National Library of Energy BETA

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  1. UO CLASSES FOR HIGH SCHOOL STUDENTS GENERAL DESCRIPTION

    E-Print Network [OSTI]

    Cina, Jeff

    that one of the UOCHSS instructors, Chris Doe, won the 4J School District's "Champion in Education" Award, UO Department of English Jane Cramer, UO Department of Political Science Chris Doe, UO Department of Psychology Hillary Nadeau, UO School of Education Clinton Sandvick, UO Department of History Nancy Shurtz, UO

  2. UO CLASSES FOR HIGH SCHOOL STUDENTS GENERAL DESCRIPTION

    E-Print Network [OSTI]

    Cina, Jeff

    that one of the UOCHSS instructors, Chris Doe, won the 4J School District's "Champion in Education" Award Department of Political Science Chris Doe, UO Department of Biology Michael Dreiling, UO Department of Education Clinton Sandvick, UO Department of History Nancy Shurtz, UO School of Law Merle Weiner, UO School

  3. UO CLASSES FOR HIGH SCHOOL STUDENTS GENERAL DESCRIPTION

    E-Print Network [OSTI]

    Cina, Jeff

    that one of the UOCHSS instructors, Chris Doe, won the 4J School District's "Champion in Education" Award Department of Political Science Chris Doe, UO Department of Biology Michael Dreiling, UO Department of Psychology Hillary Nadeau, UO School of Education Clinton Sandvick, UO Department of History COURSE SCHEDULE

  4. 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

  5. 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.

  6. 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.

  7. The UO Baker Downtown Center Getting Here and Parking

    E-Print Network [OSTI]

    ). Parking is in the lot on your left. · Fromtheeast: If on I-105 westbound, take Exit #2 and follow signs the UO Baker Downtown Center (on the corner of 10th and High). Parking is in the lot on your left Baker Downtown Center (on the corner of 10th and High). Parking is in the lot on your left. PARKING

  8. UO{sub 3} plant turnover - facility description document

    SciTech Connect (OSTI)

    Clapp, D.A.

    1995-01-01

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

  9. Vendor Control UoW 1730 (Rev. 10/07)

    E-Print Network [OSTI]

    Brown, Sally

    Vendor Control Use Only UoW 1730 (Rev. 10/07) ACCOUNTING DETAIL U.S. Taxpayer ID Number 1. Vendor 6109 requires most recipients of payments for services performed to give taxpayer identification generally withhold taxes from taxable payments to a payee who does not furnish a taxpayer identification

  10. 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.

  11. In Situ TEM Observation of Dislocation Evolutionin Polycrystalline UO2

    SciTech Connect (OSTI)

    L. F. HE; 1 M. A. KIRK; Argonne National Laboratory; J. Gan; T. R. ALLEN

    2014-10-01

    In situ transmission electron microscopy observation of polycrystalline UO2 (with average grain size of about 5 lm) irradiated with Kr ions at 600C and 800C was conducted to understand the radiation-induced dislocation evolution under the influence of grain boundaries. The dislocation evolution in the grain interior of polycrystalline UO2 was similar under Kr irradiation at different ion energies and temperatures. As expected, it was characterized by the nucleation and growth of dislocation loops at low irradiation doses, followed by transformation to extended dislocation lines and tangles at high doses. For the first time, a dislocation-denuded zone was observed near a grain boundary in the 1-MeV Kr-irradiated UO2 sample at 800C. The denuded zone in the vicinity of grain boundary was not found when the irradiation temperature was at 600C. The suppression of dislocation loop formation near the boundary is likely due to the enhanced interstitial diffusion toward grain boundary at the high temperature.

  12. RAE 2001 -UoA 44 Accounting and Finance Overview of research in Accounting and Finance

    E-Print Network [OSTI]

    Abrahams, I. David

    RAE 2001 - UoA 44 ­ Accounting and Finance Overview of research in Accounting and Finance This note gives an overview of the state of research in the field of Accounting and Finance in UK universities in the area of Accounting and Finance was submitted to the Business and Management panel (UoA 43) as part

  13. Molecular dynamics simulations of grain boundary thermal resistance in UO2

    SciTech Connect (OSTI)

    Tianyi Chen; Di Chen; Bulent H. Sencer; Lin Shao

    2014-09-01

    By means of molecular dynamics (MD) simulations, we have calculated Kaptiza resistance of UO2 with or without radiation damage. For coincident site lattice boundaries of different configurations, the boundary thermal resistance of unirradiated UO2 can be well described by a parameter-reduced formula by using boundary energies as variables. We extended the study to defect-loaded UO2 by introducing damage cascades in close vicinity to the boundaries. Following cascade annealing and defect migrations towards grain boundaries, the boundary energy increases and so does Kaptiza resistance. The correlations between these two still follow the same formula extracted from the unirradiated UO2. The finding will benefit multi-scale modeling of UO2 thermal properties under extreme radiation conditions by combining effects from boundary configurations and damage levels.

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

    SciTech Connect (OSTI)

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

    2015-01-01

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

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

    SciTech Connect (OSTI)

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

    1995-01-01

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

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

    SciTech Connect (OSTI)

    John M. Veilleux

    1999-08-01

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

  17. PUREX/UO3 Facilities deactivation lessons learned history

    SciTech Connect (OSTI)

    Gerber, M.S.

    1996-09-19

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

  18. 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

  19. Near Surface Stoichiometry in UO 2 : 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

    2015-01-01

    The mechanisms of oxygen stoichiometry variation in UO2at 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 UO2near 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 UO2have 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 bymore »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 UO2prefers to be hypostoichiometric, although the surface is near-stoichiometric.« less

  20. 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.

  1. Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets

    E-Print Network [OSTI]

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

    2014-01-01

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

  2. Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets

    E-Print Network [OSTI]

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

    2014-02-28

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

  3. 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.

  4. EMPLOYEE ACCIDENT / INCIDENT / QUALITY IMPROVEMENT REPORT UoW 1018 (Rev. 3/05)

    E-Print Network [OSTI]

    Matrajt, Graciela

    provided to affected party (originator). University Risk Management Box 351276 UWB Administrative Services the Labor and Industries (L&I) "Report of Industrial Injury or Occupational Disease." Your doctor should mail this form to the University Risk Management Office. UoW 1018 Accident/Incident Form L&I Injury

  5. Simulations of Thermal and Oxygen Transport in UO2 Fuels Marius Stan1

    E-Print Network [OSTI]

    Mihaila, Bogdan

    Simulations of Thermal and Oxygen Transport in UO2 Fuels Marius Stan1 and Bogdan Mihaila2 1 Argonne properties of nuclear fuels and predicting the behavior of nuclear fuel elements (ceramic fuel pellets or metallic fuel rods) under normal and accident conditions are major challenges for the nuclear fuel

  6. UoN2013/8694BICRICOSProvider00109J Advertising with the University of

    E-Print Network [OSTI]

    Fleming, Andrew J.

    w UoN2013/8694BICRICOSProvider00109J CONTACT Advertising with the University of Newcastle's Careers's always the right time to advertise a part-time/ casual vacancy. Students love it when employers come on on CareerHub OUR STUdENTS YOUR TAlENT It's fast, easy and cost effective. Advertising opportunities with us

  7. RAE 2001 -UoA66 Drama, Dance and Performing Arts -Overview Report 1. Research assessed

    E-Print Network [OSTI]

    Abrahams, I. David

    rise in quality. The submissions overall demonstrated an impressive diversity of types of research1 RAE 2001 - UoA66 Drama, Dance and Performing Arts - Overview Report 1. Research assessed colleges offering degree courses. The largest department submitted over forty researchers, the smallest sub

  8. In-situ TEM observation of dislocation evolution in Kr-irradiated UO2 single crystal

    SciTech Connect (OSTI)

    Lingfeng He; Mahima Gupta; Clarissa A. Yablinsky; Jian Gan; Marquis A. Kirk; Xian-Ming Bai; Janne Pakarinen; Todd R. Allen

    2013-11-01

    In-situ transmission electron microscopy (TEM) observation of UO2 single crystal irradiated with Kr ions at high temperatures was conducted to understand the dislocation evolution due to high-energy radiation. The dislocation evolution in UO2 single crystal is shown to occur as nucleation and growth of dislocation loops at low-irradiation doses, followed by transformation to extended dislocation segments and networks at high doses, as well as shrinkage and annihilation of some loops and dislocations due to high temperature annealing. Generally the trends of dislocation evolution in UO2 are similar under Kr irradiation at different ion energies and temperatures (150 keV at 600 degrees C and 1 MeV at 800 degrees C) used in this work, although the specific dislocation loop size and density are quite different. Interstitial-type dislocation loops with Burgers vector along <110> were observed in the Kr-irradiated UO2.The irradiated specimens were denuded of dislocation loops near the surface.

  9. 12/21/2011 KWarden UO Policy Library Policy Revision and Update Guidelines

    E-Print Network [OSTI]

    Oregon, University of

    12/21/2011 ­ KWarden UO Policy Library Policy Revision and Update Guidelines Any Responsible Office, Policy Statement: Development and Management. The policy refers to two types of revisions: substantive: Development and Management, which is found in the Policy Library. Minor Revision or Update A minor revision

  10. PRIOR RISK ASSESSMENT 005 Bradley / adapted by UoE URPO

    E-Print Network [OSTI]

    Bearhop, Stuart

    PRIOR RISK ASSESSMENT 005 Version 002 Author Bradley / adapted by UoE URPO Date of version 1 issue the University to undertake a prior risk assessment before commencing a new work activity involving ionising a review of existing risk assessments at appropriate intervals to ensure that they remain suitable

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

    E-Print Network [OSTI]

    Oregon, University of

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

  12. 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)

  13. 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.

  14. Strong electron correlation in UO{sub 2}{sup ?}: A photoelectron spectroscopy and relativistic quantum chemistry study

    SciTech Connect (OSTI)

    Li, Wei-Li; Jian, Tian; Lopez, Gary V.; Wang, Lai-Sheng, E-mail: lai-sheng-wang@brown.edu [Department of Chemistry, Brown University, Providence, Rhode Island 02912 (United States)] [Department of Chemistry, Brown University, Providence, Rhode Island 02912 (United States); Su, Jing [Department of Chemistry and Key Laboratory of Organic Optoelectronics and Molecular Engineering of Ministry of Education, Tsinghua University, Beijing 100084 (China) [Department of Chemistry and Key Laboratory of Organic Optoelectronics and Molecular Engineering of Ministry of Education, Tsinghua University, Beijing 100084 (China); Division of Nuclear Materials Science and Engineering, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China and Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Hu, Han-Shi; Cao, Guo-Jin; Li, Jun, E-mail: junli@tsinghua.edu.cn [Department of Chemistry and Key Laboratory of Organic Optoelectronics and Molecular Engineering of Ministry of Education, Tsinghua University, Beijing 100084 (China)] [Department of Chemistry and Key Laboratory of Organic Optoelectronics and Molecular Engineering of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2014-03-07

    The electronic structures of actinide systems are extremely complicated and pose considerable challenges both experimentally and theoretically because of significant electron correlation and relativistic effects. Here we report an investigation of the electronic structure and chemical bonding of uranium dioxides, UO{sub 2}{sup ?} and UO{sub 2}, using photoelectron spectroscopy and relativistic quantum chemistry. The electron affinity of UO{sub 2} is measured to be 1.159(20) eV. Intense detachment bands are observed from the UO{sub 2}{sup ?} low-lying (7s?{sub g}){sup 2}(5f?{sub u}){sup 1} orbitals and the more deeply bound O2p-based molecular orbitals which are separated by a large energy gap from the U-based orbitals. Surprisingly, numerous weak photodetachment transitions are observed in the gap region due to extensive two-electron transitions, suggesting strong electron correlations among the (7s?{sub g}){sup 2}(5f?{sub u}){sup 1} electrons in UO{sub 2}{sup ?} and the (7s?{sub g}){sup 1}(5f?{sub u}){sup 1} electrons in UO{sub 2}. These observations are interpreted using multi-reference ab initio calculations with inclusion of spin-orbit coupling. The strong electron correlations and spin-orbit couplings generate orders-of-magnitude more detachment transitions from UO{sub 2}{sup ?} than expected on the basis of the Koopmans’ theorem. The current experimental data on UO{sub 2}{sup ?} provide a long-sought opportunity to arbitrating various relativistic quantum chemistry methods aimed at handling systems with strong electron correlations.

  15. Bubble formation and Kr distribution in Kr-irradiated UO2

    SciTech Connect (OSTI)

    He, L. F. [Univ. of Wisconsin, Madison, WI (United States) Dept. of Engineering Physics; Valderrama, B. [Univ. of Florida, Gainesville, FL (United States) Dept. of Materials Science and Engineering; Hassan, A. -R. [Purdue Univ., West Lafayette, IN (United States) School of Nuclear Engineering; Yu, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gupta, M. [Univ. of Wisconsin, Madison, WI (United States) Dept. of Engineering Physics; Pakarinen, J. [Univ. of Wisconsin, Madison, WI (United States) Dept. of Engineering Physics; Henderson, H. B. [Univ. of Florida, Gainesville, FL (United States) Dept. of Materials Science and Engineering; Gan, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kirk, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Nelson, A. T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Manuel, M. V. [Univ. of Florida, Gainesville, FL (United States) Dept. of Materials Science and Engineering; El-Azab, A. [Purdue Univ., West Lafayette, IN (United States) School of Nuclear Engineering; Allen, T. R. [Univ. of Wisconsin, Madison, WI (United States) Dept. of Engineering Physics; Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

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

  16. 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].

  17. 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.

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

    E-Print Network [OSTI]

    Feinroth, H.

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

  19. 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.

  20. 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)

  1. 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.

  2. Multiscale simulation of xenon diffusion and grain boundary segregation in UO?

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

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

    2015-07-01

    In light water reactor fuel, gaseous fission products segregate to grain boundaries, resulting in the nucleation and growth of large intergranular fission gas bubbles. The segregation rate is controlled by diffusion of fission gas atoms through the grains and interaction with the boundaries. Based on the mechanisms established from earlier density functional theory (DFT) and empirical potential calculations, diffusion models for xenon (Xe), uranium (U) vacancies and U interstitials in UO? have been derived for both intrinsic (no irradiation) and irradiation conditions. Segregation of Xe to grain boundaries is described by combining the bulk diffusion model with a model formore »the interaction between Xe atoms and three different grain boundaries in UO? (?5 tilt, ?5 twist and a high angle random boundary), as derived from atomistic calculations. The present model does not attempt to capture nucleation or growth of fission gas bubbles at the grain boundaries. The point defect and Xe diffusion and segregation models are implemented in the MARMOT phase field code, which is used to calculate effective Xe and U diffusivities as well as to simulate Xe redistribution for a few simple microstructures.« less

  3. Synchrotron characterization of nanograined UO2 grain growth

    SciTech Connect (OSTI)

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

    2015-09-30

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

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

    SciTech Connect (OSTI)

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

    2015-09-30

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

  5. 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.

  6. THE ELECTRON AFFINITY OF UO E.B. Rudnyi, E.A. Kaibicheva, L.N. Sidorov

    E-Print Network [OSTI]

    Rudnyi, Evgenii B.

    THE ELECTRON AFFINITY OF UO 3 * E.B. Rudnyi, E.A. Kaibicheva, L.N. Sidorov Department of Chemistry keywords - negative ions, uranium oxide, electron affinity, ion - molecule equilibria, high temperature stronger than fluorides. The uranium oxide lies aside from other molecules in -1 table 1. High electron

  7. 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.

  8. 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)

  9. 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.

  10. A Study of UO2 Grain Boundary Structure and Thermal Resistance Change under Irradiation using Molecular Dynamics Simulations 

    E-Print Network [OSTI]

    Chen, Tianyi

    2013-08-02

    OF UO2 GRAIN BOUNDARY STUCTURE AND THERMAL RESISTANCE CHANGE UNDER IRRADIATION USING MOLECULAR DYNAMICS SIMULATIONS A Thesis by TIANYI CHEN Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment...] As the fuel expands to reach the cladding, the fuel cladding interaction begins to draw attention. The first thing is the fuel/cladding mechanical interaction. This phenomenon is very serious in early designed reactors since they do not have enough spaces...

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

    SciTech Connect (OSTI)

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

    2011-04-15

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

  12. 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.

  13. 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.

  14. 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.

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

    SciTech Connect (OSTI)

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

    2013-10-01

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

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

    SciTech Connect (OSTI)

    Van Brutzel, Laurent

    2008-07-01

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

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

    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

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

    SciTech Connect (OSTI)

    Margaret A. Marshall; John D. Bess

    2012-11-01

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

  19. 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.

  20. Comparison between phase field simulations and experimental data from intragranular bubble growth in UO{sub 2}

    SciTech Connect (OSTI)

    Tonks, M. R.; Biner, S. B.; Mille, P. C. [Idaho National Laboratory, P.O. Box 1625 MS 3830, Idaho Falls, ID 83415 (United States); Andersson, D. A. [MST-8, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2013-07-01

    In this work, we used the phase field method to simulate the post-irradiation annealing of UO{sub 2} described in the experimental work by Kashibe et al., 1993 [1]. The simulations were carried out in 2D and 3D using the MARMOT FEM-based phase-field modeling framework. The 2-D results compared fairly well with the experiments, in spite of the assumptions made in the model. The 3-D results compare even more favorably to experiments, indicating that diffusion in all three directions must be considered to accurate represent the bubble growth. (authors)

  1. 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

  2. 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.

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

    SciTech Connect (OSTI)

    Collins, J.L.

    2004-12-02

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

  4. 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.

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

    SciTech Connect (OSTI)

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

    2014-11-01

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

  6. UoA 27 Chemical Engineering: Overview The Panel carried out its assessment work in line with its published Criteria and Working Methods

    E-Print Network [OSTI]

    Abrahams, I. David

    UoA 27 Chemical Engineering: Overview The Panel carried out its assessment work in line with its that some chemical engineering research activity had been submitted to Unit of Assessment 26, General Engineering. The General Engineering Panel had sought cross-referral advice from the Chemical Engineering

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

    E-Print Network [OSTI]

    Cina, Jeff

    measurement, and device fabrication to design, build and study new materials and structures that have applications in solar energy harvesting and electrochemical energy storage. Chartoff, Richard - The UO Polymer relevance to developing technologies. Pluth, Michael D. - Research in the Pluth group focuses on extending

  8. 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.

  9. [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 +}.

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

    SciTech Connect (OSTI)

    Yu, Na; Klepov, Vladislav V. [Forschungszentrum Jülich GmbH, Institute for Energy and Climate Research (IEK-6), 52428 Jülich (Germany); Villa, Eric M. [Department of Chemistry, Creighton University, 2500 California Plaza, Omaha NE 68178 (United States); Bosbach, Dirk [Forschungszentrum Jülich GmbH, Institute for Energy and Climate Research (IEK-6), 52428 Jülich (Germany); Suleimanov, Evgeny V. [Department of Chemistry, Lobachevsky State University of Nizhny Novgorod, 603950 Nizhny Novgorod (Russian Federation); Depmeier, Wulf [Institut für Geowissenschaften, Universität zu Kiel, 24118 Kiel (Germany); Albrecht-Schmitt, Thomas E., E-mail: albrecht-schmitt@chem.fsu.edu [Department of Chemistry and Biochemistry, Florida State University, 102 Varsity Way, Tallahassee, FL 32306-4390 (United States); Alekseev, Evgeny V., E-mail: e.alekseev@fz-juelich.de [Forschungszentrum Jülich GmbH, Institute for Energy and Climate Research (IEK-6), 52428 Jülich (Germany); Institut für Kristallographie, RWTH Aachen University, 52066 Aachen (Germany)

    2014-07-01

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

  11. 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

  12. 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.

  13. The relative variational model - A topological view of matter and its properties: UO{sub 2}{sup {+-}x} density

    SciTech Connect (OSTI)

    Dias, M. S.; De Vasconcelos, V.; Mattos, J. R. L.; Lameiras, F. S. [Center for Development of the Nuclear Technology - CDTN, National Commission for the Nuclear Energy - CNEN, PO Box: 941, 30.161-970, Belo Horizonte, Minas Gerais (Brazil); Jordao, E. [Chemistry Engineering Dept., Campinas State Univ., FEQ/UNICAMP, Av. Albert Einstein, 500, 13083-852, Campinas, Sao Paulo (Brazil)

    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 lattice parameter and specific mass of UO{sub 2}{sup {+-}x} are depicted in dependence on the variations of the stoichiometry in the band of -0.016 {<=} x and <0.25 and temperature from 0 K up to melting point. (authors)

  14. 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. [Oak Ridge National Lab., TN (United States); Carter, J.C. [J.C. Carter Associates, Inc., Knoxville, TN (United States)

    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.

  15. 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.

  16. 2. HIGH-LOv~ JUNCTION FORY_,\\'UO AN EXPERIMENTAL STUDY OF AL-ALLOYED:'p+ JUNCT;[ONS FOR SSF SOLAR CELT.S As temperature rises en..!."

    E-Print Network [OSTI]

    del Alamo, Jesús A.

    . Luque formed. The deposited Al diss Instituto de Energia Solar {E.T,S,I.T,} phase composition given2. HIGH-LOv~ JUNCTION FORY_,\\'UO AN EXPERIMENTAL STUDY OF AL-ALLOYED:§'p+ JUNCT;[ONS FOR SSF SOLAR+pp+ bifacial SSF solar cells are used to experimentally analyse the interphase in a similar way a 5i layer

  17. 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.

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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)

  3. Pipe diffusion at dislocations in UO2

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformationJesseworkSURVEYI/OPerformancePi Day Pi Day Pi Day is anPiotr ZbiegielFUEL.P8.01

  4. 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.

  5. Modeling of Fission Gas Release in UO2

    SciTech Connect (OSTI)

    MH Krohn

    2006-01-23

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

  6. Microstructure changes and thermal conductivity reduction in UO2 following

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfate Reducing(JournalspectroscopyReport) |(Patent)Inter-Nucleon InteractionsElectron Beam2}CuO{sub3.9

  7. Migration Mechanisms of Oxygen Interstitial Clusters in UO2 (Journal

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfate Reducing(JournalspectroscopyReport) |(Patent)Inter-Nucleon InteractionsElectronArticle) |

  8. Oxidative Dissolution of UO2 in a Simulated Groundwater Containing

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTech ConnectSpeeding access to(Conference) |of lithium-ion cells (JournalSynthetic

  9. Oxidative Dissolution of UO2 in a Simulated Groundwater Containing

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTech ConnectSpeeding access to(Conference) |of lithium-ion cells

  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. UNIVERSITY NEWS AND PEOPLE OF THE UO EEP IN THE BASEMENT OF PACIFIC

    E-Print Network [OSTI]

    a small volume of water from a tube, producing a traveling vortex ring in water. The intent is to figure out how to control the swirl- ing ring. "The vortex of fluid dynamics is not understood," Johnson says

  12. OVERVIEW: UoA 22 PURE MATHEMATICS 1. Conduct of the assessment.

    E-Print Network [OSTI]

    Abrahams, I. David

    from 14 UK institutions according to their specialisations. A significant proportion was considered, there was a wide variation in the performance of institutions within these categories. This figure constituted submitted are slightly different, more were putting forward a higher proportion of staff than in 1996

  13. Materials Data on La6UO12 (SG:148) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2014-11-02

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

  14. 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

  15. 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

  16. 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.

  17. THE REDISTRIBUTION OF RUTHENIUM IN UO2 IN A TEMPERATURE GRADIENT

    E-Print Network [OSTI]

    Zhou, S.Y.

    2010-01-01

    of Metallic Fission Products in Reactor Oxide Fuels", Nucl.of metallic fission products in oxide fuel elements.metallic inclusions distributed along grain boundaries or in the central void of the fuel

  18. A grant from the UO Tom and Carol Williams Fund for Undergraduate

    E-Print Network [OSTI]

    the opportunity to create a documentary film that outlines the life of one man. My film, Carlos Enrique Marquez focuses winter term on giving a theoretical, documentary, and ethnographic understanding of the processes teaches students how to produce a short video documentary from oral history interviews. Following

  19. 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.

  20. 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

  1. Bradley / adapted by UoE URPO Date of version 1

    E-Print Network [OSTI]

    Bearhop, Stuart

    Telephone: (309) 677-2700 Fax: (309)677-3534 AUTHORIZATION FOR RELEASE OF INFORMATION I authorize Bradley

  2. 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.

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

    SciTech Connect (OSTI)

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

    2011-01-18

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

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

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfate Reducing(Journal Article) | SciTech(Journal Article)at theReport) |in QCD: Thethe

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

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfate Reducing(Journal Article) | SciTech(Journal Article)at theReport) |in QCD: Thethe(Technical

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverse (Journal Article) |Final Report Document Number(Technical Report)in the

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

    E-Print Network [OSTI]

    Naramore, Michael J

    2010-08-03

    is very prominent and porosity helps alleviate internal pressures to reduce fuel deformation. Therefore, a balance between thermal conductivity and fission gas accommodation is necessary to achieve a long lasting fuel [4]. 4> F c.> ::>oz 00 c.> 3 -l F...................................................................................................................78 APPENDIX G ..................................................................................................................82...

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

    E-Print Network [OSTI]

    Fu, Chun

    2000-01-01

    is calculated with ANS decay heat data, enables the model to be used for analysis of a large spectrum of transients and accidents. The plant model is used for analysis and prediction of a cold leg Large Break Loss-of-Coolant Accident (LBLOCA). The RELAP5/MOD3...

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

    SciTech Connect (OSTI)

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

    2009-08-01

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

  10. A literature review on the chemical and physical properties of uranyl fluoride (UO sub 2 F sub 2 )

    SciTech Connect (OSTI)

    Myers, W.L. (Los Alamos National Lab., NM (USA) Illinois Univ., Urbana, IL (USA). Dept. of Nuclear Engineering)

    1990-08-01

    This report reviews the preparation and properties of uranyl fluoride. Data are given on the crystal structure, solubility in water, specific gravity, density, specific heat, enthalpy, entropy, acidity, corrosion properties, and refractive indices. Empirical formulas are given to calculate specific gravity, density of aqueous solutions, molal volume, and refractive indices. 13 refs., 3 figs., 11 tabs.

  11. Chemistry Transfer Evaluations Before requesting a transfer evaluation, please look up your courses on the UO Transfer Credit

    E-Print Network [OSTI]

    Cina, Jeff

    Chemistry Transfer Evaluations Before requesting a transfer evaluation, please look up your courses://registrar.uoregon.edu/current_students/transfer-articulation If your Chemistry courses have not been evaluated, or if you feel the evaluation is not correct, then fill out this form and bring it and the associated documents to a Chemistry advisor. Name

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

    Office of Legacy Management (LM)

    Cornell University and Medical College Fordham University Long Island College of Medicine. Polytechnic Institute of Brooklyn Rensselaer Polytechnic Institute Rockefeller...

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

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouth DakotaRobbins and Myers CoMadison -T: Designation ofSEP 2Dr.Wm.*

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

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouth DakotaRobbins and700, 1. .&. ' , cMarchW W e e l l d d o

  15. OXIDATIVE DISSOLUTION OF BIO-U(IV)O2(S) IN PRESENCE OF NITRATE AND IRON UNDER ANAEROBIC CONDITIONS USING FLOW-THROUGH COLUMNS

    E-Print Network [OSTI]

    Pokharel, Rasesh

    2013-01-01

    UO 2 (s) to aqueous uranyl (UO 22+ ) is well known (Finneranof S. Oneidensis; (d) anaerobic phase uranyl acetate +state it occurs as the uranyl ion (UO 2 ) 2+ , which forms

  16. 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...

  17. 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.

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

    E-Print Network [OSTI]

    Cina, Jeff

    , physical measurement, and device fabrication to design, build and study new materials and structures that have applications in solar energy harvesting and electrochemical energy storage Cina, Jeffrey A solids that permits them to prepare families of new nanostructured and kinetically stable compounds

  19. Equipment List for UO Geology Field Camp, Summer 2012 Camping Equipment (We will be camping in central Oregon during the first 2-week segment,

    E-Print Network [OSTI]

    Roering, Joshua J.

    weather clothing Sun hat, sunglasses, and light weight fabrics that keep the sun off your skin Sunscreen

  20. PROCEEDINGS OF WORKSHOP ON THERMOMECHANICAL-HYDROCHEMICAL MODELING FOR A HARDROCK WASTE REPOSITORY. JULY 29-31, 1980. MARRIOTT INN, BERKELEY, CA

    E-Print Network [OSTI]

    Authors, Various

    2010-01-01

    Experi ments by the BGR in the Asse II", Proc. UoS. /FRG Bi"pp,-- Rothfuchs, T. (1979): "Asse II In-Situ Brine Migrationin the late 1960's in the Asse Salt Mine in the Federal

  1. Uranyl Sequestration: Synthesis and Structural Characterization of Uranyl Complexes with a Tetradentate Methylterephthalamide Ligand

    E-Print Network [OSTI]

    Ni, Chengbao

    2012-01-01

    Me 4 N] 8 [L(UO 2 )] 4 tetramer, formed via coordination ofAddition of KOH to the tetramer gave the corresponding2- forms an unexpected tetramer [L(UO 2 )] 48- , in which

  2. Aalborg Universitet Soshinskaya, Mariya; Graus, Wina; Guerrero, Josep M.; Quintero, Juan Carlos Vasquez

    E-Print Network [OSTI]

    Vasquez, Juan Carlos

    Technology Solutions (CERTS); Integrated Power Supply (IPS); molten carbonate fuel cell (MCFC); phosphoric acid fuel cell (PAFC); Alternating Current (AC); Direct Current (DC); use of system (UoS); research components, dual-mode switching from g

  3. Electron Microbeam Investigation of Uranium-Contaminated Soils from

    E-Print Network [OSTI]

    Zhu, Chen

    . Uranium(VI), which typically occurs in the uranyl (UO2 2+) ion or in uranyl complexes, dominates under

  4. Fate of Radionuclides in Wastewater Treatment Plants

    E-Print Network [OSTI]

    Shabani Samgh Abadi, Farzaneh

    2013-01-01

    is preferred over metallic uranium. Reactor fuel is usuallyfuel. Reactor fuel can be prepared both as UO 2 or metallic

  5. Clustering of protein families into functional subtypes using Relative Complexity Measure with reduced

    E-Print Network [OSTI]

    Yanikoglu, Berrin

    @su.sabanciuniv.edu HHO: hotu@bidmc.harvard.edu UOS: ugur@sabanciuniv.edu #12;- 2 - Abstract Background Phylogenetic

  6. Vapour Phase Hydration of Blended Oxide Magnox Waste Glasses Neil C. Hyatt,1*

    E-Print Network [OSTI]

    Sheffield, University of

    level waste (HLW) arising from the reprocessing of spent Oxide (UO2) and Magnox nuclear fuels is blended

  7. 9DSRXU 3KDVH +\\GUDWLRQ RI %OHQGHG 2[LGH 0DJQR[ :DVWH *ODVVHV Neil C. Hyatt,1*

    E-Print Network [OSTI]

    Sheffield, University of

    level waste (HLW) arising from the reprocessing of spent Oxide (UO2) and Magnox nuclear fuels is blended

  8. MODELING THE PERFORMANCE OF HIGH BURNUP THORIA AND URANIA PWR FUEL

    E-Print Network [OSTI]

    Long, Y.

    Fuel performance models have been developed to assess the performance of ThO[subscript 2]-UO[subscript 2]

  9. Evolutionary implications: myoglobin-like proteins found in ancient mic... http://www.eurekalert.org/pub_releases/2000-02/UoH-Eimp-0102100.php 1 of 2 3/5/2008 12:50 PM

    E-Print Network [OSTI]

    Alam, Maqsudul

    in oxygen transport and storage. The newly identified proteins may be the evolutionary ancestors of proteins involved in oxygen sensing as well as transport and storage. The findings, which appear in the Feb. 3 issue vapor, nitrogen, methane and ammonia that made up Earth's atmosphere for food and energy, probably

  10. Topological Analysis of Void Spaces in Tungstate Frameworks: Assessing Storage Properties for the Environmentally Important Guest Molecules and Ions: CO_2, UO_2, PuO_2, U, Pu, Sr^2+, Cs+, CH_4, and H_2

    E-Print Network [OSTI]

    Cole, Jacqueline Manina; Cramer, Alisha J.; Zeidler, Anita

    2015-07-15

    appropriate. In the ongoing search for usable materials, data mining of structure databases can provide a useful tool to identify potential candidates for the applications in hand. For example, a study on Li+ migration maps26 examined the structure... products found in nuclear waste. Waste from nuclear facilities, in the form of spent nuclear fuel, is found predominantly in the form of uranium or plutonium oxides.28 Furthermore, current efforts, especially among tungstates, are largely focused...

  11. Study of D(0) decays into K overbar (0) and K overbar (*0)

    E-Print Network [OSTI]

    Ammar, Raymond G.; Ball, S.; Baringer, Philip S.; Coppage, Don; Copty, N.; Davis, Robin E. P.

    1993-11-01

    4.6+p 9+1.1% 2.6+0.3+0.7%%uo 1.4'Po 3.9%%uo 2.7% 0.25% 3.1 %%uo K* g 1.7+0.3+0.4% 2. 3+0.7+1.0%%uo & 1.4%%uo(90% C.L.) & 2.6%(90%%uo C.L.) 0.3% 2.5% 0.25%%uo + 2. 0%%uo & 0. 13%(90% C.L.) 0.003% mode. In all cases, results were consistent...+30 80+12 372+45 6.8+0.2% 2.5+0.1% 0.91+0.07%%uo 1.63+0.06%%uo 168+15 0.85+0.03% 54+9 6500+ 171 0.32+0.02%%uo 10.1+0.3% 't is the product of the efficiency to reconstruct this made and the branching ratio of the D daughter to observable particles. Stech...

  12. Molten uranium dioxide structure and dynamics

    SciTech Connect (OSTI)

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

    2014-11-20

    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.

  13. Molten uranium dioxide structure and dynamics

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

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

    2014-11-20

    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.

  14. Molten uranium dioxide structure and dynamics

    SciTech Connect (OSTI)

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

    2014-11-21

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

  15. Molten uranium dioxide structure and dynamics

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

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

    2014-11-21

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

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

    SciTech Connect (OSTI)

    Gofryk, K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harrison, N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). National High Magnetic Field Lab. (MagLab); Jaime, M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). National High Magnetic Field Lab. (MagLab)

    2014-12-01

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

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

    E-Print Network [OSTI]

    Szigethy, Geza

    2010-01-01

    Design for the Uranyl Cation, UO 22+ ………………………. ……………….15tripodal ligands for uranyl chelation …………………….19 Figurestudies with the uranyl cation ………………………………………………. 36 Figure

  18. 4th Annual DOE-ERSP PI Meeting: Abstracts

    E-Print Network [OSTI]

    Hazen, Terry C.

    2009-01-01

    New microelectrodes measuring uranyl concentration ([UO 22suggests Ca-tricarbonato-uranyl complexes predominate underfield conditions, while calculated uranyl ion activities are

  19. On the acceleration potential in perfect fluid flow 

    E-Print Network [OSTI]

    Maestri, Raymond Rudolph

    1960-01-01

    number space. COMMENT: This theorem is tacitly understood. in this and similar discourses. It needs explicit formulation for the purposes of the following arguments. THEOREM ): If '7X U 0) V P 6. (0 U 3 )& then PROOF: ~8) Ug Uo ~P . V R 2 2 E... LEMMA 2& If $(3) ~ 0 as ( R Bp (Rp& Kp) then P ~ -; gR E' Kp 19 This lemma follows from the definition of p and from (12). 0 The conjoint of lemma 1, lemma 2, and theorem 2 states; (() - Uo ) )& Uo . VR c 0- 2 U~ Uo ? 2 ~P ) $ pP $b R Q. E. D...

  20. 4th Annual DOE-ERSP PI Meeting: Abstracts

    E-Print Network [OSTI]

    Hazen, Terry C.

    2009-01-01

    to adhere to hematite and goethite. Two adhesion-deficientUO 2 by ferrihydrite, goethite, and hematite-coated quartzof batch experiments with goethite as the electron acceptor,

  1. HIGH CURRENT D- PRODUCTION BY CHARGE EXCHANGE IN SODIUM

    E-Print Network [OSTI]

    Hooper, E.B.

    2011-01-01

    good beam optics at the 1 keV energy and below required forto relatively high energies. optics is goou. Uo)~ but the

  2. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    2015, American Physical Society Emergent Properties of the Bose-Einstein-Hubbard Condensate in UO2(+x) Conradson, Steven D. ; Durakiewicz, Tomasz Full Text Available April 2013...

  3. ELECTRON-SPIN-RESONANCE STUDIES ON PHOTO-SYNTHETIC MATERIALS

    E-Print Network [OSTI]

    Sogo, Power B.; Carter, Louise A.; Calvin, Melvin.

    2008-01-01

    for the Uo So Atomic Energy Commission UCRL-9153 ELECTRON-in part by the U. S. Atomic Energy Commission and in part by

  4. Transition projects FY 1995 multi-year program plan/fiscal year work plan WBS 1.3.1. and 7.1

    SciTech Connect (OSTI)

    Cartmell, D.B.

    1994-09-01

    This document presents a complete listing and time line of transitional projects associated with the Purex/UO3 deactivation project at the Hanford reservation.

  5. James Banfield1 Srikanth Allu2

    E-Print Network [OSTI]

    Mihaila, Bogdan

    Performance code [1] for uranium-dioxide (UO2) fuel in a heavy boiling-water reactor and is an extension reactor, which is a Heavy-Boiling Water Reactor (HBWR) with UO2 fuel. There are three rods presented to fixed temperature of 513 K, which is the saturation temperature of the heavy water in the Halden reactor

  6. Forest fires, explosions, and random trees Edward Crane

    E-Print Network [OSTI]

    Wirosoetisno, Djoko

    Forest fires, explosions, and random trees Edward Crane HIMR, UoB 13th January 2014 #12 and James Martin at the University of Oxford. Edward Crane (HIMR, UoB) Forest fires, explosions, and random trees 13th January 2014 2 / 20 #12;Overview This talk is about the mean field forest fire model

  7. Cloud K-SVD: Computing data-adaptive representations in the cloud

    E-Print Network [OSTI]

    Bajwa, Waheed U.

    Cloud K-SVD: Computing data-adaptive representations in the cloud Haroon Raja and Waheed U. Bajwa Department of Electrical and Computer Engineering, Rutgers University, Piscataway, NJ 08854 Emails: haroon a distributed algorithm, termed as cloud K-SVD, for learning a UoS structure underlying distributed data

  8. 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.

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

    SciTech Connect (OSTI)

    Gofryk, K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zapf, V. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). National High Magnetic Field Lab. (MagLab); Jaime, M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). National High Magnetic Field Lab. (MagLab)

    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. Geology of the Pontotoc North-Northwest area San Saba County, Texas 

    E-Print Network [OSTI]

    Chauvin, Aaron Lawrence

    1962-01-01

    T wg ttt ua&4 86uTps? 'KB?&? 3'o 8??? uv. Bqttesesdez (we leod 'g ?yd) deut vy6oyoa6 e~ uo ByoqmKB cl'f p pttQ e8'f ?$8 e~ go tlQL 8 ~ 8+%? ~8 e~ go Bd'$ p puc' se~ g@ct u fe+clo og pgagg etlg wf 'Pasn 89th Bs'QdUIOD uo+crKlgg SedoosoiMGQB 3'lcpg...' 0;foo'r wcuogn fd Tg Bugs ZSAggg t79$trSdg pwr{xoH EgsnofhszQ G . o 8 xodsz snld ssgci" I 8 ssot~ . 8&Rs+ ~6zng sggygnpsgg xnou Mlop uo'fQ'HTQGQxs trs go~ p3+u~oug rrrogj ssgd tlp28 p:3xgni30'2 Gtg dr~ ox' uo 3$7rgtj s~gggss uQNxQQ go dnaz6 uo...

  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. A theoretical investigation of particle motion occurring in a two-phase curvilinear flow 

    E-Print Network [OSTI]

    Redler, Kenneth Oliver

    1962-01-01

    Xl + 3 +o ?]. ?2 3 ~0+1 2Uow ?1?2 9@o? Kl + 2Uo ?2) + (9990~?lr + 3 Uow ?1?2r 3 Uod?lr - 12Uo Xl?2 + 9 Uow Kjr + 33Io? Kl?2r + ?Vow ?2r 6U w ?2r + 2V w ?2r + 3U w X12r + 3U w Kl?2r + V2?V2r+f + ( U ?V&r+~ V? ?1Kp- SUow Klr - 16Uow Kl?2r w 9 Uow... Xlr + @Vow ?1K2r + 2U w ?2r- 6Va~F + Vow+2r + 3Vo? Xlr + M2o~ Xl?2r + I, Vow ~2rp u + ( U w ?11 + W3 w Kl?2r - SVoool'r - 1SUow Xl?2r +99? Klr + Kl?2r + 6Uo? ?2r - 9Uo? 4 + 2? ?2r + )33Vo? ?l2r + 6Uow ?1?2r +3Uo?X2r~+ + QVor Klr + 3 Uo?X1X2r" gUo...

  13. 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.

  14. Development of dual phase magnesia-zirconia ceramics for light water reactor inert matrix fuel 

    E-Print Network [OSTI]

    Medvedev, Pavel

    2005-02-17

    at al., 1970) MgO chemically polished (Evans at al., 1970) UO2 as machined (Evans and Davidge,1969) UO2 chemically polished (Evans and Davidge, 1969) 25 l f E kTSR ? ?= 2.2.6 Fracture strength Results of three-point bend tests for MgO [44] and UO2... stress. The thermal shock resistance is the best for materials that have high fracture strength (? f), high thermal conductivity (k), but low elastic modulus (E) and low coefficient of thermal expansion (? l). One way to quantify thermal shock...

  15. 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.

  16. Design and Fabrication of cm-scale Tesla Turbines

    E-Print Network [OSTI]

    Krishnan, Vedavalli Gomatam

    2015-01-01

    variable torque_efficiency = (T1loss - setup.bearingloss) %torque losses T1 = rFlow.T1rotor. *torque_efficiency; eta= head_efficiency. *torque_efficiency; %= T1. /( 2*pi*Uo. *

  17. TO

    Office of Legacy Management (LM)

    42-17, Grade A. It is not presently known whether the code number refers to the uranyl nitrate which was originally ordered or tc the UO3 which was actually reoeived....

  18. Quest for Environmentally-Benign Ligands for Actinide Separations: Thermodynamic, Spectroscopic, and Structural Characterization of U(VI) Complexes with Oxa-Diamide and Related Ligands

    E-Print Network [OSTI]

    Tian, Guoxin; Advanced Light Source

    2009-01-01

    The stock solution of uranyl perchlorate was prepared bysymmetric stretching mode of uranyl in the three complexes.in Figure 8 for the free uranyl cation (UO 22+ ), the value

  19. Theoretical Investigation of Uranyl Dihydroxide: Oxo Ligand Exchange, Water Catalysis, and Vibrational Spectra

    E-Print Network [OSTI]

    Schlegel, H. Bernhard

    Theoretical Investigation of Uranyl Dihydroxide: Oxo Ligand Exchange, Water Catalysis is employed to investigate uranyl dihydroxide, UO2(OH)2, isomerization reaction energy barriers, including those occurring via proton shuttles. The ground-state structure of a uranyl dihydroxide complex

  20. Conservative behavior of uranium vs. salinity in Arctic sea ice and brine Christelle Not a,

    E-Print Network [OSTI]

    ). In natural waters U is found mostly as dis- solved uranyl carbonates (UO2(CO3)3 4 - ) under oxidizing to (1) destabilization of uranyl carbonate complexes; (2) biological uptake; and/or (3) U adsorption

  1. Effects of Sexual Dimorphism and Landscape Composition on the Trophic Behavior of Greater Prairie-

    E-Print Network [OSTI]

    Sandercock, Brett K.

    , Mario Quevedo1,2 1 Research Unit of Biodiversity, (UO/CSIC/PA), Asturias, Spain, 2 Dpt. Biologi´a de Organismos y Sistemas, A´ rea de Ecologi´a, Universidad de Oviedo, Asturias, Spain, 3 Division of Biology

  2. A Study of Selected Chemical and Biological Conditions of the Lower Trinity River and the Upper Trinity Bay 

    E-Print Network [OSTI]

    Baldauf, R. J.

    1970-01-01

    Concern over the effects of water development projects on coastal nurseries prompted the Department of Wildlife Science of Texas A&M University, with the cooperation of the UO So Bureau of Commercial Fisheries, Galveston, ...

  3. Determination of vapor liquid equilibrium for the ternary iso-propanol/atactic-polypropylene/n-heptane mixture at 105C and 140C, and the binary iso-propanol/atactic-polypropylene mixture at 85C using perturbation gas chromatography / cby Lamar Lane Joffrion 

    E-Print Network [OSTI]

    Joffrion, Lamar Lane

    1984-01-01

    Taqg cog uozzag Z H pue uosqeN v pue 'zzogga as~qua s~qs go uo~s~r. zadns quaTged stq zog savoy' sayzeqg ~aqua pue aspayzouqoe oq saqstz zoqgne aqua 'AyyeuosgTppy uoxzepunog aouaToS yeuozgeN aqua pue iuaurqzedap Guxsaauzfu3 yeoxtsa~ gSV sexag aqua go...TzsTzaqoezeqg oTmeulpomzaqZ, azngxTH EuTTapog Lzoaqg uoTgnToS LT uo11ezado TeguamTzadx3 L1 NOIssnzsIG GNv szins33 aznpaoozd uoTgezedazd umnTog pue sTeTzaqeH TeoTmaqg sngezeddv ZNHHI33dX3 suoTgeTaH oTmeuEpomzaqZ, suoTgeTa3 oTqdez604emoIqg 'ZEOHHZ, NOIzzn...

  4. Energistyrelsen 9. juni 2008 Centre for Energy, Environment and Health (CEEH)

    E-Print Network [OSTI]

    data for new technologies (Fuel cells, Electrolysis, Wind turbines ..) Present power system optimization Climate change Met. modelling DMI/NERI/UoC Risø Resulting damages and costs on regional and local

  5. Evaluation of alternative fuel cycle strategies for nuclear power generation in the 21st century

    E-Print Network [OSTI]

    Boscher, Thomas

    2005-01-01

    The deployment of fuel recycling through either CONFU (COmbined Non-Fertile and UO2 fuel) thermal watercooled reactors (LWRs) or fast ABR (Actinide Burner Reactor) reactors is compared to the Once-Through LWR reactor system ...

  6. 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.

  7. 2013-14 MCR Ambassador Application University of Oregon

    E-Print Network [OSTI]

    Oregon, University of

    (s) Minor(s) UO ID Term &Year Entered Total Number of College Credits Earned Cumulative GPA circumstances they may face during their college search. Please explain in no more than 500 words. 2. What piece

  8. 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...

  9. Collapse scenarios of WTC 1 & 2 with extension to generic tall buildings 

    E-Print Network [OSTI]

    Usmani, Asif; Flint, Graeme; Jowsey, Allan; Roben, Charlotte; Torero, Jose L

    This paper presents a summary of the author’s investigation into the collapse of tall buildings. A large number of computational analyses have been carried out at the University of Edinburgh (UoE) over the last 4 years ...

  10. i i 'i i WWW--i i http://www.icmp.lviv.ua/

    E-Print Network [OSTI]

    . Adsorption of radionuclides on ferrocyanides: quantum-chemi- cal calculations I.V.Stasyuk, I.M.Krip, R-initio) the process of adsorption of UO2+ 2 , Cs+ , Sr2+ on ferrocyanides was inves- tigated. The structure

  11. Ris Report No. Danish Atomic Energy Commission

    E-Print Network [OSTI]

    UO.-Zr Fuel Pins. F. List, Reactor Dept. and P. Knudsen, Metallurgy Dept 45 Dispersion, Electronics Dept 69 Analysis of Metallic Ores by Radioisotope-excited X-Ray Fluorescence. L. Løvborg and H

  12. High Vowel Fricativization and Chain Shift

    E-Print Network [OSTI]

    Faytak, Matthew

    2014-01-01

    Aghem – 0 Isu – U, u Oku – (v)@, uo PCR – *u Ngv@@ Kom –v Ngv Chicken mbv´ mbv` U Death ´e-v´ v:U Ey-kuo @-kv *-kuo

  13. Subsurface Biogeochemical Research (SBR) Contractor-Grantee Workshop--Abstracts

    E-Print Network [OSTI]

    Hazen, Terry C.

    2010-01-01

    UO 2 by ferrihydrite, goethite, and hematite-coated quartzFe(III) oxides (primarily goethite) were present at the timeFe(III), ferrihydrite, goethite, and hematite. Unraveling

  14. 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.

  15. Evaluation of weapons-grade mixed oxide fuel performance in U.S. Light Water Reactors using COMETHE 4D release 23 computer code 

    E-Print Network [OSTI]

    Bellanger, Philippe

    1999-01-01

    similar conventional UO? fuel. Weapons-grade MOX behavior. However, MOX fuel rods feature higher fuel centerline temperatures due to a lower thermal conductivity. Moreover, higher diffusion in MOX fuel results in a slightly higher fission gas release...

  16. Department of architecturein portlanD

    E-Print Network [OSTI]

    and stakeholders who are pioneering ecological urbanism. Our community of planners, architects, and civic leaders it easy to understand sustainable principles for creating resilient built environments recognized for educating architects who understand and practice sustainable design. The UO architecture

  17. 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...

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

    Office of Scientific and Technical Information (OSTI)

    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...

  19. A new approach to increasing diversity in engineering at the example of women in engineering 

    E-Print Network [OSTI]

    Schäfer, Andrea

    2006-01-01

    A new initiative to incorporate diversity issues into the common engineering curriculum at the University of Wollongong (UoW) in Australia is outlined and the effect on student awareness quantified. The diversity issues ...

  20. Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques NGSI Research Overview and Update on NDA Techniques

    E-Print Network [OSTI]

    A., V. Mozin, S.J. Tobin, L.W. Cambell, J.R. Cheatham, C.R. Freeman, C.J. Gesh,

    2012-01-01

    considered one of the 17x17 PWR assemblies from the NGSIplutonium signal because in a PWR spent fuel its content isspectra for a single PWR fuel pin with fresh and spent UO 2

  1. On the Disposition of Graphite Containing TRISO Particles and the Aqueous Transport of Radionuclides via Heterogeneous Geological Formations

    E-Print Network [OSTI]

    van den Akker, Bret Patrick

    2012-01-01

    element) 0.225 (compact only) 5.144 Graphite CSNF 21-PWR12-PWR 44-BWR 24-BWR UO2 21 PWR fuel assemblies 12 PWR fuel assemblies, 44

  2. 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...

  3. 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

  4. Analysis of conventional and plutonium recycle unit-assemblies for the Yankee (Rowe) PWR

    E-Print Network [OSTI]

    Mertens, Paul Gustaaf

    1971-01-01

    An analysis and comparison of Unit Conventional UO2 Fuel-Assemblies and proposed Plutonium Recycle Fuel Assemblies for the Yankee (Rowe) Reactor has been made. The influence of spectral effects, at the watergaps -and ...

  5. HFIR Plant Maintenance - August

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

    diffusion limited reaction of U3O8 to UO3. * TGA analysis of the oxidation of actual used fuel under a controlled atmosphere, conducted in REDC Cave B, correlated well with models....

  6. 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 )...

  7. Sandia Energy - EC Publications

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

    assembly. The assembly rods to be used for the tests will not be actual irradiated zirconium alloyUO2-pellet rods. Surrogate rods shall be selected that have similar mass and...

  8. Advanced LWR Nuclear Fuel Development

    Energy Savers [EERE]

    308L Steam separator and dryer: * components, 304 * welds 308L Closure studs: * alloy steel Fuel: * Cladding, Zr-2 * Fuel, UO 2 Source: R. Staehle There are many materials in a...

  9. Laboratory Experiments and their Applicability 

    E-Print Network [OSTI]

    Steinhaus, Thomas; Jahn, Wolfram

    2007-11-14

    In conjunction with the Dalmarnock Fire Tests a series of laboratory tests have been conducted at the BRE Centre for Fire Safety Engineering at the University of Edinburgh (UoE) in support of the large scale tests. These ...

  10. 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.

  11. La neutronique est l'tude de l'etat des neutrons dans la matire et des ractions qu'ils y induisent, en particulier la gnration de puissance dans les coe urs de centrales par la fission de

    E-Print Network [OSTI]

    réacteur 4.5 x x 1.7 m3 Débit d'eau à l'entrée du réacteur 1m3 s-1 Eau 190 °C en entrée 320°C en sortie de placés verticalement dans la cuve Crayon: 272 pastilles de combustible (UO2 ou Mox:PuO2 /UO2 ) Assemblage l'U235 Energie libérée (colle) lors de la fission EF = m c2 techniquement difficile car haute T pour

  12. Energy Frontier Research Center Center for Materials Science of Nuclear Fuels

    SciTech Connect (OSTI)

    Todd Allen

    2014-04-01

    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 anharmonic smoothing technique has enabled quantitative benchmarking of ab initio PDOS simulations. • Direct comparison between anharmonicity-smoothed ab initio PDOS simulations for UO2 and experimental measurements has demonstrated the need for improved understanding of UO2 at the level of phonon dispersion, and, further, that advanced lattice dynamics simulations including finite temperatures approaches will be required for handling this strongly correlated nuclear fuel. • PDOS measurements performed on polycrystalline samples have identified the phonon branches and energy ranges most highly impacted by fission-product and hyper-stoichiometry lattice defects in UO2. These measurements have revealed the broad-spectrum impact of oxygen hyper-stoichiometry on thermal transport. The reduction in thermal conductivity caused by hyper-stoichiometry is many times stronger than that caused by substitutional fission-product impurities. • Laser-based thermo-reflectance measurements on UO2 samples irradiated with light (i.e. He) ions to introduce point defects have been coupled with MD simulations and lattice parameter measurements to determine the role of uranium and oxygen point defects in reducing thermal conductivity. • A rigorous perturbation theory treatment of phonon lifetimes in UO2 based on a 3D discretization of the Brillouin zone coupled with experimentally measured phonon dispersion has been implemented that produces improved predictions of the temperature dependent thermal conductivity. • Atom probe investigations of the influence of grain boundary structure on the segregation behavior of Kr in UO2 have shown that smaller amounts of Kr are present at low angle grain boundaries than at large angle grain boundaries due to the more dense dislocation arrays associated with large angle boundaries; this observation has potentially important ramifications for thermal transport in the high burn-up rim region of light water reactor fuel. • A variable charge interatomic potential has been developed that not only provides an accurate representation of the fluorite UO2 phase, it is further capable of describing continuous stoichiometry changes from UO2 to hyper-stoichiometric UO2+x, to U4O9 and U3O7, and possibly to orthorhombic U3O8. This is the first potential that features many-body effects in all possible interactions (U-U, U-O and O-O) combined with the variable charge. • A theoretical proof has been formulated showing that it is necessary to use the so-called model C phase field approach, consisting of Cahn-Hilliard and Allen-Cahn equations, to describe void evolution in irradiated materials. This work resolved a longstanding literature controversy regarding how to model voids at the mesoscale. • A novel cluster dynamics model has been developed for the nucleation of voids and loops in UO2 under irradiation. This model is important in understanding the defect state of UO2 after irradiation and, more importantly, reveals off-stoichiometric states of irradiated UO2 that are critical for understanding the impact of irradiation on thermal transport. Personnel Successes

  13. 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.

  14. 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).

  15. 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}.

  16. Standard test method for determination of impurities in plutonium: acid dissolution, ion exchange matrix separation, and inductively coupled plasma-atomic emission spectroscopic (ICP/AES) analysis

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This specification covers blended uranium trioxide (UO3), U3O8, or mixtures of the two, powders that are intended for conversion into a sinterable uranium dioxide (UO2) powder by means of a direct reduction process. The UO2 powder product of the reduction process must meet the requirements of Specification C 753 and be suitable for subsequent UO2 pellet fabrication by pressing and sintering methods. This specification applies to uranium oxides with a 235U enrichment less than 5 %. 1.2 This specification includes chemical, physical, and test method requirements for uranium oxide powders as they relate to the suitability of the powder for storage, transportation, and direct reduction to UO2 powder. This specification is applicable to uranium oxide powders for such use from any source. 1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of th...

  17. 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.

  18. Permeability and wet-out characterization of SRIM automotive bumper beams 

    E-Print Network [OSTI]

    Morse, Christopher Todd

    1992-01-01

    types 0/CONTINUOUS HAT LAYERS (12 TOTAL) 0 DEGREE PLY MATERIAL ? TOW WEIGHT (YDS/?) LOOPS/ INCH TONS/ LOO P 0 DEG WEIGHT (OZ/YD 2) MAT TYPE NAT WEIGHT (OZ/YD"2) TOTAL LAYER WEIGHT ACTUAL STIFFNESS RATIO UO1 UO2 UO3 225 450... ACTUAL STIFFNES RATIO WOI WO2 WO3 W(N WO! 450 8 00 225 8 00 225 5 50 225 5 50 225 8 00 I 0 24 20 48 28 16 14 08 20 48 450 225 225 450 450 1 00 ! (Kl 4 50 4 50 7 00 3 84 7 68 I I 52 5 76 89( CO)J I fNUOUS CON( ll4UOUS LONI...

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Source term evaluation for postulated UF{sub 6} release accidents in gaseous diffusion plants -- Summer ventilation mode (non-seismic cases)

    SciTech Connect (OSTI)

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

    1996-12-30

    Computer models have been developed to simulate the transient behavior of aerosols and vapors as a result of a postulated accident involving the release of uranium hexafluoride (UF{sub 6}) into the process building of a gaseous diffusion plant. For the current study, gaseous UF{sub 6} is assumed to get released in the cell housing atmosphere through B-line break at 58.97 kg/s for 10 min and 30 min duration at the Paducah and Portsmouth Gaseous Diffusion Plants. The released UF{sub 6} undergoes an exothermic chemical reaction with moisture (H{sub 2}O) in the air to form hydrogen fluoride (HF) and radioactive uranyl fluoride (UO{sub 2}F{sub 2}) while it disperses throughout the process building. As part of a facility-wide safety evaluation, this study evaluated source terms consisting of UO{sub 2}F{sub 2} as well as HF during a postulated UF{sub 6} release accident in a process building. UO{sub 2}F{sub 2} mainly remains as airborne-solid particles (aerosols), and HF is in a vapor form. Some UO{sub 2}F{sub 2} aerosols are removed from the air flow due to gravitational settling. The HF and the remaining UO{sub 2}F{sub 2} are mixed with air and exhausted through the building ventilation system. The MELCOR computer code was selected for simulating aerosols and vapor transport in the process building. To characterize leakage flow through the cell housing wall, 3-D CFD tool (CFDS-FLOW3D) was used. About 57% of UO{sub 2}F{sub 2} was predicted to be released into the environment. Since HF was treated as vapor, close to 100% was estimated to get released into the environment.

  4. Characterization of Zr-Fe-Cu Alloys for an Inert Matrix Fuel for Nuclear Energy Applications 

    E-Print Network [OSTI]

    Barnhart, Brian A.

    2013-08-09

    Spectroscopy EPMA Electron Probe Microanalyzer FCCI Fuel Clad Chemical Interaction FCMI Fuel Clad Mechanical Interaction IMF Inert Matrix Fuel LFA Light Flash Analysis LLNL Lawrence Livermore National Lab LWR Light Water Reactor MRF Materials... it appealing for use in a nuclear reactor. The UO2 crystal has a fluorite structure of the CaF2 type. At beginning of life (BOL), UO2 is manufactured to be as pure as possible; as such when it is loaded into the fuel bundles it is a single phase and has...

  5. 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.

  6. Trend of Taxes on Agricultural Land in Texas and Distribution of the Farmer's Tax Dollar. 

    E-Print Network [OSTI]

    Buechel, F. A. (Frederick Anthony)

    1925-01-01

    cooperation. The officers of more than one hundred and seventy-five coun- ties responded to the questionnaire with all or part of the information de- sired. It is a source of deep regret to the writer that out of this number, ;he data from only one hundred... pap1 uoyqxodoxd ayq pus xaaX quanbasqns Aue u~ usyq 7161 U! xa~sax3 sea sasodxnd Aqunoa XOJ parha1 uo~podoxd ay& -uo~qsxapysuo:, xapun xaafi y3sa JOJ passaxau! 'qunourz? aqnlosqa ayq se IIaM ss sq3rxqsIp pa01 Aq parha1 xsq p?qo? ayq jo uorpodoxd ayq...

  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. 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.

  9. This journal is c The Royal Society of Chemistry 2010 Chem. Commun., 2010, 46, 91679169 9167 Cubic and rhombohedral heterobimetallic networks constructed from

    E-Print Network [OSTI]

    in the nuclear fuel cycle. U(VI) is typically found in the form of a linear UO2 2+ cation, called uranyl, and four to six additional donor atoms are usually found perpendicular to the uranyl axis yielding of the common oxoanions have been used to bind uranyl, and large families of uranyl oxoanion compounds are now

  10. Reduction and desymmetrisation of the uranyl dication in a macrocyclic framework 

    E-Print Network [OSTI]

    Patel, Dipti

    The transamination reaction between a Schiff base polypyrrolic macrocycle, H4Ltet/oct, where tet = tetramethyl (C38H36N8), oct = octamethyl (C42H44N8), and [UO2(THF)2{N(SiMe3)2}] results in the sole formation of mono ...

  11. Subscriber access provided by Caltech Library Services The Journal of Physical Chemistry A is published by the American Chemical

    E-Print Network [OSTI]

    Goddard III, William A.

    Addition of H 2 O and O 2 to Acetone and Dimethylsulfoxide Ligated Uranyl(V) Dioxocations Christopher M;Addition of H2O and O2 to Acetone and Dimethylsulfoxide Ligated Uranyl(V) Dioxocations Christopher M to form the reduced uranyl cation UO2 + , formally a U(V) species.8 A similar reaction of U2+ with O2

  12. Available at www.sciencedirect.com journal homepage: www.elsevier.com/locate/watres

    E-Print Network [OSTI]

    Burgos, William

    in revised form 3 April 2006 Accepted 5 April 2006 Available online 14 June 2006 Keywords: Solubility Uranyl as uranyl (UO2 2+ ) for acidic condi- tions, and forms hydroxyl and carbonato complexes with increasing p and Grambow, 1998), sorption of uranyl on geological materials (Hsi and Langmuir, 1985; Silva, 1992; Waite et

  13. Radiation Measurements 42 (2007) 10291032 www.elsevier.com/locate/radmeas

    E-Print Network [OSTI]

    2007-01-01

    M) of two uranyl nitrate compounds that have different uranium isotopic concentrations and cells were exposed to either 238U-uranyl nitrate, specific activity 0.33 Ci/g, or DU-uranyl nitrate similar Abbreviations: DU, Depleted uranium; V79, Chinese hamster lung cells; DU-UO2NO3 depleted uranium-uranyl

  14. Journal of Colloid and Interface Science 299 (2006) 4955 www.elsevier.com/locate/jcis

    E-Print Network [OSTI]

    Montes-Hernandez, German

    2006-01-01

    to the removal of uranyl ion (UO2+ 2 ) from aqueous solutions onto synthetic manganese oxide (birnessite-squares method. For initial concentration range 0.2­11.8 µM, the results showed that the uranyl removal process and thermodynamic parameters were calculated, such as maximal removed quantity of uranyl, qr,max, half-removal time

  15. Present address: Department of Geology, 245 Natural History Building, University of Illinois, 1301 W. Green Street, Urbana, Illinois 61801, U.S.A. E-mail address: jmjackso@uiuc.edu.

    E-Print Network [OSTI]

    Jackson, Jennifer M.

    Mineralogist Vol. 39, pp. 187-195 (2001) A RE-EVALUATION OF THE STRUCTURE OF WEEKSITE, A URANYL SILICATE, each of which is part of a nearly linear (UO2)2+ uranyl ion. The uranyl ions (Ur) are further. There are ten silicon atoms, each of which is tetrahedrally coordinated by oxygen atoms. The uranyl polyhedra

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

    SciTech Connect (OSTI)

    baney, Ronald; Tulenko, James

    2012-11-20

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

  17. Computer modeling of the spallation process 

    E-Print Network [OSTI]

    Walker, Wayne Claire

    1978-01-01

    that classically the particle hitting the nucleus is ab- sorbed, thon the cross section for the roverse process becomes c(EA, e) = ao(l ? V/c) for c & V (63) 0 fore &V where uo = sr and V = zz, e /r. The Coulomb field is expressed in terms of. the charge, ze...

  18. UNIVERSITY OF WASHINGTON PERSONAL DATA FORM

    E-Print Network [OSTI]

    Matrajt, Graciela

    UNIVERSITY OF WASHINGTON PERSONAL DATA FORM Name of Employee (Last, First, M.I.) Student Number (If to print this information in the directory? Yes No (County) UoW 1354 (8/01) Contact's Day Phone Contact (e.g. Dr. of Education, Dr. of Science) Employee's Signature Date *The University has requested your

  19. UNIVERSITY OF WASHINGTON PERSONAL DATA FORM

    E-Print Network [OSTI]

    Shlizerman, Eli

    UNIVERSITY OF WASHINGTON PERSONAL DATA FORM DEPARTMENT OF BIOLOGY Name of Employee (Last, First, M Is it okay to print this information in the directory? Yes No (County) UoW 1354 (10/00) Contact's Day Phone (e.g. Dr. of Education, Dr. of Science) Employee's Signature Date *The University has requested your

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

    SciTech Connect (OSTI)

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

    2010-09-01

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

  1. The United States Department of Energy (DOE) has always held the safety and reliability of the nation's nuclear reactor fleet as a top priority. Continual improvements and advancements in nuclear fuels have

    E-Print Network [OSTI]

    of the nation's nuclear reactor fleet as a top priority. Continual improvements and advancements in nuclear fuels have been instrumental in maximizing energy generation from nuclear power plants and minimizing the mechanical properties of uranium dioxide (UO2) for nuclear fuel applications. In an effort to improve

  2. Development of a RELAP5-3D three-dimensional model of a VVER-1000 Nuclear Power Plant for analysis of a large-break loss-of-coolant accident 

    E-Print Network [OSTI]

    Clarno, Kevin Taylor

    2001-01-01

    large-break loss-of-coolant accident (LB LOCA). A validated, one-dimensional control of the nuclear power plant, for the study of the effects of mixed oxide (MOX) fuel, was modified to include a standard fuel loading of UO?. The development...

  3. Molecular EcologY 7996'5' 45A57 TECHNICAL NOT E

    E-Print Network [OSTI]

    Marsh, Helene

    S H T ZoologyDepartment,lamesCookUniaersityofNorthQueensland,Townsaille,NorthQueensland,48L1,Australia was phenol/chloroform extracted (Sambrook et al. 1989), followed by ethanol precipitation (Crouse & Amorese, Brisbane,Queensland,4059,Australia. Tel.: + 61.7 38372989. E-mail: D.Tikel@mailbox.uo.edu.au @ Pge

  4. 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

  5. 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.

  6. The Power of Mesoscale Modeling... Mul$physics mesoscale simula$on provides a powerful tool for designing materials to

    E-Print Network [OSTI]

    Chen, Long-Qing

    , neutronics, geomechanics, reac+ve transport, microstructure modeling, computa+onal fluid in verba+m from Schwen, D., E. Mar/nez, and A. Caro, J. Nuclear Mater (cv) in UO2 fuel. Also shown are the switching func+on h, the order

  7. O and Pb isotopic analyses of uranium minerals by ion microprobe and UPb ages from the Cigar Lake deposit

    E-Print Network [OSTI]

    Fayek, Mostafa

    of Nuclear Engineering and Radiological Sciences, The University of Michigan, 2958A Cooley Building, 2355­30 Am, thus providing relatively accurate information regarding the timing of fluid interactions of migration of uranium and other radionuclides from a spent fuel repository because uraninite, UO2 + x (the

  8. Effective flow surface of porous materials with two populations of voids under internal pressure: I. a GTN model

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    ). Such a microstructure is typical of the highly irradiated uranium dioxide (UO2), a nuclear fuel commonly used in nuclear several studies on the mechanical behavior of highly irradiated nuclear fuels at different scales (Vincent to the effective plastic flow surface of a bi-porous material saturated by a fluid. The material under

  9. This is an author-deposited version published in: http://oatao.univ-toulouse.fr/ Eprints ID: 8615

    E-Print Network [OSTI]

    Mailhes, Corinne

    observed: nonreduced UO2 in the centre, pure metallic uranium on the external layer and an intermediate. Introduction Advanced nuclear fuel cycles are under development worldwide in order to minimise the amount future fuel cycles, recycling of actinides (Cm, Pu, Am, Np) from spent nuclear fuel is required due

  10. This is an author-deposited version published in: http://oatao.univ-toulouse.fr/ Eprints ID: 8615

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    technology for spent metallic fuel is studied in LiCl­KCl where U and Pu are recov- ered on a liquid Cd was partially reduced and three phases were observed: nonreduced UO2 in the centre, pure metallic uranium the electroreduction of solid oxides. 1. Introduction Advanced nuclear fuel cycles are under development worldwide

  11. UNIVERSITY OF WASHINGTON HONORARIA CHECKLIST

    E-Print Network [OSTI]

    Matrajt, Graciela

    UNIVERSITY OF WASHINGTON HONORARIA CHECKLIST FOR INVITED GUEST SPEAKERS PROCUREMENT SERVICES Date Number 35 QUESTIONS? Contact the Tax Office, Phone: 206-616-3003 or Email: taxofc@u.washington.edu If the person IS NOT an invited guest speaker, please use form UoW 1632 available at http://www.washington

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

    SciTech Connect (OSTI)

    Gofryk, K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jaime, M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). National High Magnetic Field Lab. (MagLab)

    2014-12-01

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

  13. Biochemistry Major Research Report Updated February 2014

    E-Print Network [OSTI]

    Cina, Jeff

    Biochemistry Major Research Report Updated February 2014 Biochemistry majors who use their year, hard copy of the report should be turned in to the mailbox for Ken Prehoda (Biochemistry Division Spokesperson) at the UO Chemsitry and Biochemistry office. Students for whom their undergraduate research

  14. Groundwater Chemistry Changes as a Result of CO2 Injection at the ZERT Field Site in Bozeman, Montana

    E-Print Network [OSTI]

    Apps, J.A.

    2010-01-01

    Co +2 Cu + Cd +2 PbSe CrO + AsSe(OH)(SeH) - FeSe MoO 4-2 UOCo +2 Cu + Cd +2 PbSe CrO + AsSe(OH)(SeH) - FeSe MoO 4-2 UO

  15. Modeling the performance of high burnup thoria and urania PWR fuel

    E-Print Network [OSTI]

    Long, Yun, 1972-

    2002-01-01

    Fuel performance models have been developed to assess the performance of ThO?-UO? fuels that can be operated to a high burnup up to 80-100MWd/kgHM in current and future Light Water Reactors (LWRs). Among the various issues ...

  16. APOLLO 16 VOICE TRANSCRIPT PERTAINING TO THE GEOLOGY OF THE LANDING SITE

    E-Print Network [OSTI]

    Rathbun, Julie A.

    * * *: {( APOLLO 16 VOICE TRANSCRIPT PERTAINING TO THE GEOLOGY OF THE LANDING SITE #12;- APOLLO 16 VOICE TRANSCRIPT Pertaining to the geology of the landIng site by N.G. Bai loey and G.E. Ulrich U.s. Geol:ogical Survey Branch of Astrogeology F]agstaff~ Arizona 1915 #12;FORM NTlS·315 UO-70

  17. 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.

  18. Please cite this article in press as: Shuffler, C., et al., Thermal hydraulic analysis for grid supported pressurized water reactor cores. Nucl. Eng. Des. (2009), doi:10.1016/j.nucengdes.2008.12.028

    E-Print Network [OSTI]

    Malen, Jonathan A.

    2009-01-01

    (average for UO2 and centerline/peak for U­ZrH1.6), and fuel rod vibrations and wear. Transient limits differences exist between the two fuel types for designs limited by rod vibrations and wear, because be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure

  19. VII. SOLAR RADIATION DATA COMPARISONS In this section some of the solar radiation data

    E-Print Network [OSTI]

    Oregon, University of

    18 VII. SOLAR RADIATION DATA COMPARISONS In this section some of the solar radiation data gathered by the UO Solar Monitoring Network is presented in tabular and pictorial form and related to similar information from other Western U.S. sites. A comparison of the amount of incident solar radiation is made us

  20. 2013-14 Comparison of Faculty Salaries BY SCHOOL/COLLEGE/DIVISION AND RANK WITH SALARY COMPRESSION, WEIGHTED AND NEW HIRE AVERAGE SALARIES

    E-Print Network [OSTI]

    2013-14 Comparison of Faculty Salaries BY SCHOOL/COLLEGE/DIVISION AND RANK WITH SALARY COMPRESSION, WEIGHTED AND NEW HIRE AVERAGE SALARIES Based on data from the Fall 2013 AAU Data Exchange (AAUDE) Faculty Salary Survey UNIVERSITY OF OREGON AND ALL PARTICIPATING PUBLIC AAU INSTITUTIONS UO Avg Salary AAU

  1. Open Archive Toulouse Archive Ouverte (OATAO) OATAO is an open access repository that collects the work of Toulouse researchers and

    E-Print Network [OSTI]

    Mailhes, Corinne

    salts (LiCl or CaCl2), for different purposes: high purity Si production [2], spent nuclear fuel processing (reduc- tion of rare earth oxides [3], UO2 [4], U3O8 [5], MOX [6], spent fuel [7], etc.), pure product is often polluted by carbides, due to the reduction of carbonates CO3 2- formed by the reaction

  2. Characterizing solution and solid-phase amorphous uranyl silicates q

    E-Print Network [OSTI]

    Illinois at Chicago, University of

    . Skanthakumar a , D. Gorman-Lewis a , M.P. Jensen a , K.L. Nagy b a Chemistry Division, Argonne National 2007 Elsevier Ltd. All rights reserved. 1. INTRODUCTION Dissolved uranium, as the uranyl ion UO2 2þ relevant conditions is severely ham- pered by its chemistry in near neutral or basic groundwater, where

  3. Diffusion model of the non-stoichiometric uranium dioxide

    SciTech Connect (OSTI)

    Moore, Emily, E-mail: emily.moore@cea.fr [CEA Saclay, DEN-DPC-SCCME, 91191 Gif-sur-Yvette Cedex (France); Guéneau, Christine, E-mail: christine.gueneau@cea.fr [CEA Saclay, DEN-DPC-SCCME, 91191 Gif-sur-Yvette Cedex (France); Crocombette, Jean-Paul, E-mail: jean-paul.crocombette@cea.fr [CEA Saclay, DEN DEN, Service de Recherches de Métallurgie Physique, 91191 Gif-sur-Yvette Cedex (France)

    2013-07-15

    Uranium dioxide (UO{sub 2}), which is used in light water reactors, exhibits a large range of non-stoichiometry over a wide temperature scale up to 2000 K. Understanding diffusion behavior of uranium oxides under such conditions is essential to ensure safe reactor operation. The current understanding of diffusion properties is largely limited by the stoichiometric deviations inherent to the fuel. The present DICTRA-based model considers diffusion across non-stoichiometric ranges described by experimentally available data. A vacancy and interstitial model of diffusion is applied to the U–O system as a function of its defect structure derived from CALPHAD-type thermodynamic descriptions. Oxygen and uranium self and tracer diffusion coefficients are assessed for the construction of a mobility database. Chemical diffusion coefficients of oxygen are derived with respect to the Darken relation and migration energies of defects are evaluated as a function of stoichiometric deviation. - Graphical abstract: Complete description of Oxygen–Uranium diffusion as a function of composition at various temperatures according to the developed Dictra model. - Highlights: • Assessment of a uranium–oxygen diffusion model with Dictra. • Complete description of U–O diffusion over wide temperature and composition range. • Oxygen model includes terms for interstitial and vacancy migration. • Interaction terms between defects help describe non-stoichiometric domain of UO{sub 2±x}. • Uranium model is separated into mobility terms for the cationic species.

  4. Uranium Oxide as a Highly Reflective Coating from 150-350 eV

    E-Print Network [OSTI]

    Hart, Gus

    1 Uranium Oxide as a Highly Reflective Coating from 150-350 eV Richard L. Sandberg, David D. Allred.byu.edu ABSTRACT We present the measured reflectances (beamline 6.3.2, ALS at LBNL) of naturally oxidized uranium incidence. These show that uranium, as UO2, can fulfill its promise as the highest known single surface

  5. Quantification of water content and speciation in natural silicic glasses (phonolite, dacite, rhyolite) by

    E-Print Network [OSTI]

    Quantification of water content and speciation in natural silicic glasses (phonolite, dacite Sciences de la Terre d'Orléans, CNRS-UO, Orléans, France Abstract The determination of total water content dissolved water content and height or area of the H2OT Raman band. Accuracy of deconvolution procedure

  6. Origin of Low Thermal Conductivity in Nuclear Fuels Quan Yin and Sergey Y. Savrasov

    E-Print Network [OSTI]

    Savrasov, Sergej Y.

    Origin of Low Thermal Conductivity in Nuclear Fuels Quan Yin and Sergey Y. Savrasov Department.41.Bm Today's nuclear fuels are based on 235 U and 239 Pu ele- ments where in a typical setup, a nuclear, the thermal conductivity of UO2 is very low, and the search for alternative materials continues

  7. Pacific Northwest Solar Radiation Data

    E-Print Network [OSTI]

    Oregon, University of

    Pacific Northwest Solar Radiation Data UO SOLAR MONITORING LAB Physics Department -- Solar Energy Center 1274 University of Oregon Eugene, Oregon 97403-1274 April 1, 1999 #12;Hourly solar radiation data can be obtained from the University of Oregon Solar Moni- toring Laboratory after obtaining permission

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

    SciTech Connect (OSTI)

    Valderrama, Billy [Univ. of Florida, Gainesville, FL (United States). Dept. of Materials Science and Engineering; He, Lingfeng [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Henderson, Hunter B. [Univ. of Florida, Gainesville, FL (United States). Dept. of Materials Science and Engineering; Pakarinen, Janne [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Jaques, Brian [Boise State Univ., Boise, Idaho (United States). Dept. of Materials Science and Engineering; Gan, Jian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Butt, Darryl P. [Boise State Univ., Boise, Idaho (United States). Dept. of Materials Science and Engineering; Allen, Todd R. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Idaho National Lab. (INL), Idaho Falls, ID (United States); Manuel, Michele V. [Univ. of Florida, Gainesville, FL (United States). Dept. of Materials Science and Engineering

    2014-12-01

    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.

  9. GEOBULLETIN SEpTEmBEr 19Th

    E-Print Network [OSTI]

    Carlson, Anders

    on the oxidation state of uranium, therefore understanding the mechanisms of UO2 oxidative corrosion is essential-classical diffusion is driven by electron transfer from multiple uranium atoms to each interstitial #12;GEOBULLETIN Synchrotron Tools for Geoscientists GeoSoilEnviroCARS (GSECARS) is a national synchrotron x-ray user facility

  10. Policy Paper 49: The Military Balance in the Middle East: An Executive Summary

    E-Print Network [OSTI]

    Cordesman, Anthony H.

    1999-01-01

    experiments in uranium enrichment and centrifuge technologyand centrifuge enrichment, purifies UO2, converts UF6, and fabricates fuel for weapons purpose. There is uraniumcentrifuge method • Al Jesira (Mosul)—mass production of UCL 4 • Al Qaim—phosphate plant for production of U308 • Akashat—uranium

  11. 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.

  12. 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.

  13. Crystallisation Within Simulated High Level Waste Borosilicate Glass Peter B. Rose, Michael I. Ojovan, Neil C. Hyatt and William E. Lee

    E-Print Network [OSTI]

    Sheffield, University of

    . INTRODUCTION In the UK, high level radioactive waste (HLW) arising from the reprocessing of spent nuclear fuel produced from waste arising from the reprocessing of Magnox nuclear fuel. Table II details the composition and La) as well as a Si-rich phase. 75/25 glass, comprising a blend of reprocessing waste derived from UO

  14. 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.

  15. 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.

  16. Characteristics of a multiple disk pump with turbulent rotor flow 

    E-Print Network [OSTI]

    Roddy, Patrick James

    1985-01-01

    of these turbomachines until the early sixties when Rice [4] designed and tested a multiple disk pump and compressor. Since then, the per- formance of these turbomachines operating with laminar rotor flow has been well documented [1-6, 7-12]. In the seventies, Bakke... the Disk Radius (LT'8) [ (8 + |/)/T + ( aH/8)]2 2T/T ruotq. znba [gy] yyzqoznqg g~otTdxa aqua. go surzag u) passaudxa aq uzo zoq. ozg uoTgo)zg 9uxuuzg aqua '(gT'2) uoTgznba uZ (9T'2) (au 2/1 2 2 ( (n) + (& ? &U)) = sz r(s)p aqua Suoyz guxod fiuz gz...

  17. Partial characterization of [alpha]-D-mannosidase deficient mutants in Saccharomyces cerevisiae 

    E-Print Network [OSTI]

    Harper, Linda Louise

    1981-01-01

    &Go Lo Lqouo IW: goaI'qnS uo CeW I86 I uaqwaoaO 33N3IDS 30 IJ3lSVW go aaubap aqua uog guawauLnbau aqua go guawIIigIng Ieigued ui Egisuan IuO Wgy sexal go aGaIIog agenpeug aqua og paggiwqnS 'd3dIJ'dH 3SIA01 VONI1 sisaql y 3It IS IA3N30 S3...VIS IR3)J30 S30AIJOJJ'dHOO'dS lJI SlNb'3lllJ J. N3IOI330 3SVOISONN'dlJ-0- 30 NOIJ. VZIJJ3J. O'JJJVHO 1@I J. 'dVd ue[[ aqua go [[e q. eqg pageo[puL suieugs quegnw uaqgo uaaqunog aqua pue [ ueg woug payonugsuoo suieuqs pLo[d ip uo pawuoguad saipnqs...

  18. Characterization of microchip capillary electrophoresis systems coupled to microdialysis for on-line monitoring

    E-Print Network [OSTI]

    Nandi, P.; Suzuki, N.; Lunte, Susan M.

    2006-10-25

    secs 5 secs F l uoresce nce U n i t Minute 7.95 mM Arg 2 4 6 8 10 12 Fl uor escence U n i t Minute 035810 Fl uor e scence U n i t Minute CHO CHO RNH 2 -X+ + NR X Naphthalene?2,3?dicarboxaldehyde ?Reduction/?absence?of?electrokinetic sampling...01234 250 nM 100 nM 25 nM Blank F l uo re sce nc e U n i t Minute F l uo re sce nc e U n i t Microdialysis DEVELOPMENT?OF?A?MICRODIALYSIS?MICROCHIP?CE?SYSTEM? FOR?ON?LINE?MONITORING Pradyot Nandi,?Nobuyuki?Suzuki,?Susan?M.?Lunte Department...

  19. Multilevel modeling of social interactions and mood in lonely and socially connected individuals: The MacArthur Social Neuroscience Studies

    E-Print Network [OSTI]

    Hawkley, L. C.; Preacher, Kristopher J.; Cacioppo, J. T.

    2007-01-01

    SUSIl asoqr aM '(OOOZ ''le ta qseqsel) NIM'lhtr puu (0002 'uop8uo3 4 '3uoaq3 'qsnq-uapneg '>lfrg) rutt se qrns sa8e>pud 611141 parpJrpap puu (0'ZI) SSdS pue (fS'S) feUStf ,o suorsral raivrau Surpnpur 'sa8eped aremryos Ie)rtsrlets lera^as ur PaPnPur 3rE sla...; '3urilas ro raulrBd uorlJEJatur Jo aJror{) lu?dDru?d uo s}urEJlsuoJ leluarurradxg'-{rrprle,r lerrSolora ur,rno1,{11errd,tr aJe pooru ;o sa8ueqrxa IeIJos Jo suorlelndtueur 'r(lrlesner JaJur ol paJrnbal aruaparard prod -ruar ar{r r...

  20. An experimental study of jet impingement on a circular cylinder 

    E-Print Network [OSTI]

    Potts, Dennis Wayne

    1984-01-01

    profile for U = 550 ft/s, x/d = 30 (radial 2) 1 30 85 Wall jet turbulence profile for U = 550 ft/s, x/d = 30 ( radial 4) 131 86 Wall jet turbulence profile for Uo5 50 f t/ s, x/ d= 30(ax i al rad i al ) 132 Spread r ate of the wall jet along the axis... of the cylinder a compar ed to a flat plate . . . . 140 95 Velocity scale of the wall jet for U = 400 ft/s, x/d = 7 141 96 Velocity scale of the wall jet for U = 400 ft/s, x/d = 15 142 97 Velocity scale of tne wall jet for Uo 4 0 0 f t / s x / d 3 0 143 9S...

  1. A new harmonic analyzer 

    E-Print Network [OSTI]

    Tchamran-Savehi, Abbas

    1958-01-01

    (3) A I f(x) Cos nx dx n o 2fl (4) B ~ ? f(x) Sin nx dx n or U~S (7) A 2 g f( U + x ) Cos n( U Q x ) S ~ U~o V~S V S (8) Bn - N f( U + x ) Sin n( U + x ) S ~ U~o are concerned require two operations: a) Multiplication of Cos nx or Sin...(x) = ? +AI Cos x + A2 Cos 2x + A3 Cos 3x +. . . + A Cos nx +. . . 1 2 3 ' ' n + Bl Sin x t B2 Sin 2x & B3 Sin 3x +. . . + Bn Sin nx +. . . or A (2) f(x) ~ ? + ( A Cos nx + Bn Sin nx ) where: A 1 n 2)I f(x) Cos nx dx (4) B 1 fl f(x) Sin nx dx 2 l1...

  2. Adequacy of the 123-group cross-section library for criticality analyses of water-moderated uranium systems

    SciTech Connect (OSTI)

    Parks, C.V.; Wright, R.Q.; Jordan, W.C. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    In a recent criticality analysis for an array of water-moderated packages containing highly enriched uranium, the 123-group cross-section library in the SCALE system was observed to have a nonconservative discrepancy of approximately 3 to 3.5% when compared with more recently developed libraries. A simple representative system of UO{sub 2}F{sub 2}-H{sub 2}O was used to identify that the problem results from a lack of resonance data for {sup 235}U. Only a single set of self-shielded cross sections, most likely corresponding to a water-moderated infinite dilute system, was provided with the original data. The UO{sub 2}F{sub 2}-H{sub 2}O study indicates that this limitation may cause nonconservative discrepancies as high as 5.5% for some water-moderated, highly enriched uranium systems. Characteristics of the systems where the discrepancy is evident are identified and discussed.

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

    SciTech Connect (OSTI)

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

    1984-05-01

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

  4. 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. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    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.

  5. 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.

  6. 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.

  7. 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.

  8. 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).

  9. 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.

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

    SciTech Connect (OSTI)

    Collin, Blaise

    2014-08-01

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

  11. AGR-2 Irradiation Test Final As-Run Report, 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.

  12. 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.

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

    SciTech Connect (OSTI)

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

    1989-12-31

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

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

    SciTech Connect (OSTI)

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

    1989-01-01

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

  15. The Origin of the Whig Party

    E-Print Network [OSTI]

    McCluggage, Robert Tyler

    1912-06-04

    strange, tne wonder lies in xne iacx %ua.z so many were iea xo uelieve contrariwise. In tracing movements wnicn imaiiy grew into aistmctive parties, ana ultimately into xne wnig party, it is necessary uo study tms perioa 01 "gooa leeamgs" so pregnant... giving information concerning the suppression of tne establisnment at Amelia island and trans- 3 mitting the papers concerned tnerewitn. Adams tnought tnat tne measures in relation to Amelia island, South America, and Spain weee "laying the foundation...

  16. POT FISHING IN THE VIRGIN ISLANDS .T. B.. Sylvest 'I' and . F. Damm nn

    E-Print Network [OSTI]

    '13\\,,011' '-. _ . _ ............. . ,0 Fig. 1 - Lo, lLon i1 f t..." u~.o lslJ.ods. St. ThlS stud) "as nl.llJ..:: In (:0\\.)t-~rJ.u0n "1, but wood ( r p sapllngs) IS gen rally us d. I'unn Is ar oval and ar> t >rmlnated 10 a right angle facing th

  17. JOURNALDEPHYSIQUEIV ColloqueC7,supplkmentauJournaldePhysique111,Volume3,novembre1993

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    ) connected to an image analyzer system, made of a host computer (SUN 3.140) and an Imaging Technology image~alaas'slreqapalou-rod.paquasa~daqITrMs~oqroede3pa-ralurs aqqpuesuTr.Juaa.xBaqq'lap~odayqoqpaqqals7Tnsa-r~e3~801oqd~ourauosA~uo E'C -saps)sq~edpaJaTUrspaqsr~od30saseurr~3suo~'~3a~aK~epuo3as #12;Granulometric and granulomorphic distributions on projected

  18. 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.

  19. Nuclear carrier business volume projections, 1980-2000

    SciTech Connect (OSTI)

    Lebo, R.G.; McKeown, M.S.; Rhyne, W.R.

    1980-05-01

    The expected number of shipments of commodities in the nuclear fuel cycle are projected for the years 1980 thru 2000. Projections are made for: yellowcake (U/sub 3/O/sub 8/); natural, enriched and reprocessed uranium hexafluoride (UF/sub 6/); uranium dioxide powder (UO/sub 2/); plutonium dioxide powder (PuO/sub 2/); fresh UO/sub 2/ and mixed oxide (MOX) fuel; spent UO/sub 2/ fuel; low-level waste (LLW); transuranic (TRU) waste; high-activity TRU waste; high-level waste (HLW), and cladding hulls. Projections are also made for non-fuel cycle commodities such as defense TRU wastes and institutional wastes, since they also are shipped by the commercial transportation industry. Projections of waste shipments from LWRs are based on the continuation of current volume reduction and solidification techniques now used by the utility industry. Projections are also made based on a 5% per year reduction in LWR waste volume shipped which is assumed to occur as a result of increased implementation of currently available volume reduction systems. This assumption results in a net 64% decrease in the total waste shipped by the year 2000. LWR waste shipment projections, and essentially all other projections for fuel cycle commodities covered in this report, are normalized to BWR and PWR generating capacity projections set forth by the Department of Energy (DOE) in their low-growth projection of April, 1979. Therefore these commodity shipment projections may be altered to comply with future changes in generating capacity projections. Projected shipments of waste from the reprocessing of spent UO/sub 2/ fuel are based on waste generation rates proposed by Nuclear Fuels Services, Allied-General Nuclear Services, Exxon Nuclear, and the DOE. Reprocessing is assumed to begin again in 1990, with mixed oxide fresh fuel available for shipment by 1991.

  20. Generic report on health effects for the US Gaseous Diffusion Plants. Sect. 8, Pt. 1

    SciTech Connect (OSTI)

    Just, R.A.; Emler, V.S.

    1984-06-01

    Toxic substances present in uranium enrichment plants include uranium hexafluoride (UF/sub 6/), hydrogen fluoride (HF), uranyl fluoride (UO/sub 2/F/sub 2/), chlorine (Cl/sub 2/), chlorine trifluoride (ClF/sub 3/), fluorine (F/sub 2/), uranium tetrafluoride (UF/sub 4/), and technetium (Tc). The current knowledge of the expected health effects of acute exposures to these substances is described. 10 references, 2 figures, 6 tables. (ACR)

  1. Accommodation of Uranium into the Garnet Structure Sergey V.Yudintsev1

    E-Print Network [OSTI]

    Utsunomiya, Satoshi

    Accommodation of Uranium into the Garnet Structure Sergey V.Yudintsev1 , Marya I. Lapina1 for uranium, the CaO ­ Fe2O3 ­ Al2O3 ­ SiO2 ­ ZrO2 ­ Gd2O3 ­ UO2 system was studied. Experiments were- corporation of U was found to be greatly dependent on the phase composition. Uranium content decreased from 18

  2. 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.

  3. Measurement of the ratio [B(D(0)?K*(?)e(+)?e)] / [B(D(0)?K(?)e(+)?e)

    E-Print Network [OSTI]

    Baringer, Philip S.

    1991-12-01

    as kaons Background from lepton fakes Background from BB events Background from K* l v Net yield Efticiency for E l+v a (Do~~-l+ v) Electrons 914+30 140+11+21 90+1+21 43+6+21 57+18+15 584+37+39 (7.2+0.2)%%uo 3.8+0.3+0.6% Muons 440+21 51+7+8 121+2+24 18...+4+9 19+6+5 23 1+23+27 (3.2+0.2)%%uo 3.3+0.3+0.6%%uo MEASUREMENT OF THE RATIO B(D ~K* e+v, )/B(D —+K e+v, ) 3399 both K* m e+v, and K* m e+v, . The efficiency for K* ir e+v, to fake K' e+v, is 70%, while the efficiency for K* m e+v, to fake K* e+v, is zero...

  4. Structure, depositional environment, and pressure characteristics of the Vicksburg formation: Javelina and East McCook fields, Hidalgo County, Texas 

    E-Print Network [OSTI]

    Hastings, John Olcott

    1984-01-01

    seM L La& Z-L?q LeM ueqasa[M aq& uaqM ' L!0 L laqS gno-w~eg e uo wna Lougad Esca Lg Eq pa~anoos ip see p La ib oloogoW Wse3 '0/6 [ 'qsn6n[f us 'adage [ sueaE uaag~nog (2 a LqeJ. ) Eep/33W 00[K Jog OLL' pues ?X?5unqsoloin ua~o [ aqua ?(w 6[OS o...62) 1k OSL6 wouy a6ueu uo[qM sqgdap ~e (?2? 0+ ?6? ) sauoqspues 6unqsqo lh ua~ol u i sunooo 'uaAamoq ' uo igonpoud (w 2LSV) 4k 000'9 l "a" o o'4 (w 068 l) 1k 0029 wo"4 q&dap a5ueu eave qoogoW gse3/eu[[anep aug ui sauogspues 6unqsqv[n 'a1esuapuo...

  5. 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.

  6. Synthesis of triglyceride by the intestinal mucosa 

    E-Print Network [OSTI]

    Buell, George Christopher

    1958-01-01

    c3r pfdD3uc.2 aucR C3rf 2so2py-r- c3uc ar- tpf2resp -srP osc cu6r Duec .o c3r erPfoc3rP.P sa aucP .o c3r .ocrPc.orR Sucre .c 7uP P3s7o if Tr.Pre rc up ?p?? c3uc -.3f-esnfu2rcsor rPcreP duf ir c3r Der2yePse sa c3r tpf2resp sa c3r ce.tpf2re.-rP er...rcsor D3sPD3ucr uo- SU? Utpf2resD3sPD3ucrR Sucre rb.-ro2r c3uc c3r Der2yePse duf p.r .o c3.P drcuisp.2 Duc3 7uP Desb.-r- if c3r 7se6 sa +seoiret uo- le.2re ?Yq? 18?R C3rf 7rer uipr cs rPcre.af SUUtpf2resD3sPD3ucr 7.c3 1s8 rPcreP sa psot 23u.o auccf u...

  7. 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.

  8. Source term evaluation for UF{sub 6} release event in feed facility at gaseous diffusion plants

    SciTech Connect (OSTI)

    Kim, S.H.; Taleyarkhan, R.P.

    1997-01-30

    An assessment of UF{sub 6} release accidents was conducted for the feed facility of a gaseous diffusion plant (GDP). Release rates from pig-tail connections were estimated from CYLIND code predictions, whereas, MELCOR was utilized for simulating reactions of UF{sub 6} with moisture and consequent transport of UO{sub 2}F{sub 2} aerosols and HF vapor through the building and to the environment. Two wind speeds were utilized. At the high end (Case 1) a wind speed of {approximately} 1 m/s (200 fpm) was assumed to flow parallel to the building length. At the low end (Case 2) to represent stagnant conditions a corresponding wind speed of 1 cm/s (2 fpm) was utilized. A further conservative assumption was made to specify no closure of crane and train doors at either end of the building. Relaxation of this assumption should provide for additional margins. Results indicated that, for the high (200 fpm) wind speed, close to 66% of the UO{sub 2}F{sub 2} aerosols and 100% of the HF gas get released to the environment over a 10-minute period. However, for the low (2 fpm) wind speed, negligible amount ({approximately} 1% UO{sub 2}F{sub 2}) of aerosols get released even over a 2 hour period.

  9. Chemistry of gaseous lower halides of uranium. Technical progress report, 1 September 1979-1 April 1980

    SciTech Connect (OSTI)

    Hildenbrand, D.L.

    1980-04-15

    The gaseous uranium species UF, UF/sub 2/, UF/sub 3/, and UF/sub 4/ were generated in effusion cell beams by vaporization of UF/sub 4/(s) under reducing conditions, and they were identified and studied by mass spectrometry. From extensive second-law studies of reaction equilibria involving these species and several reaction partners used as reference standards, the individual bond dissociation energies and standard enthalpies of formation of the U-F species were derived. Reaction entropies derived from the slope data indicate that the electronic entropies of the U-F species are substantial, and are comparable to or larger than that of atomic uranium. Additional thermochemical measurements were made to establish the properties of several Ag and Cu monohalides that have been or will be used as reference standards in the uranium halide measurements. From studies of the sublimation and decomposition of uranyl fluoride, UO/sub 2/F/sub 2/(s), the enthalpy of sublimation of UO/sub 2/F/sub 2/(g), has been determined, and another gaseous oxyfluoride, UOF/sub 4/(g), has been tentatively identified. The gaseous products of decomposition of UO/sub 2/F/sub 2/(s) observed by mass spectrometry differ from those postulated by other investigators, indicating that the mechanism of decomposition has not been clearly established. A search of the thermochemical literature on uranium halides has been completed.

  10. 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)

  11. 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.

  12. Standard specification for sintered gadolinium oxide-uranium dioxide pellets

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2008-01-01

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

  13. Criticality Safety Code Validation with LWBR’s SB Cores

    SciTech Connect (OSTI)

    Putman, Valerie Lee

    2003-01-01

    The first set of critical experiments from the Shippingport Light Water Breeder Reactor Program included eight, simple geometry critical cores built with 233UO2-ZrO2, 235UO2-ZrO2, ThO2, and ThO2-233UO2 nuclear materials. These cores are evaluated, described, and modeled to provide benchmarks and validation information for INEEL criticality safety calculation methodology. In addition to consistency with INEEL methodology, benchmark development and nuclear data are consistent with International Criticality Safety Benchmark Evaluation Project methodology.Section 1 of this report introduces the experiments and the reason they are useful for validating some INEEL criticality safety calculations. Section 2 provides detailed experiment descriptions based on currently available experiment reports. Section 3 identifies criticality safety validation requirement sources and summarizes requirements that most affect this report. Section 4 identifies relevant hand calculation and computer code calculation methodologies used in the experiment evaluation, benchmark development, and validation calculations. Section 5 provides a detailed experiment evaluation. This section identifies resolutions for currently unavailable and discrepant information. Section 5 also reports calculated experiment uncertainty effects. Section 6 describes the developed benchmarks. Section 6 includes calculated sensitivities to various benchmark features and parameters. Section 7 summarizes validation results. Appendices describe various assumptions and their bases, list experimenter calculations results for items that were independently calculated for this validation work, report other information gathered and developed by SCIENTEC personnel while evaluating these same experiments, and list benchmark sample input and miscellaneous supplementary data.

  14. 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.

  15. 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.

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

    SciTech Connect (OSTI)

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

    2011-08-01

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

  17. 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].

  18. 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.

  19. 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.

  20. 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.

  1. Cation–cation interactions and cation exchange in a series of isostructural framework uranyl tungstates

    SciTech Connect (OSTI)

    Balboni, Enrica [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, Notre Dame, IN 46556 (United States); Burns, Peter C., E-mail: pburns@nd.edu [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, Notre Dame, IN 46556 (United States); Department of Chemistry and Biochemistry, University of Notre Dame, Notre Dame, IN 46556 (United States)

    2014-05-01

    The isotypical compounds (UO{sub 2}){sub 3}(WO{sub 6})(H{sub 2}O){sub 5} (1), Ag(UO{sub 2}){sub 3}(WO{sub 6})(OH)(H{sub 2}O){sub 3} (2), K(UO{sub 2}){sub 3}(WO{sub 6})OH(H{sub 2}O){sub 4} (3), Rb(UO{sub 2}){sub 3}(WO{sub 6})(OH)(H{sub 2}O){sub 3.5} (4), and Cs(UO{sub 2}){sub 3}(WO{sub 6})OH(H{sub 2}O){sub 3} (5) were synthesized, characterized, and their structures determined. Each crystallizes in space group Cc. (1): a=12.979 (3), b=10.238 (2), c=11.302 (2), ?=102.044 (2); (2): a=13.148 (2), b=9.520 (1), c=11.083 (2), ?=101.568 (2); (3): a=13.111 (8), b=9.930 (6), c=11.242 (7), ?=101.024 (7); (4): a=12.940 (2), b=10.231 (2), c=11.259(2), ?=102.205 (2); (5): a=12.983 (3), b=10.191 (3), c=11.263 (4), ?=101.661 (4). Compounds 1–5 are a framework of uranyl and tungsten polyhedra containing cation–cation interactions. The framework has three symmetrically distinct U(VI) cations, one tungsten, sixteen to eighteen oxygen atoms, and in 2–5, one monovalent cation. Each atom occupies a general position. Each U(VI) cation is present as a typical (UO{sub 2}){sup 2+} uranyl ion in an overall pentagonal bipyramidal coordination environment. Each pentagonal bipyramid shares two equatorial edges with two other pentagonal bipyramids, forming a trimer. Trimers are connected into chains by edge-sharing with WO{sub 6} octahedra. Chains are linked through cation–cation interactions between two symmetrically independent uranyl ions. This yields a remarkably complex system of intersecting channels that extend along [0 0 1] and [?1 1 0]. The cation exchange properties of 2 and 3 were characterized at room temperature and at 140 °C. - Graphical abstract: Chains of uranium and tungsten polyhedra are connected into a three dimensional framework by cation–cation interactions occurring between two symmetrically independent uranyl pentagonal bipyramids. Monovalent cations present in channels within the structure can be exchanged by room temperature or mild hydrothermal treatments. The framework of these compounds is robust to cation exchange and heat. (yellow polyhedra=uranium pentagonal bipyramids; blue polyhedra=tungsten octahedral, purple balls=K; yellow balls=Na; grey balls=Tl). - Highlights: • Five isostructural uranyl tungstates compounds were synthesized hydrothermally. • The structures consist of a chains of uranium and tungstate polyhedral. • Chains are connected into a framework by cation–cation interactions. • Cation exchange does not alter the structural integrity of the compounds. • Cation exchange was successful at room temperature and mild hydrothermal conditions.

  2. High temperature behavior of metallic inclusions in uranium dioxide

    SciTech Connect (OSTI)

    Yang, R.L.

    1980-08-01

    The object of this thesis was to construct a temperature gradient furnace to simulate the thermal conditions in the reactor fuel and to study the migration of metallic inclusions in uranium oxide under the influence of temperature gradient. No thermal migration of molybdenum and tungsten inclusions was observed under the experimental conditions. Ruthenium inclusions, however, dissolved and diffused atomically through grain boundaries in slightly reduced uranium oxide. An intermetallic compound (probably URu/sub 3/) was formed by reaction of Ru and UO/sub 2-x/. The diffusivity and solubility of ruthenium in uranium oxide were measured.

  3. 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.

  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. Particle Decay from Giant Resonance Region of Ca-40 

    E-Print Network [OSTI]

    Youngblood, David H.; Bacher, A. D.; Brown, D. R.; Bronson, J. D.; Moss, JM; Rozsa, C. M.

    1977-01-01

    . These projections are the type in [Fig. 4(c)f. The GQR position, n threshold, and & decay thresholds are indicated. J"=2' 14-MeV group is evident in the spectrum taken at 8?=120 (g. -175') the GQR peak is con- spicuously absent (this is apparent even... thickness in the o. stacks, and the upper limit determined by window settings in the unidentified events (except at 8??=120; 135'). For the region of excitation from 10.8-12.7 MeV where E (c.m. ) is -5.7 MeV, 70%%uo of the decay is accounted...

  6. Direct Observation of Hyperfine Quenching of the (2)3p0 Level in Helium-Like Nickel 

    E-Print Network [OSTI]

    Dunford, R. W.; Liu, C. J.; Last, J.; Berrahmansour, N.; Vondrasek, R.; Church, David A.; Curtis, L. J.

    1991-01-01

    VOLUME 44, NUMBER 1 I JULY 1991 Direct observation of hyperfine quenching of the 2 I'o level in heliumlike nickel R. W. Dunford, C. J. Liu, J. Last, N. Berrah-Mansour, and R. Vondrasek Physics Division, Argonne National Laboratory, Argonne, Illinois... between Ni and 'Ni is readily observed in our experiment. The highly ionized nickel beams were obtained from the Argonne Tandem Linac (ATLAS). The ion source was charged with enriched isotopes of 'Ni (42 at. %%uo )an d Ni (58 at. %). First the Linac...

  7. e<, :,.".,, AQUATIC PLANT CONTROL

    E-Print Network [OSTI]

    US Army Corps of Engineers

    Final Report Approved For Pubiic ~lelease. Dlstrlbut!or\\ Unllfl1l1e(J Prepareo for DEPARTMENT'i\\C'lJddp lO lUt1UJ~)SH,nUl) IPL)ljjO ut' ;"J:i\\ll~)\\jnJ lOll ~;,j()P ~aUJPU apr;] jU UO')"l] 'S(isodliIl1 iFICATION OF THIS PAGE REPORT DOCUMENTATION PAGE IForm Approved OMS No. 0704-0188 la. REPORT SECURITY CLASSIFICATION 1b

  8. A contrastive study of the phonetics of Changsha dialect and Beijing Mandarin 

    E-Print Network [OSTI]

    Wu, Yiqiang

    1989-01-01

    vowels and vowel combinations in CD. Nine vowels of BM are absent from CD. They include /uo, ou, lou, a~, ]at[, uat], atj, irj, uatj/. Eleven other vowels exist only in CD: /ya, io, tti/, yai, yei, t[u, i gu, o, ie, ye/ and /ttn/ (See the table..., and it is written as /[/. TABLE 5 The Vowel Combination of CD and BM o en half close close front front back front back Un-- Un- R- Un- R-- U- R- yy Uu open ll vowel a a tail ai ai ending au au yie ya yai ye ia ia iau iau ua ua uai uai ei ei yei...

  9. A new method for the determination of cystine and cysteine in foodstuffs containing carbohydrates 

    E-Print Network [OSTI]

    Doctor, Bhupendra Pannalal

    1955-01-01

    in the absence of, "lu- cose was studied. Thin wac done oy means of recover ry experiments by ref luxinp; the standard cystine and glucose in Different, concen- trations of ac&d Xn tne camo ray, the effect of the resencc of glucose and tryptophan on tho...&abztances, especially carbohyurotes !u!o li-ids, "he 1;. tt r are relate!rely easy to re. . ove b?; extractio. ! wztn ap ropxiata or anic solvents, The forner , rene!. t . ". ". oro . . iffic lt -, roblen, Xn the case of cysts!!e anu cysteine the destructive effects...

  10. Contact Detection and Constraints Enforcement for the Simulation of Pellet/Clad Thermo-Mechanical Contact in Nuclear Fuel Rods 

    E-Print Network [OSTI]

    Lebrun-Grandié , Damien Thomas

    2014-03-05

    &M University in partial fulfillment of the requirements for the degree of DOCTOR OF PHILOSOPHY Chair of Committee, Jean C. Ragusa Co-Chair of Committee, Wolfgang Bangerth Committee Members, Marvin L. Adams Ryan G. McClarren Head of Department, Yassin A. Hassan... interaction between fuel pellets and cladding in a LWR fuel rod [1]. A number of UO2 pel- lets stacked atop each other into the cladding tube: As fabricated (I), prior to PCMI (II), after PCMI (III). . . . . . . . . . . . . . . 4 VI.1 Recursive subdivision...

  11. An apparatus for the study of superconductivity and electron transport properties of amorphous metals 

    E-Print Network [OSTI]

    Stalnaker, Hwa Sun

    1987-01-01

    p in @flem and the mass density d in ~ by the relation &1I r, i . Uo N&t(0): 4 130 x (dIIcs/dT): 8. 965 pd T T pd ivhere 'll is the gram molecular weight and dHos/dT is in. Tesla, 'K. This method of determining the density of states is preferred... homogeneous (within 0. 37o over a O. ocm diameter spherical volume). The magnet is capable of producing up to, '3. 2 Tesla ivith the operating current of 68. 4 A. The current is supplied by a Sore nson DC poiver supply to two brass sheets. These curry nts...

  12. 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.

  13. Division of Correspondence Courses. 

    E-Print Network [OSTI]

    Ribinow, S. G.

    1915-01-01

    ' "'Correspondence Courses have been transferred }to the Department of Agricultural Education, The following courses only are being offered at the present time; STUDY COURSE 15 AGP.ICULTUTill TO BE OFFERED. Ho .of Ho. of Course: . Lessons 1 Elementary Agri... Cotton Classing.and Grading, J.. B. Bagley-- 8 TOTE: Study course Uo. 140, Farm Sultry, is a general course given by Mr. T. J. C.onway, Instructor • A in Poultry Husbandry. ' There are 20 lessons and the ^ testbook used is Lewis' "Productive p o u l...

  14. ON EFFICIENT BALANCED CODES Victor R. Lesser

    E-Print Network [OSTI]

    Massachusetts at Amherst, University of

    /or specfic permission. of all n-bit binary words and the set of all (n + p) -bit codewords such that,if w) be the total number of ones in the binary word w , 0 5 U(W) 5 n,n is the length of w .Let vi(w) be the number of ones in the first i bits of w , UO(W) = O,Vn(W) = V(W);Ui(W) 5 Vi+l(W) for 0 5 i 5 n-1. Let ~("1

  15. Transmission Electron Microscopy Investigation of Krypton Bubbles in Polycrystalline CeO2

    SciTech Connect (OSTI)

    Lingfeng He; Clarissa Yablinsky; Mahima Gupta; Jian Gan; Marquis A. Kirk; Todd R. Allen

    2013-05-01

    To gain an understanding of gas bubble transport in oxide nuclear fuel, this paper uses polycrystalline CeO2, composed of both nanograins and micrograins, as a surrogate material for UO2. The CeO2 was implanted with 150-keV Kr ions up to a dose of 1 x 1016 ions/cm2 at 600 degrees C. Transmission electron microscopy characterizations of small Kr bubbles in nanograin and micrograin regions were compared. The grain boundary acted as an efficient defect sink, as evidenced by smaller bubbles and a lower bubble density in the nanograin region as compared to the micrograin region.

  16. The relative efficiencies of various catalysts in the polymerization of unsaturated hydrocarbons 

    E-Print Network [OSTI]

    Andrews, Robert Vincent

    1940-01-01

    , To ubilizs Chess i'or gasoline ym8ueCion ChsV musC bs subgsoCsg Co a oraeldz@ rsaotion. This in Cum lsags Co Che formaeion of some loe molsoulsr might oomyounQs along vdCh Cbs 6esireg yro@uoC~ pelpmerisRCion of Chsse fraoCions is assessors' fox' gasoline.... yarafftuio compounds. 'Zhs hsmperatures of distillation mare nosed, She refraohive index tsLS Csee~ bp an Ahba refraebossber, and the bro~4e number vms determined with a potassium bromide ? potassium bromaCe solution bp the ~bhod gf the Universal. Qii...

  17. An alpha scintillation spectrometer 

    E-Print Network [OSTI]

    Yates, Ralph Aaron

    1952-01-01

    scintiilation counters& and for this reason they will be of in- terest here, The amount of light produced by an ionirirg particle of a given energy varies considerably under apparently co& st ant conditions. &'<housi& the particle loses about the same.... Uranium 1'hick Uranium sources were tested to determine if any differences would exist between the oulse height distribution oi' thick Uranium and Thorium sources. Sources were prepared by placing small pieces of Uranium nitrate, UO2 (NO3)2 6H20, on a...

  18. Geology of the Schep-Panther Creek Area, Mason County, Texas 

    E-Print Network [OSTI]

    Bryant, George Frank

    1959-01-01

    ?stain send?tenn. . . . . . Se VII I . Teo oeaarroasee of the Tel go send?toss. IX. Biaberas in tbo Qergsa Crook line?toss. . . , . . . . , . SS Coataot botueoa tho Morgan Creak liaestoao snd tho Point Peek shale... thick biohera scms, yre- riousiJ iaoladed' hJ eoae ssccloBists Ba. the ysiat Posh shale sa4 bJ others ia the Saa SISS Xiaootuaec uos sopped ia the tboeis S?ea OS a separate uait sithia tbe Poiat Pash ssaber. Tbo ouyosed Riley faraaticuc uas ~ st...

  19. Evaluation of gas-phase technetium decontamination and safety related experiments during FY 1994. A report of work in progress

    SciTech Connect (OSTI)

    Simmons, D.W.; Munday, E.B.

    1995-05-01

    Laboratory activities for FY94 included: evaluation of decontamination of Tc by gas-phase techniques, evaluation of diluted ClF{sub 3} for removing U deposits, evaluation of potential hazard of wet air inlekage into a vessel containing ClF{sub 3}, planning and preparation for experiments to assess hazard of rapid reaction of ClF{sub 3} and hydrated UO{sub 2}F{sub 2} or powdered Al, and preliminary evaluation of compatibility of Tenic valve seat material.

  20. Uranyl Sequestration: Synthesis and Structural Characterization of Uranyl Complexes with a Tetradentate Methylterephthalamide Ligand

    SciTech Connect (OSTI)

    Ni, Chengbao; Shuh, David; Raymond, Kenneth

    2011-03-07

    Uranyl complexes of a bis(methylterephthalamide) ligand (LH{sub 4}) have been synthesized and characterized by X-ray crystallography. The structure is an unexpected [Me{sub 4}N]{sub 8}[L(UO{sub 2})]{sub 4} tetramer, formed via coordination of the two MeTAM units of L to two uranyl moieties. Addition of KOH to the tetramer gave the corresponding monomeric uranyl methoxide species [Me{sub 4}N]K{sub 2}[LUO{sub 2}(OMe)].

  1. The Care and Cleaning of Furniture and Furnishings. 

    E-Print Network [OSTI]

    O'Donnell, Dorothy C.

    1981-01-01

    EXTENSION SERVICE o Tho Tox~ A&M Uo;,.,;

  2. Newberry EGS Seismic Velocity Model

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

    Templeton, Dennise

    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.

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

    SciTech Connect (OSTI)

    Malony, Allen D.; Wolf, Felix G.

    2014-01-31

    The growing number of cores provided by 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

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

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTech Connect Nanomechanical switchFlue Gasinelastic scattering atNANOSTRUCTURESSNe IaUO2

  9. Vibrational spectroscopy for online monitoring of extraction solvent degradation products

    SciTech Connect (OSTI)

    Peterson, J.; Robinson, T.; Bryan, S.A.; Levitskaia, T.G. [Pacific Northwest National Laboratory, 902 Battelle Blvd, PO Box 999, MSIN P7-25 Richland, WA 99352 (United States)

    2013-07-01

    In our research, we are exploring the potential of online monitoring of the organic solvents for the flowsheets relevant to the used nuclear fuel reprocessing and tributyl phosphate (TBP)- based extraction processes in particular. Utilization of vibrational spectroscopic techniques permits the discrimination of the degradation products from the primary constituents of the loaded extraction solvent. Multivariate analysis of the spectral data facilitates development of the regression models for their quantification in real time and potentially enables online implementation of a monitoring system. Raman and FTIR spectral databases were created and used to develop the regression partial least squares (PLS) chemometric models for the quantitative prediction of HDBP (dibutyl phosphoric acid) degradation product, TBP, and UO{sub 2}{sup 2+} extraction organic product phase. It was demonstrated that both these spectroscopic techniques are suitable for the quantification of the Purex solvent components in the presence of UO{sub 2}(NO{sub 3}){sub 2}. Developed PLS models successfully predicted HDBP and TBP organic concentrations in simulated Purex solutions.

  10. A study of certain factors of the ecology of the Texas harvester ant, Pogonomyrmex barbatus var. molefaciens (Buckl.) 

    E-Print Network [OSTI]

    Mangrum, James Freed

    1954-01-01

    8?,2??LSPHL,yG ywD F?OwCY TU TE.D)OD. OC SY? G? -? -CM?TEU C3 OwD RTC(C?I SDMFYOaDEO C3 OwD A?YTK?(O?YF( FE. HDKwFETKF( 8C((D?D C3 yD?FU? ?wC wFU U?MDYBTUD. OwTU UO?.I? FE. OC SY? ?? A? ?TOO(D FE. SY? G? -? RT(UTE? C3 OwD LEOCaC(C?I SDMFYOaDEO FE.... SY? 8? 8? SCF? FE. SY? P? L? ?COODY C3 OwD RTC(C?I SDMFYO? aDEO? ?wC wFBD UwC?E KCEOTE?D. TEODYDUO FE. ?TBDE BF(?F)(D U???DUOTCEU .?YTE? OwD MYC?YDUU C3 OwD TEBDUOT?FOTCE? 4,yV2S98y42, ywD UO?.I C3 FETaF(U (TBTE? TE K(CUD FUUCKTFOTCE ?TOw UCKTF...

  11. 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.

  12. 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.

  13. Estimated critical conditions for UF{sub 4}-oil systems in fully oil-reflected spherical geometry

    SciTech Connect (OSTI)

    Plaster, M.J.

    1997-05-01

    Paraffinic oil has been exposed to UF{sub 6} gas in seal exhaust pumps and cascade equipment at the Portsmouth Gaseous Diffusion Plant. The resulting mixture is more nuclearly reactive than mixtures of UO{sub 2}F{sub 2} and H{sub 2}O and is not bounded by the subcritical mass limits presented in several nuclear criticality safety guides. The purpose of this analysis is to determine several critical parameters; specifically, (1) k{sub {infinity}} and the critical mass for several enrichments and moderation levels and (2) the mass limits for these mixtures. The estimated critical masses for the UF{sub 4}-oil systems are smaller than for the UO{sub 2}F{sub 2}-H{sub 2}O systems. The suggested mass limits for the UF{sub 4}-oil systems are 0.240, 0.280, 0.350, 0.430, and 0.670, and 1.170 kg {sup 235}U for enrichments of 100, 50, 20, 10, 5, and 3 wt.% {sup 235}U respectively.

  14. Evaluation of tecniques for controlling UF/sub 6/ release clouds in the GAT environmental chamber

    SciTech Connect (OSTI)

    Lux, C.J.

    1982-01-01

    Studies designed to characterize the reaction between UF/sub 6/ and atmospheric moisture, evaluate environmental variables of UF/sub 6/ cloud formation and ultimate cloud fate, and UF/sub 6/ release cloud control procedure have been conducted in the 1200 cu. ft. GAT environmental chamber. In earlier chamber experiments, 30 separate UF/sub 6/ release tests indicated that variations of atmospheric conditions and sample sizes had no significant effect on UO/sub 2/F/sub 2/ particle size distribution, release cloud formation, or cloud settling rates. During the past year, numerous procedures have been evaluated for accelerating UF/sub 6/ cloud knockdown in a series of 37 environmental chamber releases. Knockdown procedures included: coarse water spray; air jet; steam spray (electrostatically charged and uncharged); carbon dioxide; Freon-12; fine water mist (uncharged); boric acid mist (charged and uncharged); and an ionized dry air stream. UF/sub 6/ hydrolysis cloud settling rates monitored by a laser/powermeter densitometer, indicated the relative effectiveness of various cloud knockdown techniques. Electrostatically charged boric acid/water mist, and electrostatically ionized dry air were both found to be very effective, knocking down the UO/sub 2/F/sub 2/ release cloud particles in two to five minutes. Work to adapt these knockdown techniques for use under field conditions is continuing, taking into account recovery of the released uranium as well as nuclear criticality constraints.

  15. Nuclear criticality safety modeling of an LEU deposit

    SciTech Connect (OSTI)

    Haire, M.J.; Elam, K.R.; Jordan, W.C.; Dahl, T.L.

    1996-11-01

    The construction of the Oak Ridge Gaseous Diffusion Plant (now known as the K-25 Site) began during World War H and eventually consisted of five major process buildings: K-25, K-27, K-29, K-31, and K-33. The plant took natural (0.711% {sup 231}U) uranium as feed and processed it into both low-enriched uranium (LEU) and high-enriched uranium (HEU) with concentrations up to {approximately}93% {sup 231}U. The K-25 and K-27 buildings were shut down in 1964, but the rest of the plant produced LEU until 1985. During operation, inleakage of humid air into process piping and equipment caused reactions with gaseous uranium hexafluoride (UF{sub 6}) that produced nonvolatile uranyl fluoride (UO{sub 2}F{sub 2}) deposits. As part of shutdown, most of the uranium was evacuated as volatile UF{sub 6}. The UO{sub 2}F{sub 2} deposits remained. The U.S. Department of Energy has mitiated a program to unprove nuclear criticality safety by removing the larger enriched uranium deposits.

  16. 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.

  17. 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.

  18. 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.

  19. 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%.

  20. 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)

  1. 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.

  2. 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)

  3. "!#!#$&%'!)(0#12!#354762358@935#ACB3DFEG93H2!351 IQPSRUTWVYXa`cbed

    E-Print Network [OSTI]

    Han, Jiawei

    ¨¤o{nw Uo¡n¢wvv£ n¦n@wU@2¨¤¥¨¦§ n¨o¨nª©{mnwm©2 UoU¨ m@« ¤¨oUninn ¤¦w©{©2owv¬¨¨mw©wmumm«tinm @v{v{£¦j{¨®¯¨¨m¨¦v{nm¨¨wª °m©mwo0¤Uw {mn±³²Y´w¨wµµwm¦¨¨nwv{¨¨onwn¨¨nµew o©w@¶¨woCw¨Qv¨¨ª©{eª©{mmn@ªv0@v{v mn¨¨£5wiww{nn{wov·Qo¨vuw¤n¤}{v n¤o ©2n¤¥o¤¨¨'o¤¸w¤wvvninn¨µ±w¹jmn ªuw oom¦o¤h¨¨omwUm@oom¨jvvºiU»i¼i½'m o©w@¶¨wo´¦j¨nUoj¤Dmw2{mnomn

  4. Standard test methods for arsenic in uranium hexafluoride

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2005-01-01

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

  5. Fictional Islam: A Literary Review and Comparative Essay on Islam in Science Fiction and Fantasy 

    E-Print Network [OSTI]

    Hankins, Rebecca

    2010-01-07

    e;ua8 aqt Jo srallrm )Iruelsl Jo eJnlnJ pue d;olsq aqr Surp-re8ar suoturdo ;o dtatre,t e slee^aJ,,{serueJ puE uoIDIJ eJuaIJS JIuelsI uo eJn}eJa}ll aql Jo l\\eIAaJ V JalsDW aql saJuEApE Jru ueaq eAEq dsetue; pue u uorlJrd eJuerJS ;o aruaS- pue uou...BeD slql sluasardar terp FIrl slql ul pal)a-uuoJ ar? lnQ 'eJIl uI raqlo q)?a ^^oDl rou plp .LIno sue)rrJv "J;i;"; are ra'{-'u'e1 ",t".:.]:P pue loln)eso"rd-'luepua;ap aql 'euaSrg u;al{lroN *'1'n'ni'"1'unutr'uopard tuo4 ePares ot patdualte 'sue...

  6. 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. [and others

    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.

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

    SciTech Connect (OSTI)

    Holliday, Kiel S.; Siekhaus, Wigbert; Nelson, Art J. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, California 94551 (United States)

    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}.

  8. 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.

  9. 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.

  10. 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.

  11. Thermodynamics of formation of coffinite, USiO?

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

    Guo, Xiaofeng; Szenknect, Stéphanie; Mesbah, Adel; Labs, Sabrina; Clavier, Nicolas; Poinssot, Christophe; Ushakov, Sergey V.; Curtius, Hildegard; Bosbach, Dirk; Ewing, Rodney C.; et al

    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 coupledmore »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.« less

  12. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    SciTech Connect (OSTI)

    DOE

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

  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. Source term evaluation during seismic events in the Paducah Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Kim, S.H.; Chen, N.C.J.; Schmidt, R.W.; Taleyarkhan, R.P.

    1996-12-30

    The 00 buildings are expected to collapse (per guidance from structure evaluation) during a seismic event in which acceleration level exceeds 0.15g. All roof beams may slip off supports, and collapse. Equipment may slip off from supports and fall onto the floor. The cell floor is also supposed to collapse due to structural instability and distortion due to excessive acceleration forces. Following structure collapse, expansion joints in the process piping and joints between the piping and equipment are expected to fail. Preliminary analysis showed that converters are likely to remain intact. The UF{sub 6} gas released from the break will rapidly interact with moisture in the air to produce UO{sub 2}F{sub 2} and HF with exothermic energy released of {approximately}0.32 MJ/kg of UF{sub 6} reacted. Depending on the degree of mixing between UF{sub 6} gas, its reaction products, air and freon (R-114), there may occur a strong buoyancy force to disperse UO{sub 2}F{sub 2} aerosol particles that are subjected to the gravitational force for settling. Such a chemical reaction will also occur inside the converters. A substantial amount of UF{sub 6} must be stagnated at the bottom of the converters. At the interface between this stagnated UF{sub 6} and air, UF{sub 6} gas will diffuse into the air, undergo the chemical reaction with moisture there, and eventually be released through the break. Furthermore, lubricant oil fire in the building, if it occurs, will enhance the UF{sub 6} release into the atmosphere. The purpose of this study is to evaluate source term (UO{sub 2}F{sub 2} and HF) during such a seismic event. This study takes an approach using multiple steps as follows: (1) Source term evaluation at the break due to mixing between UF{sub 6} and air along with thermal buoyancy induced by chemical reaction energy, (2) Evaluation of additional source term from the converters in which a substantial UF{sub 6} vapor remains, and (3) Source term evaluation with lubricant oil fire.

  15. 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)

  16. Linearized Coupled Cluster Correction on the Antisymmetric Product of 1 reference orbital Geminals

    E-Print Network [OSTI]

    Boguslawski, Katharina

    2015-01-01

    We present a Linearized Coupled Cluster (LCC) correction based on an Antisymmetric Product of 1 reference orbital Geminals (AP1roG) reference state. In our LCC ansatz, the cluster operator is restricted to double and to single and double excitations as in standard single-reference CC theory. The performance of the AP1roG-LCC models is tested for the dissociation of diatomic molecules (C$_2$ and F$_2$), spectroscopic constants of the uranyl cation (UO$_2^{2+}$), and the symmetric dissociation of the H$_{50}$ hydrogen chain. Our study indicates that an LCC correction based on an AP1roG reference function is more robust and reliable than corrections based on perturbation theory, yielding spectroscopic constants that are in very good agreement with theoretical reference data.

  17. 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 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

  18. Hadron production in e(+)e(-) annihilation at ?s =29 GeV

    E-Print Network [OSTI]

    Baringer, Philip S.; Derrick, M.; Fernandez, Eduardo H.; Gan, K. K.; Fries, R.; Hyman, L.; Kooijman, P.; Loos, J. S.; Musgrave, B.; Price, L. E.; Schlereth, J.

    1987-05-01

    - greement in all c ber obtained from TOF ases was well within the er- 2642 M. DERRICK et al. 35 0 i i i » f t ] f i » ) i t i i I i i I1. CO CO CO 200 150 100 0 —, , 0 (~) 0.5 ( p ( 0.7 M~Sr~~JL 0.5 (Gev/c) 1.5 C0 x I c5 0.8 0.4 l- 0.2 I 0.0 ~'0.6 0.8 ( & K... covers 90% of 4m sr with 15 layers of wires located between radii of 0.21 and 1.03 m. The outer chamber covers 60%%uo of 4m. with two layers of drift tubes at a mean radius of 1.89 m. The average drift-chamber resolution of 200 pm, combined with the 1...

  19. The hypergeometric functions and their confluent forms 

    E-Print Network [OSTI]

    Doyle, Jack Ellsworth

    1964-01-01

    of an Ordinary Point Z. I'he, latur&: of the Solution in t&uo . leighborhoou of a Singularity l:uci&s' Conditions 4. The Solution for l. urge Values of S. Totally pvc&hs& nn &:, quot io?s 6. The Analytic Continuation of &F'1(a, b;c;z) 7. The Confluent... ~ P zn ) na zn-1 + ) q zn ( anzn n n n n (1. 4) gquating co&&f ficients, we find 2 1P + 3 2pp I 1 lqp o ll (1. 53 (1. I&) and in general that -(n-1) na (n-1) a lp + (n-2) a -Zpl +' ' '+ alpn-2 + + a 2q + a &ql +. ~ . + alrl &+ a rr (1 ~ 7...

  20. The effect of GnRH on induction of follicular development and ovulation in anovulatory and ovulatory mares 

    E-Print Network [OSTI]

    Hennington, Debra Louise

    1981-01-01

    pad faq qua!eabeunooua pue aI!oi 'aouaLged unuse gfasEw u! anaL Iaq aw bufqew pue neM aqq. aw 5u!Moqs uo3. noE queqg 'yanqui og 'F, Cfeufg puff . paydwaqge aneq y ge!fg ffe u! qua!eabeunooua pue quoddns 'aoueueaquog ufaqg uog uopbufuuaW I Cassnff... treatment and 70% of the time during GnRH treatment. Control anovulatory mares exhibited estrous behavior 17'K and 15% of the time during the two 25 U1 Q +J Ql Q CL C) c o ) O fO O O Cl col (LUj/6") SN3BOJS3908d 26 courses of sham...

  1. Solutions of the equation of radiative transfer by matrix operator techniques 

    E-Print Network [OSTI]

    Catchings, Frances Eugenia King

    1973-01-01

    X X N~xkkaR 11F 0. 0 0, 2 0. % 0, 6 0. 8 1 0 0. 8 0, 6 0. 4 0. 2 0. 0 Figure 17 Downward Radiance for uo ~ 0. 1882, A ~ 1, and ) 0 and 180 (Rayleigh). See Fig. 2 for key. 10' q 0. 1882 A 1. 0 4 Do 4-180o tran 10 ' CC C3 C) CL ~ 10-& C3... Fig. 2 for key. ac=0. 9379 A=O. O e 0 e 180 CI: C3 CZ CI: C) CL Q. I ~ ( I' I " 'T I ej X X X 1 X e~e e X +J X X X X X X X + j + x, x x + e-I + x ~x + + + + 8 + e+ e + ' e+ + + + 8 k 8 8 8 8 8 8 $&$&y~ 1 0' I 10 10a y 10...

  2. Inclusive decays of B mesons to charmonium

    E-Print Network [OSTI]

    Baringer, Philip S.

    1995-09-01

    Carlo line shape Lepton ID Total e+e 4.2% 0.9% 2.0% 1.0% 4.0% 6.3% P P 4.2% 1.0% 4.7% 6.4% 'c uncertainty into the systematicntributions the B-+J/gA branching fraction me I II I & sI i i s 1.50 5 1.0 J/g Momentum (GeV i c) 2.0 f momentum for J/@'sas a... fitted with a second-order Chebychev polyno- mial. In this case, the number of events in the y i peak is 110+18, and the excess in the y 2 region is 37+14. The world average branching fractions for y, i-+ J/Qp and y,2~ J/Qp are (27.3+1.6)%%uo and (13...

  3. A Qualitative Linear Utility Theory for Spohn’s Theory of Epistemic Beliefs

    E-Print Network [OSTI]

    Giang, Phan H.; Shenoy, Prakash P.

    2000-08-01

    of expected utility of the lottery in Figure 1 if we assume o 2 ? [o 1 .0,o 3 .3] and o 1 is the best prize and o 3 is the worst prize. Notice that when L is a simple lottery, Equation 7 can be rewritten as U([o 1 .? 1 ,o 2 .? 2 ,...,o r .? r ]) = min 1?i... + ?{?}. For A ? ? such that ?(A) < ?, the conditional dis- belief function ?(.|A) is de?ned as S3 ?(B|A)=?(B ?A) ??(A). It is easy to verify that ?(.|A) is a disbelief function, i.e., it satis?es S1 and S2. The notion of independence for Spohn?s epistemic be...

  4. 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.

  5. 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.

  6. The effect of vitamin B?? on the embryonic development of the chick 

    E-Print Network [OSTI]

    Ferguson, Thomas Morgan

    1954-01-01

    No. ?8, was used for the photomicrographs of liver sections stained with Sudan IV. 15 COMPOSITION OF PURIFIED SYNTHETIC-TYPE DIET DEFICIENT IN VITAMIN B12 TABLE I Ingredient d/o Sucrose 68.0 Soybean Protein 2i;.0 Soybean Oil 3.0 Salts IV 5....0 Inositol 1000.0 Choline Chloride 2000.0 Penicillin 33.0 Me thionine 7.5 gm Glycine U.o grn Vitamin A 10000.0 U.S.P Vitamin D3 2250.0 I.C.U 16 COMPOSITION OF PRACTICAL TYPE DIET FED TO HENS PROVIDING SOURCE OF EM3RY0S USED AS NORMAL CONTROLS TABLE...

  7. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    SciTech Connect (OSTI)

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

    2013-09-01

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

  8. 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.

  9. The pseudospectral method for third-order differential equations

    E-Print Network [OSTI]

    Huang, Weizhang; Sloan, David M.

    1992-12-01

    ]. Then the truncation error of (2.7) is given by (2.17) Eu[f] (__ l )r.+ f(ZN+n-4)( rl(2N+n-4)! O0o+rn+l,3+ln+l(X)( 7"(X))2 dx where -1 < rl < 1. Proof Let Pu+,_5(x) be the polynomial interpolating the data f(Xk),f’(Xk), 2<--k<--N-I, (2.18) f((-1), 0<= v<=ln, f)(+l), O... it is not difficult to prove the stability of the collocation method with a 0,/3 0 for (3.1)-(3.3). THEOREM 3.2 (Stability). Let uu be the pseudospectral approximation (3.4) with x O, 0 to (3.1)-(3.3). Then (3.32) N-1 Z [uU(xk, t)]Zwk<=[2+O(N-2)] max [Uo(X)l2. k=2...

  10. Engineering geology of a mudslide at Bracebridge Inlet, Bathurst Island, Northwest Territories, Canada 

    E-Print Network [OSTI]

    Mayer, Terry Ann

    1980-01-01

    AdoIoaB:qoaCqnS uoCeN OGeI uaq??0 30N3IDS 30 'd3lSW J. o aaubap aqua uog quamautnbau aug go guamIIIJ Ing Ieigued ui EgIswancuO gIIy sexal go a5a IIog aq. enpeuB aqua oy paggteqnS 'd3MIN NNV ld'd3J. slsaqJ VQ'dN'd0 ' S3IKIQJI'd'd3l iS3MHJ. 'd... possible through the kindness of the Terrain Sciences Division of the Geological Survey of Canada. Mr. Fred Alt and the Polar Continental Shelf Project, Depart- ment of Energy, Mines, and Resources of Canada, provided logistical support. Dr. Larry Dyke...

  11. 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.

  12. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    SciTech Connect (OSTI)

    Marcille, T. F.; Poston, D. I.; Kapernick, R. J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Dixon, D. D. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Fischer, G. A. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Doherty, S. P. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Engineering, Trinity College, Hartford, CT 06106 (United States)

    2006-01-20

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  13. Mass of Cu-57 

    E-Print Network [OSTI]

    Gagliardi, Carl A.; Semon, D. R.; Tribble, Robert E.; Vanausdeln, L. A.

    1986-01-01

    at temperatures of T9 0.4?7 and 0.70 with the two hypotheses. In Fig. 3(b}, we show the calculated ratio of the Cu photodisintegration rate to its beta-decay rate as a function of temperature, again ac- cording to our two hypotheses. %hen this ratio is much... PROCEDURES AND RESULTS A 76.5 MeV Li + beam from the Texas A&M cyclo- tron irradiated a target consisting of 1-52 mg/cmz ss&i (99.98%%uo purity), backed by 1.10 mg/cm of Al. Outgoing ejectiles at 7.0 deg were detected with an Enge split...

  14. Influence of loading rate on axially loaded piles in clay 

    E-Print Network [OSTI]

    Garland Ponce, Enrique Eduardo

    1984-01-01

    28. 0 0. 12 26. 4 I. nn 13. 3 NA 30. 2 HA 33. 7 Im Cul Csi I ID% UDS 4. 32 In. oo 0. 00 0. 00 I. sn S. OSMxtei 11. 17 1. 60tisxlni 0. 40 6. 6667 10 O. ze 6 666?10-1 2 IMnats. z 6. 982lxln I I. ZSIOxtn I 3. 00eoxlo-l 1. 57 11. 94... 93. 1 NA HA HA 83. 3 NA 0. 64 lb. 8 NA 62. 4 NA 0. 21 54. 9 NA sn. n I. nn sz. n t. sn uns ODS UOS CUT Cul Csl CUT CIIT Cur Cul 0. Oe 0. 00 O. nn 4. 32 0. 51 I. 82 2. 05 4. 10 7. 56 0. 7 5 0. 63 6. 6667xlo I 4. 450oxtn-t n...

  15. 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.

  16. 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.

  17. DOES CRITICAL MASS DECREASE AS TEMPERATURE INCREASES: A REVIEW OF FIVE BENCHMARK EXPERIMENTS THAT SPAN A RANGE OF ELEVATED TEMPERATURES AND CRITICAL CONFIGURATIONS

    SciTech Connect (OSTI)

    Yates, K.

    2009-06-10

    Five sets of benchmark experiments are reviewed herein that cover a diverse set of fissile system configurations. The review specifically focused on the change in critical mass of these systems at elevated temperatures and the temperature reactivity coefficient ({alpha}{sub T}) on the system. Because plutonium-based critical benchmark experiments at varying temperatures were not found at the time this review was prepared, only uranium-based systems are included, as follows: (1) HEU-SOL-THERM-010 - UO{sub 2}F{sub 2} solutions with high U{sup 235} enrichment; (2) HEU-COMP-THERM-016 - uranium-graphite blocks with low U concentration; (3) LEU-COMP-THERM-032 - water moderated lattices of UO{sub 2} with stainless steel cladding, and intermediate U{sup 235} enrichment; (4) IEU-COMP-THERM-002 - water moderated lattices of annular UO{sub 2} with/without absorbers, and intermediate U{sup 235} enrichment; and (5) LEU-COMP-THERM-026 - water moderated lattices of UO{sub 2} at different pitches, and low U{sup 235} enrichment. In three of the five benchmarks (1, 3 and 5), modeling of the critical system at room temperature is conservative compared to modeling the system at elevated temperatures, i.e., a greater fissile mass is required at elevated temperature. In one benchmark (4), there was no difference in the fissile mass between the room temperature system and the system at the examined elevated temperature. In benchmark (2), the system clearly had a negative temperature reactivity coefficient. Some of the high temperature benchmark experiments were treated with appropriate (and comprehensive) adjustments to the cross section sets and thermal expansion coefficients, while other experiments were treated with partial adjustments. Regardless of the temperature treatment, modeling the systems at room temperature was found to be conservative for the examined systems, i.e., a smaller critical mass was obtained. While the five benchmarks presented herein demonstrate that, for the conditions examined, modeling of the systems at room temperature is conservative as compared to modeling the systems at elevated temperatures, it is possible to design a system in which the critical mass at room temperature is non-conservative compared to a system at elevated temperatures. As the temperature of the systems evaluated in this review was increased, the system's overall {alpha}{sub T} was negative at elevated temperatures. Furthermore, the review demonstrates that to accurate asses the effect of increased temperature on a system's k{sub eff}, changes in fissile, moderator, cladding, and, in some cases, structural material cross sections must be combined with other factors that influence reactivity, such as volumetric thermal expansion of fissile, moderating, reflector, and other interacting media. Altering the microscopic cross sections of fissile and moderating regions for temperature changes, without adjusting the corresponding densities at elevated temperatures can lead to an incorrect assessment of the impact of elevated temperature on a fissile system.

  18. Sampling and characterization of aerosols formed in the atmospheric hydrolysis of UF/sub 6/

    SciTech Connect (OSTI)

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

    1983-01-01

    When gaseous UF/sub 6/ is released into the atmosphere, it rapidly reacts with ambient moisture to form an aerosol of uranyl fluoride and HF. As part of our Safety Analysis program, we have performed several experimental releases of UF/sub 6/ (from natural uranium) in contained volumes in order to investigate techniques for sampling and characterizing the aerosol materials. The aggregrate particle morphology and size distribution have been found to be dependent upon several conditions, including the relative humidity at the time of the release and the elapse time after the release. Aerosol composition and settling rate have been investigated using isokinetic samplers for the separate collection of UO/sub 2/F/sub 2/ and HF, and via laser spectroscopic remote sensing (Mie scatter and infrared spectroscopy). 8 references.

  19. Safety analysis of B and W Standard PWR using thorium-based fuels

    SciTech Connect (OSTI)

    Uotinen, V.O.; Carroll, W.P.; Jones, H.M.; Toops, E.C.

    1980-06-01

    A study was performed to assess the safety and licenseability of the Babcock and Wilcox standard 205-fuel assembly PWR when it is fueled with three types of thoria-based fuels denatured (/sup 233/U//sup 238/U-Th)O/sub 2/, denatured (/sup 235//U/sup 238/U-Th)O/sub 2/, and (Th-Pu)O/sub 2/. Selected transients were analyzed using typical PWR safety analysis calculational methods. The results support the conclusion that it is feasible from a safety standpoint to utilize either of the denatured urania-thoria fuels in the standard B and W plant. In addition, it appears that the use of thoria-plutonia fuels would probably also be feasible. These tentative conclusions depend on a data that is more limited than that available for UO/sub 2/ fuels.

  20. Transient Testing of Nuclear Fuels and Materials in United States

    SciTech Connect (OSTI)

    Daniel M. Wachs

    2012-12-01

    The US Department of Energy (DOE) has been engaged in an effort to develop and qualify next generation LWR fuel with enhanced performance and safety and reduced waste generation since 2010. This program, which has emphasized collaboration between the DOE, U.S. national laboratories and nuclear industry, was refocused from enhanced performance to enhanced accident tolerance following the events at Fukushima in 2011. Accident tolerant fuels have been specifically described as fuels that, in comparison with standard UO2-Zircaloy, 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, operational transients, as well as design-basis and beyond design-basis events. The program maintains an ambitious goal to insert a lead test assembly (LTA) of the new design into a commercial power reactor by 2022 .

  1. 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.

  2. 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).

  3. 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.

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

    E-Print Network [OSTI]

    Iosilevskiy, Igor

    2010-01-01

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

  5. 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.

  6. 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.

  7. 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.

  8. 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.

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2000-01-01

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

  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. 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.

  12. 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.

  13. 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.

  14. Liberalism, Religion, and the Sources of Value

    E-Print Network [OSTI]

    Blackburn, Simon

    2004-01-01

    . Chlaholm ~ fu.oom M'IO ""' S.ll Stual1 HamJIIahl,. r •uo:>m o1 u.no William J<. P'rankana SOme 8H •t' &eo~~ M ile• Wllfrfd aallarw F'orm •"" Conltnl '"' (lt\\•Cti il'leOIJ J. N. P'lndlalf Alan QewlrUI Uotaf 1\\atONtt, Alba11 Hofetadter lllaul... that it works, and that you rely on it all the time. Pop­ per may have thought that the scientific assertion that arsenic poisons 2 is hut a hold cm~jcctun·, lik<· tlw cot~j<•ctun· that an asteroid will de­ stroy lilc on earth within fi~·c hundn·d years...

  15. Analysis of Radioactive Releases During Proposed Demolition Activities for the 224-U and 224-UA Buildings

    SciTech Connect (OSTI)

    Napier, Bruce A.; Rishel, Jeremy P.; Droppo, James G.

    2009-03-31

    Atmospheric dispersion modeling has been conducted in support of the demolition of the 224-U and 224-UA buildings using estimated release rates to provide information on the location and levels of radioactive contamination that may be expected. The facilities surrounding the UO3 plant have the potential to affect dispersion patterns through various meteorological phenomena, including building wake effects. Hourly meteorological data collected over a 5-year period were used to examine the effects of wind speed, direction, and stability on projected concentrations of contaminants in air and deposited on nearby surfaces. The modeling results indicate that the radiological exposures from the planned demolition efforts will be below the designated limits for air and soil exposures.

  16. The PACSAT Communications Experiment (PCE). Final report, August 13, 1990--February 12, 1992

    SciTech Connect (OSTI)

    Not Available

    1993-02-12

    While VITA (Volunteers in Technical Assistance) is the recognized world leader in low earth orbiting (LEO) satellite technology (below 1 GHz), its involvement in communications technologies is to facilitate renewable energy technology transfer to developing countries. A communications payload was incorporated into the UoSat 2 satellite (Surrey Univ., UK), launched in 1984; a prototype satellite (PCE) was also launched Jan 1990. US DOE awarded a second grant to VITA to design and test the prototype ground stations (command and field), install field ground stations in several developing country sites, pursue the operational licensing process, and transfer the evaluation results to the design of an operating system. This report covers the principal tasks of this grant.

  17. The PACSAT Communications Experiment (PCE)

    SciTech Connect (OSTI)

    Not Available

    1993-02-12

    While VITA (Volunteers in Technical Assistance) is the recognized world leader in low earth orbiting (LEO) satellite technology (below 1 GHz), its involvement in communications technologies is to facilitate renewable energy technology transfer to developing countries. A communications payload was incorporated into the UoSat 2 satellite (Surrey Univ., UK), launched in 1984; a prototype satellite (PCE) was also launched Jan 1990. US DOE awarded a second grant to VITA to design and test the prototype ground stations (command and field), install field ground stations in several developing country sites, pursue the operational licensing process, and transfer the evaluation results to the design of an operating system. This report covers the principal tasks of this grant.

  18. Uranium dioxide electrolysis

    DOE Patents [OSTI]

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

    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.

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

    SciTech Connect (OSTI)

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

    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}.

  20. 18 years experience on UF{sub 6} handling at Japanese nuclear fuel manufacturer

    SciTech Connect (OSTI)

    Fujinaga, H.; Yamazaki, N.; Takebe, N. [Japan Nucelar Fuel Conversion Co., Ltd., Ibaraki (Japan)

    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.

  1. 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.

  2. Lifetime of 981-Kev State in Li-8 

    E-Print Network [OSTI]

    Throop, M. J.; Youngblood, David H.; Morrison, G. C.

    1971-01-01

    Laboratory Report No. AFOUL-TB-65-150, 1966 (unpublished). 'S. W. Robinson and B. D. Bent, Phys. Rev. 168, 1266 (1968). ' L. F. Chase, Jr., R. 0, Johnson, F. J, Vaughn, and E. K. %arburton, Phys. Bev. 127, 859 (1962). Chromium CorporRtlon of Amer ica... as the 2.10-1.08-MeV transition in F~8 (E?=1021.3 + 1.2 keV, 33'%%uo branch7). The accompanying 2.10-0.94-MeV transition in F"was also observed with about equal intensity, as expected. A third set of runs with a 25-cm'-coaxial detector of im- proved...

  3. Groundwater impact assessment report for the 216-U-14 Ditch

    SciTech Connect (OSTI)

    Singleton, K.M.; Lindsey, K.A.

    1994-01-01

    Groundwater impact assessments are conducted at liquid effluent receiving sites on the Hanford Site to determine hydrologic and contaminant impacts caused by discharging wastewater to the soil column. The assessments conducted are pursuant to the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-17-00A and M-17-00B, as agreed by the US Department of Energy (DOE), Washington State Department of Ecology (Ecology), and the US Environmental Protection Agency (EPA) (Ecology et al. 1992). This report assesses impacts on the groundwater and vadose zone from wastewater discharged to the 216-U-14 Ditch. Contemporary effluent waste streams of interest are 242-S Evaporator Steam Condensate and UO{sub 3}/U Plant wastewater.

  4. 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

  5. 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.

  6. Composition, stability, and measurement of reduced uranium phases for groundwater bioremediation at Old Rifle, CO

    SciTech Connect (OSTI)

    Campbell, K. M. [USGS, Menlo Park, CA (United States); Davis, J. A. [USGS, Menlo Park, CA (United States) and Lawrence Berkeley National Lab., Berkeley, CA (United States); Bargar, J. [Stanford Synchrotron Radiation Lightsource, Menlo Park, CA (United States); Giammar, D. [Washington Univ., St. Louis, MO (United States); Bernier-Latmani, R. [Ecole Polytechnique Federale de Lausanne (Switzerland). Environmental Microbiology Lab.; Kukkadapu, R. [Pacific Northwest National Lab., Richland, WA (United States); Williams, K. H. [Lawrence Berkeley National Lab., Berkeley, CA (United States); Veramani, H. [Washington Univ., St. Louis, MO (United States); Ulrich, K. U. [Washington Univ., St. Louis, MO (United States) and BGD Boden- und Grundwasserlabor GmbH Dresden (Germany); Stubbs, J. [Stanford Synchrotron Radiation Lightsource, Menlo Park, CA (United States); Yabusaki, S. [Pacific Northwest National Lab., Richland, WA (United States); Figueroa, L. [Colorado School of Mines, Golden, CO (United States); Lesher, E. [Colorado School of Mines, Golden, CO (United States); Wilkins, M. J. [Pacific Northwest National Lab., Richland, WA (United States); Peacock, A. [Haley and Aldrich, Oak Ridge, TN (United States); Long, P. E. [Pacific Northwest National Lab., Richland, WA (United States)

    2011-10-15

    Reductive biostimulation is currently being explored as a possible remediation strategy for uranium (U) contaminated groundwater, and is currently being investigated at a field site in Rifle, CO, USA. The long-term stability of the resulting U(IV) phases is a key component of the overall performance and depends upon a variety of factors, including rate and mechanism of reduction, mineral associations in the subsurface, and propensity for oxidation. To address these factors, several approaches were used to evaluate the redox sensitivity of U: measurement of the rate of oxidative dissolution of biogenic uraninite (UO{sub 2(s)}) deployed in groundwater at Rifle, characterization of a zone of natural bioreduction exhibiting relevant reduced mineral phases, and laboratory studies of the oxidative capacity of Fe(III) and reductive capacity of Fe(II) with regard to U(IV) and U(VI), respectively.

  7. Solid-state actinide acid phosphites from phosphorous acid melts

    SciTech Connect (OSTI)

    Oh, George N. [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, 156 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Burns, Peter C., E-mail: pburns@nd.edu [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, 156 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Department of Chemistry and Biochemistry, University of Notre Dame, Notre Dame, IN 46556 (United States)

    2014-07-01

    The reaction of UO{sub 3} and H{sub 3}PO{sub 3} at 100 °C and subsequent reaction with dimethylformamide (DMF) produces crystals of the compound (NH{sub 2}(CH{sub 3}){sub 2})[UO{sub 2}(HPO{sub 2}OH)(HPO{sub 3})]. This compound crystallizes in space group P2{sub 1}/n and consists of layers of uranyl pentagonal bipyramids that share equatorial vertices with phosphite units, separated by dimethylammonium. In contrast, the reaction of phosphorous acid and actinide oxides at 210 °C produces a viscous syrup. Subsequent dilution in solvents and use of standard solution-state methods results in the crystallization of two polymorphs of the actinide acid phosphites An(HPO{sub 2}OH){sub 4} (An=U, Th) and of the mixed acid phosphite–phosphite U(HPO{sub 3})(HPO{sub 2}OH){sub 2}(H{sub 2}O)·2(H{sub 2}O). ?- and ?-An(HPO{sub 2}OH){sub 4} crystallize in space groups C2/c and P2{sub 1}/n, respectively, and comprise a three-dimensional network of An{sup 4+} cations in square antiprismatic coordination corner-sharing with protonated phosphite units, whereas U(HPO{sub 3})(HPO{sub 2}OH){sub 2}(H{sub 2}O){sub 2}·(H{sub 2}O) crystallizes in a layered structure in space group Pbca that is composed of An{sup 4+} cations in square antiprismatic coordination corner-sharing with protonated phosphites and water ligands. We discuss our findings in using solid inorganic reagents to produce a solution-workable precursor from which solid-state compounds can be crystallized. - Graphical abstract: Reaction of UO{sub 3} and H{sub 3}PO{sub 3} at 100 °C and subsequent reaction with DMF produces crystals of (NH{sub 2}(CH{sub 3}){sub 2})[UO{sub 2}(HPO{sub 2}OH)(HPO{sub 3})] with a layered structure. Reaction of phosphorous acid and actinide oxides at 210 °C produces a viscous syrup and further solution-state reactions result in the crystallization of the actinide acid phosphites An(HPO{sub 2}OH){sub 4} (An=U, Th), with a three-dimensional network structure, and the mixed acid phosphite–phosphite U(HPO{sub 3})(HPO{sub 2}OH){sub 2}(H{sub 2}O){sub 2}·(H{sub 2}O) with a layered structure. - Highlights: • U(VI), U(IV) and Th(IV) phosphites were synthesized by solution-state methods. • A new uranyl phosphite structure is based upon uranyl phosphite anionic sheets. • New U and Th phosphites have framework structures.

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

    E-Print Network [OSTI]

    Igor Iosilevskiy; Victor Gryaznov

    2010-05-23

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

  9. 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 [NSTec

    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.

  10. Nuclear fuels for very high temperature applications

    SciTech Connect (OSTI)

    Lundberg, L.B.; Hobbins, R.R.

    1992-08-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  11. Cryochemical and CVD processing of shperical carbide fuels for propulsion reactors

    SciTech Connect (OSTI)

    Blair, H.T.; Carroll, D.W.; Matthews, R.B. (Los Alamos National Laboratory, MS E505, Los Alamos, New Mexico (USA))

    1991-01-10

    Many of the nuclear propulsion reactor concepts proposed for a manned mission to Mars use a coated spherical particle fuel form similar to that used in the Rover and NERVA propulsion reactors. The formation of uranium dicarbide microspheres using a cryochemical process and the coating of the UC{sub 2} spheres with zirconium carbide using chemical vapor deposition are being developed at Los Alamos National Laboratory. The cryochemical process is described with a discussion of the variables affecting the sphere formation and carbothermic reduction to produce UC{sub 2} spheres from UO{sub 2}. Emphasis is placed on minimizing the wastes produced by the process. The ability to coat particles with ZrC was recaptured, and improvements in the process and equipment were developed. Volatile organometallic precursors were investigated as alternatives to the original ZrCl{sub 4} precursor.

  12. Nuclear fuels for very high temperature applications

    SciTech Connect (OSTI)

    Lundberg, L.B.; Hobbins, R.R.

    1992-01-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  13. 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.

  14. The Structural Role of Zr within Alkali Borosilicate Glasses for Nuclear Waste Immobilisation

    SciTech Connect (OSTI)

    A Connelly; N Hyatt; K Travis; R Hand; E Maddrell; R Short

    2011-12-31

    Zirconium is a key constituent element of High Level nuclear Waste (HLW) glasses, occurring both as a fission product and a fuel cladding component. As part of a wider research program aimed at optimizing the solubility of zirconium in HLW glasses, we have investigated the structural chemistry of zirconium in such materials using X-ray Absorption Spectroscopy (XAS). Zirconium K-edge XAS data were acquired from several inactive simulant and simplified waste glass compositions, including a specimen of blended Magnox/UO{sub 2} fuel waste glass. These data demonstrate that zirconium is immobilized as (octahedral) six-fold coordinate ZrO{sub 6} species in these glasses, with a Zr-O contact distance of 2.09 {angstrom}. The next nearest neighbors of the Zr species are Si at 3.42 {angstrom} and possibly Na at 3.44 {angstrom}, no next nearest neighbor Zr could be resolved.

  15. Thermomechanical simulation of the DIAMINO irradiation experiment using the LICOS fuel design code

    SciTech Connect (OSTI)

    Bejaoui, S.; Helfer, T.; Brunon, E.; Lambert, T. [Commissariat a l'Energie Atomique - CEA, Centre de Cadarache, 13108 St-Paul-lez-Durance (France); Bendotti, S.; Neyroud, C. [Commissariat a l'Energie Atomique - CEA, Centre de Saclay, 91191 Gif sur Yvette (France)

    2013-07-01

    Two separate-effect experiments in the HFR and OSIRIS Material Test Reactors (MTRs) are currently under Post- Irradiation Examinations (MARIOS) and under preparation (DIAMINO) respectively. The main goal of these experiments is to investigate gaseous release and swelling of Am-bearing UO2-x fuels as a function of temperature, fuel microstructure and gas production rate. First, a brief description of the MARIOS and DIAMINO irradiations is provided. Then, the innovative experimental in-pile device specifically developed for the DIAMINO experiment is described. Eventually, the thermo-mechanical computations performed using the LICOS code are presented. These simulations support the DIAMINO experimental design and highlight some of the capabilities of the code. (authors)

  16. Fission Enhanced diffusion of uranium in zirconia

    E-Print Network [OSTI]

    Bérerd, N; Moncoffre, N; Sainsot, P; Faust, H; Catalette, H

    2005-01-01

    This paper deals with the comparison between thermal and Fission Enhanced Diffusion (FED) of uranium into zirconia, representative of the inner face of cladding tubes. The experiments under irradiation are performed at the Institut Laue Langevin (ILL) in Grenoble using the Lohengrin spectrometer. A thin $^{235}UO\\_2$ layer in direct contact with an oxidized zirconium foil is irradiated in the ILL high flux reactor. The fission product flux is about 10$^{11}$ ions cm$^{-2}$ s$^{-1}$ and the target temperature is measured by an IR pyrometer. A model is proposed to deduce an apparent uranium diffusion coefficient in zirconia from the energy distribution broadening of two selected fission products. It is found to be equal to 10$^{-15}$ cm$^2$ s$^{-1}$ at 480$\\circ$C and compared to uranium thermal diffusion data in ZrO$\\_2$ in the same pressure and temperature conditions. The FED results are analysed in comparison with literature data.

  17. 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.

  18. The use of radar data in the investigation of precipitation distributions and anomalous propagation 

    E-Print Network [OSTI]

    Truppi, Lawrence Ernest

    1957-01-01

    ~', ". 'f"' & 'P'3 ", , 3;CILflCIP. '. . I! l !lCI, XIII' . Qfif 30 fifCILVBIZ~Z. 'I ClI 'I V;YG VVff'3'P, dC ', 3". ff '&3Z SVXBI JO . =8: J1 2 'A V X 2%~ozbeq. eg goagqnS so&sg Enrnusl dONSIOS dO KKJSVN Jo eaxSap aqua uoJ eguaeegynbaz aqua Jo qua...Lxwnuep 30iV. '20S 3C P, ':IJGKI go eez3ap aqua cog egueurezynbez aqua go guantIIygIng Iaygmd uy sexed go a3aIIog Ieoyuaqoaff pua IamgInoyzqy aqua go Iooqog agenpezg a~ og peggILqng Icf BflHI 3 . ". fOif, fPJifWI ?C'I I'fO'fdC'Pif, "f1C...

  19. PUREX transition project case study

    SciTech Connect (OSTI)

    Jasen, W.G.

    1996-04-15

    In December 1992, the US Department of Energy (DOE) directed that the Plutonium-Uranium Extraction (PUREX) Plant be shut down and deactivated because it was no longer needed to support the nation`s production of weapons-grade plutonium. The PUREX/UO{sub 2} Deactivation Project will establish a safe and environmentally secure configuration for the facility and preserve that configuration for 10 years. The 10-year span is used to predict future maintenance requirements and represents the estimated time needed to define, authorize, and initiate the follow-on decontamination and decommissioning activities. Accomplishing the deactivation project involves many activities. Removing major hazards, such as excess chemicals, spent fuel, and residual plutonium are major goals of the project. The scope of the PUREX Transition Project is described within.

  20. The structure of the carbon black flame 

    E-Print Network [OSTI]

    Anderson, W. Kermi

    1945-01-01

    nvd1yS w.w, w. U w . ' jomvR ?Wdp jomvR ?Wdp jomvR ?Wdp w.'F w.U3F w.U*w w.U*O w.U*C w.U8O w.UCw w.Fw w..U3F w.U38 w.U8U w.U8* w.0wO w.'w ' w.*F w.U8w w.U8F w.'w ' w.'wO w.0w* w.'Uw U.ww w.U8C w.UCF w.'w0 w.'w0 w.'w8 w.'U3 U.'F w.UCw w.UCw w.0w0... w.'w0 w.'Uw w.'UO U.Fw w.UCw w.UCF w.'w0 w.'wF w.'U0 w.'w8 U.*F w.UC3 w.UC8 w.'wO w.'wF w.'UO w.0U8 ' .ww w.UC3 w.UC3 w.'w ' w .'w8 w.'UF w.0U8 '.'F w .UC8 w.UC3 w.'w ' w.'wO w.'U8 w.0U8 ' .Fw w.'wO w.'w3 w.'w8 w.0w3 w.'UC w.' 'U EO *w w.'w8 w...

  1. Mini-MITEE: Ultra Small, Ultra Light NTP Engines for Robotic Science and Manned Exploration Missions

    SciTech Connect (OSTI)

    Powell, James; Maise, George; Paniagua, John [Plus Ultra Technologies, Incorporated, Shoreham, NY 11786 (United States)

    2006-01-20

    A compact, ultra lightweight Nuclear Thermal Propulsion (NTP) engine design is described with the capability to carry out a wide range of unique and important robotic science missions that are not possible using chemical or Nuclear Electric Propulsion (NEP). The MITEE (MInature ReacTor EnginE) reactor uses hydrogeneous moderator, such as solid lithium-7 hydride, and high temperature cermet tungsten/UO2 nuclear fuel. The reactor is configured as a modular pressure tube assembly, with each pressure tube containing an outer annual shell of moderator with an inner annular region of W/UO2 cermet fuel sheets. H2 propellant flows radially inwards through the moderator and fuel regions, exiting at {approx}3000 K into a central channel that leads to a nozzle at the end of the pressure tube. Power density in the fuel region is 10 to 20 megawatts per liter, depending on design, producing a thrust output on the order of 15,000 Newtons and an Isp of {approx}1000 seconds. 3D Monte Carlo neutronic analyses are described for MITEE reactors utilizing various fissile fuel options (U-235, U-233, and Am242m) and moderators (7LiH and BeH2). Reactor mass ranges from a maximum of 100 kg for the 7LiH/U-235 option to a minimum of 28 kg for the BeH2/Am-242 m option. Pure thrust only and bi-modal (thrust plus electric power generation) MITEE designs are described. Potential unique robotic science missions enabled by the MITEE engine are described, including landing on Europa and exploring the ice sheet interior with return of samples to Earth, hopping to and exploring multiple sites on Mars, unlimited ramjet flight in the atmospheres of Jupiter, Saturn, Uranus, and Neptune and landing on, and sample return from Pluto.

  2. 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.

  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. [and others

    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. Unsteady-state material balance model for a continuous rotary dissolver

    SciTech Connect (OSTI)

    Lewis, B.E.

    1984-09-01

    The unsteady-state continuous rotary dissolver material balance code (USSCRD) is a useful tool with which to study the performance of the rotary dissolver under a wide variety of operating conditions. The code does stepwise continuous material balance calculations around each dissolver stage and the digester tanks. Output from the code consists of plots and tabular information on the stagewise concentration profiles of UO{sub 2}, PuO{sub 2}, fission products, Pu(NO{sub 3}){sub 4}, UO{sub 2}(NO{sub 3}){sub 2}, fission product nitrates, HNO{sub 3}, H{sub 2}O, stainless steel, total particulate, and total fuel in pins. Other information about material transfers, stagewise liquid volume, material inventory, and dissolution performance is also provided. This report describes the development of the code, its limitations, key operating parameters, usage procedures, and the results of the analysis of several sets of operating conditions. Of primary importance in this work was the estimation of the steady-state heavy metal inventory in a 0.5-t/d dissolver drum. Values ranging from {similar_to}12 to >150 kg of U + Pu were obtained for a variety of operating conditions. Realistically, inventories are expected to be near the lower end of this range. Study of the variation of operating parameters showed significant effects on dissolver product composition from intermittent solids feed. Other observations indicated that the cycle times for the digesters and shear feed should be closely coupled in order to avoid potential problems with off-specification product. 19 references, 14 tables.

  5. 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.

  6. 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)

  7. Kinetic studies of the [NpO? (CO?)?]?? ion at alkaline conditions using ¹³C NMR

    SciTech Connect (OSTI)

    Panasci, Adele F. [Univ. of California, Davis, CA (United States); Harley, Stephen J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zavarin, Mavrik [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Casey, William H. [Univ. of California, Davis, CA (United States)

    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.

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

    SciTech Connect (OSTI)

    Ensor, D.D.

    1997-01-01

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

  9. Effects of solution pH and complexing reagents on uranium and thorium desorption under saturated equilibrium conditions

    SciTech Connect (OSTI)

    Wang, Yug-Yea; Yu, C.

    1992-01-01

    Three contaminated bulk surface soils were used for investigating the effect of solution pH and complexing reagents on uranium and thorium desorption. At a low solution pH, the major chemical species of uranium and thorium, uranyl UO{sub 2}{sup +2}, thorium dihydroxide Th(OH){sub 2}{sup +2}, and thorium hydroxide Th(OH){sup +3}, tend to form complexes with acetates in the solution phase, which increases the fractions of uranium and thorium desorbed into this phase. At a high solution pH, important uranium and thorium species such as uranyl tricarbonate complex UO{sub 2}(CO){sub 3}{sub 3}{sup {minus}4} and thorium tetrahydroxide complex Th(OH){sub 4} tend to resist complexation with acetates. The presence of complexing reagents in solution can release radionuclides such as uranium and/or thorium from the soil to the solution by forming soluble complexes. Sodium bicarbonate (NaHCO{sub 3}) and diethylenetriaminepentaacetic acid (DTPA) are strong complex formers that released 38% to 62% of total uranium activity and 78% to 86% of total thorium activity, respectively, from the soil samples investigated. Solutions of 0.1 molar sodium nitrate (NaNO{sub 3}) and 0.1 molar sodium sulfate (Na{sub 2}SO{sub 4}) were not effective complex formers with uranium and thorium under the experimental conditions. Fractions of uranium and thorium desorbed by 0.15g/200ml humic acid ranged from 4.62% to 6.17% and 1.59% to 7.09%, respectively. This work demonstrates the importance of a knowledge of solution chemistry in investigating the desorption of radionuclides.

  10. Effects of solution pH and complexing reagents on uranium and thorium desorption under saturated equilibrium conditions

    SciTech Connect (OSTI)

    Wang, Yug-Yea; Yu, C.

    1992-08-01

    Three contaminated bulk surface soils were used for investigating the effect of solution pH and complexing reagents on uranium and thorium desorption. At a low solution pH, the major chemical species of uranium and thorium, uranyl UO{sub 2}{sup +2}, thorium dihydroxide Th(OH){sub 2}{sup +2}, and thorium hydroxide Th(OH){sup +3}, tend to form complexes with acetates in the solution phase, which increases the fractions of uranium and thorium desorbed into this phase. At a high solution pH, important uranium and thorium species such as uranyl tricarbonate complex UO{sub 2}(CO){sub 3}{sub 3}{sup {minus}4} and thorium tetrahydroxide complex Th(OH){sub 4} tend to resist complexation with acetates. The presence of complexing reagents in solution can release radionuclides such as uranium and/or thorium from the soil to the solution by forming soluble complexes. Sodium bicarbonate (NaHCO{sub 3}) and diethylenetriaminepentaacetic acid (DTPA) are strong complex formers that released 38% to 62% of total uranium activity and 78% to 86% of total thorium activity, respectively, from the soil samples investigated. Solutions of 0.1 molar sodium nitrate (NaNO{sub 3}) and 0.1 molar sodium sulfate (Na{sub 2}SO{sub 4}) were not effective complex formers with uranium and thorium under the experimental conditions. Fractions of uranium and thorium desorbed by 0.15g/200ml humic acid ranged from 4.62% to 6.17% and 1.59% to 7.09%, respectively. This work demonstrates the importance of a knowledge of solution chemistry in investigating the desorption of radionuclides.

  11. 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.

  12. Purification Testing for HEU Blend Program

    SciTech Connect (OSTI)

    Thompson, M.C. [Westinghouse Savannah River Company, AIKEN, SC (United States); Pierce, R.A.

    1998-06-01

    The Savannah River Site (SRS) is working to dispose of the inventory of enriched uranium (EU) formerly used to make fuel for production reactors. The Tennessee Valley Authority (TVA) has agreed to take the material after blending the EU with either natural or depleted uranium to give a {sup 235}U concentration of 4.8 percent low-enriched uranium will be fabricated by a vendor into reactor fuel for use in TVA reactors. SRS prefers to blend the EU with existing depleted uranium (DU) solutions, however, the impurity concentrations in the DU and EU are so high that the blended material may not meet specifications agreed to with TVA. The principal non-radioactive impurities of concern are carbon, iron, phosphorus and sulfur. Neptunium and plutonium contamination levels are about 40 times greater than the desired specification. Tests of solvent extraction and fuel preparation with solutions of SRS uranium demonstrate that the UO{sub 2} prepared from these solutions will meet specifications for Fe, P and S, but may not meet the specifications for carbon. The reasons for carbon remaining in the oxide at such high levels is not fully understood, but may be overcome either by treatment of the solutions with activated carbon or heating the UO{sub 3} in air for a longer time during the calcination step of fuel preparation.Calculations of the expected removal of Np and Pu from the solutions show that the specification cannot be met with a single cycle of solvent extraction. The only way to ensure meeting the specification is dilution with natural U which contains no Np or Pu. Estimations of the decontamination from fission products and daughter products in the decay chains for the U isotopes show that the specification of 110 MEV Bq/g U can be met as long as the activities of the daughters of U- 235 and U-238 are excluded from the specification.

  13. 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. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States)

    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.

  14. 200-UP-2 Operable Unit technical baseline report

    SciTech Connect (OSTI)

    Deford, D.H.

    1991-02-01

    This report is prepared in support of the development of a Remedial Investigation/Feasibility Study (RI/FS) Work Plan for the 200-UP-2 Operable Unit by EBASCO Environmental, Incorporated. It provides a technical baseline of the 200-UP-2 Operable Unit and results from an environmental investigation undertaken by the Technical Baseline Section of the Environmental Engineering Group, Westinghouse Hanford Company (Westinghouse Hanford). The 200-UP-2 Operable Unit Technical Baseline Report is based on review and evaluation of numerous Hanford Site current and historical reports, Hanford Site drawings and photographs and is supplemented with Hanford Site inspections and employee interviews. No field investigations or sampling were conducted. Each waste site in the 200-UP-2 Operable Unit is described separately. Close relationships between waste units, such as overflow from one to another, are also discussed. The 200-UP-2 Operable Unit consists of liquid-waste disposal sites in the vicinity of, and related to, U Plant operations in the 200 West Area of the Hanford Site. The U Plant'' refers to the 221-U Process Canyon Building, a chemical separations facility constructed during World War 2. It also includes the Uranium Oxide (UO{sub 3}) Plant, which was constructed at the same time and, like the 221-U Process Canyon Building, was later converted for other missions. Waste sites in the 200-UP-2 Operable Unit are associated with the U Plant Uranium Metal Recovery Program mission that occurred between 1952 and 1958 and the UO{sub 3} Plant's ongoing uranium oxide mission and include one or more cribs, reverse wells, french drains, septic tanks and drain fields, trenches, catch tanks, settling tanks, diversion boxes, waste vaults, and the lines and encasements that connect them. 11 refs., 1 tab.

  15. Tetrahalide Complexes of the [U(NR)(2)]2+ Ion: Synthesis, Theory, and Chlorine K-Edge X-ray Absorption Spectroscopy

    SciTech Connect (OSTI)

    Spencer, Liam P.; Yang, Ping; Minasian, Stefan G.; Jilek, Robert E.; Batista, Enrique R.; Boland, Kevin S.; Boncella, James M.; Conradson, S. D.; Clark, David L.; Hayton, Trevor W.; Kozimor, Stosh A.; Martin, Richard L.; MacInnes, Molly M.; Olson, Angela C.; Scott, Brian L.; Shuh, D. K.; Wilkerson, Marianne P.

    2013-02-13

    Synthetic routes to salts containing uranium bisimido tetrahalide anions [U(NR)(2)X-4](2-) (X = Cl-, Br-) and non-coordinating NEt4+ and PPh4+ countercations are reported. In general, these compounds can be prepared from U(NR)(2)I-2(THF)(x) (x = 2 and R = 'Bu, Ph; x = 3 and R = Me) upon addition of excess halide. In addition to providing stable coordination complexes with Cl-, the [U(NMe)(2)](2 +) cation also reacts with Br- to form stable [NEt4](2)[U(NMe)(2)Br-4] complexes. These materials were used as a platform to compare electronic structure and bonding in [U(NR)(2)](2+) with [UO2](2+). Specifically, Cl K-edge X-ray absorption spectroscopy (XAS) and both ground-state and time-dependent hybrid density functional theory (DFT and TDDFT) were used to probe U-Cl bonding interactions in [PPh4](2)[U((NBu)-Bu-t)(2)Cl-4] and [PPh4](2)[UO2Cl4]. The DFT and XAS results show the total amount of Cl 3p character mixed with the U 5f orbitals was roughly 7-10% per U-Cl bond for both compounds, which shows that moving from oxo to imido has little effect on orbital mixing between the U 5f and equatorial Cl 3p orbitals. The results are presented in the context of recent Cl K-edge XAS and DFT studies on other hexavalent uranium chloride systems with fewer oxo or imido ligands.

  16. 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)

  17. 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.

  18. Plutonium utilisation in future UK PWRs

    SciTech Connect (OSTI)

    Thomas, G. M.; Worrall, A. [Nexia Solutions Ltd. (Part of the BNFL Group of Companies), Springfield's Works, Preston, Lancashire (United Kingdom)

    2006-07-01

    Plutonium recycling in the form of Mixed Oxide (MOX) fuels is already a reality in over 30 reactors in Europe (in Belgium, Switzerland, Germany and France). Japan also plans to use MOX in approximately 30% of its reactors in the near future[1]. This paper describes potential near to mid-term disposition strategies for the United Kingdom's stockpile of plutonium. In order to be confident that MOX fuel can be utilised effectively in Pressurised Water Reactors (PWRs) in the UK, details are given of studies carried out recently at Nexia Solutions on PWR cores loaded with MOX containing typical UK plutonium isotopic compositions. Three dimensional steady state neutronic models of a standard Westinghouse four loop PWR design are constructed using state of the art tools (Studsvik of America's Core Management System[2, 3, 4]). Initially, a standard 18-month equilibrium UO{sub 2} fuel cycle is generated, followed by safety analyses and fuel performance calculations to demonstrate its feasibility. This equilibrium UO{sub 2} core is then gradually transitioned through loading patterns containing increasing MOX core loading fractions. Finally, an equilibrium MOX core loading pattern is determined. Technical safety analyses are also carried out on the transition cores and the final equilibrium scenario to ensure that all of the MOX cores are robust from a technical and safety viewpoint. Once these studies are completed the annual fuel throughputs for each scenario can be determined and used to produce options for managing the UK's plutonium stockpile. This work is part of a wider exercise currently being carried out by Nexia Solutions to explore the options for the safe disposition of the UK civil stockpile of separated PUO{sub 2}. (authors)

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

    SciTech Connect (OSTI)

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

    2003-08-01

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

  20. 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.

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

    SciTech Connect (OSTI)

    R. Finch

    2001-06-05

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

  2. Order-disorder in In{sup 3+} perovskites: The example of A(In{sub 2/3}B''{sub 1/3})O{sub 3} (A=Ba, Sr; B''=W, U)

    SciTech Connect (OSTI)

    Larregola, S.A. [Departamento de Quimica, Area de Quimica General e Inorganica, Facultad de Quimica, Bioquimica y Farmacia, Universidad Nacional de San Luis, Chacabuco y Pedernera, 5700 San Luis (Argentina)], E-mail: salarreg@unsl.edu.ar; Alonso, J.A. [Instituto de Ciencia de Materiales de Madrid, C.S.I.C., Cantoblanco, 28049 Madrid (Spain); Pinacca, R.M.; Viola, M.C.; Pedregosa, J.C. [Departamento de Quimica, Area de Quimica General e Inorganica, Facultad de Quimica, Bioquimica y Farmacia, Universidad Nacional de San Luis, Chacabuco y Pedernera, 5700 San Luis (Argentina)

    2008-10-15

    We describe the preparation and structural characterization of four In-containing perovskites from neutron powder diffraction (NPD) and X-ray powder diffraction (XRPD) data. Sr{sub 3}In{sub 2}B''O{sub 9} and Ba(In{sub 2/3}B''{sub 1/3})O{sub 3} (B''=W, U) were synthesized by standard ceramic procedures. The crystal structure of the W-containing perovskites and Ba(In{sub 2/3}U{sub 1/3})O{sub 3} have been revisited based on our high-resolution NPD and XRPD data, while for the new U-containing perovskite Sr{sub 3}In{sub 2}UO{sub 9} the structural refinement was carried out from high-resolution XRPD data. At room temperature, the crystal structure for the two Sr phases is monoclinic, space group P2{sub 1}/n, where the In atoms occupy two different sites Sr{sub 2}[In]{sub 2d}[In{sub 1/3}B''{sub 2/3}]{sub 2c}O{sub 6}, with a=5.7548(2) A, b=5.7706(2) A, c=8.1432(3) A, {beta}=90.01(1){sup o} for B''=W and a=5.861(1) A, b=5.908(1) A, c=8.315(2) A, {beta}=89.98(1){sup o} for B''=U. The two phases with A=Ba should be described in a simple cubic perovskite unit cell (S.G. Pm3-bar m) with In and B'' distributed at random at the octahedral sites, with a=4.16111(1) A and 4.24941(1) A for W and U compounds, respectively. - Graphical abstract: The structure of the new uranium-based double perovskite Sr{sub 3}In{sub 2}UO{sub 9} is described and the true symmetry of the other title compounds are revisited. The presence of long-range ordering in the Sr samples, by contrast with the Ba perovskites, is related with the smaller unit cell and B-B distances in the Sr oxides, promoting the electrostatic repulsions between highly charged W{sup 6+} and U{sup 6+} cations as driving force for the long-range B-site ordering.

  3. 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.

  4. Fissible Deposit Characterization at the Former Oak Ridge K-25 Gaseous Diffusion Plant by {sup 252}CF-Source-Driven Measurements

    SciTech Connect (OSTI)

    Hannon, T.F.; Mihalczo, J.T.; Mullens, J.A.; Uckan, T.; Valentine, T.E.; Wyatt, M.S.

    1998-05-01

    The Deposit Removal Project was undertaken with the support of the U. S. Department of Energy at the East Tennessee Technology Park (ETTP) formerly the Oak Ridge K-25 Site. The project team performed the safe removal of the hydrated uranyl fluoride (UO{sub 2}F{sub 2}) deposits from the K-29 Building of the former Oak Ridge Gaseous Diffusion Plant. The deposits had developed as a result of air leakage into UF{sub 6} gas process pipes; UO{sub 2}F{sub 2} became hydrated by moisture from the air and deposited inside the pipes. The mass, its distribution, and the hydrogen content [that is, the ratio of H to U (H/U)], were the key parameters that controlled the nuclear criticality safety of the deposits. Earlier gamma-ray spectrometry measurements in K-29 had identified the largest deposits in the building. The first and third largest deposits in the building were measured in this program. The first deposit, found in the Unit 2, Cell 7, B-Line Outlet process pipe (called the ''Hockey Stick'') was about 1,300 kg ({+-} 50% uncertainty) at 3.34 wt% {sup 235}U enrichment ({+-}50% uncertainty) and according to the gamma-ray spectroscopy was uniformly distributed. The second deposit (the third-largest deposit in the building), found in the Unit 2, Cell 6, A-Line Outlet process pipe (called the ''Tee-Pipe''), had a uranium deposit estimated to be about 240 kg ({+-} 50% uncertainty) at 3.4 wt % {sup 235}U enrichment ({+-} 20% uncertainty). Before deposit removal activities began, the Deposit Removal Project team needed to survey the inside of the pipes intrusively to assess the nuclear criticality safety of the deposits. Therefore, the spatial distribution of the deposits, the total uranium deposit mass, and the moderation level resulting from hydration of the deposits, all of which affect nuclear criticality safety were required. To perform the task safely and effectively, the Deposit Removal Project team requested that Oak Ridge National Laboratory (ORNL) characterize the two largest deposits with the {sup 252}Cf-source-driven transmission (CFSDT) technique, an active neutron interrogation method developed for use at the Oak Ridge Y-12 Plant to identify nuclear weapons components in containers. The active CFSDT measurement technique uses CFSDT time-of-flight measurements of prompt neutrons and gamma rays from an externally introduced {sup 252}Cf source.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Subsurface and surface oceanic diffusion experiments near Freeport, Texas 

    E-Print Network [OSTI]

    Berry, Alan Dale

    1981-01-01

    '0 ll U C nl 0 4i 'tl 0 0 0 J 0 0 0 Ui l4 '0 C 0 0 0 Ql S 0 n! C W 0 cC IL 0 0 ll '0 0 0 Sl 3 4 0 W 0 W 0 0 IJ C 0 S 0 U 0 C gJ 0 O' W 0 0 U lJ C g 0 0 ca 0 C N 0 0 nl 0 0 tV Cl C 0 '0 0 0 l4 '0 Ul Cl 0 C 5...U 0 0 '0 W 3 0 0 E 0 or E rt rU '0 0 rJ Ul 0 0 W W 0 0, Ul 0 Cl 0 0 w uf E C 0 + C . 0& tU 0 0'U C tU tU l 0 0 4 OD C= c c c c c c c apaog zaqaufouonpg uo uUtoqS sy uocqazuuaouoq atty 0 Ul W ul W O W O 0 IV 0...

  13. Experimental verification of a mathematical model of a semibatch biological reactor 

    E-Print Network [OSTI]

    Fox, Thomas Patrick

    1971-01-01

    (XII/SS'IN Sdl/PazrITZn p Sdl) azaII uoI:xazxI. xxlI uaMxp xxuII 3 ccJ CI Cc Ql 8 cc-c CJ CJ O CCJ CEc M Gl 0 0 0 0 0 ccc 0 0 0 CCI 0 0 CIJ CIJ CIJ CIJ cd Sl 5F CJ CJ 0 I' M H H p I I 0 Ccl 0 IA Gl 0 OJ Icc 0 IA O C CU aD C P1 CCJ...) may be written as: dC -KC C dt, sb (26) 00 8 03 O 00 0 Cl 4 ql C3 C ql 00 '4 O ql lq '0 ql '4 00 ql 'd 0 ql 4J 0 V (iqygSqN Slq/pp g, '80l) agap yeqalqaq aqSep qyug 55 dCb YKC C dt sb when C ( C + s s (27) B. nd dC s 2C Cb...

  14. Coyote-prey interactions on an intensively managed south Texas ranch 

    E-Print Network [OSTI]

    Drew, Gary Scott

    1988-01-01

    'sexaI 'fiquncg syTag wT(' uc pageoog 'ea~ qoJeasaH eqtdo) e) eq f60'l aqZ uo gg6i fiyn(' qSnoJqZ yS6( aunp woJg paxonpuoo seA& qoJeasag V3HV I((flIS PRECIPITATION IN CM CD l Dl CD 3 cD CD CD CD C p ~ CD Cl Dl O CD ~ Cn 4 O 0 3 CD C... Ch 'O CD O CD cn O 2 CD CD O O CJ D CD Cn CD ID CD 0 O T 0 n 0 0 0 D N 0 0 Z 0 CC 3 V 0 Il 0 0 IO n N 0 0 0 0 0 0 CL 0 CI 0 a~om r(ggeTgueZsqng 9S6 ( 'SuTunp auTT-souse Zsanqqnos aqq Suoge sseoueo go pasodsTp aneq o...

  15. Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

    SciTech Connect (OSTI)

    Da Cruz, D. F.; Rochman, D.; Koning, A. J. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, 1755 ZG Petten (Netherlands)

    2012-07-01

    This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {sup 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)

  16. 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)

  17. Initiation of nuclear reactions under laser irradiation of Au nanoparticles in the aqueous solution of Uranium salt

    E-Print Network [OSTI]

    A. V. Simakin; G. A. Shafeev

    2009-11-29

    Laser exposure of suspension of either gold or palladium nanoparticles in aqueous solutions of UO2Cl2 of natural isotope abundance was experimentally studied. Picosecond Nd:YAG lasers at peak power from 1011 to 1013 W/cm2 at the wavelength of 1064 and 355 nm were used as well as a visible-range Cu vapor laser at peak power of 1010 W/cm2. The composition of colloidal solutions before and after laser exposure was analyzed using atomic absorption and gamma spectroscopy between 0.06 and 1 MeV range of photon energy. A real-time gamma-spectroscopy was used to characterize the kinetics of nuclear reactions during laser exposure. It was found that laser exposure initiated nuclear reactions involving both 238U and 235U nuclei via different channels in H2O and D2O. The influence of saturation of both the liquid and nanoparticles by gaseous H2 and D2 on the kinetics of nuclear transformations was found. Possible mechanisms of observed processes are discussed.

  18. Nuclear electric propulsion /NEP/ spacecraft for the outer planet orbiter mission

    SciTech Connect (OSTI)

    Garrison, P.W.; Nock, K.T.

    1982-01-01

    The design, operating features, and a possible Neptune orbit for the spacecraft powered by the SP-100 nuclear electric propulsion (NEP) system under study by NASA and the DOE are described. The system features a reactor and a payload situated on opposite ends of a 0.5 m diam, 11 m long astromast. Mercury-ion thrusters are located beneath the reactor for side thrusting, and no contamination of the payload or obstruction of the viewing angles for scientific objectives occurs with the system, which would not degrade in performance even under high insolation during near-sun maneuvers. Results of a theoretical study of earth escapes are presented to show that an NEP powered spiral trajectory out of a 700 km Shuttle orbit and using a Triton gravity assist would be superior to departing from a 300 km orbit with a Centaur boost. The mission profile includes a 1249 kg Galileo payload. The SP-100 has a 1.4 MWth reactor with UO2 fuel tiles and weighs 19,904 kg.

  19. 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 ?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 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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  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. 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.

  6. 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.

  7. Iron (III) Matrix Effects on Mineralization and Immobilization of Actinides

    SciTech Connect (OSTI)

    Cynthia-May S. Gong; Tyler A. Sullens; Kenneth R. Czerwinski

    2006-01-01

    Abstract - A number of models for the Yucca Mountain Project nuclear waste repository use studies of actinide sorption onto well-defined iron hydroxide materials. In the case of a waste containment leak, however, a complex interaction between dissolved waste forms and failed containment vessel components can lead to immediate precipitation of migratory iron and uranyl in the silicate rich near-field environment. Use of the Fe(III) and UO22+ complexing agent acetohydroxamic acid (AHA) as a colorimetric agent for visible spectrophotometry is well-known. Using the second derivative of these spectra a distinct shift in iron complexation in the presence of silicate is seen that is not seen with uranyl or alone. Silica also decreases the ability of uranyl and ferric solutions to absorb hydroxide, hastening precipitation. These ferric silicate precipitates are highly amorphous and soluble. Precipitates formed in the presence of uranyl below ~1 mol% exhibit lower solubility than precipitates from up to 50 mol % and of uranyl silicates alone.

  8. 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)

  9. Use of Source Term and Air Dispersion Modeling in Planning Demolition of Highly Alpha-Contaminated Buildings

    SciTech Connect (OSTI)

    Droppo, James G.; Napier, Bruce A.; Rishel, Jeremy P.; Bloom, Richard W.

    2011-06-22

    The current cleanup of structures related to cold-war production of nuclear materials includes the need to demolish a number of highly alpha-contaminated structures. The process of planning for the demolition of such structures includes unique challenges related to ensuring the protection of both workers and the public. Pre-demolition modeling analyses were conducted to evaluate potential exposures resulting from the proposed demolition of a number of these structures. Estimated emission rates of transuranic materials during demolition are used as input to an air-dispersion model. The climatological frequencies of occurrence of peak air and surface exposures at locations of interest are estimated based on years of hourly meteorological records. The modeling results indicate that downwind deposition is the main operational limitation for demolition of a highly alpha-contaminated building. The pre-demolition modeling directed the need for better contamination characterization and/or different demolition methods—and in the end, provided a basis for proceeding with the planned demolition activities. Post-demolition modeling was also conducted for several contaminated structures, based on the actual demolition schedule and conditions. Comparisons of modeled and monitoring results are shown. Recent monitoring data from the demolition of a UO3 plant shows increments in concentrations that were previously identified in the pre-demolition modeling predictions; these comparisons confirm the validity and value of the pre-demolition source-term and air dispersion computations for planning demolition activities for other buildings with high levels of radioactive contamination.

  10. 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.

  11. Uranyl fluoride luminescence in acidic aqueous solutions

    SciTech Connect (OSTI)

    Beitz, J.V.; Williams, C.W. [Argonne National Lab., IL (United States). Chemistry Div.

    1996-08-01

    Luminescence emission spectra and decay rates are reported for uranyl species in acidic aqueous solutions containing HF or added NaF. The longest luminescence lifetime, 0.269 {+-} 0.006 ms, was observed from uranyl in 1 M HF + 1 M HClO{sub 4} at 296 K and decreased with increasing temperature. Based on a luminescence dynamics model that assumes equilibrium among electronically excited uranyl fluoride species and free fluoride ion, this long lived uranyl luminescence in aqueous solution is attributed primarily to UO{sub 2}F{sub 2}. Studies on the effect of added LiNO{sub 3} or Na{sub 2}WO{sub 4}{center_dot}2H{sub 2}O showed relatively weak quenching of uranyl fluoride luminescence which suggests that high sensitivity determination of the UF{sub 6} content of WF{sub 6} gas should be feasible via uranyl luminescence analysis of hydrolyzed gas samples of impure WF{sub 6}.

  12. Analysis of criticality alarm system response to an accidental criticality outside the cascade process buildings at the Portsmouth Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Negron, S.B.; Tayloe, R.W. Jr.; Dobelbower, M.C. [Battelle, Columbus, OH (United States)

    1994-07-01

    Neutron dose rates at detector positions within the X-326, X-330, and X-333 buildings were evaluated for an accidental criticality outside of each building. As fissile material bearing equipment and containers are moved to and from each building, the possibility exists for a criticality accident to occur. This analysis demonstrates that a criticality accident which occurs at any position on the access roads alongside a process building can be detected. The detectable area includes all points within the access road boundary along each face of each building. This analysis also demonstrates that the criticality alarm systems of the process buildings will respond to criticality events occurring within the tie lines connecting the process buildings. This analysis was performed using the MCNP Monte Carlo neutron-proton transport code. The radiation source is the neutron leakage spectrum of a critical solution of 4.95 percent enriched UO{sub 2}F{sub 2}-H{sub 2}O at a power level corresponding to the ANSI ANS 8.3. Standard minimum accident of concern. The evaluated neutron fluxes were converted to neutron dose rates by use of the Henderson free-in-air response functions. Critical source positions correspond to the farthest source to detector distances on the access roads along each face of the three buildings, and the centerpoint of the building tie lines. This report contains the methodology used for this study, a background on the data used, and a section about the assumptions and limits to all conclusions.

  13. {sup 252}Cf-source-correlated transmission measurements for uranyl fluoride deposit in a 24-in.-OD process pipe

    SciTech Connect (OSTI)

    Uckan, T.; Mihalczo, J.T.; Valentine, T.E.; Mullens, J.A. [Oak Ridge National Lab., TN (United States). Instrumentation and Controls Div.; Wyatt, M.S. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Hannon, T.F. [Central Engineering Services, Oak Ridge, TN (United States)

    1998-06-01

    Characterization of a hydrated uranyl fluoride (UO{sub 2}F{sub 2}{center_dot}nH{sub 2}O) deposit in a 17-ft-long, 24-in.-OD process pipe at the former Oak Ridge Gaseous Diffusion Plant was successfully performed by using {sup 252}Cf-source-correlated time-of-flight (TOF) transmission measurements. These measurements of neutrons and gamma rays through the pipe from an external {sup 2521}Cf fission source were used to measure the deposit profile and its distribution along the pipe, the hydration (or H/U), and the total uranium mass. The measurements were performed with a source in an ionization chamber on one side of the pipe and detectors on the other. Scanning the pipe vertically and horizontally produced a spatial and time-dependent radiograph of the deposit in which transmitted gamma rays and neutrons were separated in time. The cross-correlation function between the source and the detector was measured with the Nuclear Weapons Identification System. After correcting for pipe effects, the deposit thickness was determined from the transmitted neutrons and H/U from the gamma rays. Results were consistent with a later intrusive observation of the shape and the color of the deposit; i.e., the deposit was annular and was on the top of the pipe at some locations, demonstrating the usefulness of this method for deposit characterization.

  14. Newly recognized hosts for uranium in the Hanford Site vadose zone

    SciTech Connect (OSTI)

    Stubbs, Joanne E.; Veblen, Linda A.; Elbert, David; Zachara, John M.; Davis, James A.; Veblen, David R.

    2009-03-15

    Uranium contaminated sediments from the U.S. Department of Energy’s Hanford Site have been investigated using electron microscopy. Six classes of solid hosts for uranium were identified. Preliminary sediment characterization was carried out using optical petrography, and electron microprobe analysis (EMPA) was used to locate materials that host uranium. All of the hosts are fine-grained and intergrown with other materials at spatial scales smaller than the analytical volume of the electron microprobe. A focused ion beam (FIB) was used to prepare electron-transparent specimens of each host for the transmission electron microscope (TEM). The hosts were identified as: 1) metatorbernite [Cu(UO2)2(PO4)2·8H2O]; 2) coatings comprised mainly of phyllosilicates on sediment clasts; 3) an amorphous zirconium (oxyhydr)oxide found in clast coatings; 4) amorphous and poorly crystalline materials that line voids within basalt lithic fragments; 5) amorphous palagonite surrounding fragments of basaltic glass; and 6) Fe- and Mnoxides. These findings demonstrate the effectiveness of combining EMPA, FIB, and TEM to identify solid-phase contaminant hosts. Furthermore, they highlight the complexity of U geochemistry in the Hanford vadose zone, and illustrate the importance of microscopic transport in controlling the fate of contaminant metals in the environment.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    1999-01-01

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

  1. 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.

  2. 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'.

  3. 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.

  4. 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.

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01

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

  6. EPRI Cogeneration Models -- DEUS and COPE 

    E-Print Network [OSTI]

    Mauro, R.; Hu, S. D.

    1983-01-01

    .00 0.00 0.00 0.91 0.90 1.05 '''?.lItipH.rl YutlbllOIM Auxiliary Power Fractton 0.50 0.67 0.50 0.00 0.00 0.00 0.&5 0.B5 0.90 0.00 0.02 0.02 0.00 0.00 0.00 0.02 0.02 0.02 Auxiliary Helt Friction 0.03 0.0' 0.00 0.00 0.00 0.00 O.Ol 0.0' 0.03 Envtro""nul... ........................................................................... ? CJ~U' ? ,u(O ? ? ? ,,""'u ? In.,," ? ? nCHNClLOCt. C"AhClS ? ,\\.I?l. Cc,u. (I,," ? Jl("'[NUlI ? IUVlNlJU ? ......................, " , ? NO c.oClliI' 1....3. H."" 3.77. t. ? 47.1" ?......-- - ---.-.- -_ - - -_ --- -..-..? ?en I4U'ettt Z~.4I' U...

  7. Nondestructive NMR technique for moisture determination in radioactive materials.

    SciTech Connect (OSTI)

    Aumeier, S.; Gerald, R.E. II; Growney, E.; Nunez, L.; Kaminski, M.

    1998-12-04

    This progress report focuses on experimental and computational studies used to evaluate nuclear magnetic resonance (NMR) spectroscopy and magnetic resonance imaging (MRI) for detecting, quantifying, and monitoring hydrogen and other magnetically active nuclei ({sup 3}H, {sup 3}He, {sup 239}Pu, {sup 241}Pu) in Spent nuclear fuels and packaging materials. The detection of moisture by using a toroid cavity NMR imager has been demonstrated in SiO{sub 2} and UO{sub 2} systems. The total moisture was quantified by means of {sup 1}H NMR detection of H{sub 2}O with a sensitivity of 100 ppm. In addition, an MRI technique that was used to determine the moisture distribution also enabled investigators to discriminate between bulk and stationary water sorbed on the particles. This imaging feature is unavailable in any other nondestructive assay (NDA) technique. Following the initial success of this program, the NMR detector volume was scaled up from the original design by a factor of 2000. The capacity of this detector exceeds the size specified by DOE-STD-3013-96.

  8. 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.

  9. 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.

  10. 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. [Los Alamos Technical Associates, Kennewick, WA (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

    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.

  11. 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.

  12. 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)

  13. Fabrication of high exposure nuclear fuel pellets

    DOE Patents [OSTI]

    Frederickson, James R. (Richland, WA)

    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.

  14. Wastes associated with recycling spent MOX fuel into fast reactor oxide fuel

    SciTech Connect (OSTI)

    Foare, G.; Meze, F. [AREVA EP, SGN - 1, rue des Herons, 18182 Montigny-le-Bretonneux (France); McGee, D.; Murray, P.; Bader, S. [AREVA Federal Services LLC - 7207 IBM Drive, Charlotte, NC 28262 (United States)

    2013-07-01

    A study sponsored by the DOE has been performed by AREVA to estimate the process and secondary wastes produced from an 800 MTIHM/yr (initial metric tons heavy metal a year) recycling plant proposed to be built in the U.S. utilizing the COEX process and utilized some DOE defined assumptions and constraints. In this paper, this plant has been analyzed for a recycling campaign that included 89% UO{sub x} and 11% MOX UNF to estimate process and secondary waste quantities produced while manufacturing 28 MTIHM/yr of SFR fuel. AREVA utilized operational data from its backend facilities in France (La Hague and MELOX), and from recent advances in waste treatment technology to estimate the waste quantities. A table lists the volumes and types of the different final wastes for a recycling plant. For instance concerning general fission products the form of the final wastes is vitrified glass and its volume generation rate is 135 l/MTHM, concerning Iodine 129 waste its final form is synthetic rock and its volume generation rate is 0.625 l/MTIHM.

  15. Development of NF3 Deposit Removal Technology for the Portsmouth Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Scheele, R.D.; McNamara, B.K.; Rapko, B.M.; Edwards, M.K.; Kozelisky, A.E.; Daniel, R.C. [Battelle Pacific Northwest Division, PO Box 999, Battelle Blvd, Richland, Washington 99352 (United States); McSweeney, T.I.; Maharas, S.J.; Weaver, P.J.; Iwamasa, K.J. [Battelle Columbus Operations, 505 King Avenue, Columbus, Ohio 43201 (United States); Kefgen, R.B. [WASTREN, Inc., 1864 Shyville Road, Piketon, Ohio 45661 (United States)

    2006-07-01

    This paper summarizes the Battelle, Stoller, and WASTREN (BSW) team's efforts, to date, in support of the United States Department of Energy's plans to remove uranium and technetium deposits before decommissioning the Portsmouth Gaseous Diffusion Plant. The BSW team investigated nitrogen trifluoride (NF{sub 3}) as a safer yet effective alternative gaseous treatment to the chlorine trifluoride (ClF{sub 3})-elemental fluorine (F{sub 2}) treatment currently used to remove uranium and technetium deposits from the uranium enrichment cascade. Both ClF{sub 3} and F{sub 2} are highly reactive, toxic, and hazardous gases, while NF{sub 3}, although toxic [1], is no more harmful than moth balls [2]. BSW's laboratory thermo-analytical and laboratory-scale prototype studies with NF{sub 3} established that thermal NF{sub 3} can effectively remove likely and potential uranium (UO{sub 2}F{sub 2} and UF{sub 4}) and technetium deposits (a surrogate deposit material, TcO{sub 2}, and pertechnetates) by conversion to volatile compounds. Our engineering evaluations suggest that NF{sub 3}'s effectiveness could be enhanced by combining with a lesser concentration of ClF{sub 3}. BSW's and other's studies indicate compatibility with Portsmouth materials of construction (aluminum, copper, and nickel). (authors)

  16. The relative variational model: A topological view of matter and its properties: Specific heat and enthalpy

    SciTech Connect (OSTI)

    Dias, M. S.; De Vasconcelos, V.; Mattos, J. R. L. [Center for Development of the Nuclear Technology - CDTN, National Commission for the Nuclear Energy - CNEN, PO Box: 941, 30.161-970, Belo Horizonte, Minas Gerais (Brazil); Jordao, E. [Chemistry Engineering Dept., Campinas State Univ., FEQ/ UNICAMP, Av. Albert Einstein, 500, 13083-852, Campinas, Sao Paulo (Brazil)

    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 Debye's integral of the heat capacity at constant volume is analytically solved. The experimental data of the specific heat at constant pressure and the enthalpy variations are also analytically depicted by the model in the temperature range of 0 K up to the melting point. The data reductions were applied to the oxides Al{sub 2}O{sub 3} and UO{sub 2}. (authors)

  17. Sources of secondary radionuclide releases from Hanford Operations

    SciTech Connect (OSTI)

    Heeb, C.M.; Gydesen, S.P.

    1994-05-01

    This report considers Hanford facilities and operations with the potential to be secondary radionuclide release sources. Facilities that produced radionuclides or processed products of fission reactions and were not covered in previous source term reports are included in this report. The following facilities are described and any potentially significant releases from them are estimated: PUREX (1956--1972, 1983--1988) and REDOX (1952--1967)--campaigns with non-standard feed material (materials other than fuel from single-pass reactors); C PLANT (Hot Semi-Works)--pilot plant and strontium recovery; Z Plant--plutonium finishing; U and UO{sub 3} Plants--uranium recovery; 108 B Plant--tritium extraction; 300 Area Plutonium Recycle Test Reactor (PRTR); 300 Area Low Power Test Reactors; Criticality Accidents; and 400 Area Fast Flux Test Facility (FFTF). The method of analysis was to examine each facility, give a brief description of its purpose and operations, and describe the types of material the facility processed as an indication of the radionuclides it had the potential to release. Where possible, specific radionuclides are estimated and values from the original documents are reported.

  18. Deactivation completed at historic Hanford Fuels Laboratory

    SciTech Connect (OSTI)

    Gerber, M.S.

    1994-03-01

    This report discusses deactivation work which was completed as of March 31, 1994 at the 308 Fuels Development Laboratory (FDL) at the Hanford Site near Richland, Washington. The decision to deactivate the structure, formerly known as the Plutonium Fabrication Pilot Plant (PFPP), was driven by a 1980s Department of Energy (DOE) decision that plutonium fuels should not be fabricated in areas near the Site`s boundaries, as well as by changing facility structural requirements. Inventory transfer has been followed by the cleanout and stabilization of plutonium oxide (PuO{sub 2}) and enriched uranium oxide (UO{sub 2}) residues and powders in the facility`s equipment and duct work. The Hanford Site, located in southeastern Washington state, was one of America`s primary arsenals of nuclear defense production for nearly 50 years beginning in World War II. Approximately 53 metric tons of weapons grade plutonium, over half of the national supply and about one quarter of the world`s supply, were produced at Hanford between 1944 and 1989. Today, many Site buildings are undergoing deactivation, a precursor phase to decontamination and decommissioning (D&D). The primary difference between the two activities is that equipment and structural items are not removed or torn down in deactivation. However, utilities are disconnected, and special nuclear materials (SNM) as well as hazardous and pyrophoric substances are removed from structures undergoing this process.

  19. 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.

  20. 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.