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1

UoS Motor Accident Report Form COMPANY DETAILS  

E-Print Network [OSTI]

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

Sussex, University of

2

UO Policy Library Resource for  

E-Print Network [OSTI]

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

Oregon, University of

3

Cameco UO3 Materials Analysis  

SciTech Connect (OSTI)

Uranium trioxide (UO{sub 3}) was characterized using a variety of techniques to better understand its physical properties. Scanning electron microscope (SEM) images were collected to examine particle morphology, which consisted of semi-spherical particles that tended to agglomerate before sonication. Particle size analysis revealed a singular mode distribution with a mean particle size of 43.0 {micro}m. After sonication a bimodal distribution was produced with peak particle sizes at 0.226 {micro}m and 9.43 {micro}m. The O/U ratio was measured to be 3.09 by Cameco in 2009, by gravimetric analysis. X-ray diffraction (XRD) showed that the sample was mostly {gamma}-UO{sub 3} (87.1%) with a small amount of UO{sub 3} {center_dot} 0.80 H{sub 2}O (12.9%). Bulk and tap densities were determined to be 3.678 {+-} 0.2 and 4.81 {+-} 0.2 g/cm{sup 3}, respectively (crystalline density is 7.3 g/cm{sup 3}). The stoichiometry was measured to be 2.99 in 2012.

Hill, Mary Ann [Los Alamos National Laboratory; Nolen, Blake Penfield [Los Alamos National Laboratory; Wermer, Joseph R. [Los Alamos National Laboratory; Wilkerson, Marianne P. [Los Alamos National Laboratory; Fredenburg, David A. [Los Alamos National Laboratory; Wagner, Gregory L. [Los Alamos National Laboratory; Papin, Pallas A. [Los Alamos National Laboratory; Scott, Brian L. [Los Alamos National Laboratory; Guidry, Dennis Ray [Los Alamos National Laboratory

2012-07-12T23:59:59.000Z

4

UO  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched Ferromagnetism in Layeredof EnergyLeaseEnergyUNCLASSIFIED 2 1IsotopeFigure 1.

5

OXYGEN DIFFUSION IN UO2-x  

E-Print Network [OSTI]

~ 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.

Kim, K.C.

2013-01-01T23:59:59.000Z

6

Crystal fields in UO2 - revisited  

SciTech Connect (OSTI)

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.

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

7

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...

8

Vendor Control UoW 1730 (Rev. 10/07)  

E-Print Network [OSTI]

Vendor Control Use Only UoW 1730 (Rev. 10/07) ACCOUNTING DETAIL U.S. Taxpayer ID Number 1. Vendor. VENDOR'S CERTIFICATE: I hereby certify that the items and totals listed herein are proper charges

Borenstein, Elhanan

9

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...

10

PUREX/UO{sub 3} deactivation project management plan  

SciTech Connect (OSTI)

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.

Washenfelder, D.J.

1993-12-01T23:59:59.000Z

11

Spark Plasma Sintering of W-UO2 Cermets  

SciTech Connect (OSTI)

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.

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

2013-02-01T23:59:59.000Z

12

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

SciTech Connect (OSTI)

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.

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

2009-11-01T23:59:59.000Z

13

UO{sub 3} plant turnover - facility description document  

SciTech Connect (OSTI)

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.

Clapp, D.A.

1995-01-01T23:59:59.000Z

14

Density Functional Theory Calculations of Mass Transport in UO2  

SciTech Connect (OSTI)

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

Andersson, Anders D. [Los Alamos National Laboratory; Dorado, Boris [CEA; Uberuaga, Blas P. [Los Alamos National Laboratory; Stanek, Christopher R. [Los Alamos National Laboratory

2012-06-26T23:59:59.000Z

15

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

SciTech Connect (OSTI)

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.

John M. Veilleux

1999-08-01T23:59:59.000Z

16

Benchmarking of Graphite Reflected Critical Assemblies of UO2  

SciTech Connect (OSTI)

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

Margaret A. Marshall; John D. Bess

2011-11-01T23:59:59.000Z

17

PUREX/UO3 Facilities deactivation lessons learned history  

SciTech Connect (OSTI)

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

Gerber, M.S.

1996-09-19T23:59:59.000Z

18

DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL  

SciTech Connect (OSTI)

The primary goal of this investigation was to evaluate the effectiveness of the chop-leach process, with nitric acid solvent, to produce a nominally 300 g/L [U] and 1 M [H{sup +}] product solution. The results of this study show that this processing technique is appropriate for applications in which a low free acid and moderately high U content are desired. The 7.75 L of product solution, which was over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H{sup +}] in the range 0.8-1.2 M. A secondary goal was to test the effectiveness of this treatment for the removal of actinides from Zircaloy cladding to produce a low-level radioactive waste (LLW) cladding product. Analysis of the cladding shows that actinides are present in the cladding at a concentration of about 5000 {eta}Ci/g, which is about 50 times greater than the acceptable transuranium element limit in low level radioactive waste. It appears that the concentration of nitric acid used for this dissolution study (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate to completely digest the UO{sub 2} present in the spent fuel. The mass of insoluble material collected from the initial treatments with nitric acid, 340 g, was much higher than expected, and analysis of this insoluble residue showed that it contained at least 200 g U.

Kessinger, G.; Thompson, M.

2009-08-07T23:59:59.000Z

19

Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets  

E-Print Network [OSTI]

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.

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

2014-02-28T23:59:59.000Z

20

Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets  

E-Print Network [OSTI]

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.

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

2014-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

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

E-Print Network [OSTI]

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

Oregon, University of

22

f Fk66iCOP-] NBSIR 86-3422 uoL_ i 1  

E-Print Network [OSTI]

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

Oak Ridge National Laboratory

23

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

SciTech Connect (OSTI)

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.

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

2005-06-13T23:59:59.000Z

24

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...

25

UO Organizational Development and Training HUMAN RESOURCES, ORGANIZATIONAL DEVELOPMENT AND TRAINING  

E-Print Network [OSTI]

UO Organizational Development and Training HUMAN RESOURCES, ORGANIZATIONAL DEVELOPMENT AND TRAINING Organizational Development and Training HUMAN RESOURCES, ORGANIZATIONAL DEVELOPMENT AND TRAINING 5210 University with your field of choice. Explore what they have to offer members and consider learning from, and creating

Oregon, University of

26

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

SciTech Connect (OSTI)

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

Du, Shiyu [Los Alamos National Laboratory; Andersson, Anders D. [Los Alamos National Laboratory; Germann, Timothy C. [Los Alamos National Laboratory; Stanek, Christopher R. [Los Alamos National Laboratory

2012-05-02T23:59:59.000Z

27

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

SciTech Connect (OSTI)

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.

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

1984-02-01T23:59:59.000Z

28

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

SciTech Connect (OSTI)

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.

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

1997-12-08T23:59:59.000Z

29

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

SciTech Connect (OSTI)

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.

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

1998-05-11T23:59:59.000Z

30

Bubble formation and Kr distribution in Kr-irradiated UO2  

SciTech Connect (OSTI)

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.

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

31

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

DOE Patents [OSTI]

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].

Tomczuk, Z.; Miller, W.E.

1994-10-18T23:59:59.000Z

32

Oxidative Dissolution of UO2 in a Simulated Groundwater Containing Synthetic Nanocrystalline Mackinawite  

SciTech Connect (OSTI)

The long-term success of in situ reductive immobilization of uranium (U) depends on the stability of U(IV) precipitates (e.g., uraninite) under oxic 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 phases by oxygen and Fe(III). Yet, the inhibition mechanism(s) and reaction rates of uraninite (UO2) oxidative dissolution by oxic species such as oxygen in FeS-bearing systems remain largely unresolved. To address this knowledge gap, abiotic batch experiments were conducted with synthetic UO2 in the presence and absence of synthetic mackinawite (FeS) under simulated groundwater conditions of pH = 7, PO2 = 0.02 atm, and PCO2 = 0.05 atm (equivalent to total dissolved carbonate of 0.01 M). The kinetic profiles of dissolved uranium indicate that FeS inhibited UO2 dissolution for 51 hr by effectively scavenging oxygen and keeping dissolved oxygen (DO) low. During this time period, oxidation of structural Fe(II) and S(-II) of FeS were found to control the DO levels, leading to the formation of iron oxyhydroxides and elemental sulfur, respectively, as verified by X-ray diffraction (XRD), Mössbauer and X-ray absorption spectroscopy (XAS). After FeS was depleted due to oxidation, DO levels increased and UO2 oxidative dissolution occurred at an initial rate of rm = 1.2 ± 0.4 ×10-8 mol•g-1•s-1, higher than rm = 5.4 ± 0.3 ×10-9 mol•g-1•s-1 in the control experiment where FeS was absent. Soluble U(VI) products were adsorbed by iron oxyhydroxides (i.e. nanogoethite and ferrihydrite) formed from FeS oxidation, which facilitated the detachment of U(VI) surface complexes and more rapid dissolution of UO2. XAS analysis confirmed the adsorption of U(VI) species, and also showed that U(VI) was not significantly incorporated into iron oxyhydroxide structure. This work reveals that both the oxygen scavenging by FeS and the adsorption of U(VI) to FeS oxidation products may be important in U reductive immobilization systems subject to redox cycling events.

Bi, Yuqiang; Hyun, Sung Pil; Kukkadapu, Ravi K.; Hayes, Kim F.

2013-02-01T23:59:59.000Z

33

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

SciTech Connect (OSTI)

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.

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

1991-11-01T23:59:59.000Z

34

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

E-Print Network [OSTI]

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

Feinroth, H.

35

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

SciTech Connect (OSTI)

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.

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

2014-08-20T23:59:59.000Z

36

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

SciTech Connect (OSTI)

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.

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

1989-11-01T23:59:59.000Z

37

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

SciTech Connect (OSTI)

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.

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

1990-12-31T23:59:59.000Z

38

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

SciTech Connect (OSTI)

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.

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

2006-03-01T23:59:59.000Z

39

Final Version: Orbital Specificity in the Unoccupied States of UO2 from Resonant Inverse Photoelectron Spectroscopy  

SciTech Connect (OSTI)

One of the crucial questions of all actinide electronic structure determinations is the issue of 5f versus 6d character and the distribution of these components across the density of states. Here, a break-though experiment is discussed, which has allowed the direct determination of the U5f and U6d contributions to the unoccupied density of states (UDOS) in Uranium Dioxide. A novel Resonant Inverse Photoelectron (RIPES) and X-ray Emission Spectroscopy (XES) investigation of UO{sub 2} is presented. It is shown that the U5f and U6d components are isolated and identified unambiguously.

Tobin, J G; Yu, S W

2012-03-12T23:59:59.000Z

40

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)

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

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

SciTech Connect (OSTI)

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.

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

2007-06-01T23:59:59.000Z

42

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

SciTech Connect (OSTI)

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.

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

1996-12-31T23:59:59.000Z

43

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

SciTech Connect (OSTI)

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.

M. Pope; S. Bays; R. Ferrer

2008-03-01T23:59:59.000Z

44

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

SciTech Connect (OSTI)

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.

Veleckis, E.; Hoh, J.C.

1991-03-01T23:59:59.000Z

45

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

E-Print Network [OSTI]

.2 results showed a good agreement with calculations obtained with TECH-M computer program. The cladding temperatures of the MOX assembly have been compared with that of the hot UO? assembly. The peak cladding temperature of MOX assembly is about 55 K higher...

Fu, Chun

2012-06-07T23:59:59.000Z

46

Evaluation of sintering effects on SiC incorporated UO2 kernels under Ar and Ar-4%H2 environments  

SciTech Connect (OSTI)

Silicon carbide (SiC) is suggested as an oxygen getter in UO2 kernels used for TRISO particle fuels to lower oxygen potential and prevent kernel migration during irradiation. Scanning electron microscopy and X-ray diffractometry analyses performed on sintered kernels verified that internal gelation process can be used to incorporate SiC in urania fuel kernels. Sintering in either Ar or Ar-4%H2 at 1500 C lowered the SiC content in the UO2 kernels to some extent. Formation of UC was observed as the major chemical phase in the process, while other minor phases such as U3Si2C2, USi2, U3Si2, and UC2 were also identified. UC formation was presumed to be occurred by two reactions. The first was the SiC reaction with its protective SiO2 oxide layer on SiC grains to produce volatile SiO and free carbon that subsequently reacted with UO2 to form UC. The second process was direct UO2 reaction with SiC grains to form SiO, CO, and UC, especially in Ar-4%H2. A slightly higher density and UC content was observed in the sample sintered in Ar-4%H2, but the use of both atmospheres produced kernels with ~95% of theoretical density. It is suggested that incorporating CO in the sintering gas would prevent UC formation and preserve the initial SiC content.

Silva, Chinthaka M [ORNL] [ORNL; Lindemer, Terrence [Harbach Engineering and Solutions] [Harbach Engineering and Solutions; Hunt, Rodney Dale [ORNL] [ORNL; Collins, Jack Lee [ORNL] [ORNL; Terrani, Kurt A [ORNL] [ORNL; Snead, Lance Lewis [ORNL] [ORNL

2013-01-01T23:59:59.000Z

47

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

E-Print Network [OSTI]

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...

Naramore, Michael J

2010-08-03T23:59:59.000Z

48

14UO TANK,OPENING REPORT NO.5. October 20th -November 26th (37 days total; 27 working days).  

E-Print Network [OSTI]

14UO TANK,OPENING REPORT NO.5. October 20th - November 26th (37 days total; 27 working days). Since the tank was last closed the accelerator ran for 97 days.until this opening which was scheduled to replace was done during the tank-open period. We believe that there would be value in gIvIng our assessments

Chen, Ying

49

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

SciTech Connect (OSTI)

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.

Tonks, Michael R. [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

2014-06-01T23:59:59.000Z

50

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

SciTech Connect (OSTI)

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.

Wyatt, D.M.

1994-12-06T23:59:59.000Z

51

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

SciTech Connect (OSTI)

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.

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

1985-04-01T23:59:59.000Z

52

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

SciTech Connect (OSTI)

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.

S. Utsunomiya; R.C. Ewing

2005-09-08T23:59:59.000Z

53

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

SciTech Connect (OSTI)

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.

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

54

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

SciTech Connect (OSTI)

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.

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

2014-03-10T23:59:59.000Z

55

Intergranular fracture in UO{sub 2}: derivation of traction-separation law from atomistic simulations  

SciTech Connect (OSTI)

In this study, the intergranular fracture behavior of UO{sub 2} 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 ?5 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. (authors)

Zhang, Yongfeng; Millett, P.C.; Tonks, M.R.; Bai, Xian-Ming; Biner, S.B. [Fuels Modeling and Simulation Department, Idaho National Laboratory - INL, Idaho Falls, ID 83415 (United States)

2013-07-01T23:59:59.000Z

56

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

SciTech Connect (OSTI)

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.

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

2013-10-01T23:59:59.000Z

57

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

SciTech Connect (OSTI)

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.

Duncan, D.R.

1994-06-30T23:59:59.000Z

58

Report of clean out and flushing of UO{sub 3} Plant processing equipment: Revision 1  

SciTech Connect (OSTI)

The UO{sub 3} Plant went through a clean out leading to the deactivation of the facility. This clean out consisted of three phases. Phase 1 consisted of the removal of residual process material and the deactivation of most process equipment and instrumentation. Phase 2 consisted of the fixing or removal of contamination so storm water processing would be no longer required. Phase 3 consisted of the remaining activities that had to be completed before the facility was turned over to the Surplus Facility Program. Since the activities of Phase 2 and 3 were closely related, these two phases were worked simultaneously. The first part of this document summarizes the Phase 1 clean out procedures and their results. Phase 1 was completed on February 28, 1994. The second part summarizes the Phase 2/3 clean out procedures and their results. Phase 2/3 was completed before December 31, 1994. Because tanks and equipment were flushed simultaneously or in a specific sequence, the clean out processes are discussed per workplan.

Gonsalves, E.

1994-12-02T23:59:59.000Z

59

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

SciTech Connect (OSTI)

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].

Margaret A. Marshall; John D. Bess

2012-11-01T23:59:59.000Z

60

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

SciTech Connect (OSTI)

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.

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

1991-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

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

SciTech Connect (OSTI)

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.

Gary S. Groenewold; Anita K. Gianotto

2006-03-01T23:59:59.000Z

62

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

SciTech Connect (OSTI)

The gas-phase infrared spectra of discrete uranyl ([UO?]˛?) 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?ą for [UO?(CH?COCH?)?]˛? and was systematically red shifted to 1000 and 988 cm?ą 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]˛? 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?ą.

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

63

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

SciTech Connect (OSTI)

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

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

2012-10-30T23:59:59.000Z

64

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

SciTech Connect (OSTI)

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.

Collins, J.L.

2004-12-02T23:59:59.000Z

65

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

SciTech Connect (OSTI)

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

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

2014-09-02T23:59:59.000Z

66

Monte Carlo analysis of burnup-dependent plutonium concentration profiles in UO{sub 2} and MOX fuel pins  

SciTech Connect (OSTI)

The ability to accurately predict fuel performance is an essential requirement for fuel design studies. Prediction of plutonium concentration profiles in an irradiated fuel pin is important for fuel performance analysis and spent-fuel storage. The MCNP coupling with ORIGEN2 (MCWO) burnup calculation code as demonstrated in this paper can analyze the rim effect in UO{sub 2} and mixed-oxide (MOX) fuel pins. Acceptance of a code such as MCWO depends very strongly on its validation. Validation involves the benchmark of the code predictions to the in-pile experimental data and results of post-irradiation examinations (PIEs). In this paper, a validation was made by comparing the MCWO calculated results with the VIM-BURN code, which has been validated against PIE data. The validated MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. In this paper, Pu concentration (wt%) and fission power profiles versus burnup of UO{sub 2} and reactor-grade (RG)-MOX fuel pins were calculated with MCWO, and results are discussed.

Chang, G.S. [Lockheed Martin Idaho Technologies, Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

1998-09-01T23:59:59.000Z

67

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

SciTech Connect (OSTI)

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.

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

1984-12-01T23:59:59.000Z

68

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

SciTech Connect (OSTI)

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.

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

69

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

SciTech Connect (OSTI)

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.

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

1987-03-01T23:59:59.000Z

70

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

SciTech Connect (OSTI)

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.

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

71

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)

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.

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

72

UO Purchasing & Contracting Services rev: August 13, 2012 Text Requirements for Level 1 and Level 2 Signature Authority Purchase Orders and Invoices.  

E-Print Network [OSTI]

UO Purchasing & Contracting Services rev: August 13, 2012 Text Requirements for Level 1 and Level 2 the payment is due and payable (may be same person exercising L1SA)] · Either: 1. List contract # or purchase number); of 2. If you do not have a written contract or purchase order, generally describe what goods and

73

UO Purchasing & Contracting Services rev: August 19, 2014 Text Requirements for Level 1 and Level 2 Contracting Authority Purchase Orders and Invoices.  

E-Print Network [OSTI]

UO Purchasing & Contracting Services rev: August 19, 2014 Text Requirements for Level 1 and Level 2 Contracting Authority Purchase Orders and Invoices. Purchase Orders - Level 1 Contracting Authority: · L1CA [Insert on First Line of Document Text] · [name of individual exercising Level 1 Contracting Authority

74

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]

applications in solar energy harvesting and electrochemical energy storage. Chartoff, Richard - The UO Polymer and thermodynamics of quantum states of molecules embedded in a quantum environment. Lonergan, Mark C. - Research interesting electrical and electrochemical phenomena in solid-state systems. Marcus, Andrew - The Marcus group

Cina, Jeff

75

UoE Employees How to gain access to internal vacancies As a current University employee, you will be eligible to access (via the jobs site) vacancies  

E-Print Network [OSTI]

UoE Employees ­ How to gain access to internal vacancies As a current University employee, you will be eligible to access (via the jobs site) vacancies advertised internally, in addition to those advertised gain access to all vacancies (including those advertised to internal applicants only) whenever you log

Edinburgh, University of

76

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

SciTech Connect (OSTI)

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.

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

1982-08-01T23:59:59.000Z

77

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

SciTech Connect (OSTI)

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.

Finch, R. J.

1999-04-29T23:59:59.000Z

78

Measurements of the modified conversion ratio by gamma-ray spectrometry of fuel rods for water-moderated UO[sub 2] cores  

SciTech Connect (OSTI)

The modified conversion ratio is defined as the ratio of [sup 238]U captures to total fission. Gamma-ray spectrometry of irradiated fuel rods has been introduced to measure this quantity in two types of water-moderated low-enriched UO[sub 2] cores: the standard core, called the 1.42S core, and a tight-lattice core, called the 0.56S core. The water moderator-to-fuel volume ratios V[sub m]/V[sub [line integral

Nakajima, Ken; Akai, Masanori; Suzaki, Takenori (Japan Atomic Energy Research Inst., Ibaraki (Japan). Dept. of Fuel Cycle Safety Research)

1994-02-01T23:59:59.000Z

79

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)

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.

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

80

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

SciTech Connect (OSTI)

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.

Elam, K.R.

2003-10-07T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
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81

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]

. 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

del Alamo, JesĂşs A.

82

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)

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

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

83

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)

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)

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

84

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

SciTech Connect (OSTI)

Criticality safety evaluations implement requirements to proof of sufficient sub critical margins outside of the reactor environment for example in fuel fabrication plants. Basic criticality data (i.e., criticality standards) are used in the determination of sub critical margins for all processes involving plutonium or enriched uranium. There are several criticality international standards, e.g., ARH-600, which is one the US nuclear industry relies on. The French Nuclear Safety Authority (DGSNR and its advising body IRSN) has requested AREVA NP to review the criticality standards used for the evaluation of its Low Enriched Uranium fuel fabrication plants with CRISTAL V0, the recently updated French criticality evaluation package. Criticality safety is a concern for every phase of the fabrication process including UF{sub 6} cylinder storage, UF{sub 6}-UO{sub 2} conversion, powder storage, pelletizing, rod loading, assembly fabrication, and assembly transportation. Until 2003, the accepted criticality standards were based on the French CEA work performed in the late seventies with the APOLLO1 cell/assembly computer code. APOLLO1 is a spectral code, used for evaluating the basic characteristics of fuel assemblies for reactor physics applications, which has been enhanced to perform criticality safety calculations. Throughout the years, CRISTAL, starting with APOLLO1 and MORET 3 (a 3D Monte Carlo code), has been improved to account for the growth of its qualification database and for increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating a modern basic microscopic cross section set based on JEF2.2 and the comprehensive APOLLO2 and MORET 4 codes. APOLLO2 is well suited for criticality standards calculations as it includes a sophisticated self shielding approach, a P{sub ij} flux determination, and a 1D transport (S{sub n}) process. CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and the implementation of user-friendly graphical interfaces. Due to its comprehensive physical simulation and thanks to its broad qualification database with more than a thousand benchmark/calculation comparisons, CRISTAL V0 provides outstanding and reliable accuracy for criticality evaluations for configurations covering the entire fuel cycle (i.e. from enrichment, pellet/assembly fabrication, transportation, to fuel reprocessing). After a brief description of the calculation scheme and the physics algorithms used in this code package, results for the various fissile media encountered in a UO{sub 2} fuel fabrication plant will be detailed and discussed. (authors)

Doucet, M.; Durant Terrasson, L.; Mouton, J. [AREVA-NP (France)

2006-07-01T23:59:59.000Z

85

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

SciTech Connect (OSTI)

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.

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

86

Structural evolution of the double perovskites Sr{sub 2}B'UO{sub 6} (B' = Mn, Fe, Co, Ni, Zn) upon reduction: Magnetic behavior of the uranium cations  

SciTech Connect (OSTI)

Highlights: {yields} Evolution of the double perovskites Sr{sub 2}B'UO{sub 6} upon reduction were studied by XRPD. {yields} Orthorhombic (Pnma) disordered perovskites SrB'{sub 0.5-x}U{sub 0.5+x}O{sub 3} were obtained at 900 {sup o}C. {yields} U{sup 5+/4+} and Zn{sup 2+} cations are distributed at random over the octahedral positions. {yields} AFM ordering for the perovskite with B' = Zn appears below 30 K. -- Abstract: We describe the preparation of five perovskite oxides obtained upon reduction of Sr{sub 2}B'UO{sub 6} (B' = Mn, Fe, Co, Ni, Zn) with H{sub 2}/N{sub 2} (5%/95%) at 900 {sup o}C during 8 h, and their structural characterization by X-ray powder diffraction (XRPD). During the reduction process there is a partial segregation of the elemental metal when B' = Co, Ni, Fe, and the corresponding B'O oxide when B' = Mn, Zn. Whereas the parent, oxygen stoichiometric double perovskites Sr{sub 2}B'UO{sub 6} are long-range ordered concerning B' and U cations. The crystal structures of the reduced phases, SrB'{sub 0.5-x}U{sub 0.5+x}O{sub 3} with 0.37 < x < 0.27, correspond to simple, disordered perovskites; they are orthorhombic, space group Pnma (No. 62), with a full cationic disorder at the B site. Magnetic measurements performed on the phase with B' = Zn, indicate uncompensated antiferromagnetic ordering of the U{sup 5+}/U{sup 4+} sublattice below 30 K.

Pinacca, R.M., E-mail: rmp@unsl.edu.ar [Area de Quimica General e Inorganica 'Dr. Gabino F. Puelles', Departamento de Quimica, Facultad de Quimica, Bioquimica y Farmacia, Universidad Nacional de San Luis, Chacabuco y Pedernera, 5700 San Luis (Argentina); Viola, M.C.; Pedregosa, J.C. [Area de Quimica General e Inorganica 'Dr. Gabino F. Puelles', Departamento de Quimica, Facultad de Quimica, Bioquimica y Farmacia, Universidad Nacional de San Luis, Chacabuco y Pedernera, 5700 San Luis (Argentina)] [Area de Quimica General e Inorganica 'Dr. Gabino F. Puelles', Departamento de Quimica, Facultad de Quimica, Bioquimica y Farmacia, Universidad Nacional de San Luis, Chacabuco y Pedernera, 5700 San Luis (Argentina); Carbonio, R.E. [INFIQC (CONICET), Departamento de Fisicoquimica, Facultad de Ciencias Quimicas, Universidad Nacional de Cordoba, Ciudad Universitaria, X5000HUA Cordoba (Argentina)] [INFIQC (CONICET), Departamento de Fisicoquimica, Facultad de Ciencias Quimicas, Universidad Nacional de Cordoba, Ciudad Universitaria, X5000HUA Cordoba (Argentina); Lope, M.J. Martinez; Alonso, J.A. [Instituto de Ciencia de Materiales de Madrid, C.S.I.C., Cantoblanco, 28049 Madrid (Spain)] [Instituto de Ciencia de Materiales de Madrid, C.S.I.C., Cantoblanco, 28049 Madrid (Spain)

2011-11-15T23:59:59.000Z

87

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)

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.

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

88

Pipe diffusion at dislocations in UO2  

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89

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

90

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

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

Groundwater Containing Synthetic Nanocrystalline Mackinawite. Abstract: The long-term success of in situ reductive immobilization of uranium (U) depends on the stability of...

91

Modeling of UO{sub 2} oxidation in steam atmosphere  

SciTech Connect (OSTI)

Nuclear fuel oxidation is an important phenomenon affecting fission product behavior. As indicated by a number of studies, uranium dioxide shows a very wide range of nonstoichiometric states. In steam, fuel oxidation produces a hyperstoichiometric composition, changing the transport properties. Variation of stoichiometry changes diffusion coefficients for oxygen, noble gases, and fission products substantially, affecting the release of fission products.

Dobrov, B.V.; Likhanskii, V.V. [Triniti Research Center, Triniti, Moscow (Russian Federation); Ozrin, V.D. [Nuclear Safety Institute IBREA, Moscow (Russian Federation)] [and others

1997-12-01T23:59:59.000Z

92

Oxidative Dissolution of UO2 in a Simulated Groundwater Containing  

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93

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

DOE Patents [OSTI]

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}.

Tomczuk, Z.; Miller, W.E.

1992-01-01T23:59:59.000Z

94

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

E-Print Network [OSTI]

-sur-Yvette (Seine-et- Oise). (2) Quelques tentatives pour obtenir le spectre de l'ura- nium par décharge dans UF 6 n

Boyer, Edmond

95

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

SciTech Connect (OSTI)

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.

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

96

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

SciTech Connect (OSTI)

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.

Gerber, M.S.

1993-11-01T23:59:59.000Z

97

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

SciTech Connect (OSTI)

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

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

2011-01-18T23:59:59.000Z

98

(04) UO05 MFA (Creative Writing)/11 Master of Fine Arts (Creative Writing)  

E-Print Network [OSTI]

the currently proposed degree, the practical experience of conceiving, sustaining and completing a significant) will be expected to have completed papers at each level or, in the case of external or overseas applicants of the origins and management of creative processes, awareness of generic literary forms and traditions

Hickman, Mark

99

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

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100

Ligand field effects on the multiplet structure of the U4f XPS of UO2. |  

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

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Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Ab initio DFT+U Study of He Atom Incorporation into UO2 Crystals. | EMSL  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc Documentation RUCProductstwrmrAre the Effects of GlobalASCR UserProgramICarbide.Ar-CF. Ab

102

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

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

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103

Excited States and Luminescent Properties of UO2F2 and Its Solvated  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsing ZirconiaPolicy and Assistance100 tonusingdeposition. |

104

A QM/MM Study on the Aqueous Solvation of theTetrahydroxouranylate [UO(OH)]  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation Desert SouthwestTechnologies |November 2011A First LookMicroscopyComplex Ion. | EMSL QM/MM

105

A DualDisk File System: ext4 Mihai Budiu  

E-Print Network [OSTI]

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

Budiu, Mihai

106

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

E-Print Network [OSTI]

annealed at different temperatures to study the mechanism and kinetic of intermetallic phase formation [19]. One important fission product that impacts FCCI is lanthanide, which influences the interdiffusion process. Researchers have found that FCCI... is determined by the particular fuel/cladding combination and temperature before lanthanide accumulated at the interface; but dominated by the presence of lanthanides after there is enough of them. And the accumulation of lanthanide at the interphase...

Chen, Tianyi

2013-08-02T23:59:59.000Z

107

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...

108

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

E-Print Network [OSTI]

, in combination with rising demand for wood fuel and charcoal in the face of increasing energy costs

109

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

E-Print Network [OSTI]

;equilibria. A platinum effusion chamber (12 mm x 12 mm) was used with (0.5 to 1.2 mm) effusion orifice. The temperature was measured with a Pt-Pt/Rh (10 %) thermocouple, the accuracy being +4 K. Ionic currents were

Rudnyi, Evgenii B.

110

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou areDowntownRockyDeparttient,ofREQUEST FOR75' 00.955

111

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou$0.C. 20545 OCTTO:March 20,Since 5% "y$ --

112

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

SciTech Connect (OSTI)

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.

Carbajo, J.J.

2005-05-27T23:59:59.000Z

113

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]

that have applications in solar energy harvesting and electrochemical energy storage Cina, Jeffrey A research and materials science. Lonergan, Mark C. - Research in the Lonergan group blends synthesis and electrochemical phenomena in solid-state systems Marcus, Andrew - The Marcus group studies the structure

Cina, Jeff

114

Raman Investigation of The Uranium Compounds U3O8, UF4, UH3 and UO3 under Pressure at Room Temperature  

SciTech Connect (OSTI)

Our current state-of-the-art X-ray diffraction experiments are primarily sensitive to the position of the uranium atom. While the uranium - low-Z element bond (such as U-H or U-F) changes under pressure and temperature the X-ray diffraction investigations do not reveal information about the bonding or the stoichiometry. Questions that can be answered by Raman spectroscopy are (i) whether the bonding strength changes under pressure, as observed by either blue- or red-shifted peaks of the Raman active bands in the spectrum and (ii) whether the low-Z element will eventually be liberated and leave the host lattice, i.e. do the fluorine, oxygen, or hydrogen atoms form dimers after breaking the bond to the uranium atom. Therefore Raman spectra were also collected in the range where those decomposition products would appear. Raman is particularly well suited to these types of investigations due to its sensitivity to trace amounts of materials. One challenge for Raman investigations of the uranium compounds is that they are opaque to visible light. They absorb the incoming radiation and quickly heat up to the point of decomposition. This has been dealt with in the past by keeping the incoming laser power to very low levels on the tens of milliWatt range consequently affecting signal to noise. Recent modern investigations also used very small laser spot sizes (micrometer range) but ran again into the problem of heating and chemical sensitivity to the environment. In the studies presented here (in contrast to all other studies that were performed at ambient conditions only) we employ micro-Raman spectroscopy of samples situated in a diamond anvil cell. This increases the trustworthiness of the obtained data in several key-aspects: (a) We surrounded the samples in the DAC with neon as a pressure transmitting medium, a noble gas that is absolutely chemically inert. (b) Through the medium the sample is thermally heat sunk to the diamond anvils, diamond of course possessing the very best heat conductivity of any material. Therefore local heating and decomposition are avoided, a big challenge with other approaches casting doubts on their results. (c) This in turn benefits the signal/noise ratio tremendously since the Raman features of uranium-compounds are very small. The placement of the samples in DACs allows for higher laser powers to impinge on the sample spot while keeping the spot-size larger than in previous studies and keep the samples from heating up. Raman spectroscopy is a very sensitive non-invasive technique and we will show that it is even possible to distinguish the materials by their origin / manufacturer as we have studied samples from Cameco (Canada) and IBI-Labs (US-Florida) and can compare with ambient literature data for samples from Strem (US-MA) and Areva (Pierrelatte, France).

Lipp, M J; Jenei, Z; Park-Klepeis, J; Evans, W J

2011-12-15T23:59:59.000Z

115

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]

Hello "Rhythms and Rhymes" FIG student! My name is Brandon Parry and I will be your FIG Assistant, and I wanted to give Professor Kendall the chance to tell you a little more about himself: Hello. I

Oregon, University of

116

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

E-Print Network [OSTI]

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

Nottingham, University of

117

(front end fuel cycle) 2.1 (CANDU  

E-Print Network [OSTI]

, , 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

Hong, Deog Ki

118

Sodium meta-autunite colloids: Synthesis, characterization, stability  

E-Print Network [OSTI]

J.T.Baker) and crystalline uranyl nitrate, UO 2 (NO 3 ) 2 .by: mixing 0.5 mM uranyl nitrate, UO 2 (NO 3 ) 2 .6H 2 O,

Zheng, Zuoping; Wan, Jiamin; Tokunaga, Tetsu K.

2004-01-01T23:59:59.000Z

119

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]

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, [

Szigethy, Géza

2011-01-01T23:59:59.000Z

120

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

E-Print Network [OSTI]

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

Yanikoglu, Berrin

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

MODELING THE PERFORMANCE OF HIGH BURNUP THORIA AND URANIA PWR FUEL  

E-Print Network [OSTI]

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

Long, Y.

122

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

123

IWW's strategic plan at maturity was to be considered a technical resource by state and federal agencies. IWW was strategically positioned to  

E-Print Network [OSTI]

micro-hydro research through OSU Foundation to fund first joint UO JD/ OSU MS water student. #12;Metric

Escher, Christine

124

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

E-Print Network [OSTI]

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

Szigethy, Geza

2011-01-01T23:59:59.000Z

125

Investigation of Uranium Polymorphs  

SciTech Connect (OSTI)

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.

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

126

Molten uranium dioxide structure and dynamics  

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

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.

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

127

Carbon Management working with the  

E-Print Network [OSTI]

: UoR42 Energy centre 82 Appendix C29: UoR43 Fume cupboard control 83 Appendix C30: UoR44 Solar PV achievement towards target 29 5. Carbon Management Plan Financing 32 5.1 Assumptions 32 5.2 Benefits / savings ­ quantified and un-quantified 33 5.3 Additional resources 33 5.4 Financial costs and sources of funding 34 6

Reading, University of

128

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

SciTech Connect (OSTI)

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.

K. Gofryk; N. Harrison; M. Jaime

2014-12-01T23:59:59.000Z

129

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

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

extracellularly to high densities in association with an exopolymeric substance (EPS). In wild type cells, this UO2-EPS matrix exhibited glycocalyx-like properties,...

130

Electron donor-dependent radionuclide reduction and nanoparticle...  

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

size were observed, the association of UO2 nanoparticles with an exopolymeric substance (EPS) was observed and found to be independent of electron donor source. Electron...

131

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...

132

E-Print Network 3.0 - autoimmune skin blistering Sample Search...  

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

Engineering, McMaster University Collection: Materials Science 6 For additional Environmental Health and Safety needs please visit our website at: http:www.uos.harvard.edu...

133

E-Print Network 3.0 - action description memorandum Sample Search...  

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

REPAIR GOODS MEMORANDUM Summary: UNIVERSITY OF WASHINGTON RETURNREPAIR GOODS MEMORANDUM PURCHASING, Box No. 351110 UoW 1458 (Rev.8... label) Vendor Authorization Name P.O. Item...

134

Searching for the Decay of 229m Th  

E-Print Network [OSTI]

3 , UO 2 (NO 3 ) 2 (uranyl nitrate), and metallic uranium.varied from 12% for uranyl nitrate to 31% for uranium metal.

Swanberg, Erik

2012-01-01T23:59:59.000Z

135

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

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

35,346 square foot multi- storied concrete structure used for the concentration of liquid uranium solutions and UO3 powder conversion equipment. 224-UA Calcination and Loadout...

136

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

SciTech Connect (OSTI)

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.

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

137

NANCY YEN-WEN CHENG Department of Architecture, School of Architecture and Allied Arts nywc@uoregon.edu  

E-Print Network [OSTI]

of Oregon (UO) Architecture Department 2012-present Director, UO - Shanghai Xian Dai Sustainable Design1 NANCY YEN-WEN CHENG Department of Architecture, School of Architecture and Allied Arts nywc mobile: +1-541-556-4590 http://architecture.uoregon.edu/faculty/cheng AUS mobile +61 (04) 1824 3873 E D U

138

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]

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

Johnson, Robert E.

139

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

SciTech Connect (OSTI)

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.

K. Gofryk; V. Zapf; M. Jaime

2014-12-01T23:59:59.000Z

140

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

E-Print Network [OSTI]

, the thermal conductivity of UO2 is very low, and the search for alternative materials continuesOrigin of Low Thermal Conductivity in Nuclear Fuels Quan Yin and Sergey Y. Savrasov Department in a very low thermal conductivity of modern nuclear fuels. Consider semiconducting UO2 which is a main

Savrasov, Sergej Y.

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

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

SciTech Connect (OSTI)

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.

Carnahan, C.L.

1987-12-01T23:59:59.000Z

142

Electrolytic process for preparing uranium metal  

DOE Patents [OSTI]

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.

Haas, Paul A. (Knoxville, TN)

1990-01-01T23:59:59.000Z

143

Extreme Performance Scalable Operating Systems Final Progress Report (July 1, 2008 Ă?Â?Ă?¢Ă?Â?Ă?Â?Ă?Â?Ă?Â? October 31, 2011)  

SciTech Connect (OSTI)

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

Allen D. Malony; Sameer Shende

2011-10-31T23:59:59.000Z

144

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

E-Print Network [OSTI]

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 ...

Boscher, Thomas

2005-01-01T23:59:59.000Z

145

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.

146

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

SciTech Connect (OSTI)

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.

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

1991-01-01T23:59:59.000Z

147

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

E-Print Network [OSTI]

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 ...

Mertens, Paul Gustaaf

1971-01-01T23:59:59.000Z

148

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

149

2/21/11 11:08 AMOregon Quarterly Features Page 1 of 4http://www.oregonquarterly.com/winter2010/feature4.php  

E-Print Network [OSTI]

/feature4.php UO Home | Dept index Winter 2010 | Volume 90, Number 2 Donate to OQ | Past Issues:08 AMOregon Quarterly Features Page 2 of 4http://www.oregonquarterly.com/winter2010/feature4.php monochrome

Richmond, Geraldine L.

150

Small Satellite Applications of Commercial off the Shelf Radio Frequency Integrated Circuits  

E-Print Network [OSTI]

small satellite, Orbiting Satellite Carrying Amateur Radio (OSCAR) UO-14, was launched in 1990 [15]. While these amateur radio speeds may be sufficient for requirements of the aforementioned missions, they do very little towards expanding future...

Graves, John

2012-02-14T23:59:59.000Z

151

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)

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.

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

152

Characterizing solution and solid-phase amorphous uranyl silicates q  

E-Print Network [OSTI]

2007 Elsevier Ltd. All rights reserved. 1. INTRODUCTION Dissolved uranium, as the uranyl ion UO2 2Ăľ , is consid- ered a contaminant introduced into the environment near mining, processing and production

Illinois at Chicago, University of

153

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...

154

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

E-Print Network [OSTI]

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...

Lebrun-Grandié, Damien Thomas

2014-03-05T23:59:59.000Z

155

Diffusion and Adsorption of Uranyl Carbonate Species in Nanosized...  

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

whereas adsorption of CO32- and Ca2UO2(CO3)3, which attach to the surface via hydrogen bonding from a surface hydroxyl group to a carbonate oxygen, was calculated to be either...

156

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

E-Print Network [OSTI]

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

Fernandez, Jeffrey

2009-01-01T23:59:59.000Z

157

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

158

Physical Separation and Multiphase  

E-Print Network [OSTI]

- 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

Sóbester, András

159

E-Print Network 3.0 - attached chinese hamster Sample Search...  

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

uranium; V79, Chinese hamster lung cells; DU-UO2NO3 depleted uranium... - tial for DU as uranyl nitrate to induce mutations and cell trans- formation in Chinese hamster lung... in...

160

TO  

Office of Legacy Management (LM)

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

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

A density functional study of actinyl containing complexes.  

E-Print Network [OSTI]

??Density functional (DFT) methods are first used to study 22 of the most stable solution-phase UN4O12 isomers containing uranyl nitrate, UO2(NO3)2. Based on relative free… (more)

Berard, Joel J.

2008-01-01T23:59:59.000Z

162

Whole-genome transcriptional analysis of heavy metal stresses in Caulobacter crescentus  

E-Print Network [OSTI]

A concentration of 200 ?M uranyl nitrate was used forthe exception of the uranyl nitrate stock solution which wasK 2 Cr 2 O 7 ) and uranyl nitrate (UO 2 (NO 3 ) 2 ·6H 2 O).

Hu, Ping; Brodie, Eoin L.; Suzuki, Yohey; McAdams, Harley H.; Andersen, Gary L.

2005-01-01T23:59:59.000Z

163

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...

164

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...

165

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]

. 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

Svane, Axel Torstein

166

Pacific Northwest Solar Radiation Data  

E-Print Network [OSTI]

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

Oregon, University of

167

Computational evaluation of two reactor benchmark problems  

E-Print Network [OSTI]

benchmark problem . . . Fig. 2. Layouts of assembly types B and C Fig. 3. Core diagram/layout for the NEA WPPR benchmark problem . . . Fig. 4. Layouts of UOz and MOX assemblies Fig. 5. Core A effective multiplication factor. Fig. 6. Core B effective... by rod peaking factors for the MOX assembly. . . . . . . . . . . . . Fig. 12 Rod by rod peaking factors for the middle UO. assembly . . . Fig. 13. Rod by rod peaking factors for the corner UO assembly. . . . . . 30 . . . . . 3 1 . . . . . 32 Fig. 14...

Cowan, James Anthony

2012-06-07T23:59:59.000Z

168

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

SciTech Connect (OSTI)

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.

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

2005-05-11T23:59:59.000Z

169

Fundamental study on recovery uranium oxide from HEPA filters  

SciTech Connect (OSTI)

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.

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

170

Infrared Spectroscopy of Discrete Uranyl Anion Complexes  

SciTech Connect (OSTI)

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.

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

171

Infared Spectroscopy of Discrete Uranyl Anion Complexes  

SciTech Connect (OSTI)

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.

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

172

P a g e | 1 Regional Ocean Modelling  

E-Print Network [OSTI]

(external data). #12;P a g e | 4 Slide 4: Flather Condition for Shallow-Water Barotropic Flow: h/t = -Hu/x u-running) characteristic for uo-c subcritical flows. Thus, either we set "u - (g/H)1/2 h + (g/H)1/2 h for uo+c >0 always for subcritical flows. This characteristic is determined as part

173

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)

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.

Mac Donald, Philip Elsworth

2002-09-01T23:59:59.000Z

174

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

SciTech Connect (OSTI)

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.

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

2012-02-01T23:59:59.000Z

175

The burnup dependence of light water reactor spent fuel oxidation  

SciTech Connect (OSTI)

Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5).

Hanson, B.D.

1998-07-01T23:59:59.000Z

176

Surface Decontamination of System Components in Uranium Conversion Plant at KAERI  

SciTech Connect (OSTI)

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.

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

2003-02-25T23:59:59.000Z

177

EFRC CMSNF Major Accomplishments  

SciTech Connect (OSTI)

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.

D. Hurley; Todd R. Allen

2014-09-01T23:59:59.000Z

178

An economic study of the peanut industry  

E-Print Network [OSTI]

?CBCDPECYCOOW PEC rcPO UDC DCsyfrMkCn mS PEC PDUnC UO DCUnS lyD nMOayOUYe NECS UDC DClCDDCn Py UO ilUDHCDOu OPysp aCUrcPOie NEC fCrCDUY aDUsPMsC MO Py OUsp PEC aCUrcPO Mr yrC EcrnDCn aycrn mUfO Urn EUcY PECH Py HUDpCP yD Py U hUDCEycOCe NEC EUS MO mUYCn UO... MP MO PEDCOECn Urn CMPECD OPyDCn yr PEC lUDH yD HUDpCPCne vP EUO U EMfE lCCn BUYcC lyD YMBCOPysp UO lynnCD yD fDycrn UO Ur MrfDCnMCrP lyD HMJCn lCCnO Urn PEC syHHCDsMUY aCYYCP PSaC yl lCCne bUmyD NEC YUmyD DC,cMDCHCrPO Mr PEC yaCDUPMyr Mrs...

Wilkins, Charlie Smith

2013-10-04T23:59:59.000Z

179

The effect of geometry on symbology recognition  

E-Print Network [OSTI]

displays' Of the twenty geometric forms tested it was reported that the best combinations of five symbols each were 1 ) rectangle, circle, zig-zag Z, cross, and semicircle or 2) cross, semicircle, ellipse, triangle, and square. These studies led...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...

Boyless, James Andrus

2012-06-07T23:59:59.000Z

180

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

SciTech Connect (OSTI)

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.

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

1992-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

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

SciTech Connect (OSTI)

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.

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

182

c-Type Cytochrome-Dependent Formation of U(IV) Nanoparticles by Shewanella oneidensis  

SciTech Connect (OSTI)

Modern approaches for bioremediation of radionuclide contaminated environments are based on the ability of microorganisms to effectively catalyze changes in the oxidation states of metals that in turn influence their solubility. Although microbial metal reduction has been identified as an effective means for immobilizing highly-soluble uranium(VI) complexes in situ, the biomolecular mechanisms of U(VI) reduction are not well understood. Here, we show that c-type cytochromes of a dissimilatory metal reducing bacterium, Shewanella oneidensis MR-1 are essential for the reduction of U(VI) and formation of extracelluar UO2 nanoparticles. In particular, the outer membrane (OM) decaheme cytochrome MtrC, previously implicated in Mn(IV) and Fe(III) reduction, directly transferred electrons to U(VI). Additionally, deletions of mtrC and/or 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 extracellularly to high densities in association with an exopolymeric substance (EPS). In wild type cells, this UO2-EPS matrix exhibited glycocalyx-like properties, contained multiple elements of the OM, polysaccharide, and heme containing proteins. Using a novel combination of methods including synchrotron-based X-ray fluorescence microscopy and high resolution immune-electron microscopy, we demonstrate a close association of the extracellular UO2 nanoparticles with MtrC and OmcA. This is the first study to directly localize the OM-associated cytochromes with EPS, which contains biogenic UO2 nanoparticles. In the environment, such association of UO2 nanoparticles with biopolymers may exert a strong influence on subsequent behavior including susceptibility to oxidation by O2 or transport in soils and sediments.

Marshall, Matthew J.; Beliaev, Alex S.; Dohnalkova, Alice; Kennedy, David W.; Shi, Liang; Wang, Zheming; Boyanov, Maxim I.; Lai, Barry; Kemner, Kenneth M.; Mclean, Jeffrey S.; Reed, Samantha B.; Culley, David E.; Bailey, Vanessa L.; Simonson, Cody J.; Saffarini, Daad; Romine, Margaret F.; Zachara, John M.; Fredrickson, Jim K.

2006-08-08T23:59:59.000Z

183

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

SciTech Connect (OSTI)

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.

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

184

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]

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

Paris-Sud XI, Université de

185

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

E-Print Network [OSTI]

. 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

Hong, Deog Ki

186

Preparation of thorium-uranium gel spheres  

SciTech Connect (OSTI)

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.).

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

1980-01-01T23:59:59.000Z

187

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

SciTech Connect (OSTI)

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.

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

2011-01-04T23:59:59.000Z

188

Shutdown mechanisms for a hypothetical criticality accident involving HEU powder: Preliminary results  

SciTech Connect (OSTI)

This work examines the physical processes that would cause an accidental criticality involving higly enriched uranium(HEU) powder to shut down naturally. The study analyses an excursion resulting from the continous poring of slightly damp HEU powder (either UO{sub 3} or UF{sub 4} containing 1.5% water) onto a concrete floor.

Bentley, C.; Basoglu, B.; Dunn, M.; Plaster, M.; Ruggles, A.; Wilkinson, A.; Yamamoto, T.; Dodds, H. [Univ. of Tennessee, Knoxville, TN (United States)

1994-12-31T23:59:59.000Z

189

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

E-Print Network [OSTI]

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

Oregon, University of

190

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

E-Print Network [OSTI]

* * *: {( 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

Rathbun, Julie A.

191

Continued on back UNIVERSITY OF OREGON  

E-Print Network [OSTI]

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

Oregon, University of

192

Project-Role Pair user_tokens  

E-Print Network [OSTI]

OSAC Users (U) Domains (D) Roles (R) User Assignment (UA) Permission Assignment (PA) Project Ownership (PO) Project-Role Pair (PRP) Projects (P) Tokens (T) User Ownership (UO) Services (S) user_tokens token_project Groups (G) Group Ownership (GO) User Group (UG) Group Assignment (GA) token_roles PERMS

Sandhu, Ravi

193

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

E-Print Network [OSTI]

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

Weyde, Tillman

194

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

E-Print Network [OSTI]

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

Clarno, Kevin Taylor

2012-06-07T23:59:59.000Z

195

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]

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...

Bellanger, Philippe

2012-06-07T23:59:59.000Z

196

Modulation cancellation method for laser spectroscopy V. Spagnolo*a,b  

E-Print Network [OSTI]

Fisica, University and Politecnico of Bari, CNR--IFN UOS BARI, Via Amendola 173, 70126 Bari, Italy of the proposed approach are: *spagnolo@fisica.uniba.it; phone 39 080 544-2373; fax 39 080 544-2219; www.fisica

197

Curriculum vitae of Viviana Mascardi (updated to September Personal data  

E-Print Network [OSTI]

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

Mascardi, Viviana

198

n a recent TEDx talk she gave before a Portland audience, microbiologist  

E-Print Network [OSTI]

), Brown is an expert in sustainable buildings. His involvement ensures that the discoveries made test sites like the UO's sustainably built Lundquist College of Business and a Portland hospital other, with humans, and with their environment. "Buildings are complex ecosystems that are an important

Oregon, University of

199

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

SciTech Connect (OSTI)

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.

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

200

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

SciTech Connect (OSTI)

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.

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

2008-07-08T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Developing a High Thermal Conductivity Fuel with Silicon Carbide Additives  

SciTech Connect (OSTI)

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:

Ronald baney; James Tulenko

2012-11-20T23:59:59.000Z

202

Bioremediation of Uranium Plumes with Nano-scale  

E-Print Network [OSTI]

(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

Fay, Noah

203

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

SciTech Connect (OSTI)

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.

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

2009-01-26T23:59:59.000Z

204

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

E-Print Network [OSTI]

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

Azad, Abdul-Majeed

205

Synthesis of triglyceride by the intestinal mucosa  

E-Print Network [OSTI]

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....ppdroc sa c3r erNy.erdrocP ase c3r -rterr sa 9G1CGT GF lOESGAGlO& Vuf 1958 Vu'se Ayi'r2cH 5.s23rd.Pcef uo- Myce.c.so ^A & r A&MCOLAEA GF CTEXS&1LTE9L 5& COL EMCLACEM8S VI1GA8 8 9.PPrecuc.so 5f XLGTXL 1OTEACGlOLT 5ILSS 8DDesbr- uP cs Pcfpr uo- 2...

Buell, George Christopher

2013-10-04T23:59:59.000Z

206

A Multi-Modular Neutronically Coupled Power Generation System  

E-Print Network [OSTI]

The High Temperature Integrated Multi-Modular Thermal Reactor is a small modular reactor that uses an enhanced conductivity BeO-UO2 fuel with supercritical CO2 coolant to drive turbo-machinery in a direct Brayton cycle. The core consists of several...

Patel, Vishal

2012-07-16T23:59:59.000Z

207

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

SciTech Connect (OSTI)

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.

K. Gofryk; M. Jaime

2014-12-01T23:59:59.000Z

208

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

209

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

210

Water-Moderated and -Reflected Slabs of Uranium Oxyfluoride  

SciTech Connect (OSTI)

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.

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

2010-09-01T23:59:59.000Z

211

APPLIED AND ENVIRONMENTAL MICROBIOLOGY, Nov. 2005, p. 74537460 Vol. 71, No. 11 0099-2240/05/$08.00 0 doi:10.1128/AEM.71.11.74537460.2005  

E-Print Network [OSTI]

with nitrate and incubated with no electron acceptor, was used for the two time points considered and for both- ganese(IV), nitrate, nitrite, thiosulfate, sulfite, trimethylamine N-oxide (TMAO), dimethyl sulfoxide and soluble hexavalent uranyl (UO2 2 ) and chro- mate (CrO4 2 ) to less soluble and less toxic forms [U

Tebo, Brad

212

Hydrothermal synthesis, structure and thermal stability of diamine templated layered uranyl-vanadates  

E-Print Network [OSTI]

crystal structure and thermal behavior are reported herein. Experimental Synthesis Uranyl nitrate (UO2(NO31 Hydrothermal synthesis, structure and thermal stability of diamine templated layered uranyl. Murielle.rivenet@ensc-lille.fr Running Title : Diamine templated layered uranyl-vanadates. Figure for table

Paris-Sud XI, Université de

213

Nagoya University Description  

E-Print Network [OSTI]

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

Bristol, University of

214

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

SciTech Connect (OSTI)

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.

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

2000-08-01T23:59:59.000Z

215

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

SciTech Connect (OSTI)

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.

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

216

SURVEY OF THE COLUMBIA RIVER AND ITS TRIBUTARIES -Part VII  

E-Print Network [OSTI]

SURVEY OF THE COLUMBIA RIVER AND ITS TRIBUTARIES - Part VII I ^^^^'fie^BkJioJS SPECIAL SCIENTIFIC, Director Special Scientific Report - Fisheries No. UO SURVEY OF THE COLUMBIA RIVER AND ITS TRIBUTARIES PART these have been divided for con- venience into four sub-areas. On the Idaho side of the Snake River

217

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

218

IAEA-TECDOC-1450 Thorium fuel cycle --Potential  

E-Print Network [OSTI]

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

Laughlin, Robert B.

219

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

E-Print Network [OSTI]

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-

Szigethy, Geza

2010-01-01T23:59:59.000Z

220

Radiative heat transfer in porous uranium dioxide  

SciTech Connect (OSTI)

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.

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

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Reactivity initiated accident test series Test RIA 1-4 fuel behavior report. [PWR; BWR  

SciTech Connect (OSTI)

This report presents and discusses results from the final test in the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-4, conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Nine preirradiated fuel rods in a 3 x 3 bundle configuration were subjected to a power burst while at boiling water reactor hot-startup system conditions. The test resulted in estimated axial peak, radial average fuel enthalpies of 234 cal/g UO/sub 2/ on the center rod, 255 cal/g UO/sub 2/ on the side rods, and 277 cal/g UO/sub 2/ on the corner rods. Test RIA 1-4 was conducted to investigate fuel coolability and channel blockage within a bundle of preirradiated rods near the present enthalpy limit of 280 cal/g UO/sub 2/ established by the US Nuclear Regulatory Commission. The test design and conduct are described, and the bundle and individual rod thermal and mechanical responses are evaluated. Conclusions from this final test and the entire PBF RIA Test Series are presented.

Cook, B.A.; Martinson, Z.R.

1984-09-01T23:59:59.000Z

222

GEOBULLETIN SEpTEmBEr 19Th  

E-Print Network [OSTI]

are requested! If you have a news item, a request, an announcement etc. email it to geodept@geology on the oxidation state of uranium, therefore understanding the mechanisms of UO2 oxidative corrosion is essential-classical diffusion is driven by electron transfer from multiple uranium atoms to each interstitial #12;GEOBULLETIN

Carlson, Anders

223

E-Print Network 3.0 - alkaline phosphate wastes Sample Search...  

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

aqueous UO2(PO4)n 2-3n (n > 1) and mixed hydroxide-phosphate... Effects of Phosphate on Uranium(VI) Adsorption to Goethite-Coated Sand T A O C H E N G , M A R K O... -0206...

224

Radiochim. Acta 93, 265272 (2005) by Oldenbourg Wissenschaftsverlag, Mnchen  

E-Print Network [OSTI]

characterization of uranium(VI) silicate solids and associated neptunium(V) By Matthew Douglas1 , Sue B. Clark1 in revised form October 16, 2004) Uranyl / Solid solution / Spent nuclear fuel / Uranium minerals / Uranophane Summary. The uranium(VI) silicate phases urano- phane, Ca[(UO2)(SiO3OH)]2 ·5H2O, and sodium

Utsunomiya, Satoshi

225

Nanocrystalline Metals Andy Howe, Corus RD&T  

E-Print Network [OSTI]

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

Cambridge, University of

226

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

SciTech Connect (OSTI)

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.

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

227

New insights into uranium (VI) sol-gel processing  

SciTech Connect (OSTI)

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.

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

228

X-ray Absorption Spectroscopy Identifies Calcium-Uranyl-Carbonate Complexes at Environmental Concentrations  

SciTech Connect (OSTI)

Current research on bioremediation of uranium-contaminated groundwater focuses on supplying indigenous metal-reducing bacteria with the appropriate metabolic requirements to induce microbiological reduction of soluble uranium(VI) to poorly soluble uranium(IV). Recent studies of uranium(VI) bioreduction in the presence of environmentally relevant levels of calcium revealed limited and slowed uranium(VI) reduction and the formation of a Ca-UO2-CO3 complex. However, the stoichiometry of the complex is poorly defined and may be complicated by the presence of a Na-UO2-CO3 complex. Such a complex might exist even at high calcium concentrations, as some UO2-CO3 complexes will still be present. The number of calcium and/or sodium atoms coordinated to a uranyl carbonate complex will determine the net charge of the complex. Such a change in aqueous speciation of uranium(VI) in calcareous groundwater may affect the fate and transport properties of uranium. In this paper, we present the results from X-ray absorption fine structure (XAFS) measurements of a series of solutions containing 50 lM uranium(VI) and 30 mM sodium bicarbonate, with various calcium concentrations of 0-5 mM. Use of the data series reduces the uncertainty in the number of calcium atoms bound to the UO2-CO3 complex to approximately 0.6 and enables spectroscopic identification of the Na-UO2-CO3 complex. At nearly neutral pH values, the numbers of sodium and calcium atoms bound to the uranyl triscarbonate species are found to depend on the calcium concentration, as predicted by speciation calculations.

Kelly, Shelly D [Argonne National Laboratory (ANL); Kemner, Kenneth M [Argonne National Laboratory (ANL); Brooks, Scott C [ORNL

2007-01-01T23:59:59.000Z

229

Capsule HRB-15B postirradiation examination report  

SciTech Connect (OSTI)

Capsule HRB-15B design tested 184 thin graphite trays containing unbonded fuel particles to peak exposures of 6.6 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/ fast fluence, approx. 27% fissions per initial metal atom (FIMA) fissile burnup, and 6% FIMA fertile burnup at nominal time-averaged temperatures of 815 to 915/sup 0/C. The capsule tested a variety of low-enriched uranium (approx. 19.5% U-235) fissile particle types, including UC/sub 2/, UC/sub x/O/sub y/, UO/sub 2/, zirconium-buffered UO/sub 2/ (referred to in this report as UO/sub 2//sup *), and 1:1(Th,U)O/sub 2/ with both TRISO and silicon-BISO coatings. All fertile particles were ThO/sub 2/ with BISO, silicon-BISO, or TRISO coatings. The findings indicated that all TRISO particles retained virtually all of their fission product inventories, except small quantities of silver, at these irradiation temperatures, while some of the silicon-BISO particles released significant amounts of both silver and cesium. No kernel migration, pressure vessel, or outer pyrolytic carbon (OPyC) failures were observed in the fuel particles, which had total diameters of < 900 ..mu..m; however, the incidence of failed OPyC coatings was found to increase with particle size in the TRISO inert particles, which had diameters of 1000 to 1200 ..mu..m. UO/sub 2//sup */ particles exhibited no detrimental irradiation effects, but they contained pure carbon precipitates in the kernels after irradiation which were not observed in the undoped UO/sub 2/ particles. Postirradiation examination revealed no differences in the irradiation performance of three UC/sub x/O/sub y/ kernel types with varying oxygen/uranium ratios.

Ketterer, J.W.; Bullock, R.E.

1981-06-01T23:59:59.000Z

230

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

SciTech Connect (OSTI)

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.

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

231

Naturalistic interpersonal reactions to assertive and unassertive styles  

E-Print Network [OSTI]

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...

Paquette, Raymond Joseph

1988-01-01T23:59:59.000Z

232

Depleted uranium hexafluoride: Waste or resource?  

SciTech Connect (OSTI)

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.

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

233

Extraction of Uranium from Aqueous Solutions Using Ionic Liquid and Supercritical Carbon Dioxide in Conjunction  

SciTech Connect (OSTI)

Uranyl ions (UO2)2+ in aqueous nitric acid solutions can be extracted into supercritical CO2 (sc-CO2) via an imidazolium-based ionic liquid using tri-n-butylphosphate (TBP) as a complexing agent. The transfer of uranium from the ionic liquid to the supercritical fluid phase was monitored by UV/Vis spectroscopy using a high-pressure fiberoptic cell. The form of the uranyl complex extracted into the supercritical CO2 phase was found to be UO2(NO3)2(TBP)2. The extraction results were confirmed by UV/Vis spectroscopy and by neutron activation analysis. This technique could potentially be used to extract other actinides for applications in the field of nuclear waste management.

Wang, Joanna S.; Sheaff, Chrystal N.; Yoon, Byunghoon; Addleman, Raymond S.; Wai, Chien M.

2009-01-01T23:59:59.000Z

234

Fast Breeder Blanket Facility FBBF. Annual report, January 1, 1981-December 31, 1981  

SciTech Connect (OSTI)

This annual report contains a summmary of fission rate, spectra, and gamma-ray heating rate measurements made in the first blanket of the Purdue Fast Breeder Blanket Facility. The first blanket consisted of aluminum clad, natural UO/sub 2/ fuel rods with a secondary cladding of stainless steel or aluminum. The blanket was arranged in two concentric regions around the neutron source and converter regions. A neutron diffusion code, 2DB, and a Monte Carlo code, VIM, both using homogeneous cross section groups have been used to calculate the reaction rates. Calculated to experimental values for a number of important reactions are presented. A modified method of applying Bondarenko self-shielding factors to correct for the self shielding of resonance energy neutrons in aluminum, stainless steel and UO/sub 2/ has improved the agreement between the calculations and experiment, but does not account for all of the differences.

Clikeman, F M [ed.] [ed.

1982-07-01T23:59:59.000Z

235

Simulation of the thermodynamic properties of organic extraction solutions  

SciTech Connect (OSTI)

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.

Kolker, A.R.

1986-05-01T23:59:59.000Z

236

Method for fluorination of uranium oxide  

DOE Patents [OSTI]

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.

Petit, George S. (Oak Ridge, TN)

1987-01-01T23:59:59.000Z

237

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

SciTech Connect (OSTI)

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.

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

1984-05-01T23:59:59.000Z

238

Female characters in Thomas Wolfe's four major novels: Look Homeward, Angel; Of Time and the River; The Web and the Rock; and You Can't Go Home Again.  

E-Print Network [OSTI]

~ aqszeq aauy us 'BuTSSsu 'yes 'saTyp ~ Zq. Tsze~Tug eqq oq. m~ e~q. pygmy quqq. uTszq agq uo xgO au85~ mund egg Wyaas azs ueyea zeqsTs s~ pus zeqgom ttt!Tt; go tlT Sl!St: OZtt Ztt;g tttt OZBgattt'I. tlt etitttlA Za . Be l'tt' tttl1 exjujf'txoa ggf...ZgggueyT ylxs 89900'GB EzszegTT zan $$9Tlb 8rq QQTCey 8$9Aou za f sin zxioj 8 t 9+Qajtj uo'f &~ss QusSzp '8TGAalx zo f. sB zTLQ+ Gl[$ lx'f szegaszsqa 8 DiG~Gg eqg ~ 9 0". l ssx:0 o zeezsp eq. Tg GqtTt l GIMP/ $0 Aoyu j(X 9'lpga Zyeuueg g yzsqarjj 'p96j...

Sheffield, Jewell Frieda

2012-06-07T23:59:59.000Z

239

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

SciTech Connect (OSTI)

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.

Blaise Collin

2014-08-01T23:59:59.000Z

240

Semiempirical range and stopping power values for heavy ions  

E-Print Network [OSTI]

0 CO CCI CO O O O rn UJ O r mU. m ~ ~ ~ OOOOO cl N O I N 4 O' ICI N w ~ Q ~ N ~ U 0 Q If Cl CO 0' M 0' In ICI ~ ~ 0 000 ~ ~ ~ Ln or o o Lnr o UO N Z O' pCI rn N w ~ ~ ~ ~ Q 4 ~ 0 N CCI OI 0 0 r UI Cr rll m ~ ~ N 0 m... O ~ ~ N U'. r co I ~ ~ N 0 O UO CO CO Q N N ~ ~ nd NI 0 0 ILI C3 CO 0 0 t CO r UJ O' rh ~ ~ cn 0 N 'Z \\ ~ CO m rn m ~ ~ U, 0 N CO cn N N ~ ~ 0 0 U I I 4 I ? 4 4 G V 4 a a LU N IU UJ CO O Lf 0 UJ...

Schilling, Ralph Franklin, III

2012-06-07T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

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

SciTech Connect (OSTI)

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.

Collin, Blaise P.

2014-08-01T23:59:59.000Z

242

AGR-2 IRRADIATION TEST FINAL AS-RUN REPORT  

SciTech Connect (OSTI)

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.

Blaise, Collin

2014-07-01T23:59:59.000Z

243

Standard specification for uranium oxides with a 235U content of less than 5 % for dissolution prior to conversion to nuclear-grade uranium dioxide  

E-Print Network [OSTI]

1.1 This specification covers uranium oxides, including processed byproducts or scrap material (powder, pellets, or pieces), that are intended for dissolution into uranyl nitrate solution meeting the requirements of Specification C788 prior to conversion into nuclear grade UO2 powder with a 235U content of less than 5 %. This specification defines the impurity and uranium isotope limits for such urania powders that are to be dissolved prior to processing to nuclear grade UO2 as defined in Specification C753. 1.2 This specification provides the nuclear industry with a general standard for such uranium oxide powders. It recognizes the diversity of conversion processes and the processes to which such powders are subsequently to be subjected (for instance, by solvent extraction). It is therefore anticipated that it may be necessary to include supplementary specification limits by agreement between the buyer and seller. 1.3 The scope of this specification does not comprehensively cover all provisions for prevent...

American Society for Testing and Materials. Philadelphia

2005-01-01T23:59:59.000Z

244

TREKiSM Issue 40  

E-Print Network [OSTI]

~~JnJas leuo~+8N ay+ WOJj auo 4+~M 5uole 'uosqoH +Jaql~ Aq pau61s 'PJEJ ~IJ aYl Jay 5M04s a4 (OT lpJ8n6 IEuosJad AW S.8J84M (a lUO 6u~o6 s.+EYM(P lpJEaq JnOA sldJa4M (J lS14l 51 WJOj1 un jO pU1~ lP4M (q lAZPJJ auo6 AX8TE6 ay o4M a4l SPH (8 :aJe suo1lsanb BA...

1985-01-01T23:59:59.000Z

245

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

SciTech Connect (OSTI)

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.

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

246

Simulation of Low-Enriched Uranium (LEU) Burnup in Russian VVER Reactors with the HELIOS Code Package  

SciTech Connect (OSTI)

The HELIOS reactor-physics computer program system was used to simulate the burnup of UO{sub 2} fuel in three VVER reactors. The manner in which HELIOS was used in these simulations is described. Predictions of concentrations for actinides up to {sup 244}Cm and for isotopes of neodymium were compared with laboratory-measured values. Reasonable agreement between calculated and measured values was seen for experimental samples from a fuel rod in the interior of an assembly.

Murphy, B.D.; Kravchenko, J.; Lazarenko, A.; Pavlovitchev, A.; Sidorenko, V.; Chetverikov, A.

2000-03-01T23:59:59.000Z

247

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

SciTech Connect (OSTI)

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.

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

2012-06-08T23:59:59.000Z

248

2nd Annual Workshop Proceedings EC FP FIRST-Nuclides 5th  

E-Print Network [OSTI]

phases, (2) quantification of H2O2 and H2 produced by water radiolysis and (3) determination of the UO2 surface by H2O2 produced by water radiolysis. We have verified that studtite is not formed to the inhibition effect of H2 produced by water radiolysis. In these conditions, G(H2O2) and G(H2) are respectively

Paris-Sud XI, Université de

249

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

DOE Patents [OSTI]

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.

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

2002-01-01T23:59:59.000Z

250

Head-end process for the reprocessing of HTGR spent fuel  

SciTech Connect (OSTI)

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)

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

2013-07-01T23:59:59.000Z

251

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

SciTech Connect (OSTI)

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.

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

252

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

SciTech Connect (OSTI)

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.

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

253

Report of Progress with Citrus Fruits at the Beeville Sub-Station, Bee County.  

E-Print Network [OSTI]

65-109-5m TEXAS AGRICULTURAL EXPERIMENT STATIONS. -T- I . ! '? . ........ ?' ? r ; 'V ? J 'V ? ? ? BULLETIN NO. 118. February, 1 9 0 9 . REPORT Of PROGRESS WITH CITRUS FRUITS AT THE BEEVILLE SUB-STATION, BEE COUNTY S. A. WASCHKA................. ................................Stenographer. A. S. F IB5 mmmmmmmmmmmmmmmmmmmmmmmmmmmmm mmmmmmmmmmmmmmm STATE SUBSTATIONS. H. H. R IBBC8 -1X8 mmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmDirector. W. S. R X1uoMCPP0 Superintendent...

Waschka, S. A.

1909-01-01T23:59:59.000Z

254

The relative basicity and reactivity of D-mannosamine  

E-Print Network [OSTI]

$$A IIIGQB UQJOJQlu 9T)0 '. JB Uof'. reu" I Jiroro QQTIOT080 QIP, $G 90UQ qV foui'. 'PQI&STI! Tr 'TP eq ggr'I O'0 noun ~ Qcyo9pgq . [ rl oqq Uo QGUQUQTU T . TOT UT. I Iq'c'00 "" g /nlrb. 0nq eu~=9oaul!UT r IG 980q;QJT e~g u L quoeosd oq gTjrl UGI08Teq0...

Carlo, Michael John

1962-01-01T23:59:59.000Z

255

Thorium oxide slurries as blankets in fissile producing fusion- fission hybrids  

E-Print Network [OSTI]

of the blanket as related to the energy of the incident fusion neutrons. CALCULATIONAL MODEL The MARS computer code package from the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) 11 was used to determine... the aqueous homogeneous power reactor program at ORNL was begun. Significant progress was made in studies of uranium oxide (UO ) and its slurries, and in the development of equipment for circulating slurries at concentrations of several hundred grams per...

Geer, Thomas Charles

1982-01-01T23:59:59.000Z

256

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

SciTech Connect (OSTI)

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).

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

257

P a g e | 1 Regional Ocean Modelling  

E-Print Network [OSTI]

). #12;P a g e | 4 Slide 4: Flather Condition for Shallow-Water Barotropic Flow: h/t = -Hu/x u/t = -gh be shown that for shallow-water eqn: u - (g/H)1/2 h is the incoming (i.e. left-running) characteristic for uo-c subcritical flows. Thus, either we set "u - (g/H)1/2 h" to be zero

258

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)

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.

M. Chen; CM Regan; D. Noe

2006-01-09T23:59:59.000Z

259

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

SciTech Connect (OSTI)

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.

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

260

Electrochemistry of LiCl-Li2O-H2O Molten Salt Systems  

SciTech Connect (OSTI)

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.

Natalie J. Gese; Batric Pesic

2013-03-01T23:59:59.000Z

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261

Complexation of U(VI) with 1-Hydroxyethane-1,1-diphosphonicAcid (HEDPA) in Acidic to Basic Solutions  

SciTech Connect (OSTI)

Complexation of U(VI) with 1-hydroxyethane-1,1-diphosphonic acid (HEDPA) in acidic to basic solutions has been studied with multiple techniques. A number of 1:1 (UO{sub 2}H{sub 3}L), 1:2 (UO{sub 2}H{sub j}L{sub 2} where j = 4, 3, 2, 1, 0 and -1) and 2:2 ((UO{sub 2}){sub 2}H{sub j}L{sub 2} where j = 1, 0 and -1) complexes form, but the 1:2 complexes are the major species in a wide pH range. Thermodynamic parameters (formation constants, enthalpy and entropy of complexation) were determined by potentiometry and calorimetry. Data indicate that the complexation of U(VI) with HEDPA is exothermic, favored by the enthalpy of complexation. This is in contrast to the complexation of U(VI) with dicarboxylic acids in which the enthalpy term usually is unfavorable. Results from electrospray ionization mass spectrometry (ESI-MS) and {sup 31}P NMR have confirmed the presence of 1:1, 1:2 and 2:2 U(VI)-HEDPA complexes.

Reed, W A; Rao, L; Zanonato, P; Garnov, A; Powell, B A; Nash, K L

2007-01-24T23:59:59.000Z

262

Uranyl coordination environment in hydrophobic ionic liquids : an in situ investigation.  

SciTech Connect (OSTI)

Different inner-sphere coordination environments are observed for the uranyl nitrate complexes formed with octyl-phenyl-N,N-diisobutylcarbamoylmethylphosphine oxide and tributyl phosphate in dodecane and in the hydrophobic ionic liquids (ILs) [C{sub 4}mim][PF{sub 6}] and [C{sub 8}mim][N(SO{sub 2}CF{sub 3}){sub 2}]. Qualitative differences in the coordination environment of the extracted uranyl species are implied by changes in peak intensity patterns and locations for uranyl UV-visible spectral bands when the solvent is changed. EXAFS data for uranyl complexes in dodecane solutions is consistent with hexagonal bipyramidal coordination and the existence of UO{sub 2}(NO{sub 3}){sub 2}(CMPO){sub 2}. In contrast, the complexes formed when uranyl is transferred from aqueous nitric acid solutions into the ILs exhibit an average equatorial coordination number of approximately 4.5. Liquid/liquid extraction results for uranyl in both ILs indicate a net stoichiometry of UO{sub 2}(NO{sub 3})(CMPO){sup +}. The concentration of the IL cation in the aqueous phase increases in proportion to the amount of UO{sub 2}(NO{sub 3})(CMPO){sup +} in the IL phase, supporting a predominantly cation exchange mechanism for partitioning in the IL systems.

Visser, A. E.; Jensen, M. P.; Laszak, I.; Nash, K. L.; Choppin, G. R.; Roers, R. D.; Chemistry; Univ. of Alabama; Flordia State Univ.

2003-01-01T23:59:59.000Z

263

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

SciTech Connect (OSTI)

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.

Simon R. Phillpot

2012-06-08T23:59:59.000Z

264

Irradiation of SiC Clad Fuel Rods in the HFIR  

SciTech Connect (OSTI)

During 2009 and- 2010, new test capability for the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was developed that allows testing of advanced nuclear fuels and cladding under prototypic light-water-reactor (LWR) operating conditions (i.e., cladding and fuel temperatures, fuel average linear heat generation rates, and cladding fluence). For the initial experiments for this test program, ORNL teamed with commercial fuel/cladding vendors who have developed an advanced composite-wound SiC cladding material for possible use in LWRs. The first experiment, containing SiC-clad UN fuel, was inserted in HFIR in June 2010, and the second experiment, containing SiC-clad UO2 fuel, was inserted in October 2010. Two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in November 2011 at an estimated fuel burnup of approximately 10 GWd/MTHM; and two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in February 2013 at an estimated fuel burnup of approximately 20 GWd/MTHM. These capsules are currently awaiting PIE. This paper will describe the experiment, as-run operating conditions for these capsules, and current PIE plans and status.

Ott, Larry J [ORNL] [ORNL; Bell, Gary L [ORNL] [ORNL; Ellis, Ronald James [ORNL] [ORNL; McDuffee, Joel Lee [ORNL] [ORNL; Morris, Robert Noel [ORNL] [ORNL

2013-01-01T23:59:59.000Z

265

Molecular Dynamics Simulation of Thermodynamic Properties in Uranium Dioxide  

SciTech Connect (OSTI)

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

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

2014-03-01T23:59:59.000Z

266

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

SciTech Connect (OSTI)

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.

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

1997-01-30T23:59:59.000Z

267

Theory of Deep Impurity Levels in Cucl  

E-Print Network [OSTI]

, respectively, + 0.59 ~ e ~ and ?0.59 ~ e ~ . E(s,c) E(s,a) E(p, c) E(p, a) E(d,c) E'{d,c) V(x?yz,) 2.80 ?15.15 9.00 ?3.75 ?1.25 ?1.90 V(s,s) ?2.877 V(x?s,) 4.841 V(s?x,) 2.866 V(x,x) 0 V(x,y) 0 V,d ?1.980 ?5.085 1.220 quirements... than C13p Cu 3d character at top of valence band C13p character at top of valence band 75%%uo 25rob 75%%uo 25%%uo Band gap (eV) 3.25' 3.25 Photoemission peaks (eV) B C D 0.8 ?1.4 1.9?2.6b 4.9?5.2 6.0?6.3 0.6 1.9 4.8 6.4 Width...

REN, SY; Allen, Roland E.; DOW, JD; LEFKOWITZ, I.

1982-01-01T23:59:59.000Z

268

Standard specification for sintered gadolinium oxide-uranium dioxide pellets  

E-Print Network [OSTI]

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...

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

269

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

SciTech Connect (OSTI)

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.

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

270

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

SciTech Connect (OSTI)

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.

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

2011-08-01T23:59:59.000Z

271

Criticality Safety Code Validation with LWBR’s SB Cores  

SciTech Connect (OSTI)

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.

Putman, Valerie Lee

2003-01-01T23:59:59.000Z

272

In situ treatment of VOCs by recirculation technologies  

SciTech Connect (OSTI)

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.

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

273

Experimental Results for SimFuels  

SciTech Connect (OSTI)

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.

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

274

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

SciTech Connect (OSTI)

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.

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

2008-11-03T23:59:59.000Z

275

Kinematics and thermodynamics across a propagating non-stoichiometric oxidation phase front in spent fuel grains  

SciTech Connect (OSTI)

Spent fuel contains mixtures, alloy and compound, but are dominated by U and O except for some UO{sub 2} fuels with burnable poisons (gadolinia in BWR rods), the other elements evolve during reactor operation from neutron reaction and fission + fission decay events. Due to decay, chemical composition and activity of spent fuel will continue to evolve after removal from reactors. During the time interval with significant radioactivity levels relevant for a geological repository, it is important to develop models for potential chemical responses in spent fuel and potential degradation of repository. One such potential impact is the oxidation of spent fuel, which results in initial phase change of UO{sub 2} lattice to U{sub 4}O{sub 9} and the next phase change is probably to U{sub 3}O{sub 8} although it has not been observed yet below 200C. The U{sub 4}O{sub 9} lattice is nonstoichiometric with a O/U weight ratio at 2.4. Preliminary indications are that the UO{sub 2} has a O/U of 2. 4 at the time just before it transforms into the U{sub 4}O{sub 9} phase. In the oxygen weight gain versus time response, a plateau appears as the O/U approaches 2.4. Part of this plateau is due to geometrical effects of a U{sub 4}O{sub 9} phase change front propagating into UO{sub 2} grain volumes; however, this may indicate a metastable phase change delay kinetics or a diffusional related delay time until the oxygen density can satisfy stoichiometry and energy conditions for phase changes. Experimental data show a front of U{sub 4}O{sub 9} lattice structure propagating into grains of the UO{sub 2} lattice. To describe this spatially inhomogenous oxidation phase transition, as well as the expected U{sub 3}O{sub 8} phase transition from the U{sub 4}O{sub 9} lattice, lattice models are developed and spatially discontinuous kinematic and energetic expressions are derived. 9 refs.

Stout, R.B.; Kansa, E.J.; Wijesinghe, A.M.

1993-09-01T23:59:59.000Z

276

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

SciTech Connect (OSTI)

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

Rasmussen B.

2010-01-01T23:59:59.000Z

277

Pillared and open-framework uranyl diphosphonates  

SciTech Connect (OSTI)

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.

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

278

SANS study of third phase formation in the U(VI)-HNO{sub 3}/ TBP-n-dodecane system.  

SciTech Connect (OSTI)

In spite of its technological importance, third phase formation in the extraction of hexavalent actinides from nitric acid solutions into alkane solutions of tri-n-butylphosphate (TBP) has received only limited attention. The focus of the few available literature works has been primarily centered on the composition of the third phase and on the stoichiometry of the metal complexes. Very little is known, on the other hand, about the structure and morphology of the third phase species of hexavalent actinides. In the present investigation, the formation of a third phase upon extraction of U(VI) by 20% TBP in deuterated n-dodecane from nitric acid solutions was studied. Chemical analyses have shown that U(VI) exists in the third phase as a species having the composition UO{sub 2}(NO{sub 3}){sub 2}{center_dot}(TBP){sub 2}{center_dot}HNO{sub 3}. Small-angle neutron scattering measurements on TBP solutions loaded with only HNO{sub 3} or with increasing amounts of U(VI) have revealed the presence, both before and after phase splitting, of relatively large ellipsoidal aggregates with the parallel and perpendicular axes having lengths up to about 64 and 15 Angstroms, respectively. The formation of these aggregates is observed in all cases, that is, when only HNO3, only UO{sub 2}(NO{sub 3}){sub 2}, or both HNO{sub 3} and UO{sub 2}(NO{sub 3}){sub 2} are extracted by the TBP solution. Upon third phase formation, the SANS data reveal the presence of smaller aggregates in both the heavy and light organic phase.

Chiarizia, R.; Jensen, M. P.; Borkowski, M.; Ferraro, J. R.; Thiyagarajan, P.; Littrell, K. C.

2003-01-01T23:59:59.000Z

279

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

SciTech Connect (OSTI)

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.

Todd R. Allen, Director

2011-04-01T23:59:59.000Z

280

FEASIBILITY STUDY OF DUPOLY TO RECYCLE DEPLETED URANIUM.  

SciTech Connect (OSTI)

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.

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

1998-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Roughage and roughage substitutes in high concentrate finishing mixtures for beef cattle  

E-Print Network [OSTI]

different levels of roughage, showed that maximum levels of 20 to 30% cottonseed hulls, 20 to 30% coastal bermuda hay, 10 to 20/o rice hulls (ammoniated or non-ammoniated) or 10/o flax shives should be used in finishing mixtures if high gain and feed... into four uniform groups on the basis of weight and grade. These groups received four different feed mixtures as follows: all concentrate, 2 and 4%%uo oyster shell flakes and 10% ammoniated rice hulls. The second and third experiments were part of Texas...

Leigh, Jorge Eduardo

1968-01-01T23:59:59.000Z

282

Liquid Metal Bond for Improved Heat Transfer in LWR Fuel Rods  

SciTech Connect (OSTI)

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.

Donald Olander

2005-08-24T23:59:59.000Z

283

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

SciTech Connect (OSTI)

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.

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

1980-01-01T23:59:59.000Z

284

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

SciTech Connect (OSTI)

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.

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

285

Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications  

SciTech Connect (OSTI)

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.

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

2011-11-01T23:59:59.000Z

286

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

SciTech Connect (OSTI)

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)].

Ni, Chengbao; Shuh, David; Raymond, Kenneth

2011-03-07T23:59:59.000Z

287

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment  

SciTech Connect (OSTI)

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.

Rodney C. Ewing

2004-10-07T23:59:59.000Z

288

Spatially resolved U(VI) partitioning and speciation: Implications for plume scale behavior of contaminant U in the Hanford vadose zone  

SciTech Connect (OSTI)

A saline-alkaline brine containing high concentrations of U(VI) was accidentally spilled at the Hanford Site in 1951, introducing 10 tons of U into sediments under storage tank BX-102. U concentrations in the deep vadose zone and groundwater plumes increase with time, yet how the U has been migrating is not fully understood. We simulated the spill event in laboratory soil columns, followed by aging, and obtained spatially resolved U partitioning and speciation along simulated plumes. We found after aging, at apparent steady state, that the pore aqueous phase U concentrations remained surprisingly high (up to 0.022 M), in close agreement with the recently reported high U concentrations (up to 0.027 M) in the vadose zone plume (1). The pH values of aged pore liquids varying from 10 to 7, consistent with the measured pH of the field borehole sediments varying from 9.5 to 7.4 (2), from near the plume source to the plume front. The direct measurements of aged pore liquids together with thermodynamic calculations using a Pitzer approach revealed that UO{sub 2}(CO{sub 3}){sub 3} {sup 4-} is the dominant aqueous U species within the plume body (pH 8-10), while Ca{sub 2}UO{sub 2}(CO{sub 3}){sub 3} and CaUO{sub 2}(CO{sub 3}){sub 3}{sup 2-} are also significant in the plume front vicinity (pH 7-8), consistent with that measured from field borehole porewaters (3). U solid phase speciation varies at different locations along the plume flow path and even within single sediment grains, because of location dependent pore and micropore solution chemistry. Our results suggest that high geochemical stability of UO{sub 2}(CO{sub 3}){sub 3}{sup 4-} in the original carbonate and sodium rich waste solution permits its continues migration and the field observed increases of U concentrations in the vadose zone and groundwater.

Wan, Jiamin; Kim, Yongman; Tokunaga, Tetsu K.; Wang, Zheming; Dixit, Suvasis; Steefel, Carl; Saiz, Eduardo; Kunz, Martin; Tamura, Nobumichi

2009-02-01T23:59:59.000Z

289

Kapitza Resistance of the Grain Boundaries in Ceria  

SciTech Connect (OSTI)

Thermal conductivity is one of the key performance metrics of the nuclear fuels. In electrical insulators, such as most ubiquitous nuclear fuel – UO2, thermal transport is due to phonons, or lattice waves. Their propagation is impeded by any lattice defect, such as impurities or vacancies, as well as larger microstructural features: grain boundaries, dislocations and pores/bubbles. Detailed description of the phonons interactions with these features is still lacking. In this work, we elucidate the dependence of the grain boundary thermal resistance, also known as a Kapitza resistance, on the type and misorientation angle of the grain boundary in model system of CeO2.

David Bai; Jian Gan; Aleksandr Chernatynskiy

2014-06-01T23:59:59.000Z

290

Newberry EGS Seismic Velocity Model  

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

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.

Templeton, Dennise

291

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

SciTech Connect (OSTI)

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.

BD Hanson; J Abrefah; SC Marschman; SG Prussin

2000-09-08T23:59:59.000Z

292

The effect of dietary vitamin D level on the generation of protective cellular immunity against pulmonary tuberculosis  

E-Print Network [OSTI]

by inhibiting proliferation of antibody-producing B cells(77) or by an indirect effect on T helper-cell activity(39, 78). Moreover, calcitriol down-regulated expression of gamma interferon and granulocyte- macrophage colony-stimulating factor (GM-CSF) genes... ~ lymphoproliferation of peripheral blood lymphocytes taken from guinea pigs h 11 g d with H. t b 1o 9 Effect of dietary vitamin D status on ~o lymphoproliferation of guinea pig spleen cells after challenged with g~e~uos is. 10 Effect of dietary vitamin D level...

Hernandez-Frontera, Evaurely

1990-01-01T23:59:59.000Z

293

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

SciTech Connect (OSTI)

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.

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

294

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

E-Print Network [OSTI]

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

Buckley, Frank A.

1951-01-01T23:59:59.000Z

295

Facility Operations 1993 fiscal year work plan: WBS 1.3.1  

SciTech Connect (OSTI)

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).

Not Available

1992-11-01T23:59:59.000Z

296

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

SciTech Connect (OSTI)

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.

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

2011-01-12T23:59:59.000Z

297

Newberry EGS Seismic Velocity Model  

SciTech Connect (OSTI)

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.

Templeton, Dennise

2013-10-01T23:59:59.000Z

298

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents [OSTI]

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

Friedman, Horace A. (Oak Ridge, TN)

1985-01-01T23:59:59.000Z

299

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents [OSTI]

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.

Friedman, H.A.

1984-06-13T23:59:59.000Z

300

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

E-Print Network [OSTI]

), 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...

Osborn, Jeremy

2014-08-13T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

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

SciTech Connect (OSTI)

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.

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

2009-12-12T23:59:59.000Z

302

Monopole Strength in Ni-58  

E-Print Network [OSTI]

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...

Youngblood, David H.; Lui, YW.

1991-01-01T23:59:59.000Z

303

Application of horizontal wells in steeply dipping reservoirs  

E-Print Network [OSTI]

) + Rplj)/2 Np = Np(j-1) + DELTA Np R3 - Rs +[(kg/ko)(Uo/Ug)(Bo/Bg)] Gp = Gp 4-1) + DELTA NpsRp avg Rec. Fac. = Np/OOIP Rp=Gp/Np 43 IE+06 lE+05 a 1E+04 1E+03 ~m=000 ? a ? m=0, 08 ~m=0. 24 1965 1970 1975 1980 1985 1990 1995 Time (years) Fig... Major Subject: Petroleum Engineering ABSTRACT Application of Horizontal Wells in Steeply Dipping Reservoirs. (December 1995) Jose David Lopez Navarro, B. S. , Universidad de America Santafe de Bogota, Colombia Co-chairs of Advisory Committee: Dr...

Lopez Navarro, Jose David

1995-01-01T23:59:59.000Z

304

A VSP transformation technique for the determination of subsurface structure  

E-Print Network [OSTI]

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...

Malloy, Jeffrey Edward

1985-01-01T23:59:59.000Z

305

Time?resolved anisotropic coherent anti?Stokes Raman scattering: A new probe of reorientational dynamics  

E-Print Network [OSTI]

anti-Stokes Raman scattering (TRA CARS) and spontaneous Raman (TRA Raman) as probes of rota­ tional motion. II. THEORY When a sample system is illuminated by three laser beams at frequencies (Uo, (UI, and (U2, the incident fields induce a third... is described by a nonlinear sus­ ceptibility tensor containing three independent compo­ nents in the L=O subspace and six independent compo­ nents in the L = 2 subspace. First we write the molecular susceptibility MU11(0) iIi terms of the direct product...

Wan, Chaozhi; Johnson, Carey K.

1993-09-01T23:59:59.000Z

306

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

SciTech Connect (OSTI)

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

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

1995-05-01T23:59:59.000Z

307

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

SciTech Connect (OSTI)

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

Plaster, M.J.

1997-05-01T23:59:59.000Z

308

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

SciTech Connect (OSTI)

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.

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

1983-01-01T23:59:59.000Z

309

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

SciTech Connect (OSTI)

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.

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

2011-09-01T23:59:59.000Z

310

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

SciTech Connect (OSTI)

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.

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

2008-02-15T23:59:59.000Z

311

Uranium(VI) extraction by TBP in the presence of HDBP  

SciTech Connect (OSTI)

The influence of di-n-butyl phosphoric acid (HDBP) upon extraction of uranium(VI) by tri-n-butyl phosphate (TBP) from 0.5--3.0 M nitric acid solutions has been studied. It has been shown that the uranium(VI) distribution coefficient D{sub U} for extraction by 1.1 M TBP in tri-decane or xylene is increased when HDBP is present in the organic phase. For iso-molar solutions of (TBP + HDBP) with a total concentration of 0.36 M, and Uranium(VI) aqueous concentration up to 10--20 g/l, a maximum value of D{sub U} is observed when TBP/HDBP = 1; for higher U(VI) concentration the maximum gradually disappears, with D{sub U} growing monotonically with the HDBP content in the organic phase. Uranium(VI) absorption spectra for 1.1 M TBP in tri-decane or xylene, containing HDBP, provide evidence for the formation of compounds, of which composition is intermediate between uranyl nitrate--TBP disolvate and the U(VI)--HDBP complex. It is proposed that these intermediate compounds are UO{sub 2}(NO{sub 3}){sub 2}HDBP.TBP and UO{sub 2}(NO{sub 3}){sub 2}(HDBP){sub 2}.

Fedorov, Yu.S.; Zilberman, B.Ya.; Kulikov, S.M.; Blazheva, I.V.; Mishin, E.N. [V.G. Khlopin Radium Inst., Saint-Petersburg (Russian Federation); Wallwork, A.L.; Denniss, I.S.; May, I. [British Nuclear Fuels plc, Sellafield (United Kingdom); Hill, N.J. [British Nuclear Fuels plc, Risley (United Kingdom)

1999-03-01T23:59:59.000Z

312

Interfacial Complex Formation in Uranyl Extraction by Tributyl-Phosphate in Dodecane Diluent: A Molecular Dynamics Study  

SciTech Connect (OSTI)

Atomistic simulations have been carried out in a multicomponent two-phase system (aqueous and organic phases in direct contact) to investigate the interfacial molecular mechanisms leading to uranyl extractionfrom the aqueous to organic phase. The aqueous phase consists of the dissolved ions UO2^2+ and nitrate NO3-,with or without H3O+, in water to describe acidic or neutral condition; the organic phase consists of tributyl phosphate, the extractant, in dodecane as the diluent. We find that the interface facilitates the formation of various uranyl complexes, with a general formula UO2^2+(NO3-)n mTBP kH2O, with n + m + k ) 5, suggesting a 5-fold coordination. The coordination for all three molecular entities has the common feature that they all bind to the uranyl at the uranium atom with an oxygen atom in the equatorial plane perpendicular to the molecular axis of the uranyl, forming a 5-fold symmetry plane. Nitric acid has a strong effect in enhancing the formation of extractable species, which is consistent with experimental findings.

de Almeida, Valmor F [ORNL; Cui, Shengting [ORNL; Ye, Xianggui [ORNL; Khomami, Bamin [ORNL

2009-01-01T23:59:59.000Z

313

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

SciTech Connect (OSTI)

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.

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

2015-01-01T23:59:59.000Z

314

Process for continuous production of metallic uranium and uranium alloys  

DOE Patents [OSTI]

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.

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

1995-06-06T23:59:59.000Z

315

Process for continuous production of metallic uranium and uranium alloys  

DOE Patents [OSTI]

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.

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

1995-01-01T23:59:59.000Z

316

Standard test methods for arsenic in uranium hexafluoride  

E-Print Network [OSTI]

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...

American Society for Testing and Materials. Philadelphia

2005-01-01T23:59:59.000Z

317

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

SciTech Connect (OSTI)

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.

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

1981-12-01T23:59:59.000Z

318

Methodology for determining criteria for storing spent fuel in air  

SciTech Connect (OSTI)

Dry storage in an air atmosphere is a method being considered for spent light water reactor (LWR) fuel as an alternative to storage in an inert gas environment. However, methods to predict fuel integrity based on oxidation behavior of the fuel first must be evaluated. The linear cumulative damage method has been proposed as a technique for defining storage criteria. Analysis of limited nonconstant temperature data on nonirradiated fuel samples indicates that this approach yields conservative results for a strictly decreasing-temperature history. On the other hand, the description of damage accumulation in terms of remaining life concepts provides a more general framework for making predictions of failure. Accordingly, a methodology for adapting remaining life concepts to UO/sub 2/ oxidation has been developed at Pacific Northwest Laboratory. Both the linear cumulative damage and the remaining life methods were used to predict oxidation results for spent fuel in which the temperature was decreased with time to simulate the temperature history in a dry storage cask. The numerical input to the methods was based on oxidation data generated with nonirradiated UO/sub 2/ pellets. The calculated maximum allowable storage temperatures are strongly dependent on the temperature-time profile and emphasize the conservatism inherent in the linear cumulative damage model. Additional nonconstant temperature data for spent fuel are needed to both validate the proposed methods and to predict temperatures applicable to actual spent fuel storage.

Reid, C.R.; Gilbert, E.R.

1986-11-01T23:59:59.000Z

319

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

SciTech Connect (OSTI)

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.

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

1997-08-01T23:59:59.000Z

320

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

SciTech Connect (OSTI)

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.

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

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

TREKiSM Issue 34  

E-Print Network [OSTI]

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...

1984-01-01T23:59:59.000Z

322

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

SciTech Connect (OSTI)

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.

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

1980-01-01T23:59:59.000Z

323

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

SciTech Connect (OSTI)

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.

Simon Phillpot; James Tulenko

2011-09-08T23:59:59.000Z

324

Potential incorporation of transuranics into uranium phases  

SciTech Connect (OSTI)

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.

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

1999-12-07T23:59:59.000Z

325

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

SciTech Connect (OSTI)

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.

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

1996-12-30T23:59:59.000Z

326

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

SciTech Connect (OSTI)

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.

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

2006-11-01T23:59:59.000Z

327

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts  

SciTech Connect (OSTI)

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.

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

2012-04-01T23:59:59.000Z

328

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

SciTech Connect (OSTI)

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.

DOE

1997-04-01T23:59:59.000Z

329

Synthetic jets at large Reynolds number and comparison to continuous jets  

SciTech Connect (OSTI)

Experimental measurements and flow visualization of synthetic jets and similar continuous jets are described. The dimensionless stroke length necessary to form a 2-D synthetic jet is between 5 and 10, with wider-nozzle jets consistently requiring a smaller value. Synthetic jets are wider, slower and have more momentum than similar continuous jets. Synthetic jets are generated using four nozzle widths that vary by a factor of four, and the driving frequency is varied over an order of magnitude. The resultant jets are in the range 13.5 < L{sub o}/h < 80.8 and 695 < Re{sub Uo} < 14700. In spite of the large range of stroke lengths, the near-field behavior of the synthetic jets scales with L{sub o}/h.

Smith, B. L. (Barton L.); Swift, G. W. (Gregory W.)

2001-01-01T23:59:59.000Z

330

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

SciTech Connect (OSTI)

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.

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

1983-01-01T23:59:59.000Z

331

Electroslag Remelting (ESR) Slags for Removal of Radioactive Oxide Contaminants from Stainless Steel, Annual Report (1998-1999)  

SciTech Connect (OSTI)

Decontamination of radioactive contaminated stainless steel using the ESR process is investigated by conducting thermophysical and thermochemical laboratory studies on the slag. The ESR base slag investigated in this research project is 60wt%CaF{sub 2}-20wt%CaO-20wt%Al{sub 2}O{sub 3}. In this report, we present the data obtained to date on relevant slag properties, capacity to incorporate the radioactive contaminant (using CeO{sub 3}) as surrogate, simulant for PUO{sub 2} and UO{sub 2}, slag-metal partition coefficient, volatilization rate and volatile species, viscosity, electrical conductivity and surface tension as a function of temperature. The impact of these properties on the ESR decontamination process is presented.

PAL, UDAY B.

1999-08-01T23:59:59.000Z

332

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

SciTech Connect (OSTI)

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

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

2013-07-18T23:59:59.000Z

333

Numerical solution of the Navier-Stokes equations in the entry region of a straight tube  

E-Print Network [OSTI]

) aP( o '6xlt, x, q, ? Klt, x, q, c& 5 , x, q a ap~ ~o $6 I+, x, qm gf, , ?, ?-o &e) Kl ~o H. lt, x, q, ? p end (s are constants (9) Equations 1, 2, 3, and 4 are the Navier-Stokes equations and the continuity equation in cylindrical coordinates... equations become ua& + v i4 - & k + I p~u a& a~ u~+ vP =$g + ? '(W--") (io) Equations 5' and 6 ', upon similar manipulation, become $(o r) 0 u(o, r) ) (i3) V(O, I )~Q u(z. , i)- o y(z, i) = 0 (&) Q(Z, P) ~ 0 Equations 10, 11, 12, 13, and 14...

Little, James Gilbert

2012-06-07T23:59:59.000Z

334

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

SciTech Connect (OSTI)

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.

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

1991-12-31T23:59:59.000Z

335

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

SciTech Connect (OSTI)

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}.

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

1991-12-31T23:59:59.000Z

336

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

SciTech Connect (OSTI)

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.

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

2012-05-15T23:59:59.000Z

337

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

SciTech Connect (OSTI)

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.

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

338

Transient Testing of Nuclear Fuels and Materials in United States  

SciTech Connect (OSTI)

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 .

Daniel M. Wachs

2012-12-01T23:59:59.000Z

339

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

SciTech Connect (OSTI)

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)

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

340

Movement stereotypes in control-display relationships  

E-Print Network [OSTI]

dP I K~D u D ff I pl 3 m d u Q Inn 6 D pl y din f ~ P h/P II Cn I ID Im Do/D u S pi S d uo D ff I pl Pq d I d Io Q su* 7 D/spl y dP /Dmb D rmuo I 3 D apl dP f d Pushy' ll Co col D Q bo 3 D playa dpm/e Knnb D? Df/ tpl P nmml ula... to item five. . . 7 Percentage response breakdowns to item six 8 Percentage response breakdowns to item seven. . . . 9 Percentage response breakdowns to item eight. . . . . 10 Percentage response breakdowns to item nine 11 Percentage response...

Wright, Samantha Jane

2012-06-07T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Modeling of the simultaneous extraction of nitric acid and uranyl nitrate with tri-n-butyl phosphate. Application to extraction operation  

SciTech Connect (OSTI)

A mathematical model developed for the equilibrium HNO{sub 3}-UO{sub 2}(NO{sub 3}){sub 2}-tri-n-butyl phosphate (TBP)-diluent is the basis of the computation of distribution isotherms. The isotherms are used to study the influence of TBP concentration on two chosen operation parameters, distribution coefficients and number of theoretical stages, for the selected flow sheets. It is established that an increase in TBP concentration leads to a decrease in the number of theoretical stages for the extraction flow sheets but to their increase for the striping flow sheets. Given diagrams can be used to determine the efficiency of extraction processes. Agreement with available literature calculations on the number of theoretical stages supports the use of the model in the computation of distribution isotherms, of the system quoted above, in a wide range of nitric acid, uranyl nitrate, and TBP concentrations.

Comor, J.J.; Tolic, A.S.; Kopecni, M.M.; Petkovic, D.M. [Vinca Inst. of Nuclear Sciences, Belgrade (Yugoslavia). Chemical Dynamics Lab.] [Vinca Inst. of Nuclear Sciences, Belgrade (Yugoslavia). Chemical Dynamics Lab.

1999-01-01T23:59:59.000Z

342

The PACSAT Communications Experiment (PCE)  

SciTech Connect (OSTI)

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.

Not Available

1993-02-12T23:59:59.000Z

343

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

SciTech Connect (OSTI)

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.

Not Available

1993-02-12T23:59:59.000Z

344

Preliminary Investigation of Candidate Materials for Use in Accident Resistant Fuel  

SciTech Connect (OSTI)

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

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

2013-09-01T23:59:59.000Z

345

Experimental versus theoretical comparison of the effects of taper and static eccentricity on the rotordynamic coefficients of short, smooth, high-speed, liquid annular seals  

E-Print Network [OSTI]

. 24 E 0. 085 Q 0. 080 g 0. 075 Pu0 070 0. 065 uO 060 0. 055 ~0 050 10200 rpm 17400 rpm t) 24600 rpm 3-02-01 00 01 02 03 E 24k MPa EO0 0) 0, 080 s& PI5 0. 075 Pu 0. 070 C3 0. 065 a 0. 060 0. 055 ~0 050 3-02-01 00 01 02 03 EO 085 Q...) 0 080 s& F50 075 o 0. 070 0. 065 u 0. 060 0 055 ~ 0, 050 3-02 ? 01 00 0. 1 02 03 ? 0. Taper Parameter Fig. 9 Minimum radial clearance versus taper parameter for all operating condittons. 25 Stiffness The direct stiffness is used...

Lindsey, William Todd

2012-06-07T23:59:59.000Z

346

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)

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.

Shott, Gregory [NSTec

2014-08-31T23:59:59.000Z

347

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

E-Print Network [OSTI]

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.

Igor Iosilevskiy; Victor Gryaznov

2010-05-23T23:59:59.000Z

348

Uranium-contaminated soils: Ultramicrotomy and electron beam analysis  

SciTech Connect (OSTI)

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.

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

1994-02-01T23:59:59.000Z

349

Repository-relevant testing applied to the Yucca Mountain Project  

SciTech Connect (OSTI)

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.

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

1989-04-01T23:59:59.000Z

350

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

SciTech Connect (OSTI)

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.

Not Available

1986-08-01T23:59:59.000Z

351

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

SciTech Connect (OSTI)

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.

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

1983-01-01T23:59:59.000Z

352

Zirconium in the nuclear industry  

SciTech Connect (OSTI)

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).

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

1984-01-01T23:59:59.000Z

353

An accounting system for the Brazos valley cotton cooperative association  

E-Print Network [OSTI]

ttta 0 ~t tt 0 - aomca 19 ~Gt 1 S~h i the 'N~ ~Gti ~ P bit ~llt111 ~Dta - %hi 1 11 ha I t 4eposits for gas, lights, an4 water. nea m ae ~~a c~s o an~c~ 5 1 A~Ill e1 t~Satetii1 t I Cll 4&-SI1 el charged and provisioaal sales proceeds for purohase... 14; sny d1ff'?rene? 1s charge& to overpayaoats. Xn&ioation should also bo aa4e as to whether tho ovwrpsyasat wes 4uo to aa error 1n ths asrket prise or as to the extensions end basis &?¬ions. Roger&loss of tho prise ogre?4 upoa, such prise...

Simpson, William Maurice

1939-01-01T23:59:59.000Z

354

Artificial neural networks for input-output dynamic modeling of nonlinear processes  

E-Print Network [OSTI]

&at& tire behavior of n&arlinear SIS(2 and 1&IIMO pro&. esses, provi&1& d that tlrv latter operate closv vnou?h t&& dvsired operating points. In the follow'irrg &lrapters, sex&'ra) rrretlrods of non(&near modeling will be used to u&o&1&'I tl&e samv nonliu... the fnlh)&vh)r, e(tu&(t&nn: R M , V = 1 + gP, + P (1, (3) a(&cl o 11 n&1 u . . u 1?v T= (('a(v (V u st I I? tt 2 . . . n';If N is the 3I x(&1 n&atrix of parameters. ln this n&atrix u& & ls the parameter of c(u&n( ction bet&veen tth model input a...

Sarimveis, Haralambos

2012-06-07T23:59:59.000Z

355

The influence of mechanical summer pruning, row direction, and tree spacing on yield and quality of peach, Prunus persica (L.) Batsch  

E-Print Network [OSTI]

obtained in plants spaced 4. 6 m apart (Table 5). The same was true for medium-size fruits in the following year (Table 3). The percentage of 23 6 tel Ql 0 Ql hl W 6 0 S 6 S 4 IJ 4 W td L' 0 dh O CQ O Qh QJ ht IJ JDI Ql a 0 0 0 C... m m OO '0 Ql m Ql m 6 6 td N M 0 6 CO C 0 C N & N r Cc e r Uo r Ch N I Cl M 6 E E IO IU '0 4 C 0 I M O C td 0 OO JJ E 0 C '0 Ql Qj E 6 0 Ilj I-t m tc! dj I tU 6 cd Qj td IU tt! C l td m 6 Ol 6 5 0 0 6...

Raseira, Ailton

2012-06-07T23:59:59.000Z

356

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]

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.

American Society for Testing and Materials. Philadelphia

2000-01-01T23:59:59.000Z

357

Criticality experiments with fast flux test facility fuel pins  

SciTech Connect (OSTI)

A United States Department of Energy program was initiated during the early seventies at the Hanford Critical Mass Laboratory to obtain experimental criticality data in support of the Liquid Metal Fast Breeder Reactor Program. The criticality experiments program was to provide basic physics data for clean well defined conditions expected to be encountered in the handling of plutonium-uranium fuel mixtures outside reactors. One task of this criticality experiments program was concerned with obtaining data on PuO{sub 2}-UO{sub 2} fuel rods containing 20--30 wt % plutonium. To obtain this data a series of experiments were performed over a period of about twelve years. The experimental data obtained during this time are summarized and the associated experimental assemblies are described. 8 refs., 7 figs.

Bierman, S.R.

1990-11-01T23:59:59.000Z

358

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

E-Print Network [OSTI]

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...

Hennington, Debra Louise

1981-01-01T23:59:59.000Z

359

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

DOE Patents [OSTI]

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

Menlove, Howard O. (Los Alamos, NM); Stewart, James E. (Los Alamos, NM)

1986-01-01T23:59:59.000Z

360

A shipping plan for bulk petroleum products by sea-going tankers  

E-Print Network [OSTI]

&k'rs can c &i y ro'. e L'r iu c & ' cui "o ' L oii I ', Iu in&lose?1. ':oci! Soir ?. . ; Lhe '(lie pili'. r )ji i&ui'i r)i'c, Uc?in i&)? ucd I)&I)o&c [2] is, 1 foJJou-iip Lo uo 'irlicr pi&)rr by Dent:zirg, :ir' IruJI &i)son [4]. Ct deals with a ahl ip...!& SOIIOTTGV 2. l G it"ii ? o' Vti!ji, d '!!ie fin&. ciiarL in. I i; iirc. I p' ~ v id & s, i u id &&! & n f ir thc na j or sL cps I al. cn in &olving tlit pro!ii e?. T&ie pro. i, &n begins by v&, . d iiig in ' I e ini Lial info & i& ion I iles mc &tinned n...

Boyd, David George

2012-06-07T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Fully Ceramic Microencapsulated Fuel Development for LWR Applications  

SciTech Connect (OSTI)

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.

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

2012-01-01T23:59:59.000Z

362

Enhanced Accident Tolerant LWR Fuels: Metrics Development  

SciTech Connect (OSTI)

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.

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

2013-09-01T23:59:59.000Z

363

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

SciTech Connect (OSTI)

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.

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

1994-01-01T23:59:59.000Z

364

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

E-Print Network [OSTI]

(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...

Smith, Johnny Melven

1974-01-01T23:59:59.000Z

365

Uranium dioxide electrolysis  

DOE Patents [OSTI]

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.

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

2009-12-29T23:59:59.000Z

366

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

SciTech Connect (OSTI)

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

Chodak, P. III; Buksa, J.J. [Los Alamos National Lab., NM (United States). Nuclear Systems Design and Analysis Group

1997-06-01T23:59:59.000Z

367

Marketing aspects of range sheep production in Texas  

E-Print Network [OSTI]

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...

Blackwell, James Wayne

1958-01-01T23:59:59.000Z

368

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

DOE Patents [OSTI]

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.

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

1985-02-04T23:59:59.000Z

369

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

SciTech Connect (OSTI)

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.

Pritam Chakraborty; Michael R. Tonks; Giovanni Pastore

2014-09-01T23:59:59.000Z

370

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

SciTech Connect (OSTI)

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.

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

2011-10-01T23:59:59.000Z

371

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

E-Print Network [OSTI]

X6ogaag aqvag~g gofeg Zest ~sW RQ5tRXW 'gC! 8gageM $o saWap sr& ~g skusms~nbsz a~ go guouq;yygyny ~y~d anzac, go aGaygay ysayuaqasg pun gegx~ynoyg&y . a~ go g~g Sqsnyal~ s~ oq ps'q, qygqng uyAnaqg, ~g note'g sgsa~ 8VRKTi XiTigOQD 'Qgf 8 f?8...[gg ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ l49$8Eg Ue'fO jAGPIQ ~ ~ eg SU~SQQfgg QCpg UL. 8 6C IICglleg Spell $88+ +CJOY ~ ~ ~ ~ tIaqurag auoqse~g geeIg UaSIog ~ ~ ' ~ ~ ~ ' ~ Iecpag ouoyepu 9 Seger' Ze o ~ ~ ~ t t t t t UO'pMIGg SUIS@gg ez 9z ~ ~ ~ ~ ~ Iacyusg ouoqspu. -s...

Chauvin, Aaron Lawrence

1962-01-01T23:59:59.000Z

372

Low cost, self-built housing  

E-Print Network [OSTI]

! E, , ;:;; -'. ':;', LOW. CaSTp ''MLF~SUQT, . ' HQN PlQ". -:: , "A %heeled Bj' 9; K, Vatter A~ is 4o tp1e and content lbgc l I J l I LCM 4(8%')-, SgF BUXKT- HGUGXH5 f parthiL fuff~ af t, he xsqaimmnvs fox. the Centres of me X... . mlmmUOTXON I PNX' XX BRStjjLRX OF PBQKQT &O ~ . 2/1 Page %los, Page Po" . $' Papa Vo. X6 Page Uo. 20 44 ~4 41 8 The Zp Pa. ~ Struc4oae . The A Pram@ BeehmKeci Page Ho 2O Page No, 2$ Page Mo, 2C Page Fo. Pago 7lo. PX P?iGTXGaTXQH XX...

Vetter, Gale K

1955-01-01T23:59:59.000Z

373

Effect of boron and gadolinium concentration on the calculated neutron multiplication factor of U(3)O/sub 2/ fuel pins in optimum geometries  

SciTech Connect (OSTI)

The KENO-Va improved Monte Carlo criticality program is used to calculate the neutron multiplication factor for TMI-U2 fuel compositions in a variety of configurations and to display parametric regions giving rise to maximum reactivity contributions. The lattice pitch of UO/sub 2/ fuel pins producing a maximum k/sub eff/ is determined as a function of boron concentrations in the coolant for infinite and finite systems. The characteristics of U/sub 3/O/sub 8/-coolant mixtures of interest to modeling the rubble region of the core are presented. Several disrupted core configurations are calculated and comparisons made. The results should be useful to proposed defueling of the TMI-U2 reactor.

Thomas, J.T.

1984-10-01T23:59:59.000Z

374

The TMI defueling project fuel debris removal system  

SciTech Connect (OSTI)

The Three Mile Island (TMI) unit 2 pressurized water reactor loss-of-coolant accident on March 28, 1979, presented the nuclear community with many challenging remediation problems. A plethora of techniques, systems, and tools have been employed for the recovery and packaging of the postaccident configuration of the reactor core. Of particular difficulty was the removal of the fuel debris located beneath the lower core support structure. Fuel debris located beneath the lower core support structure was the result of rapid cooling of the previously molten UO{sub 2} and ZrO{sub 2}, causing formation of a ceramic like rubble. Approximately 19,100 kg of this rubble settled beneath the lower core support structure and onto the lower head of the reactor containment vessel. The development and implementation of a debris collection system based on the air lift principle proved to be an effective method for gathering the fuel debris from beneath the lower core support structure.

Burge, B. (EG and G Idaho, Inc., Idaho Falls (United States))

1992-01-01T23:59:59.000Z

375

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

SciTech Connect (OSTI)

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.

David Petti

2014-06-01T23:59:59.000Z

376

The hypergeometric functions and their confluent forms  

E-Print Network [OSTI]

of an Ordinary Point Z. I'he, latur&: of the Solution in t&uo . leighborhoou of a Singularity l:uci&s' Conditions 4. The Solution for l. urge Values of S. Totally pvc&hs& nn &:, quot io?s 6. The Analytic Continuation of &F'1(a, b;c;z) 7. The Confluent..., + + ~ +, 0 I z who tficr cO?v&. r cnt or?&! t ~ I S Saiu tO asyr ptotical ly I'c&' rc&&0?t I j&c f u?ct iun f ( z) in I j!o . Cctor 8 I wrzt to:1 8 = 02 ~ f(z) J + + + ~ ' l if' f j&r every fixe&1 positive intej&ral ? li&: z (t(z) - (a...

Doyle, Jack Ellsworth

1964-01-01T23:59:59.000Z

377

The Hanford Site: An anthology of early histories  

SciTech Connect (OSTI)

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.

Gerber, M.S.

1993-10-01T23:59:59.000Z

378

Methodology for Developing the REScheckTM Software through Version 4.2  

SciTech Connect (OSTI)

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.

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

379

United abominations: Density functional studies of heavy metal chemistry  

SciTech Connect (OSTI)

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.

Schoendorff, George

2012-04-02T23:59:59.000Z

380

Reactivity Initiated Accident Test Series Test RIA 1-1 (radial average fuel enthalpy of 285 cal/g) fuel behavior report  

SciTech Connect (OSTI)

Analyses, interpretations, and discussions of results from the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-1, conducted in the Power Burst Facility reactor are presented. Four light water reactor (LWR) type test fuel rods, two previously irradiated and two previously unirradiated, were subjected to a single power transient resulting in an estimated axial peak, radial average fuel enthalpy of 285 cal/g (335 and 315 cal/g peak fuel enthalpy near the pellet surface of the previously irradiated and unirradiated test rods, respectively). The total radial average energy deposition for the test was 365 cal/g UO2. All four test rods failed as a result of the RIA power burst. Test fuel rod behavior was assessed from instrumentation response data and post-test metallurgical observations.

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

1980-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

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

SciTech Connect (OSTI)

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.

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

1980-01-01T23:59:59.000Z

382

Validation of KENO V.a with ENDF/B-V cross sections for {sup 233}U systems  

SciTech Connect (OSTI)

The multigroup Monte Carlo code KENO V.a and the 238- and 44-energy-group ENDF/B-V cross-section libraries were validated for {sup 233}U systems. Fifty-one critical experiments involving {sup 233}UO{sub 2}(NO{sub 3}){sub 2}, {sup 233}UO{sub 2}F{sub 2}, or {sup 233}U metal were selected for the validation. The H/{sup 233}U ratios for the experiments range from 0 to 1986. Each experiment was modeled with KENO V.a, and the effective multiplication factor k{sub eff} was calculated for each system using the 44- and 238-group ENDF/B-V, the 27- and 218-group ENDF/B-IV, and the 16-group Hansen-Roach cross-section libraries. The mean calculated k{sub eff} for all experiments using the 44- and 238-group libraries is 1.0090 {+-} 0.0021 and 1.0064 {+-} 0.0020, respectively. For comparison, the mean calculated k{sub eff} using the 27-, 218-, and 16-group libraries is 1.0142 {+-} 0.0038, 1.0125 {+-} 0.0038, and 0.9991 {+-} 0.0019, respectively. In general, an improvement exists in the agreement between the calculated k{sub eff}`s and the experimental results (i.e., k{sub eff} = 1.0) obtained with the newer ENDF/B-V libraries relative to ENDF/B-IV. This study is pertinent to {sup 233}U storage outside of the reactor.

Dunn, M.E.; Basoglu, B.; Bentley, C.L.; Plaster, M.J.; Wilkinson, A.D.; Dodds, H.L. [Univ. of Tennessee, Knoxville, TN (United States). Nuclear Engineering Dept.; Haught, C. [Martin Marietta Energy Systems, Piketon, OH (United States); Yamamoto, T. [Japan Atomic Energy Research Inst., Tokai (Japan)

1995-08-01T23:59:59.000Z

383

Dissolution rates of uranium compounds in simulated lung fluid  

SciTech Connect (OSTI)

Maximum dissolution rates of uranium into simulated lung fluid from a variety of materials were measured at 37/sup 0/in the where f/sub i/ is in order to estimate clearance rates from the deep lung. A batch procedure was utilized in which samples containing as little as 10 ..mu..g of natural uranium could be tested. The materials included: products of uranium mining, milling and refining operations, coal fly ash, an environmental sample from a site exposed to multiple uranium sources, and purified samples of (NH/sub 4/)/sub 2/U/sub 2/O/sub 7/ U/sub 3/O/sub 8/, UO/sub 2/, and UF/sub 4/. Dissolution of uranium from several materials indicated the presence of more than one type of uranium compound; but in all cases, the fraction F of uranium remaining undissolved at any time t could be represented by the sum of up to three terms in the series: F = ..sigma../sub i/f/sub i/ exp (-0.693t/UPSILON/sub i/), where f/sub i/ is the initial fraction of component i with dissolution half-time epsilon/sub i/. Values of epsilon/sub i/ varied from 0.01 day to several thousand days depending on the physical and chemical form of the uranium. Dissolution occurred predominantly by formation of the (UO/sub 2/(CO/sub 3/)/sub 3/)/sup 4 -/ ion; and as a result, tetravalent uranium compounds dissolved slowly. Dissolution rates of size-separated yellow-cake aerosols were found to be more closely correlated with specific surface area than with aerodynamic diameter.

Kalkwarf, D.R.

1981-01-01T23:59:59.000Z

384

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

SciTech Connect (OSTI)

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).

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

385

THERMODYNAMIC MODEL FOR URANIUM DIOXIDE BASED NUCLEAR FUEL  

SciTech Connect (OSTI)

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.

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

386

Radiochemical analyses of several spent fuel Approved Testing Materials  

SciTech Connect (OSTI)

Radiochemical characterization data are described for UO{sub 2} and UO{sub 2} plus 3 wt% Gd{sub 2}O{sub 3} commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, {sup 14}C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program.

Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

1994-09-01T23:59:59.000Z

387

Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR  

SciTech Connect (OSTI)

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

G. S. Chang; Hongbin Zhang

2009-09-01T23:59:59.000Z

388

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)

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.

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

389

Unsteady-state material balance model for a continuous rotary dissolver  

SciTech Connect (OSTI)

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.

Lewis, B.E.

1984-09-01T23:59:59.000Z

390

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

SciTech Connect (OSTI)

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%.

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

2006-07-25T23:59:59.000Z

391

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

SciTech Connect (OSTI)

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%.

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

2006-09-15T23:59:59.000Z

392

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

SciTech Connect (OSTI)

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.

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

1996-12-31T23:59:59.000Z

393

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

SciTech Connect (OSTI)

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.

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

2012-04-01T23:59:59.000Z

394

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

SciTech Connect (OSTI)

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.

Schock, Alfred

1994-06-01T23:59:59.000Z

395

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

SciTech Connect (OSTI)

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

Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Park, Su Ki [HANARO Utilization Technology Development Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Seo, Chul Gyo [HANARO Management Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2007-07-01T23:59:59.000Z

396

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

SciTech Connect (OSTI)

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.

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

397

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

SciTech Connect (OSTI)

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.

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

1998-08-01T23:59:59.000Z

398

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

SciTech Connect (OSTI)

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.

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

1980-02-01T23:59:59.000Z

399

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

SciTech Connect (OSTI)

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.

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

1998-10-06T23:59:59.000Z

400

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

SciTech Connect (OSTI)

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.

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

2009-06-22T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
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401

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

SciTech Connect (OSTI)

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.

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

2011-09-01T23:59:59.000Z

402

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

SciTech Connect (OSTI)

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.

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

2003-08-01T23:59:59.000Z

403

Updated NGNP Fuel Acquisition Strategy  

SciTech Connect (OSTI)

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.

David Petti; Tim Abram; Richard Hobbins; Jim Kendall

2010-12-01T23:59:59.000Z

404

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

SciTech Connect (OSTI)

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.

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

1980-01-01T23:59:59.000Z

405

EDF Nuclear Power Plants Operating Experience with MOX fuel  

SciTech Connect (OSTI)

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)

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

406

Clad Degradation- Summary and Abstraction for LA  

SciTech Connect (OSTI)

The purpose of this model report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The model addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository. Earlier total system performance assessments analyzed the waste form as exposed UO{sub 2}, which was available for degradation at the intrinsic dissolution rate. Water in the waste package quickly became saturated with many of the radionuclides, limiting their release rate. In the total system performance assessments for the Viability Assessment and the Site Recommendation, cladding was analyzed as part of the waste form, limiting the amount of fuel available at any time for degradation. The current model is divided into two stages. The first considers predisposal rod failures (most of which occur during reactor operation and associated activities) and postdisposal mechanical failure (from static loading of rocks) as mechanisms for perforating the cladding. Other fuel failure mechanisms including those caused by handling or transportation have been screened out (excluded) or are treated elsewhere. All stainless-steel-clad fuel, which makes up a small percentage of the overall amount of fuel to be stored, is modeled as failed upon placement in the waste packages. The second stage of the degradation model is the splitting of the cladding from the reaction of water or moist air and UO{sub 2}. The splitting has been observed to be rapid in comparison to the total system performance assessment time steps and is modeled to be instantaneous. After the cladding splits, the rind buildup inside the cladding widens the split, increasing the diffusion area from the fuel rind to the waste package interior. This model report summarizes the component models, developed for the two stages noted above, that are used as inputs to TSPA-LA. The model concludes that less than two percent of the fuel, including all of the stainless-steel clad fuel, received at the repository is failed (perforated) upon receipt at the repository. All failed fuel is assumed to axially split upon waste package failure exposing the fuel to oxidation from the in-package environment. TSPA-LA then calculates the release of radionuclides from the exposed volume of oxidized fuel.

D. Stahl

2004-10-01T23:59:59.000Z

407

Nuclear Waste Disposal and Strategies for Predicting Long-Term Performance of Material  

SciTech Connect (OSTI)

Ceramics have been an important part of the nuclear community for many years. On December 2, 1942, an historic event occurred under the West Stands of Stagg Field, at the University of Chicago. Man initiated his first self-sustaining nuclear chain reaction and controlled it. The impact of this event on civilization is considered by many as monumental and compared by some to other significant events in history, such as the invention of the steam engine and the manufacturing of the first automobile. Making this event possible and the successful operation of this first man-made nuclear reactor, was the use of forty tons of UO2. The use of natural or enriched UO2 is still used today as a nuclear fuel in many nuclear power plants operating world-wide. Other ceramic materials, such as 238Pu, are used for other important purposes, such as ceramic fuels for space exploration to provide electrical power to operate instruments on board spacecrafts. Radioisotopic Thermoelectric Generators (RTGs) are used to supply electrical power and consist of a nuclear heat source and converter to transform heat energy from radioactive decay into electrical power, thus providing reliable and relatively uniform power over the very long lifetime of a mission. These sources have been used in the Galileo spacecraft orbiting Jupiter and for scientific investigations of Saturn with the Cassini spacecraft. Still another very important series of applications using the unique properties of ceramics in the nuclear field, are as immobilization matrices for management of some of the most hazardous wastes known to man. For example, in long-term management of radioactive and hazardous wastes, glass matrices are currently in production immobilizing high-level radioactive materials, and cementious forms have also been produced to incorporate low level wastes. Also, as part of nuclear disarmament activities, assemblages of crystalline phases are being developed for immobilizing weapons grade plutonium, to not only produce environmentally friendly products, but also forms that are proliferation resistant. All of these waste forms as well as others, are designed to take advantage of the unique properties of the ceramic systems.

Wicks, G.G.

2001-03-28T23:59:59.000Z

408

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

SciTech Connect (OSTI)

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.

R. Finch

2001-06-05T23:59:59.000Z

409

Student Progress Report: Summer 2012  

SciTech Connect (OSTI)

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.

Tucker, Lucas P [Los Alamos National Laboratory

2012-08-06T23:59:59.000Z

410

TRIMOLECULAR REACTIONS OF URANIUM HEXAFLUORIDE WITH WATER  

SciTech Connect (OSTI)

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.

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

2010-02-25T23:59:59.000Z

411

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

SciTech Connect (OSTI)

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.

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

2004-12-14T23:59:59.000Z

412

Atomistic Simulations of Mass and Thermal Transport in Oxide Nuclear Fuels  

SciTech Connect (OSTI)

In this talk we discuss simulations of the mass and thermal transport in oxide nuclear fuels. Redistribution of fission gases such as Xe is closely coupled to nuclear fuel performance. Most fission gases have low solubility in the fuel matrix, specifically the insolubility is most pronounced for large fission gas atoms such as Xe, and as a result there is a significant driving force for segregation of gas atoms to grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. The first step of the fission gas redistribution is diffusion of individual gas atoms through the fuel matrix to existing sinks, which is governed by the activation energy for bulk diffusion. Fission gas bubbles are then formed by either separate nucleation events or by filling voids that were nucleated at a prior stage; in both cases their formation and latter growth is coupled to vacancy dynamics and thus linked to the production of vacancies via irradiation or thermal events. In order to better understand bulk Xe behavior (diffusion mechanisms) in UO{sub 2{+-}x} we first calculate the relevant activation energies using density functional theory (DFT) techniques. By analyzing a combination of Xe solution thermodynamics, migration barriers and the interaction of dissolved Xe atoms with U, we demonstrate that Xe diffusion predominantly occurs via a vacancy-mediated mechanism, though other alternatives may exist in high irradiation fields. Since Xe transport is closely related to diffusion of U vacancies, we have also studied the activation energy for this process. In order to explain the low value of 2.4 eV found for U migration from independent damage experiments (not thermal equilibrium) the presence of vacancy clusters must be included in the analysis. Next a continuum transport model for Xe and U is formulated based on the diffusion mechanisms established from DFT. After combining this model with descriptions of the interaction between Xe and grain boundaries derived from separate atomistic calculations, we simulate Xe redistribution for a few simple microstructures using finite element methods (FEM), as implemented in the MOOSE framework from Idaho National Laboratory. Thermal transport together with the power distribution determines the temperature distribution in the fuel rod and it is thus one of the most influential properties on nuclear fuel performance. The fuel thermal conductivity changes as function of time due to microstructure evolution (e.g. fission gas redistribution) and compositional changes. Using molecular dynamics simulations we have studied the impact of different types of grain boundaries and fission gas bubbles on UO{sub 2} thermal conductivity.

Andersson, Anders D. [Los Alamos National Laboratory; Uberuaga, Blas P. [Los Alamos National Laboratory; Du, Shiyu [Los Alamos National Laboratory; Liu, Xiang-Yang [Los Alamos National Laboratory; Nerikar, Pankaj [IBM; Stanek, Christopher R. [Los Alamos National Laboratory; Tonks, Michael [Idaho National Laboratory; Millet, Paul [Idaho National Laboratory; Biner, Bulent [Idaho National Laboratory

2012-06-04T23:59:59.000Z

413

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

SciTech Connect (OSTI)

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.

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

2010-01-29T23:59:59.000Z

414

Improved dose estimates for nuclear criticality accidents  

SciTech Connect (OSTI)

Slide rules are improved for estimating doses and dose rates resulting from nuclear criticality accidents. The original slide rules were created for highly enriched uranium solutions and metals using hand calculations along with the decades old Way-Wigner radioactive decay relationship and the inverse square law. This work uses state-of-the-art methods and better data to improve the original slide rules and also to extend the slide rule concept to three additional systems; i.e., highly enriched (93.2 wt%) uranium damp (H/{sup 235}U = 10) powder (U{sub 3}O{sub 8}) and low-enriched (5 wt%) uranium mixtures (UO{sub 2}F{sub 2}) with a H/{sup 235}U ratio of 200 and 500. Although the improved slide rules differ only slightly from the original slide rules, the improved slide rules and also the new slide rules can be used with greater confidence since they are based on more rigorous methods and better nuclear data.

Wilkinson, A.D.; Basoglu, B.; Bentley, C.L.; Dunn, M.E.; Plaster, M.J.; Dodds, H.L. [Univ. of Tennessee, Knoxville, TN (United States). Nuclear Engineering Dept.; Haught, C.F. [Martin Marietta Utility Systems, Piketon, OH (United States); Yamamoto, T. [Japan Atomic Energy Research Inst., Tokai (Japan). Tokai Research Establishment; Hopper, C.M. [Oak Ridge National Lab., TN (United States)

1995-08-01T23:59:59.000Z

415

Polyvalent fuel treatment facility (TCP): shearing and dissolution of used fuel at La Hague facility  

SciTech Connect (OSTI)

Although many used nuclear fuel types have already been recycled, recycling plants are generally optimized for Light Water Reactor (LWR) UO{sub x} fuel. Benefits of used fuel recycling are consequently restricted to those fuels, with only limited capacity for the others like LWR MOX, Fast Reactor (FR) MOX or Research and Test Reactor (RTR) fuel. In order to recycle diverse fuel types, an innovative and polyvalent shearing and dissolving cell is planned to be put in operation in about 10 years at AREVA's La Hague recycling plant. This installation, called TCP (French acronym for polyvalent fuel treatment) will benefit from AREVA's industrial feedback, while taking part in the next steps towards a fast reactor fuel cycle development using innovative treatment solutions. Feasibility studies and R/Development trials on dissolution and shearing are currently ongoing. This new installation will allow AREVA to propose new services to its customers, in particular in term of MOX fuel, Research Test Reactors fuel and Fast Reactor fuel treatment. (authors)

Brueziere, J.; Tribout-Maurizi, A.; Durand, L.; Bertrand, N. [Recycling Business Unit, AREVA, 1 place de la coupole, 92084 Paris La defense Cedex (France)

2013-07-01T23:59:59.000Z

416

Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis  

SciTech Connect (OSTI)

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.

Gilles Youinou; R. Sonat Sen

2013-09-01T23:59:59.000Z

417

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

SciTech Connect (OSTI)

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.

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

418

Uranyl ion interaction at the water/NiO(100) interface: A predictive investigation by first-principles molecular dynamic simulations  

SciTech Connect (OSTI)

The behavior of the UO{sub 2}{sup 2+} uranyl ion at the water/NiO(100) interface was investigated for the first time using Born-Oppenheimer molecular dynamic simulations with the spin polarized DFT +U extension. A water/NiO(100) interface model was first optimized on a defect-free five layers slab thickness, proposed as a reliable surface model, with an explicit treatment of the solvent. Water molecules are adsorbed with a well-defined structure in a thickness of about 4 A above the surface. The first layer, adsorbed on nickel atoms, remains mainly in molecular form but can partly dissociate at 293 K. Considering low acidic conditions, a bidentate uranyl ion complex was characterized on two surface oxygen species (arising from water molecules adsorption on nickel atoms) with d{sub U-O{sub a{sub d{sub s{sub o{sub r{sub p{sub t{sub i{sub o{sub n}}}}}}}}}}}=2.39 A. This complex is stable at 293 K due to iono-covalent bonds with an estimated charge transfer of 0.58 electron from the surface to the uranyl ion.

Sebbari, Karim [EDF-R and D, Departement Materiaux et Mecanique des Composants, Les Renardieres, Ecuelles, 77818 Moret Sur Loing (France); Institut de Physique Nucleaire d'Orsay, Universite Paris-Sud, CNRS UMR 8608, 15 rue Georges Clemenceau, Batiment 100, 91406 Orsay Cedex (France); Roques, Jerome; Simoni, Eric [Institut de Physique Nucleaire d'Orsay, Universite Paris-Sud, CNRS UMR 8608, 15 rue Georges Clemenceau, Batiment 100, 91406 Orsay Cedex (France); Domain, Christophe [EDF-R and D, Departement Materiaux et Mecanique des Composants, Les Renardieres, Ecuelles, 77818 Moret Sur Loing (France)

2012-10-28T23:59:59.000Z

419

Radiolysis Process Model  

SciTech Connect (OSTI)

Assessing the performance of spent (used) nuclear fuel in geological repository requires quantification of time-dependent phenomena that may influence its behavior on a time-scale up to millions of years. A high-level waste repository environment will be a dynamic redox system because of the time-dependent generation of radiolytic oxidants and reductants and the corrosion of Fe-bearing canister materials. One major difference between used fuel and natural analogues, including unirradiated UO2, is the intense radiolytic field. The radiation emitted by used fuel can produce radiolysis products in the presence of water vapor or a thin-film of water (including OH• and H• radicals, O2-, eaq, H2O2, H2, and O2) that may increase the waste form degradation rate and change radionuclide behavior. H2O2 is the dominant oxidant for spent nuclear fuel in an O2 depleted water environment, the most sensitive parameters have been identified with respect to predictions of a radiolysis model under typical conditions. As compared with the full model with about 100 reactions it was found that only 30-40 of the reactions are required to determine [H2O2] to one part in 10–5 and to preserve most of the predictions for major species. This allows a systematic approach for model simplification and offers guidance in designing experiments for validation.

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

2012-07-17T23:59:59.000Z

420

Modeling Precipitation and Sorption of Al, U and Co-contaminants during Titration of Acidic Sediments in Recirculation Flow-Through Experiments  

SciTech Connect (OSTI)

We conducted batch and recirculating column titration tests with contaminated acidic sediments with controlled CO2 in the headspace, and extended the geochemical model by Gu et al. (2003, GCA) to better understand and quantify the reactions governing trace metal fate in the subsurface. The sediment titration curve showed slow pH increase due to strong buffering by Al precipitation and CO2 uptake. Assuming precipitation of basaluminite at low saturation index (SI=-4), and decreasing cation exchange selectivity coefficient (kNa\\Al=0.3), the predictions are close to the observed pH and Al; and the model explains 1) the observed Ca, Mg, and Mn concentration decrease by cation exchange with sorbed Al, and 2) the decrease of U by surface complexation with Fe hydroxides at low pH, and precipitation as liebigite (Ca2UO2(CO3)3:10H2O) at pH>5.5. Without further adjustment geochemical parameters, the model describes reasonably well previous sediment and column titration tests without CO2 in the headspace, as well as the new large column test. The apparent inhibition of U and Ni decrease in the large column can be explained by formation of aqueous carbonate complexes and/or competition with carbonate for surface sites. These results indicated that ignoring labile solid phase Al would underestimate base requirement in titration of acidic aquifers.

Tang, Guoping [ORNL; Luo, Wensui [Institute of Urban Environment, Chinese Academy of Sciences; Brooks, Scott C [ORNL; Watson, David B [ORNL; Gu, Baohua [ORNL

2013-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Diffusive Release of Uranium from Contaminated Sediments into Capillary Fringe Pore Water  

SciTech Connect (OSTI)

We investigated the dynamics of U release between pore water fractions, during river stage changes from two contaminated capillary fringe sediments. Samples were from 7.0 m and 7.6 m below ground surface (bgs) in the Hanford 300 area. Sediments were packed into columns and saturated with Hanford groundwater for three to 84 days. After specified times, > 48 µm radius (calculated) sediment pores were drained, followed by draining pores to 15 µm radius. U release in the first two weeks was similar between sediments and pore sizes with a range of 4.4 to 5.6 µM U in the 14 day sample. The 7.0 m bgs sediment U declined in the larger pores to 0.22 µM at day 84, whereas the small pores released U to 6.7 µM at day 84. The 7.6 m bgs sediment released 1.4 µM on day 84, in the large pores, but continuously released U from the smaller pores (13.2 uM on day 84). The continuous release of U has resulted in a diffusion gradient from the smaller to larger pores. The observed differences in U pore-water concentrations between the two sediment samples were attributed to co-precipitation of U with carbonates. A mineral phase in the sediments was also identified as an U-carbonate species, similar to rutherfordine [UO2(CO3)].

Rod, Kenton A.; Wellman, Dawn M.; Flury, Markus; Pierce, Eric M.; Harsh, James B.

2012-09-13T23:59:59.000Z

422

Quantifying differences in the impact of variable chemistry on equilibrium uranium(VI) adsorption properties of aquifer sediments  

SciTech Connect (OSTI)

Uranium adsorption-desorption on sediment samples collected from the Hanford 300-Area, Richland, WA varied extensively over a range of field-relevant chemical conditions, complicating assessment of possible differences in equilibrium adsorption properties. Adsorption equilibrium was achieved in 500-1000 hours although dissolved uranium concentrations increased over thousands of hours owing to changes in aqueous chemical composition driven by sediment-water reactions. A non-electrostatic surface complexation reaction, >SOH + UO22+ + 2CO32- = >SOUO2(CO3HCO3)2-, provided the best fit to experimental data for each sediment sample resulting in a range of conditional equilibrium constants (logKc) from 21.49 to 21.76. Potential differences in uranium adsorption properties could be assessed in plots based on the generalized mass-action expressions yielding linear trends displaced vertically by differences in logKc values. Using this approach, logKc values for seven sediment samples were not significantly different. However, a significant difference in adsorption properties between one sediment sample and the fines (<0.063 mm) of another could be demonstrated despite the fines requiring a different reaction stoichiometry. Estimates of logKc uncertainty were improved by capturing all data points within experimental errors. The mass-action expression plots demonstrate that applying models outside the range of conditions used in model calibration greatly increases potential errors.

Stoliker, Deborah L.; Kent, Douglas B.; Zachara, John M.

2011-09-16T23:59:59.000Z

423

Impact of uranyl-calcium-carbonato complexes on uranium(VI) adsorption to synthetic and natural sediments  

SciTech Connect (OSTI)

Adsorption on soil and sediment solids may decrease aqueous uranium concentrations and limit its propensity for migration in natural and contaminated settings. Uranium adsorption will be controlled in large part by its aqueous speciation, with a particular dependence on the presence of dissolved calcium and carbonate. Here we quantify the impact of uranyl speciation on adsorption to both goethite and sediments from the Hanford Clastic Dike and Oak Ridge Melton Branch Ridgetop formations. Hanford sediments were preconditioned with sodium acetate and acetic acid to remove carbonate grains, and Ca and carbonate were reintroduced at defined levels to provide a range of aqueous uranyl species. U(VI) adsorption is directly linked to UO{sub 2}{sup 2+} speciation, with the extent of retention decreasing with formation of ternary uranyl-calcium-carbonato species. Adsorption isotherms under the conditions studied are linear, and K{sub d} values decrease from 48 to 17 L kg{sup -1} for goethite, from 64 to 29 L kg{sup -1} for Hanford sediments, and from 95 to 51 L kg{sup -1} for Melton Branch sediments as the Ca concentration increases from 0 to 1 mM at pH 7. Our observations reveal that, in carbonate-bearing waters, neutral to slightly acidic pH values ({approx}5) and limited dissolved calcium are optimal for uranium adsorption.

Stewart, B.D. [Stanford University; Mayes, Melanie [ORNL; Fendorf, Scott [ORNL

2010-01-01T23:59:59.000Z

424

Nondestructive NMR technique for moisture determination in radioactive materials.  

SciTech Connect (OSTI)

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

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

1998-12-04T23:59:59.000Z

425

Test plan for techniques to measure and remove coatings from K West Basin fuel elements  

SciTech Connect (OSTI)

Several types of coatings have previously been visually identified on the surface of 105-K East and 105-K West Basins fuel elements. One type of coating (found only in K West Basin) in particular was found to be a thick translucent material that was often seen to be dislodged from the elements as flakes when the elements were handled during visual examinations (Pitner 1997). Subsequently it was determined (for one element only in a hot cell) that this material, in the dry condition, could easily be removed from the element using a scraping tool. The coating was identified as Al(OH){sub 3} through X-ray diffraction (XRD) analyses and to be approximately 60 {micro}m thick via scanning electron microscopy (SEM). However, brushing under water in the basin using numerous mechanical strokes failed to satisfactorily remove these coatings in their thickest form as judged by appearance. Such brushing was done with only one type of metal brush, a brush design previously found satisfactory for removing UO{sub 4}.xH{sub 2}O coatings from the elements.

Bridges, A.E.

1998-06-17T23:59:59.000Z

426

Investigation of UF/sub 6/ behavior in a fire  

SciTech Connect (OSTI)

Reactions between UF/sub 6/ and combustible gases and the potential for UF/sub 6/-filled cylinders to rupture when exposed to fire are addressed. Although the absence of kinetic data prevents specific identification and quantification of the chemical species formed, potential reaction products resulting from the release of UF/sub 6/ into a fire include UF/sub 4/, UO/sub 2/F/sub 2/, HF, C, CF/sub 4/,COF/sub 2/, and short chain, fluorinated or partially fluorinated hydrocarbons. Such a release adds energy to a fire relative to normal combustion reactions. Time intervals to an assumed point of rupture for UF/sub 6/-filled cylinders exposed to fire are estimated conservatively. Several related studies are also summarized, including a test series in which small UF/sub 6/-filled cylinders were immersed in fire resulting in valve failures and explosive ruptures. It is concluded that all sizes of UF/sub 6/ cylinders currently in use may rupture within 30 minutes when totally immersed in a fire. For cylinders adjacent to fires, rupture of the larger cylinders appears much less likely.

Williams, W.R.

1988-01-01T23:59:59.000Z

427

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

SciTech Connect (OSTI)

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.

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

1992-12-01T23:59:59.000Z

428

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

SciTech Connect (OSTI)

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.

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

1992-12-01T23:59:59.000Z

429

Aquifer restoration at in-situ leach uranium mines: evidence for natural restoration processes  

SciTech Connect (OSTI)

Pacific Northwest Laboratory conducted experiments with aquifer sediments and leaching solution (lixiviant) from an in-situ leach uranium mine. The data from these laboratory experiments and information on the normal distribution of elements associated with roll-front uranium deposits provide evidence that natural processes can enhance restoration of aquifers affected by leach mining. Our experiments show that the concentration of uranium (U) in solution can decrease at least an order of magnitude (from 50 to less than 5 ppM U) due to reactions between the lixiviant and sediment, and that a uranium solid, possibly amorphous uranium dioxide, (UO/sub 2/), can limit the concentration of uranium in a solution in contact with reduced sediment. The concentrations of As, Se, and Mo in an oxidizing lixiviant should also decrease as a result of redox and precipitation reactions between the solution and sediment. The lixiviant concentrations of major anions (chloride and sulfate) other than carbonate were not affected by short-term (less than one week) contact with the aquifer sediments. This is also true of the total dissolved solids level of the solution. Consequently, we recommend that these solution parameters be used as indicators of an excursion of leaching solution from the leach field. Our experiments have shown that natural aquifer processes can affect the solution concentration of certain constituents. This effect should be considered when guidelines for aquifer restoration are established.

Deutsch, W.J.; Serne, R.J.; Bell, N.E.; Martin, W.J.

1983-04-01T23:59:59.000Z

430

Reactivity impact of {sup 16}O thermal elastic-scattering nuclear data for some numerical and critical benchmark systems  

SciTech Connect (OSTI)

The thermal neutron-elastic-scattering cross-section data for {sup 16}O used in various modern evaluated-nuclear-data libraries were reviewed and found to be generally too high compared with the best available experimental measurements. Some of the proposed revisions to the ENDF/B-VII.0 {sup 16}O data library and recent results from the TENDL system increase this discrepancy further. The reactivity impact of revising the {sup 16}O data downward to be consistent with the best measurements was tested using the JENDL-3.3 {sup 16}O cross-section values and was found to be very small in MCNP5 simulations of the UO{sub 2} and reactor-recycle MOX-fuel cases of the ANS Doppler-defect numerical benchmark. However, large reactivity differences of up to about 14 mk (1400 pcm) were observed using {sup 16}O data files from several evaluated-nuclear-data libraries in MCNP5 simulations of the Los Alamos National Laboratory HEU heavy-water solution thermal critical experiments, which were performed in the 1950's. The latter result suggests that new measurements using HEU in a heavy-water-moderated critical facility, such as the ZED-2 zero-power reactor at the Chalk River Laboratories, might help to resolve the discrepancy between the {sup 16}O thermal elastic-scattering cross-section values and thereby reduce or better define its uncertainty, although additional assessment work would be needed to confirm this. (authors)

Kozier, K. S.; Roubtsov, D. [AECL, Chalk River Laboratories, Chalk River, ON (Canada); Plompen, A. J. M.; Kopecky, S. [EC-JRC, Inst. for Reference Materials and Measurements, Retieseweg 111, 2440 Geel (Belgium)

2012-07-01T23:59:59.000Z

431

Roadmap for development of an advanced head-end reactor  

SciTech Connect (OSTI)

A novel dry treatment process for used nuclear fuel (UNF) using nitrogen dioxide is being developed to remove volatile and semi-volatile fission products and convert the monolithic fuel material to a fine powder suitable as a feed to many different separations processes. The process may be considered an advanced form of voloxidation, which was envisioned to remove tritium from the fuel prior to introduction of the fuel into the aqueous separations systems, where subsequent separation of tritium from the water would be difficult and expensive. The product from NO{sub 2} reaction can be selectively chosen to be U{sub 3}O{sub 8}, UO{sub 3}, or a nitrate by adjusting the processing conditions; all products are generated at temperatures lower than those used in standard voloxidation. All the fundamental tenants of the process have been successfully demonstrated as a proof of principle, and many aspects have been corroborated multiple times at laboratory scale. The goal of this roadmap is to define the activities required to develop the process to a technology-readiness level sufficient to an engineering-scale implementation. (authors)

Del Cul, G.D.; Johnson, J.A.; Spencer, B.B.; Collins, E.D. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831-6243 (United States)

2013-07-01T23:59:59.000Z

432

On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel  

SciTech Connect (OSTI)

Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

Perret, G.; Jordan, K. A. [Paul Scherrer Institut, Villigen, 5232 (Switzerland)

2011-07-01T23:59:59.000Z

433

Measurement of {sup 235}U content and flow of UF{sub 6} using delayed neutrons or gamma rays following induced fission  

SciTech Connect (OSTI)

Feasibility experiments conducted at Pacific Northwest National Laboratory demonstrate that either delayed neutrons or energetic gamma rays from short-lived fission products can be used to monitor the blending of UF{sub 6} gas streams. A {sup 252}Cf neutron source was used to induce {sup 235}U fission in a sample, and delayed neutrons and gamma rays were measured after the sample moved {open_quotes}down-stream.{close_quotes} The experiments used a UO{sub 2} powder that was transported down the pipe to simulate the flowing UF{sub 6} gas. Computer modeling and analytic calculation extended the test results to a flowing UF{sub 6} gas system. Neutron or gamma-ray measurements made at two downstream positions can be used to indicate both the {sup 235}U content and UF{sub 6} flow rate. Both the neutron and gamma-ray techniques have the benefits of simplicity and long-term reliability, combined with adequate sensitivity for low-intrusion monitoring of the blending process. Alternatively, measuring the neutron emission rate from (a, n) reactions in the UF{sub 6} provides an approximate measure of the {sup 235}U content without using a neutron source to induce fission.

Stromswold, D.C.; Peurrung, A.J.; Reeder, P.L.; Perkins, R.W.

1996-06-01T23:59:59.000Z

434

THE PENA BLANCA NATURAL ANALOGUE PERFORMANCE ASSESSMENT MODEL  

SciTech Connect (OSTI)

The Nopal I uranium mine in the Sierra Pena Blanca, Chihuahua, Mexico serves as a natural analogue to the Yucca Mountain repository. The Pena Blanca Natural Analogue Performance Assessment Model simulates the mobilization and transport of radionuclides that are released from the mine and transported to the saturated zone. The Pena Blanca Natural Analogue Performance Assessment Model uses probabilistic simulations of hydrogeologic processes that are analogous to the processes that occur at the Yucca Mountain site. The Nopal I uranium deposit lies in fractured, welded, and altered rhyolitic ash-flow tuffs that overlie carbonate rocks, a setting analogous to the geologic formations at the Yucca Mountain site. The Nopal I mine site has the following analogous characteristics as compared to the Yucca Mountain repository site: (1) Analogous source--UO{sub 2} uranium ore deposit = spent nuclear fuel in the repository; (2) Analogous geology--(i.e. fractured, welded, and altered rhyolitic ash-flow tuffs); (3) Analogous climate--Semiarid to arid; (4) Analogous setting--Volcanic tuffs overlie carbonate rocks; and (5) Analogous geochemistry--Oxidizing conditions Analogous hydrogeology: The ore deposit lies in the unsaturated zone above the water table.

G. Saulnier and W. Statham

2006-04-16T23:59:59.000Z

435

Speciation of plutonium and other metals under UREX process conditIONS  

SciTech Connect (OSTI)

The extractability of various Pu and Np species into tri-n-butyl phosphate (TBP) was investigated. The concentration effects of aceto-hydroxamic acid, nitric acid and nitrate on the distribution ratio of U, Pu and Np were investigated. The considerable ability of AHA to form complexes with the studied elements even under strong acidic conditions was found. While the difference in the extraction of uranyl in the presence and absence of AHA is minimal, extraction yields of Pu and Np decrease significantly. The UV-Vis-NIR and FT-IR spectroscopic investigations of uranium, plutonium, and neptunium species in the presence and absence of AHA in both aqueous and organic extraction phase were also performed. Spectroscopic analysis showed that the organic phase can contain a substantial amount of metal-hydroxamate species. A solvated ternary complex of uranium UO{sub 2}.AHA.NO{sub 3}.2TBP was observed only after prolonged contact between the aqueous and organic phases, whereas the plutonium hydroxamate species, presumably Pu(AHA){sub x}(NO{sub 3}){sub 4-x}.2TBP, appeared in the organic phase after a four minute extraction. (authors)

Paulenova, Alena; Tkac, Peter [Radiation Center, Oregon State University 100 Radiation Center, Corvallis, OR 97331-5903 (United States); Matteson, Brent S. [Department of Chemistry, Oregon State University 100 Radiation Center, Corvallis, OR 97331-5903 (United States)

2007-07-01T23:59:59.000Z

436

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

SciTech Connect (OSTI)

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

Wai, Chien M.

2003-09-10T23:59:59.000Z

437

Development of alternate extractant systems for fast reactor fuel cycle  

SciTech Connect (OSTI)

Due to the limitations of TBP in processing of high burn-up, Pu-rich fast reactor fuels, there is a need to develop alternate extractants for fast reactor fuel processing. In this context, our Centre has been examining the suitability of alternate tri-alkyl phosphates. Third phase formation in the extraction of Th(IV) by TBP, tri-n-amyl phosphate (TAP) and tri-2-methyl-butyl phosphate (T2MBP) from nitric acid media has been investigated under various conditions to derive conclusions on their application for extraction of Pu at macro levels. The chemical and radiolytic degradation of tri-n-amyl-phosphate (TAP) diluted in normal paraffin hydrocarbon (NPH) in the presence of nitric acid has been investigated by the measurement of plutonium retention in organic phase. The potential application of room temperature ionic liquids (RTILs) for reprocessing of spent nuclear fuel has been explored. Extraction of uranium (VI) and palladium (II) from nitric acid medium by commercially available RTIL and tri-n-butyl phosphate solution in RTIL have been studied and the feasibility of electrodeposition of uranium as uranium oxide (UO{sub 2}) and palladium (II) as metallic palladium from the loaded organic phase have been demonstrated. This paper describes results of the above studies and discusses the suitability of the systems for fast reactor fuel reprocessing. (authors)

Vasudeva Rao, P.R.; Suresh, A.; Venkatesan, K.A.; Srinivasan, T.G.; Raj, Baldev [Indira Gandhi Centre for Atomic Research, Kalpakkam - 603 102 (India)

2007-07-01T23:59:59.000Z

438

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

SciTech Connect (OSTI)

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

Not Available

2011-06-22T23:59:59.000Z

439

Biogeochemical Processes Controlling Microbial Reductive Precipitation of Radionuclides  

SciTech Connect (OSTI)

This project is focused on elucidating the principal biogeochemical reactions that govern the concentrations, chemical speciation, and distribution of the redox sensitive contaminants uranium (U) and technetium (Tc) between the aqueous and solid phases. The research is designed to provide new insights into the under-explored areas of competing geochemical and microbiological oxidation-reduction reactions that govern the fate and transport of redox sensitive contaminants and to generate fundamental scientific understanding of the identity and stoichiometry of competing microbial reduction and geochemical oxidation reactions. These goals and objectives are met through a series of hypothesis-driven tasks that focus on (1) the use of well-characterized microorganisms and synthetic and natural mineral oxidants, (2) advanced spectroscopic and microscopic techniques to monitor redox transformations of U and Tc, and (3) the use of flow-through experiments to more closely approximate groundwater environments. The results are providing an improved understanding and predictive capability of the mechanisms that govern the redox dynamics of radionuclides in subsurface environments. For purposes of this poster, the results are divided into three sections: (1) influence of Ca on U(VI) bioreduction; (2) localization of biogenic UO{sub 2} and TcO{sub 2}; and (3) reactivity of Mn(III/IV) oxides.

Fredrickson, James K.; Brooks, Scott C.

2004-03-17T23:59:59.000Z

440

Use of a permeable biological reaction barrier for groundwater remediation at a uranium mill tailings remedial action (UMTRA) site  

SciTech Connect (OSTI)

Previous work at the University of New Mexico and elsewhere has shown that sulfate reducing bacteria are capable of reducing uranium from the soluble +6 oxidation state to the insoluble +4 oxidation state. This chemistry forms the basis of a proposed groundwater remediation strategy in which microbial reduction would be used to immobilize soluble uranium. One such system would consist of a subsurface permeable barrier which would stimulate microbial growth resulting in the reduction of sulfate and nitrate and immobilization of metals while permitting the unhindered flow of ground water through it. This research investigated some of the engineering considerations associated with a microbial reducing barrier such as identifying an appropriate biological substrate, estimating the rate of substrate utilization, and identifying the final fate of the contaminants concentrated in the barrier matrix. The performance of batch reactors and column systems that treated simulated plume water was evaluated using cellulose, wheat straw, alfalfa hay, sawdust, and soluble starch as substrates. The concentrations of sulfate, nitrate, and U(VI) were monitored over time. Precipitates from each system were collected and the precipitated U(IV) was determined to be crystalline UO{sub 2}(s) by X-ray Diffraction. The results of this study support the proposed use of cellulosic substrates as candidate barrier materials.

Thombre, M.S.; Thomson, B.M.; Barton, L.L. [Univ. of New Mexico, Albuquerque, NM (United States)

1997-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

Uranium Extraction From Laboratory Synthesized, Uranium-Doped Hydrous Ferric Oxides  

SciTech Connect (OSTI)

The extractability of uranium (U) from synthetic hydrous ferric oxides has been shown to decrease as a function of mineral ripening, consistent with the hypothesis that the ripening process decrease contaminant lability. To evaluate this process, three hydrous ferric oxide (HFO) suspensions were co-precipitated with uranyl (UO22+) and maintained at pH 7.0 ± 0.1. Uranyl was added to the HFO post-precipitation in a fourth suspension. Two suspensions also contained either co-precipitated silicate (Si-U-HFO) or phosphate (P-U-HFO). After precipitation of the HFOs, at time intervals of one week, one month, six months, one year, and 2 years, aliquots of the suspensions were contacted with five solutions for a range of time. The extracts were analyzed for U and iron (Fe). The results are consistent with the hypothesis that U and Fe extractability will decrease as the mineral phase ripens. All extracting solutions exhibited some degree of selectivity for U, as the proportional extraction of U exceeded that for congruent dissolution. Micro X-ray diffraction analysis indicates the transformation from an amorphous phase to a material containing substantial proportions of crystalline goethite and hematite, except the P-U-HFO which remained primarily amorphous. Further analysis of the co-precipitates by the Mössbauer technique and scanning electron microscopy (SEM) provides further evidence of mineralogic ripening

Smith, Steven C.; Douglas, Matthew; Moore, Dean A.; Kukkadapu, Ravi K.; Arey, Bruce W.

2009-03-01T23:59:59.000Z

442

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

SciTech Connect (OSTI)

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

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

2013-11-06T23:59:59.000Z

443