Sample records for uo uo uo

  1. UoS Motor Accident Report Form COMPANY DETAILS

    E-Print Network [OSTI]

    Sussex, University of

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

  2. UO Policy Library Resource for

    E-Print Network [OSTI]

    Oregon, University of

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

  3. UO

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

    success in this work will deepen our fundamental understanding of the nuclear energy materials. Background: The actinide (U, Np, Pu) oxides, nitrides, and carbides are of...

  4. OXYGEN DIFFUSION IN UO2-x

    E-Print Network [OSTI]

    Kim, K.C.

    2013-01-01T23:59:59.000Z

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

  5. Crystal fields in UO2 - revisited

    SciTech Connect (OSTI)

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

    2009-01-01T23:59:59.000Z

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

  6. Spectroscopic Studies of the Several Isomers of UO3

    SciTech Connect (OSTI)

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

    2013-09-26T23:59:59.000Z

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

  7. Investigations into the Polymorphs and Hydration Products of UO3

    SciTech Connect (OSTI)

    Sweet, Lucas E.; Buck, Edgar C.; Henager, Charles H.; Hu, Shenyang Y.; Meier, David E.; Peper, Shane M.; Schwantes, Jon M.; Su, Yin-Fong; Sams, Robert L.; Blake, Thomas A.; Johnson, Timothy J.; Kulp, Thomas J.; Sommers, Ricky L.; Sugar, Joshua D.; Chames, Jeffrey D.

    2012-04-27T23:59:59.000Z

    This work focuses on progress in gaining a better understanding of the polymorphic nature of the UO{sub 3} and UO{sub 3}-water system; one of several important materials associated with the nuclear fuel cycle. The UO{sub 3}-water system is complex and has not been fully characterized, even though these species are common throughout the fuel cycle. For example, most production schemes for UO{sub 3} result in a mixture of up to six different polymorphic phases, and small differences in these conditions will affect phase genesis that ultimately results in measureable changes to the end product. Here we summarize our efforts to better characterize the UO{sub 3}-water system with optical techniques for the purpose of developing some predictive capability of estimating process history and utility, e.g. for polymorphic phases of unknown origin. Specifically, we have investigated three industrially relevant production pathways of UO{sub 3} and discovered a previously unknown low temperature route to {beta}-UO{sub 3}. Powder x-ray diffraction and optical spectroscopies were utilized in our characterization of the UO{sub 3}-water system. Pure phases of UO{sub 3}, its hydrolysis products and starting materials were used to establish optical spectroscopic signatures for these compounds. Preliminary aging studies were conducted on the {alpha}- and {gamma}-phases of UO{sub 3}.

  8. advanced doped uo2: Topics by E-print Network

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

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

  9. First-principles study of defects and phase transition in UO2...

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

    defects and phase transition in UO2. First-principles study of defects and phase transition in UO2. Abstract: The electronic properties, structure and phase transformation of UO2...

  10. advanced uo2 fuel: Topics by E-print Network

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

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

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

    SciTech Connect (OSTI)

    Washenfelder, D.J.

    1993-12-01T23:59:59.000Z

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

  12. Spark Plasma Sintering of W-UO2 Cermets

    SciTech Connect (OSTI)

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

    2013-02-01T23:59:59.000Z

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

  13. 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.

    2009-11-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Clapp, D.A.

    1995-01-01T23:59:59.000Z

    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.

  15. 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

    2014-10-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2014-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1995-01-01T23:59:59.000Z

    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).

  18. Cation-Cation Interactions in [(UO2)2(OH)n](4-n) Complexes. ...

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

    and (UO2)2(OH)5- shows that the CCIs and bridging hydroxo between the dioxo-uranium units are mainly electrostatic in nature. Citation: Odoh SO, N Govind, G...

  19. INTERPRETATION OF TRACER SURFACE DIFFUSION EXPERIMENTS ON UO2 ROLES OF GAS AND SOLID TRANSPORT PROCESSES

    E-Print Network [OSTI]

    Olander, D.R.

    2013-01-01T23:59:59.000Z

    hydrogen. This extrapolation was made using Eq [2) and the values of the trans- port properties (hydrogen in the MK experiment means that the UO is probably not stoichiometric. That this transport property

  20. Modeling thermal conductivity in UO2 with BeO additions as a function of microstructure

    E-Print Network [OSTI]

    direction. Ó 2009 Elsevier B.V. All rights reserved. 1. Introduction Uranium dioxide (UO2) is the most gradients which affect heat removal and overall reactor performance. These thermal gradients strongly

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

    SciTech Connect (OSTI)

    John M. Veilleux

    1999-08-01T23:59:59.000Z

    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.

  2. Benchmarking of Graphite Reflected Critical Assemblies of UO2

    SciTech Connect (OSTI)

    Margaret A. Marshall; John D. Bess

    2011-11-01T23:59:59.000Z

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

  3. PUREX/UO3 Facilities deactivation lessons learned history

    SciTech Connect (OSTI)

    Gerber, M.S.

    1996-09-19T23:59:59.000Z

    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

  4. 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

    2014-02-28T23:59:59.000Z

    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.

  5. 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

    2014-01-01T23:59:59.000Z

    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.

  6. 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

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

    SciTech Connect (OSTI)

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

    2005-06-13T23:59:59.000Z

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

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

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

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

  9. UO Retirement Recognition Reception Pap Reception Hall, J.S. Museum of Art

    E-Print Network [OSTI]

    Oregon, University of

    UO Retirement Recognition Reception Papé Reception Hall, J.S. Museum of Art Thursday, June 16, 2011 Information Services 29½ Years Richard Edwards Early Childhood CARES 11 Years Kayla Hinds Architecture will not be attending the retirement reception. They asked that we announce their retirement. Robert Barton Theatre Arts

  10. f Fk66iCOP-] NBSIR 86-3422 uoL_ i 1

    E-Print Network [OSTI]

    Oak Ridge National Laboratory

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

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

    SciTech Connect (OSTI)

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

    1984-02-01T23:59:59.000Z

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

  13. 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)

    2014-03-07T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1997-12-08T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1998-05-11T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2015-01-01T23:59:59.000Z

    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.

  17. Computational Investigation of the Formation of Hyper-stoichiometric Uranium Dioxide (UO{sub 2+x})

    SciTech Connect (OSTI)

    Skomurski, Frances; Becker, Udo; Ewing, Rodney [Geological Sciences, University of Michigan, 2534 C.C. Little Building, 1100 North University Ave., Ann Arbor, MI, 48109 (United States)

    2007-07-01T23:59:59.000Z

    Understanding the mechanisms behind the formation of hyper-stoichiometric UO{sub 2} phases is important because oxidation of uranium atoms upon the addition of excess oxygen to the UO{sub 2} structure leads to volume changes that increase the susceptibility of spent fuel to corrosion. While a variety of diffraction and spectroscopic studies have been used to investigate structural changes as UO{sub 2} oxidizes to U{sub 4}O{sub 9}, the effect of interstitial oxygen on the charge distribution of uranium in hyper-stoichiometric UO{sub 2} remains inconclusive. In this study, quantum mechanical techniques were used to model the effects of interstitial oxygen on the structure and charge distribution of atoms in a simplified U{sub 4}O{sub 9} unit cell. A density functional theory-based approach was used to optimize the geometry and charge distribution of a variety of U{sub 4}O{sub 9} starting models with different U{sup 4+}, U{sup 5+} and U{sup 6+} charge configurations. Results from our calculations suggest that the formation of one U{sup 5+} per addition of interstitial oxygen at a perpendicular bisector site is favorable; this oxidation event is accompanied by partial reduction of the interstitial oxygen atom. Deflection of two lattice oxygen atoms along the body diagonal of the cubic site surrounding the U{sup 5+} is also observed upon the addition of one interstitial oxygen atom. Structural and bond length data are compared with experimental data whenever possible. (authors)

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

    SciTech Connect (OSTI)

    Wittman, Richard S.; Buck, Edgar C.

    2012-09-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1994-10-18T23:59:59.000Z

    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].

  20. 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)

    1994-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1991-11-01T23:59:59.000Z

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

  2. 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

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

    SciTech Connect (OSTI)

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

    1989-11-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1990-12-31T23:59:59.000Z

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

  5. Assessment of structures and stabilities of defect clusters and surface energies predicted by nine interatomic potentials for UO2

    SciTech Connect (OSTI)

    Stephen A. Taller; Xian-Ming Bai

    2013-11-01T23:59:59.000Z

    The irradiation in nuclear reactors creates many point defects and defect clusters in uranium dioxide (UO2) and their evolution severely degrades the thermal and mechanical properties of the nuclear fuels. Previously many empirical interatomic potentials have been developed for modeling defect production and evolution in UO2. However, the properties of defect clusters and extended defects are usually not fitted into these potentials. In this work nine interatomic potentials for UO2 are examined by using molecular statics and molecular dynamics to assess their applicability in predicting the properties of various types of defect clusters in UO2. The binding energies and structures for these defect clusters have been evaluated for each potential. In addition, the surface energies of voids of different radii and (1 1 0) flat surfaces predicted by these potentials are also evaluated. It is found that both good agreement and significant discrepancies exist for these potentials in predicting these properties. For oxygen interstitial clusters, these potentials predict significantly different defect cluster structures and stabilities; For defect clusters consisting of both uranium and oxygen defects, the prediction is in better agreement; The surface energies predicted by these potentials have significant discrepancies, and some of them are much higher than the experimentally measured values. The results from this work can provide insight on interpreting the outcome of atomistic modeling of defect production using these potentials and may provide guidelines for choosing appropriate potential models to study problems of interest in UO2.

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

    SciTech Connect (OSTI)

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

    2014-08-20T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2006-03-01T23:59:59.000Z

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

  8. 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)

  9. 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

    2007-06-01T23:59:59.000Z

    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.

  10. Nuclear data uncertainties by the PWR MOX/UO{sub 2} core rod ejection benchmark

    SciTech Connect (OSTI)

    Pasichnyk, I.; Klein, M.; Velkov, K.; Zwermann, W.; Pautz, A. [Boltzmannstr. 14, D-85748 Garching b. Muenchen (Germany)

    2012-07-01T23:59:59.000Z

    Rod ejection transient of the OECD/NEA and U.S. NRC PWR MOX/UO{sub 2} core benchmark is considered under the influence of nuclear data uncertainties. Using the GRS uncertainty and sensitivity software package XSUSA the propagation of the uncertainties in nuclear data up to the transient calculations are considered. A statistically representative set of transient calculations is analyzed and both integral as well as local output quantities are compared with the benchmark results of different participants. It is shown that the uncertainties in nuclear data play a crucial role in the interpretation of the results of the simulation. (authors)

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    M. Pope; S. Bays; R. Ferrer

    2008-03-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Veleckis, E.; Hoh, J.C.

    1991-03-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Naramore, Michael J

    2010-08-03T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Naramore, Michael J

    2010-08-03T23:59:59.000Z

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

  16. Forest transitions and ecosystem services in Zimbabwe Supervisors: Dr Casey Ryan (UoE), Dr Isla Grundy (University of Zimbabwe)

    E-Print Network [OSTI]

    Forest transitions and ecosystem services in Zimbabwe Supervisors: Dr Casey Ryan (UoE), Dr Isla and a variety of other ecosystem services. However the expansion of agricultural land and the curing of tobacco is accelerating deforestation and forest degradation rates. These structural changes to the ecosystem threaten

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

    E-Print Network [OSTI]

    Chen, Tianyi

    2013-08-02T23:59:59.000Z

    showing formation of intermetallic phases ........................................................................................................... 12 Figure 2.1 The thermal conductivity of UO2 as a function on temperature .................... 15... in the fissile materials creates localized heating in the fuel element, which is transferred to a coolant, and then be used to produce mechanical energy and ultimately electricity. Understanding the properties and behaviors of these fissile material compounds...

  18. 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

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

    SciTech Connect (OSTI)

    Michael R Tonks; Yongfeng Zhang; Xianming Bai

    2014-06-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Wyatt, D.M.

    1994-12-06T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1985-04-01T23:59:59.000Z

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

  2. The Effect Of Ion Implantation Of Selected Dopants On Some Of The Electrical Properties Of UO2

    SciTech Connect (OSTI)

    Haire, M. J.; von Roedern, R. J.; Meek, T. T.; Tesmer, J.; Wetteland, C.

    2003-02-25T23:59:59.000Z

    The United States Department of Energy (DOE) has {approx}1 billion pounds of surplus depleted uranium (i.e., uranium tails) from uranium gaseous diffusion enrichment facilities. Rather than treating this material as waste, DOE is investigating potential beneficial uses for this uranium. Of the many possible uses, uranium dioxide (UO2) has properties that make it an equal to or better than conventional photovoltaic (e.g., solar cell) materials. For example, the electronic bandgap of UO2 occurs at an efficiency equal to that of GaS and Si, and it has five radiation adsorption peaks instead of one. This paper describes the experimental work being conducted to develop urania photovoltaic devices.

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

    SciTech Connect (OSTI)

    S. Utsunomiya; R.C. Ewing

    2005-09-08T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-01-01T23:59:59.000Z

    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.

  5. 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

    2014-03-10T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2013-10-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Duncan, D.R.

    1994-06-30T23:59:59.000Z

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

  8. Molecular dynamics simulations of intergranular fracture in UO2 with nine empirical interatomic potentials

    SciTech Connect (OSTI)

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

    2014-09-01T23:59:59.000Z

    The intergranular fracture behavior of UO2 was studied using molecular dynamics simulations with a bicrystal model. The anisotropic fracture behavior due to the different grain boundary characters was investigated with the View the MathML source symmetrical tilt S5 and the View the MathML source symmetrical tilt S3 ({1 1 1} twin) grain boundaries. Nine interatomic potentials, seven rigid-ion plus two core–shell ones, were utilized to elucidate possible potential dependence. Initiating from a notch, crack propagation along grain boundaries was observed for most potentials. The S3 boundary was found to be more prone to fracture than the S5 one, indicated by a lower energy release rate associated with the former. However, some potential dependence was identified on the existence of transient plastic deformation at crack tips, and the results were discussed regarding the relevant material properties including the excess energies of metastable phases and the critical energy release rate for intergranular fracture. In general, local plasticity at crack tips was observed in fracture simulations with potentials that predict low excess energies for metastable phases and high critical energy release rates for intergranular fracture.

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

    SciTech Connect (OSTI)

    Margaret A. Marshall; John D. Bess

    2012-11-01T23:59:59.000Z

    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].

  10. WIMS/PANTHER analysis of UO{sub 2}/MOX cores using embedded super-cells

    SciTech Connect (OSTI)

    Knight, M.; Bryce, P. [EDF Energy, Barnett Way, Barnwood, Gloucester (United Kingdom); Hall, S. [Advanced Modelling and Computation Group, Imperial College, London (United Kingdom)

    2012-07-01T23:59:59.000Z

    This paper describes a method of analysing PWR UO{sub 2}MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

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

    SciTech Connect (OSTI)

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

    1991-07-01T23:59:59.000Z

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

  12. Dissolution Behaviour of UO{sub 2} in Anoxic Conditions: Comparison of Ca-Bentonite and Boom Clay

    SciTech Connect (OSTI)

    Mennecart, Thierry; Cachoir, Christelle; Lemmens, Karel [SCK-CEN, Boeretang 200, MOL, 2400 (Belgium)

    2007-07-01T23:59:59.000Z

    In order to determine in how far the clay properties influence the dissolution of spent fuel, experiments were carried out with depleted UO{sub 2} in the presence of either compacted dry Ca-bentonite with Boom Clay groundwater (KB-BCW) or compacted dry Boom Clay with Boom Clay groundwater (BC-BCW). The leach tests were performed at 25 deg. C in anoxic atmosphere for 2 years. The U concentrations in the clay water were followed during these 2 years, and the amount of U in the clay was determined after 2 years in order to determine the UO{sub 2} dissolution rate. The uranium concentration after 0.45 {mu}m filtration was 50 times higher in the Boom Clay with Boom Clay water (2.0 x 10{sup -7} mol.L{sup -1}) than in Ca-bentonite with Boom Clay water (6.5 x 10{sup -9} mol.L{sup -1}), probably due to colloid formation in the Boom Clay system. Most released uranium was found in the clay. The fraction of uranium, dissolved from the UO{sub 2} pellet and found on the clay represents about 42 % of total uranium release in the system BC-BCW and more than 76 % in the system KB-BCW. The higher uranium retention of Boom Clay goes together with a higher dissolution rate. Global dissolution rates were estimated at about 2.0 x 10{sup -2} {mu}g.cm{sup -2}.d{sup -1} for the BCBCW system and 3.4 x 10{sup -3} {mu}g.cm{sup -2}.d{sup -1} for the KB-BCW system. This is not much lower than for similar tests with spent fuel, reported in literature. (authors)

  13. 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

    2012-05-30T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Gary S. Groenewold; Anita K. Gianotto

    2006-03-01T23:59:59.000Z

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

  15. Vibrational Spectroscopy of Mass-Selected [UO?(ligand)n]²? Complexes in the Gas Phase: Comparison with Theory

    SciTech Connect (OSTI)

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

    2006-03-18T23:59:59.000Z

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

  16. Evaluation of UF{sub 6}-to-UO{sub 2} conversion capability at commercial nuclear fuel fabrication facilities.

    SciTech Connect (OSTI)

    Ranek, N. L.; Monette, F. A.

    2001-06-08T23:59:59.000Z

    This report examines the capabilities of existing commercial nuclear fuel fabrication facilities to convert depleted uranium hexafluoride (UF{sub 6}) to uranium oxide (UO{sub 2}). The U.S. Department of Energy (DOE) needs this information to determine whether using such capacity to convert DOE's inventory of depleted UF{sub 6} to a more stable form is a reasonable alternative that should be considered in the site-specific environmental impact statement for construction and operation of depleted UF{sub 6} conversion facilities. Publicly available information sources were consulted to ascertain the information summarized in this report. For domestic facilities, the information summarized includes currently operating capacity to convert depleted UF{sub 6} to UO{sub 2}; transportation distances from depleted UF{sub 6} storage locations near Oak Ridge, Tennessee, Portsmouth, Ohio, and Paducah, Kentucky, to the facilities; and regulatory requirements applicable to nuclear fuel fabrication and transportation of depleted UF{sub 6}. The report concludes that the total currently operating capability of U.S. commercial nuclear fuel fabricators to convert UF{sub 6} to UO{sub 2} is approximately 5,200 metric tons of UF{sub 6} per annum (tUF{sub 6}/a). This total includes 666 tUF{sub 6}/a scheduled for shutdown by the end of 2001. However, only about 300 tUF{sub 6}/a of this capacity could be confirmed as being possibly available to DOE. The report also provides some limited descriptions of the capabilities of foreign fuel fabrication plants to convert UF{sub 6} to uranium oxide forms.

  17. 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

    2012-10-30T23:59:59.000Z

    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

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

    SciTech Connect (OSTI)

    Uberuaga, Blas P. [Los Alamos National Laboratory

    2012-05-02T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Collins, J.L.

    2004-12-02T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1984-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-09-02T23:59:59.000Z

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

  2. 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

    2014-11-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1987-03-01T23:59:59.000Z

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

  4. 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 approach to the synthesis of extended solids that permits them to prepare families of new nanostructured

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

    SciTech Connect (OSTI)

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

    2010-05-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2013-09-23T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Cheong, W.J.; Keech, P.G.; Wren, J.C.; Shoesmith, D.W.; Qin, Z. [The University of Western Ontario, London, Ontario, N6A 5B7 (Canada)

    2007-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1982-08-01T23:59:59.000Z

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

  9. 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)

    2014-07-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Metz, Volker; Bohnert, Elke; Kelm, Manfred; Schild, Dieter; Kienzler, Bernhard [Institute for Radioactive Waste Disposal (FZK-INE), Forschungszentrum Karlsruhe / Research Center Karlsruhe, Helmholtz-Platz, Eggenstein-Leopoldshafen, D-76344 (Germany); Reinhardt, Juergen; Buchmeiser, Michael R. [Leibniz-Institut fuer Oberflaechenmodifizierung, IOM, Permoserstr. 15, Leipzig, D-04318 (Germany)

    2007-07-01T23:59:59.000Z

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

  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

    2014-03-01T23:59:59.000Z

    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

    2007-08-03T23:59:59.000Z

    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. Oxidative corrosion of spent UO{sub 2} fuel in vapor and dripping groundwater at 90{degree}C.

    SciTech Connect (OSTI)

    Finch, R. J.

    1999-04-29T23:59:59.000Z

    Corrosion of spent UO{sub 2} fuel has been studied in experiments conducted for nearly six years. Oxidative dissolution in vapor and dripping groundwater at 90 C occurs via general corrosion at fuel-fragment surfaces. Dissolution along fuel-grain boundaries is also evident in samples contacted by the largest volumes of groundwater, and corroded grain boundaries extend at least 20 or 30 grains deep (> 200 {micro}m), possibly throughout millimeter-sized fragments. Apparent dissolution of fuel along defects that intersect grain boundaries has created dissolution pits that are 50 to 200 nm in diameter. Dissolution pits penetrate 1-2 {micro}m into each grain, producing a ''worm-like'' texture along fuel-grain-boundaries. Sub-micrometer-sized fuel shards are common between fuel grains and may contribute to the reactive surface area of fuel exposed to groundwater. Outer surfaces of reacted fuel fragments develop a fine-grained layer of corrosion products adjacent to the fuel (5-15 {micro}m thick). A more coarsely crystalline layer of corrosion products commonly covers the fine-grained layer, the thickness of which varies considerably among samples (from less than 5 {micro}m to greater than 40 {micro}m). The thickest and most porous corrosion layers develop on fuel fragments exposed to the largest volumes of groundwater. Corrosion-layer compositions depend strongly on water flux, with uranyl oxy-hydroxides predominating in vapor experiments, and alkali and alkaline earth uranyl silicates predominating in high drip-rate experiments. Low drip-rate experiments exhibit a complex assemblage of corrosion products, including phases identified in vapor and high drip-rate experiments.

  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)

    1996-12-30T23:59:59.000Z

    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.

    2003-10-07T23:59:59.000Z

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

    SciTech Connect (OSTI)

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

    2013-10-15T23:59:59.000Z

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

  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. [Department of Nuclear Engineering, KyungHee University, 1732 Deokyoungdaero, Giheung-gu, Yongin, Gyeonggi-do, 446-701 (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    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. Comet whole-core solution to a stylized 3-dimensional pressurized water reactor benchmark problem with UO{sub 2}and MOX fuel

    SciTech Connect (OSTI)

    Zhang, D.; Rahnema, F. [Georgia Inst. of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States)

    2012-07-01T23:59:59.000Z

    A stylized pressurized water reactor (PWR) benchmark problem with UO{sub 2} and MOX fuel was used to test the accuracy and efficiency of the coarse mesh radiation transport (COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000 fuel pins. The COMET code was used to compute the core eigenvalue and assembly and pin power distributions for three core configurations. In these calculations, a set of tensor products of orthogonal polynomials were used to expand the neutron angular phase space distribution on the interfaces between coarse meshes. The COMET calculations were compared with the Monte Carlo code MCNP reference solutions using a recently published an 8-group material cross section library. The comparison showed both the core eigenvalues and assembly and pin power distributions predicated by COMET agree very well with the MCNP reference solution if the orders of the angular flux expansion in the two spatial variables and the polar and azimuth angles on the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in the pin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within 3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMET can achieve accuracy comparable to Monte Carlo. It was also found that COMET's computational speed is 450 times faster than MCNP. (authors)

  20. 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

    2012-04-11T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2009-01-15T23:59:59.000Z

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

  2. ladedqsleasatl uo!lelSluam!radx3

    E-Print Network [OSTI]

    Standiford, Richard B.

    , timber management regime Estimates of timber net value change due to wildfire are sensitive: at the time oflhis study were with the Station's unit studying fire management planning and economics, Riverside.She has been with the Station since 1976. DARlA A. CAIN, research forester, earned a bachelor

  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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for RenewableSpeedingBiomassPPPOPetroleum ReservesThrust AreasDepartment ofFUEL.P8.01

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

    SciTech Connect (OSTI)

    Gerber, M.S.

    1997-11-25T23:59:59.000Z

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

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

    under oxic conditions. Field and laboratory studies have implicated iron sulfide minerals as redox buffers or oxidant scavengers that may slow oxidation of reduced U(VI) solid...

  6. UO Annual Report 2011 Page 1 2011 Annual Report

    E-Print Network [OSTI]

    . CONDITIONS NOT MET Student Performance Criteria SPC 9: Non-Western Traditions progress to ensure that we are meeting 2009 SPC A. 9. Historical Traditions and Global Culture. SPC 13: Human Diversity The visiting team report states

  7. UoS PhD Studentship! A University of Sheffield PhD studentship within the framework of Project

    E-Print Network [OSTI]

    Dixon, Peter

    Sunshine is available, held jointly at the Solar Physics and Space Plasma Research Centre (SP2 RC), School

  8. Lattice anisotropy, electronic and chemical structures of uranyl carbonate, UO2CO3, from

    E-Print Network [OSTI]

    Boyer, Edmond

    localization function mapping, oxygen atoms are found to preferentially bind with uranium for one sublattice Binary and ternary uranium natural and synthetic compounds are known and characterized (mainly X the electron band structure, the equation of state (EOS) as well as the properties of chemi- cal bonding

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

    SciTech Connect (OSTI)

    Gerber, M.S.

    1993-11-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-03-28T23:59:59.000Z

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

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

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1992-01-01T23:59:59.000Z

    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}.

  12. A DualDisk File System: ext4 Mihai Budiu

    E-Print Network [OSTI]

    Budiu, Mihai

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

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

    E-Print Network [OSTI]

    Fu, Chun

    2000-01-01T23:59:59.000Z

    A RELAP5/MOD3.2 model of a VVER-1000/MODEL V320 nuclear power plant, Balakovo Unit 4, was updated, improved and validated on the basis of an input deck prepared by the Kurchatov Institute of Moscow. The RELAP5 model includes both the primary...

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

    Office of Legacy Management (LM)

    Institute Rockefeller Institute'for Medical Research University of Rochester Case School of Applied Science, Ohio State University University of Cincinnati University of...

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

    E-Print Network [OSTI]

    Fu, Chun

    2000-01-01T23:59:59.000Z

    Secondary System Parameters and Calculated Values. . 38 Vl The Sequence of Events during LBLOCA. 46 VII Experimental Design and Calculated Temperatures. . . 57 NOMENCLATURE ATWS BOL CR DBA ECCS FLC HA HP HPIS IAEA I&C Anticipated Transient... postulated accidents such as loss of coolant accident (LOCA) and operational transients such as anticipated transients without scram (ATWS), loss of off-site power, loss of feedwater, loss of flow, and turbine trip. RELAP5 is a one-dimensional code which...

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

    E-Print Network [OSTI]

    Chen, Tianyi

    2013-08-02T23:59:59.000Z

    Our study is focused on the behavior of grain boundaries in uranium dioxide system under irradiation conditions. The research can be seen as two parts: the study of interaction of the grain boundary and the damage cascade, and the calculation...

  17. 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

    2009-08-01T23:59:59.000Z

    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.

  18. 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling7 AugustAFRICAN3uj:'I,\ W C 5% "y$ -- :I_.

  19. 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling CorpNew 1325.8.Enaineer District andP.Dr.Wm.*

  20. 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

  1. Biosciences Undergraduate Research at Nottingham School of Biosciences, UoN, 2009. 1 Assessing the link between forest composition and soil nutrient

    E-Print Network [OSTI]

    Nottingham, University of

    threat of global climate change, concern is growing about the consequence of large scale biodiversity exudates or changes in growth, increasing litter input or nutrient uptake. Nutrient cycling and the plant and therefore more efficient nutrient cycling. The Kamchatka peninsula in Far East Russia makes up the northern

  2. 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.

    2005-05-27T23:59:59.000Z

    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.

  3. Hello "Rhythms and Rhymes" FIG student! My name is Brandon Parry and I will be your FIG Assistant this fall at the UO. When fall

    E-Print Network [OSTI]

    Oregon, University of

    , riding my bike and hanging out with my friends. Naturally, being the FA for "Rhythms and Rhymes", I have, and created, in order for people to assert their individual and group identities. The words we use

  4. (front end fuel cycle) 2.1 (CANDU

    E-Print Network [OSTI]

    Hong, Deog Ki

    , , Phosphorus . 2.2. (U3O8) . U235 . UO2 ( ) UF6 ( ) . 2235 U235 UF6 . 2.2.2. UF6 UO2 UF6 UO2 UF4 UF4 UF6 . (1) : (4) : UF4 UF6 . UF4 1600 500 . UF4 UF4 UO2 . UF6 , , 150

  5. 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

    2010-01-01T23:59:59.000Z

    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

  6. The Influence of the Linker Geometry in Bis(3-hydroxy-N-methyl-pyridin-2-one) Ligands on Solution-Phase Uranyl Affinity

    E-Print Network [OSTI]

    Szigethy, Géza

    2011-01-01T23:59:59.000Z

    M. S. Murali, K. L. Nash, Solv. Extr. Ion Exch. 2001, 19,dimers of the form [UO 2 L 2 (solv. )] 2 as opposed to thesterically-induced [UO 2 (L)(solv. )] 2 dimer formation, [

  7. Ligand field effects on the multiplet structure of the U4f XPS...

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

    Ligand field effects on the multiplet structure of the U4f XPS of UO2. Ligand field effects on the multiplet structure of the U4f XPS of UO2. Abstract: Ab initio, fully...

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

    E-Print Network [OSTI]

    Ni, Chengbao

    2012-01-01T23:59:59.000Z

    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

  9. Winter 2012 Prof. Carol Silverman Office: 321 Condon Off. hrs. M 4-5, W 12-1:00 PM

    E-Print Network [OSTI]

    in the UO Bookstore are: Sacco, Joe. Safe Area Gorazde: The War in Eastern Bosnia 1992-95. Fantagraphics

  10. 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

  11. 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]

  12. The Influence of Linker Geometry on Uranyl Complexation by Rigidly-Linked Bis(3-hydroxy-N-methyl-pyridin-2-one)

    E-Print Network [OSTI]

    Szigethy, Geza

    2011-01-01T23:59:59.000Z

    to the formation of UO 2 L(solv. ) complexes (where L is theUO 2 (bis-Me-3,2-HOPO)(solv. ) (right), tabulated in Tablethe uranyl cation. The U–O solv distances also show little

  13. Page 1 of 2 Risk Management March 2012 Risk Management

    E-Print Network [OSTI]

    Page 1 of 2 Risk Management March 2012 Risk Management Supervisor's Vehicle Incident Report-Employee Volunteer Drivers License No. State of Issue UO Certification Date UO Driver Training Date Vehicle Information: License No. Make Model Year UO Vehicle # Motorpool Outside Rental Personal Vehicle Incident

  14. 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 process and the Brownian CRT. Edward Crane (HIMR, UoB) Forest fires, explosions, and random trees 13th

  15. Investigation of Uranium Polymorphs

    SciTech Connect (OSTI)

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

    2011-08-01T23:59:59.000Z

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

  16. 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)

    2014-11-20T23:59:59.000Z

    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.

  17. 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)

    2014-11-20T23:59:59.000Z

    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.

  18. 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

    2014-11-20T23:59:59.000Z

    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

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

    SciTech Connect (OSTI)

    K. Gofryk; N. Harrison; M. Jaime

    2014-12-01T23:59:59.000Z

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

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

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

    characterize the coordination environments of U incorporated in three Fe-(hydr)oxide minerals: goethite, magnetite, and hematite. The simulations provided information on U-O and...

  1. Crystallographic controls on uranyl binding at the quartz/water...

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

    controls on uranyl binding at the quartzwater interface. Abstract: Molecular dynamics methods were used to simulate UO2(OH)20 binding to pairs of oxo sites on three...

  2. A Coordinate Gradient Descent Method for Linearly Constrained ...

    E-Print Network [OSTI]

    2 (1992),X#" -5s . 29%$ ? uo, Z .-Q. and Tseng, P., Error bounds and convergence analysis of feasible descent methods: a general approach, Ann. Oper. Res.

  3. anhydrous li uranyl: Topics by E-print Network

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

    crystal structure and thermal behavior are reported herein. Experimental Synthesis Uranyl nitrate (UO2(NO31 Hydrothermal synthesis, structure and thermal stability of...

  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.

    1994-09-01T23:59:59.000Z

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

  5. 4th Annual DOE-ERSP PI Meeting: Abstracts

    E-Print Network [OSTI]

    Hazen, Terry C.

    2009-01-01T23:59:59.000Z

    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,

  6. RADIOACTIVE ELEMENT REMOVAL FROM WATER USING GRAPHENE OXIDE (GO) 

    E-Print Network [OSTI]

    Concklin, Joshua Paul

    2013-12-19T23:59:59.000Z

    the UO2 and create a water soluble salt of uranyl nitrate (UO2?(NO3)2). This salt will then be dissolved in deionized water to produce a 1ppm solution of uranium. Once the sample solution has been produced, it should be stored in a glass container....84E+04 counts per second ? Total activity expected from sample = 9.57E+04 counts per second A solution containing uranium was created by treating depleted reactor fuel (UO2) with Nitric Acid (HNO3) to create uranyl nitrate (UO2·(NO3)2) which...

  7. Short Curriculum Vitae of Kyriacos Petratos

    E-Print Network [OSTI]

    Petratos, Kyriacos

    of Crete (UoC). The set-up comprises an X-ray rotating anode generator (RU-3HR, Rigaku), a crystal cryo-cooling system (X-stream, Rigaku) and a system of `imaging-plate' photon detector (Mar300, Mar Department, UoC. Participation in committees for the supervision of graduate students' progress

  8. SUSS-EX CLUB STEERING GROUP The 21st meeting will be held at 17.00 on Friday 11th November 2011

    E-Print Network [OSTI]

    Sussex, University of

    /6/11) [b] UoS 50th birthday (10/9/11) `Sussex through the ages' ~Bob Benewick, Jennifer Platt & David Smith (late summer & December /January) ~ Jennifer Platt [e] Christmas Party (December) ~ Sue Bullock [g 2010 Jennifer Platt July 2010 Ken Wheeler September 2010 David Smith November 2010 Charles Goldie

  9. 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. [National Research Center Kurchatov Institute (Russian Federation); Stogov, Yu. V., E-mail: YVStogov@mephi.ru [National Research Nuclear University MEPhI (Russian Federation)

    2014-12-15T23:59:59.000Z

    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.

  10. s.haszeldine@ed.ac.uk Scottish Centre for Carbon Storage, Petrobras 2008 1 University of Edinburgh (est 1583)

    E-Print Network [OSTI]

    Haszeldine, Stuart

    , and HWU Chem Eng Legislation and regulation UoE Europa Institute, Dundee CEPMLP Energy systems & Social, environmental impacts (HWU, BGS UoE) Licensing and regulation Enhanced hydrocarbon recovery (HWU #12;s for Carbon Storage, Petrobras 2008 9 Sci-tech. background Geoscience, Power plant Capture, Pipeline Legal

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

    SciTech Connect (OSTI)

    K. Gofryk; V. Zapf; M. Jaime

    2014-12-01T23:59:59.000Z

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

  12. JOURNAL OF GEOPHYSICAL RESEARCH, VOL. 97, NO. A9, PAGES 13,911-13,914, SEPTEMBER 1, 1992 Comment on "Dayside Pickup Oxygen Ion Precipitation at Venus and Mars'

    E-Print Network [OSTI]

    Johnson, Robert E.

    on "Dayside Pickup Oxygen Ion Precipitation at Venus and Mars' Spatial Distributions, Energy Deposition; McGrath and Johnson, 1987]and by locally generated"pickup ions" [e.g., Kozyra et al., 1982; Ishimoto a hemisphericallyaveragedyield for ejection of oxygen atoms (O) from an oxygen exosphere, Y0= [rr(T> Uo)+ (6/·2)(otSn)/Uo]/o'd (1

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

    SciTech Connect (OSTI)

    Carnahan, C.L.

    1987-12-01T23:59:59.000Z

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

  14. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1990-01-01T23:59:59.000Z

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

  15. Extreme Performance Scalable Operating Systems Final Progress Report (July 1, 2008 ���¢�������� October 31, 2011)

    SciTech Connect (OSTI)

    Allen D. Malony; Sameer Shende

    2011-10-31T23:59:59.000Z

    This is the final progress report for the FastOS (Phase 2) (FastOS-2) project with Argonne National Laboratory and the University of Oregon (UO). The project started at UO on July 1, 2008 and ran until April 30, 2010, at which time a six-month no-cost extension began. The FastOS-2 work at UO delivered excellent results in all research work areas: * scalable parallel monitoring * kernel-level performance measurement * parallel I/0 system measurement * large-scale and hybrid application performance measurement * onlne scalable performance data reduction and analysis * binary instrumentation

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

    SciTech Connect (OSTI)

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

    2009-01-05T23:59:59.000Z

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

  17. Physical Separation and Multiphase

    E-Print Network [OSTI]

    Sóbester, András

    - research into CVD and HVOF coatings for subsea choke valve applications. s US Navy - understanding the processes of charge generation in gear contacts as a predictive maintenance tool. s DRA/UoS - corrosion

  18. workshop.report.indd

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

    3( 66*:4-01,(56(-&-(6&1+-?-'+:()1::'+)03( Figure 7. (Left) Phase fi eld calculation using empirical free energy model (after SY Hu et al) & (Right) Irradiated UO2...

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

    E-Print Network [OSTI]

    Boscher, Thomas

    2005-01-01T23:59:59.000Z

    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 ...

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

    SciTech Connect (OSTI)

    Broczkowski, Michael E.; Noel, Jamie J.; Shoesmith, David W. [Chemistry, The University of Western Ontario, 1151 Richmond St., London, N6A 5B7 (Canada)

    2007-07-01T23:59:59.000Z

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

  1. 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,

    2012-01-01T23:59:59.000Z

    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

  2. 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

    2012-01-01T23:59:59.000Z

    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

  3. Microsoft PowerPoint - MOX Adventure_Reactor Subcommittee_Tamara...

    National Nuclear Security Administration (NNSA)

    MOX Fuel at Duke Energy MOX Fuel and NMMSS Page 3 MOX Fuel - General MOX fuel pellets from former weapons plutonium Blend of 5% PuO 2 with 95% depleted UO 2 Like...

  4. 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

  5. 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

    2013-08-14T23:59:59.000Z

    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.

  6. anti-ganglioside gd2 antibodies: Topics by E-print Network

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

    Dillon; Barry; M. Gray 3 Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets CERN Preprints Summary: A transversal mapping of the Gd concentration was measured in...

  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

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

    E-Print Network [OSTI]

    Mertens, Paul Gustaaf

    1971-01-01T23:59:59.000Z

    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 ...

  9. Changes in U(VI) speciation upon sorption onto montmorillonite from aqueous and organic solutions

    SciTech Connect (OSTI)

    Chisholm-Brause, C.; Morris, D.E.; Eller, P.G.; Buscher, T.; Conradson, S.D.

    1991-01-01T23:59:59.000Z

    The speciation of UO{sub 2}{sup 2+} and UO{sub 2}{sup 2+} Tributylphosphate (TBP) mixtures has been investigated in solution and intercalated with the reference smectite clay SAz-1 using x-ray absorption, Raman, and luminescence spectroscopies. Neither aquated UO{sub 2}{sup 2+} nor its TBP complex undergoes any detectable changes in uranium oxidation state on intercalation. Further, at the pH values employed in this work, there is no evidence for hydrolysis of the uranium species to generate dimeric or higher order uranium oligomers. However, we do find indications that the structures of the solution complexes are altered on intercalation, particularly for the UO{sub 2}{sup 2+}/TBP system. In addition, several lines of evidence suggest that, at the loading levels used in this study, the uranyl species may be interacting with two or more spectroscopically distinguishable sites on SAz-1. 29 refs., 3 figs., 2 tabs.

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

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

    pRad User Program pRad-uo@lanl.gov P-25 Subatomic Physics P-Division LANSCE pRad logo Los Alamos National Laboratory has used high energy protons as a probe in flash...

  11. Laboratory Experiments and their Applicability 

    E-Print Network [OSTI]

    Steinhaus, Thomas; Jahn, Wolfram

    2007-11-14T23:59:59.000Z

    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 ...

  12. MFR PAPER 1031 Traw ls and traps capture

    E-Print Network [OSTI]

    a carapace II idlh of 7 inches and a \\\\eighl of more than 2.5 pounds (WillieI'. 1966), A cammer· ci al prl,\\uo, of ,\\ ,hart ,upph prl)UULCr "I red .:rab

  13. Annual Report 2008 -- Office of the Chief Financial Officer (OCFO)

    E-Print Network [OSTI]

    Fernandez, Jeffrey

    2009-01-01T23:59:59.000Z

    C R A D A - Small Business CRADA • Othei Total CooperativeInformation ($K) uo.ooo DOEM&O CRADA WFO Program (WN| BT FY2006 FY2007 T DOEM&O CRADA WFO Program (WN| Universities

  14. Department of architecturein portlanD

    E-Print Network [OSTI]

    recognized for educating architects who understand and practice sustainable design. The UO architecture program is rated in the top three for sustainable design education based on surveys of U.S. architectural . . . . . . . . . . . . . 24 For further information . . . . . . . . 26 Arch architecture

  15. 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.

  16. 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...

  17. Identification and Characterization of UndA-HRCR-6, an Outer...

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

    MR-1 mutant’s ability to reduce solid phase ferrihydrite at 40% of that for MR-1 wild type, (ii) increased extracellular formation of UO2 associated with the outer membrane...

  18. c-Type Cytochrome-Dependent Formation of U(IV) Nanoparticles...

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

    of mtrC andor omcA significantly affected the in vivo U(VI) reduction rate relative to wild type MR-1. Similar to the wild type, the mutants accumulated UO2 nanoparticles...

  19. 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

    2014-03-05T23:59:59.000Z

    As fission process heats up the fuel rods, UO2 pellets stacked on top of each other swell both radially and axially, while the surrounding Zircaloy cladding creeps down, so that the pellets eventually come into contact with the clad...

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

    E-Print Network [OSTI]

    Svane, Axel Torstein

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

  1. 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

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

    E-Print Network [OSTI]

    Hazen, Terry C.

    2010-01-01T23:59:59.000Z

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

  3. Thermodynamics of Uranyl Minerals: Enthalpies of Formation of Uranyl Oxide Hydrates

    SciTech Connect (OSTI)

    K. Kubatko; K. Helean; A. Navrotsky; P.C. Burns

    2005-05-11T23:59:59.000Z

    The enthalpies of formation of seven uranyl oxide hydrate phases and one uranate have been determined using high-temperature oxide melt solution calorimetry: [(UO{sub 2}){sub 4}O(OH){sub 6}](H{sub 2}O){sub 5}, metaschoepite; {beta}-UO{sub 2}(OH){sub 2}; CaUO{sub 4}; Ca(UO{sub 2}){sub 6}O{sub 4}(OH){sub 6}(H{sub 2}O){sub 8}, becquerelite; Ca(UO{sub 2}){sub 4}O{sub 3}(OH){sub 4}(H{sub 2}O){sub 2}; Na(UO{sub 2})O(OH), clarkeite; Na{sub 2}(UO{sub 2}){sub 6}O{sub 4}(OH){sub 6}(H{sub 2}O){sub 7}, the sodium analogue of compreignacite and Pb{sub 3}(UO{sub 2}){sub 8}O{sub 8}(OH){sub 6}(H{sub 2}O){sub 2}, curite. The enthalpy of formation from the binary oxides, {Delta}H{sub f-ox}, at 298 K was calculated for each compound from the respective drop solution enthalpy, {Delta}H{sub ds}. The standard enthalpies of formation from the elements, {Delta}H{sub f}{sup o}, at 298 K are -1791.0 {+-} 3.2, -1536.2 {+-} 2.8, -2002.0 {+-} 3.2, -11389.2 {+-} 13.5, -6653.1 {+-} 13.8, -1724.7 {+-} 5.1, -10936.4 {+-} 14.5 and -13163.2 {+-} 34.4 kJ mol{sup -1}, respectively. These values are useful in exploring the stability of uranyl oxide hydrates in auxiliary chemical systems, such as those expected in U-contaminated environments.

  4. Ab Initio Molecular Dynamics Study of the Reaction between Th+ Jia Zhou and H. Bernhard Schlegel*

    E-Print Network [OSTI]

    Schlegel, H. Bernhard

    explored by Jackson et al.15-17 in a study of the reactivity of Th+ , U+ , ThO+ , UO+ , and UO2 in providing additional information. Uranium is one of the most well-studied actinides, and in particular) to examine O-H and N-O bond activation by uranium ions.18-20 In addition to uranium and its oxides, thorium

  5. Production of cerium oxide microsheres by an internal gelation sol-gel process

    E-Print Network [OSTI]

    Wegener, Jeffrey J.

    2010-01-14T23:59:59.000Z

    Initiative DSC Differential Scanning Calorimetry DOE Department of Energy HMTA Hexamethylenetetramine MOX Mixed Oxide NERI Nuclear Energy Research Initiative ORNL Oak Ridge National Laboratory PUREX Plutonium and Uranium Extraction RTV Room... Page Figure 1 Dried UO2 spheres (~1000 ?m diameter, left) and sintered UO2 spheres (500 to 532 ?m diameter, right) ........................................... 11 Figure 2 Diagram of Oak Ridge internal gelation sol-gel system...

  6. Fundamental study on recovery uranium oxide from HEPA filters

    SciTech Connect (OSTI)

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

    1993-12-31T23:59:59.000Z

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

  7. Production of cerium oxide microsheres by an internal gelation sol-gel process 

    E-Print Network [OSTI]

    Wegener, Jeffrey J.

    2010-01-14T23:59:59.000Z

    Initiative DSC Differential Scanning Calorimetry DOE Department of Energy HMTA Hexamethylenetetramine MOX Mixed Oxide NERI Nuclear Energy Research Initiative ORNL Oak Ridge National Laboratory PUREX Plutonium and Uranium Extraction RTV Room... Page Figure 1 Dried UO2 spheres (~1000 ?m diameter, left) and sintered UO2 spheres (500 to 532 ?m diameter, right) ........................................... 11 Figure 2 Diagram of Oak Ridge internal gelation sol-gel system...

  8. Infrared Spectroscopy of Discrete Uranyl Anion Complexes

    SciTech Connect (OSTI)

    Gary S. Groenewold; Anita K. Gianotto; Michael E. McIlwain; Michael J. Van Stipdonk; Michael Kullman; Travis J. Cooper; David T. Moore; Nick Polfer; Jos Oomens; Ivan Infante; Lucas Visscher; Bertrand Siboulet; Wibe A. de Jong

    2007-12-01T23:59:59.000Z

    The Free-Electron Laser for Infrared Experiments, FELIX, was used to study the wavelength-resolved multiphoton dissociation of discrete, gas phase uranyl (UO22+) complexes containing a single anionic ligand (A), with or without ligated solvent molecules (S). The apparent uranyl antisymmetric and symmetric stretching frequencies were measured for complexes with general formula [UO2A(S)n]+, where A was either hydroxide, methoxide or acetate, S was water, ammonia, acetone or acetonitrile, and n = 0-2. The values for the antisymmetric stretching frequency for uranyl ligated with only an anion ([UO2A]+) were as low or lower than measurements for [UO2]2+ ligated with as many as five strong neutral donor ligands, and are comparable to solution phase values. This result was surprising because initial DFT calculations using B3LYP predicted values that were 30 – 40 cm-1 higher, consistent with intuition but not with the data. Modification of the basis set and use of alternative functionals improved computational accuracy for the methoxide and acetate complexes, but calculated values for the hydroxide were greater than the measurement regardless of the computational method used. Attachment of a neutral donor ligand S to [UO2A]+ produced [UO2AS]+, which resulted only very modest changes to the uranyl frequency, and did not universally shift values lower. DFT calculations for [UO2AS]+ were in accord with trends in the data, and showed that attachment of the solvent was accommodated by weakening of the U-anion bond as well as the uranyl. When uranyl frequencies were compared for [UO2AS]+ species having different solvent neutrals, values decreased with increasing neutral nucleophilicity.

  9. Infared Spectroscopy of Discrete Uranyl Anion Complexes

    SciTech Connect (OSTI)

    Groenewold, G. S.; Gianotto, Anita K.; McIIwain, Michael E.; Van Stipdonk, Michael J.; Kullman, Michael; Moore, David T.; Polfer, Nick; Oomens, Jos; Infante, Ivan A.; Visscher, Lucas; Siboulet, Bertrand; De Jong, Wibe A.

    2008-01-24T23:59:59.000Z

    The Free-Electron Laser for Infrared Experiments (FELIX) w 1 as used to study the wavelength-resolved multiple photon photodissociation of discrete, gas phase uranyl (UO2 2 2+) complexes containing a single anionic ligand (A), with or without ligated solvent molecules (S). The uranyl antisymmetric and symmetric stretching frequencies were measured for complexes with general formula [UO2A(S)n]+, where A was either hydroxide, methoxide, or acetate; S was water, ammonia, acetone, or acetonitrile; and n = 0-3. The values for the antisymmetric stretching frequency for uranyl ligated with only an anion ([UO2A]+) were as low or lower than measurements for [UO2]2+ ligated with as many as five strong neutral donor ligands, and are comparable to solution phase values. This result was surprising because initial DFT calculations predicted values that were 30–40 cm-1 higher, consistent with intuition but not with the data. Modification of the basis sets and use of alternative functionals improved computational accuracy for the methoxide and acetate complexes, but calculated values for the hydroxide were greater than the measurement regardless of the computational method used. Attachment of a neutral donor ligand S to [UO2A]+ produced [UO2AS]+, which produced only very modest changes to the uranyl antisymmetric stretch frequency, and did not universally shift the frequency to lower values. DFT calculations for [UO2AS]+ were in accord with trends in the data, and showed that attachment of the solvent was accommodated by weakening of the U-anion bond as well as the uranyl. When uranyl frequencies were compared for [UO2AS]+ species having different solvent neutrals, values decreased with increasing neutral nucleophilicity.

  10. Catalysts for the hydrolysis of thiophosphate triesters 

    E-Print Network [OSTI]

    Picot, Alexandre

    2005-02-17T23:59:59.000Z

    ............................................................ Coordination reaction of 66 with uranyl nitrate...................................... NMR spectra of 66 in the presence of increasing amount of UO 2 2+ ....... Coordination chemistry of UO 2 (II... decontaminants during and after World War II. However, the active chlorines would rapidly decompose upon storage and large excesses of bleach were needed for the oxidation of the agents. As these solutions were also corrosive to most surfaces and ineffective...

  11. Free energies and mechanisms of water exchange around Uranyl from first principles molecular dynamics

    SciTech Connect (OSTI)

    Atta-Fynn, Raymond; Bylaska, Eric J.; De Jong, Wibe A.

    2012-02-01T23:59:59.000Z

    From density functional theory (DFT) based ab initio (Car-Parrinello) metadynamics, we compute the activation energies and mechanisms of water exchange between the first and second hydration shells of aqueous Uranyl (UO{sub 2}{sup 2+}) using the primary hydration number of U as the reaction coordinate. The free energy and activation barrier of the water dissociation reaction [UO{sub 2}(OH{sub 2}){sub 5}]{sup 2+}(aq) {yields} [UO{sub 2}(OH{sub 2})4]{sup 2+}(aq) + H{sub 2}O are 0.7 kcal and 4.7 kcal/mol respectively. The free energy is in good agreement with previous theoretical (-2.7 to +1.2 kcal/mol) and experimental (0.5 to 2.2 kcal/mol) data. The associative reaction [UO{sub 2}(OH{sub 2}){sub 5}]{sup 2+}(aq) + H{sub 2}O {yields} [UO{sub 2}(OH{sub 2})6]{sup 2+}(aq) is short-lived with a free energy and activation barrier of +7.9 kcal/mol and +8.9 kca/mol respectively; it is therefore classified as associative-interchange. On the basis of the free energy differences and activation barriers, we predict that the dominant exchange mechanism between [UO{sub 2}(OH{sub 2}){sub 5}]{sup 2+}(aq) and bulk water is dissociative.

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

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth

    2002-09-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Todd Allen

    2014-04-01T23:59:59.000Z

    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

  14. Surface Decontamination of System Components in Uranium Conversion Plant at KAERI

    SciTech Connect (OSTI)

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

    2003-02-25T23:59:59.000Z

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

  15. EFRC CMSNF Major Accomplishments

    SciTech Connect (OSTI)

    D. Hurley; Todd R. Allen

    2014-09-01T23:59:59.000Z

    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.

  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

    2003-01-01T23:59:59.000Z

    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. Influence of zirconium and niobium on cathodic deposition of uranium dioxide from alkali chloride melts

    SciTech Connect (OSTI)

    Komarov, V.E.; Borodina, N.P.; Martem`yanova, Z.S. [Institute of High-Temperature Electrochemistry, Ekaterinburg (Russian Federation)

    1995-07-01T23:59:59.000Z

    Electrocrystallization of uranium dioxide from molten chloride electrolytes in the presence of zirconium(IV) and niobium(V) was studied by voltammetry. Zirconium(IV) was found to react with uranium dioxide according to exchange mechanism to form (1 - x)UO{sub 2}{center_dot}xZrO{sub 2} solid solutions. Niobium(IV), a product of cathodic reduction of niobium(V), enters into the exchange reaction with uranium dioxide to yield (1 - y)UO{sub 2{center_dot}y}NbO{sub 2} solid solutions. In the case of simultaneous presence of Nb(V) and Zr(IV) in electrolyte, ternary (1 - x - y)UO{sub 2 {center_dot}x}ZrO{sub 2{center_dot}y}NbO{sub 2} solid solutions are obtained at the cathode surface. Nucleation of the solid solutions phase was shown to occurs at the most active sites of the crystalline precipitate of uranium dioxide.

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

    SciTech Connect (OSTI)

    Tsai, C.

    1981-11-01T23:59:59.000Z

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

  19. Effects of air oxidation on the dissolution rate of LWR spent fuel

    SciTech Connect (OSTI)

    Gray, W.J.; Thomas, L.E.; Einziger, R.E.

    1992-11-01T23:59:59.000Z

    Dissolution rates for air-oxidized spent fuel were measured in flowthrough tests. Results from two types of specimens, separated grains and multigrain particles, both in oxidized (U[sub 4]O[sub 9+x]) and unoxidized (UO[sub 2]) conditions indicated only minor effects of oxidation on the surface-area-normalized rates. Similar results were obtained for unirradiated specimens in three different oxidation states (UO[sub 2], U[sub 3]O[sub 7], and U[sub 3]O[sub 8]). These observations have important practical implications for disposal of spent fuel in a geologic repository as well as implications regarding the oxidative dissolution mechanism of UO[sub 2] fuel.

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

    SciTech Connect (OSTI)

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

    2014-01-01T23:59:59.000Z

    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.

  1. The effect of geometry on symbology recognition

    E-Print Network [OSTI]

    Boyless, James Andrus

    1979-01-01T23:59:59.000Z

    comparisions of these sub-divisions and their advantages and disadvantages' ~dl C ' . H 'df 1d (19649 9 6 d researchers have performed studies using color (Erikson, 1952; Cohen and Senders, 1952; and Muller, 1955) ~ Among these researchers a cons nsus...AaTA fiue 1e Tte1ap 1sa[[errrs aq1 uJaosrp o1 1o[rd aq1 aJTnbaJ uot1eurJogut go sadfi1 q1oH srUa1sfis fieydstp pue s1uaurnJ1sut 1geJoJre aq1 rrroJQ pa~taoaJ st uo rlerUJogut 1oaJTpuZ '1geJoJ&e aq1 go 1uarUuoJznua TeuJa1xa aq1 rUoJg pawTaoaJ st uoT, 1errr...

  2. New Catalytic DNA Biosensors for Radionuclides and Metal ions

    SciTech Connect (OSTI)

    Lu, Yi

    2005-06-01T23:59:59.000Z

    In vitro selection for DNAzymes that are catalytically active with UO22+ ions as the metal cofactor has been completed. The 10th generation pool of DNA was cloned and sequenced. A total of 84 clones were sequenced and placed into families based on sequence alignments. Selected members of each family were 5-labeled with 32P and amplified using PCR. Activity assays were conducted using the isotopically labeled DNAzymes in order to determine which sequences were the most active. The secondary structures of the two most active sequences, called Clone 13 and Clone 39, were determined using the computer program Mfold. A cleavage rate of approximately 1 min-1 in the presence of 10 uM UO22+ was observed for both clones. Clone 39 was determined to be the best candidate for truncation to create a trans-cleaving DNAzyme, based on its secondary structure. An enzyme strand, called 39E, and a substrate strand, called 39DS, were designed by truncating the cis-cleaving DNAzyme. An alternative enzyme strand, called 39Ec, was also assayed with the 39DS substrate. This strand was designed so that the two binding arms were perfectly complimentary, unlike 39E, which formed three mismatched base pairs with 39DS. Both 39E and 39Ec were found to be active, with a rate of approximately 1 min-1 in the presence of 10 uM UO22+. A preliminary UO22+ binding curve was obtained for the 39Ec/39DS trans-cleaving system. The enzyme is active with UO22+ concentrations as low as 1 nM. Based on the preliminary binding curve data, the apparent UO22+ binding constant is approximately 330 nM, and kmax is approximately 1 min-1.

  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)

    1996-12-30T23:59:59.000Z

    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. Greening academia: Developing sustainable waste management at Higher Education Institutions

    SciTech Connect (OSTI)

    Zhang, N. [School of Civil Engineering and the Environment, University of Southampton, University Rd., Highfield, Southampton, Hampshire SO17 1BJ (United Kingdom); Williams, I.D., E-mail: idw@soton.ac.uk [School of Civil Engineering and the Environment, University of Southampton, University Rd., Highfield, Southampton, Hampshire SO17 1BJ (United Kingdom); Kemp, S. [School of Civil Engineering and the Environment, University of Southampton, University Rd., Highfield, Southampton, Hampshire SO17 1BJ (United Kingdom); Smith, N.F. [Estates and Facilities Management, University of Southampton, University Rd., Highfield, Southampton, Hampshire SO17 1BJ (United Kingdom)

    2011-07-15T23:59:59.000Z

    Higher Education Institutions (HEIs) are often the size of small municipalities. Worldwide, the higher education (HE) sector has expanded phenomenally; for example, since the 1960s, the United Kingdom (UK) HE system has expanded sixfold to >2.4 million students. As a consequence, the overall production of waste at HEIs throughout the world is very large and presents significant challenges as the associated legislative, economic and environmental pressures can be difficult to control and manage. This paper critically reviews why sustainable waste management has become a key issue for the worldwide HE sector to address and describes some of the benefits, barriers, practical and logistical problems. As a practical illustration of some of the issues and problems, the four-phase waste management strategy developed over 15 years by one of the largest universities in Southern England - the University of Southampton (UoS) - is outlined as a case study. The UoS is committed to protecting the environment by developing practices that are safe, sustainable and environmentally friendly and has developed a practical, staged approach to manage waste in an increasingly sustainable fashion. At each stage, the approach taken to the development of infrastructure (I), service provision (S) and behavior change (B) is explained, taking into account the Political, Economic, Social, Technological, Legal and Environmental (PESTLE) factors. Signposts to lessons learned, good practice and useful resources that other institutions - both nationally and internationally - can access are provided. As a result of the strategy developed at the UoS, from 2004 to 2008 waste costs fell by around Pounds 125k and a recycling rate of 72% was achieved. The holistic approach taken - recognizing the PESTLE factors and the importance of a concerted ISB approach - provides a realistic, successful and practical example for other institutions wishing to effectively and sustainably manage their waste.

  5. DECAY HEAT CONDITIONS OF CURRENT AND NEXT GENERATION REACTORS

    E-Print Network [OSTI]

    Choe, JongSoo 1985-

    2012-05-04T23:59:59.000Z

    expects to operate the VHTR by 2021 (NGNP, A Report to Congress, 2008). Advanced Burner 3 Reactors (ABR) are Sodium-Cooled Fast Reactors (SFR) which are fast neutron spectrum and closed fuel cycle system reactors. Its management of actinides... enriched uranium dioxide(UO2 ) less than 5 wt% and gadolinia-uranium dioxide(Gd,UO2). The cladding material is ZIRLO which is a zirconium based alloy for improved corrosion resistance (US-APWR, 2011). The operation power of the ABWR is 3926 MWth. Its...

  6. Congrs "Matriaux 2006", Colloque "Matrise des microstructures des matriaux", 13-17 nov. 2006, Dijon. Actes dits sur DVD, ISBN 978-2-9528-1400-3.

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    France par une conversion en voie sèche d'UF6 gazeux. Le procédé comporte deux étapes : hydrolyse en UO2F combustible nucléaire peut être produite par le procédé de conversion en voie sèche d'UF6 gazeux. Ce procédé réalise successivement une hydrolyse de l'UF6 gazeux en poudre de difluorure d'uranyle UO2F2, puis une

  7. , 227 240 . Half-life Radioactive decay Reaction with 2200 m/s neutrons

    E-Print Network [OSTI]

    Hong, Deog Ki

    . 700-800 U2N3 UN2 . 1.2.5. 1.5 . , . UF4 UF6 UO2 HF . UO3 + 4HF Boils at 1 atm UF6 Black ~1427 8.95 UF4 Green 1036 1457 6.70 U4F17 Black 430 Disp. 6.94 U2F9 390 Disp. 7.06 UF5 White 348 Disp. 6.45 UF6 Colorless 64.05 56.54 5.06 UCl3 Olive green 837 1657 5.51 UCl4 Dark

  8. Preparation of thorium-uranium gel spheres

    SciTech Connect (OSTI)

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

    1980-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-01-04T23:59:59.000Z

    Sludge disposition will be managed in two phases under the K Basin Sludge Treatment Project. The first phase is to retrieve the sludge that currently resides in engineered containers in the K West (KW) Basin pool at ~10 to 18°C. The second phase is to retrieve the sludge from interim storage in the sludge transport and storage containers (STSCs) and treat and package it in preparation for eventual shipment to the Waste Isolation Pilot Plant. The work described in this report was conducted to gain insight into how sludge may change during long-term containerized storage in the STSCs. To accelerate potential physical and chemical changes, the tests were performed at temperatures and oxygen partial pressures significantly greater than those expected in the T Plant canyon cells where the STSCs will be stored. Tests were conducted to determine the effects of 50°C oxygenated water exposure on settled quiescent uraninite (UO2) slurry and a full simulant of KW containerized sludge to determine the effects of oxygen and heat on the composition and mechanical properties of sludge. Shear-strength measurements by vane rheometry also were conducted for UO2 slurry, mixtures of UO2 and metaschoepite (UO3•2H2O), and for simulated KW containerized sludge. The results from these tests and related previous tests are compared to determine whether the settled solids in the K Basin sludge materials change in volume because of oxidation of UO2 by dissolved atmospheric oxygen to form metaschoepite. The test results also are compared to determine if heating or other factors alter sludge volumes and to determine the effects of sludge composition and settling times on sludge shear strength. It has been estimated that the sludge volume will increase with time because of a uranium metal ? uraninite ? metaschoepite oxidation sequence. This increase could increase the number of containers required for storage and increase overall costs of sludge management activities. However, the volume might decrease because of decreases in the water-volume fraction caused by sludge solid reactions, compaction, or intergrowth and recrystallization of metaschoepite. In that case, fewer STSCs may be needed, but the shear strength would increase, and this could challenge recovery by water jet erosion and require more aggressive retrieval methods. Overall, the tests described herein indicate that the settled solids volume remains the same or decreases with time. The only case for which the sludge solids volumes increase with time is for the expansion factor attendant upon the anoxic corrosion of uranium metal to produce UO2 and subsequent reaction with oxygen to form equimolar UO2.25 and UO3•2H2O.

  10. Capacitance-based prover for gas flow meters

    E-Print Network [OSTI]

    Pipkins, Sean Patrick

    1995-01-01T23:59:59.000Z

    ;?= enthalpy of the air at the inlet. H Uo =U +RT Uo =Cv(T To)+RT, (36) 24 Equation (35) can be written as c (T-T)=[c (T?? T)+RT. , ?] Il ? ' n j (37) Solving Equation (37) for T gives the equation used to calculate the temperature profiles of the air.... . . . . Relation of Capacitance Change to Molar Flow Rate. DiIferential Method Integral Method THEORETICAL CALCULATIONS. Calculated C. Experimental C. Thermal Expansion Effects . . Capacitance Change Due to Length Change Capacitance Change Due to Radius...

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

    E-Print Network [OSTI]

    Apps, J.A.

    2010-01-01T23:59:59.000Z

    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

  12. Nagoya University Description

    E-Print Network [OSTI]

    Bristol, University of

    is a friendly, bustling industrial port city, and is the site of Toyota HQ (whose plant you can visit to watch their amazing robot workforce). Don't miss the festivals at Atsuta-jing, one of the most important Shinto for a JASSO Scholarship. Apply Once you have completed the UoB online application and been allocated a space

  13. Structural similarities between biogenic uraninites produced by phylogenetically and metabolically diverse bacteria.

    SciTech Connect (OSTI)

    Sharp, Jonathan; Schofield, Eleanor J.; Veeramani, Harish; Suvorova, Elena; Kennedy, David W.; Marshall, Matthew J.; Mehta, Apurva; Bargar, John R.; Bernier-Latmani, Rizlan

    2009-11-01T23:59:59.000Z

    While the product of microbial uranium reduction is often reported to be“UO2”, a comprehensive characterization including stoichiometry and unit cell determination is available for only one Shewanella species. Here, we compare the products of batch uranyl reduction by a collection of dissimilatory metal- and sulfate-reducing bacteria of the genera Shewanella, Geobacter, Anaeromyxobacter, and Desulfovibrio under similar laboratory conditions. Our results demonstrate that U(VI) bioreduction by this assortment of commonly studied, environmentally relevant bacteria leads to the precipitation of uraninite with a composition between UO2.00 and UO2.075, regardless of phylogenetic or metabolic diversity. Coupled analyses, including electron microscopy, X-ray absorption spectroscopy, and powder diffraction, confirm that structurally and chemically analogous uraninite solids are produced. These biogenic uraninites have particle diameters of about 2-3 nm and lattice constants consistent with UO2.0 and exhibit a high degree of intermediate-range order. Results indicate that phylogenetic and metabolic variability within delta- and gamma-proteobacteria has little effect on nascent biouraninite structure or crystal size under the investigated conditions.

  14. Advanced Fuel Cycle Scenarios with AP1000 PWRs and VHTRs and Fission Spectrum Uncertainties

    E-Print Network [OSTI]

    Cuvelier, Marie-Hermine

    2012-07-16T23:59:59.000Z

    evaluations are indeed intensively used in reactors' calculations. Discrepancies higher than 10% have been computed among nuclear data libraries for energies above 8MeV for 235U. TRU arising from a 3wt% 235U-enriched UO2-fueled AP1000 were incinerated in a...

  15. 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.

    2008-07-08T23:59:59.000Z

    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.

  16. Analysis of a proposed fuel freezing mechanism in a rod bundle 

    E-Print Network [OSTI]

    Nguyen-Wayne, David Loc

    1983-01-01T23:59:59.000Z

    void thermite fuel simulant (80 w/o UO2, 20 w/o Mo at about 3470 K, approximately 330 K superheat) and stainless steel walls. The main system variables in these tests were the initial wall temperatures and the fuel downward injection pressures...

  17. Quantifying submarine groundwater discharge in the coastal zone via multiple methods

    E-Print Network [OSTI]

    South Florida Water Management District, USA i University of Western Australia, Australia j Department, Tallahassee, FL 32306, USA b Isotope Hydrology Section, International Atomic Energy Agency, Austria c, Turkey n Marine Environment Laboratory, International Atomic Energy Agency, Monaco o U.O. 4.17 of the G

  18. HMSC Sustainability Committee Meeting Minutes: January 13, 2009

    E-Print Network [OSTI]

    for Sustainability Committee Oregon Climate Dialog/ National Teach-in Recycle Mania Earth Tub Industrial audit about tax and energy rebates and home energy conservation. Actions: Devin will get a blurb from. Devin concluded that UO and OSU both had excellent websites, and that the University of Pennsylvania

  19. LCLS USERS' ORGANIZATION CHARTER AND BY-LAWS A. Terms and definitions

    E-Print Network [OSTI]

    Wechsler, Risa H.

    LCLS USERS' ORGANIZATION CHARTER AND BY-LAWS A. Terms and definitions Several definitions and acronyms used in this document are defined below LCLS Linac Coherent Light Source LCLS/UO LCLS Users' Organization LCLS/UOEC LCLS Users' Organization Executive Committee SLAC Stanford Linear Accelerator Center

  20. IAEA-TECDOC-1450 Thorium fuel cycle --Potential

    E-Print Network [OSTI]

    Laughlin, Robert B.

    in liquid metal cooled fast breeder reactor (LMFBR) and for neutron flux flattening of the initial core neutron reactor has been recognized. Several experimental and prototype power reactors were successfully operated during the mid 1950s to the mid 1970s using (Th, U)O2 and (Th, U)C2 fuels in high temperature gas

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

    E-Print Network [OSTI]

    Szigethy, Geza

    2010-01-01T23:59:59.000Z

    Murali, M. S. ; Nash, K. L. Solv. Extr. Ion Exch. 2001, 19,D. C. ; Raymond, K. N. Solv. Extr. Ion Exch. 2004, 22, (22)DMF) and UO 2 (bis-Me-3,2-HOPO)(solv) tabulated in Table 2-

  2. Thermal-Hydraulic Analysis of Seed-Blanket Unit Duplex Fuel Assemblies with VIPRE-01

    E-Print Network [OSTI]

    McDermott, Patrick 1987-

    2012-11-15T23:59:59.000Z

    and blanket unit (SBU) configuration, where the seed region contains standard UO2 fuel, and the blanket region contains an inert matrix (Pu,Np,Am)O2-MgO-ZrO2 fuel. The research efforts of this thesis are first to consider the higher burnup effects on DUPLEX...

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

    SciTech Connect (OSTI)

    Liu, Xiang-yand [Los Alamos National Laboratory; Uberuaga, Blas P [Los Alamos National Laboratory; Nerikar, Pankaj [Los Alamos National Laboratory; Sickafus, Kurt E [Los Alamos National Laboratory; Stanek, Chris R [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Long, Yun, 1972-

    2002-01-01T23:59:59.000Z

    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 ...

  5. Latest Neoproterozoic to Mid-Cambrian age for the main deformation phases of the Transantarctic Mountains: new stratigraphic and isotopic constraints from the Pensacola Mountains, Antarctica

    E-Print Network [OSTI]

    Rowell, A. J.; Van Schmus, W. R.; Storey, B. C.; Fetter, A. H.; Evans, K. R.

    2001-03-01T23:59:59.000Z

    and monitored by analysis of NBS SRM-983 radiogenic Pb. Uranium was loaded on the same filament with the Pb and analysed as UO 2 + ; uranium fractionation was monitored using NBS SRM U-500. Uncertainties in U/Pb ratios due to uncertainties in fractionation...

  6. Complexation of the uranyl ion with the aminomethylenediphosphonates MAMDP and AMDP

    SciTech Connect (OSTI)

    Bollinger, J.E.; Roundhill, D.M. [Tulane Univ., New Orleans, LA (United States)

    1994-12-21T23:59:59.000Z

    The use of uranium as a nuclear energy source has made it a valuable mineral resource over the past forty years. Processing uranium generally involves leaching the metal as the uranyl ion (UO{sub 2}){sup 2+} from ore, followed by solvent extraction, precipitation or batch adsorption onto ion exchange-type resins. Uranium reserves exist also in the form of UO{sub 2}(CO{sub 3}){sub 3}{sup 4{minus}} dissolved in seawater, and although these concentrations are very low they represent globally some 4.9 x 10{sup 9} tons. In the interest of tapping this resource efforts toward developing more efficient and selective means of UO{sub 2}{sup 2+} ion sequestering have been underway for some time. The authors have measured the stability and protonation constants of the different complexes formed in aqueous solution between the UO{sub 2}{sup 2+} ion and the compounds N,N{prime}-dimethylaminomethylenebis(phosphonic acid) (MAMDP) and aminomethylenebis(phosphonic acid) (AMDP). From these data one can evaluate the potential for compounds of this type to be useful as uranyl ion sequestering agents.

  7. 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

    1999-01-01T23:59:59.000Z

    The COMETHE 4D Release 23 computer code was used to evaluate the thermal, chemical and mechanical performance of weapons-grade MOX fuel irradiated under U.S. light water reactor typical conditions. Comparisons were made to and UO? fuels exhibited...

  8. 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

  9. 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

  10. 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

  11. Radiative heat transfer in porous uranium dioxide

    SciTech Connect (OSTI)

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

    1992-12-01T23:59:59.000Z

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

  12. For the Meyer Fund for Sustainable Development and the University of Oregon Department of Physics and the Solar Radiation Monitoring Laboratory

    E-Print Network [OSTI]

    Oregon, University of

    and the Solar Radiation Monitoring LaboratoryG:SourcesforBackgroundInformation© Useful Web Sites: UO Solar Radiation Monitoring Laboratory Website: http://solardata.uoregon.edu/Educational Solar Radiation Basics Solar Electric Lesson Plans o What is a KiloWatt Hour? o Experiments

  13. 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

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

    SciTech Connect (OSTI)

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

    2010-09-01T23:59:59.000Z

    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.

  15. NEAMSUpdateQuarterly report for January March 2013 Published May 2013 Nuclear Energy

    E-Print Network [OSTI]

    Kemner, Ken

    of fracture behavior in UO2 were performed (page 2). }} A new model for fuel restructuring by void/pore, and structural mechanics modules were integrated in SHARP to perform high-fidelity, multiphysics simulations (page 3). }} Collaborations with Russia and Euratom continue to assess models of thermal fluid flows

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

    SciTech Connect (OSTI)

    baney, Ronald; Tulenko, James

    2012-11-20T23:59:59.000Z

    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. 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

  18. Nanocrystalline Metals Andy Howe, Corus RD&T

    E-Print Network [OSTI]

    Cambridge, University of

    for single phase polygonal structures · Concentrate on refinement of dual phase structures ­ Dual Phase & with the emphasis on steel! Andy Howe, Corus RD&T Super Bainite Workshop, 6/05/2010 #12;Outline · Ultra-fine Ferrite iron to whatever strength you want! FIB-cut and imaged sub-micron IF steel Corus ­ UoManchester +FEI

  19. Continued on back UNIVERSITY OF OREGON

    E-Print Network [OSTI]

    Oregon, University of

    Continued on back UNIVERSITY OF OREGON MOBILE TECHNOLOGY ACCESS AND PAYMENT OPTION REQUEST Please complete this form to apply for access to mobile technology (e.g., cell phones, smart phones, etc device use. Information on the UO policies regarding access to mobile technology and payment options can

  20. Research Strategy February 2014

    E-Print Network [OSTI]

    Paxton, Anthony T.

    3.1 Priority areas of research Marine Renewable Energy The marine group has an established track in each discipline. 3. The School has a number of strategic international network connectionsA27) and Town & Country Planning (UoA31). The outcome is summarised in Table 1. Table 1 Percentage

  1. An economic analysis of investment in agricultural land

    E-Print Network [OSTI]

    Cherry, Russell Cather

    1966-01-01T23:59:59.000Z

    . DED KCATXOM ~isieia dedioar-Xa teats a mi quixida esposa . sou tuba amabile a@ada y dsfueruo incesagte yo uo pudiexa habet texmioado mi maestxia ~bar X ~ XNTPtODUCTXOM AMD STATIPIEMT OF THE PROBLEM a ~ o 0 o o 0 ~ XXe THE THEORETXCAL ~RK ~ 'a o oo...

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

    SciTech Connect (OSTI)

    K. Gofryk; M. Jaime

    2014-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2009-01-26T23:59:59.000Z

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

  4. 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

    1999-01-01T23:59:59.000Z

    The COMETHE 4D Release 23 computer code was used to evaluate the thermal, chemical and mechanical performance of weapons-grade MOX fuel irradiated under U.S. light water reactor typical conditions. Comparisons were made to and UO? fuels exhibited...

  5. Design and Optimization of Neuro-Fuzzy-Based Recognition of Musical Rhythm Patterns

    E-Print Network [OSTI]

    Weyde, Tillman

    Design and Optimization of Neuro-Fuzzy-Based Recognition of Musical Rhythm Patterns Tillman Weyde framework to support computer models and applications has not yet been established. Musical Pattern@uos.de Abstract The task of recognizing patterns and assigning rhythmic structure to unquan- tized musical input

  6. Bioremediation of Uranium Plumes with Nano-scale

    E-Print Network [OSTI]

    Fay, Noah

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

  7. Timetable of events Knowledge exchange seminars 23 to 26 september 2013

    E-Print Network [OSTI]

    Reading, University of

    Estate Valuation Methods Welcome and Introduction Michael Newey President RICS seminar 9:30 Eliane Monetti USP Valuation Methods in Brazil seminar 10:00 Fernando Burone UB Urban Land Valuations of Development Project in Argentina seminar 10:30 Coffee break 11:00 Real Estate Valuation Methods Neil Crosby Uo

  8. Curriculum vitae of Viviana Mascardi (updated to September Personal data

    E-Print Network [OSTI]

    Mascardi, Viviana

    rock carvings. #12;Role: Principal investigator, UO coordinator Funding source and programme: MIUR FIRB: Mobilit`a Intelligente Ecosostenibile (MIE) Role: Participant Funding source and programme: MIUR, progetto for Reliable Large-Scale Software Systems (BETTY) Role: Participant Funding source and programme: ICT COST

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

    SciTech Connect (OSTI)

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

    2000-08-01T23:59:59.000Z

    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.

  10. 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

  11. 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)

    2013-07-15T23:59:59.000Z

    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.

  12. Morphological and ultrastructural study of extrusion texturized defatted soy flour

    E-Print Network [OSTI]

    Kazemzadeh, Massoud

    1980-01-01T23:59:59.000Z

    and the speed of the cutting knii'e. Moist extrudate is then dried, ;:cooled, and, packaged. Dried, textured products contain 6 ? 8'fo mois- ture, 50-53%%uo protein, 1'%%d fat, 3% fiber, and 5 ? 6% ash. Den- sities may range from 9-25 1b/cu. ft. (Ziemba...

  13. ENDF/B-V, ENDF/B-VI, AND ENDF/B-VII.0 RESULTS FOR THE DOPPLER-DEFECT BENCHMARK (U)

    SciTech Connect (OSTI)

    MOSTELLER, RUSSELL D. [Los Alamos National Laboratory

    2007-02-09T23:59:59.000Z

    A set of computational benchmarks for the Doppler reactivity defect has been specified for an infinite array of identical fuel pin cells containing normal or enriched UO{sub 2} fuel, reactor-recycle mixed-oxide (MOX) fuel, or weapons-grade MOX fuel. The Doppler coefficient of reactivity, as well as the Doppler defect, can be computed for each of the cells. The MCNP5 Monte Carlo code was used to perform calculations for these benchmarks using cross sections derived from the ENDF/B-V, ENDF/B-VI, and ENDF/B-VII.0 nuclear data sets. The Doppler coefficients obtained from the three data sets exhibit very similar behavior. The Doppler coefficient for UO{sub 2} fuel becomes less negative with increasing enrichment, with a generally asymptotic shape. The Doppler coefficient for the reactor-recycle MOX becomes less negative with increasing PuO{sub 2} content but exhibits less curvature than that for UO{sub 2} fuel. The Doppler coefficient for weapons-grade MOX shows a pronounced shoulder between 1 wt.% and 2 wt.% PuO{sub 2}, with a nearly constant value thereafter. The Doppler coefficient for heavily loaded MOX fuel, whether reactor-recycle or weapons-grade, is significantly more negative than that for highly enriched UO{sub 2} fuel.

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

    E-Print Network [OSTI]

    Azad, Abdul-Majeed

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

  15. *EH&S has hire date, date of birth, employee's gender and hours of employment on file University of Washington

    E-Print Network [OSTI]

    Wilcock, William

    /Parking Lot: Room: Other: Incident Details: Classification (Please select level and check an item below) Uo Violence: Patient, Staff, Visitors h Radiation h Motor Vehicles h Non-human Primates h Drugs h None h Other Housekeeping h Inclement Weather h Slippery/Uneven surface h Ergonomics Issues h Sharp Objects h Hot Objects h

  16. Proceedings of IMECE04 2004 ASME International Mechanical Engineering Congress

    E-Print Network [OSTI]

    Motta, Arthur T.

    of uranium dioxide (UO2) encased in long thin tubes (called fuel cladding) made of zirconium alloys as coolant to remove the heat from the core and produce power. The fuel cladding has the functions procedures of Zr alloys so that corrosion of the zirconium alloy cladding tubing in nuclear reactors

  17. CX-011566: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

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

  18. Oldest and Largest The Weather

    E-Print Network [OSTI]

    one spot to fourth place in the U,S. News and World Report college rankings this year but remained in FILGs This year, 30 ~ambridge stu- CMI, Page 17 FRANK DABEK-THE TECH Pamela ThUo and her sons ~icholas

  19. A strategic partnership with the University of Exeter A logo representing an equal partenership or collaboration should be displayed at

    E-Print Network [OSTI]

    Bristol, University of

    Example B A strategic partnership with the University of Exeter · A logo representing an equal partenership or collaboration should be displayed at the bottom left of the letter. The logo can be smaller or equal size to the UoB logo but MUST NOT be larger (see Example A). A brief caption should also

  20. 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 ...

  1. Yucca Mountain Project - Argonne National Laboratory annual progress report, FY 1994

    SciTech Connect (OSTI)

    Bates, J.K.; Fortner, J.A.; Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Buck, E.C.; Wolf, S.F.

    1995-02-01T23:59:59.000Z

    This document reports on the work done by the Nuclear Waste Management Section of the Chemical Technology Division (CMT), Argonne National Laboratory, in the period October 1993-September 1994. Studies have been performed to evaluate the performance of nuclear waste glass and spent fuel samples under unsaturated conditions (low volume water contact) that are likely to exist in the Yucca Mountain environment being considered as a potential site for a high-level waste repository. Tests with simulated waste glasses have been in progress for over eight years and demonstrate that actinides from initially fresh glass surfaces will be released as a result of the spallation of reacted glass layers from the surface, as the small volume of water passes over the waste form. Studies are also underway to evaluate the performance of spent fuel samples and unirradiated UO{sub 2} in projected repository conditions. Tests with UO{sub 2} have been ongoing for nine years and show that the oxidation of UO{sub 2} occurs rapidly, and the resulting paragenetic sequence of secondary phases that form on the sample surface is similar to that observed in natural analogues. The reaction of spent fuel samples under conditions similar to those used with UO{sub 2} have been in progress for nearly two years, and the results suggest that spent fuel follows the same reaction progress as UO{sub 2}. The release of individual fission products and transuranic elements was not congruent, with the release being controlled by the formation of small particles or colloids that are suspended in solution and transported away from the waste form. The reaction progress depends on the composition of the spent fuel samples used and, likely, on the composition of the groundwater that contacts the waste form.

  2. New insights into uranium (VI) sol-gel processing

    SciTech Connect (OSTI)

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

    1990-01-01T23:59:59.000Z

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

  3. Modeling and analyses of postulated UF{sub 6} release accidents in gaseous diffusion plant

    SciTech Connect (OSTI)

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

    1995-10-01T23:59:59.000Z

    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. 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}). 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. In the postulated accident scenario, {approximately}7900 kg (17,500 lb) of hot UF{sub 6} vapor is released over a 5 min period from the process piping into the atmosphere of a large 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. MELCOR model was first used to develop a single volume representation of a process building and its results were compared with those from past lumped parameter models specifically developed for studying UF{sub 6} release accidents. Preliminary results indicate that MELCOR predicted results (using a lumped formulation) are comparable with those from previously developed models.

  4. 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

    1984-01-01T23:59:59.000Z

    Tg uoTquaqaz (T-u) p)aTZ (pter 'uozqzsoduroo uragsIs aqua pue aseqd ozzamEyod aqua qqtrr uoTgoezazut naqz uodn quapuadap saT&zooyarr ge urrrnyoo aqua Suts "accus 'suoTqeqsnqzad asaqg suotqeqznqzad urntzqtyTnba (y-u) asneo yyTrr aSueqo uox...&zooyaA aSeqd rruzaoyg avS uaqrr . (/e/ 'ae) = . "&r a?qa (t-u) x (~-u) azzs go sarrtqerrzsap maaqqosT uotgdsos go xtsqem = ?([ s, ?'I aqua go so&oars (y-u) x T e = ?Z (-, -u) x (y-u) azas go xzzqem Zqaquapz ('Le/ "Be) = '"?q pue ZV "?q = "r&V qeqg qons...

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

    SciTech Connect (OSTI)

    Collin, Blaise P.

    2014-08-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Blaise, Collin

    2014-07-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Chauvin, Aaron Lawrence

    1962-01-01T23:59:59.000Z

    UTequneg uo qg . Iecgg9g QUGQseQlfg u'gBQUnog de@ i q8 'I' I r' 8 ~ ~ e . w ~ o e ~, s ~ i . ~ e ~~ggggg yp 1 ~, ~ ~ 0, 'I . ~ ~ 0 0 I 4 \\ 1 lk, , ~ 0 S ~ ~ 1 OggggQggQQ- gg ~ ~ ~ 5 fl 4 ~ ~ '. ' . ~ '. ; irajueegXsguag-greg ~ ~ 4, ~ ~ 4... -xoa gsam~nos a~ ~ ggyydn oueyg aqua yo query wsamqgxou a~ uo paqeaoy sy eaxe ps' qgxog-~xoH aogoguod aqua, DVHZ Sgg 5 r Apngs yo asm ~ ay yaaaas ~ spry duo e~ sg~ gym' eyuopa~ ~ Sao~' appose Imp yy~g$08 QQB SS~lgl 5$PClg QQ'f0+ Q~ '~ gpgOg g...

  8. Repellents to prevent cattle browsing of pine seedlings

    E-Print Network [OSTI]

    Duncan, Don Arlen

    1959-01-01T23:59:59.000Z

    Tm ZzaA ssA Zqyaywm. c". pqZ gso~qv ~ 'ZBTTsq=om qsmqsu 'BuTTfmzq zus Buys?ozq syqqev usqq usqqo eaeuss moTg sBomsTr Buy~aaB syqqso Zq pseuss sar~ cq. oqd: &~ssqo uo Zqyqs~~crz s~? ~o fiqTuopem sqq. puc 'eBuyqTsae Bssp go aBsquoozacT q. eaqDyu sctq...~ygi Zq paBccrsp azar' ppo~l; go sBv~aso r Zquo ~~~ cueqaoB qaz~ooB Zq;;aZozqgap ossa sax~ Bus eq. qqqaz Zq - zo uaqqyq a&as eBuZTSase or q. ZBuqe stqq go ssznoa sqq Buymg mIog Z. toquaAuy pgaTg aq, uo sB mp qyqqzu Zus Z~q, . og spam azaw suoTeTAo~ zou...

  9. Effect of Helium Accumulation on the Spent Fuel Microstructure

    SciTech Connect (OSTI)

    Ferry, Cecile [Department of Physico-chemistry, Commissariat a l'Energie Atomique, CEA-Saclay, Gif-sur-Yvette, 91191 (France); Piron, Jean-Paul [Commisariat a l'Energie Atomique, CEA - Cadarache, Saint-Paul Lez Durance, 13108 (France); Stout, Ray [Rho Beta Sigma Affaires, Livermore, CA, CA 94550 (United States)

    2007-07-01T23:59:59.000Z

    In a nuclear spent fuel repository, the aqueous rapid release of radio-activity from exposed spent fuel surfaces will depend on the pellet microstructure at the arrival time of water into the disposal container. Research performed on spent fuel evolution in a closed system has shown that the evolution of microstructure under disposal conditions should be governed by the cumulated {alpha}-decay damage and the subsequent helium behavior. The evolution of fission gas bubble characteristics under repository conditions has to be assessed. In UO{sub 2} fuels with a burnup of 47.5 GWd/t, the pressure in fission gas bubbles, including the pressure increase from {alpha}-decay helium atoms, is not expected to reach the critical bubble pressure that will cause failure, thus micro-cracking in UO{sub 2} spent fuel grains is not expected. (authors)

  10. Experimental Verification of a Cracked Fuel Mechanical Model

    SciTech Connect (OSTI)

    Williford, R. E.

    1982-12-01T23:59:59.000Z

    This report describes the results of a series of laboratory experiments conducted to independently verify a model that describes the nonlinear mechanical behavior of cracked fuel in pelletized UO{sub 2}/Zircaloy nuclear fuel rods under normal operating conditions. After a brief description of the analytical model, each experiment is discussed in detail. Experiments were conducted to verify the general behavior and numerical values for the three primary independent modelling parameters (effective crack roughness, effective gap roughness, and total crack length), and to verify the model predictions that the effective Young's moduli for cracked fuel systems were substantially less than those for solid UO{sub 2} pellets. In general, the model parameters and predictions were confirmed, and new insight was gained concerning the complexities of cracked fuel mechanics.

  11. Fuel System Compatibility Issues for Prometheus-1

    SciTech Connect (OSTI)

    DC Noe; KB Gibbard; MH Krohn

    2006-01-20T23:59:59.000Z

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

  12. 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.

    1984-05-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Takahashi, Yuya; Nakamura, Hitoshi; Yamada, Akira; Mizuguchi, Koji; Fujita, Reiko [Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan)

    2013-07-01T23:59:59.000Z

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

  14. Depleted uranium hexafluoride: Waste or resource?

    SciTech Connect (OSTI)

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

    1995-07-01T23:59:59.000Z

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

  15. DU-AGG pilot plant design study

    SciTech Connect (OSTI)

    Lessing, P.A.; Gillman, H.

    1996-07-01T23:59:59.000Z

    The Idaho National Engineering Laboratory (INEL) is developing new methods to produce high-density aggregate (artificial rock) primarily consisting of depleted uranium oxide. The objective is to develop a low-cost method whereby uranium oxide powder (UO[sub 2], U[sub 3]O[sub ]8, or UO[sub 3]) can be processed to produce high-density aggregate pieces (DU-AGG) having physical properties suitable for disposal in low-level radioactive disposal facilities or for use as a component of high-density concrete used as shielding for radioactive materials. A commercial company, G-M Systems, conducted a design study for a manufacturing pilot plant to process DU-AGG. The results of that study are included and summarized in this report. Also explained are design considerations, equipment capacities, the equipment list, system operation, layout of equipment in the plant, cost estimates, and the proposed plan and schedule.

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

    SciTech Connect (OSTI)

    Blaise Collin

    2014-08-01T23:59:59.000Z

    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.

  17. Naturalistic interpersonal reactions to assertive and unassertive styles

    E-Print Network [OSTI]

    Paquette, Raymond Joseph

    1988-01-01T23:59:59.000Z

    sinai awaJ)xa jo asneoaq pa&oaias Ai[ensn os(e aJa~ suogen(ena aug 6uiqew sgoa jqns auj. uogJasse jo s(ana[ &ua~ajjrp Ran 6ugeJgsuowap s(apow qgie sauaos pake(d a[oJ gJous paztlgn peq qojeasaJ sno[naJd aqi aa&Jasse aug uo uo[gjasse jo )oedwl aui... weeks ago, then failed to return them at the next class, thus forcing you to take notes on scrap paper. Now he/she is asking to borrow your notes again. Suppose that the person who borrowed your notes were someone you had only met in class and did...

  18. Simulation of the thermodynamic properties of organic extraction solutions

    SciTech Connect (OSTI)

    Kolker, A.R.

    1986-05-01T23:59:59.000Z

    A method is proposed for the simulation of the activity coefficients of the components, the excess volume, the heat of mixing, and other excess thermodynamic functions of organic extraction solutions. The method is based on a search in an assigned region for parameters of the NRTL equations of local composition for which the state of the solution satisfies the requirements of chemical thermodynamics, as well as the assigned recovery criteria. The following binary systems of the solvent-extractant, and solvent-solvate types have been simulated according to the program developed on an ES-1033 computer: C6H/sub 14/-TBP, CHC1/sub 3/-TBP, CC1/sub 4/-TBP, UO/sub 2/(NO/sub 3/)/sub 2/ X 2TBP-TBP, and CC1/sub 4/-UO/sub 2/(NO/sub 3/)/sub 2/ X 2TBP.

  19. Method for fluorination of uranium oxide

    DOE Patents [OSTI]

    Petit, George S. (Oak Ridge, TN)

    1987-01-01T23:59:59.000Z

    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.

  20. Synthesis of triglyceride by the intestinal mucosa

    E-Print Network [OSTI]

    Buell, George Christopher

    1958-01-01T23:59:59.000Z

    FE CT OF A MU CO SA L HO MO GE NA TE PR EP AR ED IN 92$ GL YC ER OL ON scuO! vO i?I C\\ivOm oIT\\ m 00rnrn o[>.rn QMo lJ-qQ)UOaoW W!coO ? CO...l i b r a r y A & M COLLEGE OF TEXAS A&MCOLAEA GF CTEXS&1LTE9L 5& COL EMCLACEM8S VI1GA8 8 9.PPrecuc.so 5f XLGTXL 1OTEACGlOLT 5ILSS Ayid.ccr- cs c3r Xeu-yucr A23ssp sa c3r 8te.2ypcyeup uo- Vr23uo.2up 1spprtr sa CrnuP .o Duec.up aypa...

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

    E-Print Network [OSTI]

    Baringer, Philip S.

    1991-12-01T23:59:59.000Z

    as a muon increases from 0.8%%uo at 1.4 GeV/c to 1.4%%uo at 2.0 GeV, and stays constant beyond. III. DETECTION OF D ~K e+v, D 's are required to be decay products of D*+'s from the reaction D*+~D m+. D*+ candidates are formed from three... and transverse polarizations [14]. We have checked our background estimates by using M(Ksvr )(Gev) FIG. 2. The histogram is the Kzm candidate mass for the decay mode D ~K* e+v, . The solid line is a fit to the data using a Breit-Wigner line shape for the signal...

  2. Feasibility Study of MOX Fuel Online Burnup Analysis

    SciTech Connect (OSTI)

    Dennis, M.L.; Usman, S. [University of Missouri-Rolla, 222 Fulton Hall, 1870 Miner Circle, Rolla, MO 65409-0170 (United States)

    2006-07-01T23:59:59.000Z

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

  3. 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)

    2013-07-01T23:59:59.000Z

    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.

  4. High-harmonic XUV source for time- and angle-resolved photoemission spectroscopy

    SciTech Connect (OSTI)

    Dakovski, Georgi L [Los Alamos National Laboratory; Li, Yinwan [Los Alamos National Laboratory; Durakiewicz, Tomasz [Los Alamos National Laboratory; Rodriguez, George [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

    We present a laser-based apparatus for visible pump/XUV probe time- and angle-resolved photoemission spectroscopy (TRARPES) utilizing high-harmonic generation from a noble gas. Femtosecond temporal resolution for each selected harmonic is achieved by using a time-delay-compensated monochromator (TCM). The source has been used to obtain photoemission spectra from insulators (UO{sub 2}) and ultrafast pump/probe processes in semiconductors (GaAs).

  5. 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

  6. Head-end process for the reprocessing of HTGR spent fuel

    SciTech Connect (OSTI)

    Chen, J.; Wen, M. [Institute of Nuclear and New Energy Technology, Tsinghua University, Bejing 10084 (China)

    2013-07-01T23:59:59.000Z

    The reprocessing of HTGR spent fuels is in favor of the sustainable development of nuclear energy to realize the maximal use of nuclear resource and the minimum disposal of nuclear waste. The head-end of HTGR spent fuels reprocessing is different from that of the LWR spent fuels reprocessing because of the difference of spent fuel structure. The dismantling of the graphite spent fuel element and the highly effective dissolution of fuel kernel is the most difficult process in the head end of the reprocessing. Recently, some work on the head-end has been done in China. First, the electrochemical method with nitrate salt as electrolyte was studied to disintegrate the graphite matrix from HTGR fuel elements and release the coated fuel particles, to provide an option for the head-end technology of reprocessing. The results show that the graphite matrix can be effectively separated from the coated particle without any damage to the SiC layer. Secondly, the microwave-assisted heating was applied to dissolve the UO{sub 2} kernel from the crashed coated fuel particles. The ceramic UO{sub 2} as the solute has a good ability to absorb the microwave energy. The results of UO{sub 2} kernel dissolution from crushed coated particles by microwave heating show that the total dissolution percentage of UO{sub 2} is more than 99.99% after 3 times cross-flow dissolution with the following parameters: 8 mol/L HNO{sub 3}, temperature 100 Celsius degrees, initial ratio of solid to liquid 1.2 g/ml. (authors)

  7. 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)

    1989-01-01T23:59:59.000Z

    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.

  8. 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

    1989-12-31T23:59:59.000Z

    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.

  9. Use of Savannah River Site facilities for blend down of highly enriched uranium

    SciTech Connect (OSTI)

    Bickford, W.E.; McKibben, J.M.

    1994-02-01T23:59:59.000Z

    Westinghouse Savannah River Company was asked to assess the use of existing Savannah River Site (SRS) facilities for the conversion of highly enriched uranium (HEU) to low enriched uranium (LEU). The purpose was to eliminate the weapons potential for such material. Blending HEU with existing supplies of depleted uranium (DU) would produce material with less than 5% U-235 content for use in commercial nuclear reactors. The request indicated that as much as 500 to 1,000 MT of HEU would be available for conversion over a 20-year period. Existing facilities at the SRS are capable of producing LEU in the form of uranium trioxide (UO{sub 3}) powder, uranyl nitrate [UO{sub 2}(NO{sub 3}){sub 2}] solution, or metal. Additional processing, and additional facilities, would be required to convert the LEU to uranium dioxide (UO{sub 2}) or uranium hexafluoride (UF{sub 3}), the normal inputs for commercial fuel fabrication. This study`s scope does not include the cost for new conversion facilities. However, the low estimated cost per kilogram of blending HEU to LEU in SRS facilities indicates that even with fees for any additional conversion to UO{sub 2} or UF{sub 6}, blend-down would still provide a product significantly below the spot market price for LEU from traditional enrichment services. The body of the report develops a number of possible facility/process combinations for SRS. The primary conclusion of this study is that SRS has facilities available that are capable of satisfying the goals of a national program to blend HEU to below 5% U-235. This preliminary assessment concludes that several facility/process options appear cost-effective. Finally, SRS is a secure DOE site with all requisite security and safeguard programs, personnel skills, nuclear criticality safety controls, accountability programs, and supporting infrastructure to handle large quantities of special nuclear materials (SNM).

  10. TREKiSM Issue 43

    E-Print Network [OSTI]

    1985-01-01T23:59:59.000Z

    .JO 'snJ02V"~ IIAaQ GmT IS uO lqau9Q A'~I 04. U&A& o. SnolAqo oq Pln04s { y ·Z ' I -78- *******Reader Survey ******* {c We realize that you new folk have Just sent one of these suckers, {c but would appreciate your feedback anyway, as we've added some...

  11. 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, K. J.; Cacioppo, J. T.

    2007-01-01T23:59:59.000Z

    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...

  12. 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)

    2002-01-01T23:59:59.000Z

    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.

  13. Nuclear carrier business volume projections, 1980-2000

    SciTech Connect (OSTI)

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

    1980-05-01T23:59:59.000Z

    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.

  14. Characterization Methodology for Decommissioning Low and Intermediate Level Fissile Nuclide Contaminated Buried Soils and Process Piping Using Photon Counting

    E-Print Network [OSTI]

    Pritchard, Megan L

    2014-05-03T23:59:59.000Z

    Tellurium CFR Code of Federal Regulations Ci Curie cm Centimeter cpm Counts per Minute CZT Cadmium Zinc Telluride D&D Decontamination and Decommissioning DU Depleted Uranium EAF European Activation File ENDF Evaluated Nuclear Data... an abundance of depleted uranium ( ? 0.96 wt. % 235U in UO2), thorium, and radium. These nuclides emit gammas and produce high count rates in a windowed NaI. Detection of these nuclides is pertinent for waste management and radiological exposure purposes...

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

    SciTech Connect (OSTI)

    Valdez, James A [Los Alamos National Laboratory; Byler, Darrin D [Los Alamos National Laboratory

    2012-06-08T23:59:59.000Z

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

  16. In situ treatment of VOCs by recirculation technologies

    SciTech Connect (OSTI)

    Siegrist, R.L.; Webb, O.F.; Ally, M.R.; Sanford, W.E. [Oak Ridge National Lab., TN (US); Kearl, P.M.; Zutman, J.L. [Oak Ridge National Lab., Grand Junction, CO (US)

    1993-06-01T23:59:59.000Z

    The project described herein was conducted by Oak Ridge National Laboratory (ORNL) to identify processes and technologies developed in Germany that appeared to have near-term potential for enhancing the cleanup of volatile organic compound (VOC) contaminated soil and groundwater at DOE sites. Members of the ORNL research team identified and evaluated selected German technologies developed at or in association with the University of Karlsruhe (UoK) for in situ treatment of VOC contaminated soils and groundwater. Project activities included contacts with researchers within three departments of the UoK (i.e., Applied Geology, Hydromechanics, and Soil and Foundation Engineering) during fall 1991 and subsequent visits to UoK and private industry collaborators during February 1992. Subsequent analyses consisted of engineering computations, groundwater flow modeling, and treatment process modeling. As a result of these project efforts, two processes were identified as having near-term potential for DOE: (1) the vacuum vaporizer well/groundwater recirculation well and (2) the porous pipe/horizontal well. This document was prepared to summarize the methods and results of the assessment activities completed during the initial year of the project. The project is still ongoing, so not all facets of the effort are completely described in this document. Recommendations for laboratory and field experiments are provided.

  17. A Spectroscopic Study of the effect of Ligand Complexation on the Reduction of Uranium(VI) by Anthraquinone-2,6-disulfonate (AH2DS)

    SciTech Connect (OSTI)

    Wang, Zheming; Wagnon, Ken B.; Ainsworth, Calvin C.; Liu, Chongxuan; Rosso, Kevin M.; Fredrickson, Jim K.

    2008-11-03T23:59:59.000Z

    In this project, the reduction rate of uranyl complexes with hydroxide, carbonate, EDTA, and Desferriferrioxamine B (DFB) by anthraquinone-2,6-disulfonate (AH2DS), a potential electron shuttle for microbial reduction of metal ions (Newman and Kolter 2000), is studied by stopped-flow kinetics techniques under anoxic atmosphere. The apparent reaction rates varied with ligand type, solution pH, and U(VI) concentration. For each ligand, a single largest kobs within the studied pH range was observed, suggesting the influence of pH-dependent speciation on the U(VI) reduction rate. The maximum reaction rate found in each case followed the order of OH- > CO32- > EDTA > DFB, consistent with the same trend of the thermodynamic stability of the uranyl complexes and ionic sizes of the ligands. Increasing the stability of uranyl complexes and ligand size decreased the maximum reduction rate. The pH-dependent rates were modeled using a second-order rate expression that was assumed to be dependent on a single U(VI) complex and AH2DS species. By quantitatively comparing the calculated and measured apparent rate constants as a function of pH, species AHDS3- was suggested as the primary reductant in all cases examined. Species UO2CO3(aq) , UO2HEDTA-, and (UO2)2(OH)22+ were suggested as the principal electron acceptors among the U(VI) species mixture in carbonate, EDTA, and hydroxyl systems, respectively.

  18. 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.

    1997-01-30T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Simon R. Phillpot

    2012-06-08T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2008-01-01T23:59:59.000Z

    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...

  1. 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

    2006-01-09T23:59:59.000Z

    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.

  2. 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.

    2012-08-22T23:59:59.000Z

    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.

  3. Development of a high-temperature oven for the 28 GHz electron cyclotron resonance ion source

    SciTech Connect (OSTI)

    Ohnishi, J., E-mail: ohnishi@riken.jp; Higurashi, Y.; Kidera, M.; Ozeki, K.; Nakagawa, T. [RIKEN Nishina Center, Wako, Saitama 351-0198 (Japan)] [RIKEN Nishina Center, Wako, Saitama 351-0198 (Japan)

    2014-02-15T23:59:59.000Z

    We have been developing the 28 GHz ECR ion source in order to accelerate high-intensity uranium beams at the RIKEN RI-beam Factory. Although we have generated U{sup 35+} beams by the sputtering method thus far, we began developing a high-temperature oven with the aim of increasing and stabilizing the beams. Because the oven method uses UO{sub 2}, a crucible must be heated to a temperature higher than 2000?°C to supply an appropriate amount of UO{sub 2} vapor to the ECR plasma. Our high-temperature oven uses a tungsten crucible joule-heated with DC current of approximately 450 A. Its inside dimensions are ?11 mm × 13.5 mm. Since the crucible is placed in a magnetic field of approximately 3 T, it is subject to a magnetic force of approximately 40 N. Therefore, we used ANSYS to carefully design the crucible, which was manufactured by machining a tungsten rod. We could raise the oven up to 1900?°C in the first off-line test. Subsequently, UO{sub 2} was loaded into the crucible, and the oven was installed in the 28 GHz ECR ion source and was tested. As a result, a U{sup 35+} beam current of 150 ?A was extracted successfully at a RF power of approximately 3 kW.

  4. Helium Behavior in Oxide Nuclear Fuels: First Principles Modeling

    SciTech Connect (OSTI)

    D. Gryaznov; S. Rashkeev; E. A. Kotomin; E. Heifets; Y. Zhukovskii

    2010-10-01T23:59:59.000Z

    UO2 and (U, Pu)O2 solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein. We present density functional theory (DFT) calculations for UO2, PuO2 and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO2 remains semiconducting with He in the octahedral position while UO2 requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.

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

    SciTech Connect (OSTI)

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

    2015-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-08-01T23:59:59.000Z

    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.

  7. Synthesis of triglyceride by the intestinal mucosa 

    E-Print Network [OSTI]

    Buell, George Christopher

    1958-01-01T23:59:59.000Z

    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...

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

    SciTech Connect (OSTI)

    Natalie J. Gese; Batric Pesic

    2013-03-01T23:59:59.000Z

    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.

  9. 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.

    1988-04-01T23:59:59.000Z

    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.

  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 [Mitsubishi Heavy Industries, Ltd., 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan); Kosaka, Yuji [Nuclear Development Corporation, 622-12 Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Arakawa, Yasushi [The Kansai Electric Power Co., Inc., 8 Yokota, 13 Goichi, Mihama-cho, Mikata-gun, Fukui, 919-1141 (Japan)

    2007-07-01T23:59:59.000Z

    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. Analysis of Reactor Physics Experiment for the Irradiated LWR MOX Fuels

    SciTech Connect (OSTI)

    Katsuyuki; Kawashima; Toru, Yamamoto; Katsuichiro, Kamimura [Japan Nuclear Energy Safety Organization, Tokyu Reit Toranomon Bldg. 7F, 3-17-1, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2006-07-01T23:59:59.000Z

    As an important part to validate the LWR core neutronics analysis methods, Japan Nuclear Energy Safety Organization (JNES) has been participating in the REBUS international program and performing analyses and evaluations of the reactor physics experiment data including the irradiated fuels. In REBUS program, physics experiments were performed at the VENUS critical test facility in SCK/CEN, Belgium, in which the five core configurations were tested. In each core configuration, the central part of the 3.3%- and 4%-enriched UO{sub 2} fuel core was replaced by the test bundle of (1)fresh MOX fuel, (2)medium-burnup MOX fuel, (3)high-burnup MOX fuel, (4)fresh UO{sub 2} fuel, and (5)irradiated UO{sub 2} fuel. Measured parameters are critical water level, water level reactivity, fission rate distributions, and neutron flux distributions. In this paper, the results of the fresh MOX and medium-burnup MOX core critical experiments and analysis are presented. The medium-burnup MOX fuel used in this test is 6.9% in initial fissile enrichment and the burnup averaged over the fuel rods is 20 GWd/t. In the core critical analysis, a continuous energy Monte Carlo code MVP was used with the JENDL-3.2 nuclear data library as well as a deterministic analysis code SRAC. The calculated results are compared with the experimental ones. (authors)

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

    SciTech Connect (OSTI)

    Putman, Valerie Lee

    2003-01-01T23:59:59.000Z

    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.

  13. YUCCA Mountain Project - Argonne National Laboratory, Annual Progress Report, FY 1997 for activity WP 1221 unsaturated drip condition testing of spent fuel and unsaturated dissolution tests of glass.

    SciTech Connect (OSTI)

    Bates, J. K.; Buck, E. C.; Emery, J. W.; Finch, R. J.; Finn, P. A.; Fortner, J.; Hoh, J. C.; Mertz, C.; Neimark, L. A.; Wolf, S. F.; Wronkiewicz, D. J.

    1998-09-18T23:59:59.000Z

    This document reports on the work done by the Nuclear Waste Management Section of the Chemical Technology Division of Argonne National Laboratory in the period of October 1996 through September 1997. Studies have been performed to evaluate the behavior of nuclear waste glass and spent fuel samples under the unsaturated conditions (low-volume water contact) that are likely to exist in the Yucca Mountain environment being considered as a potential site for a high-level waste repository. Tests with actinide-doped waste glasses, in progress for over 11 years, indicate that the transuranic element release is dominated by colloids that continuously form and span from the glass surface. The nature of the colloids that form in the glass and spent fuel testing programs is being investigated by dynamic light scattering to determine the size distribution, by autoradiography to determine the chemistry, and by zeta potential to measure the electrical properties of the colloids. Tests with UO{sub 2} have been ongoing for 12 years. They show that the oxidation of UO{sub 2} occurs rapidly, and the resulting paragenetic sequence of secondary phases forming on the sample surface is similar to that observed for uranium found in natural oxidizing environments. The reaction of spent fuel samples in conditions similar to those used with UO{sub 2} have been in progress for over six years, and the results suggest that spent fuel forms many of the same alteration products as UO{sub 2}. With spent fuel, the bulk of the reaction occurs via a through-grain reaction process, although grain boundary attack is sufficient to have reacted all of the grain boundary regions in the samples. New test methods are under development to evaluate the behavior of spent fuel samples with intact cladding: the rate at which alteration and radionuclide release occurs when water penetrates fuel sections and whether the reaction causes the cladding to split. Alteration phases have been formed on fine grains of UO{sub 2} in contact with small volumes of water within a several month period when the radiolysis product H{sub 2}O{sub 2} is added to the groundwater solution. The test setup has been mocked up for operation with spent fuel in the hot-cell.

  14. Pillared and open-framework uranyl diphosphonates

    SciTech Connect (OSTI)

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

    2011-09-15T23:59:59.000Z

    The hydrothermal reactions of uranium trioxide, uranyl acetate, or uranyl nitrate with 1,4-benzenebisphosphonic acid in the presence of very small amount of HF at 200 deg. C results in the formation of three different uranyl diphosphonate compounds, [H{sub 3}O]{sub 2}{l_brace}(UO{sub 2}){sub 6}[C{sub 6}H{sub 4}(PO{sub 3})(PO{sub 2}OH)]{sub 2}[C{sub 6}H{sub 4}(PO{sub 2}OH){sub 2}]{sub 2}[C{sub 6}H{sub 4}(PO{sub 3}){sub 2}]{r_brace}(H{sub 2}O){sub 2} (Ubbp-1), [H{sub 3}O]{sub 4}{l_brace}(UO{sub 2}){sub 4}[C{sub 6}H{sub 4}(PO{sub 3}){sub 2}]{sub 2}F{sub 4}{r_brace}.H{sub 2}O (Ubbp-2), and {l_brace}(UO{sub 2})[C{sub 6}H{sub 2}F{sub 2}(PO{sub 2}OH){sub 2}(H{sub 2}O){r_brace}{sub 2}.H{sub 2}O (Ubbp-3). The crystal structures of these compounds were determined by single crystal X-ray diffraction experiments. Ubbp-1 consists of UO{sub 7} pentagonal bipyramids that are bridged by the phosphonate moieties to form a three-dimensional pillared structure. Ubbp-2 is composed of UO{sub 5}F{sub 2} pentagonal bipyramids that are bridged through the phosphonate oxygen atoms into one-dimensional chains that are cross-linked by the phenyl spacers into a pillared structure. The structure of Ubbp-3 is a three-dimensional open-framework with large channels containing water molecules with internal dimensions of approximately 10.9x10.9 A. Ubbp-1 and Ubbp-2 fluoresce at room temperature. - Graphical Abstract: Illustration of the three-dimensional open-framework structure of {l_brace}(UO{sub 2})[C{sub 6}H{sub 2}F{sub 2}(PO{sub 2}OH){sub 2}(H{sub 2}O){r_brace}{sub 2}.H{sub 2}O viewed along the c-axis. The structure is constructed from UO{sub 7} units, pentagonal bipyramids=green, oxygen=red, phosphorus=magenta, carbon=black, hydrogen=white. Highlights: > The influence of the uranyl salt anions and pH were critically examined in relation to structural variation. > The acetate and nitrate counter ions of uranyl may be acting as structure directing agents. > The use of rigid phenyl spacer yield a three-dimensional network of pillared structures of uranyl diphosphonates that fluoresce. > The fluorination of the phenyl ring under hydrothermal condition. > The large voids in this structure are suggestive of potential applications in sorption, separation of gases and in catalytic processes.

  15. 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)

    2014-05-01T23:59:59.000Z

    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.

  16. Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

    SciTech Connect (OSTI)

    Todd R. Allen, Director

    2011-04-01T23:59:59.000Z

    The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center’s investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center’s research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

  17. FEASIBILITY STUDY OF DUPOLY TO RECYCLE DEPLETED URANIUM.

    SciTech Connect (OSTI)

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

    1998-02-01T23:59:59.000Z

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

  18. Towards the reanalysis of void coefficients measurements at proteus for high conversion light water reactor lattices

    SciTech Connect (OSTI)

    Hursin, M.; Koeberl, O.; Perret, G. [Paul Scherrer Institut PSI, 5232 Villigen (Switzerland)

    2012-07-01T23:59:59.000Z

    High Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resources thanks to a higher breeding ratio than standard LWR. Their uses together with the current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and the US today. However, one of the issues related to HCLWR is their void reactivity coefficient (VRC), which can be positive. Accurate predictions of void reactivity coefficient in HCLWR conditions and their comparisons with representative experiments are therefore required. In this paper an inter comparison of modern codes and cross-section libraries is performed for a former Benchmark on Void Reactivity Effect in PWRs conducted by the OECD/NEA. It shows an overview of the k-inf values and their associated VRC obtained for infinite lattice calculations with UO{sub 2} and highly enriched MOX fuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with the libraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spread of results for voided conditions is found for the high content MOX fuel. The spread of eigenvalues for the moderated and voided UO{sub 2} fuel are about 200 pcm and 700 pcm, respectively. The standard deviation for the VRCs for the UO{sub 2} fuel is about 0.7% while the one for the MOX fuel is about 13%. This work shows that an appropriate treatment of the unresolved resonance energy range is an important issue for the accurate determination of the void reactivity effect for HCLWR. A comparison to experimental results is needed to resolve the presented discrepancies. (authors)

  19. High temperature behavior of metallic inclusions in uranium dioxide

    SciTech Connect (OSTI)

    Yang, R.L.

    1980-08-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Bryant, George Frank

    1959-01-01T23:59:59.000Z

    ?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...

  1. Facility Operations 1993 fiscal year work plan: WBS 1.3.1

    SciTech Connect (OSTI)

    Not Available

    1992-11-01T23:59:59.000Z

    The Facility Operations program is responsible for the safe, secure, and environmentally sound management of several former defense nuclear production facilities, and for the nuclear materials in those facilities. As the mission for Facility Operations plants has shifted from production to support of environmental restoration, each plant is making a transition to support the new mission. The facilities include: K Basins (N Reactor fuel storage); N Reactor; Plutonium-Uranium Reduction Extraction (PUREX) Plant; Uranium Oxide (UO{sub 3}) Plant; 300 Area Fuels Supply (N Reactor fuel supply); Plutonium Finishing Plant (PFP).

  2. Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Fortner, Jeffrey A.; Kropf, A. Jeremy; Jerden, James L.; Cunnane, James C. [Chemical Engineering, Argonne National Laboratory, CMT/205, 9700 S. Cass Avenue, Argonne, IL, 60439 (United States)

    2007-07-01T23:59:59.000Z

    Performance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron X-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO{sub 2} matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO{sub 2}{sup 2+}) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of {approx} 50 micrometers. We find evidence of a thin ({approx} 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO{sub 2}{sup +}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. (authors)

  3. Monopole Strength in Ni-58

    E-Print Network [OSTI]

    Youngblood, David H.; Lui, YW.

    1991-01-01T23:59:59.000Z

    PHYSICAL REVIEW C VOLUME 44, NUMBER 5 Monopole strength in Ni NOVEMBER 1991 D. H. Youngblood and Y.-W. Lui Cyclotron Institute, Texas AdkM Uni Uersi ty, College Station, Texas 77843 (Received 20 June 1991) Differential cross-section data from... strength is locat- ed nearer the quadrupole (for Ca [3] and Si [4] at vir- tually the same energy). Only two reports of substantial strength in lighter nuclei are in the literature. Lui et al. [4] reported 66%%uo of the EO energy-weighted sum rule...

  4. A VSP transformation technique for the determination of subsurface structure

    E-Print Network [OSTI]

    Malloy, Jeffrey Edward

    1985-01-01T23:59:59.000Z

    offset is 305 m, and reflector depth is 3000 m. . . . . . . . . . . . . . . . . , . . . . . . . 13 5 Plot showing normalized coverage as a function of dip of the reflector. Source offset is 305 m. Reflector depth is 2000 m, 6 Plot showing normalized... Geometry showing significance of as. 10 Physical significance of angle uo. 20 22 11 Stacking fold vs. bin offset for a reflector at 2000 m with a dip of 15'. Source offset is 305 m. 23 12 Stack'ing fold vs. bin offset for a reflector at 2000 m with a...

  5. trPd_ ~ ---THE uNIVERSITY OF TENNESSEE

    E-Print Network [OSTI]

    Cui, Yan

    #12;· · · · · ·· · · · · · · · · · · ........llOlll·utM DJ IGllR U*JSUl'I N··PYN-.at IJWD918CKDAU · I 7 F .. 2 , l'l$u JU .m , a f';' #12;~\\\\\\\\\\\\\\\\ l\\ 1 1liti\\\\\\\\\\III\\I\\l!;~IIiilll\\ 1 I!1 1!!I 1111111111111111111:1:'.ll!i;mJ11f·l ···t ·hO·Oooo-··o·uo ...... TO SEEK THE GOAL ... #12;I ~ . ' ;__I - J

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

    SciTech Connect (OSTI)

    Rodney C. Ewing

    2004-10-07T23:59:59.000Z

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

  7. 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.

    2011-01-12T23:59:59.000Z

    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.

  8. The use of the principle of expected value in engineering design

    E-Print Network [OSTI]

    Golondzinier, Theodore Matthew

    1970-01-01T23:59:59.000Z

    ? 2xp. + p - 2a o x)]dx. 20 2 3 (3. 4) Let '2 ? ~ L. o~2v By factoring the x terms and completing the square the integral be- comes L ] exp[- ? ((x - (u + a o )) + 2a uo + a o ]dx, (3. 5) 2 2 2 2 4 22 3 3 3 L exp[ ? a o + a u] ) exp... constant M (0) the moment generating function of x the n moment as given by the power series expansion of th the moment generating function m 0 p( ~ ) the mean the mode the number of times a function is simulated the discrete probability...

  9. Nuclear Energy Research Initiative. Development of a Stabilized Light Water Reactor Fuel Matrix for Extended Burnup

    SciTech Connect (OSTI)

    BD Hanson; J Abrefah; SC Marschman; SG Prussin

    2000-09-08T23:59:59.000Z

    The main objective of this project is to develop an advanced fuel matrix capable of achieving extended burnup while improving safety margins and reliability for present operations. In the course of this project, the authors improve understanding of the mechanism for high burnup structure (HBS) formation and attempt to design a fuel to minimize its formation. The use of soluble dopants in the UO{sub 2} matrix to stabilize the matrix and minimize fuel-side corrosion of the cladding is the main focus.

  10. Metal chelates of N,N'-dihydroxyethyl-N,N'-ethylenediaminedisuccinic acid and selected metal ions

    E-Print Network [OSTI]

    Hampton, Joan Martiner

    1972-01-01T23:59:59.000Z

    The approximate volumes of standard hydrochloric acid required to neutralize basic groups of the ligand, po- tassium nitrate to maintain an ionic strength of 0. 10, and distilled water to bring the volume to 50 ml, were intro- duced into the titration cell...n iS. 4" 16 zlo 18. 79 18, 56 zs. lp EDDS 12. 9q 1. 4. ozq 18. 45q Ig. ozq zz. oq EDDS, N, N'-ethylenediaminedisuccinic acid. t = 25 C and 0 u f. l unlesS otherwise stated, aFrom Colman-Porter and Monk , u~O. From Jogephs, w = 0. 15...

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

    SciTech Connect (OSTI)

    Donald Olander

    2005-08-24T23:59:59.000Z

    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.

  12. Assessment of light water reactor fuel damage during a reactivity initiated accident

    SciTech Connect (OSTI)

    MacDonald, P.E.; Seiffert, S.L.; Martinson, Z.R.; McCardell, R.K.; Owen, D.E.; Fukuda, S.K.

    1980-01-01T23:59:59.000Z

    This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalpy of approximately 140 cal/g UO/sub 2/. Volume expansion of previously irradiated fuel upon melting may cause deformation and rupture of the cladding, and coolant channel blockage at higher peak enthalpies.

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

    DOE Patents [OSTI]

    Friedman, H.A.

    1984-06-13T23:59:59.000Z

    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.

  14. Status of ANL out-of-pile investigations of severe accident phenomena for liquid metal reactors

    SciTech Connect (OSTI)

    Spencer, B.W.; Marchaterre, J.F.; Anderson, R.P.; Armstrong, D.R.; Baker, L.; Cho, D.H.; Gabor, J.D.; Pedersen, D.R.; Sienicki, J.J.; Stein, R.P.

    1986-01-01T23:59:59.000Z

    Research addressing LMFBR whole core accidents has been terminated, and there is now emphasis on quantifying reactivity feedbacks, and in particular enhancing negative feedback, so that advanced LMR designs will provide inherently safe operation. The status of recent HCDA-related laboratory research performed at ANL, up to the time that such activities were no longer needed to support CRBR licensing, is described. Included are descriptions of programs addressing sodium channel voiding, fuel sweepout, fuel dispersal and plugging, boiled-up pool, UO/sub 2//sodium FCI, and debris coolability. Descriptions of recent investigations involving the metal fuel/sodium system are also included.

  15. Fiscal Year 2010 Summary Report on the Epsilon-Metal Phase as a Waste Form for 99 Tc

    SciTech Connect (OSTI)

    Strachan, Denis M.; Crum, Jarrod V.; Buck, Edgar C.; Riley, Brian J.; Zumhoff, Mac R.

    2010-09-30T23:59:59.000Z

    Epsilon metal (?-metal) is generated in nuclear fuel during irradiation. This metal consists of Pd, Ru, Rh, Mo, and some Te. These accumulate at the UO2 grain boundaries as small (ca 5 µm) particles. These metals have limited solubility in the acid used to dissolve fuel during reprocessing and in typical borosilicate glass. These must be treated separately to improve overall waste loading in glass. This low solubility and their survival in 2 Gy-old natural reactors led us to investigate them as a waste form for the immobilization of 99Tc and 107Pd, two very long-lived isotopes.

  16. Computer modeling of the spallation process

    E-Print Network [OSTI]

    Walker, Wayne Claire

    1978-01-01T23:59:59.000Z

    ) de. (62) Note here that S(E) is the entropy of a nucleus with an energy between E andE+ dE. An allowance must be made at this point for the affect that the Coulomb field will have on the cross section of the reverse process. If we assume... 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...

  17. Viscosity of many-component glasses

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Arrigoni, Benjamin M.; Schweiger, Michael J.

    2009-06-01T23:59:59.000Z

    The effect of composition on the viscosity of multicomponent glasses was expressed as a function of temperature and composition for three composition regions containing various subsets of Al2O3, B2O3, Bi2O3, CaO, Cr2O3, F, Fe2O3, K2O, Li2O, MgO, MnO, Na2O, NiO, P2O5, SiO2, UO2, and ZrO2. Limits of applicability of the composition models are discussed.

  18. A generalized land use study of the San Jacinto River watershed of Texas

    E-Print Network [OSTI]

    Buckley, Frank A.

    1951-01-01T23:59:59.000Z

    Pi ? ? ft o ] 00 I to jco jco j ? co 03 ? 5 ^ O aS ?? ?? ?p U Pi ? ? ft O 4? CQ ? U O aS ?? ?* 43 U Pi ? ? ft O BG uo CM n CM c> o 2 0 t - cr... ?P = Q ?* ^ft ? = -p ? c3 ^ O > T i ? n3 *H ? <35 J i JL, DO * CQ CJ -PCO O ciJ o CO ft O 4= P i ? O O o CO ? Q

  19. Multi-spectral glucose sensing using a polarimetric differencing technique

    E-Print Network [OSTI]

    Michael, Mathew

    1995-01-01T23:59:59.000Z

    - cpf 4'g2+ Z I2$M2KD2 Kv& s+a + Z Uo KLI2 s+b Figure 3. 2: Closed Loop System Diagram. d&s -rotation due to glucose, &ter -compensation rotation for Faraday Rotator, I -light intensity, E, -detector responsivity, M -modulation depth, Ko... 4. 1 Sequential Multi-Spectral Experimental System. . . . . . . . . 22 4. 2 4. 3(a) Optical Train (Vector Representation). Detector Output of a Non-Optically Active Sample (P=O) . . . . . 24 . . . . 25 4. 3(b) Detector Output of an Optically...

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

    SciTech Connect (OSTI)

    Malony, Allen D. [Department of Computer and Information Science, University of Oregon] [Department of Computer and Information Science, University of Oregon; Wolf, Felix G. [Juelich Supercomputing Centre, Forschungszentrum Juelich] [Juelich Supercomputing Centre, Forschungszentrum Juelich

    2014-01-31T23:59:59.000Z

    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.

  1. 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.

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

    E-Print Network [OSTI]

    Stalnaker, Hwa Sun

    1987-01-01T23:59:59.000Z

    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...

  3. Transuranic Transmutation and Criticality Calculation Sensitivity to Heterogeneous Lattice Effects - 12391

    SciTech Connect (OSTI)

    Barbaras, Sean A. [United States Military Academy, West Point, New York 10996 (United States); Knight, Travis W. [University of South Carolina, Columbia, South Carolina 29208 (United States)

    2012-07-01T23:59:59.000Z

    Using Mixed Oxide (MOX) fuel in traditional Pressurized Water Reactor (PWR) assemblies has been researched at length and has shown to provide the benefit of transmutation and targets the amount and toxicity of high level waste needed to be managed. Advanced MOX concepts using enriched Uranium Dioxide (UO{sub 2}) are required for multiple recycling of plutonium. The use of MOX and ordinary UO{sub 2} fuel in the same assembly as well as unfueled rods and assembly edge effects contrasts with the unit cell computational assumption of a uniform infinite array of rods. While a deterministic method of calculating the Dancoff factor has traditionally been employed in fuel assembly analysis due to the lighter computational and modeling requirements, this research seeks to determine the validity of the uniform, infinite lattice assumption with respect to Dancoff factor and determine the magnitude of the impact of nonuniform lattice effects on fuel assembly criticality calculations as well as transuranic isotope production and transmutation. This research explored the pin-to-pin interaction in a non-uniform lattice of MOX fuel rods and UO{sub 2} fuel rods through the impact of the calculated Dancoff factors from the deterministic method used in SCALE versus the Monte Carlo method used in the code DANCOFF-MC. Using the Monte Carlo method takes into account the non-uniform lattice effects of having neighboring fuel rods with different cross-sectional spectra whereas the Dancoff factor calculated by SCALE assumes a uniform, infinite lattice of one fuel rod type. Differences in eigenvalue calculations as a function of burnup are present between the two methods of Dancoff factor calculation. The percent difference is greatest at low burnup and then becomes smaller throughout the cycle. Differences in the transmutation rate of transuranic isotopes in the MOX fuel are also present between the Dancoff factor calculation methods. The largest difference is in Pu-239, Pu-242, and Am-241 composition whereas U-238, Pu-242, and Pu-238 composition was not changed by taking into account the non-homogenous lattice effects. Heterogeneous lattice effects do change the calculated eigenvalue and transmutation rate in a non-uniform lattice of MOX fuel rods and UO{sub 2} fuel. However, the uncertainty in the ENDF data used by SCALE in these calculations is large enough that the infinite lattice assumption remains valid. (authors)

  4. Actinide speciation in glass leach-layers: An EXAFS study

    SciTech Connect (OSTI)

    Biwer, B.M.; Soderholm, L. [Argonne National Lab., IL (United States); Greegor, R.B. [Boeing Co., Seattle, WA (United States); Lytle, F.W. [EXAFS Co., Pioche, NV (United States)

    1996-12-31T23:59:59.000Z

    Uranium L{sub 3} X-ray absorption data were obtained from two borosilicate glasses, which are considered as models for radioactive wasteforms, both before and after leaching. Surface sensitivity to uranium speciation was attained by a novel application of simultaneous fluorescence and electron-yield detection. Changes in speciation are clearly discernible, from U(VI) in the bulk to (UO{sub 2}){sup 2+}-uranyl in the leach layer. The leach-layer uranium concentration variations with leaching times are also determined from the data.

  5. Neutronics evaluation of the replacement fuel for the Texas A&M Nuclear Science Center TRIGA Reactor

    E-Print Network [OSTI]

    Shelton, Gary Ross

    1989-01-01T23:59:59.000Z

    compares with the calculated value of 1. 4 x 10 n /(cm s). These results were considered sufficient to justify using the same procedure to model the core with the BeO-UOs-Er fuel. The graphite sections of the U-ZrHr s-Er fuel rods were not used in the Be...O- UOr-Er model since it was believed the additional fuel loading made possible by their removal would be of more value. After several calculations with the UOr mass fraction varying from 21. 5 to 55. 32 percent, control rod reactivity worths were...

  6. Trace Fission Product Ratios for Nuclear Forensics Attribution of Weapons-Grade Plutonium from Fast Breeder Reactor Blankets

    E-Print Network [OSTI]

    Osborn, Jeremy

    2014-08-13T23:59:59.000Z

    ), whereas that used in an FBR blanket fuel is depleted uranium (0.25 atom percent 235U). The energy production in the FBR core is from the seed fuel subassemblies containing mixed oxides (MOX) of PuO2 and UO2. A plot of fast and thermal neutron energy... of the program involves a fleet of fast breeder reactors. The stage two fast breeder reactors, beginning with the PFBR, will be fueled with reactor-grade plutonium and depleted uranium from the reprocessed spent fuel of stage one reactors and will breed more...

  7. Development of Self-Interrogation Neutron Resonance Densitometry (SINRD) to Measure the Fissile Content in Nuclear Fuel

    E-Print Network [OSTI]

    Lafleur, Adrienne

    2011-10-21T23:59:59.000Z

    BU Burnup BWR Boiling Water Reactor C/S Containment and Surveillance DU Depleted Uranium Fuel FA Fuel Assembly FC Fission Chambers IAEA International Atomic Energy Agency IE Initial Enrichment LEU Low Enriched Uranium Fuel (UO2) LWR Light... as the removal of 50% or more of the irradiated fuel pins from an assembly. The spent fuel pins removed can be replaced with nothing (except water) or replaced with non-irradiated dummy pins, such as iron or depleted uranium (DU) pins. SIMULATED CASES...

  8. Density Functional Studies on the Complexation and Spectroscopy of Uranyl Ligated with Acetonitrile and Acetone Derivatives

    SciTech Connect (OSTI)

    Schoendorff, George E.; Windus, Theresa L.; De Jong, Wibe A.

    2009-12-12T23:59:59.000Z

    The coordination of nitrile (acetonitrile, propionitrile, and benzonitrile) and carbonyl (formaldehyde, ethanal, and acetone) ligands to the uranyl dication (UO22+) has been examined using density functional theory (DFT) utilizing relativistic effective core potentials (RECPs). Complexes containing up to six ligands have been modeled for all ligands except formaldehyde, for which no minimum could be found. A comparison of relative binding energies indicates that five coordinate complexes are predominant while a six coordinate complex involving propionitrile ligands might be possible. Additionally, the relative binding energy and the weakening of the uranyl bond is related to the size of the ligand and, in general, nitriles bind more strongly to uranyl than carbonyls.

  9. Newberry EGS Seismic Velocity Model

    SciTech Connect (OSTI)

    Templeton, Dennise

    2013-10-01T23:59:59.000Z

    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.

  10. Short stories of Julio Cortazar: English translations of contemporary Latin American fiction.

    E-Print Network [OSTI]

    Dean, Frances F

    1968-01-01T23:59:59.000Z

    = ior hd s willingness uo support icy initial inquiici"-: in thi. , endeavor. T a!a grate . 1 to Dr. , Jones a'id to Kiss Deverly Jones foz ti!eir. Fou111e taslc of proof ? reading all st 0? 7 es, ? ot only in J-'ngl i. sii 'but B1 so in Snani sh... autl'. ors. There are mary con. emporary write os of Iiatin Amo ica who certainly must have a potentla:~ bu" heretofore largely unsuspected appeal for North American readers. Hy way of illustrating the above remarks, I wi sh to offer a selection...

  11. DECAY HEAT CONDITIONS OF CURRENT AND NEXT GENERATION REACTORS 

    E-Print Network [OSTI]

    Choe, JongSoo 1985-

    2012-05-04T23:59:59.000Z

    nitrate solution con- taining methanol and an additive, spherical droplets are produced by a vibration dropping technique. The diameter of uranyl nitrate droplet is determined by the combination of the #3;ow rate of metal solution and the frequency...2+Ar) IPyC coating (C3H6+Ar) SiC coating (CH3SiCl3+H2) OPyC coating (C3H6+Ar) Burnable poison Graphite block Graphite sleeve Fuel rod Fuel compact TRISO coated particle UO2 particle Uranyl nitrate solution Overcoat particle Fig. 2. Flow...

  12. Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications

    SciTech Connect (OSTI)

    Yang Zhong; Robert C. O'Brien; Steven D. Howe; Nathan D. Jerred; Kristopher Schwinn; Laura Sudderth; Joshua Hundley

    2011-11-01T23:59:59.000Z

    The feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

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

    SciTech Connect (OSTI)

    Ni, Chengbao; Shuh, David; Raymond, Kenneth

    2011-03-07T23:59:59.000Z

    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)].

  14. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    SciTech Connect (OSTI)

    Phillpot, Simon; Tulenko, James

    2011-09-08T23:59:59.000Z

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  15. U(VI) reduction to mononuclear U(VI) by desulfitobacterium spp.

    SciTech Connect (OSTI)

    Fletcher, K. E.; Boyanov, M. I.; Thomas, S. H.; Wu, Q.; Kemner, K. M.; Loffler, F. E. (Biosciences Division); (Georgia Inst. of Tech.)

    2010-06-15T23:59:59.000Z

    The bioreduction of U(VI) to U(IV) affects uranium mobility and fate in contaminated subsurface environments and is best understood in Gram-negative model organisms such as Geobacter and Shewanella spp. This study demonstrates that U(VI) reduction is a common trait of Gram-positive Desulfitobacterium spp. Five different Desulfitobacterium isolates reduced 100 {mu}M U(VI) to U(IV) in <10 days, whereas U(VI) remained soluble in abiotic and heat-killed controls. U(VI) reduction in live cultures was confirmed using X-ray absorption near-edge structure (XANES) analysis. Interestingly, although bioreduction of U(VI) is almost always reported to yield the uraninite mineral (UO{sub 2}), extended X-ray absorption fine structure (EXAFS) analysis demonstrated that the U(IV) produced in the Desulfitobacterium cultures was not UO{sub 2}. The EXAFS data indicated that the U(IV) product was a phase or mineral composed of mononuclear U(IV) atoms closely surrounded by light element shells. This atomic arrangement likely results from inner-sphere bonds between U(IV) and C/N/O- or P/S-containing ligands, such as carbonate or phosphate. The formation of a distinct U(IV) phase warrants further study because the characteristics of the reduced material affect uranium stability and fate in the contaminated subsurface.

  16. 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

    1997-08-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

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

    1995-06-06T23:59:59.000Z

    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.

  18. 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)

    1995-01-01T23:59:59.000Z

    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.

  19. Alteration of Coffinite (USiO{sub 4}) Under Reducing and Oxidizing Conditions

    SciTech Connect (OSTI)

    Deditius, Artur Piotr; Utsunomiya, Satoshi; Ewing, Rodney C. [Geological Sciences, University of Michigan, 1100 N. University Ave., Ann Arbor, MI, 48109 (United States)

    2007-07-01T23:59:59.000Z

    Samples of natural coffinite (USiO{sub 4}.nH{sub 2}O) from Grants uranium region, New Mexico were investigated in order to understand the alteration process of coffinite under reducing and oxidizing conditions. Alteration of the primary coffinite under reducing conditions was promoted by organic acids, and as a result, secondary coffinite precipitated. Subsequently oxidizing fluids altered the coffinite, and (Na,K)-boltwoodite [(Na,K)(UO{sub 2})(SiO{sub 3}OH)(H{sub 2}O){sub 1.5}] and jachymovite [(UO{sub 2})(SO{sub 4})(OH){sub 14}(H{sub 2}O){sub 13}] precipitated with no rare earth elements. Based on the charge balance calculation, we suggest that the amount of U{sup 6+} in the coffinite is less than 0.2 [apfu] and U{sup 6+} is accommodated in the structure via substitution: U{sup 4+} + Si{sup 4+} {r_reversible} U{sup 6+} + 2(OH){sup -}. The high and variable totals for electron microprobe analyses indicate that H{sub 2}O is not an essential component in coffinite structure. The U-Pb ages of coffinite formation vary from 36.6-0 Ma suggesting that the coffinite has precipitated continuously in this period and organic matter can preserve reducing conditions even when oxidizing conditions dominate. (authors)

  20. Standard test methods for arsenic in uranium hexafluoride

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2005-01-01T23:59:59.000Z

    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...

  1. Selective elution of uranium from amidoxime polymer. I

    SciTech Connect (OSTI)

    Hirotsu, T.; Katoh, S.; Sugasaka, K.; Takai, N.; Seno, M.; Itagaki, T.

    1987-07-01T23:59:59.000Z

    The separative elution of uranium from an amidoxime polymer was examined by the column method with hydrochloric acid solutions. The amidoxime polymer was immersed in seawater for 40 d for preparation of an uranium-loaded polymer sample for the elution experiments; the metal ions adsorbed were Mg(II), Ca(II), Fe(III), Ni(II), Cu(II), and Zn(II) as well as UO/sub 2/(VI). It was found from the pH dependence of elution extent by a batch method that the order of elution pH values is Fe(III) < UO/sub 2/(VI) < Cu(II) < Ni(II) < Zn(II) < Ca(II) < Mg(II). In the elution by a column method, Mg(II), Ca(II), Zn(II), and Ni(II) were eluted completely by 0.1 M HCl and the eluate of enriched uranium was obtained by a succeeding elution with 0.5 or 1 M HCl. This eluate contained Cu(II) and Fe(III), which could be removed in the succeeding step. The elution treatment with hydrochloric acid solutions hardly affected the adsorptivity for uranium in seawater. It was suggested that the elution of uranium with hydrochloric acid solutions from amidoxime polymers is satisfactorily applicable to uranium elution in the recovery of uranium from seawater with amidoxime polymers.

  2. Aerosols Generated by Free Fall Spills of Powders and Solutions in Static Air

    SciTech Connect (OSTI)

    Sutter, S. L.; Johnston, J. W.; Mishima, J.

    1981-12-01T23:59:59.000Z

    Safety assessments and environmental impact statements for nuclear fuel cycle facilities require an estimation of potential airborne releases. Aerosols generated by accidents are being investigated to develop the source terms for these releases. The lower boundary accidental release event would be a free fall spill of powders or liquids in static air. Experiments measured the mass airborne and particle size distribution of these aerosols for various source sizes and spill heights. Two powder and liquid sources were used: Ti02 and uo2; and aqueous uranine (sodium fluorescein) and uranyl nitrate solutions. Spill height and source size were significant in releases of both powders and liquids. For the source powders used (l "m uo2 and 1.7 "m Ti0 2, quantities from 25 g to 1000 g, and fall heights of 1 m and 3m), the maximum source airborne was 0.12%. The maximum source airborne was an order of magnitude less for the liquids (with source quantities ranging from 125 to 1000 cc at the same fall heights). The median aerodynamic equivalent diameters for collected airborne powder ranged from 6 to 26.5 "m; liquids ranged from 4.1 to 34 "m. All of the spills produced a significant fraction of respirable particles 10 ~m and less.

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

    SciTech Connect (OSTI)

    Carmack, W.J.; Husser, D.L.; Mohr, T.C.; Richardson, W.C. [BWX Technologies, PO Box 785, Lynchburg, VA 24505-0785 (United States)

    2004-02-04T23:59:59.000Z

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

  4. Assessment of molten debris freezing in a severe RIA in-pile test. [PWR; BWR

    SciTech Connect (OSTI)

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

    1980-01-01T23:59:59.000Z

    An understanding of the freezing of molten debris on cold core structures following a hypothetical core meltdown accident in a light water reactor (LWR) is of importance to reactor safety analysis. The purpose of the present investigation was to analyze the transient freezing of the molten debris produced in a severe reactivity initiated accident (RIA) scoping test, designated RIA-ST-4, which was performed in the Power Burst Facility and simulated a BWR control rod drop accident. In the RIA-ST-4 experiment, a single, unirradiated, 20 wt % enriched, UO/sub 2/ fuel rod contained within a Zircaloy flow shroud was subjected to a single power burst which deposited a total energy of about 700 cal/g UO/sub 2/. This energy deposition is well above what is possible in a commercial LWR during a hypothetical control rod drop (BWR) or ejection (PWR) accident. However, the performance of such an in-pile test has provided important information regarding molten debris movement, relocation, and freezing on cold walls.

  5. Comparison of GAP-3 and GAP-4 experiments with conduction freezing calculations. [LMFBR

    SciTech Connect (OSTI)

    Sienicki, J.J.; Spencer, B.W.

    1983-01-01T23:59:59.000Z

    Experiments GAP-3 and GAP-4 were performed at ANL to investigate the ability of molten fuel to penetrate downward through the narrow channels separating adjacent subassemblies during an LMFBR hypothetical core disruptive accident. Molten fuel-metal mixtures (81% UO/sub 2/, 19% Mo) at an initial temperature of 3470/sup 0/K generated by a thermite reaction were injected downward into 1 m long rectangular test sections (gap thickness = 0.43 cm, channel width = 20.3 cm) initially at 1170/sup 0/K simulating the nominal Clinch River Breeder Reactor intersubassembly gap. In the GAP-3 test, a prolonged reaction time of approx. 15 s resulted in segregation of the metallic Mo and oxidic UO/sub 2/ constituents within the reaction vessel prior to injection. Consequently, Mo entered the test section first and froze, forming a complete plug at a penetration distance of 0.18 m. In GAP-4, the reaction time was reduced to approx. 3 s and the constituents remained well mixed upon injection with the result that the leading edge penetration distance increased to 0.35 m. Posttest examination of the cut-open test sections has revealed the existence of stable insulating crusts upon the underlying steel walls with melting and ablation of the walls only very localized.

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

    SciTech Connect (OSTI)

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

    1988-12-01T23:59:59.000Z

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

  7. 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

    2011-09-01T23:59:59.000Z

    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.

  8. 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)

    2013-05-15T23:59:59.000Z

    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}.

  9. 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.

    2015-01-01T23:59:59.000Z

    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.

  10. A calibration method for lateral forces for use with colloidal probe force microscopy cantilevers

    SciTech Connect (OSTI)

    Quintanilla, M. A. S.; Goddard, D. T. [Nexia Solutions Ltd., Springfields, Salwick, Preston, Lancashire PR4 0XJ (United Kingdom)

    2008-02-15T23:59:59.000Z

    A calibration method is described for colloidal probe cantilevers that enables friction force measurements obtained using lateral force microscopy (LFM) to be quantified. The method is an adaptation of the lever method of Feiler et al. [A. Feiler, P. Attard, and I. Larson, Rev. Sci. Instum. 71, 2746 (2000)] and uses the advantageous positioning of probe particles that are usually offset from the central axis of the cantilever. The main sources of error in the calibration method are assessed, in particular, the potential misalignment of the long axis of the cantilever that ideally should be perpendicular to the photodiode detector. When this is not taken into account, the misalignment is shown to have a significant effect on the cantilever torsional stiffness but not on the lateral photodiode sensitivity. Also, because the friction signal is affected by the topography of the substrate, the method presented is valid only against flat substrates. Two types of particles, 20 {mu}m glass beads and UO{sub 3} agglomerates attached to silicon tapping mode cantilevers were used to test the method against substrates including glass, cleaved mica, and UO{sub 2} single crystals. Comparisons with the lateral compliance method of Cain et al. [R. G. Cain, S. Biggs, and N. W. Page, J. Colloid Interface Sci. 227, 55 (2000)] are also made.

  11. Proliferation resistance of small modular reactors fuels

    SciTech Connect (OSTI)

    Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

    2013-07-01T23:59:59.000Z

    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.

  12. 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)

    2013-07-01T23:59:59.000Z

    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.

  13. TREKiSM Issue 34

    E-Print Network [OSTI]

    1984-01-01T23:59:59.000Z

    Drive Lewisville, Texas 75067 · CRYPTOGRA~I ANSWERS: ' 1I"j.LasAw aSJnOJ uO~S~LLOJ ~ JO 6u~4:).awos uo wlI 'w~rll IIi'" UeJ~a 4:).~'" aSJnOJ uO~S~LLOJ E uo d~4SaJeds e 5.:).1 isauog ':).aUELd e :).,US~ s~4:)' lng ll "~# lIi:).~ paJn:> I l:)'~ d...ND SOURCE . AL.L UflOiEO 1 TEO ,. ..H 'lfR1TING "NO ART IS THE WORI( OF THE EDITOR . ....... StJ8S[['IU IQN RATES; FO~Il"HU~RS CF r:NSA IN US, {t {t ~~::~"~ !+O~_~~!:~=~s~~g'1JA I~~~.: !AT~~~FA~~N~~: {t {t ' !~~U...

  14. Potential incorporation of transuranics into uranium phases

    SciTech Connect (OSTI)

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

    1999-12-07T23:59:59.000Z

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

  15. 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.

    1996-12-30T23:59:59.000Z

    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.

  16. Yucca Mountain project : FY 2006 annual report for waste form testingactivities.

    SciTech Connect (OSTI)

    Ebert, W. L.; Fortner, J. A.; Guelis, A. V.; Cunnane, J. C.

    2006-11-01T23:59:59.000Z

    This report describes the experimental work performed at Argonne National Laboratory (Argonne) during fiscal year (FY 2006) under the Bechtel SAIC Company, LLC (BSC) Memorandum Purchase Order (MPO) contract number B004210CM3X. Because this experimental work is focused on the dissolution and precipitation behavior of neptunium, the report also includes, or incorporates by reference, earlier results that are relevant to presenting a more complete understanding of the likely behavior of neptunium under experimental conditions relevant to the Yucca Mountain repository. Important results relevant to the technical bases, validations, and conservatisms in current source term models are summarized. The CSNF samples were observed to corrode following the general contour of the surface rather than via (for instance) grain boundary attack. This supports the current approach of estimating the effective surface area of corroding CSNF based on the geometric surface area of fuel pellet fragments. It was observed that the neptunium and plutonium concentrations in corroded CSNF samples were somewhat higher at and near the corrosion front (i.e., at the interface between the alteration product ''rind'' layer and the underlying fuel) than in the bulk fuel. The neptunium and plutonium at the corrosion front and in the uranyl alteration layer were found to be in the quadravalent (4+) oxidation state. The uranyl phases that constitute most of the alteration rind were depleted in neptunium relative to the bulk fuel: neptunium concentrations in the uranyl alteration rind were less than 20% of that in the parent fuel. Homogeneous precipitation tests have shown that solids precipitate from a 1 x 10{sup -4} M Np(V) solution over the temperature range of 200-280 C, but no evidence was found that any solids precipitated from the same solution at 150 C through 289 days. The solids formed in the homogeneous precipitation tests were predominantly a Np(IV)-bearing phase, probably NpO{sub 2}. The presence of UO{sub 2} resulted in the rapid precipitation at room temperature of similar amounts of Np(IV)- and Np(V)-bearing phases, probably NpO{sub 2} and Np{sub 2}O{sub 5}. Although the UO{sub 2} is presumed to act as a reducing agent for Np(V) that leads to the precipitation of a Np(IV)-bearing phase, the observed formation of a Np(V)-bearing phase suggests that the UO{sub 2} also catalyzes Np{sub 2}O5 precipitation under these test conditions.

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

    SciTech Connect (OSTI)

    DOE

    1997-04-01T23:59:59.000Z

    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.

  18. 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

    2012-04-01T23:59:59.000Z

    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.

  19. Marketing aspects of range sheep production in Texas

    E-Print Network [OSTI]

    Blackwell, James Wayne

    1958-01-01T23:59:59.000Z

    Ma]er sub)sots Agricultural goonIELo? LISRRRY I 4 M COLLEGE OF TEXlg iV:~"tiffin A% 'OI, : QV . W't~~': 3. .:"%'9 P-'-. QD'~7'iXC4 31i 'i'm'. A3 tune~ Ozbnttte~ to the Graduate School of tlat l~icultuxal and Hechanica1 College ef 'exaa fn ce... of Cased. ttes / c3 ~ Head Dapartawnt August, 195S Vtikl B'l% -'W' W~~'. ~ ". . i'P uO 9~~"'988 &8 RpD~CXSVJ. QB CO . :1' ~ iBTWig . ~ . "~sr @ha g~avw Zrs~Q a. 1k@ '. Em a~& !s, "~led'a M helotry? wt ~ t-e i'd ~ 'MpO~' ~ 8100 NRi~/ 3 QSBk8 8' 8...

  20. Multiscale numerical methods for partial differential equations using limited global information and their applications

    E-Print Network [OSTI]

    Jiang, Lijian

    2009-05-15T23:59:59.000Z

    form div(?(S)k?p) = f, (2.2) where the total mobility?(S) is given by?(S) = ?w(S)+?o(S) andf is a source term. The saturation dynamics affects the flow equations. One can derive the equation 8 describing the dynamics of the saturation ?S ?t +div(F) = 0..., (2.3) where F = ufw(S), with fw(S), the fractional flow of water, given by fw = ?w/(?w + ?o), and the total velocity u by: u = uw +uo = ??(S)k?p. (2.4) In the presence of capillary effects, an additional diffusion term is present in (2.3). If krw = S...

  1. Anhydrous ferric chloride as an alkylation catalyst: The condensation of 2-methylpropene and benzene, preparation and identification of several fractions.

    E-Print Network [OSTI]

    Mommessin, Pierre Robert

    1950-01-01T23:59:59.000Z

    Lg T: I snoop /~qua ep ? polio mn'[IunIp snoop. &qua 'op:zonI. xzg uo'oq 'oppxc+uad snzoqdsoqd 'pgoa o-?zngIos 'ap-=onIg ue3ozpZq sa qons 'sos. . Ie'=- plo'-' 1 c acTTa e Jd aqua uQ suoqgT oogppq ot )gmo Te qf~ $ese J IIT+ sane&Ia g&qq. emI~ amos... -uesezo. Qq~ , T eosazuao, qq. . ~ aueguq Pua . ue~::nod ~o uotgus -ueP too ev", PetPuq s Pixatsvnoty, Puu 'B?og 'Butuatip Q' "g J. elQ J euagnq-u erg qgtsp. Qe. ~-. aqqa eZV. SgtuSa Zattmtg Ouasueo. -tfiouo--. , see-tap pu- euez, :aqtfi. ?nq...

  2. The isolation of an unidentified factor from yeast extract for the formate-pyruvate exchange reaction in streptococcus faecalis

    E-Print Network [OSTI]

    Chen, Chi-sin

    1962-01-01T23:59:59.000Z

    % t ~ ~ ~ I ~ ' ~ ~ 1 ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ Qggg(gg -emmyao ccIsea g xssag paa p~?@ eggygeqtatI ~ sEwggasxg eh'$$os ga seyptgs +~qsqg Zj osogao oIs04 Q'dI~~ ~'pI4eqctp et@ tackle ccsolao oIsex QVl !g &'A&eqsp e% ++V oagko~ qq'1 aqua ccrc 4~08$ et...X; xoq, oag eqq go uoggaog~xn8 aqua cog smsqo, ~ ' . ' ~ ' ~ ~ ~ ~ ~ a~sex eousqoxs uog qqgn soqoag eqq go exnpeooz8 usga, soT~xnd ps~T ~ ~ ~ ~ \\ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ sxq goy o, uuqo~ qqTn sogosg eqg go exnpsoaxd uoTqmg~xm~. ~ puooog ~ I...

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

    DOE Patents [OSTI]

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

    1985-02-04T23:59:59.000Z

    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.

  4. PUREX transition project case study

    SciTech Connect (OSTI)

    Jasen, W.G.

    1996-04-15T23:59:59.000Z

    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.

  5. Control of the Pyrimidine biosynthetic pathway in different strains of Salmonella typhimurium

    E-Print Network [OSTI]

    Smith, Johnny Melven

    1974-01-01T23:59:59.000Z

    (2 (S121) 22 20 1. 1 1. 1 ~ able B. 1 r'si n Wrlo iype Levels of 0 I Case aq c B pacific fctiv'ty Specific Rc ivity Fold 0 repress'on LT2 (51) 119 17H u, i a, u. . ~l) 112 BB S+rains sere oro. . n on minimal media '. . ) S+raina were gro... ioe 3, 2 192 3. 0 a) 5tz'a '. ns w re crown nn minimal media:upplement d wi. "5 uraci 1, 5) 5t-sins growth were grown on 10 ug per ml of uracil and calle were harvested after had -e-, 'ad fc- 00 min. were grown on 50 uo per ml n u~ a . il and c...

  6. Information media used by cotton farmers in producing cotton in a ten-county area of North Central Texas

    E-Print Network [OSTI]

    Anwarul Karim, A. M

    1965-01-01T23:59:59.000Z

    M'Zl QK'~~644~ . Ki@84~p COU&lry B@4CGCBp 'BK+~ ~. KCiFS/3'l&6"8? P8+56f'8 &VXIV'". 50 thigh "80~43QX CQGCKL ~ '. NBQS Q CAt. 4' ?iles: M 49K~~' '%FAN 9X' ~~A';M, "VEST RK4R 5F~'5 hQVSX&7, gMgk': 4&F3, 'Gm:. M~4$M ~~C g". "~Q4QX' VW' 'sZ 5L~XL~'~&9 j l... on M O'Ia Used by Xoixnpea" Uoa ton Farmers fo? 8fz . Ukffeientt Btrapa ixx Uo'Nan Prodcc'tfan e . e" ~ e ~ o ~ ~ r ~ ' 6g Xnformabfon Illadfa Used by addle Age, UQbl;on farmers- fax 8', UJ. Xxex'Gv:g 8taePG e n. Uabtaon Prado''1 Qn a o a -. a...

  7. Uranium dioxide electrolysis

    DOE Patents [OSTI]

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

    2009-12-29T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Ferguson, Thomas Morgan

    1954-01-01T23:59:59.000Z

    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...

  9. Helium Migration Mechanisms in Polycrystalline Uranium Dioxide

    SciTech Connect (OSTI)

    Martin, Guillaume; Desgardin, Pierre; Sauvage, Thierry; Barthe, Marie-France [CERI, CNRS, 3 A rue de la Ferollerie, ORLEANS, 45071 (France); Garcia, Philippe; Carlot, Gaelle [DEN/DEC/SESC/LLCC, CEA Cadarache, Saint Paul Lez Durance, 13108 (France)

    2007-07-01T23:59:59.000Z

    This study aims at identifying the release mechanisms of helium in uranium dioxide. Two sets of polycrystalline UO{sub 2} sintered samples presenting different microstructures were implanted with {sup 3}He ions at concentrations in the region of 0.1 at.%. Changes in helium concentrations were monitored using two Nuclear Reaction Analysis (NRA) techniques based on the {sup 3}He(d,{alpha}){sup 1}H reaction. {sup 3}He release is measured in-situ during sample annealing at temperatures ranging between 700 deg. C and 1000 deg. C. Accurate helium depth profiles are generated after each annealing stage. Results that provide data for further understanding helium release mechanisms are discussed. It is found that helium diffusion appears to be enhanced above 900 deg. C in the vicinity of grain boundaries possibly as a result of the presence of defects. (authors)

  10. Synroc-D Type Ceramics Produced by Hot Isostatic Pressing and Cold Crucible Melting for Immobilisation of (Al, U) Rich Nuclear Waste

    SciTech Connect (OSTI)

    Vance, Eric R.; La Robina, Michael; Li, Huijun; Davis, Joel [Institute of Materials and Engineering Science, Australian Nuclear Science and Technology Organisation, Menai, NSW 2234 (Australia)

    2007-07-01T23:59:59.000Z

    A synroc-D ceramic consisting mostly of spinel, hollandite, pyrochlore-structured CaUTi{sub 2}O{sub 7}, UO{sub 2}, and Ti-rich regions shows promise for immobilisation of a HLW containing mainly Al and U, together with fission products. Ceramics with virtually zero porosities and waste loadings of 50-60 wt% on an oxide basis were prepared by cold crucible melting (CCM) at {approx}1500 deg. C, and also by subsolidus hot isostatic pressing (HIP) at 1100 deg. C to prevent volatile losses. PCT leaching test values for Cs were < 13 g/L, with all other normalised elemental extractions being well below 1 g/L. (authors)

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

    SciTech Connect (OSTI)

    Pritam Chakraborty; Michael R. Tonks; Giovanni Pastore

    2014-09-01T23:59:59.000Z

    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.

  12. Uranium-contaminated soils: Ultramicrotomy and electron beam analysis

    SciTech Connect (OSTI)

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

    1994-02-01T23:59:59.000Z

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

  13. Repository-relevant testing applied to the Yucca Mountain Project

    SciTech Connect (OSTI)

    Bates, J.K.; Gerding, T.J.; Veleckis, E.

    1989-04-01T23:59:59.000Z

    A repository environment poses a challenge to developing a testing program because of the diverse nature of conditions that may exist at a given time during the life of the repository. A starting point is to identify whether any potential waste-water contact modes are particularly deleterious to the waste form performance, and whether any interactions between materials present in the waste package environment need to be accounted for during modeling the waste form reaction. The Unsaturated Test method in one approach that has been developed by the Yucca Mountain Project (YMP) to investigate the above issues, and a description of results that have been obtained during the testing of glass and unirradiated UO{sub 2} are the subject of this report. 10 refs., 7 figs., 4 tabs.

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

    SciTech Connect (OSTI)

    Not Available

    1986-08-01T23:59:59.000Z

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

  15. 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.

    1983-01-01T23:59:59.000Z

    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.

  16. SOLUBILITY OF URANIUM AND PLUTONIUM IN ALKALINE SAVANNAH RIVER SITE HIGH LEVEL WASTE SOLUTIONS

    SciTech Connect (OSTI)

    King, W.; Hobbs, D.; Wilmarth, B.; Edwards, T.

    2010-03-10T23:59:59.000Z

    Five actual Savannah River Site tank waste samples and three chemically-modified samples were tested to determine solubility limits for uranium and plutonium over a one year time period. Observed final uranium concentrations ranged from 7 mg U/L to 4.5 g U/L. Final plutonium concentrations ranged from 4 {micro}g Pu/L to 12 mg Pu/L. Actinide carbonate complexation is believed to result in the dramatic solubility increases observed for one sample over long time periods. Clarkeite, NaUO{sub 2}(O)OH {center_dot} H{sub 2}O, was found to be the dominant uranium solid phase in equilibrium with the waste supernate in most cases.

  17. Recovery of uranium from seawater; 15: Development of amidoxime resins with high sedimentation velocity for passively driver fluidized bed adsorbers

    SciTech Connect (OSTI)

    Egawa, Hiroaki; Kabay, N.; Jyo, A.; Hirono, Masaki; Shuto, Taketomi (Kumamoto Univ. (Japan). Dept. of Applied Chemistry)

    1994-03-01T23:59:59.000Z

    In order to design the amidoxime resins (RNH) suitable for circulating fluidized bed adsorbers, RNH were prepared from precursory acrylonitrile-divinylbenzene copolymer beads of different particle sizes, and chemical and physical properties of the resulting RNH were evaluated. Specific surface areas, pore structures, swelling ratios, and anion and cation-exchange capacities of RNH are little affected by the particle size, while their sedimentation velocities in water increase with an increase in particle size as expected from fluid dynamics. Although the uptake of uncomplexed uranyl ion from a uranyl nitrate solution (0.01 M) was not influenced by the particle size, the uranium uptake from seawater decreases with an increase in the particle size, indicating that the particle diffusion of the bulky complexed species UO[sub 2](CO[sub 3])[sub 3][sup 4[minus

  18. Zirconium in the nuclear industry

    SciTech Connect (OSTI)

    Franklin, D.G.; Adamson, R.B.

    1984-01-01T23:59:59.000Z

    This book examines the properties of Zircaloy-2, Zircaloy-4, and Zr-2.5Nb with regard to their use as structural materials in nuclear reactors. Topics considered include refinement and fabrication (extractive metallurgy, zirconium and hafnium separation, electron-beam remelting, pressure tube development, cold working and heat treatments), basic metallurgical studies (etching, strain anisotropy, fuel cladding, anneal hardening, recrystallization, hydrides in zirconium alloy tubes), texture and irradiation creep (microstructure, ultrasonic velocity, in-reactor creep, fuel rods, deformation), irradiation growth (proton and neutron bombardment, high-fluence irradiation growth), corrosion (ZrO/sub 2/ films, aqueous corrosion kinetics, corrosive effects of lithium hydroxide, oxidation films, hydridation), fracture studies (stress-corrosion cracking, hydrogen cracking), and high-temperature and transient effects (cladding deformation in LOCA, high-temperature behavior of fuel rods, steam oxidation kinetics, dissolution of solid UO/sub 2/ by molten Zircaloy-4).

  19. 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

    2000-01-01T23:59:59.000Z

    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.

  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)

    1991-12-31T23:59:59.000Z

    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. 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)

    1991-12-31T23:59:59.000Z

    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}.

  2. 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)

    2006-01-20T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Hennington, Debra Louise

    1981-01-01T23:59:59.000Z

    6u[ge[nwt. gs uoy ajit. goagga qou si ynq saioads uaqqo po sa[ewag pue sauew ui uo[ge[nno age[nw[gs og g[asg[ Eq pasn uaaq os [e seq u[douq. opeuo5 o iuoiuouo uewnH asn s|. I 5u[wo[[og. unsoo uaq. go [[LM suo[ge[ -nno a[d[g[nw hei]g si goezq. xa... [iu[onpau iq Eouafoigga bu[paauq panoudwf u[douqopeuob otuoluoqo uewnH 'sauew oi[oEo ui uofqe[nno bufqe[enwfqs uog antqoayga X[46[4 si 004 'uanazoH 'sauew ue[no[L[og age[nwigs og sqdwagqe snouawnu aqua oq uoigippe u[ HLIug qgce sasuoq u( uocge...

  4. Music in the Aegean Bronze Age

    E-Print Network [OSTI]

    Younger, John G.

    1998-01-01T23:59:59.000Z

    ; s.E7.a8 5 r'ir cEtOl:5I HE EX EE€fif€UO g & Ffr F; tal..0) :>, ...1 (g v -i.v(hX 4.- >\\aAH L PV * oAO dJO\\ Sc\\ ^ k ..u) r'\\ i{ ^u V.X Hvd.l3 oaE 3 ln6z q'$ e!tE 3E vg:vL!ts (|i dE \\o a* E h' S; +: -r:+E ;E:I J. H .\\c.l$'J st s6 .E .X oqI E 9-94 P=u 3 -:th6 tr O. do--: Eegs€ir -E999.=b.,^'; 2 >,+ x -cizq: h iE q6bo'";itD E = F :'7 E io .= r= .= ;i Y ^-.oa4!.:!- ':-66*FE q! gsf i...

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

    SciTech Connect (OSTI)

    R. L. Williamson; D. A. Knoll

    2009-09-01T23:59:59.000Z

    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.

  6. Plastic deformation in oxide ceramics. Progress report, January 1-December 31, 1980

    SciTech Connect (OSTI)

    Heuer, A. H.

    1980-09-01T23:59:59.000Z

    Effects of O/U ratio on UO/sub 2/ single crystals oriented either to suppress or promote )100) < 110 > slip are described. For both orientations, a change from crystallographic to noncrystallographic slip is observed with composition and temperature and correlated with the presence of Willis defects. Research on the kinetics and mechanism of formation of crystallographic shear (CS) planes in reduced rutile (TiO/sub 2-x/) is also described. Two orientations of CS planes, )132) and )121), are present, although prior work had suggested that only one orientation would be present for the particular value of x studied. In-situ reductions in the environmental stage of the HVEM have also begun. Precipitation in nonstoichiometric spinels was also investigated.

  7. 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.

    2009-08-31T23:59:59.000Z

    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.

  8. 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.

    1980-01-01T23:59:59.000Z

    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.

  9. Results of the GAP-4 experiment on molten-fuel drainage through intersubassembly gap geometry. [LMFBR

    SciTech Connect (OSTI)

    Spencer, B.W.; Vetter, D.; Wesel, R.; Sienicki, J.J.

    1983-01-01T23:59:59.000Z

    One of the key issues in assessment of the meltout phase of a hypothetical core disruptive accident in the LMFBR system involves the timing and paths for dispersal of molten fuel from the disrupted core. A program of experiments is underway at Argonne National Laboratory to investigate molten fuel penetration through these postulated escape paths. The purpose of the GAP-4 test was to examine the penetration distances of molten fuel flowing through the flat, narrow channels representing the intersubassembly gap geometry. In the experiment design, the gap geometry was selected to be two-dimensional on the basis that the gap volume in a reactor design would be interconnected and continuous. The molten fuel used in these tests was a mixture of UO/sub 2/ (81%) and molybdenum (19%) which was generated by an exothermic thermite reaction at a temperature of approx. 3470 K.

  10. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    SciTech Connect (OSTI)

    Hartini, Entin, E-mail: entin@batan.go.id; Andiwijayakusuma, Dinan, E-mail: entin@batan.go.id [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)

    2014-09-30T23:59:59.000Z

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

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

    SciTech Connect (OSTI)

    Daniel M. Wachs

    2012-12-01T23:59:59.000Z

    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 .

  12. On the analysis method of effective delayed neutron fraction at thermal neutron systems

    SciTech Connect (OSTI)

    Nakajima, K.; Unesaki, H. [Research Reactor Inst., Kyoto Univ., Asashiro-Nishi 2, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan)

    2006-07-01T23:59:59.000Z

    The effective delayed neutron fraction (beta-effective) was numerically analyzed with different analysis methods, and their effects on the results were investigated. The cores investigated in this study were light-water moderated low enriched UO{sub 2} lattices, of which the beta-effective had been reported. The effects of transport/diffusion calculation, energy group collapsing, and change of nuclear data library were studied. The study showed that the diffusion calculation with coarse group cross section gave smaller beta-effective than the transport one with fine group cross section, although the difference was not so large, about 2%. On the other hand, the change of nuclear data library from JENDL-3.3 to ENDF/B-VI.8 gave a significant difference, over than 4%. In comparisons with the experiments, it was indicated that the delayed neutron data in JENDL-3.3 are more reliable than those in ENDF/B-VI.8. (authors)

  13. 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)

    2013-07-01T23:59:59.000Z

    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)

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

    E-Print Network [OSTI]

    Igor Iosilevskiy; Victor Gryaznov

    2010-05-23T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Campbell, Kate M.; Davis, J. A.; Bargar, John R.; Giammar, Daniel E.; Bernier-Latmani, Rizlan; Kukkadapu, Ravi K.; Williams, K. H.; Veramani, H.; Ulrich, Kai-Uwe; Stubbs, J. B.; Yabusaki, Steven B.; Figueroa, Linda A.; Lesher, Emily; Wilkins, Michael J.; Peacock, Aaron D.; Longg, P. E.

    2011-03-26T23:59:59.000Z

    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 (UO2(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.

  16. 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

    2014-08-31T23:59:59.000Z

    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.

  17. Fully Ceramic Microencapsulated Fuel Development for LWR Applications

    SciTech Connect (OSTI)

    Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Terrani, Kurt A [ORNL; Voit, Stewart L [ORNL

    2012-01-01T23:59:59.000Z

    The concept, fabrication, and key feasibility issues of a new fuel form based on the microencapsulated (TRISO-type) fuel which has been specifically engineered for LWR application and compacted within a SiC matrix will be presented. This fuel, the so-called fully ceramic microencapsulated fuel is currently undergoing development as an accident tolerant fuel for potential UO2 replacement in commercial LWRs. While the ability of this fuel to facilitate normal LWR cycle performance is an ongoing effort within the program, this will not be a focus of this paper. Rather, key feasibility and performance aspects of the fuel will be presented including the ability to fabricate a LWR-specific TRISO, the need for and route to a high thermal conductivity and fully dense matrix that contains neutron poisons, and the performance of that matrix under irradiation and the interaction of the fuel with commercial zircaloy clad.

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

    SciTech Connect (OSTI)

    Schoendorff, George

    2012-04-02T23:59:59.000Z

    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.

  19. The PACSAT Communications Experiment (PCE)

    SciTech Connect (OSTI)

    Not Available

    1993-02-12T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1993-02-12T23:59:59.000Z

    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.

  1. Nuclear fuels for very high temperature applications

    SciTech Connect (OSTI)

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

    1992-08-01T23:59:59.000Z

    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.

  2. 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))

    1991-01-10T23:59:59.000Z

    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.

  3. Nuclear fuels for very high temperature applications

    SciTech Connect (OSTI)

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

    1992-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    David Petti

    2014-06-01T23:59:59.000Z

    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.

  5. 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.

    1980-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1994-01-01T23:59:59.000Z

    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.

  7. Action Sheet 36 Final Report

    SciTech Connect (OSTI)

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

    2012-02-24T23:59:59.000Z

    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.

  8. 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)

    2014-07-01T23:59:59.000Z

    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.

  9. The Hanford Site: An anthology of early histories

    SciTech Connect (OSTI)

    Gerber, M.S.

    1993-10-01T23:59:59.000Z

    This report discusses the following topics: Memories of War: Pearl Harbor and the Genesis of the Hanford Site; safety has always been promoted at the Hanford Site; women have an important place in Hanford Site history; the boom and bust cycle: A 50-year historical overview of the economic impacts of Hanford Site Operations on the Tri-Cities, Washington; Hanford`s early reactors were crucial to the sites`s history; T-Plant made chemical engineering history; the UO{sub 3} plant has a long history of service. PUREX Plant: the Hanford Site`s Historic Workhorse. PUREX Plant Waste Management was a complex challenge; and early Hanford Site codes and jargon.

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

    SciTech Connect (OSTI)

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

    2011-10-01T23:59:59.000Z

    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.

  11. Research Programme for the 660 Mev Proton Accelerator Driven MOX-Plutonium Subcritical Assembly

    E-Print Network [OSTI]

    Barashenkov, V S; Buttseva, G L; Dudarev, S Yu; Polanski, A; Puzynin, I V; Sissakian, A N

    2000-01-01T23:59:59.000Z

    The paper presents a research programme of the Experimental Acclerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton acceletator operating at the Laboratory of Nuclear Problems of the JINR, Dubna. MOX fuel (25% PuO_2 + 75% UO_2) designed for the BN-600 reactor use will be adopted for the core of the assembly. The present conceptual design of the experimental subcritical assembly is based on a core of a nominal unit capacity of 15 kW (thermal). This corresponds to the multiplication coefficient k_eff = 0.945, energetic gain G = 30 and the accelerator beam power 0.5 kW.

  12. MOX Reprocessing at Tokai Reprocessing Plant

    SciTech Connect (OSTI)

    Taguchi, Katsuya; Nagaoka, Shinichi; Yamanaka, Atsushi; Nakamura, Yoshinobu; Omori, Eiichi [Tokai Reprocessing Technology Development Center, Japan Atomic Energy Agency 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan); SATO, Takehiko; MIURA, Nobuyuki [Nuclear Fuel Cycle Technology Development Directorate, Japan Atomic Energy Agency 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)

    2007-07-01T23:59:59.000Z

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

  13. 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

    2013-09-01T23:59:59.000Z

    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.

  14. 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

    2012-05-15T23:59:59.000Z

    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.

  15. The market for beef among commercial eating establishments

    E-Print Network [OSTI]

    Larey, Claudie Eugene

    1962-01-01T23:59:59.000Z

    then sh vn ths "ulcc?zuo ci 'uh"-' r. oEx run'8 . :Rjrek~x&'ui13ig difx . ::-&t Ue Gs beef Crades, ' The buyer was aoherl 'uo ri conc, the c?us they p ?;=' or d irc'~ thee lrioturea end giVC reaanna fOr hie ChaiCe. ESm -. ~ctnua Was laioaled with the as...;ni. , is -. rte SSle 0, renlo ie nst the Orirn~- Dusine 0, - 'r. ?:, 'ir. - =:;='5: ' i30:. Ri dining roonec is;. =rtnont onti vsriety store restsorsnts, ctxr, . t i 0 . t I-. teria lt. n-d, son. ttes sn well se ciioing oogz of "m'oos t q:" s 0~ olobn. Xrlitlr...

  16. Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01T23:59:59.000Z

    The preceding paper described designs and analyses of thermionic reactors employing full-core-length single-cell converters with their heated emitters located on the outside of their internally cooled collectors, and it presented results of detailed parametric analyses which illustrate the benefits of this unconventional design. The present paper describes the fabrication and testing of full-length prototypical converters, both unfueled and fueled, and presents parametric results of electrically heated tests. The unfueled converter tests demonstrated the practicality of operating such long converters without shorting across a 0.3-mm interelectrode gap. They produced a measured peak output of 751 watts(e) from a single diode and a peak efficiency of 15.4%. The fueled converter tests measured the parametric performance of prototypic UO(subscript 2)-fueled converters designed for subsequent in-pile testing. They employed revolver-shaped tungsten elements with a central emitter hole surrounded by six fuel chambers. The full-length converters were heated by a water-cooled RF-induction coil inside an ion-pumped vacuum chamber. This required development of high-vacuum coaxial RF feedthroughs. In-pile test rules required multiple containment of the UO (subscript 2)-fuel, which complicated the fabrication of the test article and required successful development of techniques for welding tungsten and other refractory components. The test measured a peak power output of 530 watts(e) or 7.1 watts/cm (superscript 2) at an efficiency of 11.5%. There are three copies in the file. Cross-Reference a copy FSC-ESD-217-94-529 in the ESD files with a CID #8574.

  17. Thermodynamic stabilities of U(VI) minerals: Estimated and observed relationships

    SciTech Connect (OSTI)

    Finch, R.J. [Univ. of Manitoba, Winnipeg, Manitoba (Canada)

    1996-12-31T23:59:59.000Z

    Gibbs free energies of formation ({Delta}G{degree}{sub f}) for several structurally related U(VI) minerals are estimated by summing the Gibbs energy contributions from component oxides. The estimated {Delta}G{degree}{sub f} values are used to construct activity-activity (stability) diagrams, and the predicted stability fields are compared with observed mineral occurrences and reaction pathways. With some exceptions, natural occurrences agree well with the mineral stability fields estimated for the systems SiO{sub 2}-CaO-UO{sub 3}-H{sub 2}O and CO{sub 2}-CaO-UO{sub 3}H{sub 2}O, providing confidence in the estimated thermodynamic values. Activity-activity diagrams are sensitive to small differences in {Delta}G{degree}{sub f} values, and mineral compositions must be known accurately, including structurally bound H{sub 2}O. The estimated {Delta}G{degree}{sub f} values are not considered reliable for a few minerals for two major reasons: (1) the structures of the minerals in question are not closely similar to those used to estimate the {Delta}G{sub f}* values of the component oxides, and/or (2) the minerals in question are exceptionally fine grained, leading to large surface energies that increase the effective mineral solubilities. The thermodynamic stabilities of uranium(VI) minerals are of interest for understanding the role of these minerals in controlling uranium concentrations in oxidizing groundwaters associated with uranium ore bodies, uranium mining and mill tailings and geological repositories for nuclear waste.

  18. 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)

    2014-04-21T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2012-10-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2009-06-22T23:59:59.000Z

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

  1. Fabrication Characteristics of Large Grain DUPIC Fuel Using SIMFUEL

    SciTech Connect (OSTI)

    Park, Geun IL; Lee, Jung Won; Lee, Jae Won; Yang, Myung Seung; Song, Kee Chan [Korea Atomic Energy Research Institute, 150-1 Duckjin-Dong, 1045 Daedeokdaero, Yuseong, Daejeon Korea, 305-353 (Korea, Republic of)

    2007-07-01T23:59:59.000Z

    Fabrication characteristics to improve the density and grain size of DUPIC fuel with relation to its fuel performance were experimentally evaluated using SIMFUEL as a surrogate for an actual spent PWR fuel due to the high radioactivity of a spent fuel. Hence, SIMFUELs with a burn-up of 35,000 MWd/tU and 60,000 MWd/tU were used to investigate the influence of fission products contents as an impurity on the fuel powder properties and on the density and grain size of a simulated DUPIC pellet. In order to improve the densification and grain growth of the simulated DUPIC fuel, the effect of the addition of sintering aids was investigated. The specific surface area of the OREOX powders was increased with an increase of the impurities by the dissolved oxides in UO{sub 2} among the impurity groups. The specific surface area of the powders milled after the OXREOX treatment was slightly higher than the UO{sub 2} powder used for a nuclear power plant, thus resulting in sintered pellets with a higher than 95% T.D. (theoretical density). The grain size of the sintered pellets was significantly decreased with increasing amount of the metallic and oxide precipitates. However, on adding the sintering aids such as TiO{sub 2} and Nb{sub 2}O{sub 5}, the grain size of the sintering aids-doped pellets was greatly improved by up to around 3 times that of the raw pellets and their sintered density was also increased by up to 2%. (authors)

  2. 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

    1998-08-01T23:59:59.000Z

    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.

  3. REACTIVITY INITIATED ACCIDENT TEST SERIES TEST RIA 1-4 EXPERIMENT PREDICTIONS

    SciTech Connect (OSTI)

    Fukuda, S. K.; Martinson, Z. R.

    1980-02-01T23:59:59.000Z

    The results of the pretest analyses for Test RIA 1-4 are presented. Test RIA 1-4 consists of a 3x3 array of previously irradiated MAP! fuel rods. The rods have 5.7% enriched UO{sub 2} fuel in zircaloy-4 cladding with an average burnup of 5300 MWd/t. The objective for Test RIA 1-4 is to provide information regarding loss-of-coolable fuel rod geometry following RIA event for a radial-average peak fuel enthalpy equivalent to the present licensing criteria of 1172 J/g (280 cal/g UO{sub 2}). Radial averaged peak fuel enthalpies of 1172 J/g (280 cal/g) 1077 J/g {257 cal/g), and 978 J/g (234 cal/g) for the corner, side, and center fuel rods, respectively, are planned to be achieved during a 2.7 ms reactor period power burst. The results of the FRAP-T5 analyses indicate that all nine rods will fail within 26 ms from the start of the power burst due to pellet-cladding mechanical interaction. All of the rods will undergo partial fuel melting. All rods will operate under extended film boiling (>30 sec) conditions and about 70% of the cladding length is expected to be molten. Approximately 15% of the cladding thickness will be oxided. Fuel swelling due to fission gas release and melting combined with fuel and cladding fragmentation, will probably produce a complete coolant flow blockage within the flow shroud.

  4. Syntheses and structures of three f-element selenite/hydroselenite compounds

    SciTech Connect (OSTI)

    Burns, Wendy L. [Department of Chemistry, Northwestern University, 2145 Sheridan Road, Evanston, IL 60208-3113 (United States); Ibers, James A., E-mail: ibers@chem.northwestern.ed [Department of Chemistry, Northwestern University, 2145 Sheridan Road, Evanston, IL 60208-3113 (United States)

    2009-06-15T23:59:59.000Z

    The selenite/hydroselenite compounds Ce(SeO{sub 3})(HSeO{sub 3}), Tb(SeO{sub 3})(HSeO{sub 3}).2H{sub 2}O, and Cs[U(SeO{sub 3})(HSeO{sub 3})].3H{sub 2}O were synthesized by hydrothermal means at 453 K from the reaction of CeO{sub 2} or Tb{sub 4}O{sub 7} or UO{sub 2} with SeO{sub 2} and CsCl (as a mineralizer). Ce(SeO{sub 3})(HSeO{sub 3}) crystallizes in the non-centrosymmetric orthorhombic space group Pca2{sub 1}. The structure comprises a two-dimensional network of interconnected CeO{sub 10} bicapped distorted square antiprisms and SeO{sub 3} trigonal pyramids. Tb(SeO{sub 3})(HSeO{sub 3}).2H{sub 2}O crystallizes in the non-centrosymmetric orthorhombic space group P2{sub 1}2{sub 1}2{sub 1}. The structure features a two-dimensional layer of interconnected TbO{sub 8} distorted square antiprisms and SeO{sub 3} trigonal pyramids. Cs[U(SeO{sub 3})(HSeO{sub 3})].3H{sub 2}O crystallizes in the centrosymmetric monoclinic space group P2{sub 1}/n. The structure consists of two-dimensional layers of interconnected UO{sub 7} pentagonal bipyramids and SeO{sub 3} trigonal pyramids. The layers in all three structures are held together by hydrogen-bonding networks. - Graphical abstract: Structure of Ce[U(SeO{sub 3})(HSeO{sub 3})].3H{sub 2}O (Cs, purple; U, black; Se, blue; O, red; O{sub w}, green; H, gray).

  5. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-07-25T23:59:59.000Z

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on {approx}0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford ''New Production Reactor'', later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory's (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in Plutonium-239 concentration from about 97 to 99 atom%.

  6. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-09-15T23:59:59.000Z

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on ~0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford “New Production Reactor,” later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory’s (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in Plutonium-239 concentration from about 97 to 99 atom%.

  7. Density Functional Theory Study of the Complexation of the Uranyl Dication with Anionic Phosphate Ligands with and without Water Molecules

    SciTech Connect (OSTI)

    Jackson, Virgil E.; Gutowski, Keith E.; Dixon, David A.

    2013-09-12T23:59:59.000Z

    The structures, vibrational frequencies and energetics of anhydrous and hydrated complexes of UO2 2+ with the phosphate anions H2PO4 ?, HPO4 2?, and PO4 3? were predicted at the density functional theory (DFT) and MP2 molecular orbital theory levels as isolated gas phase species and in aqueous solution by using self-consistent reaction field (SCRF) calculations with different solvation models. The geometries and vibrational frequencies of the major binding modes for these complexes are compared to experiment where possible and good agreement is found. The uranyl moiety is nonlinear in many of the complexes, and the coordination number (CN) 5 in the equatorial plane is the predominant binding motif. The phosphates are found to bind in both monodentate and bidentate binding modes depending on the charge and the number of water molecules. The SCRF calculations were done with a variety of approaches, and different SCRF approaches were found to be optimal for different reaction types. The acidities of HxPO4 3?x in HxPO4 3?x(H2O)4, x = 0?3 complexes were calculated with different SCRF models and compared to experiment. Phosphate anions can displace water molecules from the first solvation shell at the uranyl exothermically. The addition of water molecules can cause the bonding of H2PO4 ? and HPO4 2? to change from bidentate to monodentate exothermically while maintaining CN 5. The addition of water can generate monodentate structures capable of cross-linking to other uranyl phosphates to form the types of structures found in the solid state. [UO2(HPO4)(H2O)3] is predicted to be a strong base in the gas phase and in aqueous solution. It is predicted to be a much weaker acid than H3PO4 in the gas phase and in solution.

  8. 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

    2012-04-01T23:59:59.000Z

    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.

  9. THERMODYNAMIC MODEL FOR URANIUM DIOXIDE BASED NUCLEAR FUEL

    SciTech Connect (OSTI)

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

    2007-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Lewis, B.E.

    1984-09-01T23:59:59.000Z

    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.

  11. 200-UP-2 Operable Unit technical baseline report

    SciTech Connect (OSTI)

    Deford, D.H.

    1991-02-01T23:59:59.000Z

    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.

  12. Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

    SciTech Connect (OSTI)

    Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Boer, B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); SCK-CEN, Boertang 200, BE-2400 Mol (Belgium)

    2012-07-01T23:59:59.000Z

    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 retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

  13. Neutronic Analysis of the Burning of Transuranics in Fully Ceramic Micro-Encapsulated Tri-Isotropic Particle-Fuel in a PWR

    SciTech Connect (OSTI)

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

    2012-11-01T23:59:59.000Z

    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 UO2 and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO2 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 is 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.

  14. 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

    2012-04-01T23:59:59.000Z

    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.

  15. 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)

    2006-07-01T23:59:59.000Z

    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)

  16. 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.

    1997-01-01T23:59:59.000Z

    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.

  17. Closed ThUOX Fuel Cycle for LWRs with ADTT (ATW) Backend for the 21st Century

    SciTech Connect (OSTI)

    Beller, D.E.; Sailor, W.C.; Venneri, F.

    1998-10-06T23:59:59.000Z

    A future nuclear energy scenario with a closed, thorium-uranium-oxide (ThUOX) fuel cycle and new light water reactors (TULWRs) supported by Accelerator Transmutation of Waste (ATW) systems could provide several improvements beyond today's once-through, UO{sub 2}-fueled nuclear technology. A deployment scenario with TULWRs plus ATWs to burn the actinides produced by these LWRs and to close the back-end of the ThUOX fuel cycle was modeled to satisfy a US demand that increases linearly from 80 GWe in 2020 to 200 GWe by 2100. During the first 20 years of the scenario (2000-2020), nuclear energy production in the US declines from today's 100 GWe to about 80 GWe, in accordance with forecasts of the US DOE's Energy Information Administration. No new nuclear systems are added during this declining nuclear energy period, and all existing LWRs are shut down by 2045. Beginning in 2020, ATWs that transmute the actinides from existing LWRs are deployed, along with TULWRs and additional ATWs with a support ratio of 1 ATW to 7 TULWRs to meet the energy demand scenario. A final mix of 174 GWe from TULWRs and 26 GWe from ATWs provides the 200 GWe demand in 2100. Compared to a once-through LWR scenario that meets the same energy demand, the TULWR/ATW concept could result in the following improvements: depletion of natural uranium resources would be reduced by 50%; inventories of Pu which may result in weapons proliferation will be reduced in quantity by more than 98% and in quality because of higher neutron emissions and 50 times the alpha-decay heating of weapons-grade plutonium; actinides (and possibly fission products) for final disposal in nuclear waste would be substantially reduced; and the cost of fuel and the fuel cycle may be 20-30% less than the once-through UO{sub 2} fuel cycle.

  18. Purification Testing for HEU Blend Program

    SciTech Connect (OSTI)

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

    1998-06-01T23:59:59.000Z

    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.

  19. 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)

    1991-12-31T23:59:59.000Z

    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.

  20. Study of Chemical Changes in Uranium Oxyfluoride Particles Progress Report June 2008 - February 2009

    SciTech Connect (OSTI)

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

    2009-02-25T23:59:59.000Z

    The present study aims to demonstrate how knowledge of time-dependent changes in uranium oxyfluoride particles can benefit particle analyses for environmental sampling. Environmental sampling depends upon laboratory analysis of nuclear material that has often been exposed to the environment after it was produced. It is therefore important to understand how those environmental conditions might have changed the chemical composition of the material over time. To investigate this, we prepared a set of uranium oxyfluoride particles at the Institute for Reference Materials and Measurements (IRMM-DG Joint Research Centre of the European Commission, Belgium). These UO{sub 2}F{sub 2} particles were prepared from the release and subsequent hydrolysis of UF{sub 6} gas, and were stored at LLNL in environmental chambers, set to different humidity, temperature and lighting conditions. An experimental plan was drafted to assess the number of analyses required to track the changes in particle composition, morphology, and structure. Due to its high spatial resolution and excellent transmission, the NanoSIMS secondary ion mass spectrometer at LLNL was found to be the optimal tool to measure individual oxyfluoride particles. This was confirmed by our participation in the inter-laboratory measurement campaign for particle analysis (NUSIMEP-6), organized by the IRMM in June last year. The reported uranium isotope ratios demonstrated the precision and accuracy of the NanoSIMS and ims 3f SIMS measurements at LLNL, and provided a high degree of confidence that the new measurements on the UO{sub 2}F{sub 2} samples will be of comparable high quality. As fluorine is known to be a chemically-sensitive compound, we measured the intensity of the fluorine secondary ions relative to the ions generated by the matrix to evaluate the rate of particle degradation under different environmental conditions. A relative sensitivity factor was empirically determined to convert these measurements to absolute fluorine concentrations. Additional measurements in selected uranium compounds were carried out to account for variations in matrix composition. Because of the complexity of both the SIMS instruments, as well as the nature of the samples, we spent a substantial amount of time on instrument training and instrument set up. The latest NanoSIMS measurements on the freshly-prepared UO{sub 2}F{sub 2} particles however, showed that we are on the right track when it comes to determining the chemical changes in individual uranium particles. At PNNL, several optical techniques including cryogenic laser-induced time-resolved U(VI) fluorescence micro-spectroscopy, Raman spectroscopy, and Fourier transform infrared spectroscopy will be applied to investigate molecular transformations of the particles. As a control, dynamic SIMS measurements will also be performed on a subset of the samples sent to PNNL.

  1. 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

    2003-08-01T23:59:59.000Z

    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.

  2. 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

    2011-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1980-01-01T23:59:59.000Z

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

    SciTech Connect (OSTI)

    R. Finch

    2001-06-05T23:59:59.000Z

    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.

  5. 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.

    2010-01-29T23:59:59.000Z

    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.

  6. Student Progress Report: Summer 2012

    SciTech Connect (OSTI)

    Tucker, Lucas P [Los Alamos National Laboratory

    2012-08-06T23:59:59.000Z

    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. Drying results of K-Basin fuel element 3128W (run 2)

    SciTech Connect (OSTI)

    Abrefah, J.; Klinger, G.S.; Oliver, B.M.; Marshman, S.C.; MacFarlan, P.J.; Ritter, G.A. [Pacific Northwest National Lab., Richland, WA (United States); Flament, T.A. [Numatec Hanford Corp., Richland, WA (United States)

    1998-07-01T23:59:59.000Z

    An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-East Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of N-Reactor spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from an open K-East canister (3128W) during the first fuel selection campaign conducted in 1995, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. Although it was judged to be breached during in-basin (i.e., K-Basin) examinations, visual inspection of this fuel element in the hot cell indicated that it was likely intact. Some scratches on the coating covering the cladding were identified before the furnace test. The drying test was conducted in the Whole Element Furnace Testing System located in G-Cell within the PTL. This test system is composed of three basic systems: the in-cell furnace equipment, the system gas loop, and the analytical instrument package. Element 3128W was subjected to the drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step. Results of the Pressure Rise and Gas Evolution Tests suggest that most of the free water in the system was released during the extended CVD cycle (68 hr versus 8 hr for the first run). An additional {approximately}0.34 g of water was released during the subsequent HVD phase, characterized by multiple water release peaks, with a principle peak at {approximately}180 C. This additional water is attributed to decomposition of a uranium hydrate (UO{sub 4}{center_dot}4H{sub 2}O/UO{sub 4}{center_dot}2H{sub 2}O) coating that was observed to be covering the surface of the fuel element to a thickness of {approximately}1.6 mg/cm{sup 2}. A limited quantity of hydrogen ({approximately}9 mg) was also released during HVD, mainly at temperatures above 300 C, likely from hydride decomposition.

  8. Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

    SciTech Connect (OSTI)

    Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

    2004-12-14T23:59:59.000Z

    Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows for surface alloying well in excess of the thermodynamically dictated solubility limits, an effect that is particularly relevant to this research due to the negligible solubility of boron and gadolinium in zirconium. University of Wisconsin is performing the near surface materials characterization and analysis, aiding Sandia in process optimization, and promoting educational activities. Westinghouse is performing process manufacturability and scale-up analysis and is performing autoclave testing of the surface treated samples. The duration of this NERI project is 2 years, from 9/2002 to 9/2004.

  9. TRIMOLECULAR REACTIONS OF URANIUM HEXAFLUORIDE WITH WATER

    SciTech Connect (OSTI)

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

    2010-02-25T23:59:59.000Z

    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.

  10. 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.

    1998-05-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    S. Bays; H. Zhang; M. Pope

    2009-05-01T23:59:59.000Z

    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.

  12. EDF Nuclear Power Plants Operating Experience with MOX fuel

    SciTech Connect (OSTI)

    Thibault, Xavier [EDF Generation, Tour EDF Part Dieu - 9 rue des Cuirassiers B.P.3181 - 69402 Lyon Cedex 03 (France)

    2006-07-01T23:59:59.000Z

    EDF started Plutonium recycling in PWR in 1987 and progressively all the 20 reactors, licensed in using MOX fuel, have been loaded with MOX assemblies. At the origin of MOX introduction, these plants operated at full power in base load and the core management limited the irradiation time of MOX fuel assemblies to 3 annual cycles. Since 1995 all these reactors can operate in load follow mode. Since that time, a large amount of experience has been accumulated. This experience is very positive considering: - Receipt, handling, in core behaviour, pool storage and shipment of MOX fuel; - Operation of the various systems of the plant; - Environment impact; - Radioprotection; - Safety file requirements; - Availability for the grid. In order to reduce the fuel cost and to reach a better adequacy between UO{sub 2} fuel reprocessing flow and plutonium consumption, EDF had decided to improve the core management of MOX plants. This new core management call 'MOX Parity' achieves parity for MOX and UO{sub 2} assemblies in term of discharge burn-up. Compared to the current MOX assembly the Plutonium content is increased from 7,08% to 8,65% (equivalent to natural uranium enriched to respectively 3,25% and 3,7%) and the maximum MOX assembly burn-up moves from 42 to 52 GWd/t. This amount of burn-up is obtained from loading MOX assemblies for one additional annual cycle. Some, but limited, adaptations of the plant are necessary. In addition a new MOX fuel assembly has been designed to comply with the safety criteria taking into account the core management performances. These design improvements are based on the results of an important R and D program including numerous experimental tests and post-irradiated fuel examinations. In particular, envelope conditions compared to MOX Parity neutronic solicitations has been extensively investigated in order to get a full knowledge of the in reactor fuel behavior. Moreover, the operating conditions of the plant have been evaluated in many details and finally no important impact is anticipated. The industrial maturity of plutonium recycling activities is fully demonstrated and a new progress can be done with a complete confidence. The licensing process of 'MOX Parity' core management is in progress and its implementation on the 20 PWR is now expected at mid 2007. (author)

  13. Experimental studies and thermodynamic modelling of volatilities of uranium, plutonium, and americium from their oxides and from their oxides interacted with ash

    SciTech Connect (OSTI)

    Krikorian, O.H.; Ebbinghaus, B.B.; Adamson, M.G.; Fontes, A.S. Jr.; Fleming, D.L.

    1993-09-15T23:59:59.000Z

    The purpose of this study is to identify the types and amounts of volatile gaseous species of U, Pu, and Am that are produced in the combustion chamber offgases of mixed waste oxidation processors. Primary emphasis is on the Rocky Flats Plant Fluidized Bed Incinerator. Transpiration experiments have been carried out on U{sub 3}O{sub 8}(s), U{sub 3}O{sub 8} interacted with various ash materials, PuO{sub 2}(s), PuO{sub 2} interacted with ash materials, and a 3%PuO{sub 2}/0.06%AmO{sub 2}/ash material, all in the presence of steam and oxygen, and at temperatures in the vicinity of 1,300 K. UO{sub 3}(g) and UO{sub 2}(OH){sub 2}(g) have been identified as the uranium volatile species and thermodynamic data established for them. Pu and Am are found to have very low volatilities, and carryover of Pu and Am as fine dust particulates is found to dominate over vapor transport. The authors are able to set upper limits on Pu and Am volatilities. Very little lowering of U volatility is found for U{sub 3}O{sub 8} interacted with typical ashes. However, ashes high in Na{sub 2}O (6.4 wt %) or in CaO (25 wt %) showed about an order of magnitude reduction in U volatility. Thermodynamic modeling studies were carried out that show that for aluminosilicate ash materials, it is the presence of group I and group II oxides that reduces the activity of the actinide oxides. K{sub 2}O is the most effective followed by Na{sub 2}O and CaO for common ash constituents. A more major effect in actinide activity lowering could be achieved by adding excess group I or group II oxides to exceed their interaction with the ash and lead to direct formation of alkali or alkaline earth uranates, plutonates, and americates.

  14. Updated NGNP Fuel Acquisition Strategy

    SciTech Connect (OSTI)

    David Petti; Tim Abram; Richard Hobbins; Jim Kendall

    2010-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Gilles Youinou; R. Sonat Sen

    2013-09-01T23:59:59.000Z

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

  16. A study of the relationship between recreational user-day visits and the physical and economic characteristics of Texas water impoundment areas

    E-Print Network [OSTI]

    Jones, Ronnie D

    1966-01-01T23:59:59.000Z

    ILTsy ~ QLLkeC %ltd'708 XW$@~p4. Che . ccQKSPgetds Gx KBBQp' @?oee ~ +~ 3. sx'gN ~S the RgK'9. ~ EtQS t ~ e&G, %@GC G BtM e VBAhLXX P, QQ'O~~~V~s dNXV A@ NQ 84 )J? p I& Fig. 1. --Location of Standard Metropolitan Statistical Areas of' Texas. 6 g UO...EigoX'iG~44 QQE9pSZ'2, Bono 9 'C&MQ 8'RSCG VMS ZMW$. C4'G8, 3. n@Q 8& Z QX'8@8 y Zn QQ88. ggfn CLn~~ '5hn ~&on& pnztl. e+3e~ . @Ments. on +@8 gLvav. 'ao eneh Zeeitme ea t@W'8'Q~l' y MS' ~~i@f'@3. 3?e @ho 64kkn '@8K'5 +G3. 38$'58d; Ltp' CRS sGLL Gonso...

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

    SciTech Connect (OSTI)

    Jason Hales; Various

    2014-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Higginson, M.C.

    1994-10-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Lloyd, R.

    1980-10-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Ikeuchi, Hirotomo; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Kondo, Yoshikazu; Noguchi, Yoshikazu [PESCO Co.Ltd. (Korea, Republic of)

    2013-07-01T23:59:59.000Z

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

  2. 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. [Los Alamos National Lab., NM (United States). Safeguards Science and Technology Group] [and others

    1996-09-01T23:59:59.000Z

    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.

  3. 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

    2014-10-01T23:59:59.000Z

    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.

  4. Incorporation of radionuclides in the alteration phases of spent nuclear fuel.

    SciTech Connect (OSTI)

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

    1999-08-25T23:59:59.000Z

    Alteration may be expected for spent nuclear fuel exposed to groundwater under oxidizing conditions such as that which exist at the proposed nuclear waste repository at Yucca Mountain, Nevada. The actinide elements released during the corrosion of spent fuel may be incorporated into the structures of secondary U{sup 6+} phases. The incorporation of transuranics into the crystal structures of the alteration products may significantly decrease their mobility. A series of precipitation tests were conducted at 90 C to determine the potential incorporation of Ce{sup 4+} and Nd{sup 3+} (surrogates for Pu{sup 4+} and Am{sup 3+}, respectively) into uranyl phase. Dehydrated schoepite (UO{sub 3}{center_dot}0.8-1.0HP{sub 2}O) was produced by hydrolysis of a uranium oxyacetate solution containing either cerium or neodymium. ICP-MS analysis of the leachant, leachate, and solid phase reaction products which were dissolved in a HNO{sub 3} solution indicates that 26 ppm of Ce was incorporated into dehydrated schoepite. ICP-MS results from the Nd-doped tests indicate significant neodymium incorporation as well, however, the heterogeneous distribution of Nd in the solid phase noted during the AEM/EELS examination implies that neodymium may not incorporate into the structure of dehydrated schoepite.

  5. Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

    SciTech Connect (OSTI)

    Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

    1993-01-01T23:59:59.000Z

    The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO{sub 2} pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel.

  6. Structural model of uramarsite

    SciTech Connect (OSTI)

    Rastsvetaeva, R. K., E-mail: rast@ns.crys.ras.ru [Russian Academy of Sciences, Shubnikov Institute of Crystallography (Russian Federation); Sidorenko, G. A. [All-Russia Research Institute of Mineral Resources (VIMS) (Russian Federation); Ivanova, A. G. [Russian Academy of Sciences, Shubnikov Institute of Crystallography (Russian Federation); Chukanov, N. V. [Russian Academy of Sciences, Institute of Problems of Chemical Physics (Russian Federation)

    2008-09-15T23:59:59.000Z

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

  7. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    SciTech Connect (OSTI)

    Perret, G.; Pattupara, R. M. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Girardin, G. [Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland); Chawla, R. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland)

    2012-07-01T23:59:59.000Z

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

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

    SciTech Connect (OSTI)

    Franceschini, F.; Lahoda, E.; Wenner, M. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)] [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Lindley, B. [University of Cambridge (United Kingdom)] [University of Cambridge (United Kingdom); Fiorina, C. [Polytechnic of Milan (Italy)] [Polytechnic of Milan (Italy); Phillips, C. [Energy Solutions, Richland, WA (United States)] [Energy Solutions, Richland, WA (United States)

    2013-07-01T23:59:59.000Z

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

  9. 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

    2010-01-01T23:59:59.000Z

    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 ...

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

    SciTech Connect (OSTI)

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

    1991-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1992-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1992-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Not Available

    1986-06-01T23:59:59.000Z

    The uranium hexafluoride (UF/sub 6/) release of January 4, 1986, at the Sequoyah Fuels Corporation facility has been reviewed by a NRC Lessons-Learned Group. A Model 48Y cylinder containing UF/sub 6/ ruptured upon being heated after it was grossly overfilled. The Uf/sub 6/ released upon rupture of the cylinder reacted with airborne moisture to produce hydrofluoric acid (HF) and uranyl fluoride (UO/sub 2/F/sub 2/). One individual died from exposure to airborne HF and several others were injured. There were no significant immediate effects from exposure to uranyl fluoride. This report of the Lessons-Learned Group presents discussions and recommendations on the process, operation and design of the facility, as well as on the responses of the licensee, NRC, and other local, state and federal agencies to the incident. It also provides recommendations in the areas of NRC licensing and inspection of fuel facility and certain other NMSS licensees. The implementation of some recommendations will depend on decisions to be made regarding the scope of NRC responsibilities with respect to those aspects of the design and operation of such facilities that are not directly related to radiological safety.

  14. Uranyl fluoride luminescence in acidic aqueous solutions

    SciTech Connect (OSTI)

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

    1996-08-01T23:59:59.000Z

    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}.

  15. Assessing the Role of Iron Sulfides in the Long Term Sequestration of Uranium by Sulfate-Reducing Bacteria

    SciTech Connect (OSTI)

    Hayes, Kim F.; Bi, Yuqiang; Carpenter, Julian; Hyng, Sung Pil; Rittmann, Bruce E.; Zhou, Chen; Vannela, Raveender; Davis, James A.

    2014-01-01T23:59:59.000Z

    This overarching aim of this project was to identify the role of biogenic and synthetic iron-sulfide minerals in the long-term sequestration of reduced U(IV) formed under sulfate-reducing conditions when subjected to re-oxidizing conditions. The work reported herein was achieved through the collaborative research effort conducted at Arizona State University (ASU) and the University of Michigan (UM). Research at ASU, focused on the biogenesis aspects, examined the biogeochemical bases for iron-sulfide production by Desulfovibrio vulgaris, a Gram-negative bacterium that is one of the most-studied strains of sulfate-reducing bacteria. A series of experimental studies were performed to investigate comprehensively important metabolic and environmental factors that affect the rates of sulfate reduction and iron-sulfide precipitation, the mineralogical characteristics of the iron sulfides, and how uranium is reduced or co-reduced by D. vulagaris. FeS production studies revealed that controlling the pH affected the growth of D. vulgaris and strongly influenced the formation and growth of FeS solids. In particular, lower pH produced larger-sized mackinawite (Fe1+xS). Greater accumulation of free sulfide, from more sulfate reduction by D. vulgaris, also led to larger-sized mackinawite and stimulated mackinawite transformation to greigite (Fe3S4) when the free sulfide concentration was 29.3 mM. On the other hand, using solid Fe(III) (hydr)oxides as the iron source led to less productivity of FeS due to their slow and incomplete dissolution and scavenging of sulfide. Furthermore, sufficient free Fe2+, particularly during Fe(III) (hydr)oxide reductions, led to the additional formation of vivianite [Fe3(PO4)2•8(H2O)]. The U(VI) reduction studies revealed that D. vulgaris reduced U(VI) fastest when accumulating sulfide from concomitant sulfate reduction, since direct enzymatic and sulfide-based reductions of U(VI) occurred in parallel. The UO2 produced in presence of ferrous iron was poorly crystalline. At UM, laboratory-scale reactor studies were performed to assess the potential for the predominant abiotic reductants formed under sulfate reducing conditions (SRCs) to: (1) reduce U(VI) in contaminated groundwater sediments), and (2) inhibit the re-oxidation of U(IV) species, and in particular, uraninite (UO2(s)). Under SRCs, mackinawite and aqueous sulfide are the key reductants expected to form. To assess their potential for abiotic reduction of U(VI) species, a series of experiments were performed in which either FeS or S(-II) was added to solutions of U(VI), with the rates of conversion to U(IV) solids monitored as a function of pH, and carbonate and calcium concentration. In the presence of FeS and absence of oxygen or carbonate, U(IV) was completely reduced uraninite. S(-II) was also found to be an effective reductant of aqueous phase U(VI) species and produced uraninite, with the kinetics and extent of reduction depending on geochemical conditions. U(VI) reduction to uraninite was faster under higher S(-II) concentrations but was slowed by an increase in the dissolved Ca or carbonate concentration. Rapid reduction of U(VI) occurred at circumneutral pH but virtually no reduction occurred at pH 10.7. In general, dissolved Ca and carbonate slowed abiotic U(VI) reduction by forming stable Ca-U(VI)-carbonate soluble complexes that are resistant to reaction with aqueous sulfide. To investigate the stability of U(IV) against re-oxidation in the presence of iron sulfides by oxidants in simulated groundwater environments, and to develop a mechanistic understanding the controlling redox processes, continuously-mixed batch reactor (CMBR) and flow-through reactor (CMFR) studies were performed at UM. In these studies a series of experiments were conducted under various oxic groundwater conditions to examine the effectiveness of FeS as an oxygen scavenger to retard UO2 dissolution. The results indicate that FeS is an effective oxygen scavenger, and can lower the rate of oxidative dissolution of UO2 by over an order of magnitude compared to

  16. Submersion Criticality Safety Analysis of Tungsten-Based Fuel for Nuclear Power and Propulsion Applications

    SciTech Connect (OSTI)

    A.E. Craft; R. C. O'Brien; S. D. Howe; J. C. King

    2014-07-01T23:59:59.000Z

    The Center for Space Nuclear Research (CSNR) is developing tungsten-encapsulated fuels for space nuclear applications. Aims to develop NTP fuels that are; Affordable Low impact on production and testing environment Producible on a large scale over suitable time period Higher-performance compared to previous graphite NTP fuel elements Space nuclear reactors remain subcritical before and during launch, and do not go critical until required by its mission. A properly designed reactor will remain subcritical in any launch abort scenario, where the reactor falls back to Earth and becomes submerged in terrestrial material. Submersion increases neutron reflection and thermalizes the neutrons, which typically increases the reactivity of the core. This effect is usually very significant for fast-spectrum reactors. This research provided a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor. Determine the submersion behavior of a reactor fueled by tungsten-based fuel. Considered fuel compositions with varying: Rhenium content (wt% rhenium in tungsten) Fuel loading fractions (UO2 vol%)

  17. Preconcentration of uranium in seawater with heterocyclic azo dyes supported on silica gel

    SciTech Connect (OSTI)

    Ueda, K.; Koshino, Y.; Yamamoto, Y.

    1985-11-01T23:59:59.000Z

    The chelating adsorbents, heterocyclic azo dyes supported on silica gel, were prepared and their adsorption behaviors of metal ions were investigated. The 1-(2-pyridylazo)-2-naphthol(PAN)-SG and 2-(2-thiazolylazo)-p-cresol(TAC)-SG show greater affinity for UO/sub 2/(II) and ZrO(II), compared with the other metal ions like Cu, Cd, Fe and alkaline earths. Trace uranyl can be quantitatively retained on the column of the gels at neutral pH region and flowrate 3-4 ml/min. The uranyl retained is easily eluted from the column bed with a mixture of acetone and nitric acid (9:1 v/v) and determined by spectrophotometry using Arsenazo-III. Matrix components in seawater do not interfere and the spiked recovery of uranyl in artificial seawater was found to be average 98.6%, with the relative standard deviation of 1.08%. Both gels were applied to the determination of uranium in seawater with satisfactory results. 16 references, 3 figures, 3 tables.

  18. Helium/solid powder O-ring leakage correlation experiments

    SciTech Connect (OSTI)

    Leisher, W.B.; Weissman, S.H.; Tallant, D.R.; Kubo, M.

    1983-01-01T23:59:59.000Z

    We have developed a method to test powder leakage that has passed O-ring seals. To validate this method we have spiked a test fixture with 98 ng of U and recovered 130 +- 25 ng of U. We did not detect U at a detection limit of 26 ng in a fixture which was treated as a blank. This method has been applied to the leakage of UO/sub 2/ powder passing the type of EPDM O-ring seals used in a SNM shipping cask belonging to PNC. Considering the three experimental tests in which no or very small quantities of U were detected as effective blank test, it appears that the level of external contamination is negligible. Therefore, we believe that the U quantities greater than 26 ng (6 tests) passed the primary O-ring seal. From this limited quantity of data, we observe no apparent correlation between the amount of U measured and either helium leak rate or equivalent tube diameter. The data for the 130/sup 0/C tests indicate the possibility of a U/time relationship; however, more data are needed for verification.

  19. 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.

    2011-06-22T23:59:59.000Z

    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.

  20. 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)

    1994-08-01T23:59:59.000Z

    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.