3 concentrations of 5 and 10 mol.%.Gd2O3. The fuel configuration consisting of 5 mol.% UO2 was found to have a total mass of 2761 kg and a thrust to weight ratio of 4.10 and required a coolant channel surface area to fueled volume ratio of approximately 15.0 in order to keep the centerline temperature below 3000 K. The configuration consisting of 10 mol.% Gd2O3 required a surface area to volume ratio of approximately 12.2 to cool the reactor to a peak temperature of 3000 K and had a total mass of 3200 kg and a thrust to weight ratio of 3.54. It is not known yet what concentration of Gd2O3 is required to maintain fuel stability at 3000 K; however, both reactors offer the potential for operations at 25,000 lb, and at a specific impulse which may range from 900 to 950 seconds.

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

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

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

    2015-04-13

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

  14. Experimental Results for SimFuels

    SciTech Connect (OSTI)

    Buck, Edgar C.; Casella, Andrew M.; Skomurski, Frances N.; MacFarlan, Paul J.; Soderquist, Chuck Z.; Wittman, Richard S.; Mcnamara, Bruce K.

    2012-08-22

    Assessing the performance of Spent (or Used) Nuclear Fuel (UNF) in geological repository requires quantification of time-dependent phenomena that may influence its behavior on a time-scale up to millions of years. A high-level waste repository environment will be a dynamic redox system because of the time-dependent generation of radiolytic oxidants and reductants and the corrosion of Fe-bearing canister materials. One major difference between used fuel and natural analogues, including unirradiated UO2, is the intense radiolytic field. The radiation emitted by used fuel can produce radiolysis products in the presence of water vapor or a thin-film of water that may increase the waste form degradation rate and change radionuclide behavior. To study UNF, we have been working on producing synthetic UO2 ceramics, or SimFuels that can be used in testing and which will contain specific radionuclides or non-radioactive analogs so that we can test the impact of radiolysis on fuel corrosion without using actual spent fuel. Although, testing actual UNF would be ideal for understanding the long term behavior of UNF, it requires the use of hot cells and is extremely expensive. In this report, we discuss, factors influencing the preparation of SimFuels and the requirements for dopants to mimic the behavior of UNF. We have developed a reliable procedure for producing large grain UO2 at moderate temperatures. This process will be applied to a series of different formulations.

  15. On the photo and thermally stimulated luminescence properties of U doped SrBPO{sub 5}

    SciTech Connect (OSTI)

    Kumar, Mithlesh Mohapatra, M.; Natarajan, V.

    2014-12-15

    Highlights: Synthesis of SrBPO{sub 5}:U phosphor by solid state route. Confirmed the stabilization of uranium as UO{sub 2}{sup 2+}. Evaluation of order of kinetics and trap parameters of the system. ESR-TSL correlation of the observed glow peak. Probable mechanism proposed for the TSL glow peak. - Abstract: Un-doped and uranium doped SrBPO{sub 5} samples were synthesized using solid-state reaction route and investigated for their photo and luminescence properties. Photoluminescence (PL) spectrum of uranium doped sample showed five peaks at 502, 524, 547, 569 and 597 nm. The average frequency of symmetric stretching of O=U=O in the ground electronic state was found to be about 757 cm{sup ?1}. PL decay time measurements on the system confirmed the stabilization of uranium as UO{sub 2}{sup 2+} in the matrix. Thermally stimulated luminescence (TSL) measurements carried out on gamma irradiated SrBPO{sub 5}:U sample showed a glow peak at 390 K, whose spectral characteristics was found to be typical of UO{sub 2}{sup 2+}. The trap parameters were evaluated using different heating rate method. Room temperature EPR data suggested the formation of borate and oxygen based radical centers in the gamma-irradiated sample. Detailed EPR-TSL correlation studies confirmed the destruction of the oxygen radical to be responsible for the observed glow peak.

  16. Final assessment of MOX fuel performance experiment with Japanese PWR specification fuel in the HBWR

    SciTech Connect (OSTI)

    Fujii, Hajime; Teshima, Hideyuki; Kanasugi, Katsumasa; Kosaka, Yuji; Arakawa, Yasushi

    2007-07-01

    In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pu-fissile enrichment of about 6 wt% had been irradiated in the HBWR from 1995 to 2006. The peak burn-up of MOX pellet achieved 72 GWd/tM. In this test, fuel centerline temperature, rod internal pressure, stack length and cladding length were measured for MOX fuel and UO{sub 2} fuel as reference. MOX fuel temperature is confirmed to have no significant difference in comparison with UO{sub 2}, taking into account of adequate thermal conductivity degradation due to PuO{sub 2} addition and burn-up development. And the measured fuel temperature agrees well with FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly greater than UO{sub 2} based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behavior. MOX fuel swelling rate agrees well with solid swelling rate in the literature. Cladding elongation data shows onset of PCMI in high power region. (authors)

  17. Subcellular-level resolution MALDI-MS imaging of maize leaf metabolites by MALDI-linear ion trap-Orbitrap mass spectrometer

    SciTech Connect (OSTI)

    Korte, Andrew R.; Yandeau-Nelson, Marna D.; Nikolau, Basil J.; Lee, Young Jin

    2015-01-25

    A significant limiting factor in achieving high spatial resolution for matrix-assisted laser desorption ionization-mass spectrometry (MALDI-MS) imaging is the size of the laser spot at the sample surface. We present modifications to the beam-delivery optics of a commercial MALDI-linear ion trap-Orbitrap instrument, incorporating an external Nd:YAG laser, beam-shaping optics, and an aspheric focusing lens, to reduce the minimum laser spot size from ~50 ?m for the commercial configuration down to ~9 ?m for the modified configuration. This improved system was applied for MALDI-MS imaging of cross sections of juvenile maize leaves at 5-?m spatial resolution using an oversampling method. There are a variety of different metabolites including amino acids, glycerolipids, and defense-related compounds were imaged at a spatial resolution well below the size of a single cell. Such images provide unprecedented insights into the metabolism associated with the different tissue types of the maize leaf, which is known to asymmetrically distribute the reactions of C4 photosynthesis among the mesophyll and bundle sheath cell types. The metabolite ion images correlate with the optical images that reveal the structures of the different tissues, and previously known and newly revealed asymmetric metabolic features are observed.

  18. The use of DRIFTS-MS and kinetic studies to determine the role of acetic acid in the palladium-catalyzed vapor-phase synthesis of vinyl acetate

    SciTech Connect (OSTI)

    Augustine, S.M.; Blitz, J.P. (Quantum Chemical Corp., Cincinnati, OH (United States))

    1993-07-01

    Supported palladium catalyzes the synthesis of vinyl acetate (VA) by oxyacetylation of ethylene. Alkali promoters increase activity and selectivity. The role of acetic acid (HOAc) in these processes is not well understood. Activation energy studies show that HOAc alters the catalyst site and lowers the reaction barrier to VA formation. After correction for this effect, the kinetics reveal that as a reagent HOAc is zero order. This is probably due to a strong adsorption of HOAc and Pd which forms the catalyst active phase. Detailed spectroscopic studies support this conclusion. The surface processes on a supported vinyl acetate catalyst were studied using a method which couples diffuse reflectance infrared Fourier transform spectroscopy (DRIFTS) with mass spectrometry (MS). The DRIFTS-MS technique combines the capability of selectively analyzing IR-active surface species with sensitive detection of transient reaction products. By comparing the catalyst with mixtures of palladium acetate powder physically dispersed in potassium chloride, it is determined that the active phase on the catalyst is a form of palladium acetate. Compound formation is consistent with the strong chemisorption of HOAc on Pd. Kinetic analysis of temperature-programmed reaction(TPRxn) data suggests that Pd metal or metal oxide adjacent to the active site is important in the reaction mechanism. 25 refs., 10 figs., 2 tabs.

  19. Subcellular-level resolution MALDI-MS imaging of maize leaf metabolites by MALDI-linear ion trap-Orbitrap mass spectrometer

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

    Korte, Andrew R.; Yandeau-Nelson, Marna D.; Nikolau, Basil J.; Lee, Young Jin

    2015-01-25

    A significant limiting factor in achieving high spatial resolution for matrix-assisted laser desorption ionization-mass spectrometry (MALDI-MS) imaging is the size of the laser spot at the sample surface. We present modifications to the beam-delivery optics of a commercial MALDI-linear ion trap-Orbitrap instrument, incorporating an external Nd:YAG laser, beam-shaping optics, and an aspheric focusing lens, to reduce the minimum laser spot size from ~50 μm for the commercial configuration down to ~9 μm for the modified configuration. This improved system was applied for MALDI-MS imaging of cross sections of juvenile maize leaves at 5-μm spatial resolution using an oversampling method. Theremore » are a variety of different metabolites including amino acids, glycerolipids, and defense-related compounds were imaged at a spatial resolution well below the size of a single cell. Such images provide unprecedented insights into the metabolism associated with the different tissue types of the maize leaf, which is known to asymmetrically distribute the reactions of C4 photosynthesis among the mesophyll and bundle sheath cell types. The metabolite ion images correlate with the optical images that reveal the structures of the different tissues, and previously known and newly revealed asymmetric metabolic features are observed.« less

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

    SciTech Connect (OSTI)

    Rasmussen B.

    2010-01-01

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

  1. Uranium diphosphonates templated by interlayer organic amines

    SciTech Connect (OSTI)

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

    2013-02-15

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

  2. On the formation of carbonyl sulfide in the reduction of sulfur dioxide by carbon monoxide on lanthanum oxysulfide catalyst: A study by XPS and TPR/MS

    SciTech Connect (OSTI)

    Lau, N.T.; Fang, M. [Hong Kong Univ. of Science and Technology, Clear Water Bay (Hong Kong). Applied Technology Center] [Hong Kong Univ. of Science and Technology, Clear Water Bay (Hong Kong). Applied Technology Center

    1998-10-25

    Both the X-ray photoelectron spectroscopy (XPS) and temperature-programmed reaction, coupled with mass spectrometry (TPR/MS), are used to study the formation of carbonyl sulfide in the reduction of sulfur dioxide on lanthanum oxysulfide catalyst. It was found that the lattice sulfur of the oxysulfide is released and reacts with carbon monoxide to form carbonyl sulfide when the oxysulfide is heated. The oxysulfide is postulated to form sulfur vacancies at a temperature lower than that for the formation of carbonyl sulfide and atomic sulfur is released in the process. The atomic sulfur can either enter the gas phase and leave the oxysulfide catalyst or react with carbon monoxide to form carbonyl sulfide.

  3. Profiling of adrenocorticotropic hormone and arginine vasopressin in human pituitary gland and tumor thin tissue sections using droplet-based liquid-microjunction surface-sampling-HPLC–ESI-MS–MS

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

    Kertesz, Vilmos; Calligaris, David; Feldman, Daniel R.; Changelian, Armen; Laws, Edward R.; Santagata, Sandro; Agar, Nathalie Y. R.; Van Berkel, Gary J.

    2015-06-18

    Described here are the results from the profiling of the proteins arginine vasopressin (AVP) and adrenocorticotropic hormone (ACTH) from normal human pituitary gland and pituitary adenoma tissue sections using a fully automated droplet-based liquid microjunction surface sampling-HPLC-ESI-MS/MS system for spatially resolved sampling, HPLC separation, and mass spectral detection. Excellent correlation was found between the protein distribution data obtained with this droplet-based liquid microjunction surface sampling-HPLC-ESI-MS/MS system and those data obtained with matrix assisted laser desorption ionization (MALDI) chemical imaging analyses of serial sections of the same tissue. The protein distributions correlated with the visible anatomic pattern of the pituitary gland.more » AVP was most abundant in the posterior pituitary gland region (neurohypophysis) and ATCH was dominant in the anterior pituitary gland region (adenohypophysis). The relative amounts of AVP and ACTH sampled from a series of ACTH secreting and non-secreting pituitary adenomas correlated with histopathological evaluation. ACTH was readily detected at significantly higher levels in regions of ACTH secreting adenomas and in normal anterior adenohypophysis compared to non-secreting adenoma and neurohypophysis. AVP was mostly detected in normal neurohypophysis as anticipated. This work demonstrates that a fully automated droplet-based liquid microjunction surface sampling system coupled to HPLC-ESI-MS/MS can be readily used for spatially resolved sampling, separation, detection, and semi-quantitation of physiologically-relevant peptide and protein hormones, such as AVP and ACTH, directly from human tissue. In addition, the relative simplicity, rapidity and specificity of the current methodology support the potential of this basic technology with further advancement for assisting surgical decision-making.« less

  4. A user's guide to the GoldSim/BLT-MS integrated software package:a low-level radioactive waste disposal performance assessment model.

    SciTech Connect (OSTI)

    Knowlton, Robert G.; Arnold, Bill Walter; Mattie, Patrick D.

    2007-03-01

    Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in the assessment of radioactive waste disposal and at the time of this publication is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. In countries with small radioactive waste programs, international technology transfer program efforts are often hampered by small budgets, schedule constraints, and a lack of experienced personnel. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available software codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission (NRC) and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, revitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a credible and solid computational platform for constructing probabilistic safety assessment models. This document is a reference users guide for the GoldSim/BLT-MS integrated modeling software package developed as part of a cooperative technology transfer project between Sandia National Laboratories and the Institute of Nuclear Energy Research (INER) in Taiwan for the preliminary assessment of several candidate low-level waste repository sites. Breach, Leach, and Transport-Multiple Species (BLT-MS) is a U.S. NRC sponsored code which simulates release and transport of contaminants from a subsurface low-level waste disposal facility. GoldSim is commercially available probabilistic software package that has radionuclide transport capabilities. The following report guides a user through the steps necessary to use the integrated model and presents a successful application of the paradigm of renewing legacy codes for contemporary application.

  5. Identification of volatile butyl rubber thermal-oxidative degradation products by cryofocusing gas chromatography/mass spectrometry (cryo-GC/MS).

    SciTech Connect (OSTI)

    Smith, Jonell Nicole; White, Michael Irvin; Bernstein, Robert; Hochrein, James Michael

    2013-02-01

    Chemical structure and physical properties of materials, such as polymers, can be altered as aging progresses, which may result in a material that is ineffective for its envisioned intent. Butyl rubber formulations, starting material, and additives were aged under thermal-oxidative conditions for up to 413 total days at up to 124 %C2%B0C. Samples included: two formulations developed at Kansas City Plant (KCP) (%236 and %2310), one commercially available formulation (%2321), Laxness bromobutyl 2030 starting material, and two additives (polyethylene AC-617 and Vanax MBM). The low-molecular weight volatile thermal-oxidative degradation products that collected in the headspace over the samples were preconcentrated, separated, and detected using cryofocusing gas chromatography mass spectrometry (cryo-GC/MS). The majority of identified degradation species were alkanes, alkenes, alcohols, ketones, and aldehydes. Observations for Butyl %2310 aged in an oxygen-18 enriched atmosphere (18O2) were used to verify when the source of oxygen in the applicable degradation products was from the gaseous environment rather than the polymeric mixture. For comparison purposes, Butyl %2310 was also aged under non-oxidative thermal conditions using an argon atmosphere.

  6. Gilbert_etal_MS_FIGS

    Office of Scientific and Technical Information (OSTI)

    ... Magnetite nanoparticles can be formed by microbial activity in anaerobic subsurface ... J. F.; Nord, G. L.; Phillips, E. J. P., Anaerobic production of magnetite by a ...

  7. ARM - Instrument - ptr-ms

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

    Aerosols This instrument measures low concentrations of volatile organic compounds-important precursors in aerosol formation-in the atmosphere by positively ionizing the...

  8. Gilbert_etal_MS_FIGS

    Office of Scientific and Technical Information (OSTI)

    ... The absorption spectra were normalized for the energy dependence of incident beam intensity by the division of a gold grid electron yield signal acquired simultaneously with the ...

  9. LA-10733-MS UC-15

    Office of Scientific and Technical Information (OSTI)

    ... This equation shows how the plutonium mass fractions change with time. The denominator is a normalization factor because all +he fractions must sum to unity. Now one can use the ...

  10. MS_07_Number_14.pdf

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

    Mergedsounding VAP: Recent Enhancements David Troyan 1 , Michael Jensen 1 , David Turner 2 , Mark Miller 1 , Jennifer Delamere 3 , Eli Mlawer 3 , Gerald Mace 4 Author Affiliations ...

  11. Ms. Maria Galanti Site Coordinator

    Energy Savers [EERE]

    I AND II OF THE REMOVAL OF THE X-760 CHEMICAL ENGINEERING BUILDING AT THE PORTSMOUTH ... I and II ofthe Removal of the X-760 Chemical Engineering Building," dated May 26, ...

  12. Ms. Maria Galanti Site Coordinator

    Office of Environmental Management (EM)

    ... uranium LEU low-enriched uranium NCP ... Office OMB Office of Management and Budget OSDC ... United States Code WAC waste acceptance criteria This page is ...

  13. PHYSOR 2014 Template MS Word

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    ... 23 energy groups and on-the-fly pin-homogenized cross-sections generated from a 252-group ENDF BVII.0 SCALE library has ... calculation with a memory usage of 2.5 TB. The runtime ...

  14. PHYSOR 2014 Template MS Word

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    1 / 21 AP1000 ® PWR REACTOR PHYSICS ANALYSIS WITH VERA-CS AND KENO-VI - PART II: POWER DISTRIBUTION F. Franceschini Westinghouse Electric Co. LLC, Cranberry Township, Pennsylvania, USA francef@westinghouse.com A. T. Godfrey, J. C. Gehin Oak Ridge National Laboratory godfreyat@ornl.gov ABSTRACT Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Ap- plications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics

  15. Ms. Maria Galanti Site Coordinator

    Office of Environmental Management (EM)

    ......... 8 Figure 5. X-633-1 Pump House slab and transformer pads - piping ... The X-633 RCW Complex was comprised of the X-633-1 Pump House, which included a ...

  16. PHYSOR 2014 Template MS Word

    Office of Scientific and Technical Information (OSTI)

    ... The value thus obtained for this parameter is negative, -2.6 pcmK, which compares well with typical values for the LWRs. The coolant temperature coefficient of reactivity is also ...

  17. ARM - Datastreams - irt200ms

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

    KS (Extended) SGP E10 Browse Data Tyro, KS (Extended) retired SGP E11 Browse Data Byron, OK (Extended) SGP E12 Browse Data Pawhuska, OK (Extended) SGP E13 Browse Data Lamont, OK...

  18. PHYSOR 2014 Template MS Word

    Office of Scientific and Technical Information (OSTI)

    ... Cycle 1. This activity has led to the creation of a very rigorous quarter core model of ... the long com- putational runtimes and large data storage requirements for this analysis. ...

  19. MS_08_15.pdf

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

    and Changes to the Merged Sounding VAP David Troyan and Mike Jensen, BNL Jim Mather and Sally McFarlane, PNNL Eli Mlawer and Jennifer Delamere, AER, Inc. Mark Miller, Rutger University Dave Turner, University of Wisconsin Jay Mace, University of Utah Merged Sounding Data Availability SGP: NSA: TWP Manus: TWP Nauru: TWP Darwin: 2000 - 2005 2004 - 2007 2006 2004 - 2006 2002 - 2006 Two versions of Merged Sounding exist. The original version is continuing to be run for all permanent sites; the

  20. LA-10733-MS UC-15

    Office of Scientific and Technical Information (OSTI)

    Los Alamos National Laboratory Los Alamos, New Mexico 87545 Q l S T R l E U T l 0 M " m D O (tm) (tm) ls CONTENTS ABSTRACT 1 I. INTRODUCTION 1 II. STATEMENT OF THE PROBLEM 2 ...

  1. Bulk and surface controlled diffusion of fission gas atoms

    SciTech Connect (OSTI)

    Andersson, Anders D.

    2012-08-09

    Fission gas retention and release impact nuclear fuel performance by, e.g., causing fuel swelling leading to mechanical interaction with the clad, increasing the plenum pressure and reducing the gap thermal conductivity. All of these processes are important to understand in order to optimize operating conditions of nuclear reactors and to simulate accident scenarios. Most fission gases have low solubility in the fuel matrix, which is especially pronounced for large fission gas atoms such as Xe and Kr, and as a result there is a significant driving force for segregation of gas atoms to extended defects such as grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. Several empirical or semi-empirical models have been developed for fission gas release in nuclear fuels, e.g. [1-6]. One of the most commonly used models in fuel performance codes was published by Massih and Forsberg [3,4,6]. This model is similar to the early Booth model [1] in that it applies an equivalent sphere to separate bulk UO{sub 2} from grain boundaries represented by the sphere circumference. Compared to the Booth model, it also captures trapping at grain boundaries, fission gas resolution and it describes release from the boundary by applying timedependent boundary conditions to the circumference. In this work we focus on the step where fission gas atoms diffuse from the grain interior to the grain boundaries. The original Massih-Forsberg model describes this process by applying an effective diffusivity divided into three temperature regimes. In this report we present results from density functional theory calculations (DFT) that are relevant for the high (D{sub 3}) and intermediate (D{sub 2}) temperature diffusivities of fission gases. The results are validated by making a quantitative comparison to Turnbull's [8-10] and Matzke's data [12]. For the intrinsic or high temperature regime we report activation energies for both Xe and Kr diffusion in UO{sub 2{+-}x}, which compare favorably to available experiments. This is an extension of previous work [13]. In particular, it applies improved chemistry models for the UO{sub 2{+-}x} nonstoichiometry and its impact on the fission gas activation energies. The derivation of these models follows the approach that used in our recent study of uranium vacancy diffusion in UO{sub 2} [14]. Also, based on the calculated DFT data we analyze vacancy enhanced diffusion mechanisms in the intermediate temperature regime. In addition to vacancy enhanced diffusion we investigate species transport on the (111) UO{sub 2} surface. This is motivated by the formation of small voids partially filled with fission gas atoms (bubbles) in UO{sub 2} under irradiation, for which surface diffusion could be the rate-limiting transport step. Diffusion of such bubbles constitutes an alternative mechanism for mass transport in these materials.

  2. Validation of the WATEQ4 geochemical model for uranium

    SciTech Connect (OSTI)

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

    1983-09-01

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

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

    SciTech Connect (OSTI)

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

    2015-04-13

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

  4. Natural radionuclides in groundwaters from J-13 well at the Nevada Test Site

    SciTech Connect (OSTI)

    Laul, J.C.; Maiti, T.C.

    1990-04-01

    The concentrations of U-238 and Th-232 chain members are extremely low in J-13 water, suggesting that their concentrations in groundwaters are largely governed by sorption/desorption processes. Relative to radon (gas), uranium, thorium, radium, and polonium radionuclides are highly sorbed in a tuffaceous matrix, and the retardation factors range from 10{sup 2} to 10{sup 5}. Uranium, unlike Th, is in the +6 state and is soluble as carbonate complex (UO{sub 2}CO{sub 3}), and the aquifer`s environment is oxidizing. There is no colloidal effect down to <0.10 {mu}m. 15 refs., 1 fig., 2 tabs.

  5. Natural radionuclides in groundwater from J-13 well at the Nevada test site

    SciTech Connect (OSTI)

    Laul, J.C.; Maiti, T.C.

    1990-10-01

    This paper discusses the concentrations of U-238 and Th-232 chain members which are extremely low in J-13 water, suggesting that their concentrations in groundwaters are largely governed by sorption/desorption processes. Relative to radon (gas), uranium, thorium, radium, and polonium radionuclides are highly sorbed in a tuffaceous matrix, and the retardation factors range from 10{sup 2} to 10{sup 5}. Uranium, unlike Th, is in the + 6 state and is soluble as carbonate complex (UO{sub 2}CO{sub 3}), and the aquifer`s environment is oxidizing. There is no colloidal effect down to {lt}0.10 {mu}m.

  6. Proton Radiography at Los Alamos National Laboratory (pRad)

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

    pRad at LANL P-Division | LANSCE >> pRad Home About News Movies Proposals Contacts Team Leader Dale Tupa 505.665.1820 Project Leader Andy Saunders 505.665.3090 Area Manager (interim) Eron Kerstiens 505.667.3618 Fesseha Mariam 505.667.3546 Christopher Morris 505.667.5652 Frans Trouw 505.665.7575 pRad User Program pRad-uo@lanl.gov P-25 Group Leader Melynda Brooks 505.667.6909 P-25 Deputy Group Leader Frans Trouw 505.665.7575 P-25 Subatomic Physics P-Division LANSCE pRad logo Los Alamos

  7. SSRL HEADLINES August 2013

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

    2 - August 2013 View the Archives **Note for Outlook users: For easier reading, please click the bar at the top of this message that reads "This message was converted to plain text" and select "Display as HTML."** Science Highlights thumbnail Nanoparticulate FeS as an Effective Redox Buffer to Prevent Uraninite (UO2) Oxidation - Contacts: Yuqiang Bi and Kim F. Hayes, University of Michigan A major concern in the nuclear age is the contamination of soils and groundwater with

  8. Microsoft Word - (ThU)O2-manuscript-as accepted

    Office of Scientific and Technical Information (OSTI)

    Theoretical Investigation of Thermodynamic Stability and Mobility of the Oxygen Vacancy in ThO2-UO2 Solid Solutions B. Liu1, *, D.S. Aidhy1, Y. Zhang1, 2, W.J. Weber2,1 * 1Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831, USA 2Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee 37996, USA Corresponding Author: Bin Liu Materials Science and Technology Division Oak Ridge National Laboratory Oak Ridge, TN

  9. The Purpose and Value of Successful Technology Demonstrations … The Energy Independence and Security Act of 2007 Demonstrations

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

    Start with the End in Mind - Utility of the Near Future by Steve Pullins, Team Leader, DOE/NETL Modern Grid Strategy Some of my grid friends and I have been discussing the emotional staying power of the utility industry to deploy a Smart Grid over a very long period, maybe the next 15 to 20 years. For me, this raises a very interesting question about vision. Over the last three years, we have seen a few Utility of the Future (UoF) efforts at utilities as they formulate an over-the-horizon vision

  10. Photostat Price S /

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

    Photostat Price S / . p d Microfilm Price $ /- 80 Available from the Office of Technical Services Department of Commerce Washington 25, D. C. A. ifetallurgi c a l Pro.1 ect PHYSICS rnSEARR u E. Fermi, Division Director; G a l e Young, Section Chief * * * . - 1 I - t khCALC'ULATIOM OF TEIE CRITICAL SIZE AND MULTIPUCATIQ! , . - . - L C O N S T A N T OF A H@dOGENBOUS UO2 - DZO MIXTURFS E . P. Nigner, A. M. Ileinberg, J, Stephenson February 11, 1944 The roultiplication constant w d optimal

  11. BISON Enhanced | Department of Energy

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

    Enhanced BISON Enhanced January 29, 2013 - 10:42am Addthis Pin-scale Code Development A mechanistic, smeared fuel cracking model for UO2 has been implemented in BISON and tested with simulations of IFA-432 Rod 1, an experiment conducted in the Halden reactor. ("Smeared" refers to the fact that cracks are represented in aggregate, rather than as discrete, individual cracks.) Failure to account for fuel cracking can result in temperature predictions that are off by as much as 200°C at

  12. SAVANNAH RIVER SITE INCIPIENT SLUDGE MIXING IN RADIOACTIVE LIQUID WASTE STORAGE TANKS DURING SALT SOLUTION BLENDING

    SciTech Connect (OSTI)

    Leishear, R.; Poirier, M.; Lee, S.; Steeper, T.; Fowley, M.; Parkinson, K.

    2011-01-12

    This paper is the second in a series of four publications to document ongoing pilot scale testing and computational fluid dynamics (CFD) modeling of mixing processes in 85 foot diameter, 1.3 million gallon, radioactive liquid waste, storage tanks at Savannah River Site (SRS). Homogeneous blending of salt solutions is required in waste tanks. Settled solids (i.e., sludge) are required to remain undisturbed on the bottom of waste tanks during blending. Suspension of sludge during blending may potentially release radiolytically generated hydrogen trapped in the sludge, which is a safety concern. The first paper (Leishear, et. al. [1]) presented pilot scale blending experiments of miscible fluids to provide initial design requirements for a full scale blending pump. Scaling techniques for an 8 foot diameter pilot scale tank were also justified in that work. This second paper describes the overall reasons to perform tests, and documents pilot scale experiments performed to investigate disturbance of sludge, using non-radioactive sludge simulants. A third paper will document pilot scale CFD modeling for comparison to experimental pilot scale test results for both blending tests and sludge disturbance tests. That paper will also describe full scale CFD results. The final paper will document additional blending test results for stratified layers in salt solutions, scale up techniques, final full scale pump design recommendations, and operational recommendations. Specifically, this paper documents a series of pilot scale tests, where sludge simulant disturbance due to a blending pump or transfer pump are investigated. A principle design requirement for a blending pump is UoD, where Uo is the pump discharge nozzle velocity, and D is the nozzle diameter. Pilot scale test results showed that sludge was undisturbed below UoD = 0.47 ft{sup 2}/s, and that below UoD = 0.58 ft{sup 2}/s minimal sludge disturbance was observed. If sludge is minimally disturbed, hydrogen will not be released. Installation requirements were also determined for a transfer pump which will remove tank contents, and which is also required to not disturb sludge. Testing techniques and test results for both types of pumps are presented.

  13. LANSCE | Lujan Center | People | Zoe Fisher

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

    Gus Sinnis 505.667.6069 Deputy Leader Fredrik Tovesson 505.665.9652 Deputy Leader & Experimental Area Manager Charles Kelsey 505.665.5579 Experiment Coordinator Victor Fanelli 505.667.8755 User Program Administration lujan-uo@lanl.gov Administrative Assistant Julie Quintana-Valdez 505.665.5390 People Instrument Scientists Zoë Fisher | PCS Biographical Sketch Dr. Zoë Fisher is Staff Scientist II in the Bioscience Division of Los Alamos National Laboratory. She is also the instrument

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

  15. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    SciTech Connect (OSTI)

    Rodney C. Ewing

    2004-10-07

    Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

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

    DOE Patents [OSTI]

    Friedman, H.A.

    1984-06-13

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

  17. Liquid Metal Bond for Improved Heat Transfer in LWR Fuel Rods

    SciTech Connect (OSTI)

    Donald Olander

    2005-08-24

    A liquid metal (LM) consisting of 1/3 weight fraction each of Pb, Sn, and Bi has been proposed as the bonding substance in the pellet-cladding gap in place of He. The LM bond eliminates the large AT over the pre-closure gap which is characteristic of helium-bonded fuel elements. Because the LM does not wet either UO2 or Zircaloy, simply loading fuel pellets into a cladding tube containing LM at atmospheric pressure leaves unfilled regions (voids) in the bond. The HEATING 7.3 heat transfer code indicates that these void spaces lead to local fuel hot spots.

  18. Thin-layer chromatography of metal ions complexed with anils. V. Detection, separation, and determination

    SciTech Connect (OSTI)

    Upadhyay, R.K.; Tewari, A.P.

    1980-01-01

    p-Diethylaminoanil of phenylglyoxal, a bidentate ligand, was used for complexation with Hg(II), UO/sub 2/(II), Au(III), Pt(IV), Mg(II), Bi(II), Sb(III), and Be(II) ions. The chelates were characterized by their analysis, molar conductance, and infrared spectra. TLC detection, separation, and determination of these complexes on starch-bound silica gel layers were studied. Long persisting dark colors of the complexes rendered the spots self-descernible and no locating agent was required. A minimum of four complexes could be resolved and identified. Errors in the determinations and maximum separation limits were also deduced. 3 tables.

  19. Microsoft Word - Tebo FINAL Report140825.docx

    Office of Scientific and Technical Information (OSTI)

    TEC H N ICA L REPORT DOE STI Product/Report Number DE-SC0005324 STI Product Title M anganese Redox M ediation o f UO 2 Stability and U ranium Fate in the Subsurface: M olecular and M eter Scale Dynam ics STI Product Type and Reporting Period Final Report for June 1, 2010 - May 31, 2013 Date of Issuance or Publication August 2014 Author Bradley M. Tebo Originating Research Organization Oregon Health & Science University Portland, Oregon 97239 Information Category Unclassified Unlimited

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

    DOE Patents [OSTI]

    Friedman, Horace A. (Oak Ridge, TN)

    1985-01-01

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

  1. DOUBLE-BAKED, SELF-CHANNELLING ELECTRODE

    DOE Patents [OSTI]

    Piper, R.D.; Leifield, R.F.

    1963-03-12

    A method is given for making an electrode for use in the electrolytic reduction of uranium oxides to uranium metal in a fused salt electrolyte. Uranlum oxide such as UO/sub 2/ is mixed with somewhat less than the stoichiometric amount of carbon needed for the reduction, and the mixture is baked and crushed to make a nonspherical material. The latter is then mixed with a carbon binder sufficient to satisfy stoichiometry, pressed into a shape such as a cylinder, and baked. (AEC)

  2. Performance Refactoring of Instrumentation, Measurement, and Analysis Technologies for Petascale Computing. The PRIMA Project

    SciTech Connect (OSTI)

    Malony, Allen D.; Wolf, Felix G.

    2014-01-31

    The growing number of cores provided by todays high-end computing systems present substantial challenges to application developers in their pursuit of parallel efficiency. To find the most effective optimization strategy, application developers need insight into the runtime behavior of their code. The University of Oregon (UO) and the Juelich Supercomputing Centre of Forschungszentrum Juelich (FZJ) develop the performance analysis tools TAU and Scalasca, respectively, which allow high-performance computing (HPC) users to collect and analyze relevant performance data even at very large scales. TAU and Scalasca are considered among the most advanced parallel performance systems available, and are used extensively across HPC centers in the U.S., Germany, and around the world. The TAU and Scalasca groups share a heritage of parallel performance tool research and partnership throughout the past fifteen years. Indeed, the close interactions of the two groups resulted in a cross-fertilization of tool ideas and technologies that pushed TAU and Scalasca to what they are today. It also produced two performance systems with an increasing degree of functional overlap. While each tool has its specific analysis focus, the tools were implementing measurement infrastructures that were substantially similar. Because each tool provides complementary performance analysis, sharing of measurement results is valuable to provide the user with more facets to understand performance behavior. However, each measurement system was producing performance data in different formats, requiring data interoperability tools to be created. A common measurement and instrumentation system was needed to more closely integrate TAU and Scalasca and to avoid the duplication of development and maintenance effort. The PRIMA (Performance Refactoring of Instrumentation, Measurement, and Analysis) project was proposed over three years ago as a joint international effort between UO and FZJ to accomplish these objectives: (1) refactor TAU and Scalasca performance system components for core code sharing and (2) integrate TAU and Scalasca functionality through data interfaces, formats, and utilities. As presented in this report, the project has completed these goals. In addition to shared technical advances, the groups have worked to engage with users through application performance engineering and tools training. In this regard, the project benefits from the close interactions the teams have with national laboratories in the United States and Germany. We have also sought to enhance our interactions through joint tutorials and outreach. UO has become a member of the Virtual Institute of High-Productivity Supercomputing (VI-HPS) established by the Helmholtz Association of German Research Centres as a center of excellence, focusing on HPC tools for diagnosing programming errors and optimizing performance. UO and FZJ have conducted several VI-HPS training activities together within the past three years.

  3. Performance Refactoring of Instrumentation, Measurement, and Analysis Technologies for Petascale Computing: the PRIMA Project

    SciTech Connect (OSTI)

    Malony, Allen D.; Wolf, Felix G.

    2014-01-31

    The growing number of cores provided by todays high-end computing systems present substantial challenges to application developers in their pursuit of parallel efficiency. To find the most effective optimization strategy, application developers need insight into the runtime behavior of their code. The University of Oregon (UO) and the Juelich Supercomputing Centre of Forschungszentrum Juelich (FZJ) develop the performance analysis tools TAU and Scalasca, respectively, which allow high-performance computing (HPC) users to collect and analyze relevant performance data even at very large scales. TAU and Scalasca are considered among the most advanced parallel performance systems available, and are used extensively across HPC centers in the U.S., Germany, and around the world. The TAU and Scalasca groups share a heritage of parallel performance tool research and partnership throughout the past fifteen years. Indeed, the close interactions of the two groups resulted in a cross-fertilization of tool ideas and technologies that pushed TAU and Scalasca to what they are today. It also produced two performance systems with an increasing degree of functional overlap. While each tool has its specific analysis focus, the tools were implementing measurement infrastructures that were substantially similar. Because each tool provides complementary performance analysis, sharing of measurement results is valuable to provide the user with more facets to understand performance behavior. However, each measurement system was producing performance data in different formats, requiring data interoperability tools to be created. A common measurement and instrumentation system was needed to more closely integrate TAU and Scalasca and to avoid the duplication of development and maintenance effort. The PRIMA (Performance Refactoring of Instrumentation, Measurement, and Analysis) project was proposed over three years ago as a joint international effort between UO and FZJ to accomplish these objectives: (1) refactor TAU and Scalasca performance system components for core code sharing and (2) integrate TAU and Scalasca functionality through data interfaces, formats, and utilities. As presented in this report, the project has completed these goals. In addition to shared technical advances, the groups have worked to engage with users through application performance engineering and tools training. In this regard, the project benefits from the close interactions the teams have with national laboratories in the United States and Germany. We have also sought to enhance our interactions through joint tutorials and outreach. UO has become a member of the Virtual Institute of High-Productivity Supercomputing (VI-HPS) established by the Helmholtz Association of German Research Centres as a center of excellence, focusing on HPC tools for diagnosing programming errors and optimizing performance. UO and FZJ have conducted several VI-HPS training activities together within the past three years.

  4. A Scoping Analysis Of The Impact Of SiC Cladding On Late-Phase Accident Progression Involving CoreConcrete Interaction

    SciTech Connect (OSTI)

    Farmer, M. T.

    2015-11-01

    The overall objective of the current work is to carry out a scoping analysis to determine the impact of ATF on late phase accident progression; in particular, the molten-core concrete interaction portion of the sequence that occurs after the core debris fails the reactor vessel and relocates into containment. This additional study augments previous work by including kinetic effects that govern chemical reaction rates during core-concrete interaction. The specific ATF considered as part of this study is SiC-clad UO2.

  5. Proton Radiography at Los Alamos National Laboratory (pRad)

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

    pRad at LANL P-Division | LANSCE >> pRad Home About News Movies Proposals Contacts Team Leader Dale Tupa 505.665.1820 Project Leader Andy Saunders 505.665.3090 Area Manager (interim) Eron Kerstiens 505.667.3618 Fesseha Mariam 505.667.3546 Christopher Morris 505.667.5652 Frans Trouw 505.665.7575 pRad User Program pRad-uo@lanl.gov P-25 Group Leader Melynda Brooks 505.667.6909 P-25 Deputy Group Leader Frans Trouw 505.665.7575 P-25 Subatomic Physics P-Division LANSCE pRad logo Los Alamos

  6. Thermodynamic assessment of the U-Y-O system

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

    McMurray, Jake W; Shin, Dongwon; Besmann, Theodore M

    2015-01-01

    A CALPHAD assessment of the U-Y-O system is developed. To represent the YO2 compound in the compound energy formalism (CEF) for U1-yYyO2 x, the lattice stability was calculated using density functional theory (DFT) while a partially ionic liquid sub-lattice model is used to describe the liquid phase. A Gibbs function for the stoichiometric rhombohedral UY6O12 phase is proposed. Models representing the phases in the U-O and Y-O systems taken from the literature along with the phases that appear in the U-Y-O ternary are combined to form a unified assessment.

  7. Newberry EGS Seismic Velocity Model

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

    Templeton, Dennise

    2013-10-01

    We use ambient noise correlation (ANC) to create a detailed image of the subsurface seismic velocity at the Newberry EGS site down to 5 km. We collected continuous data for the 22 stations in the Newberry network, together with 12 additional stations from the nearby CC, UO and UW networks. The data were instrument corrected, whitened and converted to single bit traces before cross correlation according to the methodology in Benson (2007). There are 231 unique paths connecting the 22 stations of the Newberry network. The additional networks extended that to 402 unique paths crossing beneath the Newberry site.

  8. The AN neutron transport by nodal diffusion

    SciTech Connect (OSTI)

    Barbarino, A.; Tomatis, D.

    2013-07-01

    The two group diffusion model combined to a nodal approach in space is the preferred scheme for the industrial simulation of nuclear water reactors. The main selling point is the speed of computation, allowing a large number of parametric studies. Anyway, the drawbacks of the underlying diffusion equation may arise with highly heterogeneous interfaces, often encountered in modern UO{sub 2} and MO{sub x} fuel loading patterns, and boron less controlled systems. This paper aims at showing how the simplified AN transport model, equivalent to the well known SPN, can be implemented in standard diffusion codes with minor modifications. Some numerical results are illustrated. (authors)

  9. Microsoft PowerPoint - MOX Adventure_Reactor Subcommittee_Tamara Reavis

    National Nuclear Security Administration (NNSA)

    MOX Adventure Tamara Reavis May 2015 Page 2 Overview of Presentation > Characteristics of MOX Fuel  MOX Fuel at Duke Energy  MOX Fuel and NMMSS Page 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 uranium  Fission power comes primarily from plutonium (Pu 239 ) instead of uranium (U 235 )  Other than the material of the fuel pellets, MOX and uranium fuel

  10. A survey of Existing V&V, UQ and M&S Data and Knowledge Bases in Support of the Nuclear Energy - Knowledge base for Advanced Modeling and Simulation (NE-KAMS)

    SciTech Connect (OSTI)

    Hyung Lee; Rich Johnson, Ph.D.; Kimberlyn C. Moussesau

    2011-12-01

    The Nuclear Energy - Knowledge base for Advanced Modeling and Simulation (NE-KAMS) is being developed at the Idaho National Laboratory in conjunction with Bettis Laboratory, Sandia National Laboratories, Argonne National Laboratory, Oak Ridge National Laboratory, Utah State University and others. The objective of this consortium is to establish a comprehensive knowledge base to provide Verification and Validation (V&V) and Uncertainty Quantification (UQ) and other resources for advanced modeling and simulation (M&S) in nuclear reactor design and analysis. NE-KAMS will become a valuable resource for the nuclear industry, the national laboratories, the U.S. NRC and the public to help ensure the safe operation of existing and future nuclear reactors. A survey and evaluation of the state-of-the-art of existing V&V and M&S databases, including the Department of Energy and commercial databases, has been performed to ensure that the NE-KAMS effort will not be duplicating existing resources and capabilities and to assess the scope of the effort required to develop and implement NE-KAMS. The survey and evaluation have indeed highlighted the unique set of value-added functionality and services that NE-KAMS will provide to its users. Additionally, the survey has helped develop a better understanding of the architecture and functionality of these data and knowledge bases that can be used to leverage the development of NE-KAMS.

  11. Molecular dynamics simulations of uranyl adsorption and structure on the basal surface of muscovite

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

    Teich-McGoldrick, Stephanie L.; Greathouse, Jeffery A.; Cygan, Randall T.

    2014-02-05

    Anthropogenic activities have led to an increased concentration of uranium on the Earth’s surface and potentially in the subsurface with the development of nuclear waste repositories. Uranium is soluble in groundwater, and its mobility is strongly affected by the presence of clay minerals in soils and in subsurface sediments. We use molecular dynamics simulations to probe the adsorption of aqueous uranyl (UO22+) ions onto the basal surface of muscovite, a suitable proxy for typically ultrafine-grained clay phases. Model systems include the competitive adsorption between potassium counterions and aqueous ions (0.1 M and 1.0 M UO2Cl2 , 0.1 M NaCl). Wemore » find that for systems with potassium and uranyl ions present, potassium ions dominate the adsorption phenomenon. Potassium ions adsorb entirely as inner-sphere complexes associated with the ditrigonal cavity of the basal surface. Uranyl ions adsorb in two configurations when it is the only ion species present, and in a single configuration in the presence of potassium. Finally, the majority of adsorbed uranyl ions are tilted less than 45° relative to the muscovite surface, and are associated with the Si4Al2 rings near aluminum substitution sites.« less

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

    SciTech Connect (OSTI)

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

    2014-12-01

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

  13. Chemical processing programs. Monthly status report, April 1986

    SciTech Connect (OSTI)

    Not Available

    1986-04-01

    During the month of April, 99 metric tonnes uranium (MTU's) of zircaloy-clad N-Reactor fuel were charged to the PUREX dissolvers; bringing the FYTD total to 684 MTU's, 115 MTU's ahead of the 1060 commitment schedule. PUREX solvent extraction was shut down April 14 and the plant entered into a planned maintenance period to effect repairs and perform process chemical flushes to maintain acceptable waste losses and production specification. The Plutonium Oxide Conversion (N)-Cell bi-monthly nuclear material and accountability inventory, initiated in March, was completed satisfactorily in April. UO/sub 3/ Plant initiated the second fiscal year 1986 campaign. During April, 46 MTU's of UO/sub 3/ were shipped to FMPC, bringing the FYTD shipment total to 456 MTU's vs a plan of 490 MTU's. Design and procurement activities for the PUREX Aqueous Make-Up (AMU) chemical containment upgrades continued on schedule during April. The Remote Mechanical C (RMC) line began processing feed for its first fiscal year (FY) 1986 campaign on April 5, 1986. The Plutonium Reclamation Facility (PRF) maintenance outage upgrades are one and one half weeks behind schedule. Functional Design Criteria for B609, RMC Ventilation Improvement (FY 1988 GPP) has been completed. The updated Ten Year Shipping Forecast has been complete and sent to DOE-RL.

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

    DOE Patents [OSTI]

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

    1995-06-06

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

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

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

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

    2014-12-01

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

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

    DOE Patents [OSTI]

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

    1995-01-01

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

  17. Validating mass spectrometry measurements of nuclear materials via a non-contact volume analysis method of ion sputter craters

    SciTech Connect (OSTI)

    Willingham, David G.; Naes, Benjamin E.; Fahey, Albert J.

    2015-01-01

    A combination of secondary ion mass spectrometry, optical profilometry and a statistically-driven algorithm was used to develop a non-contact volume analysis method to validate the useful yields of nuclear materials. The volume analysis methodology was applied to ion sputter craters created in silicon and uranium substrates sputtered by 18.5 keV O- and 6.0 keV Ar+ ions. Sputter yield measurements were determined from the volume calculations and were shown to be comparable to Monte Carlo calculations and previously reported experimental observations. Additionally, the volume calculations were used to determine the useful yields of Si+, SiO+ and SiO2+ ions from the silicon substrate and U+, UO+ and UO2+ ions from the uranium substrate under 18.5 keV O- and 6.0 keV Ar+ ion bombardment. This work represents the first steps toward validating the interlaboratory and cross-platform performance of mass spectrometry for the analysis of nuclear materials.

  18. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    SciTech Connect (OSTI)

    Carmack, W.J.; Husser, D.L.; Mohr, T.C.; Richardson, W.C.

    2004-02-04

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  19. Contaminant Leach Testing of Hanford Tank 241-C-104 Residual Waste

    SciTech Connect (OSTI)

    Cantrell, Kirk J.; Snyder, Michelle M.V.; Wang, Guohui; Buck, Edgar C.

    2015-07-01

    Leach testing of Tank C-104 residual waste was completed using batch and column experiments. Tank C-104 residual waste contains exceptionally high concentrations of uranium (i.e., as high as 115 mg/g or 11.5 wt.%). This study was conducted to provide data to develop contaminant release models for Tank C-104 residual waste and Tank C-104 residual waste that has been treated with lime to transform uranium in the waste to a highly insoluble calcium uranate (CaUO4) or similar phase. Three column leaching cases were investigated. In the first case, C-104 residual waste was leached with deionized water. In the second case, crushed grout was added to the column so that deionized water contacted the grout prior to contacting the waste. In the third case, lime was mixed in with the grout. Results of the column experiments demonstrate that addition of lime dramatically reduces the leachability of uranium from Tank C-104 residual waste. Initial indications suggest that CaUO4 or a similar highly insoluble calcium rich uranium phase forms as a result of the lime addition. Additional work is needed to definitively identify the uranium phases that occur in the as received waste and the waste after the lime treatment.

  20. Investigation of the long-term performance of betafite and zirconolite in hydrothermal veins from Adamello, Italy

    SciTech Connect (OSTI)

    Lumpkin, G.R.; Day, R.A.; McGlinn, P.J.; Payne, T.E.; Giere, R.; Williams, C.T.

    1999-07-01

    Betafite and zirconolite occur in Ti-rich hydrothermal veins emplaced within dolomite marble in the contact aureole of the Adamello batholith, northern Italy. Zirconolite contains up to 18 wt% ThO{sub 2} and 24 wt% UO{sub 2}, and exhibits strong compositional zoning. Some zirconolite grains were corroded by the hydrothermal fluid. Betafite, the Ti-rich member of the pyrochlore group, often occurs as overgrowths on zirconolite. The betafite is weakly zoned and contains 29--34 wt% UO{sub 2}. In terms of end-members, betafite contains approximately 50 mole percent CaUTi{sub 2}O{sub 7} and is the closest known natural composition to the pyrochlore phase proposed for use in titanate waste forms. Amorphization and volume expansion of the betafite caused cracks to form in the enclosing silicate mineral grains. Backscattered electron images reveal that betafite was subsequently altered along crystal rims, particularly near the cracks. EPMA data reveal little difference in composition between altered and unaltered areas, except for lower totals, suggesting that alteration is primarily due to hydration. The available evidence demonstrates that both betafite and zirconolite retained actinides for approximately 40 million years after the final stage of vein formation. During this time, betafite and zirconolite accumulated a total alpha-decay dose of 3--4 x 10{sup 16} and 0.2--2 x 10{sup 16} {alpha}/mg, respectively.

  1. Proliferation resistance of small modular reactors fuels

    SciTech Connect (OSTI)

    Polidoro, F.; Parozzi, F.; Fassnacht, F.; Kuett, M.; Englert, M.

    2013-07-01

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

  2. Long-term, low-temperature oxidation of PWR spent fuel: Interim transition report

    SciTech Connect (OSTI)

    Einziger, R.E.; Buchanan, H.C.

    1988-05-01

    Since some of the fuel rods will be breached and eventually most of the cladding will corrode, exposing fuel, one factor influencing the ability of spent fuel to retain radionuclides is its oxidation state in the expected moist air atmosphere. Oxidation of the fuel could split the cladding, exposing additional fuel and changing the leaching characteristics. Thermodynamically, there is no reason why UO{sub 2} should not oxidize completely to UO{sub 3} at repository temperatures. The underlying uncertainty is the rate of oxidation. Extrapolation of higher temperature data indicates that insufficient oxidation to convert all of the fuel to U{sub 3}O{sub 8} will occur during the first 10,000 years. However, lower oxidation states, such as U{sub 4}O{sub 9} and U{sub 3}O{sub 7}, might form. To date, the tests have run between 3200 and 4100 hours out of a planned 16,000-hour duration. Some preliminary conclusions can be drawn: (1) Moisture content of the air has no significant effect on oxidation rate, (2) the data have an uncertainty of 15 to 20%, which must be accounted for in the interpretation of single sample tests, and (3) below 175{degree}C, the oxidation rate is dependent on the particle size in the sample. The smaller particles oxidize more rapidly. 19 refs., 23 figs., 7 tabs.

  3. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment. [BWR; PWR

    SciTech Connect (OSTI)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO/sub 2/ fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm/sup 3//s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO/sub 2/ fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%.

  4. Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

    SciTech Connect (OSTI)

    Michael A. Pope; R. Sonat Sen; Brian Boer; Abderrafi M. Ougouag; Gilles Youinou

    2011-09-01

    The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

  5. PROCESS FOR PRODUCING URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-10-22

    A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ convented to UF/sub 6/ by reaction with a fluorinating agent. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reaction chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. The oven is then swept clean of hydrogen and the water vapor formed by means of nitrogen and then while continuing to maintain the temperature between 400 and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion to uranium tetrafluoride, the temperature of the reaction chamber is lowered to ahout 400 deg C, and elemental fluorine is used as the fluorinating agent for the conversion of UF/sub 4/ into UF/sub 6/. The fluorine gas is passed into the chamber, and the UF/sub 6/ formed passes out and is delivered to a condenser.

  6. Molecular dynamics simulations of uranyl adsorption and structure on the basal surface of muscovite

    SciTech Connect (OSTI)

    Teich-McGoldrick, Stephanie L.; Greathouse, Jeffery A.; Cygan, Randall T.

    2014-02-05

    Anthropogenic activities have led to an increased concentration of uranium on the Earth’s surface and potentially in the subsurface with the development of nuclear waste repositories. Uranium is soluble in groundwater, and its mobility is strongly affected by the presence of clay minerals in soils and in subsurface sediments. We use molecular dynamics simulations to probe the adsorption of aqueous uranyl (UO22+) ions onto the basal surface of muscovite, a suitable proxy for typically ultrafine-grained clay phases. Model systems include the competitive adsorption between potassium counterions and aqueous ions (0.1 M and 1.0 M UO2Cl2 , 0.1 M NaCl). We find that for systems with potassium and uranyl ions present, potassium ions dominate the adsorption phenomenon. Potassium ions adsorb entirely as inner-sphere complexes associated with the ditrigonal cavity of the basal surface. Uranyl ions adsorb in two configurations when it is the only ion species present, and in a single configuration in the presence of potassium. Finally, the majority of adsorbed uranyl ions are tilted less than 45° relative to the muscovite surface, and are associated with the Si4Al2 rings near aluminum substitution sites.

  7. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect (OSTI)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

  8. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    SciTech Connect (OSTI)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO/sub 2/ oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO/sub 2/ pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs.

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

    SciTech Connect (OSTI)

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

    2013-05-15

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

  10. Spectral indices measurements using miniature fission chambers at the MINERVE zero-power reactor at CEA using calibration data obtained at the BR1 reactor at SCK.CEN

    SciTech Connect (OSTI)

    De lanaute, N. Blanc; Mellier, F.; Lyoussi, A.; Domergue, C.; Di Salvo, J. [CEA, DEN, DER, SPEX, F-13108 St Paul Les Durance, (France); Borms, L.; Wagemans, J. [CEN SCK, Belgian Nucl Res Ctr, B-2400 Mol, (Belgium)

    2012-08-15

    Spectral indices measurements performed in 2004 at the CEA MINERVE facility loaded with the R-UO{sub 2} lattice, using calibration data acquired at the SCK center dot CEN BR1 facility in 2001, resulted in ambivalent conclusions. On one hand, spectral indices involving only fissile isotopes gave consistent discrepancies between calculation and experiment. On the other hand, spectral indices involving both fissile and fertile isotopes, in particular the {sup 238}U(n, f)/{sup 235}U(n, f) spectral index, showed inconsistent results depending on the type of calibration data used. For different reasons, no definitive explanation was given at that time. In 2009, the preparation of the AMMON program at the EOLE facility motivated the manufacturing of a new set of detectors. At the same time, the re-installation of the R1-UO{sub 2} lattice in MINERVE provided the opportunity to carry out again a spectral indices measurement campaign. Nevertheless, although the isotopic compositions of active deposits were better known than previously, the comparison between experimental results and calculations still lead to inconsistent discrepancies. In April 2010, a new calibration series conducted again at the BR1 facility allowed the CEA to reanalyze the spectral indices measurements performed in 2009. With these very latest calibration data, experimental values of spectral indices finally matched calculations within the uncertainty margins. This paper also sums up the work that has been achieved to explain the incoherencies observed in 2004. (authors)

  11. Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

    SciTech Connect (OSTI)

    Williford, R.E.; Mohr, C.L.; Lanning, D.D.; Cunningham, M.E.; Rausch, W.N.; Bradley, E.R.

    1980-04-01

    The effects of the thermally-induced cracking and subsequent relocation of UO/sub 2/ fuel pellets on the thermal and mechanical behavior of light-water reactor fuel rods during irradiation are quantified in this report. Data from the Nuclear Regulatory Commission/Pacific Northwest Laboratory Halden experiments on instrumented fuel assemblies (IFA) IFA-431, IFA-432, and IFA-513 are analyzed. Beginning-of-life in-reactor measurements of fuel center temperatures, linear heat ratings, and cladding axial elongations are used in a new model to solve for the effective thermal conductivity and elastic moduli of the cracked fuel column. The effective thermal conductivity and elastic moduli for the cracked fuel were found to be significantly reduced from the values for solid UO/sub 2/ pellets. The calculated fuel-cladding gap remained relatively constant (closed) with respect to power level, indicating that the fuel fragments do not retreat from the cladding when the power/temperature is reduced. Recommendations are made pertaining to the work required to further refine the model. 30 refs., 81 figs., 8 tabs.

  12. Evaluation of fission gas release in high-burnup light water reactor fuel rods

    SciTech Connect (OSTI)

    Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D. )

    1993-05-01

    Research to define the behavior of Zircaloy-clad light water reactor (LWR) UO[sub 2] fuel irradiated to high burnup levels was conducted as part of the High Burnup Effects Program (HBEP). The HBEP was a 12-yr program that ultimately acquired, characterized, irradiated, and examined after irradiation 82 LWR fuel rods ranging in rod-average fuel burnup from 22 to 69 MWd/kgM with a peak pellet burnup of 83 MWd/kg M. A principal emphasis of the HBEP was to evaluate the effect of high burnup on fission gas release. It was confirmed that fission gas release remained as dependent on design and irradiation history parameters at high burnup levels as at low to moderate burnup levels. One observed high-burnup effect was the development of a burnup-dependent microstructure at the fuel pellet surface when pellet-edge burnup exceeded 65 MWd/kgM. This low-temperature rim region' was characterized by a loss of optically definable grain structure, a high volume of porosity, and diffusion of fission gas from the UO[sub 2] matrix to the porosity. Although the rim region has the potential for enhanced fission gas release, it is concluded that no significant enhancement of rod-average fission gas release at high burnup levels was observed for the examined fuel rods.

  13. Direct Experimental Evaluation of the Grain Boundaries Gas Content in PWR fuels: New Insight and Perspective of the ADAGIO Technique

    SciTech Connect (OSTI)

    Pontillon, Y.; Noirot, J.; Caillot, L.

    2007-07-01

    Over the last decades, many analytical experiments (in-pile and out-of-pile) have underlined the active role of the inter-granular gases on the global fuel transient behavior under accidental conditions such as RIA and/or LOCA. In parallel, the improvement of fission gas release modeling in nuclear fuel performance codes needs direct experimental determination/validation regarding the local gas distribution inside the fuel sample. In this context, an experimental program, called 'ADAGIO' (French acronym for Discriminating Analysis of Accumulation of Inter-granular and Occluded Gas), has been initiated through a joint action of CEA, EDF and AREVA NP in order to develop a new device/technique for quantitative and direct measurement of local fission gas distribution within an irradiated fuel pellet. ADAGIO technique is based on the fact that fission gas inventory (intra and inter-granular parts) can be distinguished by controlled fuel oxidation, since grain boundaries oxidize faster than the bulk. The purpose of the current paper is to present both the methodology and the associated results of the ADAGIO program performed at CEA. It has been divided into two main parts: (i) feasibility (UO{sub 2} and MOX fuels), (ii) application on high burn up UO{sub 2} fuel. (authors)

  14. Clay-sewage sludge co-pyrolysis. A TG-MS and Py-GC study on potential advantages afforded by the presence of clay in the pyrolysis of wastewater sewage sludge

    SciTech Connect (OSTI)

    Ischia, Marco; Maschio, Roberto Dal; Grigiante, Maurizio; Baratieri, Marco

    2011-01-15

    Wastewater sewage sludge was co-pyrolyzed with a well characterized clay sample, in order to evaluate possible advantages in the thermal disposal process of solid waste. Characterization of the co-pyrolysis process was carried out both by thermogravimetric-mass spectrometric (TG-MS) analysis, and by reactor tests, using a lab-scale batch reactor equipped with a gas chromatograph for analysis of the evolved gas phase (Py-GC). Due to the presence of clay, two main effects were observed in the instrumental characterization of the process. Firstly, the clay surface catalyzed the pyrolysis reaction of the sludge, and secondly, the release of water from the clay, at temperatures of approx. 450-500 deg. C, enhanced gasification of part of carbon residue of the organic component of sludge following pyrolysis. Moreover, the solid residue remaining after pyrolysis process, composed of the inorganic component of sludge blended with clay, is characterized by good features for possible disposal by vitrification, yielding a vitreous matrix that immobilizes the hazardous heavy metals present in the sludge.

  15. Microsoft Word - Fischer_MS.doc

    Office of Scientific and Technical Information (OSTI)

    Challenges to modern magnetic microscopies Research of magnetism in low dimensions has not ... has been awarded with the Noble Prize Physics in 2007, but has also tremendously ...

  16. Microsoft Word - benzyl-ms-Revisednoyellow.doc

    Office of Scientific and Technical Information (OSTI)

    The Reaction of bis(1,2,4-tri-t-butylcyclopentadienyl)ceriumbenzyl, Cp' 2 CeCH 2 Ph with Methylhalides: a Metathesis Reaction that does not proceed by a Metathesis Transition State. Evan L. Werkema, a Richard A. Andersen,* a Laurent Maron, *b and Odile Eisenstein *c a) Department of Chemistry and Chemical Sciences Division of Lawrence Berkeley National Laboratory, University of California, Berkeley, California 94720-1460. b) LPCNO, Université de Toulouse, INSA, UPS, LPCNO, 135 avenue de

  17. To: Ms. Dorothy Riehle, FOIA Officer,

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

    8 (Released July 10, 1998) Energy Information Administration DOE/EIA-0202(98/3Q) Distribution Category UC-950 Short-Term Energy Outlook July 1998 Energy Information Administration Office of Energy Markets and End Use U.S. Department of Energy Washington, DC 20585 This report was prepared by the Energy Information Administration, the independent statistical and analytical agency within the U.S. Department of Energy. The information contained herein should be attributed to the Energy Information

  18. LA-8318-MS Informal Report I

    Office of Scientific and Technical Information (OSTI)

    ... TRACT CELLS USING FLOW - SYSTEM CELL - ANALYSIS TECHNIQUES, SENSING OF ENVIRONMENTAL ... EVENTS IN THE CELL CYCLE. CELL CULTURE CONGRESS. 1975. 1ST INTERNATIONAL, ...

  19. LA-8318-MS Informal Report I

    Office of Scientific and Technical Information (OSTI)

    ... D. (E-3) 2-75 QUANTITATION OF CELL FUSION BY TWENTY-ONE STRAINS OF NEWCASTLE DISEASE VIRUS USING FLOW MICROFLUOROMETRY. J. GEN. VIROL., V.41. P.27-36. 1978. CRAM, L. SCOTT (H-10) ...

  20. WSRC-MS-99-00210

    Office of Scientific and Technical Information (OSTI)

    included standard machining, gas-tungsten arc (GTA) welding, pinch welding, and casting. Each assembly is small enough to hold in two hands and weighs 44 pounds or less....

  1. LA-9252-MS UC-70a

    Office of Legacy Management (LM)

    ... managing subdivision of land, and Los Alamos County has ... Natural gas for Los Alamos County is purchased from DOE and ... for DOE Operations; Requirements for Radiation ...

  2. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    SciTech Connect (OSTI)

    Field, Kevin G.; Howard, Richard H.

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys, hence promoting FCCI between the fuel-clad systems. The other factor was to develop a test bed where multiple candidate alloys could be evaluated within a single irradiation test train, thereby reducing overall costs and increasing efficiency in alloy screening efforts. A collaboration between ORNL and INL was developed to facilitate the completion of the test bed for FCCI testing. The report highlights the activities related to the development of the ATF-1 ORNL FCCI rodlets for irradiation in INL’s ATR as part of the on-going ATF-1 experiments.

  3. AREVA NP Cr{sub 2}O{sub 3}-doped fuel development for BWRs

    SciTech Connect (OSTI)

    Delafoy, C.; Dewes, P.; Miles, T.

    2007-07-01

    The search for improvements in nuclear fuel cycle economics results in increasing demands for fuel discharged burnup and reliability, plant maneuverability and power up-rating. To achieve these objectives without any reduction of safety margins, fuel design and materials that enable enhanced performance capabilities have been developed or are under investigations. Research on fuel pellets focuses on the modification of the microstructure to increase fission product retention and pellet mechanical compliance. Currently, production of the desired large grain viscoplastic UO{sub 2} fuel microstructures has been extensively investigated by AREVA NP through the use of doping elements. This track is nowadays a worldwide working field. In this area, AREVA NP has launched the development of a new UO{sub 2} fuel pellet obtained by optimum chromium oxide doping. The purpose of this paper is first to present the current results with the AREVA NP optimized chromia doped fuel and to discuss the key advantages in terms of fuel performance for BWR applications. In particular, the development relies on ramp testing results, fuel temperature and fission gas release values acquired at high burnup and high power levels. Second, the paper focuses on the qualification process implemented by AREVA NP to assess the margins of the optimized Cr{sub 2}O{sub 3}-doped UO{sub 2} fuel towards safety criteria at high burnup and the risk of PCI failure, as well as to develop calculation tools to support design. The driving force in this qualification plan is to gain the accurate knowledge of the optimized doped fuel behavior under normal, transient and anticipated accident conditions. To support this effort, irradiation campaigns are under progress in PWR and BWR plants to cover a wide range of existing operating conditions and to anticipate future demands. Considering only the BWR part, the program has successfully run since 2005 and is designed to obtain data up to high burnup, at least 70 GWd/tU. The aim is to define the range of operational conditions for application of chromia-doped fuel in combination with LTP2 non-liner cladding as an alternative to the present standard Fe-enhanced Zr liner cladding. (authors)

  4. "FERC423",2007,1,195,"Alabama Power Co",3,"Barry","AL","C","application/vnd.ms-excel","Coal","BIT",45,"IM","SU","County Unknown",999,"MINA PRIBBENOW",289050,22.732,0.5,5.2,217.3

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

    7,1,195,"Alabama Power Co",3,"Barry","AL","C","application/vnd.ms-excel","Coal","BIT",45,"IM","SU","County Unknown",999,"MINA PRIBBENOW",289050,22.732,0.5,5.2,217.3 "FERC423",2007,1,195,"Alabama Power Co",3,"Barry","AL","C","application/vnd.ms-excel","Coal","BIT",45,"IM","SU","County

  5. CONTINUOUS PROCESS FOR PREPARING URANIUM HEXAFLUORIDE FROM URANIUM TETRAFLUORIDE AND OXYGEN

    DOE Patents [OSTI]

    Adams, J.B.; Bresee, J.C.; Ferris, L.M.

    1961-11-21

    A process for preparing UF/sub 6/ by reacting UF/sub 4/ and oxygen is described. The UF/sub 4/ and oxygen are continuously introduced into a fluidized bed of UO/sub 2/F/sub 2/ at a temperature of 600 to 900 deg C. The concentration of UF/sub 4/ in the bed is maintained below 25 weight per cent in order to avoid sintering and intermediate compound formation. By-product U0/sub 2/F/sub 2/ is continuously removed from the top of the bed recycled. In an alternative embodiment heat is supplied to the reaction bed by burning carbon monoxide in the bed. The product UF/sub 6/ is filtered to remove entrained particles and is recovered in cold traps and chemical traps. (AEC)

  6. PROCESS FOR PRODUCING URANIUM HALIDES

    DOE Patents [OSTI]

    Murphree, E.V.

    1957-10-29

    A process amd associated apparatus for producing UF/sub 4/ from U/sub 3/ O/sub 8/ by a fluidized'' technique are reported. The U/sub 3/O/sub 8/ is first reduced to UO/sub 2/ by reaction with hydrogen, and the lower oxide of uranium is then reacted with gaseous HF to produce UF/sub 4/. In each case the reactant gas is used, alone or in combination with inert gases, to fluidize'' the finely divided reactant solid. The complete setup of the plant equipment including bins, reactor and the associated piping and valving, is described. An auxiliary fluorination reactor allows for the direct production of UF/sub 6/ from UF/sub 4/ and fluorine gas, or if desired, UF/sub 4/ may be collected as the product.

  7. Comparison studies of head-end reprocessing using three LWR fuels

    SciTech Connect (OSTI)

    Goode, J.H.; Stacy, R.G.; Vaughen, V.C.A.

    1980-06-01

    The removal of {sup 3}H by voloxidation and the dissolution behavior of two PWR and one BWR fuels were compared in hot-cell studies. The experiments showed that >99% of the {sup 3}H contained in the irradiated UO{sub 2} was volatilized by oxidation in air at 753{sup 0}K (480{sup 0}C). The oxidation did not affect the dissolution of the uranium and plutonium in 7 M HNO{sub 3} (0.02 to 0.03% insoluble plutonium) but did create a fission-product residue that was two to three times more insoluble. From 40 to 69% of the ternary fission-product {sup 3}H was found in the Zircaloy cladding of the fuel rods. Voloxidation had little effect on the {sup 3}H held in the Zircaloy cladding; oxidation for 6 h at 753{sup 0}K released only 0.05% of the {sup 3}H.

  8. Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect (OSTI)

    David Petti

    2014-06-01

    Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germany produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO2 particle fuel up to about 10% FIMA and 1150°C, UO2 fuel is known to have limitations because of CO formation and kernel migration at the high burnups, power densities, temperatures, and temperature gradients that may be encountered in the prismatic modular HTGRs. With uranium oxycarbide (UCO) fuel, the kernel composition is engineered to prevent CO formation and kernel migration, which are key threats to fuel integrity at higher burnups, temperatures, and temperature gradients. Furthermore, the recent poor fuel performance of UO2 TRISO fuel pebbles measured in Chinese irradiation testing in Russia and in German pebbles irradiated at 1250°C, and historic data on poorer fuel performance in safety testing of German pebbles that experienced burnups in excess of 10% FIMA [1] have each raised concern about the use of UO2 TRISO above 10% FIMA and 1150°C and the degree of margin available in the fuel system. This continues to be an active area of study internationally.

  9. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  10. Steam Oxidation of FeCrAl and SiC in the Severe Accident Test Station (SATS)

    SciTech Connect (OSTI)

    Pint, Bruce A.; Unocic, Kinga A.; Terrani, Kurt A.

    2015-08-01

    Numerous research projects are directed towards developing accident tolerant fuel (ATF) concepts that will enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the U.S. program, the high temperature steam oxidation performance of ATF solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012 [1-3] and this facility continues to support those efforts in the ATF community. Compared to the current UO2/Zr-based alloy fuel system, alternative cladding materials can offer slower oxidation kinetics and a smaller enthalpy of oxidation that can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [4-5]. Thus, steam oxidation behavior is a key aspect of the evaluation of ATF concepts. This report summarizes recent work to measure steam oxidation kinetics of FeCrAl and SiC specimens in the SATS.

  11. Thermodynamic properties of uranium dioxide

    SciTech Connect (OSTI)

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

    1981-04-01

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

  12. Direct fissile assay of enriched uranium using random self-interrogation and neutron coincidence response

    DOE Patents [OSTI]

    Menlove, Howard O.; Stewart, James E.

    1986-01-01

    Apparatus and method for the direct, nondestructive evaluation of the .sup.235 U nuclide content of samples containing UF.sub.6, UF.sub.4, or UO.sub.2 utilizing the passive neutron self-interrogation of the sample resulting from the intrinsic production of neutrons therein. The ratio of the emitted neutron coincidence count rate to the total emitted neutron count rate is determined and yields a measure of the bulk fissile mass. The accuracy of the method is 6.8% (1.sigma.) for cylinders containing UF.sub.6 with enrichments ranging from 6% to 98% with measurement times varying from 3-6 min. The samples contained from below 1 kg to greater than 16 kg. Since the subject invention relies on fast neutron self-interrogation, complete sampling of the UF.sub.6 takes place, reducing difficulties arising from inhomogeneity of the sample which adversely affects other assay procedures.

  13. Direct fissile assay of enriched uranium using random self-interrogation and neutron coincidence response

    DOE Patents [OSTI]

    Menlove, H.O.; Stewart, J.E.

    1985-02-04

    Apparatus and method for the direct, nondestructive evaluation of the /sup 235/U nuclide content of samples containing UF/sub 6/, UF/sub 4/, or UO/sub 2/ utilizing the passive neutron self-interrogation of the sample resulting from the intrinsic production of neutrons therein. The ratio of the emitted neutron coincidence count rate to the total emitted neutron count rate is determined and yields a measure of the bulk fissile mass. The accuracy of the method is 6.8% (1sigma) for cylinders containing UF/sub 6/ with enrichments ranging from 6% to 98% with measurement times varying from 3-6 min. The samples contained from below 1 kg to greater than 16 kg. Since the subject invention relies on fast neutron self-interrogation, complete sampling of the UF/sub 6/ takes place, reducing difficulties arising from inhomogeneity of the sample which adversely affects other assay procedures. 4 figs., 1 tab.

  14. United abominations: Density functional studies of heavy metal chemistry

    SciTech Connect (OSTI)

    Schoendorff, George

    2012-04-02

    Carbonyl and nitrile addition to uranyl (UO{sup 2}{sup 2+}) are studied. The competition between nitrile and water ligands in the formation of uranyl complexes is investigated. The possibility of hypercoordinated uranyl with acetone ligands is examined. Uranyl is studied with diactone alcohol ligands as a means to explain the apparent hypercoordinated uranyl. A discussion of the formation of mesityl oxide ligands is also included. A joint theory/experimental study of reactions of zwitterionic boratoiridium(I) complexes with oxazoline-based scorpionate ligands is reported. A computational study was done of the catalytic hydroamination/cyclization of aminoalkenes with zirconium-based catalysts. Techniques are surveyed for programming for graphical processing units (GPUs) using Fortran.

  15. A study on a voloxidizer with an oxygen concentration controller for a scale-up DESIGN

    SciTech Connect (OSTI)

    Kim, Young-Hwan; Yoon, Ji-Sup; Park, Byung-Suk; Jung, Jae-Hoo

    2007-07-01

    For a oxidation of UO{sub 2} pellets of tens/kg in a vol-oxidizer, the existing devices take a long time, also, for their scale-up to an engineering scale, we need the optimum oxygen concentration with an maximum oxidation efficiency. In this study, we attained the optimum oxygen concentration to shorten the oxidation time of a simulation fuel using a vol-oxidizer with an oxygen concentration controller and sensor. We compared the characteristics of a galvanic sensor with a zirconium oxide (ZrO{sub 2}) one. The simulation fuel was manufactured with 14 metallic oxides, and used at a mass of 500 g HM/batch. At 500 deg. C, the galvanic and zirconium oxide sensors measured the oxidation time for the simulation fuel. Also, the oxidation time of the simulation fuel was measured according to a change of the oxygen concentration with the selected sensor, and the sample was analyzed. (authors)

  16. Development of FLUOREX Process as a Progressive LWR Reprocessing System

    SciTech Connect (OSTI)

    Sasahira, Akira; Kani, Yuko; Iino, Kenji; Hoshino, Kuniyoshi; Kawamura, Fumio

    2007-07-01

    New LWR fuel reprocessing technology named FLUOREX, the hybrid process of fluoride volatility and solvent extraction, proposed here is suitable for future thermal/fast reactors (coexistence) cycle. Recently we developed the flame reactor which was the main equipment to reduce volume of the spent nuclear fuel in order to down-size the PUREX like purification system. Newly constructed flame reactor, having 300 gU/h to 1 kgU/h production rate, was employed to examine the fluorination efficiency of UO{sub 2}. It was successfully demonstrated that we can control the residual amount of U within a range of 2 to 10 % by changing the excess amount of F{sub 2} gas during fluorination. This amount indicated that spent LWR fuel would be reduced to 8 to 15% by fluorination. (authors)

  17. Reactivity control mechanisms for a HPLWR fuel assembly

    SciTech Connect (OSTI)

    Schlagenhaufer, Marc; Schulenberg, Thomas; Vogt, Bastian

    2007-07-01

    A parametric study of different reactivity control mechanisms has been performed for the cross section of a single fuel assembly of a High Performance Light Water Reactor using the Monte Carlo code MCNP5. The fuel temperature feedback, known as the Doppler Effect, and the coolant density feedback have been determined for fresh UO{sub 2} fuel in a large range of fuel and coolant temperatures. The local shutdown reactivity of different control rods with different absorber materials has been predicted. The neutron flux inside the control rods, the power profile in the fuel pins with and without control rods and the coolant density coefficient have been evaluated for future core optimization. Methods to improve the power profile with additional absorbers mounted outside the fuel cluster have been studied exemplarily. (authors)

  18. MOX Reprocessing at Tokai Reprocessing Plant

    SciTech Connect (OSTI)

    Taguchi, Katsuya; Nagaoka, Shinichi; Yamanaka, Atsushi; Nakamura, Yoshinobu; Omori, Eiichi; SATO, Takehiko; MIURA, Nobuyuki

    2007-07-01

    In March 2007, the first reprocessing of the 'Type B' MOX spent fuels of the Prototype Advanced Thermal Reactor FUGEN was initiated at Tokai Reprocessing Plant as a plant-scale demonstration of MOX fuel reprocessing. The operation was advanced satisfactorily and it has been confirmed that the MOX fuels as well as UO{sub 2} fuels can be reprocessed safely. Some characteristics of MOX fuels on reprocessing, such as properties of undissolved residue affecting the clarification process, are becoming visible. Reprocessing of the 'Type B' MOX fuels will be continued for several more years from now on, further investigations on solubility of fuels, characteristics of undissolved residues, progress of solvent degradation and so on will be continued. (authors)

  19. Partitioning of minor actinides from PUREX raffinate by the TODGA process

    SciTech Connect (OSTI)

    Magnusson, D.; Christiansen, B.; Glatz, J.P.; Malmbeck, R.; Serrano Purroy, D.; Modolo, G.; Sorel, C.

    2007-07-01

    A genuine High Active Raffinate (HAR) was produced from small scale PUREX reprocessing of a UO{sub 2} spent fuel solution as feed for a subsequent TODGA/TBP process. In this process, efficient recovery of the trivalent Minor Actinides (MA) actinides could be demonstrated using a hot cell set-up of 32 centrifugal contactor stages. The feed decontamination factors obtained for Am and Cm were in the range of 4 x 10{sup 4} which corresponds to a recovery of more than 99.99 % in the product fraction. Trivalent lanthanides and Y were co-extracted, otherwise only a small part of the Ru ended up in the product. The collected actinide/lanthanide fraction can be used as feed for a SANEX (separation actinides from lanthanides) with some modification of the acidity depending on the extracting molecule. (authors)

  20. Application of the DART Code for the Assessment of Advanced Fuel Behavior

    SciTech Connect (OSTI)

    Rest, J.; Totev, T.

    2007-07-01

    The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO{sub 2} fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)

  1. Removal of dissolved actinides from alkaline solutions by the method of appearing reagents

    DOE Patents [OSTI]

    Krot, Nikolai N.; Charushnikova, Iraida A.

    1997-01-01

    A method of reducing the concentration of neptunium and plutonium from alkaline radwastes containing plutonium and neptunium values along with other transuranic values produced during the course of plutonium production. The OH.sup.- concentration of the alkaline radwaste is adjusted to between about 0.1M and about 4M. [UO.sub.2 (O.sub.2).sub.3 ].sup.4- ion is added to the radwastes in the presence of catalytic amounts of Cu.sup.+2, Co.sup.+2 or Fe.sup.+2 with heating to a temperature in excess of about 60.degree. C. or 85.degree. C., depending on the catalyst, to coprecipitate plutonium and neptunium from the radwaste. Thereafter, the coprecipitate is separated from the alkaline radwaste.

  2. Preliminary Investigation of Candidate Materials for Use in Accident Resistant Fuel

    SciTech Connect (OSTI)

    Jason M. Harp; Paul A. Lessing; Blair H. Park; Jakeob Maupin

    2013-09-01

    As part of a Collaborative Research and Development Agreement (CRADA) with industry, Idaho National Laboratory (INL) is investigating several options for accident resistant uranium compounds including silicides, and nitrides for use in future light water reactor (LWR) fuels. This work is part of a larger effort to create accident tolerant fuel forms where changes to the fuel pellets, cladding, and cladding treatment are considered. The goal fuel form should have a resistance to water corrosion comparable to UO2, have an equal to or larger thermal conductivity than uranium dioxide, a melting temperature that allows the material to stay solid under power reactor conditions, and a uranium loading that maintains or improves current LWR power densities. During the course of this research, fuel fabricated at INL will be characterized, irradiated at the INL Advanced Test Reactor, and examined after irradiation at INL facilities to help inform industrial partners on candidate technologies.

  3. Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Kontogeorgakos, D.; Derstine, K.; Wright, A.; Bauer, T.; Stevens, J.

    2013-06-01

    The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO2 particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.

  4. 18 years experience on UF{sub 6} handling at Japanese nuclear fuel manufacturer

    SciTech Connect (OSTI)

    Fujinaga, H.; Yamazaki, N.; Takebe, N.

    1991-12-31

    In the spring of 1991, a leading nuclear fuel manufacturing company in Japan, celebrated its 18th anniversary. Since 1973, the company has produced over 5000 metric ton of ceramic grade UO{sub 2} powder to supply to Japanese fabricators, without major accident/incident and especially with a successful safety record on UF{sub 6} handling. The company`s 18 years experience on nuclear fuel manufacturing reveals that key factors for the safe handling of UF{sub 6} are (1) installing adequate facilities, equipped with safety devices, (2) providing UF{sub 6} handling manuals and executing them strictly, and (3) repeating on and off the job training for operators. In this paper, equipment and the operation mode for UF{sub 6} processing at their facility are discussed.

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

    SciTech Connect (OSTI)

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

    1991-12-31

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

  6. Release of UF/sub 6/ from a ruptured model 48Y cylinder at Sequoyah Fuels Corporation Facility: lessons-learned report

    SciTech Connect (OSTI)

    Not Available

    1986-08-01

    The uranium hexafluoride (UF/sub 6/) release of January 4, 1986, at the Sequoyah Fuels Corporation facility has been reviewed by a NRC Lessons-Learned Group. A Model 48Y cylinder containing UF/sub 6/ ruptured upon being heated after it was grossly overfilled. The UF/sub 6/ released upon rupture of the cylinder reacted with airborne moisture to produce hydrofluoric acid (HF) and uranyl fluoride (UO/sub 2/F/sub 2/). One individual died from exposure to airborne HF and several others were injured. There were no significant immediate effects from exposure to uranyl fluoride. This supplement report contains NRC's response to the recommendations made in NUREG-1198 by the Lessons Learned Group. In developing a response to each of the recommendations, the staff considered actions that should be taken: (1) for the restart of the Sequoyah Fuels Facility; (2) to make near-term improvement; and (3) to improve the regulatory framework.

  7. URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

    DOE Patents [OSTI]

    Handwerk, J.H.; Noland, R.A.; Walker, D.E.

    1957-09-10

    In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

  8. Conversion of radioactive ferrocyanide compounds to immobile glasses

    DOE Patents [OSTI]

    Schulz, W.W.; Dressen, A.L.

    1975-11-21

    A method is described for converting complex radioactive ferrocyanide compounds of /sup 134/Cs and /sup 137/Cs to immobile glass that is resistant to leaching by water. The /sup 134/Cs and /sup 137/Cs are separated from nuclear waste solutions by precipitation from alkaline solutions by the addition of a soluble Ni/sup 2 +/, Zn/sup 2 +/, Cu/sup 2 +/, Co/sup 2 +/, UO/sub 2//sup 2 +/, or Mn/sup 2 +/ and K/sub 4/Fe(CN)/sub 6/. The dried, finely ground precipitate is mixed with Na/sub 2/CO/sub 3/ and a mixture of (a) basalt and B/sub 2/O/sub 3/ or (b) SiO/sub 2/ and CaO, melted, and allowed to solidify. (BLM)

  9. M3FT-15OR0202237: Submit Report on Results From Initial Coating Layer Development For UN TRISO Particles

    SciTech Connect (OSTI)

    Jolly, Brian C.; Lindemer, Terrence; Terrani, Kurt A.

    2015-02-01

    In support of fully ceramic matrix (FCM) fuel development, coating development work has begun at the Oak Ridge National Laboratory (ORNL) to produce tri-isotropic (TRISO) coated fuel particles with UN kernels. The nitride kernels are used to increase heavy metal density in these SiC-matrix fuel pellets with details described elsewhere. The advanced gas reactor (AGR) program at ORNL used fluidized bed chemical vapor deposition (FBCVD) techniques for TRISO coating of UCO (two phase mixture of UO2 and UCx) kernels. Similar techniques were employed for coating of the UN kernels, however significant changes in processing conditions were required to maintain acceptable coating properties due to physical property and dimensional differences between the UCO and UN kernels.

  10. Analysis of molten debris freezing and wall erosion during a severe RIA test. [PWR; BWR

    SciTech Connect (OSTI)

    El-Genk, M.S.; Moore, R.L.

    1980-01-01

    A one-dimensional physical model was developed to study the transient freezing of the molten debris layer (a mixture of UO/sub 2/ fuel and zircaloy cladding) produced in a severe reactivity initiated accident in-pile test and deposited on the inner surface of the test shroud wall. The wall had a finite thickness and was cooled along its outer surface by coolant bypass flow. Analyzed are the effects of debris temperature, radiation cooling at the debris layer surface, zircaloy volume ratio within the debris, and initial wall temperature on the transient freezing of the debris layer and the potential melting of the wall. The governing equations of this two-component, simultaneous freezing and melting problem in a finite geometry were solved using a one-dimensional finite element code based on the method of weighted residuals.

  11. EPRI/B and W cooperative program on PWR fuel-rod performance. Final report

    SciTech Connect (OSTI)

    Papazoglou, T.P.; Davis, H.H.

    1983-03-01

    Zircaloy-4 fuel cladding specimens were irradiated in a fueled and non-fueled condition for two and four cyles of irradiation, respectively, in the Oconee 2 reactor. The purpose of this long-term surveillance program was to study the in-reactor performance of four Zircaloy-4 cladding types with distinctly different properties, in combination with two types of UO/sub 2/ fuel pellets. The cladding types included Sandvik Special Metals tubing in the cold-worked/stress relieved and cold-worked/recrystallized conditions, and German VDM cladding with two different anneal temperatures. The fuel pellets included a conventional densifying pellet type, and a special (shorter) stable pellet type intended to reduce pellet-clad mechanical interaction. The irradiation growth and creep under compressive stress of the above cladding types were studied and followed up to fluences of 1.3 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV).

  12. How are the energy waves blocked on the way from hot to cold?

    SciTech Connect (OSTI)

    Bai, Xianming; He, Lingfeng; Khafizov, Marat; Yu, Jianguo; Chernatynskiy, Aleksandr

    2013-07-18

    Representing the Center for Materials Science of Nuclear Fuel (CMSNF), this document is one of the entries in the Ten Hundred and One Word Challenge. As part of the challenge, the 46 Energy Frontier Research Centers were invited to represent their science in images, cartoons, photos, words and original paintings, but any descriptions or words could only use the 1000 most commonly used words in the English language, with the addition of one word important to each of the EFRCs and the mission of DOE energy. The mission of CMSNF to develop an experimentally validated multi-scale computational capability for the predictive understanding of the impact of microstructure on thermal transport in nuclear fuel under irradiation, with ultimate application to UO2 as a model system

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

    SciTech Connect (OSTI)

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

    2012-05-15

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

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

    SciTech Connect (OSTI)

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

    2005-05-15

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

  15. 2010-2011 Seminar Calendar

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

    ---2011 S eminar Calendar for CSTEC CSTEC S eminars i n 1 504 G GBL o n s elected T hursdays. R efreshments a t 4 :00pm. S eminar a t 4 :15pm. *MSE/CSTEC S eminars i n 1 670 C SE o n s elected F ridays. R efreshments a t 3 :30pm. S eminar a t 3 :45pm. Sept. 2 3 Zhaohui Z hong GGBL 1 504 University o f M ichigan "Photocurrent G eneration a t C arbon N anotube/Polymer H eterojunctions" Oct. 2 2* Yi L uo CSE 1 670 Carnegie M ellon "Efficient P olymeric P hotovoltaics w ith 3

  16. Methodology for Developing the REScheckTM Software through Version 4.2

    SciTech Connect (OSTI)

    Bartlett, Rosemarie; Connell, Linda M.; Gowri, Krishnan; Lucas, R. G.; Schultz, Robert W.; Taylor, Zachary T.; Wiberg, John D.

    2009-08-31

    This report explains the methodology used to develop Version 4.2 of the REScheck software developed for the 1992, 1993, and 1995 editions of the MEC, and the 1998, 2000, 2003, and 2006 editions of the IECC, and the 2006 edition of the International Residential Code (IRC). Although some requirements contained in these codes have changed, the methodology used to develop the REScheck software for these five editions is similar. REScheck assists builders in meeting the most complicated part of the code?the building envelope Uo-, U-, and R-value requirements in Section 502 of the code. This document details the calculations and assumptions underlying the treatment of the code requirements in REScheck, with a major emphasis on the building envelope requirements.

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

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

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

  18. Current State of Knowledge of Water Radiolysis Effects on Spent Nuclear Fuel Corrosion

    SciTech Connect (OSTI)

    Christensen, H.; Sunder, S.

    2000-07-15

    Literature data on the effect of water radiolysis products on spent-fuel oxidation and dissolution are reviewed. Effects of gamma radiolysis, alpha radiolysis, and dissolved O{sub 2} or H{sub 2}O{sub 2} in unirradiated solutions are discussed separately. Also, the effect of carbonate in gamma-irradiated solutions and radiolysis effects on leaching of spent fuel are reviewed. In addition, a kinetic model for calculating the corrosion rates of UO{sub 2} in solutions undergoing radiolysis is discussed. The model gives good agreement between calculated and measured corrosion rates in the case of gamma radiolysis and in unirradiated solutions containing dissolved oxygen or hydrogen peroxide. However, the model fails to predict the results of alpha radiolysis. In a recent study, it was shown that the model gave good agreement with measured corrosion rates of spent fuel exposed in deionized water. The applications of radiolysis studies for geologic disposal of used nuclear fuel are discussed.

  19. A practical strategy for reducing the future security risk of United States spent nuclear fuel

    SciTech Connect (OSTI)

    Chodak, P. III; Buksa, J.J.

    1997-06-01

    Depletion calculations show that advanced oxide (AOX) fuels can be used in existing light water reactors (LWRs) to achieve and maintain virtually any desired level of US (US) reactor-grade plutonium (R-Pu) inventory. AOX fuels are composed of a neutronically inert matrix loaded with R-Pu and erbium. A 1/2 core load of 100% nonfertile, 7w% R-Pu AOX and 3.9 w% UO{sub 2} has a net total plutonium ({sup TOT}Pu) destruction rate of 310 kg/yr. The 20% residual {sup TOT}Pu in discharged AOX contains > 55% {sup 242}Pu making it unattractive for nuclear explosive use. A three-phase fuel-cycle development program sequentially loading 60 LWRs with 100% mixed oxide, 50% AOX with a nonfertile component displacing only some of the {sup 238}U, and 50% AOX, which is 100% nonfertile, could reduce the US plutonium inventory to near zero by 2050.

  20. Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

    SciTech Connect (OSTI)

    R. L. Williamson

    2011-08-01

    A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

  1. Modeling the influence of bubble pressure on grain boundary separation and fission gas release

    SciTech Connect (OSTI)

    Pritam Chakraborty; Michael R. Tonks; Giovanni Pastore

    2014-09-01

    Grain boundary (GB) separation as a mechanism for fission gas release (FGR), complementary to gas bubble interlinkage, has been experimentally observed in irradiated light water reactor fuel. However there has been limited effort to develop physics-based models incorporating this mechanism for the analysis of FGR. In this work, a computational study is carried out to investigate GB separation in UO2 fuel under the effect of gas bubble pressure and hydrostatic stress. A non-dimensional stress intensity factor formula is obtained through 2D axisymmetric analyses considering lenticular bubbles and Mode-I crack growth. The obtained functional form can be used in higher length-scale models to estimate the contribution of GB separation to FGR.

  2. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  3. Uranium dioxide electrolysis

    DOE Patents [OSTI]

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

    2009-12-29

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

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

    SciTech Connect (OSTI)

    Yoshida, Thomas M.; Meyers, Lisa

    2015-11-10

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

  5. Multi-scale modeling of microstructure dependent intergranular brittle fracture using a quantitative phase-field based method

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

    Chakraborty, Pritam; Zhang, Yongfeng; Tonks, Michael R.

    2015-12-07

    In this study, the fracture behavior of brittle materials is strongly influenced by their underlying microstructure that needs explicit consideration for accurate prediction of fracture properties and the associated scatter. In this work, a hierarchical multi-scale approach is pursued to model microstructure sensitive brittle fracture. A quantitative phase-field based fracture model is utilized to capture the complex crack growth behavior in the microstructure and the related parameters are calibrated from lower length scale atomistic simulations instead of engineering scale experimental data. The workability of this approach is demonstrated by performing porosity dependent intergranular fracture simulations in UO2 and comparing themore » predictions with experiments.« less

  6. Action Sheet 36 Final Report

    SciTech Connect (OSTI)

    Kips, R E; Kristo, M J; Hutcheon, I D

    2012-02-24

    Pursuant to the Arrangement between the European Commission DG Joint Research Centre (EC-JRC) and the Department of Energy (DOE) to continue cooperation on research, development, testing, and evaluation of technology, equipment, and procedures in order to improve nuclear material control, accountancy, verification, physical protection, and advanced containment and surveillance technologies for international safeguards, dated 1 September 2008, the IRMM and LLNL established cooperation in a program on the Study of Chemical Changes in Uranium Oxyfluoride Particles under IRMM-LLNL Action Sheet 36. The work under this action sheet had 2 objectives: (1) Achieve a better understanding of the loss of fluorine in UO{sub 2}F{sub 2} particles after exposure to certain environmental conditions; and (2) Provide feedback to the EC-JRC on sample reproducibility and characteristics.

  7. Implementation of a spark plasma sintering facility in a hermetic glovebox for compaction of toxic, radiotoxic, and air sensitive materials

    SciTech Connect (OSTI)

    Tyrpekl, V. E-mail: vaclav.tyrpekl@gmail.com; Berkmann, C.; Holzhäuser, M.; Köpp, F.; Cologna, M.; Somers, J.; Wangle, T.

    2015-02-15

    Spark plasma sintering (SPS) is a rapidly developing method for densification of powders into compacts. It belongs to the so-called “field assisted sintering techniques” that enable rapid sintering at much lower temperatures than the classical approaches of pressureless sintering of green pellets or hot isostatic pressing. In this paper, we report the successful integration of a SPS device into a hermetic glovebox for the handling of highly radioactive material containing radioisotopes of U, Th, Pu, Np, and Am. The glovebox implantation has been facilitated by the replacement of the hydraulic system to apply pressure with a compact electromechanical unit. The facility has been successfully tested using UO{sub 2} powder. Pellets with 97% of the theoretical density were obtained at 1000 °C for 5 min, significantly lower than the ∼1600 °C for 5-10 h used in conventional pellet sintering.

  8. Mechanistic prediction of fission-product release under normal and accident conditions: key uncertainties that need better resolution. [PWR; BWR

    SciTech Connect (OSTI)

    Rest, J.

    1983-09-01

    A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO/sub 2/-base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles.

  9. Enhancing the ABAQUS Thermomechanics Code to Simulate Steady and Transient Fuel Rod Behavior

    SciTech Connect (OSTI)

    R. L. Williamson; D. A. Knoll

    2009-09-01

    A powerful multidimensional fuels performance capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth , gap heat transfer, and gap/plenum gas behavior during irradiation. The various modeling capabilities are demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multi-pellet fuel rod, during both steady and transient operation. Computational results demonstrate the importance of a multidimensional fully-coupled thermomechanics treatment. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermo-mechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

  10. Update of Energy Efficiency Requirements for Manufactured Homes

    SciTech Connect (OSTI)

    Conner, Craig C.; Dillon, Heather E.; Lucas, Robert G.; Early, Chris; Lubliner, Michael

    2004-03-22

    Energy efficiency requirements were developed for manufactured (mobile) homes, which are regulated by the U.S. Department of Housing and Urban Development (HUD). A life-cycle cost analysis from the homeowner's perspective was used to establish parameters for a least-cost home in a large number of cities. Economic, financial, and energy-efficiency measures for the life-cycle cost analysis were selected. The resulting energy-efficiency levels were aggregated to the existing HUD zones and expressed as a maximum overall home U-value (thermal transmittance) requirement for the building envelope. The proposed revised standard's costs, benefits, and net value to the consumer were quantified. This analysis updates a similar effort completed in 1992, which was the basis for the existing HUD code Uo requirements.

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

    SciTech Connect (OSTI)

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

    2015-03-03

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

  12. THERMODYNAMIC MODEL FOR URANIUM DIOXIDE BASED NUCLEAR FUEL

    SciTech Connect (OSTI)

    Thompson, Dr. William T.; Lewis, Dr. Brian J; Corcoran, E. C.; Kaye, Dr. Matthew H.; White, S. J.; Akbari, F.; Higgs, Jamie D.; Thompson, D. M.; Besmann, Theodore M; Vogel, S. C.

    2007-01-01

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

  13. Static electric dipole polarizabilities of An{sup 5+/6+} and AnO{sub 2}{sup +/2+} (An = U, Np, and Pu) ions

    SciTech Connect (OSTI)

    Parmar, Payal E-mail: kipeters@wsu.edu Peterson, Kirk A. E-mail: kipeters@wsu.edu; Clark, Aurora E. E-mail: kipeters@wsu.edu

    2014-12-21

    The parallel components of static electric dipole polarizabilities have been calculated for the lowest lying spin-orbit states of the penta- and hexavalent oxidation states of the actinides (An) U, Np, and Pu, in both their atomic and molecular diyl ion forms (An{sup 5+/6+} and AnO{sub 2}{sup +/2+}) using the numerical finite-field technique within a four-component relativistic framework. The four-component Dirac-Hartree-Fock method formed the reference for MP2 and CCSD(T) calculations, while multireference Fock space coupled-cluster (FSCC), intermediate Hamiltonian Fock space coupled-cluster (IH-FSCC) and Kramers restricted configuration interaction (KRCI) methods were used to incorporate additional electron correlation. It is observed that electron correlation has significant (?5 a.u.{sup 3}) impact upon the parallel component of the polarizabilities of the diyls. To the best of our knowledge, these quantities have not been previously reported and they can serve as reference values in the determination of various electronic and response properties (for example intermolecular forces, optical properties, etc.) relevant to the nuclear fuel cycle and material science applications. The highest quality numbers for the parallel components (?{sub zz}) of the polarizability for the lowest ? levels corresponding to the ground electronic states are (in a.u.{sup 3}) 44.15 and 41.17 for UO{sub 2}{sup +} and UO{sub 2}{sup 2+}, respectively, 45.64 and 41.42 for NpO{sub 2}{sup +} and NpO{sub 2}{sup 2+}, respectively, and 47.15 for the PuO{sub 2}{sup +} ion.

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

    SciTech Connect (OSTI)

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

    2015-03-03

    We present a detailed mechanism of, and the effect of homogeneous strains on, the migration of uranium vacancies in UO2. Vacancy migration pathways and barriers are identified using density functional theory and the effect of uniform strain fields are accounted for using the dipole tensor approach. We report complex migration pathways and noncubic symmetry associated with the uranium vacancy in UO2 and show that these complexities need to be carefully accounted for to predict the correct diffusion behavior of uranium vacancies. We show that under homogeneous strain fields, only the dipole tensor of the saddle with respect to the minimum is required to correctly predict the change in the energy barrier between the strained and the unstrained case. Diffusivities are computed using kinetic Monte Carlo simulations for both neutral and fully charged state of uranium single and divacancies. We calculate the effect of strain on migration barriers in the temperature range 8001800 K for both vacancy types. Homogeneous strains as small as 2% have a considerable effect on diffusivity of both single and divacancies of uranium, with the effect of strain being more pronounced for single vacancies than divacancies. In contrast, the response of a given defect to strain is less sensitive to changes in the charge state of the defect. Further, strain leads to anisotropies in the mobility of the vacancy and the degree of anisotropy is very sensitive to the nature of the applied strain field for strain of equal magnitude. Our results indicate that the influence of strain on vacancy diffusivity will be significantly greater when single vacancies dominate the defect structure, such as sintering, while the effects will be much less substantial under irradiation conditions where divacancies dominate.

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

    SciTech Connect (OSTI)

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

    1993-08-01

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

  16. Production plant separator system conceptual design

    SciTech Connect (OSTI)

    Ng, E.; Kan, T.

    1994-12-31

    A full conceptual design has been completed for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant capable of producing {approximately}1700 metric tons of enriched uranium per year (MTU/y). This plant is the first step in the deployment of AVLIS enrichment technology, which will provide inexpensive, dependable, and environmentally safe uranium enrichment services to utility customers. Previous issues of the ISAM Semiannual Report describe other major systems in the plant, namely the laser, feed and product systems. This article describes the design of the separator system. The separator system is a a key component in the plant. After the feed conversion system converts uranium trioxide (UO{sub 3}) to a uranium-iron alloy, the alloy enters the separator system. In the separator, and intense electron beam vaporizes uranium metal in a vacuum chamber. In the laser system, fixed-frequency copper-vapor lasers pump tunable dye lasers. These precisely tuned dye lasers then selectively excite and ionize uranium-235 atoms in the vapor stream, leaving the uranium-238 atoms untouched. The photo-ions of uranium-235 are then drawn to an electrically biased collector, producing the enriched product stream. The remaining vapor flows through, producing the depleted tails stream. Both product and tails streams are continuously removed from the separator pod as flowing liquid uranium metal. Withdrawal containers are used to collect separately the enriched and depleted uranium. The enriched product will be converted by fuel fabricators to uranium dioxide (UO{sub 2}) and used to fabricate reactor fuel assemblies for utility customers.

  17. A nuclear criticality safety assessment of the loss of moderation control in 2 1/2 and 10-ton cylinders containing enriched UF{sub 6}

    SciTech Connect (OSTI)

    Newvahner, R.L.; Pryor, W.A.

    1991-12-31

    Moderation control for maintaining nuclear criticality safety in 2 {1/2}-ton, 10-ton, and 14-ton cylinders containing enriched uranium hexafluoride (UF{sub 6}) has been used safely within the nuclear industry for over thirty years, and is dependent on cylinder integrity and containment. This assessment evaluates the loss of moderation control by the breaching of containment and entry of water into the cylinders. The first objective of this study was to estimate the required amounts of water entering these large UF{sub 6} cylinders to react with, and to moderate the uranium compounds sufficiently to cause criticality. Hypothetical accident situations were modeled as a uranyl fluoride (UO{sub 2}F{sub 2}) slab above a UF{sub 6} hemicylinder, and a UO{sub 2}F{sub 2} sphere centered within a UF{sub 6} hemicylinder. These situations were investigated by computational analyses utilizing the KENO V.a Monte Carlo Computer Code. The results were used to estimate both the masses of water required for criticality, and the limiting masses of water that could be considered safe. The second objective of the assessment was to calculate the time available for emergency control actions before a criticality would occur, i.e., a {open_quotes}safetime{close_quotes}, for various sources of water and different size openings in a breached cylinder. In the situations considered, except the case for a fire hose, the safetime appears adequate for emergency control actions. The assessment shows that current practices for handling moderation controlled cylinders of low enriched UF{sub 6}, along with the continuation of established personnel training programs, ensure nuclear criticality safety for routine and emergency operations.

  18. Design of Mega-Voltage X-ray Digital Radiography and Computed Tomography Performance Phantoms

    SciTech Connect (OSTI)

    Aufderheide, M B; Martz, H E; Curtin, M

    2009-06-22

    A number of fundamental scientific questions have arisen concerning the operation of high-energy DR and CT systems. Some of these questions include: (1) How deeply can such systems penetrate thickly shielded objects? (2) How well can such systems distinguish between dense and relatively high Z materials such as lead, tungsten and depleted uranium and lower Z materials such as steel, copper and tin? (3) How well will such systems operate for a uranium material which is an intermediate case between low density yellowcake and high density depleted uranium metal? These questions have led us to develop a set of phantoms to help answer these questions, but do not have any direct bearing on any smuggling concern. These new phantoms are designed to allow a systemic exploration of these questions by gradually varying their compositions and thicknesses. These phantoms are also good probes of the blurring behavior of radiography and tomography systems. These phantoms are composed of steel ({rho} assumed to be 7.8 g/cc), lead ({rho} assumed to be 11.4 g/cc), tungsten ({rho} assumed to be 19.25 g/cc), uranium oxide (UO{sub 3}) ({rho} assumed to be 4.6 g/cc), and depleted uranium (DU) ({rho} assumed to be 18.9 g/cc). There are five designed phantoms described in this report: (1) Cylindrical shells of Tungsten and Steel; (2) Depleted Uranium Inside Tungsten Hemi-cube Shells; (3) Nested Spherical Shells; (4) UO{sub 3} Cylinder; and (5) Shielded DU Sphere.

  19. Solvent extraction of thorium(IV), uranium(VI), and europium(III) with lipophilic alkyl-substituted pyridinium salts. Final report for subcontract 9-XZ2-1123E-1, June 1, 1992--December 1, 1995

    SciTech Connect (OSTI)

    Ensor, D.D.

    1997-01-01

    In the treatment of high level nuclear wastes, aromatic pyridinium salts which are radiation-resistant are desired for the extraction of actinides and lanthanides. The solvent extraction of Th{sup +4}, UO{sub 2}{sup +2}, and Eu{sup +3} by three aromatic extractants, 3,5-didodecylpyridinium nitrate (35PY), 2,6-didodecylpyridinium nitrate (26PY), and 1-methyl-3,5-didodecyl-pyridinium iodide (1M35PY) has been studied in nitric acid media. The general order of extractability of the three extractants in toluene was 1M35PY>> 26PY > 35PY. The overall extraction efficiency of the metal ions was Th{sup +4} >UO{sub 2}{sup +2} > Eu{sup +3}. The extraction of HNO{sub 3}, which was competitive with the extraction of metal ions, was quantitatively investigated by NaOH titration and UV spectrometry. The loading capacity suggested that the extracted species in the organic phase for thorium was (R{sub 4}N{sup +}){sub 2}Th(NO{sub 3}{sup -}){sub 6}, where R{sub 4}N{sup +} denotes 1M35PY. A comparison of 1M35PY to the well-characterized extractant, Aliquat-336, an aliphatic ammonium salt was made. At the same extractant concentration, 1M35PY extracted thorium more efficiently than Aliquat-336 at high acidity. Thorium could be readily stripped with dilute nitric acid from 1M35PY. After irradiation of 0.1M 1M35PY with {sup 60}Co at 40R/min for 48 hours, no change in the extraction efficiency of thorium was observed.

  20. Kinetic studies of the [NpO? (CO?)?]?? ion at alkaline conditions using C NMR

    SciTech Connect (OSTI)

    Panasci, Adele F.; Harley, Stephen J.; Zavarin, Mavrik; Casey, William H.

    2014-04-21

    Carbonate ligand-exchange rates on the [NpO? (CO?)?]?? ion were determined using a saturation-transfer C nuclear magnetic resonance (NMR) pulse sequence in the pH range of 8.1 ? pH ? 10.5. Over the pH range 9.3 ? pH ? 10.5, which compares most directly with previous work of Stout et al.,1 we find an average rate, activation energy, enthalpy, and entropy of k298ex = 40.6(4.3) s?, Ea =45.1(3.8) kJ mol?, ?H = 42.6(3.8) kJ mol?, and ?S = -72(13) J mol? K?, respectively. These activation parameters are similar to the Stout et al. results at pH 9.4. However, their room-temperature rate at pH 9.4, k298ex = 143(1.0) s?, is ~3 times faster than what we experimentally determined at pH 9.3: k298ex = 45.4(5.3) s?. Our rates for [NpO? (CO?)?]?? are also faster by a factor of ~3 relative to the isoelectronic [UO?(CO?)?]?? as reported by Brucher et al.2 of k298ex = 13(3) s?. Consistent with results for the [UO?(CO?)?]?? ion, we find evidence for a proton-enhanced pathway for carbonate exchange for the [NpO?(CO?)?]?? ion at pH < 9.0.

  1. ASSESSMENT OF POSSIBLE CYCLE LENGTHS FOR FULLY-CERAMIC MICRO-ENCAPSULATED FUEL-BASED LIGHT WATER REACTOR CONCEPTS

    SciTech Connect (OSTI)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal Pasamehmetoglu; Francesco Venneri

    2012-04-01

    The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

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

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

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

    2015-03-03

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

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

    SciTech Connect (OSTI)

    Baybas, Demet; Ulusoy, Ulvi

    2012-10-15

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

  4. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect (OSTI)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  5. Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

    SciTech Connect (OSTI)

    Degueldre, Claude; Wiesenack, Wolfgang

    2008-07-01

    An yttria stabilised zirconia doped with plutonia and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er){sub y}Pu{sub x}Zr{sub 1-y}O{sub 2-{xi}} where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O{sub 2-{xi}} (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia- IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO{sub 2}. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O{sub 2} fuels. The properties of the spent fuel pellets are presented focusing on the once-through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO{sub 2} in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a 'burn and bury' strategy. (authors)

  6. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    SciTech Connect (OSTI)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  7. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect (OSTI)

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

  8. Ms. Chris Andres, Chief Bureau of Federal Facilities Division...

    Office of Legacy Management (LM)

    ... The technical exchange meeting and the Surface Geophysics Report provided the basis for developing the new data acquisition plan that was submitted to NDEP in October 2011. The ...

  9. revised MS A5-ROR text+figures

    Office of Scientific and Technical Information (OSTI)

    ... Eur J Neurosci. 1997;9:2687-2701. 21. Hamilton BA, Frankel WN, Kerrebrock AW, Hawkins TL, FitzHugh W, Kusumi K, Russell LB, Mueller KL, van Berkel V, Birren BW, Kruglyak L, Lander ...

  10. Questions? Contact Ms. Debbie Cutler at cutlerd@osti.gov.

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

    that makes it easy to search a wide range of ... Using translation tools, WWE allows users to enter searches ... of choice, and the search terms are automatically ...

  11. MS4 & IP Connection - PowerPoint Presentation

    SciTech Connect (OSTI)

    Lemke, Terrill W.

    2015-05-20

    An overview of Los Alamos National Lab’s (LANL) storm sewer system, the presentation includes information on the management storm sewer permit and environmental programs and projects of LANL institutions.

  12. Magnetic moment of {sup 43m}S

    SciTech Connect (OSTI)

    Daugas, J. M.; Gaudefroy, L.; Meot, V.; Morel, P.; Rosse, B.; Hass, M.; Kumar, V.; Angelique, J. C.; Simpson, G. S.; Balabanski, D. L.; Fiori, E.; Georgiev, G.; Lozeva, R.; Force, C.; Grevy, S.; Stodel, Ch.; Thomas, J. C.; Kameda, D.; Matea, I.; Singh, B. S. Nara

    2008-11-11

    The gyromagnetic factor of the isomeric state of {sup 43}S has been measured using the Time Dependent Perturbed Angular Distribution (TDPAD) technique. The isomer was produced and spin aligned via the fragmentation of a 60 AMeV {sup 48}Ca beam at the GANIL facility. The deduced magnetic moment confirms the 7/2{sup -} spin/parity of the isomeric state and shows, for the first time, the intruder nature of the ground state. Comparison of the experimental values with Shell Model and mean-field based calculations were performed revealing a pronounced ground state deformation and a quasi-spherical isomeric state. A new isomeric state has been observed in the {sup 42}P.

  13. Ms. Kimberly Krizanovic U.S. Department of Energy

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

    ... has likely already been agreed on by the auditor and auditee based on the contents and the ... of 2011. We are not aware of any communication from the DOE to its award recipients ...

  14. Prepared Statement for Ms. Patricia K. Vincent-Collawn Chairman...

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

    ... Technological obsolescence will drive continuous investment. The electric industry is willing and able to make the needed investments- but they need to be well-planned and costs ...

  15. Ms. Paula Call NEPA Document Manager US Department of Energy

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

    ... the EA Scoping Period for Land Con veyance of 4,413 acres ... to Richland utilities and the future natural gas pipeline. ... Also, the Port of Benton has requirements for ordinary ...

  16. Questions? Contact Ms. Debbie Cutler at cutlerd@osti.gov.

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

    and open websites from around the globe. Using translation tools, WWE allows users to enter searches in their native language and find results across other languages. While WWE...

  17. Climate Mag_27JUN2013_ms07022013.indd

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

    ... melt rapidly and the ice upstream of the shelves ows more ... of fugitive methane (CH 4 ) and hydrocarbon leaks. ... that greenhouse gas emissions from coal power plants ...

  18. U-190: Microsoft Security Bulletin MS12-037- Critical

    Broader source: Energy.gov [DOE]

    This security update resolves one publicly disclosed and twelve privately reported vulnerabilities in Internet Explorer.

  19. Climate Mag_27JUN2013_ms07022013.indd

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

    ability to process and store carbon. 19 PartnersCollaborators Our partners include the United States Geological Survey, Michigan State University, the University of New Mexico,...

  20. Climate Mag_27JUN2013_ms07022013.indd

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

    CLIMATE AND IMPACT RESEARCH at Los Alamos National Laboratory Climate Research and National Security Los Alamos National Laboratory is truly a national security science laboratory, tackling some of the world's most challenging science and engineering issues. We are interested in the potential future impacts of climate change on global security, such as the coastal e ects of sea level rise, increased number of extreme storms, and the consequences of extensive regional tree mortality. Gaining a