National Library of Energy BETA

Sample records for uo uo uo

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

    Office of Scientific and Technical Information (OSTI)

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

  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. 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

  4. 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

  5. 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):

  6. 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.

  7. 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.

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

    SciTech Connect (OSTI)

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

    2009-11-01

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

  9. 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.

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

    SciTech Connect (OSTI)

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

    2015-01-01

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

  11. 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).

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. 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:

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. 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.

  5. 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.

  6. $sup 18$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.

  7. 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.

  8. 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].

  9. 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].

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

    DOE Patents [OSTI]

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

    1994-01-01

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

  11. 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)

  12. 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.

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

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

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

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

    SciTech Connect (OSTI)

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

    2007-06-01

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

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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)

  20. 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.

  1. 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}.

  2. 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.

  3. 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.

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

    SciTech Connect (OSTI)

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

    2014-03-10

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

  5. 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.

  6. 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.

  7. 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)

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

    SciTech Connect (OSTI)

    Uberuaga, Blas Pedro; Andersson, David A.

    2015-10-01

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

  9. 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.

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

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

    SciTech Connect (OSTI)

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

    2014-11-01

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

  16. 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.

  17. 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.

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

    Office of Legacy Management (LM)

    H-S I-H HATIOHAL LEAth~~Y~~OF' OtUO ' Industrial Hygiene No. P.O. Box 158)At.' He&by~kation Sample Nos. ? Sk. 0 qq! Cinchnail 31;Obio Type of SampleCI" lz -- HEALTH AND SAFETY DIVISIQN Analyze for u Method &An.ly,i, ;i __ = - . -- Industrial Hygiene No. P.O. Box 158 Mt.~He&lly Qq99 Q' ' - Ciacian& 31. 01 Sample Nos. 992' I HtAL I H ANU SAFt H-S 1-H J MATI~ LEID-WAIJY. OF OIUO station rio Type of S&h d' r dQsr CYS ..- . . -.. ..- -. -- :TY DIVISICJN Analyze for d t I I I

  19. 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).

  20. [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 +}.

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

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

    SciTech Connect (OSTI)

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

    2014-03-01

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

  3. 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.

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

    SciTech Connect (OSTI)

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

    1996-12-30

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

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. 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

  10. 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

  11. 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.

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

    SciTech Connect (OSTI)

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

    2007-11-01

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

  13. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO? fuel and APMT-Steel Clad/UN-U?Si? fuel concepts

    SciTech Connect (OSTI)

    Unal, Cetin; Galloway, Jack D.

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO? ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U?Si? fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

  14. 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.

  15. 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)

  16. [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.

  17. 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...

  18. 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.

  19. 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.

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

    Office of Scientific and Technical Information (OSTI)

    One strategy to remediate U contamination in the subsurface is the immobilization of U via injection of an electron donor, e.g., acetate, which leads to stimulation of the ...

  1. 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].

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

    Office of Scientific and Technical Information (OSTI)

    Have feedback or suggestions for a way to improve these results? Save Share this Record Citation Formats MLA APA Chicago Bibtex Export Metadata Endnote Excel CSV XML Save to My ...

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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}.

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

    Office of Legacy Management (LM)

    CT Perrine Field, FL Bloomington, IN Layfayette, IN Note 1 South Bend, IN Note 2 Cambridge, MA Medford, MA Baltimore, MD . 1 NAME (Continued) University of Michigan St. Louis...

  12. 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.

  13. 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.

  14. BROADER National Security Missions

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

    Metal Chips (U) Uranium Trioxide (UO 3 ) UO 2 (NO 3 ) 2 Ur anyl Nitrate Ammonium Uranyl Carbonate (NH 4 ) 2 UO 2 (CO 3 ) 4 DEVELOP NEW NATIONAL SECURITY MISSIONS Y-12 has...

  15. 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.

  16. 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)

  17. 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.

  18. CASL Plan of Record 2 (1/11-6/11)

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

    for UO2 Assembly 23 Figure 18. UO 2 assembly peaking factor 23 TABLES Table 1 Analytic measurement techniques and uncertainty 1 11 Table 2. Average of Results for SF95 14...

  19. 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.

  20. 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

  1. 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.

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

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

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

  3. 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

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

    Office of Scientific and Technical Information (OSTI)

    Greenspan, 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 ...

  5. 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.

  6. 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

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

    Office of Scientific and Technical Information (OSTI)

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

  8. U Plant Ancillary Facility Demolition A Department of Energy...

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

    multi-storied structure used for the conversion of uranium solutions to UO3 powder. 203-UX Concentrated Uranium Storage Tank Enclosure: Two outdoor concrete enclosures containing...

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

    SciTech Connect (OSTI)

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

    2014-12-01

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

  10. 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.

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

    SciTech Connect (OSTI)

    Carnahan, C.L.

    1987-12-01

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

  12. 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.

  13. 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,

  14. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1990-01-01

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

  15. END POINT IMPLEMENTATION EXAMPLES End Point Implementation Examples

    Office of Environmental Management (EM)

    ... When the deactivated UO3 Facility no longer had accountable quantities of SNM, vital equipment, or classified information, DOE Order 5632.2A, Physical Protection of Special Nuclear ...

  16. 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....

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

    SciTech Connect (OSTI)

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

    2006-02-21

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

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

    Office of Scientific and Technical Information (OSTI)

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

  19. 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...

  20. 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...

  1. 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

  2. MARMOT Enhanced

    Broader source: Energy.gov [DOE]

    To develop mechanistic models for fuel thermal conductivity, the Fuel team used supercells up to 55 nm long to determine the thermal conductivity of UO2 with Xe incorporated.

  3. 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 µM. 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.

  4. 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...

  5. 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 )...

  6. 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 ...

  7. 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.

  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. NEAMS Update

    Office of Environmental Management (EM)

    July - September 2014 Nuclear Energy ANL/NEAMS-14/4 Overview } } BISON version 1.1 was released in September, and the BISON Assessment Report was updated. } } BISON was given a better capability for modelling hydrogen behavior in Zircaloy cladding. } } A sensitivity analysis of the gap heat transfer model for UO 2 fuel at high burnup was completed. } } The microstructure-based UO 2 fracture model was improved by using the fission gas bubble density at grain boundaries to inform the fracture

  10. 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

  11. 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.

  12. 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).

  13. 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}.

  14. 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 Center’s 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.

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

    SciTech Connect (OSTI)

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

    2014-07-28

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

  16. 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.

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

    SciTech Connect (OSTI)

    Tsai, C.

    1981-11-01

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

  18. 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.

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

    SciTech Connect (OSTI)

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

    2015-09-11

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

  20. 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.

  1. Infrared Multiphoton Dissociation Spectroscopy of a Gas-Phase Complex of Uranyl and 3-Oxa-Glutaramide: An Extreme Red-Shift of the [O=U=O]²? 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)?²? (TMOGA = tetramethyl-3-oxa-glutaramide) prepared by electrospray ionization was characterized by infrared multiphoton dissociation (IRMPD) spectroscopy. The IRMPD spectrum from 700–1800 cm?¹ 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 (<1%) and a maximum deviation of 21 cm?¹ (<2%). The only IR peak assigned to the linear uranyl moiety was the asymmetric ?? mode, which appeared at 965 cm?¹ and was predicted by DFT as 953 cm?¹. This ?? frequency is red-shifted relative to bare uranyl, UO?²?, by ca. 150 cm?¹ 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)?²?. 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)?²?, 950 cm?¹, is essentially the same as that for UO?(TMOGA)?²?, 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)?²?, NpO?(TMOGA)?²? and PuO?(TMOGA)?²?, are comparable. This similarity is discussed in the context of the relationship between ?? and intrinsic actinide-oxygen bond energies in actinyl complexes.

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

    SciTech Connect (OSTI)

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

    2015-10-01

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

  3. 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.

  4. 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 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

  5. 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

  6. Oxidation and crystal field effects in uranium

    SciTech Connect (OSTI)

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

    2015-07-06

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

  7. 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.

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

    SciTech Connect (OSTI)

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

    2014-01-22

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

  10. 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.

  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 1000ºC, 1300ºC, and 1600°C for 1 hour to produce a Kr-bubble dominated microstructure. The results of this work indicate a strong dependence of Kr concentration as a function of grain boundary structure. Temperature also influences grain boundary chemistry with greater Kr concentration evident at higher temperatures, resulting in a reduced Kr concentration in the bulk. While Kr migration is active at elevated temperatures, no changes in grain size or texture were observed in the irradiated UO2 samples.

  12. 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...

  13. 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.

  14. 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.

  15. 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

  16. 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.

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

    SciTech Connect (OSTI)

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

    2008-07-08

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

  18. 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.

  19. 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.

  20. 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.

  1. 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?²?) 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 Å) 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.

  2. 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.

  3. 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)

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

  5. 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).

  6. 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.

  7. 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.

  8. 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.

  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. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

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

    DOE Patents [OSTI]

    Herrmann, Steven D. (Idaho Falls, ID); Mariani, Robert D. (Idaho Falls, ID)

    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.

  17. Microsoft Word - DOE-ID-14-061 Cal-Berkeley EC B3-6.doc

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

    61 SECTION A. Project Title: Selective ligands for uranyl via combinatorial peptoid libraries: A synthetic, structural, thermodynamic and computational study - University of California Berkeley SECTION B. Project Description University of California Berkeley will conduct research aimed at developing a fundamental understanding of how donor ligands bind to the uranyl ion, UO2 2+, with the longer-term goal of using this information to tackle selective recognition of uranyl in aqueous solution.

  18. 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

  19. 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

  20. 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

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

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

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

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

    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,

  3. LANSCE | User Resources

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

    From the initial proposal process to the completion of the experiment, LANSCE provides its users with resources critical to their experiements and their experience. Lujan Resources WNR Resources Submit a proposal for beam time Visit Registration Schedules Experiment Reports User Satisfaction Survey Reviews Submit a proposal for beam time Visit Registration Schedules Experiment Reports User Satisfaction Survey Reviews User Program Administration lujan-uo@lanl.gov Ph: 505.667.6069 User Program

  4. 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 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 Department of Energy, National Nuclear Security Administration nnsa.energy.gov

  5. 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

  6. 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.

  7. NEAMS Update

    Office of Environmental Management (EM)

    January - March 2014 Nuclear Energy ANL/NEAMS-14/2 Overview } } A fully three-dimensional smeared cracking model has been implemented and tested in BISON (page 2). } } DAKOTA-BISON was used to study the parameters that govern heat transfer across the fuel-cladding (page 2). } } Calculations of grain boundary mobility in UO 2 have been extended to high temperatures (page 2). } } Mesh adaptivity is being employed in MARMOT simulations to increase computational efficiency (page 3). } } Molecular

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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.

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

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

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

    2015-04-13

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

  16. 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.

  17. 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)

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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].

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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)

  10. The Purpose and Value of Successful Technology Demonstrations Â… 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

  11. 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.

  12. 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)

  13. Multiscale Speciation of U and Pu at Chernobyl, Hanford, Los Alamos,

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

    McGuire AFB, Mayak, and Rocky Flats | Stanford Synchrotron Radiation Lightsource Multiscale Speciation of U and Pu at Chernobyl, Hanford, Los Alamos, McGuire AFB, Mayak, and Rocky Flats Friday, June 26, 2015 X-ray fluorescence maps of (clockwise from upper right) Ga, U, Ca, Pu, Ti, and K in a 350 micron PuO2-UO2 composite particle produced by the fire that consumed a nuclear armed BOMARC missile at McGuire AFB in 1960, measured with the two micron focused x-ray beam at SSRL. EST June 2,

  14. 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 today’s 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.

  15. 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 today’s 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.

  16. 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.

  17. 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.

  18. A Scoping Analysis Of The Impact Of SiC Cladding On Late-Phase Accident Progression Involving Core–Concrete 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.

  19. 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.

  20. LANSCE | Users

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

    Rosen Scholar Rosen Prize Users dotline User Information Planning for Arrival: Contact the User Office several weeks before arriving at LANSCE. The LANSCE User Office will assist you in preparing for your visit to LANSCE. Plan to arrive one day prior to your scheduled beam time. User Office Contacts Division Office Ph: 505.667.5051 Lujan Center User Office lujan-uo@lanl.gov WNR User Program Administrator Tanya Herrera Ph: 505.667.6797 User Office Main Desk Email: lansce-user-office@lanl.gov Ph:

  1. 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 Fesseha Mariam 505.667.3546 Christopher Morris 505.667.5652 Frans Trouw 505.665.7575 P-25 Group Leader Melynda Brooks 505.667.6909 P-25 Deputy Group Leader Frans Trouw 505.665.7575 pRad User Program pRad-uo@lanl.gov P-25 Subatomic Physics P-Division LANSCE pRad logo Los Alamos National Laboratory has used high energy protons as a

  2. 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.

  3. 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 Fesseha Mariam 505.667.3546 Christopher Morris 505.667.5652 Frans Trouw 505.665.7575 P-25 Group Leader Melynda Brooks 505.667.6909 P-25 Deputy Group Leader Frans Trouw 505.665.7575 pRad User Program pRad-uo@lanl.gov P-25 Subatomic Physics P-Division LANSCE pRad logo Los Alamos National Laboratory has used high energy protons as a

  4. 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.

  5. BISON Enhanced | Department of Energy

    Energy Savers [EERE]

    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

  6. 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.

  7. DOE - Office of Legacy Management -- Spencer Chemical Co - KS 0-01

    Office of Legacy Management (LM)

    KS 0-01 FUSRAP Considered Sites Site: SPENCER CHEMICAL CO. (KS.0-01 ) Eliminated from further consideration under FUSRAP - an AEC licensed operation Designated Name: Not Designated Alternate Name: Jayhawk Works KS.0-01-1 Location: Pittsburg , Kansas KS.0-01-1 Evaluation Year: 1985 KS.0-01-2 Site Operations: Processed enriched uranium (UF-6) and scrap to produce primarily uranium dioxide (UO-2) under AEC licenses. KS.0-01-3 KS.0-01-4 Site Disposition: Eliminated - No Authority - AEC licensed

  8. DOE - Office of Legacy Management -- Spencer Chemical Co - MO 0-01

    Office of Legacy Management (LM)

    MO 0-01 FUSRAP Considered Sites Site: SPENCER CHEMICAL CO. (MO.0-01) Eliminated from further consideration under FUSRAP - an AEC licensed operation Designated Name: Not Designated Alternate Name: Jayhawk Works MO.0-01-1 Location: Joplin , Missouri MO.0-01-1 Evaluation Year: 1985 MO.0-01-2 Site Operations: Processed enriched uranium (UF-6) and scrap to produce primarily uranium dioxide (UO-2) under AEC licenses. MO.0-01-3 MO.0-01-4 Site Disposition: Eliminated - No Authority MO.0-01-2 Radioactive

  9. TRXP REPORT M i3UREAU 08 HINES, RWO, N,sVADA ON OCTOBIGt 9,1962

    Office of Legacy Management (LM)

    TRXP REPORT M i3UREAU 08 HINES, RWO, N,sVADA ON OCTOBIGt 9,1962 3. Marehell W. E. Shaw OBJBcTIvE OFTRIP Infometion had been reoelved that the Bureau of Xinee Station was actively engaged.ln studying the eleotrowinnlng oi uranium from 602. It was dealred to eee the equipment used and deter- mine the statue of the projeot. SUMaaRp. CONCLUSIONS AND REC~ATIONS 1. A relatively simple cell that could be coaled up has bean designed for the suoceaefu~ pxyQction of aoaleeoed uraniua from UO2 feed at

  10. 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

  11. 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.

  12. 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.

  13. 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}.

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

    SciTech Connect (OSTI)

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

    2015-10-01

    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.

  15. 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.

  16. 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.

  17. 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

  18. Thermodynamics of formation of coffinite, USiO?

    SciTech Connect (OSTI)

    Guo, Xiaofeng; Szenknect, Stéphanie; Mesbah, Adel; Labs, Sabrina; Clavier, Nicolas; Poinssot, Christophe; Curtius, Hildegard; Bosbach, Dirk; Ewing, Rodney C.; Burns, Peter C.; Dacheux, Nicolas; Navrotsky, Alexandra

    2015-05-26

    Coffinite, USiO?, is an important U(IV) mineral, but its thermodynamic properties are not well-constrained. In this work, two different coffinite samples were synthesized under hydrothermal conditions and purified from a mixture of products. The enthalpy of formation was obtained by high temperature oxide melt solution calorimetry. Coffinite is energetically metastable with respect to a mixture of UO? (uraninite) and SiO? (quartz) by 25.6 ± 3.9 kJ/mol. Its standard enthalpy of formation from the elements at 25 °C is -1,970.0 ± 4.2 kJ/mol. Decomposition of the two samples was characterized by X-ray diffraction and by thermogravimetry and differential scanning calorimetry coupled with mass spectrometric analysis of evolved gases. Coffinite slowly decomposes to U?O? and SiO? starting around 450 °C in air and thus has poor thermal stability in the ambient environment. The energetic metastability explains why coffinite cannot be synthesized directly from uraninite and quartz but can be made by low temperature precipitation in aqueous and hydrothermal environments. These thermochemical constraints are in accord with observations of the occurrence of coffinite in nature and are relevant to spent nuclear fuel corrosion.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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%.

  7. 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.

  8. 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)

  9. 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.

  10. 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)

  11. 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.

  12. 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.

  13. 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 O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.

  14. 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)

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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.

  6. 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}.

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. 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.

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

    SciTech Connect (OSTI)

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

    2012-05-15

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

  13. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect (OSTI)

    Shott, Gregory

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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.

  1. 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).

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

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

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

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

  8. I

    Office of Legacy Management (LM)

    ?am-3 . ,' .*. . - yp: -.* : .- ., ._ ' Yi * <. ? :+". thfa prcbputir. 80,UUU lb. of tmmiuu, J.m,cDu lb. of 3wukdlw crper' tiwu 5.8 t&i8 l atr:irur ral u&d i.Wttd&?# Bir;n8 i;orammant end rUl rid nrtrlcial by uo&utboFlwd putqlm. ). The ~&&a, ' 8m ;altielJ 79 p-rmlt arrgora ted and ttw tap t.ha aikalini~, . L pokotlal brlf)r, bU88M 8-i .ii.i co# sat8 awtaet wltb the mBtmtl8a. aada q*iast fb a8v0-*..u @ow +.ta p-?Y h&al. . .; . ' 6 G.. ..*... . ,,z.. ,. ..*,::

  9. 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.

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

    DOE Patents [OSTI]

    Krot, Nikolai N. (Chelomiya St., 2, Apartment 15, Moscow, RU); Charushnikova, Iraida A. (Svoboda St., Bldg. 7A, Apartment 15, Moscow, RU)

    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.

  11. 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.

  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. 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)

  14. 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

  15. 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.

  16. 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.

  17. Kinetic studies of the [NpO? (CO?)?]?? ion at alkaline conditions using ¹³C 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 ¹³C 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?¹, Ea =45.1(±3.8) kJ mol?¹, ?H = 42.6(±3.8) kJ mol?¹, and ?S = -72(±13) J mol?¹ K?¹, 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?¹, is ~3 times faster than what we experimentally determined at pH 9.3: k298ex = 45.4(±5.3) s?¹. 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?¹. 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.

  18. 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

  19. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect (OSTI)

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

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  20. 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.

  1. 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.

  2. 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)

  3. 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.

  4. 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)

  5. 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.

  6. 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.

  7. 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.

  8. Development of LWR Fuels with Enhanced Accident Tolerance

    SciTech Connect (OSTI)

    Lahoda, Edward J.; Boylan, Frank A.

    2015-10-30

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U3Si2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO2 pellets. Pellets and powders of UO2, UN, and U3Si2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO2. Cold spray application of either the amorphous steel or the Ti2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.

  9. Irradiation Planning for Fully-Ceramic Micro-encsapsulated fuel in ATR at LWR-relevant conditions: year-end report on FY-2011

    SciTech Connect (OSTI)

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

    2011-09-01

    This report presents the estimation of required ATR irradiation levels for the DB-FCM fuel design (fueled with Pu and MAs). The fuel and assembly designs are those considered in a companion report [R. S. Sen et al., FCR&D-2011- 00037 or INL/EXT-11-23269]. These results, pertaining to the DB-FCM fuel, are definitive in as much as the design of said fuel is definitive. In addition to the work performed, as required, for DB-FCM fuel, work has started in a preliminary fashion on single-cell UO2 and UN fuels. These latter activities go beyond the original charter of this project and although the corresponding work is incomplete, significant progress has been achieved. However, in this context, all that has been achieved is only preliminary because the corresponding fuel designs are neither finalized nor optimized. In particular, the UO2 case is unlikely to result in a viable fuel design if limited to enrichment at or under 20 weight % in U-235. The UN fuel allows reasonable length cycles and is likely to make an optimal design possible. Despite being limited to preliminary designs and offering only preliminary conclusions, the irradiation planning tasks for UO2 and UN fuels that are summarized in this report are useful to the overall goal of devising and deploying FCM-LWR fuel since the methods acquired and tested in this project and the overall procedure for planning will be available for planning tests for the finalized fuel design. Indeed, once the fuel design is finalized and the expected burnup level is determined, the methodology that has been assembled will allow the prompt finalization of the neutronic planning of the irradiation experiment and would provide guidance on the expected experimental performance of the fuel. Deviations from the expected behavior will then have to be analyzed and the outcome of the analysis may be corrections or modifications for the assessment models as well as, possibly, fuel design modifications, and perhaps even variation of experimental control for future experimental phases. Besides the prediction of irradiation times, preliminary work was carried out on other aspects of irradiation planning. In particular, a method for evaluating the interplay of depletion, material performance modeling and irradiation is identified by reference to a companion report. Another area that was addressed in a preliminary fashion is the identification and selection of a strategy for the physical and mechanical design of the irradiation experiments. The principal conclusion is that the similarity between the FCM fuel and the fuel compacts of the Next Generation Nuclear Plant prismatic design are strong enough to warrant using irradiation hardware designs and instrumentation adapted from the AGR irradiation tests. Modifications, if found necessary, will probably be few and small, except as pertains to the water environment and its implications on the use of SiC cladding or SiC matrix with no additional cladding.

  10. Student Progress Report: Summer 2012

    SciTech Connect (OSTI)

    Tucker, Lucas P

    2012-08-06

    The Los Alamos SOURCES 4C code has been benchmarked for alpha particle beam problems and common neutron source materials (e.g. those containing plutonium or beryllium), but little benchmarking has been performed for more exotic isotopic neutron sources or uranium mixtures. This work extends SOURCES 4C benchmarking effort. Experimental data was found in the literature for several isotopic neutron sources, namely Am/Be, Am/F, Am/B, Cm/Be, {sup 238}Pu/{sup 13}C, {sup 252}Cf, and Am/Li. SOURCES 4C simulations were run for each of these materials and the output was used to develop a source term for use in MCNP, which allowed other physical effects such as down scattering and multiplication to be accounted for. Neutron emission rate and energy spectra results were compared for these sources, generally yielding order-of-magnitude agreement for the neutron emission rate and qualitative agreement for the shape of the neutron energy spectra. An exception was the neutron energy spectrum calculated for {sup 238}Pu/{sup 13}C whose primary peak was calculated to be 1 MeV higher than was measured. The accuracy of SOURCES is highly dependent on an accurate material definition. This discrepancy is likely due to inhomogeneity of the source materials, which cannot be simulated by SOURCES or MCNP, and chemical impurities not reported by the experimentalist. The results of the Am/Li calculation demonstrate that even small impurities are capable of dramatically changing the results. The neutron emission rates of numerous uranium compounds were also calculated with SOURCES and benchmarked with experimentally determined values found in the literature. The calculated results were similar to the experimental results with less than 10% error for the following compounds: uranyl fluoride, uranyl nitrate, UO{sub 3}, UO{sub 2}F{sub 2}, UF{sub 4}, UF{sub 6}, and U-metal of less than 90% enrichment. This work demonstrates the robustness of SOURCES as a tool for calculating neutron emission rates and energy spectra.

  11. TRIMOLECULAR REACTIONS OF URANIUM HEXAFLUORIDE WITH WATER

    SciTech Connect (OSTI)

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

    2010-02-25

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

  12. Uranium Fate in Hanford Sediment Altered by Simulated Acid Waste Solutions

    SciTech Connect (OSTI)

    Gartman, Brandy N.; Qafoku, Nikolla; Szecsody, James E.; Kukkadapu, Ravi K.; Wang, Zheming; Wellman, Dawn M.; Truex, Michael J.

    2015-07-24

    Many aspects of U(VI) behavior in sediments that are previously exposed to acidic waste fluids for sufficiently long times to induce significant changes in pH and other physical, mineralogical and chemical properties, are not well documented in the literature. For this reason, we conducted a series of macroscopic batch experiments combined with a variety of bulk characterization studies (Mössbauer and laser spectroscopy), micro-scale inspections (µ-XRF), and molecular scale interrogations (XANES) with the objectives to: i) determine the extent of U(VI) partitioning to Hanford sediments previously exposed to acidic waste simulants (pH = 2 and pH = 5) and under neutral conditions (pH = 8) at varying background solution concentrations (i.e., NaNO3); ii) determine micron-scale solid phase associated U valence state and phase identity; and iii) provide information for a plausible conceptual model of U(VI) attenuation under waste plume acidic conditions. The results of the batch experiments showed that the acid pre-treated sediment had high affinity for aqueous U(VI), which was removed from solution via two pH dependent and apparently different mechanisms (adsorption at pH = 2 and precipitation at pH = 5). The micro-scale inspections and XANES analyses confirmed that high concentration areas were rich mainly in U(VI), demonstrating that most of the added U(VI) was not reduced to U(IV). The laser spectroscopy data showed that uranyl phosphates {e.g. metaautunite [Ca(UO2)2(PO4)2•10-12H2O] and phosphuranylite [KCa(H3O)3(UO2)7(PO4)4O4•8(H2O)]} were present in the sediments. They also showed clear differences between the U bearing phases in the experiments conducted in the presence or absence of air. The data generated from these experiments will help in a better understanding of the reactions and processes that have a significant effect and/or control U mobility.

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

    SciTech Connect (OSTI)

    Icenhour, A.S.

    2003-09-10

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

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

    SciTech Connect (OSTI)

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

    2010-01-29

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

  15. Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

    SciTech Connect (OSTI)

    Marshall, Margaret A.; Bess, John D.

    2015-02-01

    The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965 have been evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study “power plants for the production of electrical power in space vehicles.” The third configuration in this series of experiments was a beryllium-reflected assembly of stainless-steel-clad, highly enriched uranium (HEU)-O2 fuel mockup of a potassium-cooled space power reactor. Reactivity measurements cadmium ratio spectral measurements and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The fuel tube effect and neutron moderating and absorbing material worth measurements are the focus of this paper. Additionally, a measurement of the worth of potassium filling the core region was performed but has not yet been evaluated Pellets of 93.15 wt.% enriched uranium dioxide (UO2) were stacked in 30.48 cm tall stainless steel fuel tubes (0.3 cm tall end caps). Each fuel tube had 26 pellets with a total mass of 295.8 g UO2 per tube. 253 tubes were arranged in 1.506-cm triangular lattice. An additional 7-tube cluster critical configuration was also measured but not used for any physics measurements. The core was surrounded on all side by a beryllium reflector. The fuel effect worths were measured by removing fuel tubes at various radius. An accident scenario was also simulated by moving outward twenty fuel rods from the periphery of the core so they were touching the core tank. The change in the system reactivity when the fuel tube(s) were removed/moved compared with the base configuration was the worth of the fuel tubes or accident scenario. The worth of neutron absorbing and moderating materials was measured by inserting material rods into the core at regular intervals or placing lids at the top of the core tank. Stainless steel 347, tungsten, niobium, polyethylene, graphite, boron carbide, aluminum and cadmium rods and/or lid worths were all measured. The change in the system reactivity when a material was inserted into the core is the worth of the material.

  16. Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, January-March 1979

    SciTech Connect (OSTI)

    Steindler, M J; Ader, M; Barletta, R E

    1980-01-01

    In the program on pyrochemical and dry processing methods (PDPM) for nuclear fuel, corrosion testing of refractory metals and alloys, graphite, and SiC in PDPM environments was done. A tungsten-metallized Al/sub 2/O/sub 3/-3% Y/sub 2/O/sub 3/ crucible was successfully fabricated. Tungsten microstructure of a plasma-sprayed tungsten crucible was stabilized by nickel infiltration and heat treatment. Solubility measurements of Th in Cd and Cd-Mg alloys were continued, as were experiments to study the reduction of high-fired ThO/sub 2/. Work on the fused salt electrolysis of CaO also was continued. The method of coprocessing of U and Pu by a salt transport process was modified. Tungsten-coated molybdenum crucibles were fabricated. The proliferation resistance of chloride volatility processing of thorium-based fuels is being evaluated by studying the behavior of fission product elements during chlorination of U and Th. Thermodynamic analysis of the phase relationships in the U-Pu-Zn system was initiated. The Pyro-Civex reprocessing method is being reviewed. Reactivity of UO/sub 2/ and PuO/sub 2/ with molten equimolar NaNO/sub 3/-KNO/sub 3/ is being studied along with the behavior of selected fission product elements. Work was continued on the reprocessing of actinide oxides by extracting the actinides from a bismuth solution. Rate of dissolution of UO/sub 2/ microspheres in LiCl/AlCl/sub 3/ was measured. Nitriding rates of Th and U dissolved in molten tin were measured. In work on the encapsulation of radioactive waste in metal, leach rates of a simulated waste glass were studied. Rates of dissolution of metals (potential barrier materials) in aqueous media are being studied. In work on the transport properties of nuclear waste in geologic media, the adsorption of iodate by hematite as a function of pH and iodate concentration was measured. The migration behavior of cesium in limestone was studied in relation to the cesium concentration and pH of simulated groundwater solutions.

  17. Atomistic Simulations of Mass and Thermal Transport in Oxide Nuclear Fuels

    SciTech Connect (OSTI)

    Andersson, Anders D.; Uberuaga, Blas P.; Du, Shiyu; Liu, Xiang-Yang; Nerikar, Pankaj; Stanek, Christopher R.; Tonks, Michael; Millet, Paul; Biner, Bulent

    2012-06-04

    In this talk we discuss simulations of the mass and thermal transport in oxide nuclear fuels. Redistribution of fission gases such as Xe is closely coupled to nuclear fuel performance. Most fission gases have low solubility in the fuel matrix, specifically the insolubility is most pronounced for large fission gas atoms such as Xe, and as a result there is a significant driving force for segregation of gas atoms to grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. The first step of the fission gas redistribution is diffusion of individual gas atoms through the fuel matrix to existing sinks, which is governed by the activation energy for bulk diffusion. Fission gas bubbles are then formed by either separate nucleation events or by filling voids that were nucleated at a prior stage; in both cases their formation and latter growth is coupled to vacancy dynamics and thus linked to the production of vacancies via irradiation or thermal events. In order to better understand bulk Xe behavior (diffusion mechanisms) in UO{sub 2{+-}x} we first calculate the relevant activation energies using density functional theory (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, though other alternatives may exist in high irradiation fields. 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 a continuum transport model for Xe and U is formulated based on the diffusion mechanisms established from DFT. After combining this model with descriptions of the interaction between Xe and grain boundaries derived from separate atomistic calculations, we simulate Xe redistribution for a few simple microstructures using finite element methods (FEM), as implemented in the MOOSE framework from Idaho National Laboratory. Thermal transport together with the power distribution determines the temperature distribution in the fuel rod and it is thus one of the most influential properties on nuclear fuel performance. The fuel thermal conductivity changes as function of time due to microstructure evolution (e.g. fission gas redistribution) and compositional changes. Using molecular dynamics simulations we have studied the impact of different types of grain boundaries and fission gas bubbles on UO{sub 2} thermal conductivity.

  18. Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

    SciTech Connect (OSTI)

    Shwageraus, Eugene; Zhao Xianfeng; Driscoll, Michael J.; Hejzlar, Pavel; Kazimi, Mujid S.; Herring, J. Stephen

    2004-07-15

    A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with {sup 235}U is necessary, and the {sup 235}U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO{sub 2}-UO{sub 2}) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of {sup 233}U from the {sup 232}Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the {sup 233}U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial {sup 235}U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO{sub 2} fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.

  19. Updated NGNP Fuel Acquisition Strategy

    SciTech Connect (OSTI)

    David Petti; Tim Abram; Richard Hobbins; Jim Kendall

    2010-12-01

    A Next Generation Nuclear Plant (NGNP) fuel acquisition strategy was first established in 2007. In that report, a detailed technical assessment of potential fuel vendors for the first core of NGNP was conducted by an independent group of international experts based on input from the three major reactor vendor teams. Part of the assessment included an evaluation of the credibility of each option, along with a cost and schedule to implement each strategy compared with the schedule and throughput needs of the NGNP project. While credible options were identified based on the conditions in place at the time, many changes in the assumptions underlying the strategy and in externalities that have occurred in the interim requiring that the options be re-evaluated. This document presents an update to that strategy based on current capabilities for fuel fabrication as well as fuel performance and qualification testing worldwide. In light of the recent Pebble Bed Modular Reactor (PBMR) project closure, the Advanced Gas Reactor (AGR) fuel development and qualification program needs to support both pebble and prismatic options under the NGNP project. A number of assumptions were established that formed a context for the evaluation. Of these, the most important are: • Based on logistics associated with the on-going engineering design activities, vendor teams would start preliminary design in October 2012 and complete in May 2014. A decision on reactor type will be made following preliminary design, with the decision process assumed to be completed in January 2015. Thus, no fuel decision (pebble or prismatic) will be made in the near term. • Activities necessary for both pebble and prismatic fuel qualification will be conducted in parallel until a fuel form selection is made. As such, process development, fuel fabrication, irradiation, and testing for pebble and prismatic options should not negatively influence each other during the period prior to a decision on reactor type. • Additional funding will be made available beginning in fiscal year (FY) 2012 to support pebble bed fuel fabrication process development and fuel testing while maintaining the prismatic fuel schedule. Options for fuel fabrication for prismatic and pebble bed were evaluated based on the credibility of each option, along with a cost and schedule to implement each strategy. The sole prismatic option is Babcock and Wilcox (B&W) producing uranium oxycarbide (UCO) tristructural-isotropic (TRISO) fuel particles in compacts. This option finishes in the middle of 2022 . Options for the pebble bed are Nuclear Fuel Industries (NFI) in Japan producing uranium dioxide (UO2) TRISO fuel particles, and/or B&W producing UCO or UO2 TRISO fuel particles. All pebble options finish in mid to late 2022.

  20. Deep Burn Fuel Cycle Integration: Evaluation of Two-Tier Scenarios

    SciTech Connect (OSTI)

    S. Bays; H. Zhang; M. Pope

    2009-05-01

    The use of a deep burn strategy using VHTRs (or DB-MHR), as a means of burning transuranics produced by LWRs, was compared to performing this task with LWR MOX. The spent DB-MHR fuel was recycled for ultimate final recycle in fast reactors (ARRs). This report summarizes the preliminary findings of the support ratio (in terms of MWth installed) between LWRs, DB-MHRs and ARRs in an equilibrium “two-tier” fuel cycle scenario. Values from literature were used to represent the LWR and DB-MHR isotopic compositions. A reactor physics simulation of the ARR was analyzed to determine the effect that the DB-MHR spent fuel cooling time on the ARR transuranic consumption rate. These results suggest that the cooling time has some but not a significant impact on the ARRs conversion ratio and transuranic consumption rate. This is attributed to fissile worth being derived from non-fissile or “threshold-fissioning” isotopes in the ARR’s fast spectrum. The fraction of installed thermal capacity of each reactor in the DB-MHR 2-tier fuel cycle was compared with that of an equivalent MOX 2-tier fuel cycle, assuming fuel supply and demand are in equilibrium. The use of DB-MHRs in the 1st-tier allows for a 10% increase in the fraction of fleet installed capacity of UO2-fueled LWRs compared to using a MOX 1st-tier. Also, it was found that because the DB-MHR derives more power per unit mass of transuranics charged to the fresh fuel, the “front-end” reprocessing demand is less than MOX. Therefore, more fleet installed capacity of DB-MHR would be required to support a given fleet of UO2 LWRs than would be required of MOX plants. However, the transuranic deep burn achieved by DB-MHRs reduces the number of fast reactors in the 2nd-tier to support the DB-MHRs “back-end” transuranic output than if MOX plants were used. Further analysis of the relative costs of these various types of reactors is required before a comparative study of these options could be considered complete.

  1. Unexpected Actinyl Cation-Directed Structural Variation in Neptunyl(VI) A-Type Tri-lacunary Heteropolyoxotungstate Complexes

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

    Berg, John M.; Gaunt, Andrew J.; May, Iain; Pugmire, Alison L.; Reilly, Sean D.; Scott, Brian L.; Wilkerson, Marianne P.

    2015-04-22

    A-type tri-lacunary heteropolyoxotungstate anions (e.g., [PW9O34]9-, [AsW9O34]9-, [SiW9O34]10- and [GeW9O34]10-) are multi-dentate oxygen donor ligands that readily form sandwich complexes with actinyl cations ({UO2}2+, {NpO2}+, {NpO2}2+ & {PuO2}2+) in near neutral/slightly alkaline aqueous solutions. Two or three actinyl cations are sandwiched between two trilacunary anions, with additional cations (Na+, K+ or NH4 +) also often held within the cluster. Studies thus far have indicated that it is these additional +I cations, rather than the specific actinyl cation, that direct the structural variation in the complexes formed. We now report the structural characterization of the neptunyl (VI) cluster complex (NH4)13 [Na(NpO2)2(A-α-more » PW9O34)2]·12H2O. The anion in this complex, [Na(NpO2)2(PW9O34)2]13-, contains one Na+ cation and two {NpO2}2+ cations held between two [PW9O34]9- anions – with an additional partial occupancy NH4 + or {NpO2}2+ cation also present. In the analogous uranium (VI) system, under similar reaction conditions that includes an excess of NH4Cl in the parent solution, it was previously shown that [(NH4)2(UVIO2)2(A-PW9O34)2]12- is the dominant species in both solution and the crystallized salt. Spectroscopic studies provide further proof of differences in the observed chemistry for the {NpO2}2+/[PW9O34]9- and {UO2}2+/[PW9O34]9- systems, both in solution and in solid state complexes crystallized from comparable salt solutions. The work revealed that varying the actinide element (Np vs. U) can indeed measurably impact structure and complex stability in the cluster chemistry of actinyl (VI) cations with A-type tri-lacunary heteropolyoxotungstate anions.« less

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

    SciTech Connect (OSTI)

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

    2010-03-15

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

  3. Stress Analysis of Coated Particle Fuel in the Deep-Burn Pebble Bed Reactor Design

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-05-01

    High fuel temperatures and resulting fuel particle coating stresses can be expected in a Pu and minor actinide fueled Pebble Bed Modular Reactor (400 MWth) design as compared to the ’standard’ UO2 fueled core. The high discharge burnup aimed for in this Deep-Burn design results in increased power and temperature peaking in the pebble bed near the inner and outer reflector. Furthermore, the pebble power in a multi-pass in-core pebble recycling scheme is relatively high for pebbles that make their first core pass. This might result in an increase of the mechanical failure of the coatings, which serve as the containment of radioactive fission products in the PBMR design. To investigate the integrity of the particle fuel coatings as a function of the irradiation time (i.e. burnup), core position and during a Loss Of Forced Cooling (LOFC) incident the PArticle STress Analysis code (PASTA) has been coupled to the PEBBED code for neutronics, thermal-hydraulics and depletion analysis of the core. Two deep burn fuel types (Pu with or without initial MA fuel content) have been investigated with the new code system for normal and transient conditions including the effect of the statistical variation of thickness of the coating layers.

  4. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  5. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    SciTech Connect (OSTI)

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  6. A Combined Neutronic-Thermal Hydraulic Model of CERMET NTR Reactor

    SciTech Connect (OSTI)

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

    2011-02-01

    Abstract. Two different CERMET fueled Nuclear Thermal Propulsion reactors were modeled to determine the optimum coolant channel surface area to volume ratio required to cool a 25,000 lbf rocket engine operating at a specific impulse of 940 seconds. Both reactor concepts were computationally fueled with hexagonal cross section fuel elements having a flat-to-flat distance of 3.51 cm and containing 60 vol.% UO2 enriched to 93wt.%U235 and 40 vol.% tungsten. Coolant channel configuration consisted of a 37 coolant channel fuel element and a 61 coolant channel model representing 0.3 and 0.6 surface area to volume ratios respectively. The energy deposition from decelerating fission products and scattered neutrons and photons was determined using the MCNP monte carlo code and then imported into the STAR-CCM+ computational fluid dynamics code. The 37 coolant channel case was shown to be insufficient in cooling the core to a peak temperature of 3000 K; however, the 61 coolant channel model shows promise for maintaining a peak core temperature of 3000 K, with no more refinements to the surface area to volume ratio. The core was modeled to have a power density of 9.34 GW/m3 with a thrust to weight ratio of 5.7.

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

    SciTech Connect (OSTI)

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

    1991-03-01

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

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

    SciTech Connect (OSTI)

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

    2014-07-01

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

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

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

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

    2014-04-15

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

  10. Radiolysis Process Model

    SciTech Connect (OSTI)

    Buck, Edgar C.; Wittman, Richard S.; Skomurski, Frances N.; Cantrell, Kirk J.; McNamara, Bruce K.; Soderquist, Chuck Z.

    2012-07-17

    Assessing the performance of spent (used) nuclear fuel 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 (including OH• and H• radicals, O2-, eaq, H2O2, H2, and O2) that may increase the waste form degradation rate and change radionuclide behavior. H2O2 is the dominant oxidant for spent nuclear fuel in an O2 depleted water environment, the most sensitive parameters have been identified with respect to predictions of a radiolysis model under typical conditions. As compared with the full model with about 100 reactions it was found that only 30-40 of the reactions are required to determine [H2O2] to one part in 10–5 and to preserve most of the predictions for major species. This allows a systematic approach for model simplification and offers guidance in designing experiments for validation.

  11. The radiation induced chemistry of uranyl cation in aqueous carbonate –bicarbonate solutions as followed by NMR spectroscopy

    SciTech Connect (OSTI)

    McNamara, Bruce K.; Snow, Lanee A.; Soderquist, Chuck Z.; Sinkov, Sergei I.; Cho, Herman M.; Friese, Judah I.

    2006-05-01

    Alpha radiation induced formation of hydrogen peroxide in carbonate ?bicarbonate media was followed by 13C NMR using dissolved [233UO2(13CO3)3]4- as the alpha source (Dalpha= 12.1 Gy/hr). Between the pH region between 5.9 and 11.6 hydrogen peroxide causes a varied speciation of the uranyl carbonates that is a function of the uranium, carbonate and the hydrogen peroxide concentrations. It is shown that the speciation of the peroxy carbonates (or other species) formed in solution by titration with hydrogen peroxide are common to those formed by hydrogen peroxide generated by radiolysis. The radiolysis experiment was carried out above pH = 9.96 to minimize the loss of 13CO2 over a 2800 hr period. Radiolytic generation of hydrogen peroxide was followed by formation of a uranyl peroxy carbonate complex and complex formation accelerated for about 1200 hours. Complex formation was observed to terminate at a concentration between 1x10-4 and 5x10-4 M. It is assumed that either a steady state H2O2 production rate was established in solution or that some limiting feature of the experiment was responsible for slowing the yield of product.

  12. The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

    SciTech Connect (OSTI)

    Mitenkova, E. F.; Novikov, N. V.; Blokhin, A. I.

    2012-07-01

    The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

  13. Severe Accident Test Station Activity Report

    SciTech Connect (OSTI)

    Pint, Bruce A.; Terrani, Kurt A.

    2015-06-01

    Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

  14. Use of ion conductors in the pyrochemical reduction of oxides

    DOE Patents [OSTI]

    Miller, W.E.; Tomczuk, Z.

    1994-02-01

    An electrochemical process and electrochemical cell for reducing a metal oxide are provided. First the oxide is separated as oxygen gas using, for example, a ZrO[sub 2] oxygen ion conductor anode and the metal ions from the reduction salt are reduced and deposited on an ion conductor cathode, for example, sodium ion reduced on a [beta]-alumina sodium ion conductor cathode. The generation of and separation of oxygen gas avoids the problem with chemical back reaction of oxygen with active metals in the cell. The method also is characterized by a sequence of two steps where an inert cathode electrode is inserted into the electrochemical cell in the second step and the metallic component in the ion conductor is then used as the anode to cause electrochemical reduction of the metal ions formed in the first step from the metal oxide where oxygen gas formed at the anode. The use of ion conductors serves to isolate the active components from chemically reacting with certain chemicals in the cell. While applicable to a variety of metal oxides, the invention has special importance for reducing CaO to Ca[sup o] used for reducing UO[sub 2] and PuO[sub 2] to U and Pu. 2 figures.

  15. AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

    SciTech Connect (OSTI)

    Michael L. Abbott; Keith A. Daum

    2011-08-01

    This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).

  16. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R.

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  17. Reference reactor module for NASA's lunar surface fission power system

    SciTech Connect (OSTI)

    Poston, David I; Kapernick, Richard J; Dixon, David D; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  18. The relative variational model: A topological view of matter and its properties: Thermal expansion

    SciTech Connect (OSTI)

    Dias, M. S.; De Vasconcelos, V.; Mattos, J. R. L.; Jordao, E.

    2012-07-01

    Formal definitions of convergence, connected-ness and continuity were established to characterize and describe the crystalline solid and its properties as a unified notion in the topological space. The crystalline solid is a previously empty space that has been filled with atoms and phonons, i.e., the crystal is built with packages of matter and energy in a regular and orderly repetitive pattern along three orthogonal dimensions of the space. The spatial occupation of the atom in the crystal structure is determined by its mean vibrational volume. Thus, the changes of volume and the changes of internal energy are intrinsically linked. In fact, physical and material properties are the interdependent and bijective quantifications associated with variations of the internal energy. These properties are modeled by means of an intrinsic and invariable form function: the Relative Variational Model. In this paper, the experimental data of the thermal expansion for the oxides Al{sub 2}O{sub 3} and UO{sub 2} were analytically depicted by means of this model in the temperature range of 0 K up to the melting point. (authors)

  19. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect (OSTI)

    Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  20. Advanced Triso fuels with zirconium carbide for high temperature reactors

    SciTech Connect (OSTI)

    Lobach, Sergiy Y.; Knight, Travis W.; Jacob, Norman P.; Athon, Clifton E.

    2007-07-01

    There are several options for the advanced TRISO fuel: one is primarily replacement SiC with ZrC and the other is a concept involving a thin ZrC layer coating on the kernel, which is then enclosed in usual TRISO coatings. An effort at modeling, fabrication and testing of an advanced TRISO coated UO{sub 2} fuel particle design incorporating an added layer of ZrC over the fuel kernel is under investigation. The objectives of the coated particle development program are to define the essentials of a production route for the manufacture of kernels and coated particles and to identify the important process parameters that determine the particle properties. Still, the integrity of the ZrC coating is important, but not the main goal. The primary purpose of a ZrC coating examination in this study is to determine hot it serves as an oxygen getter to limit CO production and hence pressure buildup that would stress coatings leading to failure. This additional ZrC coating also aids in retaining fission products within the kernel, and carbon diffusion in the particle is limited hence kernel migration rates are slowed. The combined result being that failure rates of coated particles should decrease. (authors)

  1. Mass balance and separation factor of actinides through series process test on pyro-process

    SciTech Connect (OSTI)

    Kitawaki, Shinichi; Fukushima, Mineo; Yahagi, Noboru; Kurata, Masaki

    2007-07-01

    An electrolysis test in a sequential condition was performed using U-Pu alloy and liquid Cd as an anode and a cathode material, respectively, in which the anode and cathode was changed three times. Mass balance of U and Pu after three-time electrolysis was determined from both the chemical analysis of salt and anode residue, and the weight increase in the cathode. Approximately 95% of U and 100% of Pu was detected even after three times electrolysis. The loss of U is more significant than that of Pu. The mass balance was also evaluated for the intermediate steps of the sequence. More than 90% of U and Pu with respect to the initial amount were constantly caught up with in each step. The separation factor of U/Pu in three cathodes were varied 1.62-2.06. The chemical form of the anode residue was UO, PuOCl and PuO{sub 2}, which can be converted to the chloride by a reaction with ZrCl{sub 4}. (authors)

  2. Predicting Stability Constants for Uranyl Complexes Using Density Functional Theory

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

    Vukovic, Sinisa; Hay, Benjamin P.; Bryantsev, Vyacheslav S.

    2015-04-02

    The ability to predict the equilibrium constants for the formation of 1:1 uranyl:ligand complexes (log K1 values) provides the essential foundation for the rational design of ligands with enhanced uranyl affinity and selectivity. We also use density functional theory (B3LYP) and the IEFPCM continuum solvation model to compute aqueous stability constants for UO22+ complexes with 18 donor ligands. Theoretical calculations permit reasonably good estimates of relative binding strengths, while the absolute log K1 values are significantly overestimated. Accurate predictions of the absolute log K1 values (root mean square deviation from experiment < 1.0 for log K1 values ranging from 0more » to 16.8) can be obtained by fitting the experimental data for two groups of mono and divalent negative oxygen donor ligands. The utility of correlations is demonstrated for amidoxime and imide dioxime ligands, providing a useful means of screening for new ligands with strong chelate capability to uranyl.« less

  3. TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

    SciTech Connect (OSTI)

    Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

    2014-04-01

    As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

  4. Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

    SciTech Connect (OSTI)

    Jason Hales; Various

    2014-06-01

    The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

  5. Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi nuclear power station

    SciTech Connect (OSTI)

    Ikeuchi, Hirotomo; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro; Kondo, Yoshikazu; Noguchi, Yoshikazu

    2013-07-01

    For the decommissioning of the Fukushima-Daiichi Nuclear Power Station (1F), the characterization of fuel-debris in cores of Units 1-3 is necessary. In this study, typical phases of the in-vessel fuel-debris were estimated using a thermodynamic equilibrium (TDE) calculation. The FactSage program and NUCLEA database were applied to estimate the phase equilibria of debris. It was confirmed that the TDE calculation using the database can reproduce the phase separation behavior of debris observed in the Three Mile Island accident. In the TDE calculation of 1F, the oxygen potential [G(O{sub 2})] was assumed to be a variable. At low G(O{sub 2}) where metallic zirconium remains, (U,Zr)O{sub 2}, UO{sub 2}, and ZrO{sub 2} were found as oxides, and oxygen-dispersed Zr, Fe{sub 2}(Zr,U), and Fe{sub 3}UZr{sub 2} were found as metals. With an increase in zirconium oxidation, the mass of those metals, especially Fe{sub 3}UZr{sub 2}, decreased, but the other phases of metals hardly changed qualitatively. Consequently, (U,Zr)O{sub 2} is suggested as a typical phase of oxide, and Fe{sub 2}(Zr,U) is suggested as that of metal. However, a more detailed estimation is necessary to consider the distribution of Fe in the reactor pressure vessel through core-melt progression. (authors)

  6. Westinghouse Hanford Company FY 1995 Materials Management Plan (MMP)

    SciTech Connect (OSTI)

    Higginson, M.C.

    1994-10-01

    The safe and sound operation of facilities and storage of nuclear material are top priorities within Hanford`s environmental management, site restoration mission. The projected materials estimates, based on the Materials Management Plan (MMP) assumptions outlined below, were prepared for Department of Energy (DOE) use in long-range planning. The Hanford MMP covers the period FY 1995 through FY 2005, as directed by DOE. All DOE Richland Operations (RL) Office facilities are essentially funded by the Office of Transition and Facilities Management, Environmental Restoration and Waste Management (EM). These facilities include PUREX, the UO{sub 3} plant, N-Reactor, T-Plant, K-Basins, FFTF, PFP and the 300 Area Fuel Fabrication facilities. Currently DP provides partial funding for the latter two facilities. Beginning in FY 1996 (in accordance with DOE-HQ MMP assumptions), EM will fund expenses related to the storage, monitoring, and safeguarding of all Special Nuclear Material (SNM) in the PFP. Ownership and costs related to movement and/or stabilization of that material will belong to EM programs (excluding NE material). It is also assumed that IAEA will take over inventory validation and surveillance of EM owned SNM at this time (FY 1996).

  7. Fuel elements of research reactor CM

    SciTech Connect (OSTI)

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  8. Supercritical Fluid Extraction of Radionuclides - A Green Technology for Nuclear Waste Management

    SciTech Connect (OSTI)

    Wai, Chien M.

    2003-09-10

    Supercritical fluid carbon dioxide (SF-CO2) is capable of extracting radionuclides including cesium, strontium, uranium, plutonium and lanthanides directly from liquid and solid samples with proper complexing agents. Of particular interest is the ability of SF-CO2 to dissolve uranium dioxide directly using a CO2-soluble tri-nbutylphosphate- nitric acid (TBP-HNO3) extractant to form a highly soluble UO2(NO3)2(TBP)2 complex that can be transported and separated from Cs, Sr, and other transition metals. This method can also dissolve plutonium dioxide in SF-CO2. The SF-CO2 extraction technology offers several advantages over conventional solvent-based methods including ability to extract radionuclides directly from solids, easy separation of solutes from CO2, and minimization of liquid waste generation. Potential applications of the SF-CO2 extraction technology for nuclear waste treatment and for reprocessing of spent nuclear fuels will be discussed. Information on current demonstrations of the SF-CO2 technology by nuclear companies and research organizations in different countries will be reviewed.

  9. Conceptual designs of NDA instruments for the NRTA system at the Rokkasho Reprocessing Plant

    SciTech Connect (OSTI)

    Li, T.K.; Klosterbuer, S.F.; Menlove, H.O.

    1996-09-01

    The authors are studying conceptual designs of selected nondestructive assay (NDA) instruments for the near-real-time accounting system at the rokkasho Reprocessing Plant (RRP) of Japan Nuclear Fuel Limited (JNFL). The JNFL RRP is a large-scale commercial reprocessing facility for spent fuel from boiling-water and pressurized-water reactors. The facility comprises two major components: the main process area to separate and produce purified plutonium nitrate and uranyl nitrate from irradiated reactor spent fuels, and the co-denitration process area to combine and convert the plutonium nitrate and uranyl nitrate into mixed oxide (MOX). The selected NDA instruments for conceptual design studies are the MOX-product canister counter, holdup measurement systems for calcination and reduction furnaces and for blenders in the co-denitration process, the isotope dilution gamma-ray spectrometer for the spent fuel dissolver solution, and unattended verification systems. For more effective and practical safeguards and material control and accounting at RRP, the authors are also studying the conceptual design for the UO{sub 3} large-barrel counter. This paper discusses the state-of-the-art NDA conceptual design and research and development activities for the above instruments.

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

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

    Uberuaga, Blas Pedro; Andersson, David A.

    2015-10-01

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

  11. Environmental Controls on the Activity of Aquifer Microbial Communities in the 300 Area of the Hanford Site

    SciTech Connect (OSTI)

    Konopka, Allan; Plymale, Andrew E.; Carvajal, Denny A.; Lin, Xueju; McKinley, James P.

    2013-11-06

    Aquifer microbes in the 300 Area of the Hanford Site in southeastern Washington State, USA are periodically exposed to U(VI) concentrations that can range up to 10 ?M in small sediment fractures. Assays of 35 H-leucine incorporation indicated that both sediment-associated and planktonic microbes were metabolically active, and that organic C was growth-limiting in the sediments. Although bacteria suspended in native groundwater retained high activity when exposed to 100 ?M U(VI), they were inhibited by U(VI) < 1 ?M in synthetic groundwater that lacked added bicarbonate. Chemical speciation modeling suggested that positively-charged species and particularly (UO2)3(OH)5+ rose in concentration as more U(VI) was added to synthetic groundwater, but that carbonate complexes dominated U(VI) speciation in natural groundwater. U toxicity was relieved when increasing amounts of bicarbonate were added to synthetic groundwater containing 4.5 ?M U(VI). Pertechnetate, an oxyanion that is another contaminant of concern at the Hanford Site, was not toxic to groundwater microbes at concentrations up to 125 ?M.

  12. Structural model of uramarsite

    SciTech Connect (OSTI)

    Rastsvetaeva, R. K.; Sidorenko, G. A.; Ivanova, A. G.; Chukanov, N. V.

    2008-09-15

    The structural model of uramarsite, a new mineral of the uran-mica family from the Bota-Burum deposit (South Kazakhstan), is determined using a single-crystal X-ray diffraction analysis. The parameters of the triclinic unit cell are as follows: a = 7.173(2) A, b = 7.167(5) A, c = 9.30(1) A, {alpha} = 90.13(7){sup o}, {beta} = 90.09(4){sup o}, {gamma} = 89.96(4){sup o}, and space group P1. The crystal chemical formula of uramarsite is: (UO{sub 2}){sub 2}[AsO{sub 4}][PO{sub 4},AsO{sub 4}][NH{sub 4}][H{sub 3}O] . 6H{sub 2}O (Z = 1). Uramarsite is the second ammonium-containing mineral of uranium and an arsenate analogue of uramphite. In the case of uramarsite, the lowering of the symmetry from tetragonal to triclinic, which is accompanied by a triclinic distortion of the tetragonal unit cell, is apparently caused by the ordering of the As and P atoms and the NH{sub 4}, H{sub 3}O, and H{sub 2}O groups.

  13. Investigation of biologically-designed metal-specific chelators for potential metal recovery and waste remediation applications.

    SciTech Connect (OSTI)

    Criscenti, Louise Jacqueline; Ockwig, Nathan W.

    2009-01-01

    Bacteria, algae and plants produce metal-specific chelators to capture required nutrient or toxic trace metals. Biological systems are thought to be very efficient, honed by evolutionary forces over time. Understanding the approaches used by living organisms to select for specific metals in the environment may lead to design of cheaper and more effective approaches for metal recovery and contaminant-metal remediation. In this study, the binding of a common siderophore, desferrioxamine B (DFO-B), to three aqueous metal cations, Fe(II), Fe(III), and UO{sub 2}(VI) was investigated using classical molecular dynamics. DFO-B has three acetohydroxamate groups and a terminal amine group that all deprotonate with increasing pH. For all three metals, complexes with DFO-B (-2) are the most stable and favored under alkaline conditions. Under more acidic conditions, the metal-DFO complexes involve chelation with both acetohydroxamate and acetylamine groups. The approach taken here allows for detailed investigation of metal binding to biologically-designed organic ligands.

  14. Promises and Challenges of Thorium Implementation for Transuranic Transmutation - 13550

    SciTech Connect (OSTI)

    Franceschini, F.; Lahoda, E.; Wenner, M.; Lindley, B.; Fiorina, C.; Phillips, C.

    2013-07-01

    This paper focuses on the challenges of implementing a thorium fuel cycle for recycle and transmutation of long-lived actinide components from used nuclear fuel. A multi-stage reactor system is proposed; the first stage consists of current UO{sub 2} once-through LWRs supplying transuranic isotopes that are continuously recycled and burned in second stage reactors in either a uranium (U) or thorium (Th) carrier. The second stage reactors considered for the analysis are Reduced Moderation Pressurized Water Reactors (RMPWRs), reconfigured from current PWR core designs, and Fast Reactors (FRs) with a burner core design. While both RMPWRs and FRs can in principle be employed, each reactor and associated technology has pros and cons. FRs have unmatched flexibility and transmutation efficiency. RMPWRs have higher fuel manufacturing and reprocessing requirements, but may represent a cheaper solution and the opportunity for a shorter time to licensing and deployment. All options require substantial developments in manufacturing, due to the high radiation field, and reprocessing, due to the very high actinide recovery ratio to elicit the claimed radiotoxicity reduction. Th reduces the number of transmutation reactors, and is required to enable a viable RMPWR design, but presents additional challenges on manufacturing and reprocessing. The tradeoff between the various options does not make the choice obvious. Moreover, without an overarching supporting policy in place, the costly and challenging technologies required inherently discourage industrialization of any transmutation scheme, regardless of the adoption of U or Th. (authors)

  15. Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

    SciTech Connect (OSTI)

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.

    2013-07-01

    This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

  16. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    SciTech Connect (OSTI)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

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

    DOE Patents [OSTI]

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

    2001-01-01

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

  18. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    SciTech Connect (OSTI)

    G. Pastore; L.P. Swiler; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; L. Luzzi; P. Van Uffelen; R.L. Williamson

    2014-10-01

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  19. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

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

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; Novascone, Stephen R.; Perez, Danielle M.; Spencer, Benjamin W.; Luzzi, Lelio; Uffelen, Paul Van; Williamson, Richard L.

    2014-10-12

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertaintymore » in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.« less

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

    SciTech Connect (OSTI)

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

    1980-06-01

    The effects of the thermally-induced cracking and subsequent relocation of UO2 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 primary assumptions of the new model are that (1) the cracked fuel is in a hydrostatic state of stress in the (r,theta) plane, and that (2) there is no axial slipping between fuel and cladding. Three basic parameters are used to describe the cracked fuel: (1) the crack pattern, (2) the crack roughness, and (3) the fuel surface (gap) roughness. Recommendations are made on refining the model.

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

    SciTech Connect (OSTI)

    Margaret A. Marshall; John D. Bess

    2009-11-01

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

  2. U-234/U-238 ratio: Qualitative estimate of groundwater flow in Rocky Flats monitoring wells

    SciTech Connect (OSTI)

    Laul, J.C.

    1994-02-01

    Groundwater movement through various pathways is the primary mechanism for the transport of radionuclides and trace elements in a water/rock interaction. About three dozen wells, installed in the Rocky Flats Plant (RFP) Solar Evaporation Ponds (SEP) area, are monitored quarterly to evaluate the extent of any lateral and downgradient migration of contaminants from the Solar Evaporation Ponds: 207-A; 207-B North, 207-B Center, and 207-B South; and 207-C. The Solar Ponds are the main source for the various contaminants: radionuclides (U-238, U-234, Pu-239, 240 and Am-241); anions; and trace metals to groundwaters. The U-238 concentrations in Rocky Flats groundwaters vary from <0.2 to 69 pCi/I (IpCi = 3 ug). However, the activity U-234/U-238 ratios are low and range mostly 1.2 to 2.7. The low activity ratios can be interpreted to suggest that the groundwaters are moving slow (UO{sub 2}(CO{sub 3}){sub 2}{sup 2{minus}}, because of the predominant bicarbonate medium.

  3. Modeling of the performance of weapons MOX fuel in light water reactors

    SciTech Connect (OSTI)

    Alvis, J.; Bellanger, P.; Medvedev, P.G.; Peddicord, K.L.; Gellene, G.I.

    1999-05-01

    Both the Russian Federation and the US are pursing mixed uranium-plutonium oxide (MOX) fuel in light water reactors (LWRs) for the disposition of excess plutonium from disassembled nuclear warheads. Fuel performance models are used which describe the behavior of MOX fuel during irradiation under typical power reactor conditions. The objective of this project is to perform the analysis of the thermal, mechanical, and chemical behavior of weapons MOX fuel pins under LWR conditions. If fuel performance analysis indicates potential questions, it then becomes imperative to assess the fuel pin design and the proposed operating strategies to reduce the probability of clad failure and the associated release of radioactive fission products into the primary coolant system. Applying the updated code to anticipated fuel and reactor designs, which would be used for weapons MOX fuel in the US, and analyzing the performance of the WWER-100 fuel for Russian weapons plutonium disposition are addressed in this report. The COMETHE code was found to do an excellent job in predicting fuel central temperatures. Also, despite minor predicted differences in thermo-mechanical behavior of MOX and UO{sub 2} fuels, the preliminary estimate indicated that, during normal reactor operations, these deviations remained within limits foreseen by fuel pin design.

  4. Analysis of fission gas release in LWR fuel using the BISON code

    SciTech Connect (OSTI)

    G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

  5. Novel Accident-Tolerant Fuel Meat and Cladding

    SciTech Connect (OSTI)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01

    A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas release and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.

  6. Top Ten Reasons for DEOX as a Front End to Pyroprocessing

    SciTech Connect (OSTI)

    B.R. Westphal; K.J. Bateman; S.D. Herrmann

    2007-11-01

    A front end step is being considered to augment chopping during the treatment of spent oxide fuel by pyroprocessing. The front end step, termed DEOX for its emphasis on decladding via oxidation, employs high temperatures to promote the oxidation of UO2 to U3O8 via an oxygen carrier gas. During oxidation, the spent fuel experiences a 30% increase in lattice structure volume resulting in the separation of fuel from cladding with a reduced particle size. A potential added benefit of DEOX is the removal of fission products, either via direct release from the broken fuel structure or via oxidation and volatilization by the high temperature process. Fuel element chopping is the baseline operation to prepare spent oxide fuel for an electrolytic reduction step. Typical chopping lengths range from 1 to 5 mm for both individual elements and entire assemblies. During electrolytic reduction, uranium oxide is reduced to metallic uranium via a lithium molten salt. An electrorefining step is then performed to separate a majority of the fission products from the recoverable uranium. Although DEOX is based on a low temperature oxidation cycle near 500oC, additional conditions have been tested to distinguish their effects on the process.[1] Both oxygen and air have been utilized during the oxidation portion followed by vacuum conditions to temperatures as high as 1200oC. In addition, the effects of cladding on fission product removal have also been investigated with released fuel to temperatures greater than 500oC.

  7. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    SciTech Connect (OSTI)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

    The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

  8. O/M RATIO MEASUREMENT IN PURE AND MIXED OXIDE FULES - WHERE ARE WE NOW?

    SciTech Connect (OSTI)

    J. RUBIN; ET AL

    2000-12-01

    The oxygen-to-metal (O/M) ratio is one of the most critical parameters of nuclear fuel fabrication, and its measurement is closely monitored for manufacturing process control and to ensure the service behavior of the final product. Thermogravimetry is the most widely used method, the procedure for which has remained largely unchanged since its development some thirty years ago. It was not clear to us, however, that this method is still the optimum one in light of advances in instrumentation, and in the current regulatory environment, particularly with regard to waste management and disposal. As part of the MOX fuel fabrication program at Los Alamos, we conducted a comprehensive review of methods for O/M measurements in UO{sub 2}, PuO{sub 2} and mixed oxide fuels for thermal reactors. A concerted effort was made to access information not available in the open literature. We identified approximately thirty five experimental methods that (a) have been developed with the intent of measuring O/M, (b) provided O/M indirectly by suitable reduction of the measured data, or (c) could provide O/M data with suitable data reduction or when combined with other methods. We will discuss the relative strengths and weaknesses of these methods in their application to current routine and small-lot production environment.

  9. Chalcogen-based aerogels as a multifunctional platform for remediation of radioactive iodine

    SciTech Connect (OSTI)

    Riley, Brian J.; Chun, Jaehun; Ryan, Joseph V.; Matyas, Josef; Li, Xiaohong S.; Matson, Dean W.; Sundaram, S. K.; Strachan, Denis M.; Vienna, John D.

    2011-12-22

    Aerogels employing chalcogen-based (i.e., S, Se, and/or Te) structural units and interlinking metals are termed chalcogels and have many emerging applications. Here, chalcogels are discussed in the context of nuclear fuel reprocessing and radioactive waste remediation. Motivated by previous work on removal of heavy metals in aqueous solution, we explored the application of germanium sulfide chalcogels as a sorbent for gas-phase I2 based on Pearson's Hard/Soft Acid-Base (HSAB) principle. This work was driven by a significant need for high-efficiency sorbents for I-129, a long-lived isotope evolved during irradiated UO2 nuclear fuel reprocessing. These chalcogel compositions are shown to possess an affinity for iodine gas, I2(g), at various concentrations in air and the affinity is attributed to a strong chemical attraction between the chalcogen and I2(g), according to the HSAB principle. The high sorption efficiency is facilitated by the high porosity as well as the exceptionally large surface area of the chalcogels.

  10. One pass core design of a super fast reactor

    SciTech Connect (OSTI)

    Liu, Qingjie; Oka, Yoshiaki

    2013-07-01

    One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

  11. Characterization of decontamination and decommissioning wastes expected from the major processing facilities in the 200 Areas

    SciTech Connect (OSTI)

    Amato, L.C.; Franklin, J.D.; Hyre, R.A.; Lowy, R.M.; Millar, J.S.; Pottmeyer, J.A.; Duncan, D.R.

    1994-08-01

    This study was intended to characterize and estimate the amounts of equipment and other materials that are candidates for removal and subsequent processing in a solid waste facility when the major processing and handling facilities in the 200 Areas of the Hanford Site are decontaminated and decommissioned. The facilities in this study were selected based on processing history and on the magnitude of the estimated decommissioning cost cited in the Surplus Facilities Program Plan; Fiscal Year 1993 (Winship and Hughes 1992). The facilities chosen for this study include B Plant (221-B), T Plant (221-T), U Plant (221-U), the Uranium Trioxide (UO{sub 3}) Plant (224-U and 224-UA), the Reduction Oxidation (REDOX) or S Plant (202-S), the Plutonium Concentration Facility for B Plant (224-B), and the Concentration Facility for the Plutonium Finishing Plant (PFP) and REDOX (233-S). This information is required to support planning activities for current and future solid waste treatment, storage, and disposal operations and facilities.

  12. Predicting Stability Constants for Uranyl Complexes Using Density Functional Theory

    SciTech Connect (OSTI)

    Vukovic, Sinisa; Hay, Benjamin P.; Bryantsev, Vyacheslav S.

    2015-04-02

    The ability to predict the equilibrium constants for the formation of 1:1 uranyl:ligand complexes (log K1 values) provides the essential foundation for the rational design of ligands with enhanced uranyl affinity and selectivity. We also use density functional theory (B3LYP) and the IEFPCM continuum solvation model to compute aqueous stability constants for UO22+ complexes with 18 donor ligands. Theoretical calculations permit reasonably good estimates of relative binding strengths, while the absolute log K1 values are significantly overestimated. Accurate predictions of the absolute log K1 values (root mean square deviation from experiment < 1.0 for log K1 values ranging from 0 to 16.8) can be obtained by fitting the experimental data for two groups of mono and divalent negative oxygen donor ligands. The utility of correlations is demonstrated for amidoxime and imide dioxime ligands, providing a useful means of screening for new ligands with strong chelate capability to uranyl.

  13. Simplified treatment of exact resonance elastic scattering model in deterministic slowing down equation

    SciTech Connect (OSTI)

    Ono, M.; Wada, K.; Kitada, T.

    2012-07-01

    Simplified treatment of resonance elastic scattering model considering thermal motion of heavy nuclides and the energy dependence of the resonance cross section was implemented into NJOY [1]. In order to solve deterministic slowing down equation considering the effect of up-scattering without iterative calculations, scattering kernel for heavy nuclides is pre-calculated by the formula derived by Ouisloumen and Sanchez [2], and neutron spectrum in up-scattering term is expressed by NR approximation. To check the verification of the simplified treatment, the treatment is applied to U-238 for the energy range from 4 eV to 200 eV. Calculated multi-group capture cross section of U-238 is greater than that of conventional method and the increase of the capture cross sections is remarkable as the temperature becomes high. Therefore Doppler coefficient calculated in UO{sub 2} fuel pin is calculated more negative value than that on conventional method. The impact on Doppler coefficient is equivalent to the results of exact treatment of resonance elastic scattering reported in previous studies [2-7]. The agreement supports the validation of the simplified treatment and therefore this treatment is applied for other heavy nuclide to evaluate the Doppler coefficient in MOX fuel. The result shows that the impact of considering thermal agitation in resonance scattering in Doppler coefficient comes mainly from U-238 and that of other heavy nuclides such as Pu-239, 240 etc. is not comparable in MOX fuel. (authors)

  14. Pebble Bed Boiling Water Reactor Concept With Superheated Steam

    SciTech Connect (OSTI)

    Tsiklauri, G.; Newman, D.; Meriwether, G.; Korolev, V. [Pacific Northwest National Laboratory, P.O. Box 999 Richland, WA 99352 (United States)

    2002-07-01

    An Advanced Nuclear Reactor concept is presented which extends Boiling Water Reactor technology with micro-fuel elements (MFE) and produces superheated steam. A nuclear plant with MFE is highly efficient and safe, due to ceramic-clad nuclear fuel. Water is used as both moderator and coolant. The fuel consists of spheres of about 1.5 mm diameter of UO{sub 2} with several external coatings of different carbonaceous materials. The outer coating of the particles is SiC, manufactured with chemical vapor disposition (CVD) technology. Endurance of the integrity of the SiC coating in water, air and steam has been demonstrated experimentally in Germany, Russia and Japan. This paper describes a result of a preliminary design and analysis of 3750 MWt (1500 MWe) plant with standard pressure of 16 MPa, which is widely achieved in the vessel of pressurized-water type reactors. The superheated steam outlet temperature of 550 deg. C elevates the steam cycle to high thermal efficiency of 42%. (authors)

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

    SciTech Connect (OSTI)

    Not Available

    2011-06-22

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

  16. Use of ion conductors in the pyrochemical reduction of oxides

    DOE Patents [OSTI]

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

    1994-01-01

    An electrochemical process and electrochemical cell for reducing a metal oxide are provided. First the oxide is separated as oxygen gas using, for example, a ZrO.sub.2 oxygen ion conductor anode and the metal ions from the reduction salt are reduced and deposited on an ion conductor cathode, for example, sodium ion reduced on a .beta.-alumina sodium ion conductor cathode. The generation of and separation of oxygen gas avoids the problem with chemical back reaction of oxygen with active metals in the cell. The method also is characterized by a sequence of two steps where an inert cathode electrode is inserted into the electrochemical cell in the second step and the metallic component in the ion conductor is then used as the anode to cause electrochemical reduction of the metal ions formed in the first step from the metal oxide where oxygen gas formed at the anode. The use of ion conductors serves to isolate the active components from chemically reacting with certain chemicals in the cell. While applicable to a variety of metal oxides, the invention has special importance for reducing CaO to Ca.degree. used for reducing UO.sub.2 and PuO.sub.2 to U and Pu.

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

    SciTech Connect (OSTI)

    Lloyd, R.

    1980-10-01

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

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

    SciTech Connect (OSTI)

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

    2009-04-01

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

  19. Fabrication of high exposure nuclear fuel pellets

    DOE Patents [OSTI]

    Frederickson, James R.

    1987-01-01

    A method is disclosed for making a fuel pellet for a nuclear reactor. A mixture is prepared of PuO.sub.2 and UO.sub.2 powders, where the mixture contains at least about 30% PuO.sub.2, and where at least about 12% of the Pu is the Pu.sup.240 isotope. To this mixture is added about 0.3 to about 5% of a binder having a melting point of at least about 250.degree. F. The mixture is pressed to form a slug and the slug is granulated. Up to about 4.7% of a lubricant having a melting point of at least about 330.degree. F. is added to the granulated slug. Both the binder and the lubricant are selected from a group consisting of polyvinyl carboxylate, polyvinyl alcohol, naturally occurring high molecular weight cellulosic polymers, chemically modified high molecular weight cellulosic polymers, and mixtures thereof. The mixture is pressed to form a pellet and the pellet is sintered.

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

    SciTech Connect (OSTI)

    Contescu, Cristian I

    2006-01-01

    This report supports the effort for development of small scale fabrication of UCO (a mixture of UO{sub 2} and UC{sub 2}) fuel kernels for the generation IV high temperature gas reactor program. In particular, it is focused on optimization of dispersion conditions of carbon black in the broths from which carbon-containing (UO{sub 2} {center_dot} H{sub 2}O + C) gel spheres are prepared by internal gelation. The broth results from mixing a hexamethylenetetramine (HMTA) and urea solution with an acid-deficient uranyl nitrate (ADUN) solution. Carbon black, which is previously added to one or other of the components, must stay dispersed during gelation. The report provides a detailed description of characterization efforts and results, aimed at identification and testing carbon black and surfactant combinations that would produce stable dispersions, with carbon particle sizes below 1 {micro}m, in aqueous HMTA/urea and ADUN solutions. A battery of characterization methods was used to identify the properties affecting the water dispersability of carbon blacks, such as surface area, aggregate morphology, volatile content, and, most importantly, surface chemistry. The report introduces the basic principles for each physical or chemical method of carbon black characterization, lists the results obtained, and underlines cross-correlations between methods. Particular attention is given to a newly developed method for characterization of surface chemical groups on carbons in terms of their acid-base properties (pK{sub a} spectra) based on potentiometric titration. Fourier-transform infrared (FTIR) spectroscopy was used to confirm the identity of surfactants, both ionic and non-ionic. In addition, background information on carbon black properties and the mechanism by which surfactants disperse carbon black in water is also provided. A list of main physical and chemical properties characterized, samples analyzed, and results obtained, as well as information on the desired trend or range of values generally associated with better dispersability, is provided in the Appendix. Special attention was given to characterization of several surface-modified carbon blacks produced by Cabot Corporation through proprietary diazonium salts chemistry. As demonstrated in the report, these advanced carbons offer many advantages over traditional dispersions. They disperse very easily, do not require intensive mechanical shearing or sonication, and the particle size of the dispersed carbon black aggregates is in the target range of 0.15-0.20 {micro}m. The dispersions in water and HMTA/urea solutions are stable for at least 30 days; in conditions of simulated broth, the dispersions are stable for at least 6 hours. It is proposed that the optimization of the carbon black dispersing process is possible by replacing traditional carbon blacks and surfactants with surface-modified carbon blacks having suitable chemical groups attached on their surface. It is recognized that the method advanced in this report for optimizing the carbon black dispersion process is based on a limited number of tests made in aqueous and simulated broth conditions. The findings were corroborated by a limited number of tests carried out with ADUN solutions by the Nuclear Science and Technology Division at Oak Ridge National Laboratory (ORNL). More work is necessary, however, to confirm the overall recommendation based on the findings discussed in this report: namely, that the use of surface-modified carbon blacks in the uranium-containing broth will not adversely impact the chemistry of the gelation process, and that high quality uranium oxicarbide (UCO) kernels will be produced after calcination.

  1. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    SciTech Connect (OSTI)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no significant failure is to be expected for the reference fuel particle during normal operation. It was found, however, that the sensitivity of the coating stress to the CO production in the kernel was large. The CO production is expected to be higher in DB fuel than in UO2 fuel, but its exact level has a high uncertainty. Furthermore, in the fuel performance analysis transient conditions were not yet taken into account. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high transuranic [TRU] content and high burn-up). Accomplishments of this work include: •Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel. •Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Uranium. •Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Modified Open Cycle Components. •Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Americium targets.

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

    SciTech Connect (OSTI)

    Chen Zhu

    2006-06-15

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

  3. O and H diffusion in uraninite: Implications for fluid-uraninite interactions, nuclear waste disposal, and nuclear forensics

    SciTech Connect (OSTI)

    Fayek, Mostafa [University of Manitoba, Canada; Anovitz, Lawrence {Larry} M [ORNL; Cole, David [Ohio State University; Bostick, Debra A [ORNL

    2011-01-01

    Diffusion coefficients for oxygen and hydrogen were determined from a series of natural uraninite H2O experiments between 50 and 700 C. Under hydrous conditions there are two diffusion mechanisms: (1) an initial extremely fast-path diffusion mechanism that overprinted the oxygen isotopic composition of the entire crystals regardless of temperature and (2) a slower volume-diffusive mechanism dominated by defect clusters that displace or eject nearest neighbor oxygen atoms to form two interstitial sites and two partial vacancies, and by vacancy migration. Using the volume diffusion coefficients in the temperature range of 400 600 C, diffusion coefficients for oxygen can be represented by D = 1.90e5 exp (123,382 J/RT) cm2/s and for temperatures between 100 and 300 C the diffusion coefficients can be represented by D = 1.95e10 exp (62484 J/ RT) cm2/s, where the activation energies for uraninite are 123.4 and 62.5 kJ/mol, respectively. Hydrogen diffusion in uraninite appears to be controlled by similar mechanisms as oxygen. Using the volume diffusion coefficients for temperatures between 50 and 700 C, diffusion coefficients for hydrogen can be represented by D = 9.28e6 exp (156,528 J/RT) cm2/s for temperatures between 450 and 700 C and D = 1.39e14 exp (34518 J/RT) cm2/s for temperatures between 50 and 400 C, where the activation energies for uraninite are 156.5 and 34.5 kJ/mol, respectively. Results from these new experiments have implications for isotopic exchange during natural UO2 water interactions. The exceptionally low d18O values of natural uraninites (i.e. 32& to 19.5&) from unconformity-type uranium deposits in Saskatchewan, in conjunction with theoretical and experimental uraninite water and UO3 water fractionation factors, suggest that primary uranium mineralization is not in oxygen isotopic equilibrium with coeval clay and silicate minerals. The low d18O values have been interpreted as resulting from the low temperature overprinting of primary uranium mineralization in the presence of relatively modern meteoric fluids having d18O values of ca. 18&, despite petrographic and U Pb isotope data that indicate limited alteration. Our data show that the anomalously low oxygen isotopic composition of the uraninite from the Athabasca Basin can be due to meteoric water overprinting under reducing conditions, and meteoric water or groundwater can significantly affect the oxygen isotopic composition of spent nuclear fuel in a geologic repository, with minimal change to the chemical composition or texture. Moreover, the rather fast oxygen and hydrogen diffusion coefficients for uraninite, especially at low temperatures, suggest that oxygen and hydrogen diffusion may impart characteristic isotopic signals that can be used to track the route of fissile material.

  4. Actinide Foil Production for MPACT Research

    SciTech Connect (OSTI)

    Beller, Denis

    2012-10-31

    Sensitive fast-neutron detectors are required for use in lead slowing down spectrometry (LSDS), an active interrogation technique for used nuclear fuel assay for Materials Protection, Accounting, and Controls Technologies (MPACT). During the past several years UNLV sponsored a research project at RPI to investigate LSDS; began development of fission chamber detectors for use in LSDS experiments in collaboration with INL, LANL, and Oregon State U.; and participated in a LSDS experiment at LANL. In the LSDS technique, research has demonstrated that these fission chamber detectors must be sensitive to fission energy neutrons but insensitive to thermal-energy neutrons. Because most systems are highly sensitive to large thermal neutron populations due to the well-known large thermal cross section of 235U, even a miniscule amount of this isotope in a fission chamber will overwhelm the small population of higher-energy neutrons. Thus, fast-fission chamber detectors must be fabricated with highly depleted uranium (DU) or ultra-pure thorium (Th), which is about half as efficient as DU. Previous research conducted at RPI demonstrated that the required purity of DU for assay of used nuclear fuel using LSDS is less than 4 ppm 235U, material that until recently was not available in the U.S. In 2009 the PI purchased 3 grams of ultra-depleted uranium (uDU, 99.99998% 238U with just 0.2 ���± 0.1 ppm 235U) from VNIIEF in Sarov, Russia. We received the material in the form of U3O8 powder in August of 2009, and verified its purity and depletion in a FY10 MPACT collaboration project. In addition, chemical processing for use in FC R&D was initiated, fission chamber detectors and a scanning alpha-particle spectrometer were developed, and foils were used in a preliminary LSDS experiment at a LANL/LANSCE in Sept. of 2010. The as-received U3O8 powder must be chemically processed to convert it to another chemical form while maintaining its purity, which then must be used to electro-deposit U or UO2 in extremely thin layers (1 to 2 mg/cm2) on various media such as films, foils, or discs. After many months of investigation and trials in FY10 and 11, UNLV researchers developed a new method to produce pure UO2 deposits on foils using a unique approach, which has never been demonstrated, that involves dissolution of U3O8 directly into room temperature ionic liquid (RTIL) followed by electrodeposition from the RTIL-uDU solution (Th deposition from RTIL had been previously demonstrated). The high-purity dissolution of the U3O8 permits the use of RTIL solutions for deposition of U on metal foils in layers without introducing contamination. In FY10 and early FY11 a natural U surrogate for the uDU was used to investigate this and other techniques. In this research project UNLV will deposit directly from RTIL to produce uDU and Th foils devoid of possible contaminants. After these layers have been deposited, they will be examined for purity and uniformity. UNLV will complete the development and demonstration of the RTIL technology/ methodology to prepare uDU and Th samples for use in constructing fast-neutron detectors. Although this material was purchased for use in research using fast-fission chamber detectors for active inspection techniques for MPACT, it could also contribute to R&D for other applications, such as cross section measurements or neutron spectroscopy for national security

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

    SciTech Connect (OSTI)

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

    2012-07-31

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

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

    SciTech Connect (OSTI)

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

    2009-05-29

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

  7. Observation of Radiolytic Field Alteration of the Uranyl Cation in Bicarbonate Solution

    SciTech Connect (OSTI)

    Snow, Lanee A.; McNamara, Bruce K.; Sinkov, Sergey I.; Cho, Herman M.; Friese, Judah I.

    2006-12-01

    In previous work we demonstrated that radiolysis of uranyl tris carbonate in near neutral pH to alkaline carbonate solutions, could be followed by 13C NMR. Radiolysis of the complex produced novel uranyl peroxo carbonate solution state species, whose structures depended on the pH and radiolytic dose rate. In this work, we investigate speciation of the uranyl carbonate trimer which is predominant in bicarbonate solution near pH 5.9. We observe radiolytically derived speciation to different mixed peroxy carbonate species than seen in the higher pH solutions. Auto radiolysis of uranium (VI) carbonate solutions between pH 5.9 and 7.2 is shown to alter the uranium speciation over relatively short periods of time and was followed by 13C NMR and visible spectrophotometry, using dissolved 233(UO2)3(CO3)6 6- both as the radiolysis source (D= 14.9 Gy/hr) and as a trap for the newly formed hydrogen peroxide. Direct addition of hydrogen peroxide to solutions of the uranyl-carbonate trimer is shown to reproduce the 13 C NMR signatures of the complexe(s) formed by radiolysis, but additionally a variety of new complexes are revealed. Ratios of H2O2/trimer < 1.5 produced a uranyl peroxo carbonate adduct, that is shown to be common to the radiolytically produced species. Ratios of H2O2/ trimer >1 resulted in formation of stable higher order peroxo carbonate complexes. The 13C NMR signatures and visible spectra of these complexes are described here. Rigorous characterization of the species is an ongoing effort.

  8. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect (OSTI)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly insertion into a commercial reactor within the desired timeframe (by 2022).

  9. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    SciTech Connect (OSTI)

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

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  10. Results of the Triggered TROI Steam Explosion Experiments with a Narrow Interaction Vessel

    SciTech Connect (OSTI)

    Kim, J.H.; Park, I.K.; Min, B.T.; Hong, S.W.; Hong, S.H.; Song, J.H.; Kim, H.D.

    2006-07-01

    The effect of the interaction vessel geometry has been studied on the energetics of a steam explosion in the TROI experiment. The interaction vessel was 30 cm in diameter (1-D geometry). Two types of corium composition were used as a melt. One was spontaneously non-explosive 80 : 20 corium (UO{sub 2} : ZrO{sub 2}) and the other was spontaneously explosive 70 : 30 eutectic corium. A test with 80 : 20 corium was carried out without an external triggering. Another test with 80 : 20 corium was also carried out with an external trigger. In addition, two tests with 70 : 30 corium were carried out with an external trigger. The external trigger was applied just before the contact between the melt and the bottom of the interaction vessel. This time was thought to be the triggering time of a spontaneous steam explosion. The external trigger was a chemical explosive of PETN 1.0 g. However, none of these tests led to steam explosions even with an external triggering. Since eutectic corium led to spontaneous or triggered steam explosions in a previous test using a 60 cm wide interaction vessel (3-D geometry), it is quite probable that a geometry effect of the interaction vessel could exist. The reason for no steam explosions in the narrow (1-D) interaction vessel is believed to be a relatively high void fraction in the vessel when compared with the 3-D vessel. Due to the high void fraction, a steam explosion could not propagate to the surroundings of the melt where the water was depleted. (authors)

  11. Rapid Computer Aided Ligand Design and Screening of Precious Metal Extractants from TRUEX Raffinate with Experimental Validation

    SciTech Connect (OSTI)

    Clark, Aurora Sue; Wall, Nathalie; Benny, Paul

    2015-11-16

    Rhodium is the most extensively used metal in catalytic applications (i.e., oxidation of ammonia to nitric acid, automobile catalytic converters) and occurs in mixed ores with platinum group metals (PGMs) in the earth’s crust in low concentrations (0.4 - 10 ppb). It is resistant to aerial oxidation and insoluble in all acids, including aqua regia making classical purification methods time-consuming and inefficient. To ensure adequate purity, several precipitation and dissolution steps are necessary during separation. Low abundance, high demand, and extensive processing make rhodium the most expensive of all PGMs. From alternative sources, rhodium is also produced in sufficient quantities (0.47 kg per ton initial heavy metal (tIHM)) during the fission of U-235 in nuclear reactors along with other PGMs (i.e., Ag, Pd, Ru). A typical power water reactor operating with UO2 fuel after cooling can generate PGMs in quantities greater than found in the earth’s crust (0.5-2 kg/tIHM). This currently untapped supply of PGMs has the potential to yield $5,000-30,000/tIHM. It is estimated that by the year 2030, the amount of rhodium generated in reactors could exceed natural reserves. Typical SNF processing removes the heavier lanthanides and actinides and can leave PGMs at ambient temperatures in aqueous acidic (Cl? or NO3?; pH < 1) solutions at various activities. While the retrieval of these precious metals from SNF would minimize waste generation and improve resource utilization, it has been difficult to achieve thus far.

  12. Application of wavelet scaling function expansion continuous-energy resonance calculation method to MOX fuel problem

    SciTech Connect (OSTI)

    Yang, W.; Wu, H.; Cao, L.

    2012-07-01

    More and more MOX fuels are used in all over the world in the past several decades. Compared with UO{sub 2} fuel, it contains some new features. For example, the neutron spectrum is harder and more resonance interference effects within the resonance energy range are introduced because of more resonant nuclides contained in the MOX fuel. In this paper, the wavelets scaling function expansion method is applied to study the resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are utilized within resonance energy, which means that it's capable to solve problems with serious resonance interference effects without iteration calculations. Therefore, this method adapts to treat the MOX fuel resonance calculation problem natively. Furthermore, plutonium isotopes have fierce oscillations of total cross-section within thermal energy range, especially for {sup 240}Pu and {sup 242}Pu. To take thermal resonance effect of plutonium isotopes into consideration the wavelet scaling function expansion continuous-energy resonance calculation code WAVERESON is enhanced by applying the free gas scattering kernel to obtain the continuous-energy scattering source within thermal energy range (2.1 eV to 4.0 eV) contrasting against the resonance energy range in which the elastic scattering kernel is utilized. Finally, all of the calculation results of WAVERESON are compared with MCNP calculation. (authors)

  13. Reactions of aluminum with uranium fluorides and oxyfluorides

    SciTech Connect (OSTI)

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

    1991-12-31

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

  14. Capsule HRB-15A postirradiation examination report

    SciTech Connect (OSTI)

    Ketterer, J.W.; Kania, M.J.; Bullock, R.E.; Siman-Tov, I.I.

    1984-07-01

    Capsule HRB-15A successfully irradiated a variety of LHTGR fuel particle types to peak exposures of 6.5 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/ fast fluence, 29% FIMA fissile burnup and 6.4% FIMA fertile burnup, and a time-average irradiation temperature of 1100/sup 0/C. The capsule was irradiated in HFIR for 6 months. The HRB-15A fuel rods were in excellent condition following irradiation. Candidate TRISO-coated LEU UC/UO/sub 2/ fuel particles performed adequately for the HTGR steam cycle application. The SiC coatings on the two UC/sub 0/ /sub 5/O/sub 1/ /sub 5/ TRISO batches were least retentive of Ag-110m during irradiation. Severe degradation of the SiC layers by fission product attack had occurred in 70% of these particles with much deeper penetration than in other TRISO batches. Large diameter (nominal 600 ..mu..m kernel) ThO/sub 2/ particles were judged unacceptable for use in an all-TRISO separable fuel system because of high OPyC and SiC failure levels. ZrC barrier layers in two ZrC-TRISO-coated fissile particle batches showed superior resistance to fission products relative to SiC. However, one of the ZrC-TRISO batches was poorly retentive of Ag and Eu, and fuel rods containing both batches had the highest end-of-life fission gas release values. Performance models for defective particles were shown to be overly conservative. Failure rate of OPyC coatings on TRISO particles was found to increase with the particle diameter.

  15. Am phases in the matrix of a U–Pu–Zr alloy with Np, Am, and rare-earth elements

    SciTech Connect (OSTI)

    Janney, Dawn E.; Kennedy, J. Rory; Madden, James W.; O’Holleran, Thomas P.

    2015-01-01

    Phases and microstructures in the matrix of an as-cast U-Pu-Zr alloy with 3 wt% Am, 2% Np, and 8% rare-earth elements were characterized by scanning and transmission electron microscopy. The matrix consists primarily of two phases, both of which contain Am: ?-(U, Np, Pu, Am) (~70 at% U, 5% Np, 14% Pu, 1% Am, and 10% Zr) and ?-(U, Np, Pu, Am)Zr2 (~25% U, 2% Np, 10-15% Pu, 1-2% Am, and 55-60 at% Zr). These phases are similar to those in U-Pu-Zr alloys, although the Zr content in ?-(U, Np, Pu, Am) is higher than that in ?-(U, Pu) and the Zr content in ?-(U, Np, Pu, Am)Zr2 is lower than that in ?-UZr2. Nanocrystalline actinide oxides with structures similar to UO2 occurred in some areas, but may have formed by reactions with the atmosphere during sample handling. Planar features consisting of a central zone of ?-(U, Np, Pu, Am) bracketed by zones of ?-(U, Np, Pu, Am)Zr2 bound irregular polygons ranging in size from a few micrometers to a few tens of micrometers across. The rest of the matrix consists of elongated domains of ?-(U, Np, Pu, Am) and ?-(U, Np, Pu, Am)Zr2. Each of these domains is a few tens of nanometers across and a few hundred nanometers long. The domains display strong preferred orientations involving areas a few hundred nanometers to a few micrometers across.

  16. Exploring high temperature phenomena related to post-detonation using an electric arc

    SciTech Connect (OSTI)

    Dai, Z. R. Crowhurst, J. C.; Grant, C. D.; Knight, K. B.; Tang, V.; Chernov, A. A.; Cook, E. G.; Lotscher, J. P.; Hutcheon, I. D.

    2013-11-28

    We report a study of materials recovered from a uranium-containing plasma generated by an electric arc. The device used to generate the arc is capable of sustaining temperatures of an eV or higher for up to 100??s. Samples took the form of a 4??m-thick film deposited onto 8 pairs of 17??m-thick Cu electrodes supported on a 25??m-thick Kapton backing and sandwiched between glass plates. Materials recovered from the glass plates and around the electrode tips after passage of an arc were characterized using scanning and transmission electron microscopy. Recovered materials included a variety of crystalline compounds (e.g., UO{sub 2}, UC{sub 2}, UCu{sub 5},) as well as mixtures of uranium and amorphous glass. Most of the materials collected on the glass plates took the form of spherules having a wide range of diameters from tens of nanometers to tens of micrometers. The composition and size of the spherules depended on location, indicating different chemical and physical environments. A theoretical analysis we have carried out suggests that the submicron spherules presumably formed by deposition during the arc discharge, while at the same time the glass plates were strongly heated due to absorption of plasma radiation mainly by islands of deposited metals (Cu, U). The surface temperature of the glass plates is expected to have risen to ?2300?K thus producing a liquefied glass layer, likely diffusions of the deposited metals on the hot glass surface and into this layer were accompanied by chemical reactions that gave rise to the observed materials. These results, together with the compact scale and relatively low cost, suggest that the experimental technique provides a practical approach to investigate the complex physical and chemical processes that occur when actinide-containing material interacts with the environment at high temperature, for example, during fallout formation following a nuclear detonation.

  17. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    SciTech Connect (OSTI)

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) 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 can 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. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

  18. Waste analysis plan for confirmation or completion of Tank Farms backlog waste designation. Revision 1

    SciTech Connect (OSTI)

    Not Available

    1993-10-01

    On January 23, 1992, waste management problems in the Tank Farms were acknowledged through an Unusual Occurrence (UO) Report No. RL-WHC-TANKFARM-19920007 (DOE-RL 1992). On March 10, 1993, the Washington State Department of Ecology (Ecology) issued Order 93NM-201 (Order) to the US Department of Energy, Richland Operations Office (DOE-RL) and the Westinghouse Hanford Company (Westinghouse Hanford) asserting that ``DOE-RL and Westinghouse Hanford have failed to designate approximately 2,000 containers of solid waste in violation of WAC 173-303170(l)(a) and the procedures of WAC 173-303-070`` (Ecology 1993). On June 30, 1993, a Settlement Agreement and Order Thereon (Settlement Agreement) among Ecology, DOE-RL, and Westinghouse Hanford was approved by the Pollution Control Hearings Board (PCHB). Item 3 of the Settlement Agreement requires that DOE-RL and Westinghouse Hanford submit a waste analysis plan (WAP) for the waste subject to the Order by September 1, 1993 (PCHB 1993). This WAP satisfies the requirements of Item 3 of the Order as amended per the Settlement Agreement. Item 3 states: ``Within forty (40) calendar days of receipt of this Order, DOE-RL and WHC provide Ecology with a waste analysis plan for review and approval detailing the established criteria and procedures for waste inspection, segregation, sampling, designation, and repackaging of all containers reported in item No. 1. The report shall include sampling plan criteria for different contaminated media, i.e., soils, compactable waste, high-efficiency particular air (HEPA) filters, etc., and a schedule for completing the work within the time allowed under this Order.``

  19. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S.

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

  20. Chemical Engineering Division fuel cycle programs. Quarterly progress report, July-September 1978

    SciTech Connect (OSTI)

    Steindler, M.J.; Ader, M.; Barletta, R.E.

    1980-01-01

    Fuel cycle work included hydraulic performance and extraction efficiency of eight-stage centrifugal contactors, flowsheet for the Aralex process, Ru and Zr extraction in a miniature centrifugal contactor, study of Zr aging in the organic phase and its effect on Zr extraction and hydraulic testing of the 9-cm-ID contactor. Work for predicting accident consequences in LWR fuel processing covered the relation between energy input (to subdivide a solid) and the modes of particle size frequency distribution. In the pyrochemical and dry processing program corrosion-testing materials for containment vessels and equipment for studying carbide reactions in bismuth is under way. Analytical studies have been made of salt-transport processes; efforts to spin tungsten crucibles 13 cm dia continue, and other information on tungsten fabrication is being assembled; the process steps of the chloride volatility process have been demonstrated and the thoria powder product used to produce oxide pellets; solubility of UO/sub 2/, PuO/sub 2/, and fission products in molten alkali nitrates is being investigated; work was continued on reprocessing actinide oxides by extracting the actinides into ammonium chloroaluminate from bismuth; the preparation of thorium-uranium carbide from the oxide is being studied as a means of improving the oxide reactivity; studies are in progress on producing uranium metal and decontaminated ThO/sub 2/ by the reaction of (Th,U)O/sub 2/ solid solution in molten salts containing ThCl/sub 4/ and thorium metal chips. In the molten tin process, no basic thermodynamic or kinetic factors have been found that may limit process development.

  1. Chemical Engineering Division fuel cycle programs. Quarterly progress report, October-December 1978

    SciTech Connect (OSTI)

    Steindler, M J; Ader, M; Barletta, R E

    1980-01-01

    In the program on pyrochemical and dry processing methods (PDPM) for nuclear fuel, tungsten crucibles were successfully spun for use in laboratory-scale experiments. Corrosion testing of refractory metals and alloys in PDPM environments was done. Ceramic substrates were successfully coated with tungsten. Solubility measurements were made to determine Cd/Mg alloy composition and temperature at which dissolved Th will precipitate. Experiments were started to study the reduction of high-fired ThO/sub 2/ with Ca in a molten metal-molten salt system. Work on the fused salt electrolysis of CaO was started. Equipment for determining phase diagrams for U-Cu-Mg system was set up. The reaction of UO/sub 2/ with molten equimolar NaNO/sub 3/-KNO/sub 3/ was studied as part of a project to identify chemically feasible nonaqueous fuel reprocessing methods. Work was continued on development of a flowsheet for reprocessing actinide oxides by extracting actinides into ammonium chloro-aluminate (and alternative salts) from a bismuth solution. Preparation of Th, U, and Pu nitrides after dissolution of spent fuel elements in molten tin is being studied. Leach rates of glass beads, pulverized beads, and beads encapsulated in a lead matrix with no protective envelope were studied. A method (employing no pressure or vacuum systems) of encapsulating various solid wastes in a lead metal matrix was developed and tested. A preliminary integration was made of earlier data on effects of impacts on metal-matrix waste forms.Leach migration experiments were compared with conventional infiltration experiments as methods of evaluating geologic formations as barriers to nuclide migration. The effect of the streaming potential on the rates of transport of radioactive I/sup -/ and Na/sup +/ through kaolinite columns was measured, as well as adsorption of iodide and iodate by several compounds; implications of the results upon the disposal of radioactive iodine are discussed.

  2. Radionuclide release from PWR fuels in a reference tuff repository groundwater

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1985-03-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high-level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: fuel rod sections split open to expose bare fuel particles; rod sections with water-tight end fittings with a 2.5-cm long by 150-{mu}m wide slit through the cladding; rod sections with water-tight end fittings and two 200-{mu}m-diameter holes through the cladding; and undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested on deionized water. Selected initial results are also given for Turkey Point fuel specimens tested on J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO{sub 2} spent fuel matrix dissolves. Fractional release of {sup 137}Cs and {sup 99}Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water. 8 references, 7 figures, 9 tables.

  3. Radionuclide release from PWR fuels in a reference tuff repository groundwater subsquently changed to Radionuclide release from PWR fuels in J-13 well water

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1985-04-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: (1) fuel rod sections split open to expose bare fuel particles; (2) rod sections with water-tight end fittings with a 2.5-cm long by 150-{mu}m wide slit through the cladding; (3) rod sections with water-tight end fittings and two 200-{mu}m diameter holes through the cladding; and (4) undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested in deionized water. Selected initial results are also given for Turkey Point fuel specimens tested in J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO{sub 2} spent fuel matrix dissolves. Fractional release of {sup 137}Cs and {sup 99}Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water.

  4. Optimization of multi-group cross sections for fast reactor analysis

    SciTech Connect (OSTI)

    Chin, M. R.; Manalo, K. L.; Edgar, C. A.; Paul, J. N.; Molinar, M. P.; Redd, E. M.; Yi, C.; Sjoden, G. E.

    2013-07-01

    The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO{sub 2}-UO{sub 2} with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)

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

    SciTech Connect (OSTI)

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

    2013-06-01

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

  6. Investigation of the effects of soluble boron tracking on coupled CTF / NEM, LWR simulations

    SciTech Connect (OSTI)

    Biery, M.; Avramova, M.; Ivanov, K.

    2013-07-01

    The primary objective of this study is to evaluate the effects of introducing a boron tracking capability to the COBRA-TF / NEM code coupling. The Pennsylvania State University (PSU) versions of COBRA-TF - CTF, and Nodal Expansion Method (NEM) codes are utilized. Previous implementations of the CTF / NEM coupled code had no capability to model soluble boron feedback effects due to boron transport. This study builds upon the validation and qualification efforts of the boron tracking model implementation in CTF by modeling the boron feedback calculated by the CTF boron tracking model in NEM. The core model chosen for this study is the Purdue MOX/UO{sub 2} core model used in the 2007 OECD/NRC code benchmark study. Following the implementation of an explicit online coupling scheme and accompanying k-search routine, the newly coupled CTF / NEM code version with boron tracking is compared to prior results of the non-boron tracking CTF / NEM code version at steady-state hot full power and hot zero power conditions. It was found that the boron tracking model exhibited little influence on the hot zero power result as expected due to a smaller heat flux, which does not significantly change the moderator density and boron concentration as the moderator travels up the axial core length. Meanwhile the boron tracking model had a much greater impact on the hot full power results, predicting the critical inlet boron concentration to be 9.9 ppm below the non-boron tracking result due to greater and more rapid changes in boron concentration corresponding to the reduction in moderator density from being more rapidly heated. (authors)

  7. A Uranium Bioremediation Reactive Transport Benchmark

    SciTech Connect (OSTI)

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

    2015-06-01

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

  8. Elevated concentrations of U and co-occurring metals in abandoned mine wastes in a northeastern Arizona Native American community

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

    Blake, Johanna M.; Avasarala, Sumant; Artyushkova, Kateryna; Ali, Abdul -Mehdi S.; Brearley, Adrian J.; Shuey, Christopher; Robinson, Wm. Paul; Nez, Christopher; Bill, Sadie; Lewis, Johnnye; et al

    2015-07-09

    The chemical interactions of U and co-occurring metals in abandoned mine wastes in a Native American community in northeastern Arizona were investigated using spectroscopy, microscopy and aqueous chemistry. The concentrations of U (67–169 μg L–1) in spring water samples exceed the EPA maximum contaminant limit of 30 μg L–1. Elevated U (6,614 mg kg–1), V (15,814 mg kg–1), and As (40 mg kg–1) concentrations were detected in mine waste solids. Spectroscopy (XPS and XANES) solid analyses identified U (VI), As (-I and III) and Fe (II, III). Linear correlations for the release of U vs V and As vs Femore » were observed for batch experiments when reacting mine waste solids with 10 mM ascorbic acid (~pH 3.8) after 264 h. The release of U, V, As, and Fe was at least 4-fold lower after reaction with 10 mM bicarbonate (~pH 8.3). These results suggest that U–V mineral phases similar to carnotite [K2(UO2)2V2O8] and As–Fe-bearing phases control the availability of U and As in these abandoned mine wastes. Elevated concentrations of metals are of concern due to human exposure pathways and exposure of livestock currently ingesting water in the area. This study contributes to understanding the occurrence and mobility of metals in communities located close to abandoned mine waste sites.« less

  9. Elevated concentrations of U and co-occurring metals in abandoned mine wastes in a northeastern Arizona Native American community

    SciTech Connect (OSTI)

    Blake, Johanna M.; Avasarala, Sumant; Artyushkova, Kateryna; Ali, Abdul -Mehdi S.; Brearley, Adrian J.; Shuey, Christopher; Robinson, Wm. Paul; Nez, Christopher; Bill, Sadie; Lewis, Johnnye; Hirani, Chris; Pacheco, Juan S. Lezama; Cerrato, José M.

    2015-07-09

    The chemical interactions of U and co-occurring metals in abandoned mine wastes in a Native American community in northeastern Arizona were investigated using spectroscopy, microscopy and aqueous chemistry. The concentrations of U (67–169 ?g L–1) in spring water samples exceed the EPA maximum contaminant limit of 30 ?g L–1. Elevated U (6,614 mg kg–1), V (15,814 mg kg–1), and As (40 mg kg–1) concentrations were detected in mine waste solids. Spectroscopy (XPS and XANES) solid analyses identified U (VI), As (-I and III) and Fe (II, III). Linear correlations for the release of U vs V and As vs Fe were observed for batch experiments when reacting mine waste solids with 10 mM ascorbic acid (~pH 3.8) after 264 h. The release of U, V, As, and Fe was at least 4-fold lower after reaction with 10 mM bicarbonate (~pH 8.3). These results suggest that U–V mineral phases similar to carnotite [K2(UO2)2V2O8] and As–Fe-bearing phases control the availability of U and As in these abandoned mine wastes. Elevated concentrations of metals are of concern due to human exposure pathways and exposure of livestock currently ingesting water in the area. This study contributes to understanding the occurrence and mobility of metals in communities located close to abandoned mine waste sites.

  10. Assessment of Possible Cycle Lengths for Fully Encapsulated Microstructure fueled light water reactor Concepts

    SciTech Connect (OSTI)

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

    2013-02-01

    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 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. Another mandatory constraint is use of low enriched uranium (at or below 20 w/o). The feasibility of using this fuel is assessed by looking at two factors: cycle lengths and fuel material failure rates. Other considerations (e.g., safety parameters such as reactivity coefficients, feedback, etc.) were not considered at this stage of the study. The study includes the examination of increases in the TRISO kernel sizes without changing the thickness of any of the coating layers. In addition, cases where the buffer layer thickness is allowed to vary are also considered. The study shows that a naïve use of UO2 (even up to 20 w/o enrichment) 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. The increase of fissile inventory can be accomplished through multiple means, including higher particle packing fraction, higher enrichment, larger fuel kernel sizes, and the use of higher density fuels (that contain a higher number of U atoms per unit volume). In this study, starting with the recognized highest packing fraction practically achievable (44%), combinations of the other means 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.

  11. THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

    SciTech Connect (OSTI)

    Croft, Stephen; Favalli, Andrea; Swinhoe, Martyn T.

    2012-06-19

    Verification of commercial low enriched uranium light water reactor fuel takes place at the fuel fabrication facility as part of the overall international nuclear safeguards solution to the civilian use of nuclear technology. The fissile mass per unit length is determined nondestructively by active neutron coincidence counting using a neutron collar. A collar comprises four slabs of high density polyethylene that surround the assembly. Three of the slabs contain {sup 3}He filled proportional counters to detect time correlated fission neutrons induced by an AmLi source placed in the fourth slab. Historically, the response of a particular collar design to a particular fuel assembly type has been established by careful cross-calibration to experimental absolute calibrations. Traceability exists to sources and materials held at Los Alamos National Laboratory for over 35 years. This simple yet powerful approach has ensured consistency of application. Since the 1980's there has been a steady improvement in fuel performance. The trend has been to higher burn up. This requires the use of both higher initial enrichment and greater concentrations of burnable poisons. The original analytical relationships to correct for varying fuel composition are consequently being challenged because the experimental basis for them made use of fuels of lower enrichment and lower poison content than is in use today and is envisioned for use in the near term. Thus a reassessment of the correction factors is needed. Experimental reassessment is expensive and time consuming given the great variation between fuel assemblies in circulation. Fortunately current modeling methods enable relative response functions to be calculated with high accuracy. Hence modeling provides a more convenient and cost effective means to derive correction factors which are fit for purpose with confidence. In this work we use the Monte Carlo code MCNPX with neutron coincidence tallies to calculate the influence of Gd{sub 2}O{sub 3} burnable poison on the measurement of fresh pressurized water reactor fuel. To empirically determine the response function over the range of historical and future use we have considered enrichments up to 5 wt% {sup 235}U/{sup tot}U and Gd weight fractions of up to 10 % Gd/UO{sub 2}. Parameterized correction factors are presented.

  12. Infinitely many solutions of a quasilinear elliptic problem with an oscillatory potential

    SciTech Connect (OSTI)

    Omari, P.; Zanolin, F.

    1996-12-31

    Let {Omega} be a bounded domain in IR{sup N}, with N {ge} 1, having a smooth boundary {partial_derivative}{Omega}. We denote by A the quasilinear elliptic second order differential operator defined by Au+div(a({vert_bar}{del}{sub u}{vert_bar}{sup 2}){del}{sub u}). We suppose that the function a:[O,+{infinity}{r_arrow}O, +{infinity}] is of class C{sup 1} and satisfies the following ellipticity and growth conditions of Leray-Lions type (cf. e.g. [22]): there are constants {gamma}, {Lambda} > O, K {epsilon} [O,1] and p {epsilon}[1, +{infinity}]such that, for every s > O, {lambda}(K + S){sup p-2} {le} a(s{sup 2}){le}{Lambda} (K+S){sup p-2}({lambda}-1/2) a(s){le}a{prime}(s) s {le}{Gamma} a(s). Hence, we can define, for each s {ge} O, the function A(s) = {integral}{sub O}{sup s} a({xi})d{xi}. Let us consider the Dirichlet problem -Au={mu}(x)g(u) + h(x) in {Omega}, u=O on {partial_derivative}{Omega}, where g: IR {r_arrow} IR is continuous and {mu}, h {epsilon} L{sup {infinity}}({infinity}), with {mu}{sub O} = ess inf{sub {Omega}}{sub {mu}} > O. We also set G(s) = {integral}{sub O}{sup s}g({integral})d{integral}, for all s {epsilon} IR. By a solution of (1.3) we mean a function u {epsilon} W{sub O}{sup 1,p} ({Omega}) {intersection} L{sup {infinity}} ({Omega}) such that {integral}{sub {Omega}} a({vert_bar}{del}{sub u}{vert_bar}{sup 2}){del}{sub u}{del}{sub wdx}= {integral}{sub {Omega}} {mu}g(u)wdx + {integral}{sub {Omega}} hwdx, for every w {epsilon} W{sub O}{sup 1,p}({Omega}), where p is the exponent which appears in (1.1). The aim of this paper is to prove the existence of infinitely many solutions of problem (1.3) when the potential G(s) exhibits an oscillatory behaviour at infinity. 22 refs.

  13. CRITICAL CONFIGURATION FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH AND 7-TUBE CLUSTERS)

    SciTech Connect (OSTI)

    Margaret A. Marshall

    2012-05-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first two experiments in the series were evaluated in HEU-COMP-FAST-001 (SCCA-FUND-EXP-001) and HEU-COMP-FAST-002 (SCCA-FUND-EXP-002). The first experiment had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The third set of experiments in the series, performed in mid-1963, which is studied in this evaluation, used beryllium reflectors. The beryllium reflected system was the preferred reactor configuration for this application because of the small thickness of the reflector. The two core configurations had the 253 fuel tubes at a 1.506-cm triangular lattice and arranged in 7-tube clusters. The experiments have been determined to represent acceptable benchmark experiments. Information for this evaluation was compiled from published reports on all three parts of the experimental series (Reference 1-5) and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

  14. Extraction of uranium: comparison of stripping with ammonia vs. strong acid

    SciTech Connect (OSTI)

    Moldovan, B.; Grinbaum, B.; Efraim, A.

    2008-07-01

    Following extraction of uranium in the first stage of solvent extraction using a tertiary amine, typically Alamine 336, the stripping of the extracted uranium is accomplished either by use of an aqueous solution of (NH{sub 4}){sub 2}SO{sub 4} /NH{sub 4}OH or by strong-acid stripping using 400-500 g/L H{sub 2}SO{sub 4}. Both processes have their merits and determine the downstream processing. The classical stripping with ammonia is followed by addition of strong base, to precipitate ammonium uranyl sulfate (NH{sub 4}){sub 2}UO{sub 2}(SO{sub 4}){sub 2}, which yields finally the yellow cake. Conversely, stripping with H{sub 2}SO{sub 4}, followed by oxidation with hydrogen peroxide yields uranyl oxide as product. At the Cameco Key Lake operation, both processes were tested on a pilot scale, using a Bateman Pulsed Column (BPC). The BPC proved to be applicable to both processes. It met the process criteria both for extraction and stripping, leaving less than 1 mg/L of U{sub 3}O{sub 8} in the raffinate, and product solution had the required concentration of U{sub 3}O{sub 8} at high flux and reasonable height of transfer unit. In the Key Lake mill, each operation can be carried out in a single column. The main advantages of the strong-acid stripping over ammonia stripping are: (1) 60% higher flux in the extraction, (2) tenfold higher concentration of the uranium in the product solution, and (3) far more robust process, with no need of pH control in the stripping and no need to add acid to the extraction in order to keep the pH above the point of precipitation of iron compounds. The advantages of the ammoniacal process are easier stripping, that is, less stages needed to reach equilibrium and lower concentration of modifier needed to prevent the creation of a third phase. (authors)

  15. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

  16. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    Jubin, R. T.; Strachan, D. M.; Ilas, G.; Spencer, B. B.; Soelberg, N. R.

    2014-09-01

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is then discussed with respect to what is known in the literature about their behavior in a reprocessing facility. The context for the evaluation in this document is a UO2-based fuel processed through an aqueous-based reprocessing system with a TBP-based solvent extraction chemistry. None of these elements form sufficiently volatile compounds in the context of the reprocessing facility to be of regulatory concern.

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

    SciTech Connect (OSTI)

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

    1984-06-19

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

  18. Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

    SciTech Connect (OSTI)

    Bruce G. Schnitzler

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.

  19. Development of an Immobilisation Technology for Radioactive Waste Solution from Mo-99 Production

    SciTech Connect (OSTI)

    Sizgek, G.D.; Sizgek, E.

    2006-07-01

    Australian Nuclear Science and Technology Organisation (ANSTO) developed a method to immobilize the Intermediate Level Liquid Waste (ILLW) arising from its Mo-99 production process. The immobilisation process involves impregnation of waste solution into ceramic precursor powders, drying, calcining and consolidation (Hot Isostatic Pressing, HIP) to produce final ceramic waste form. Ceramic precursor powder is produced by spray drying of a sol-gel based colloidal dispersion. These free-flowing, microspherical, 20-80 microns, precursors have porosity of 40-50%. An in-house custom designed and manufactured microwave-heated and mechanically fluidized mixer-drier was used for impregnation of the precursor powder with the simulated waste (Depleted Uranyl Nitrate Hexahydrate, DUNH, and inactive Cs, Sr nitrates as fission products) and drying. During impregnation an evaporation rate of 1 l/h water per kW microwave energy in steady state was achieved by matching the feed rate of DUNH to produce equivalent of 35% UO{sub 2} loading. It was demonstrated that the tuned microwave energy can be delivered to the mixer-drier during the entire impregnation process within very low reflection values. The samples of the waste loaded free-flowing powder were subsequently calcined at 750 deg. C under reducing atmosphere for thermal denitration and mineral phase nucleation. Calcined powders were filled into cans. After evacuation and sealing, the cans were isostatically pressed at 1260 deg. C. The consolidated ceramic waste form produced from the DUNH run has been assessed by durability and material characterization tests. Successful confirmation of each processing step at pilot and/or plant scale, has led to the design and construction of the overall process at full scale (equivalent of processing 8 kg U per batch) in a simulated hot-cell mock-up plant. The constructed plant mainly consists of a Microwave-heated Mechanical Fluidized Bed (MWMFB) mixer-drier a fluidized bed calciner, an off gas unit, material transfer/holding and can filling units. Performance of the overall process by integrating each of the processing steps and material transfer operations are currently being tested with inactive simulants from the point of remote operated plant design perspective. Definitive design of a hot-cell production system has been initiated in parallel to the mock-up plant tests. This paper describes the results of both full-scale DUNH impregnation run and inactive mock-up plant tests in developing unique Mo-99 waste immobilisation technology. (authors)

  20. NEAMS update quarterly report for January - March 2012.

    SciTech Connect (OSTI)

    Bradley, K.S.; Hayes, S.; Pointer, D.; Summers, R.; Sadasivan, P.; Sun, X.; Bernholdt, D.; Miller, M.; Stewart, J.

    2012-05-10

    Quarterly highlights are: (1) The integration of Denovo and AMP was demonstrated in an AMP simulation of the thermo-mechanics of a complete fuel assembly; (2) Bison was enhanced with a mechanistic fuel cracking model; (3) Mechanistic algorithms were incorporated into various lower-length-scale models to represent fission gases and dislocations in UO2 fuels; (4) Marmot was improved to allow faster testing of mesoscale models using larger problem domains; (5) Component models of reactor piping were developed for use in Relap-7; (6) The mesh generator of Proteus was updated to accept a mesh specification from Moose and equations were formulated for the intermediate-fidelity Proteus-2D1D module; (7) A new pressure solver was implemented in Nek5000 and demonstrated to work 2.5 times faster than the previous solver; (8) Work continued on volume-holdup models for two fuel reprocessing operations: voloxidation and dissolution; (9) Progress was made on a pyroprocessing model and the characterization of pyroprocessing emission signatures; (10) A new 1D groundwater waste transport code was delivered to the used fuel disposition (UFD) campaign; (11) Efforts on waste form modeling included empirical simulation of sodium-borosilicate glass compositions; (12) The Waste team developed three prototypes for modeling hydride reorientation in fuel cladding during very long-term fuel storage; (13) A benchmark demonstration problem (fission gas bubble growth) was modeled to evaluate the capabilities of different meso-scale numerical methods; (14) Work continued on a hierarchical up-scaling framework to model structural materials by directly coupling dislocation dynamics and crystal plasticity; (15) New 'importance sampling' methods were developed and demonstrated to reduce the computational cost of rare-event inference; (16) The survey and evaluation of existing data and knowledge bases was updated for NE-KAMS; (17) The NEAMS Early User Program was launched; (18) The Nuclear Regulatory Commission (NRC) Office of Regulatory Research was introduced to the NEAMS program; (19) The NEAMS overall software quality assurance plan (SQAP) was revised to version 1.5; and (20) Work continued on NiCE and its plug-ins and other utilities, such as Cubit and VisIt.

  1. Performance of Trasuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Interim Report, Including Void Reactivity Evaluation

    SciTech Connect (OSTI)

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

    2011-03-01

    The current focus of the Deep Burn Project is on once-through burning of transuranice (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 would be 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 calculations have been performed using the DRAGON-4 code in order 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 containing typical UO2 and MOX fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Loading of TRU-only FCM fuel into a pin without significant quantities of uranium challenges the design from the standpoint of several key reactivity parameters, particularly void reactivity, and to some degree, the Doppler coefficient. These unit cells, while providing an indication of how a whole core of similar fuel would behave, also provide information of how individual pins of TRU-only FCM fuel would influence the reactivity behavior of a heterogeneous assembly. If these FCM fuel pins are included in a heterogeneous assembly with LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance of the TRU-only FCM fuel pins may be preserved. A configuration such as this would be similar to CONFU assemblies analyzed in previous studies. Analogous to the plutonium content limits imposed on MOX fuel, some amount of TRU-only FCM pins in an otherwise-uranium fuel assembly may give acceptable reactivity performance. Assembly calculations will be performed in future work to explore the design options for heterogeneous assemblies of this type and their impact on reactivity coefficients.

  2. Rational Ligand Design for U(VI) and Pu(IV)

    SciTech Connect (OSTI)

    Szigethy, Geza

    2009-08-12

    Nuclear power is an attractive alternative to hydrocarbon-based energy production at a time when moving away from carbon-producing processes is widely accepted as a significant developmental need. Hence, the radioactive actinide power sources for this industry are necessarily becoming more widespread, which is accompanied by the increased risk of exposure to both biological and environmental systems. This, in turn, requires the development of technology designed to remove such radioactive threats efficiently and selectively from contaminated material, whether that be contained nuclear waste streams or the human body. Raymond and coworkers (University of California, Berkeley) have for decades investigated the interaction of biologically-inspired, hard Lewis-base ligands with high-valent, early-actinide cations. It has been established that such ligands bind strongly to the hard Lewis-acidic early actinides, and many poly-bidentate ligands have been developed and shown to be effective chelators of actinide contaminants in vivo. Work reported herein explores the effect of ligand geometry on the linear U(IV) dioxo dication (uranyl, UO{sub 2}{sup 2+}). The goal is to utilize rational ligand design to develop ligands that exhibit shape selectivity towards linear dioxo cations and provides thermodynamically favorable binding interactions. The uranyl complexes with a series of tetradentate 3-hydroxy-pyridin-2-one (3,2-HOPO) ligands were studied in both the crystalline state as well as in solution. Despite significant geometric differences, the uranyl affinities of these ligands vary only slightly but are better than DTPA, the only FDA-approved chelation therapy for actinide contamination. The terepthalamide (TAM) moiety was combined into tris-beidentate ligands with 1,2- and 3,2-HOPO moieties were combined into hexadentate ligands whose structural preferences and solution thermodynamics were measured with the uranyl cation. In addition to achieving coordinative saturation, these ligands exhibited increased uranyl affinity compared to bis-Me-3,2-HOPO ligands. This result is due in part to their increased denticity, but is primarily the result of the presence of the TAM moiety. In an effort to explore the relatively unexplored coordination chemistry of Pu(IV) with bidentate moieties, a series of Pu(IV) complexes were also crystallized using bidentate hydroxypyridinone and hydroxypyrone ligands. The geometries of these complexes are compared to that of the analogous Ce(IV) complexes. While in some cases these showed the expected structural similarities, some ligand systems led to significant coordination changes. A series of crystal structure analyses with Ce(IV) indicated that these differences are most likely the result of crystallization condition differences and solvent inclusion effects.

  3. Overview of the U.S. DOE Accident Tolerant Fuel Development Program

    SciTech Connect (OSTI)

    Jon Carmack; Frank Goldner; Shannon M. Bragg-Sitton; Lance L. Snead

    2013-09-01

    The United States Fuel Cycle Research and Development Advanced Fuels Campaign has been given the responsibility to conduct research and development on enhanced accident tolerant fuels with the goal of performing a lead test assembly or lead test rod irradiation in a commercial reactor by 2022. The Advanced Fuels Campaign has defined fuels with enhanced accident tolerance as those that, in comparison with the standard UO2-Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations and operational transients, as well as design-basis and beyond design-basis events. This paper provides an overview of the FCRD Accident Tolerant Fuel program. The ATF attributes will be presented and discussed. Attributes identified as potentially important to enhance accident tolerance include reduced hydrogen generation (resulting from cladding oxidation), enhanced fission product retention under severe accident conditions, reduced cladding reaction with high-temperature steam, and improved fuel-cladding interaction for enhanced performance under extreme conditions. To demonstrate the enhanced accident tolerance of candidate fuel designs, metrics must be developed and evaluated using a combination of design features for a given LWR design, potential improvements to that design, and the design of an advanced fuel/cladding system. The aforementioned attributes provide qualitative guidance for parameters that will be considered for fuels with enhanced accident tolerance. It may be unnecessary to improve in all attributes and it is likely that some attributes or combination of attributes provide meaningful gains in accident tolerance, while others may provide only marginal benefits. Thus, an initial step in program implementation will be the development of quantitative metrics. A companion paper in these proceedings provides an update on the status of establishing these quantitative metrics for accident tolerant LWR fuel.1 The United States FCRD Advanced Fuels Campaign has embarked on an aggressive schedule for development of enhanced accident tolerant LWR fuels. The goal of developing such a fuel system that can be deployed in the U.S. LWR fleet in the next 10 to 20 years supports the sustainability of clean nuclear power generation in the United States.

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

    SciTech Connect (OSTI)

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

    2004-10-06

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

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

    SciTech Connect (OSTI)

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

    2014-10-01

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

  6. THE PENA BLANCA NATURAL ANALOGUE PERFORMANCE ASSESSMENT MODEL

    SciTech Connect (OSTI)

    G.J. Saulnier Jr; W. Statham

    2006-03-10

    The Nopal I uranium mine in the Sierra Pena Blanca, Chihuahua, Mexico serves as a natural analogue to the Yucca Mountain repository. The Pena Blanca Natural Analogue Performance Assessment Model simulates the mobilization and transport of radionuclides that are released from the mine and transported to the saturated zone. the Pena Blanca Natural Analogue Model uses probabilistic simulations of hydrogeologic processes that are analogous to the processes that occur at the Yucca Mountain site. The Nopal I uranium deposit lies in fractured, welded, and altered rhyolitic ash flow tuffs that overlie carbonate rocks, a setting analogous to the geologic formations at the Yucca Mountain site. The Nopal I mine site has the following characteristics as compared to the Yucca Mountain repository site. (1) Analogous source: UO{sub 2} uranium ore deposit = spent nuclear fuel in the repository; (2) Analogous geologic setting: fractured, welded, and altered rhyolitic ash flow tuffs overlying carbonate rocks; (3) Analogous climate: Semiarid to arid; (4) Analogous geochemistry: Oxidizing conditions; and (5) Analogous hydrogeology: The ore deposit lies in the unsaturated zone above the water table. The Nopal I deposit is approximately 8 {+-} 0.5 million years old and has been exposed to oxidizing conditions during the last 3.2 to 3.4 million years. The Pena Blanca Natural Analogue Model considers that the uranium oxide and uranium silicates in the ore deposit were originally analogous to uranium-oxide spent nuclear fuel. The Pena Blanca site has been characterized using field and laboratory investigations of its fault and fracture distribution, mineralogy, fracture fillings, seepage into the mine adits, regional hydrology, and mineralization that shows the extent of radionuclide migration. Three boreholes were drilled at the Nopal I mine site in 2003 and these boreholes have provided samples for lithologic characterization, water-level measurements, and water samples for laboratory analysis of the saturated zone water chemistry. The results of the field investigations and laboratory analyses of rock and water samples collected at Nopal I are used to calibrate the Pena Blanca Natural Analogue Model.

  7. Uranium Transport Modeling

    SciTech Connect (OSTI)

    Bostick, William D.

    2008-01-15

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

  8. MELCOR 1.8.5 modeling aspects of fission product release, transport and deposition an assessment with recommendations.

    SciTech Connect (OSTI)

    Gauntt, Randall O.

    2010-04-01

    The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged light water reactor fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and mixed oxide (MOX) fuels. This paper discusses the synthesis of these findings in the MELCOR severe accident code. Based on recent assessments of MELCOR 1.8.5 fission product release modeling against the Phebus FPT-1 test and on observations from the ISP-46 exercise, modifications to the default MELCOR 1.8.5 release models are recommended. The assessments identified an alternative set of Booth diffusion parameters recommended by ORNL (ORNL-Booth), which produced significantly improved release predictions for cesium and other fission product groups. Some adjustments to the scaling factors in the ORNL-Booth model were made for selected fission product groups, including UO{sub 2}, Mo and Ru in order to obtain better comparisons with the FPT-1 data. The adjusted model, referred to as 'Modified ORNL-Booth,' was subsequently compared to original ORNL VI fission product release experiments and to more recently performed French VERCORS tests, and the comparisons was as favorable or better than the original CORSOR-M MELCOR default release model. These modified ORNL-Booth parameters, input to MELCOR 1.8.5 as 'sensitivity coefficients' (i.e. user input that over-rides the code defaults) are recommended for the interim period until improved release models can be implemented into MELCOR. For the case of ruthenium release in air-oxidizing conditions, some additional modifications to the Ru class vapor pressure are recommended based on estimates of the RuO{sub 2} vapor pressure over mildly hyperstoichiometric UO{sub 2}. The increased vapor pressure for this class significantly increases the net transport of Ru from the fuel to the gas stream. A formal model is needed. Deposition patterns in the Phebus FPT-1 circuit were also significantly improved by using the modified ORNL-Booth parameters, where retention of lower volatile Cs{sub 2}MoO{sub 4} is now predicted in the heated exit regions of the FPT-1 test, bringing down depositions in the FPT-1 steam generator tube to be in closer alignment with the experimental data. This improvement in 'RCS' deposition behavior preserves the overall correct release of cesium to the containment that was observed even with the default CORSOR-M model. Not correctly treated however is the release and transport of Ag to the FPT-1 containment. A model for Ag release from control rods is presently not available in MELCOR. Lack of this model is thought to be responsible for the underprediction by a factor of two of the total aerosol mass to the FPT-1 containment. It is suggested that this underprediction of airborne mass led to an underprediction of the aerosol agglomeration rate. Underprediction of the agglomeration rate leads to low predictions of the aerosol particle size in comparison to experimentally measured ones. Small particle size leads low predictions of the gravitational settling rate relative to the experimental data. This error, however, is a conservative one in that too-low settling rate would result in a larger source term to the environment. Implementation of an interim Ag release model is currently under study. In the course of this assessment, a review of MELCOR release models was performed and led to the identification of several areas for future improvements to MELCOR. These include upgrading the Booth release model to account for changes in local oxidizing/reducing conditions and including a fuel oxidation model to accommodate effects of fuel stoichiometry. Models such as implemented in the French ELSA code and described by Lewis are considered appropriate for MELCOR. A model for ruthenium release under air oxidizing conditions is also needed and should be included as part of a fuel oxidation model since fuel stoichiometry is a fundamen

  9. NEAMS Update. Quarterly Report for October - December 2011.

    SciTech Connect (OSTI)

    Bradley, K.

    2012-02-16

    The Advanced Modeling and Simulation Office within the DOE Office of Nuclear Energy (NE) has been charged with revolutionizing the design tools used to build nuclear power plants during the next 10 years. To accomplish this, the DOE has brought together the national laboratories, U.S. universities, and the nuclear energy industry to establish the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program. The mission of NEAMS is to modernize computer modeling of nuclear energy systems and improve the fidelity and validity of modeling results using contemporary software environments and high-performance computers. NEAMS will create a set of engineering-level codes aimed at designing and analyzing the performance and safety of nuclear power plants and reactor fuels. The truly predictive nature of these codes will be achieved by modeling the governing phenomena at the spatial and temporal scales that dominate the behavior. These codes will be executed within a simulation environment that orchestrates code integration with respect to spatial meshing, computational resources, and execution to give the user a common 'look and feel' for setting up problems and displaying results. NEAMS is building upon a suite of existing simulation tools, including those developed by the federal Scientific Discovery through Advanced Computing and Advanced Simulation and Computing programs. NEAMS also draws upon existing simulation tools for materials and nuclear systems, although many of these are limited in terms of scale, applicability, and portability (their ability to be integrated into contemporary software and hardware architectures). NEAMS investments have directly and indirectly supported additional NE research and development programs, including those devoted to waste repositories, safeguarded separations systems, and long-term storage of used nuclear fuel. NEAMS is organized into two broad efforts, each comprising four elements. The quarterly highlights October-December 2011 are: (1) Version 1.0 of AMP, the fuel assembly performance code, was tested on the JAGUAR supercomputer and released on November 1, 2011, a detailed discussion of this new simulation tool is given; (2) A coolant sub-channel model and a preliminary UO{sub 2} smeared-cracking model were implemented in BISON, the single-pin fuel code, more information on how these models were developed and benchmarked is given; (3) The Object Kinetic Monte Carlo model was implemented to account for nucleation events in meso-scale simulations and a discussion of the significance of this advance is given; (4) The SHARP neutronics module, PROTEUS, was expanded to be applicable to all types of reactors, and a discussion of the importance of PROTEUS is given; (5) A plan has been finalized for integrating the high-fidelity, three-dimensional reactor code SHARP with both the systems-level code RELAP7 and the fuel assembly code AMP. This is a new initiative; (6) Work began to evaluate the applicability of AMP to the problem of dry storage of used fuel and to define a relevant problem to test the applicability; (7) A code to obtain phonon spectra from the force-constant matrix for a crystalline lattice has been completed. This important bridge between subcontinuum and continuum phenomena is discussed; (8) Benchmarking was begun on the meso-scale, finite-element fuels code MARMOT to validate its new variable splitting algorithm; (9) A very computationally demanding simulation of diffusion-driven nucleation of new microstructural features has been completed. An explanation of the difficulty of this simulation is given; (10) Experiments were conducted with deformed steel to validate a crystal plasticity finite-element code for bodycentered cubic iron; (11) The Capability Transfer Roadmap was completed and published as an internal laboratory technical report; (12) The AMP fuel assembly code input generator was integrated into the NEAMS Integrated Computational Environment (NiCE). More details on the planned NEAMS computing environment is given; and (13) The NEAMS program website (neams.ener

  10. Characterization of cores from an in-situ recovery mined uranium deposit in Wyoming: Implications for post-mining restoration

    SciTech Connect (OSTI)

    WoldeGabriel, G.; Boukhalfa, H.; Ware, S. D.; Cheshire, M.; Reimus, P.; Heikoop, J.; Conradson, S. D.; Batuk, O.; Havrilla, G.; House, B.; Simmons, A.; Clay, J.; Basu, A.; Christensen, J. N.; Brown, S. T.; DePaolo, D. J.

    2014-10-08

    In-situ recovery (ISR) of uranium (U) from sandstone-type roll-front deposits is a technology that involves the injection of solutions that consist of ground water fortified with oxygen and carbonate to promote the oxidative dissolution of U, which is pumped to recovery facilities located at the surface that capture the dissolved U and recycle the treated water. The ISR process alters the geochemical conditions in the subsurface creating conditions that are more favorable to the migration of uranium and other metals associated with the uranium deposit. There is a lack of clear understanding of the impact of ISR mining on the aquifer and host rocks of the post-mined site and the fate of residual U and other metals within the mined ore zone. We performed detailed petrographic, mineralogical, and geochemical analyses of several samples taken from about 7 m of core of the formerly the ISR-mined Smith Ranch–Highland uranium deposit in Wyoming. We show that previously mined cores contain significant residual uranium (U) present as coatings on pyrite and carbonaceous fragments. Coffinite was identified in three samples. Core samples with higher organic (> 1 wt.%) and clay (> 6–17 wt.%) contents yielded higher 234U/238U activity ratios (1.0–1.48) than those with lower organic and clay fractions. The ISR mining was inefficient in mobilizing U from the carbonaceous materials, which retained considerable U concentrations (374–11,534 ppm). This is in contrast with the deeper part of the ore zone, which was highly depleted in U and had very low 234U/238U activity ratios. This probably is due to greater contact with the lixiviant (leaching solution) during ISR mining. EXAFS analyses performed on grains with the highest U and Fe concentrations reveal that Fe is present in a reduced form as pyrite and U occurs mostly as U(IV) complexed by organic matter or as U(IV) phases of carbonate complexes. Moreover, U–O distances of ~ 2.05 Å were noted, indicating the potential formation of other poorly defined U(IV/VI) species. We also noted a small contribution from Udouble bond; length as m-dashO at 1.79 Å, which indicates that U is partially oxidized. There is no apparent U–S or U–Fe interaction in any of the U spectra analyzed. However, SEM analysis of thin sections prepared from the same core material reveals surficial U associated with pyrite which is probably a minor fraction of the total U present as thin coatings on the surface of pyrite. Our data show the presence of different structurally variable uranium forms associated with the mined cores. U associated with carbonaceous materials is probably from the original U mobilization that accumulated in the organic matter-rich areas under reducing conditions during shallow burial diagenesis. U associated with pyrite represents a small fraction of the total U and was likely deposited as a result of chemical reduction by pyrite. Our data suggest that areas rich in carbonaceous materials had limited exposure to the lixiviant solution, continue to be reducing, and still hold significant U resources. Because of their limited access to fluid flow, these areas might not contribute significantly to post-mining U release or attenuation. Areas with pyrite that are accessible to fluids seem to be more reactive and could act as reductants and facilitate U reduction and accumulation, limiting its migration.

  11. Methodology for Developing the REScheckTM Software through Version 4.4.3

    SciTech Connect (OSTI)

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

    2012-09-01

    The Energy Policy Act of 1992 (EPAct, Public Law 102-486) establishes the 1992 Model Energy Code (MEC), published by the Council of American Building Officials (CABO), as the target for several energy-related requirements for residential buildings (CABO 1992). The U.S. Department of Housing and Urban Development (HUD) and the U.S. Department of Agriculture (via Rural Economic and Community Development [RECD] [formerly Farmers Home Administration]) are required to establish standards for government-assisted housing that “meet or exceed the requirements of the Council of American Building Officials Model Energy Code, 1992.” CABO issued 1992, 1993, and 1995 editions of the MEC (CABO 1992, 1993, and 1995). Effective December 4, 1995, CABO assigned all rights and responsibilities for the MEC to the International Code Council (ICC). The first edition of the ICC’s International Energy Conservation Code (ICC 1998) issued in 1998 therefore replaced the 1995 edition of the MEC. The 1998 IECC incorporates the provisions of the 1995 MEC and includes the technical content of the MEC as modified by approved changes from the 1995, 1996, and 1997 code development cycles. The ICC subsequently issued the 2000 edition of the IECC (ICC 1999). Many states and local jurisdictions have adopted one edition of the MEC or IECC as the basis for their energy code. In a Federal Register notice issued January 10, 2001 (FR Vol. 99, No. 7, page 1964), the U.S. Department of Energy (DOE) concluded that the 1998 and 2000 editions of the IECC improve energy efficiency over the 1995 MEC. DOE has previously issued notices that the 1993 and 1995 MEC also improved energy efficiency compared to the preceding editions. To help builders comply with the MEC and IECC requirements, and to help HUD, RECD, and state and local code officials enforce these code requirements, DOE tasked Pacific Northwest National Laboratory (PNNL) with developing the MECcheck™ compliance materials. In November 2002, MECcheck was renamed REScheck™ to better identify it as a residential code compliance tool. The “MEC” in MECcheck was outdated because it was taken from the Model Energy Code, which has been succeeded by the IECC. The “RES” in REScheck is also a better fit with the companion commercial product, COMcheck™. The easy-to-use REScheck compliance materials include a compliance and enforcement manual for all the MEC and IECC requirements and three compliance approaches for meeting the code’s thermal envelope requirements?prescriptive packages, software, and a trade-off worksheet (included in the compliance manual). The compliance materials can be used for single-family and low-rise multifamily dwellings. The materials allow building energy efficiency measures (such as insulation levels) to be “traded off” against each other, allowing a wide variety of building designs to comply with the code. This report explains the methodology used to develop Version 4.4.3 of the REScheck software developed for the 1992, 1993, and 1995 editions of the MEC, and the 1998, 2000, 2003, 2006, 2007, 2009, and 2012 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 editions is similar. Beginning with REScheck Version 4.4.0, support for 1992, 1993, and 1995 MEC and the 1998 IECC is no longer included, but those sections remain in this document for reference purposes. 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.

  12. 300 Area Uranium Stabilization Through Polyphosphate Injection: Final Report

    SciTech Connect (OSTI)

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

    2009-06-30

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

  13. Redwing: A MOOSE application for coupling MPACT and BISON

    SciTech Connect (OSTI)

    Frederick N. Gleicher; Michael Rose; Tom Downar

    2014-11-01

    Fuel performance and whole core neutron transport programs are often used to analyze fuel behavior as it is depleted in a reactor. For fuel performance programs, internal models provide the local intra-pin power density, fast neutron flux, burnup, and fission rate density, which are needed for a fuel performance analysis. The fuel performance internal models have a number of limitations. These include effects on the intra-pin power distribution by nearby assembly elements, such as water channels and control rods, and the further limitation of applicability to a specified fuel type such as low enriched UO2. In addition, whole core neutron transport codes need an accurate intra-pin temperature distribution in order to calculate neutron cross sections. Fuel performance simulations are able to model the intra-pin fuel displacement as the fuel expands and densifies. These displacements must be accurately modeled in order to capture the eventual mechanical contact of the fuel and the clad; the correct radial gap width is needed for an accurate calculation of the temperature distribution of the fuel rod. Redwing is a MOOSE-based application that enables coupling between MPACT and BISON for transport and fuel performance coupling. MPACT is a 3D neutron transport and reactor core simulator based on the method of characteristics (MOC). The development of MPACT began at the University of Michigan (UM) and now is under the joint development of ORNL and UM as part of the DOE CASL Simulation Hub. MPACT is able to model the effects of local assembly elements and is able calculate intra-pin quantities such as the local power density on a volumetric mesh for any fuel type. BISON is a fuel performance application of Multi-physics Object Oriented Simulation Environment (MOOSE), which is under development at Idaho National Laboratory. BISON is able to solve the nonlinearly coupled mechanical deformation and heat transfer finite element equations that model a fuel element as it is depleted in a nuclear reactor. Redwing couples BISON and MPACT in a single application. Redwing maps and transfers the individual intra-pin quantities such as fission rate density, power density, and fast neutron flux from the MPACT volumetric mesh to the individual BISON finite element meshes. For a two-way coupling Redwing maps and transfers the individual pin temperature field and axially dependent coolant densities from the BISON mesh to the MPACT volumetric mesh. Details of the mapping are given. Redwing advances the simulation with the MPACT solution for each depletion time step and then advances the multiple BISON simulations for fuel performance calculations. Sub-cycle advancement can be applied to the individual BISON simulations and allows multiple time steps to be applied to the fuel performance simulations. Currently, only loose coupling where data from a previous time step is applied to the current time step is performed.

  14. Fate of Contaminants in Contact with West Valley Grouts

    SciTech Connect (OSTI)

    Fuhrmann,M.; Gillow, J.

    2009-07-01

    The objective of the work described here is to determine to what extent a variety of contaminants, including fission products, actinides, and RCRA elements are sequestered by the two grout formulations. The conceptual model for this study is as follows: a large mass of grout having been poured into a high-level waste tank is in the process of aging and weathering for thousands of years. The waste remaining in the tank will contain radionuclides and other contaminants, much of which will adhere to tank walls and internal structures. The grout will encapsulate the contaminants. Initially the grout will be well sequestered, but over time rainwater and groundwater will gain access to it. Ultimately, the grout/waste environment will be an open system. In this condition water will move through the grout, exposing it to O{sub 2} and CO{sub 2} from the air and HCO{sub 3}{sup -} from the groundwater. Thus we are considering an oxic environment containing HCO{sub 3}{sup -}. Initially the solubility of many contaminants, but not all, will be constrained by chemistry dominated by the grout, primarily by the high pH, around 11.8. This is controlled and buffered by the portland cement and blast furnace slag components of the grout, which by themselves maintain a solution pH of about 12.5. Slowly the pH will diminish as Ca(OH){sub 2} and KOH dissolve, are carried away by water, and CaCO{sub 3} forms. As these conditions develop, the behavior of these elements comes into question. In our conceptual model, although the grout is formulated to provide some reducing capacity, in order to be conservative this mechanism is not considered. In addition to solubility constraints imposed by pH, the various contaminants may be incorporated into a variety of solid phases. Some may be incorporated into newly forming compounds as the grout sets and cures. Others (like soddyite, (UO{sub 2}){sub 2}SiO{sub 4}(H{sub 2}O){sub 2}) are the result of slower reactions but may become important over time as contaminants are exposed to evolving chemistry in the grout. Still other solid phases may form from reactions between the waste and grout components, not only the cementitious materials, but also the additives used in the grout. Another process that may exert some control on contaminant concentrations is adsorption onto solids within the grout. These may be additives such as the fluorapatite or zeolite that are substantial percentages of the grouts or they may be minerals, typically Ca-Al-Si materials, that form in the grout system as cement sets. In addition, as the grout weathers over time, CaCO{sub 3} minerals, such as calcite and aragonite, will form as a rind on the grout and as a fracture filling mineral. Some contaminants are likely to be incorporated into these minerals, to a greater or lesser extent, as they precipitate. For some elements, such as U, there is a significant literature exploring the incorporation into CaCO{sub 3}, but for others there is essentially no information. This is also the case for much of the chemical regime of the grouts. Initial conditions are at pH values around 12 and information is often sparse.

  15. Promoting Uranium Immobilization by the Activities of Microbial Phosphatases

    SciTech Connect (OSTI)

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

    2007-04-19

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

  16. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    SciTech Connect (OSTI)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and Commercialization. The activities performed during the feasibility assessment phase include laboratory scale experiments; fuel performance code updates; and analytical assessment of economic, operational, safety, fuel cycle, and environmental impacts of the new concepts. The development and qualification stage will consist of fuel fabrication and large scale irradiation and safety basis testing, leading to qualification and ultimate NRC licensing of the new fuel. The commercialization phase initiates technology transfer to industry for implementation. Attributes for fuels with enhanced accident tolerance include improved reaction kinetics with steam and slower hydrogen generation rate, while maintaining acceptable cladding thermo-mechanical properties; fuel thermo-mechanical properties; fuel-clad interactions; and fission-product behavior. These attributes provide a qualitative guidance for parameters that must be considered in the development of fuels and cladding with enhanced accident tolerance. However, quantitative metrics must be developed for these attributes. To initiate the quantitative metrics development, a Light Water Reactor Enhanced Accident Tolerant Fuels Metrics Development Workshop was held October 10-11, 2012, in Germantown, Maryland. This document summarizes the structure and outcome of the two-day workshop. Questions regarding the content can be directed to Lori Braase, 208-526-7763, lori.braase@inl.gov.

  17. Integrated Biomass Gasification with Catalytic Partial Oxidation for Selective Tar Conversion

    SciTech Connect (OSTI)

    Zhang, Lingzhi; Wei, Wei; Manke, Jeff; Vazquez, Arturo; Thompson, Jeff; Thompson, Mark

    2011-05-28

    Biomass gasification is a flexible and efficient way of utilizing widely available domestic renewable resources. Syngas from biomass has the potential for biofuels production, which will enhance energy security and environmental benefits. Additionally, with the successful development of low Btu fuel engines (e.g. GE Jenbacher engines), syngas from biomass can be efficiently used for power/heat co-generation. However, biomass gasification has not been widely commercialized because of a number of technical/economic issues related to gasifier design and syngas cleanup. Biomass gasification, due to its scale limitation, cannot afford to use pure oxygen as the gasification agent that used in coal gasification. Because, it uses air instead of oxygen, the biomass gasification temperature is much lower than well-understood coal gasification. The low temperature leads to a lot of tar formation and the tar can gum up the downstream equipment. Thus, the biomass gasification tar removal is a critical technology challenge for all types of biomass gasifiers. This USDA/DOE funded program (award number: DE-FG36-O8GO18085) aims to develop an advanced catalytic tar conversion system that can economically and efficiently convert tar into useful light gases (such as syngas) for downstream fuel synthesis or power generation. This program has been executed by GE Global Research in Irvine, CA, in collaboration with Professor Lanny Schmidt's group at the University of Minnesota (UoMn). Biomass gasification produces a raw syngas stream containing H2, CO, CO2, H2O, CH4 and other hydrocarbons, tars, char, and ash. Tars are defined as organic compounds that are condensable at room temperature and are assumed to be largely aromatic. Downstream units in biomass gasification such as gas engine, turbine or fuel synthesis reactors require stringent control in syngas quality, especially tar content to avoid plugging (gum) of downstream equipment. Tar- and ash-free syngas streams are a critical requirement for commercial deployment of biomass-based power/heat co-generation and biofuels production. There are several commonly used syngas clean-up technologies: (1) Syngas cooling and water scrubbing has been commercially proven but efficiency is low and it is only effective at small scales. This route is accompanied with troublesome wastewater treatment. (2) The tar filtration method requires frequent filter replacement and solid residue treatment, leading to high operation and capital costs. (3) Thermal destruction typically operates at temperatures higher than 1000oC. It has slow kinetics and potential soot formation issues. The system is expensive and materials are not reliable at high temperatures. (4) In-bed cracking catalysts show rapid deactivation, with durability to be demonstrated. (5) External catalytic cracking or steam reforming has low thermal efficiency and is faced with problematic catalyst coking. Under this program, catalytic partial oxidation (CPO) is being evaluated for syngas tar clean-up in biomass gasification. The CPO reaction is exothermic, implying that no external heat is needed and the system is of high thermal efficiency. CPO is capable of processing large gas volume, indicating a very compact catalyst bed and a low reactor cost. Instead of traditional physical removal of tar, the CPO concept converts tar into useful light gases (eg. CO, H2, CH4). This eliminates waste treatment and disposal requirements. All those advantages make the CPO catalytic tar conversion system a viable solution for biomass gasification downstream gas clean-up. This program was conducted from October 1 2008 to February 28 2011 and divided into five major tasks. - Task A: Perform conceptual design and conduct preliminary system and economic analysis (Q1 2009 ~ Q2 2009) - Task B: Biomass gasification tests, product characterization, and CPO tar conversion catalyst preparation. This task will be conducted after completing process design and system economics analysis. Major milestones include identification of syngas cleaning requirements for proposed system

  18. Material Property Correlations: Comparisons between FRAPCON-3.4, FRAPTRAN 1.4, and MATPRO

    SciTech Connect (OSTI)

    Luscher, Walter G.; Geelhood, Kenneth J.

    2010-08-01

    The U.S. Nuclear Regulatory Commission (NRC) uses the computer codes FRAPCON-3 and FRAPTRAN to model steady state and transient fuel behavior, respectively, in regulatory analysis. In order to effectively model fuel behavior, material property correlations must be used for a wide range of operating conditions (e.g. temperature and burnup). In this sense, a 'material property' is a physical characteristic of the material whose quantitative value is necessary in the analysis process. Further, the property may be used to compare the benefits of one material versus another. Generally speaking, the material properties of interest in regulatory analysis of nuclear fuel behavior are mechanical or thermodynamic in nature. The issue of what is and is not a 'material property' will never be universally resolved. In this report, properties such as thermal conductivity are included. Other characteristics of the material (e.g. fission gas release) are considered 'models' rather than properties, and are discussed elsewhere. Still others (e.g., neutron absorption cross-section) are simply not required in this specific analysis. The material property correlations for the FRAPCON-3 and FRAPTRAN computer codes were documented in NUREG/CR-6534 and NUREG/CR-6739, respectively. Some of these have been modified or updated since the original code documentation was published. The primary purpose of this report is to consolidate the current material property correlations used in FRAPCON-3 and FRAPTRAN into a single document. Material property correlations for oxide fuels, including uranium dioxide (UO2) and mixed oxide (MOX) fuels, are described in Section 2. Throughout this document, the term MOX will be used to describe fuels that are blends of uranium and plutonium oxides, (U,Pu)O2. The properties for uranium dioxide with other additives (e.g., gadolinia) are also discussed. Material property correlations for cladding materials and gases are described in Sections 3 and 4, respectively. In addition to describing the material property correlations used in the subroutines of FRAPCON-3 and FRAPTRAN, this report also provides a variety of comparisons between material property correlations and data. Although they are frequently identical, comparisons are made between the material property correlations used in the FRAPCON-3 and FRAPTRAN codes. Comparisons are also made between the material property correlations used in MATPRO, a compilation of fuel and cladding material property correlations with an extensive history of used with various fuel performance and severe accident codes. For a number of reasons, consistency between the material property correlations in FRAPCON-3, FRAPTRAN, and MATPRO has never been complete. However, the current versions of FRAPCON-3 and FRAPTRAN use a relatively consistent set of correlations for the properties that are used by both codes. The material property correlations in the most recent version of MATPRO are documented in Volume 4 of NUREG/CR-6150. In addition to comparison of the various correlations, correlation-to-data comparisons are also made with FRAPCON-3, FRAPTRAN, and MATPRO. All comparisons made in this report are based on the material property correlations used in the most recent version of the FRAPCON-3 and FRAPTRAN codes, FRAPCON-3.4 and FRAPTRAN 1.4. The source code for each material property correlation discussed will be provided for FRAPCON-3.4 and FRAPTRAN 1.4 (see appendix) as well as a range of applicability and an estimate of uncertainty where possible.

  19. Advanced Aqueous Separation Systems for Actinide Partitioning

    SciTech Connect (OSTI)

    Nash, Ken; Martin, Leigh; Lumetta, Gregg

    2015-04-02

    One of the most challenging aspects of advanced processing of used nuclear fuel is the separation of transplutonium actinides from fission product lanthanides. This separation is essential if actinide transmutation options are to be pursued in advanced fuel cycles, as lanthanides compete with actinides for neutrons in both thermal and fast reactors, thus limiting efficiency. The separation is difficult because the chemistry of Am3+ and Cm3+ is nearly identical to that of the trivalent lanthanides (Ln3+). The prior literature teaches that two approaches offer the greatest probability of devising a successful group separation process based on aqueous processes: 1) the application of complexing agents containing ligand donor atoms that are softer than oxygen (N, S, Cl-) or 2) changing the oxidation state of Am to the IV, V, or VI state to increase the essential differences between Am and lanthanide chemistry (an approach utilized in the PUREX process to selectively remove Pu4+ and UO22+ from fission products). The latter approach offers the additional benefit of enabling a separation of Am from Cm, as Cm(III) is resistant to oxidation and so can easily be made to follow the lanthanides. The fundamental limitations of these approaches are that 1) the soft(er) donor atoms that interact more strongly with actinide cations than lanthanides form substantially weaker bonds than oxygen atoms, thus necessitating modification of extraction conditions for adequate phase transfer efficiency, 2) soft donor reagents have been seen to suffer slow phase transfer kinetics and hydro-/radiolytic stability limitations and 3) the upper oxidation states of Am are all moderately strong oxidants, hence of only transient stability in media representative of conventional aqueous separations systems. There are examples in the literature of both approaches having been described. However, it is not clear at present that any extant process is sufficiently robust for application at the scale necessary for commercial fuel processing supporting transmutation of transplutonium elements. This research project continued basic themes investigated by this research group during the past decade. In the Fuel Cycle Research and Development program at DOE, the current favorite process for accomplishing the separation of trivalent actinides from fission product lanthanides is the TALSPEAK process. TALSPEAK is a solvent extraction method (developed at Oak Ridge National Lab in the 1960s) based on the combination of a cation exchanging extractant (e.g., HDEHP), an actinide-selective aminopolycarboxylate complexing agent (e.g., DTPA), and a carboxylic acid buffer to control pH in the range of 3-4. Considerable effort has been expended in this research group during the past 8 years to elaborate the details of TALSPEAK in the interest of developing improved approaches to the operation of TALSPEAK-like systems. In this project we focused on defining aggregation phenomena in conventional TALSPEAK separations, on supporting the development of Advanced TALSPEAK processes, on profiling the aqueous complexation kinetics of lanthanides in TALSPEAK relevant aqueous media, on the design of new diglycolamide and N-donor extractants, and on characterizing cation-cation complexes of pentavalent actinides.

  20. Enhancing BWR Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect (OSTI)

    Gray S. Chang

    2008-07-01

    Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu /Pu. For future advanced nuclear systems, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. As a result, 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. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm3) to the top (0.35 g/cm3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. We concluded that the concept of MARA, which involves the use of transuranic nuclides (237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy rennaissance.

  1. Advanced Fuels for LWRs: Fully-Ceramic Microencapsulated and Related Concepts FY 2012 Interim Report

    SciTech Connect (OSTI)

    R. Sonat Sen; Brian Boer; John D. Bess; Michael A. Pope; Abderrafi M. Ougouag

    2012-03-01

    This report summarizes the progress in the Deep Burn project at Idaho National Laboratory during the first half of fiscal year 2012 (FY2012). The current focus of this work is on Fully-Ceramic Microencapsulated (FCM) fuel containing low-enriched uranium (LEU) uranium nitride (UN) fuel kernels. UO2 fuel kernels have not been ruled out, and will be examined as later work in FY2012. Reactor physics calculations confirmed that the FCM fuel containing 500 mm diameter kernels of UN fuel has positive MTC with a conventional fuel pellet radius of 4.1 mm. The methodology was put into place and validated against MCNP to perform whole-core calculations using DONJON, which can interpolate cross sections from a library generated using DRAGON. Comparisons to MCNP were performed on the whole core to confirm the accuracy of the DRAGON/DONJON schemes. A thermal fluid coupling scheme was also developed and implemented with DONJON. This is currently able to iterate between diffusion calculations and thermal fluid calculations in order to update fuel temperatures and cross sections in whole-core calculations. Now that the DRAGON/DONJON calculation capability is in place and has been validated against MCNP results, and a thermal-hydraulic capability has been implemented in the DONJON methodology, the work will proceed to more realistic reactor calculations. MTC calculations at the lattice level without the correct burnable poison are inadequate to guarantee zero or negative values in a realistic mode of operation. Using the DONJON calculation methodology described in this report, a startup core with enrichment zoning and burnable poisons will be designed. Larger fuel pins will be evaluated for their ability to (1) alleviate the problem of positive MTC and (2) increase reactivity-limited burnup. Once the critical boron concentration of the startup core is determined, MTC will be calculated to verify a non-positive value. If the value is positive, the design will be changed to require less soluble boron by, for example, increasing the reactivity hold-down by burnable poisons. Then, the whole core analysis will be repeated until an acceptable design is found. Calculations of departure from nucleate boiling ratio (DNBR) will be included in the safety evaluation as well. Once a startup core is shown to be viable, subsequent reloads will be simulated by shuffling fuel and introducing fresh fuel. The PASTA code has been updated with material properties of UN fuel from literature and a model for the diffusion and release of volatile fission products from the SiC matrix material . Preliminary simulations have been performed for both normal conditions and elevated temperatures. These results indicated that the fuel performs well and that the SiC matrix has a good retention of the fission products. The path forward for fuel performance work includes improvement of metallic fission product release from the kernel. Results should be considered preliminary and further validation is required.

  2. Multi-Application Small Light Water Reactor Final Report

    SciTech Connect (OSTI)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as cogeneration, water desalination or district heating were not addressed directly in the economic analyses since these depend more on local conditions, demand and economy and can not be easily generalized. Current economic performance experience and available cost data were used. The preliminary cost estimate, based on a concept that could be deployed in less than a decade, is: (1) Net Electrical Output--1050 MWe; (2) Net Station Efficiency--23%; (3) Number of Power Units--30; (4) Nominal Plant Capacity Factor--95%; (5) Total capital cost--$1241/kWe; and (6) Total busbar cost--3.4 cents/kWh. The project includes a testing program that has been conducted at Oregon State University (OSU). The test facility is a 1/3-height and 1/254.7 volume scaled design that will operate at full system pressure and temperature, and will be capable of operation at 600 kW. The design and construction of the facility have been completed. Testing is scheduled to begin in October 2002. The MASLWR conceptual design is simple, safe, and economical. It operates at NSSS parameters much lower than for a typical PWR plant, and has a much simplified power generation system. The individual reactor modules can be operated as on/off units, thereby limiting operational transients to startup and shutdown. In addition, a plant can be built in increments that match demand increases. The ''pull and replace'' concept offers automation of refueling and maintenance activities. Performing refueling in a single location improves proliferation resistance and eliminates the threat of diversion. Design certification based on testing is simplified because of the relatively low cost of a full-scale prototype facility. The overall conclusion is that while the efficiency of the power generation unit is much lower (23% versus 30%), the reduction in capital cost due to simplification of design more than makes up for the increased cost of nuclear fuel. The design concept complies with the safety requirements and criteria. It also satisfies the goals for modularity, standard plant design, certification before construction, c

  3. Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

    SciTech Connect (OSTI)

    B. Boer; R. S. Sen; M. A. Pope; A. M. Ougouag

    2011-09-01

    The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO{sub 2} kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 {micro}mm kernel diameter, 100 {micro}mm buffer, 35 {micro}mm IPyC, 35 {micro}mm SiC, 40 {micro}mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, results in a 8.0 x 10{sup -2} failure probability. For a 'best-estimate' FGR fraction of 50% and a more modest burnup target level of 500 MWd/kg ,the failure probability drops below 2.0 x 10{sup -5}, the typical performance of TRISO fuel made under the German HTR research program. An optimization study on particle design shows improved performance if the buffer size is increased from 100 to 120 {micro}mm while reducing the OPyC layer. The presence of the latter layer does not provide much benefit at high burnup levels (and fast fluence levels). Normally the shrinkage of the OPyC would result in a beneficial compressive force on the SiC coating. However, at high fluence levels the shrinkage is expected to turn into swelling, resulting in the opposite effect. However, this situation is different when the SiC-matrix, in which the particles are embedded, is also considered: the OPyC swelling can result in a beneficial compressive force on the SiC coating since outward displacement of the OPyC outer surface is inhibited by the presence of the also-swelling SiC matrix. Taking some credit for this effect by adopting a 5 {micro}mm SiC-matrix layer, the optimized particle (100 {micro}mm buffer and 10 {micro}mm OPyC), gives a failure probability of 1.9 x 10{sup -4} for conservative conditions. During a LOCA transient, assuming core re-flood in 30 seconds, the temperature of the coated particle can be expected to be about 200K higher than nominal temperature (900K). For this event the particle failure fraction for a conservative case is 1.0 x 10{sup -2}, for the optimized particle design. For a FGR-fraction of 50% this value reduces to 6.4 x 10{sup -4}.