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1

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

2

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

3

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

4

Simple but Stronger UO, Double but Weaker UNMe Bonds: The Tale Told by Cp2UO and Cp2UNR  

E-Print Network [OSTI]

Chapter 1. (23) ?H(reaction) for eq 1 = {Cp 2 U}{O} + {Ph 2 C}{NMe} ({Cp 2 U}{NMe} + {Ph 2 C}{O)) where theUNMe Bonds: The Tale Told by Cp 2 UO and Cp 2 UNR. Nomi

Barros, Noemi; LPCNO, CNRS-UPS-INSA, INSA Toulouse; Institut Charles Gerhardt, CNRS, Universite Montpellier; Laboratoire de Chimie et Physique Quantiques, CNRS, IRSAMC, Universite Paul Sabatier

2007-01-01T23:59:59.000Z

5

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

6

Sorption of 237Np by UO2 under Repository Conditions  

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

237 Np by UO 2 under Repository Conditions M. Jonathan Haire E. V. Zakharova T. V. Kazakovskaya Oak Ridge National Laboratory Institute of Physical Chemistry Institute of Experimental Physics Oak Ridge, Tennessee 37831-6166 Moscow, Russia, 117915 Sarov, Russia, 607190 Phone: (865) 574-7141 Phone: 7 095 335 1742 Phone: 7 42796 73369 e-mail: hairemj@ornl.gov e-mail: zakharova@ipc.rssi.ru e-mail: kaz@astra.vniief.ru Abstract - The primary radioisotope contributor to the calculated long-term radiation dose to the public at the Yucca Mountain spent nuclear fuel (SNF) repository site boundary is neptunium-237 ( 237 Np). Russian experiments have shown that Np(V) and Np(IV) are sorbed onto UO 2 . If Np were sorbed by UO 2 in spent fuel rather than being transported to the site

7

Thermodynamics of fission products in UO2+-x  

SciTech Connect (OSTI)

The stabilities of selected fission products - Xe, Cs, and Sr - are investigated as a function of non-stoichiometry x in UO{sub 2{+-}x}. In particular, density functional theory (OFT) is used to calculate the incorporation and solution energies of these fission products at the anion and cation vacancy sites, at the divacancy, and at the bound Schottky defect. In order to reproduce the correct insulating state of UO{sub 2}, the DFT calculations are performed using spin polarization and with the Hubbard U tenn. In general, higher charge defects are more soluble in the fuel matrix and the solubility of fission products increases as the hyperstoichiometry increases. The solubility of fission product oxides is also explored. CS{sub 2}O is observed as a second stable phase and SrO is found to be soluble in the UO{sub 2} matrix for all stoichiometries. These observations mirror experimentally observed phenomena.

Nerikar, Pankaj V [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

8

Simulations of Xe and U diffusion in UO2  

SciTech Connect (OSTI)

Diffusion of xenon (Xe) and uranium (U) in UO{sub 2} is controlled by vacancy mechanisms and under irradiation the formation of mobile vacancy clusters is important. Based on the vacancy and cluster diffusion mechanisms established from density functional theory (DFT) calculations, we derive continuum thermodynamic and diffusion models for Xe and U in UO{sub 2}. In order to capture the effects of irradiation, vacancies (Va) are explicitly coupled to the Xe and U dynamics. Segregation of defects to grain boundaries in UO{sub 2} is described by combining the bulk diffusion model with models of the interaction between Xe atoms and vacancies with grain boundaries, which were derived from atomistic calculations. The diffusion and segregation models were implemented in the MOOSE-Bison-Marmot (MBM) finite element (FEM) framework and the Xe/U redistribution was simulated for a few simple microstructures.

Andersson, Anders D. [Los Alamos National Laboratory; Vyas, Shyam [Los Alamos National Laboratory; Tonks, Michael R. [Idaho National Laboratory; Casillas, Luis [Los Alamos National Laboratory; Uberuaga, Blas P. [Los Alamos National Laboratory; Millett, Paul [Idaho National Laboratory

2012-09-10T23:59:59.000Z

9

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 ceramicmetallic (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 WUO2 and nuclear cermet production via SPS.

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

2013-02-01T23:59:59.000Z

10

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 UO2F22H2O to produce the volatile UF6 at temperatures between 100 and 500?C. Thermogravimetric reaction profiles are described that reflect changes in the uranium oxidation state and discrete chemical speciation. Differences in the onset temperatures for each system indicate that NF3-substrate interactions are important for the temperature at which NF3 reacts: U metal > UO3 > UO2 > UO2F2 > UF4 and in fact may indicate different fluorination mechanisms for these various substrates. These studies demonstrate that NF3 is a potential replacement fluorinating agent in the existing nuclear fuel cycle and in oft-proposed actinide volatility reprocessing.

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

2009-11-01T23:59:59.000Z

11

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

12

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

13

Design Study of 300 MWt PWR Fueled With UO{sub 2} Coated Fuel Particle  

SciTech Connect (OSTI)

A neutronic design was performed for 300 MWt Pressurized Water Reactor (PWR) with UO{sub 2} compacts made of coated fuel particles (CFP) comparing that with sintered pellets made of UO{sub 2} powder as ordinary fuel type. UO{sub 2} CFP type was enriched 4.8% of {sup 235}U and UO{sub 2} ordinary type was enriched 5% of {sup 235}U. Both reactors were operated with single batch refueling system with a cycle period of 3 years. The purpose of the design was to investigate the applicability of UO{sub 2} CFP type to PWR comparing with UO{sub 2} ordinary type that commonly used for PWR. The calculation was done with SRAC (Standard Reactor Analysis Code) computer code and nuclear library of JENDL-33. The results of calculation showed that k-effective for both type of fuel could be maintained at critical condition for 3 years operation without refueling. The k-effective and the Doppler coefficients have been calculated for all types of fuel at 600 K and 900 K degrees. The results of calculation showed that for all types of fuel Doppler coefficient was negative, which was good for inherent safety characteristic. The size optimization design showed that the active core dimensions of UO{sub 2} CFP type reactor was about 2 times larger than the UO{sub 2} ordinary type reactor. (authors)

Abu Khalid Rivai; Ferhat Aziz; Minoru Takahashi [Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

2006-07-01T23:59:59.000Z

14

Bestimmung der dichte, offenen porositt, porengrssenverteilung und spezifischen oberflache von UO2-Tabletten  

Science Journals Connector (OSTI)

Basically, the quantitative characterization of structural parameters of UO2 pellets is necessary for optimizing their microstructure with respect to in-service performance. Dimensional behaviour (swelling, densification, plasticity) as well as fission gas and fission product release are dependent on density, open porosity, pore size distribution and specific surface area of the pellets. Measurement techniques for one or several of these properties are reported together with typical results on UO2 pellets from the AUC conversion process.

W. Drr; H. Assmann; G. Maier; J. Steven

1979-01-01T23:59:59.000Z

15

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

16

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

17

Probable leaching mechanisms for UO/sub 2/ and spent fuel  

SciTech Connect (OSTI)

The oxidation and dissolution mechanisms for UO/sub 2/ and spent fuel will be quite similar based on this preliminary work with electrochemical leaching of UO/sub 2/ and spent fuel. In solutions containing oxygen or other oxidizing species, the UO/sub 2/ surface will be rapidly oxidized and dissolved following the transformation of uranium from U(IV) to U(VI). The hydrolysis of dissolved uranyl ions forms solid UO/sub 3/ hydrates or related complex compounds deposited onto the UO/sub 2/ surface, or other surfaces, as thin or thick coatings. Depending on the pH, temperature, and time, the various kinds of porosity and the mechanical properties of the hydrate coatings will control the dissolution rate. The effects of radiation, in terms of generation of H/sub 2/O/sub 2/, will enhance the dissolution kinetics. Electrochemical methods may be useful for determining the surface conditions, dissolution rate, and accelerated dissolution behavior for NO/sub 2/ and spent fuel. Electrochemial methods can rapidly generate much information in terms of dissolution rate and surface film properties - such as thickness, porosity, and oxidation state - in-situ during the leaching process.

Wang, R.; Katayama, Y.B.

1980-01-01T23:59:59.000Z

18

MARMOT simulations of Xe segregation to grain boundaries in UO2  

SciTech Connect (OSTI)

Diffusion of Xe and U in UO{sub 2} is controlled by vacancy mechanisms and under irradiation the formation of mobile vacancy clusters is important. We derive continuum thermodynamic and diffusion models for Xe and U in UO{sub 2} based on the vacancy and cluster diffusion mechanisms established from recent density functional theory (DFT) calculations. Segregation of defects to grain boundaries in UO{sub 2} is described by combining the diffusion model with models of the interaction between Xe atoms and vacancies with grain boundaries derived from separate atomistic calculations. The diffusion and segregation models are implemented in the MOOSE/MARMOT (MBM) finite element (FEM) framework and we simulate Xe redistribution for a few simple microstructures. In this report we focus on segregation to grain boundaries. The U or vacancy diffusion model as well as the coupled diffusion of vacancies and Xe have also been implemented, but results are not included in this report.

Andersson, Anders D. [Los Alamos National Laboratory; Tonks, Michael [Idaho National Laboratory; Casillas, Luis [Los Alamos National Laboratory; Millett, Paul [Idaho National Laboratory; Vyas, Shyam [Los Alamos National Laboratory; Uberuaga, Blas P. [Los Alamos National Laboratory; Nerikar, Pankaj [IBM

2012-06-20T23:59:59.000Z

19

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

20

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

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

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

22

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

SciTech Connect (OSTI)

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

Wittman, Richard S.; Buck, Edgar C.

2012-09-01T23:59:59.000Z

23

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

24

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), Mssbauer 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 molg-1s-1, higher than rm = 5.4 0.3 10-9 molg-1s-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

25

Fluorination behavior of UO{sub 2}F{sub 2} in the presence of F{sub 2} and O{sub 2}  

SciTech Connect (OSTI)

To apply the fluoride volatility process to the spent nuclear fuel, fluorination of UO{sub 2} by fluorine has been studied. In this reaction, it is possible that the U-O-F compounds, such as UO{sub 2}F{sub 2}, are produced. Therefore, study of such compounds is useful in order to know the fluorination behavior of UO{sub 2}. This paper presents the fluorination behavior of UO{sub 2}F{sub 2} in the presence of F{sub 2} and O{sub 2}, analyzed by thermogravimetry and differential thermal analysis (TG-DTA) method using anti-corrosion type differential thermo-balance. In fluorine gas, exothermic peaks appeared and volatilization of UF{sub 6}. In oxygen gas, only slowly pace decomposition was measured from UO{sub 22} to UF{sub 6} and UO{sub 3}. (authors)

Matsuda, Minoru; Sato, Nobuaki; Kirishima, Akira; Tochiyama, Osamu [Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, 2-1-1 Katahira Aoba-ku, Sendai, Miyagi, 980-8577 (Japan)

2007-07-01T23:59:59.000Z

26

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

27

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

28

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

29

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

30

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

31

STEPNCMillUoA: a CNC system based on STEP-NC and Function Block architecture  

Science Journals Connector (OSTI)

STEPNCMillUoA is the prototype of a new CNC system that utilises the STEP-NC data model and IEC 61499. STEP-NC provides a high-level data model and enables feature-based machining whereas the enabled layered Function Block architecture simplifies the design of the CNC controller. The architecture layers are responsible for data processing, storage and execution. The object-oriented Model-View-Control design pattern supports the system architecture and the design framework, in which simulation of the machining becomes natural and inherent part of the design process, with seamless transition from simulation to actual machining. This system possesses interoperability, portability, re-configurability and distribution characteristics. The system was tested through simulation and actual machining.

Mohamad Minhat; Xun Xu; Valeriy Vyatkin

2009-01-01T23:59:59.000Z

32

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

33

Modeling the Pyrochemical Reduction of Spent UO2 Fuel in a Pilot-Scale Reactor  

SciTech Connect (OSTI)

A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be only dependent upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

Steven D. Herrmann; Michael F. Simpson

2006-08-01T23:59:59.000Z

34

Spent-fuel special-studies progress report: probable mechanisms for oxidation and dissolution of single-crystal UO/sub 2/ surfaces  

SciTech Connect (OSTI)

Due to the complexity of the structural, microstructural and compositional characteristics of spent fuel, basic leaching and dissolution mechanisms were studied with UO/sub 2/ matrix material, specifically with single-crystal UO/sub 2/, to isolate individual contributory factors. The effects of oxidation and oxidation-dissolution were investigated in different oxidation conditions, such as in air, oxygenated solutions and deionized water containing H/sub 2/O/sub 2/. In addition, the effects of temperature on dissolution of UO/sub 2/ were studied in autoclaves at 75 and 150/sup 0/C. Also, oxidation and dissolution measurements were investigated via electrochemical methods to determine if those techniques could be applied to the characterization of leaching and dissolution of spent fuel in a hot cell. Finally, the effects of radiation were explored since the radiolysis of water may create a localized oxidizing condition at or near the spent fuel-solution interface, even in neutral or reducing conditions as commonly found in deep geological environments. The oxidation and oxidation-dissolution mechanisms for UO/sub 2/ are proposed as follows: The UO/sub 2/ surface is first oxidized in solution to form a UO/sub 2+x/ surface layer several angstroms thick. This oxidized surface has a high dissolution rate since the UO/sub 2+x/ reacts with the dissolved O/sub 2/, or H/sub 2/O/sub 2/, to form uranyl complex ions in a U(VI) state. As the uranyl ions exceed the solubility limits in solution, they become hydrolyzed to form solid deposits and suspended particles of UO/sub 3/ hydrates. The thickness and porosity of the deposited UO/sub 3/ hydrate surface-film is dependent on temperature, pH and deposition time. A long-term dissolution rate is then determined by the nature of the surface film, such as porosity, solubility and mechanical properties.

Wang, R.

1981-03-01T23:59:59.000Z

35

Fission gas and iodine release measured up to 15 GWd/t UO/sub 2/ burnup  

SciTech Connect (OSTI)

A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO/sub 2/ burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO/sub 2/ pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup.

Appelhans, A.D.

1983-01-01T23:59:59.000Z

36

Effects of rare earth oxides and UO2 + x on the structure and the mechanical properties of Zircaloy  

Science Journals Connector (OSTI)

Tension test specimens of Zircaloy have been annealed with simulated fission products, as CeO2, La2O3, Nd2O3, Y2O3 or mixtures of these rare earth oxides and UO2 + x at 350C up to 10 000 hours and at 500 or 700C up to 2000 hours. The long term effects have been studied by tension tests, scanning electron microscopy, X-ray diffraction and metallography. Annealing of Zircaloy at 700C with rare earth oxides generally leads to total embrittlement. There exists a gradation of efficacy which becomes obvious when the results of the annealing series at 500C are compared. Rare earth oxides in mixtures with UO2 + x cause improportional intense reductions of ductility. The structural characteristics of specimens lead to thermodynamic considerations of the probable reaction mechanism.

Fritz Holub

1985-01-01T23:59:59.000Z

37

Accelerated Dissolution Process of the Spent Fuel (UO{sub 2}) under Repository Conditions  

SciTech Connect (OSTI)

Nowadays, nuclear energy is one of the options for developed countries in order to maintain the demand of electric energy. One of the key problems associated with kind of energy generation is the residual waste formed after a fuel cycle (spent nuclear fuel). The thermal treatment received in the reactor and there composition renders these materials very difficult to characterize and thus exhaustive studies are required to obtain knowledge that will help to build a complete, reliable and very safety underground facility. In this way, the option known as the Deep Geological Repository (DGR) is under development by each country taking part in the nuclear energy industry. The unique pathway for the migration to the biosphere of the radionuclide, actinide and lanthanides content in the spent fuel pellet (UO{sub 2}) after the closing of the deep geological repository is by a water transport phenomena. It is a fundamental question to know how much time they will spend on their trip and the first step is the rate of liberation of these radionuclides from the spent fuel pellet. In this way the matrix dissolution rate of the spent fuel pellet, which is not dependent on the specific surface area after normalization by the initial value is a key parameter to begin the performance assessment for any deep geological repository. The specific surface value is, following the Matrix Alteration Model (MAM) sensitivity analysis, one of the most important parameters controlling the radionuclides liberation. In this way, several measurements were carried out to obtain values in different conditions for different sieves of UO{sub 2} powder treated as fresh fuel. First of all, the specific surface area was measured with a multi-point isothermal procedure with N{sub 2} and Kr for both. The values obtained are presented in order to obtain a general law for the rate of evolution with the particle size. These data are part of a bigger project about the complete description of the spent fuel analogues, which are very useful for obtaining new dissolution rates for spent nuclear fuel under repository simulated conditions. (authors)

Iglesias, Eduardo; Quinones, Javier; Rodriguez, Nieves [Energy, CIEMAT, Avda. Complutense 22, Madrid, 28040 (Spain)

2008-07-01T23:59:59.000Z

38

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

39

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

40

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

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

2014-03-10T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
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41

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

42

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

Chaitanya Deo; Davis Adnersson; Corbett Battaile; Blas uberuaga

2012-10-30T23:59:59.000Z

43

THE FATE OF THE EPSILON PHASE IN UO2 OF THE OKLO NATURAL FISSON REACTORS  

SciTech Connect (OSTI)

In spent nuclear fuel (SNF), the micron- to nano-sized epsilon phase (Mo-Ru-Pd-Tc-Rh) is an important host of {sup 99}Tc which has a long half life (2.13 x 10{sup 5} years) and can be an important contributor to dose in safety assessments of nuclear waste repositories. In order to examine the occurrence and the fate of the epsilon phase during the corrosion of SNF over long time periods, samples of uraninite from the Oklo natural reactors ({approx}2.0 Ga) have been investigated using transmission electron microscopy (TEM). Because essentially all of the {sup 99}Tc has decayed to {sup 99}Ru, this study focuses on 4d-elements of the epsilon phase. Samples were obtained from the research collection at University of Michigan representing reactor zone (RZ) 10 (836, 819,687) and from RZ 13 (864,910). Several phases with 4d-metals have been identified within UO{sub 2} matrix at the scale of 50-700 nm; fioodite, PdBi{sub 2}, with trace amounts of As, Fe, and Te, and palladodymite or rhodarsenide, (Pd,Rh){sub 2}As. The most abundant 4d-metal phase is ruthenarsinite, (Ru,Ni)As, which has a representative composition: As, 59.9; Coy 2.5; Ni, 5.2; Ru, 18.6; Rh, 8.4; Pd, 3.1; Sb, 2.4 in atomic%. Ruthenarsenite nanoparticles are typically surrounded by Pb-rich domains, galena in most cases; whereas, some particles reveal a complexly zoned composition within the grain, such as a Pb-rich domain at the core and enrichment of Ni, Co, and As at the rim. Some ruthenarsenites and Rh-Bi-particles are embedded in surrounding alteration products, e.g., chlorite, adjacent to uraninite (no further than {approx}5 {micro}m). A few of those particles are still coated by a Pb-rich layer. Based on these results, the history that epsilon phases have experienced can be described as follows: (1) The original epsilon phase was changed to, in most cases, ruthenarsenite, by As-rich fluids with other trace metals. Dissolution and a simultaneous precipitation may be responsible for the phase change. (2) All Mo and most of the Tc were released from the epsilon phase. Galena precipitated surrounding the 4d-metal phases. (3) Once the uraninite matrix has dissolved, the epsilon nanoparticles were released and ''captured'' within alteration phases that are immediately adjacent to the uraninite.

S. Utsunomiya; R.C. Ewing

2005-09-08T23:59:59.000Z

44

In-situ repairs of pipelines using metal arc welding under oil (MAW-UO) aided by eddy current crack detection  

Science Journals Connector (OSTI)

Metal arc welding under oil (MAW-UO) is a new process developed to make in-situ internal repairs of in-service oil industry pipelines tanks and vessels without the need to evacuate the service from the containing fluid. High nickel alloy welding wires were used to produce a tough relatively soft austenitic weld metal; with reduced weld metal hardness porosity residual strain and cracking susceptibility. Eddy current sensors were able to detect cracks under oil which then can be repaired in-situ using MAW-UO. The in-situ under oil crack detection and arc weld repair process will be described.

2012-01-01T23:59:59.000Z

45

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

46

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 51016 He2+/cm2 at low-temperature (< 200 C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 m thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 m) in the sample subjected to 51016 He2+/cm2, the highest fluence reached, while similar features were not detected at 91015 He2+/cm2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55 % for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.

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

47

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

48

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

49

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

50

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

51

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

52

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

Office of Legacy Management (LM)

75' 75' 00.955 L' E&nr Ph. S. W.. Wahhgt~n. D.C. 200242174, TIkpbnc (202) 48a60uo 7117-03.87.cdy.43 23 September 1987 CR CA.d M r. Andrew Wallo, III, NE-23 Division of Facility & Site Decoaunissioning Projects U.S. Department of Energy Germantown, Maryland 20545 Dear M r. Wallo: ELIMINATION RECOMMENDATION -- COLLEGES AND UNIVERSITIES M /4.0-03 kl 77.0% I - The attached elimination reconunendation was prepared in accordance rlL.0~ with your suggestion during our meeting on 22 September. The recommendation flO.o-02 includes 26 colleges and universities identified.in Enclosure 4 to Aerospace letter subject: Status of Actions - FUSRAP Site List, dated N0.03. 27 May 1987; three institutions (Tufts College, University of Virginia, rJCDCJ/ and the University of Washington) currently identified on the FUSRAP

53

Vapor-deposited /sup 235/UO/sub 2/ layers for an ultra-high-sensitivity fission counter  

SciTech Connect (OSTI)

After evaluating the properties of uranium oxide coatings prepared by electrodeposition, painting and physical vapor deposition, the vapor deposition method was selected as being preferable for preparing coatings on aluminum electrodes having a total area of 5 m/sup 2/. The electrodes were used in an experimental fission chamber designed at the Oak Ridge National Laboratory for use as a neutron flux monitor the Clinch River Breeder Reactor. Initial testing of the Ultra-High Sensitivity Fission Counter (UHSFC) indicated that a tenfold increase in sensitivity was achieved as compared to commercially available fission counters. Techniques used in vapor coating and characterizing the /sup 235/UO/sub 2/ deposits on the large-area curved substrates are described.

Adair, H.L.; Byrum, B.L.; Dailey, J.M.; Gibson, J.R.

1982-01-01T23:59:59.000Z

54

The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel  

SciTech Connect (OSTI)

The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.

David Petti; John Maki

2005-02-01T23:59:59.000Z

55

In situ Raman monitoring of He{sup 2+} irradiation induced damage in a UO{sub 2} ceramic  

SciTech Connect (OSTI)

The in situ Raman probing of a UO{sub 2} ceramic in [Ar/H{sub 2}, 95/5] gas atmosphere followed by exposure to He{sup 2+} ionic irradiation coming from a cyclotron accelerator was implemented. It was observed that the growth of Raman defect bands exhibits a unique kinetic nicely modelized by a simple direct impact model, and with an annealing rate constant of 5.6 10{sup ?4} 4 10{sup ?5} s{sup ?1} for an ionic flow of 50 nA and an ions-beam induced sample heating of 170 10 C. Also, it was observed that the Ar plasma induced by the ions-beam is a sensitive probe of the presence of the ions-beam.

Guimbretire, G.; Canizars, A.; Duval, F.; Raimboux, N.; Omne, R.; Ammar, M. R.; Simon, P. [CNRS/UPR3079 CEMHTI, 45071 Orlans Cedex 2 et Universit d'Orlans, 45067 Orlans Cedex 2 (France)] [CNRS/UPR3079 CEMHTI, 45071 Orlans Cedex 2 et Universit d'Orlans, 45067 Orlans Cedex 2 (France); Desgranges, L. [CEA/DEN/DEC Bat 352 Cadarache, 13108 Saint Paul lez Durance (France)] [CEA/DEN/DEC Bat 352 Cadarache, 13108 Saint Paul lez Durance (France); Caraballo, R.; Jgou, C. [CEA/DTCD/SECM/LMPA, Marcoule 30207 Bagnols Sur Ceze (France)] [CEA/DTCD/SECM/LMPA, Marcoule 30207 Bagnols Sur Ceze (France)

2013-07-22T23:59:59.000Z

56

The gas-phase bis-uranyl nitrate complex [(UO2)2(NO3)5]-: infrared spectrum and structure  

SciTech Connect (OSTI)

The infrared spectrum of the bis-uranyl nitrate complex [(UO2)2(NO3)5]- was measured in the gas phase using multiple photon dissociation (IRMPD). Intense absorptions corresponding to the nitrate symmetric and asymmetric vibrations, and the uranyl asymmetric vibration were observed. The nitrate v3 vibrations indicate the presence of nitrate in a bridging configuration bound to both uranyl cations, and probably two distinct pendant nitrates in the complex. The coordination environment of the nitrate ligands and the uranyl cations were compared to those in the mono-uranyl complex. Overall, the uranyl cation is more loosely coordinated in the bis-uranyl complex [(UO2)2(NO3)5]- compared to the mono-complex [UO2(NO3)3]-, as indicated by a higher O-U-O asymmetric stretching (v3) frequency. However, the pendant nitrate ligands are more strongly bound in the bis-complex than they are in the mono-uranyl complex, as indicated by the v3 frequencies of the pendant nitrate, which are split into nitrosyl and O-N-O vibrations as a result of bidentate coordination. These phenomena are consistent with lower electron density donation per uranyl by the nitrate bridging two uranyl centers compared to that of a pendant nitrate in the mono-uranyl complex. The lowest energy structure predicted by density functional theory (B3LYP functional) calculations was one in which the two uranyl molecules bridged by a single nitrate coordinated in a bis-bidentate fashion. Each uranyl molecule was coordinated by two pendant nitrate ligands. The corresponding vibrational spectrum was in excellent agreement with the IRMPD measurement, confirming the structural assignment.

Groenewold, G. S.; van Stipdonk, Michael J.; Oomens, Jos; De Jong, Wibe A.; McIIwain, Michael E.

2011-12-01T23:59:59.000Z

57

The gas-phase bis-uranyl nitrate complex [(UO2)(2)(NO3)(5)](-): infrared spectrum and structure  

SciTech Connect (OSTI)

The infrared spectrum of the bis-uranyl nitrate complex [(UO{sub 2}){sub 2}(NO{sub 3}){sub 5}]{sup -} was measured in the gas phase using multiple photon dissociation (IRMPD). Intense absorptions corresponding to the nitrate symmetric and asymmetric vibrations, and the uranyl asymmetric vibration were observed. The nitrate nu3 vibrations indicate the presence of nitrate in a bridging configuration bound to both uranyl cations, and probably two distinct pendant nitrates in the complex. The coordination environment of the nitrate ligands and the uranyl cations were compared to those in the mono-uranyl complex. Overall, the uranyl cation is more loosely coordinated in the bis-uranyl complex [(UO{sub 2}){sub 2}(NO{sub 3}){sub 5}]{sup -} compared to the mono-complex [UO{sub 2}(NO{sub 3}){sub 3}]{sup -}, as indicated by a higher O-U-O asymmetric stretching (nu3) frequency. However, the pendant nitrate ligands are more strongly bound in the bis-complex than they are in the mono-uranyl complex, as indicated by the {nu}{sub 3} frequencies of the pendant nitrate, which are split into nitrosyl and O-N-O vibrations as a result of bidentate coordination. These phenomena are consistent with lower electron density donation per uranyl by the nitrate bridging two uranyl centers compared to that of a pendant nitrate in the mono-uranyl complex. The structure was calculated using density functional theory (B3LYP functional), which produced a structure in which the two uranyl molecules bridged by a single nitrate coordinated in a bis-bidentate fashion. Each uranyl molecule was coordinated by two pendant nitrate ligands. The corresponding vibrational spectrum was in excellent agreement with the IRMPD measurement, confirming the structural assignment.

Gary S. Groenewold; Michael J. van Stipdonk; Jos Oomens; Wibe de Jong; Michael E. McIlwain

2011-12-01T23:59:59.000Z

58

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

59

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

60

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

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61

Complexation study of NpO{sub 2}{sup +} and UO{sub 2}{sup 2+} ions with several organic ligands in aqueous solutions of high ionic strength  

SciTech Connect (OSTI)

The acid dissociation constants, pK{sub a}, and the stability constants for NpO{sub 2}{sup +} and UO{sub 2}{sup 2+} have been measured for certain organic ligands [acetate, {alpha}-hydroxyisobutyrate, lactate, ascorbate, oxalate, citrate, EDTA, 8-hydroxyquinoline, 1, 10-phenanthroline, and thenoyltrifluoroacetone] in 5 m (NaCl) ionic strength solution. The pK{sub a} values were determined by potentiometry or spectrometry. These methods, as well as solvent extraction with {sup 233}U and {sup 237}Np radiotracers, were used to measure the stability constants of the 1:1 and 1:2 complexes of dioxo cations. These constants were used to estimate the concentrations required to result in 10 % competition with hydrolysis in the 5 m NaCl solution. Such estimates are of value in assessing the solubility from radioactive waste of AnO{sub 2}{sup +} and AnO{sub 2}{sup 2+} in brine solutions in contact with nuclear waste in a salt-bed repository.

Borkowski, M.; Lis, S.; Choppin, G.R. [Florida State Univ., Tallahassee, FL (United States)

1995-09-01T23:59:59.000Z

62

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

63

[Ni(H{sub 2}O){sub 4}]{sub 3}[U(OH,H{sub 2}O)(UO{sub 2}){sub 8}O{sub 12}(OH){sub 3}], crystal structure and comparison with uranium minerals with U{sub 3}O{sub 8}-type sheets  

SciTech Connect (OSTI)

The new U(VI) compound, [Ni(H{sub 2}O){sub 4}]{sub 3}[U(OH,H{sub 2}O)(UO{sub 2}){sub 8}O{sub 12}(OH){sub 3}], was synthesized by mild hydrothermal reaction of uranyl and nickel nitrates. The crystal-structure was solved in the P-1 space group, a=8.627(2), b=10.566(2), c=12.091(4) A and alpha=110.59(1), beta=102.96(2), gamma=105.50(1){sup o}, R=0.0539 and wR=0.0464 from 3441 unique observed reflections and 151 parameters. The structure of the title compound is built from sheets of uranium polyhedra closely related to that in beta-U{sub 3}O{sub 8}. Within the sheets [(UO{sub 2})(OH)O{sub 4}] pentagonal bipyramids share equatorial edges to form chains, which are cross-linked by [(UO{sub 2})O{sub 4}] and [UO{sub 4}(H{sub 2}O)(OH)] square bipyramids and through hydroxyl groups shared between [(UO{sub 2})(OH)O{sub 4}] pentagonal bipyramids. The sheets are pillared by sharing the apical oxygen atoms of the [(UO{sub 2})(OH)O{sub 4}] pentagonal bipyramids with the oxygen atoms of [NiO{sub 2}(H{sub 2}O){sub 4}] octahedral units. That builds a three-dimensional framework with water molecules pointing towards the channels. On heating [Ni(H{sub 2}O){sub 4}]{sub 3}[U(OH,H{sub 2}O)(UO{sub 2}){sub 8}O{sub 12}(OH){sub 3}] decomposes into NiU{sub 3}O{sub 10}. - Graphical abstract: The framework of [Ni(H{sub 2}O){sub 4}]{sub 3}[U(OH,H{sub 2}O)(UO{sub 2}){sub 8}O{sub 12}(OH){sub 3}] built from uranium polyhedra sheets pillared by Ni-centered octahedra.

Rivenet, Murielle, E-mail: Murielle.rivenet@ensc-lille.f [Unite de Catalyse et de Chimie du Solide, Equipe Chimie du Solide, UCCS UMR CNRS 8181, USTL, ENSC-B.P. 90108, 59652 Villeneuve d'Ascq Cedex (France); Vigier, Nicolas; Roussel, Pascal; Abraham, Francis [Unite de Catalyse et de Chimie du Solide, Equipe Chimie du Solide, UCCS UMR CNRS 8181, USTL, ENSC-B.P. 90108, 59652 Villeneuve d'Ascq Cedex (France)

2009-04-15T23:59:59.000Z

64

APS Long Range Schedule FY1999  

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

9 Beamline Operations Schedule 9 Beamline Operations Schedule JAN Monday Tuesday Wednesday Thursday Friday Saturday Sunday Week 1 18 19 20 21 22 23 24 0000 - 0800 MS MS MS MS UO UO UO 0800 - 1600 MS MS MS UO UO UO UO 1600 - 2400 MS MS MS UO UO UO UO Week 2 25 26 27 28 29 30 31 0000 - 0800 UO UO UO UO UO UO UO 0800 - 1600 UO MS UO UO UO UO UO 1600 - 2400 UO UO UO UO UO UO UO FEB 1 2 3 4 5 6 7 0000 - 0800 UO MS MS UO UO UO UO 0800 - 1600 MS MS UO UO UO UO UO 1600 - 2400 MS MS UO UO UO UO UO Week 4 8 9 10 11 12 13 14 0000 - 0800 UO UO UO UO UO UO UO 0800 - 1600 UO MS UO UO* UO UO UO 1600 - 2400 UO UO UO UO UO UO UO Week 5 15 16 17 18 19 20 21 0000 - 0800 UO MS MS UO UO U O UO

65

THE GRADUATE GUIDE #UoBgraduation  

E-Print Network [OSTI]

attend, seeing the ambition and energy of recent graduates and hearing the experiences and great stories

Birmingham, University of

66

Pipe diffusion at dislocations in UO2  

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

the fuel grain. Dislo- cations may also provide pathways for enhanced diffusion of fis- sion products, in particular, dislocations pinned to fission product precipitates may...

67

Leaching of Irradiated Candu UO2 Fuel  

Science Journals Connector (OSTI)

An assessment of the concept to dispose of spent, irradiated nuclear fuel in an underground repository requires information on the rates of radionuclide leaching from the fuel matrix and of fuel matrix dissolu...

T. T. Vandergraaf; L. H. Johnson

1980-01-01T23:59:59.000Z

68

Structure and dynamics of complexes of the uranyl ion with nonamethylimidodiphosphoramide (NIPA). 2. NMR studies of complexes (UO/sub 2/(NIPA)/sub 2/X)(CIO/sub 4/)/sub 2/ with X = H/sub 2/O, MeOH, EtOH, or Me/sub 2/CO  

SciTech Connect (OSTI)

The /sup 31/P and /sup 1/H spectra at -90/sup 0/C of the title uranyl complex ions (prepared as solutions of the solid perchlorates in inert anhydrous organic solvents (CH/sub 3/NO/sub 2/, CH/sub 2/Cl/sub 2/)) reveal a pentacoordinated arrangement of two symmetrically doubly bonded NIPA molecules and one solvent molecule about the uranyl group. In the case of (UO/sub 2/(NIPA)/sub 2/(EtOH))(ClO/sub 4/)/sub 2/, an intermolecular exchange between bound and free ethanol molecules is observed above -75/sup 0/C upon addition of ethanol to a solution of the complex. The observed rate law, k/sub inter/ = kK(EtOH)/(1 + K(EtOH) is accounted for by the existence of an outer-sphere complex (UO/sub 2//sup 2 +/(NIPA)/sub 2/(EtOH))EtOH in fast equilibrium (K) with the initial complex and free ethanol. The rate-determining step (k) consists of an outer-sphere to inner-sphere interchange of ethanol molecules. The thermodynamic and kinetic parameters are K(25/sup 0/C) = 15.8 dm/sup 3/ mol/sup -1/, k(25/sup 0/C) = 1.0 x 10/sup 4/s/sup -1/, ..delta..H and ..delta..H/sub inter//sup + +/ = -4.8 and 7.6 kcal mol/sup -1/, and ..delta..S and ..delta..S/sub inter//sup + +/ = 10.7 and -14.7 eu. A second exchange takes place at higher temperatures (above -30/sup 0/C) yielding full dynamic equivalence of the phosphorus nuclei of the coordinated NIPA molecules. The observed rate law k/sub intra/ = k/sub ex/(1 + K(EtOH)) reveals that the internal rearrangement of NIPA molecules occurs on the complex ion (UO/sub 2/(NIPA)/sub 2/(EtOH))/sup 2 +/ but not on the outer-sphere complex: k/sub ex/(25/sup 0/C) = 0.91 x 10/sup 3/s/sup -1/, ..delta..H/sub intra//sup + +/ = 10.6 kcal mol/sup -1/ and ..delta..S/sub intra//sup + +/ = -9.4 eu. Possible mechanisms for this exchange are discussed. 5 figures, 2 tables.

Rodehueser, L.; Rubini, P.R.; Bokolo, K.; Delpuech, J.J.

1982-03-01T23:59:59.000Z

69

APS Long Range Schedule FY1998  

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

8 Beamline Operations Schedule 8 Beamline Operations Schedule JAN Monday Tuesday Wednesday Thursday Friday Saturday Sunday Week 1 5 6 7 8 9 10 11 0000 - 0800 SM SM SM MS MS MS MS 0800 - 1600 SM SM* MS MS MS MS MS 1600 - 2400 SM SM MS MS MS MS MS Week 2 12 13 14 15 16 17 18 0000 - 0800 MS MS UO UO UO UO UO 0800 - 1600 MS* UO UO UO UO UO UO 1600 - 2400 MS UO UO UO UO UO UO Week 3 19 20 21 22 23 24 25 0000 - 0800 UO MS UO UO UO UO UO 0800 - 1600 MS UO UO UO UO UO UO 1600 - 2400 MS UO UO UO UO UO UO JAN/FEB 26 27 28 29 30 31 1 0000 - 0800 UO MS UO UO UO UO UO 0800 - 1600 MS SOM UO UO UO UO UO 1600 - 2400 MS UO UO UO UO UO UO Week 5 2 3 4 5 6 7 8 0000 - 0800 UO MS UO UO UO UO UO

70

APS Long Range Schedule 2000  

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

APS FY2000 Beamline Operations Schedule APS FY2000 Beamline Operations Schedule JAN Monday Tuesday Wednesday Thursday Friday Saturday Sunday Week 1 10 11 12 13 14 15 16 0000-0800 SM SM SM SM SM MS MS 0800 - 1600 SM SM SM SM MS MS MS 1600 - 2400 SM SM SM SM MS MS MS Week 2 17 18 19 20 21 22 23 0000 - 0800 MS MS MS UO UO UO UO 0800 - 1600 MS MS UO UO UO UO UO 1600 - 2400 MS MS UO UO UO UO UO Week 3 24 25 26 27 28 29 30 0000 - 0800 UO UO UO UO UO UO UO 0800 - 1600 UO MS UO UO UO UO UO 1600 - 2400 UO UO UO UO UO UO UO JAN/FEB 31 1 2 3 4 5 6 0000 - 0800 UO MS MS UO UO UO UO 0800 - 1600 MS MS UO UO UO UO UO 1600 - 2400 MS MS UO UO UO UO UO Week 5 7 8 9 10 11 12 13 0000 - 0800 UO UO UO UO UO UO UO

71

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

SciTech Connect (OSTI)

In May 1997, a historic deactivation project at the PUREX (Plutonium URanium EXtraction) facility at the Hanford Site in south-central Washington State concluded its activities (Figure ES-1). The project work was finished at $78 million under its original budget of $222.5 million, and 16 months ahead of schedule. Closely watched throughout the US Department of Energy (DOE) complex and by the US Department of Defense for the value of its lessons learned, the PUREX Deactivation Project has become the national model for the safe transition of contaminated facilities to shut down status.

Gerber, M.S.

1997-11-25T23:59:59.000Z

72

Modeling of Fission Gas Release in UO2  

SciTech Connect (OSTI)

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

MH Krohn

2006-01-23T23:59:59.000Z

73

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...

74

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

75

APS Long Range Schedule FY1997  

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

7 Beamline Operations Schedule 7 Beamline Operations Schedule January Monday Tuesday Wednesday Thursday Friday Saturday Sunday 6 7 8 9 10 11 12 00:00-08:00 MS MS UO UO UO UO UO 08:00-16:00 MS UO UO UO SV UO UO 16:00-24:00 MS UO UO UO UO UO UO 13 14 15 16 17 18 19 00:00-08:00 UO UO UO UO UO MS MS 08:00-16:00 UO SV SV Contingency UO MS MS MS 16:00-24:00 UO UO UO UO MS MS MS 20 21 22 23 24 25 26 00:00-08:00 MS UO UO UO UO UO MS 08:00-16:00 UO UO SV Contingency UO UO MS MS 16:00-24:00 UO UO UO UO UO MS MS 27 28 29 30 31 1 2 00:00-08:00 MS SM SM SM SM SM SM 08:00-16:00 MS SM SM SM SM SM SM 16:00-24:00 MS SM SM SM SM SM SM FEB Monday Tuesday Wednesday Thursday Friday Saturday Sunday

76

(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

77

f Fk66iCOP-] NBSIR 86-3422 uoL_ i 1  

E-Print Network [OSTI]

performance (up 6% in capacity, no change in efficiency) at the expense of reduced high-temperature heating. The advantages claimed for NARMs in this application are improved thermodynamic efficiency as a result of gliding refrigerant temperatures in the evaporator and condenser, and capacity modulation through composition shifting

Oak Ridge National Laboratory

78

Solubility of Pu, Np, and U from Spent UO2?Fuel Under Inert/Reducing Conditions  

Science Journals Connector (OSTI)

The overall objective of this program is to improve the scientific understanding of processes that control the release of radioactive species especially actinides from spent fuel inside a disposal canister. The Swedish concept has focused on deep burial in the rock with an iron?lined Cu?canister. Corrosion of the canister iron insert will consume any residual oxygen and provide actively reducing conditions in any fluid phase. Therefore an investigation of the solubility of different radionuclides under actively reducing conditions) (Fe2+/H2) has been performed. The solubility of U Np and Pu is measured as a function of time for three different conditions: Ar atmosphere H2 atmosphere and H2 atmosphere with Fe(II) in solution.

Yngve Albinsson; Virginia Oversby; Arvid degaard?Jensen; Lars Werme

2003-01-01T23:59:59.000Z

79

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

80

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

E-Print Network [OSTI]

, increases fission gas re- lease during irradiation and stores energy in the fuel, decreasing safety margins fission rate. Regardless of the spe- cific nuclear fuel design, the fuel is subject to temperature of Technology, 771 Ferst Dr. Atlanta, GA 30332-0245, USA b Department of Mechanical Engineering, Florida A

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.


81

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

82

Electrochemical behavior of liquid Sb anode system for electrolytic reduction of UO2  

Science Journals Connector (OSTI)

Electrolytic reduction of metal oxides is a key technique of pyroprocessing, the combination of several electrochemical processes to...16]. The spent nuclear fuels are mainly composed of metal oxides including U...

Sung-Wook Kim; Wooshin Park; Hun Suk Im

2014-09-01T23:59:59.000Z

83

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.

84

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

E-Print Network [OSTI]

for the continuous (a) and dispersed (b) types [16]. 2.3 Silicon Carbide as a High Conductivity Additive Solomon et al. explored the feasibility of increasing the thermal conductivity of oxide fuels by the addition of a second, higher thermal conductivity solid... methodology used to restrict the CO or SiO gases. All processing, therefore, must take place below this temperature. Because of ! "# Table 2.3. Samples used in the thermal conductivity measurements $%&'()&*! $(+!%,-.&/! 0...

Naramore, Michael J

2010-08-03T23:59:59.000Z

85

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

86

Fission product retention in TRISCO coated UO sub 2 particle fuels subjected to HTR simulated core heating tests  

SciTech Connect (OSTI)

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600{degree}C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800{degree}C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800{degree}C and above may exist. 6 refs., 6 figs., 4 tabs.

Baldwin, C.A.; Kania, M.J.

1990-11-01T23:59:59.000Z

87

Atom Probe Tomography | EMSL  

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

Engineered nanocrystalline... First-principles study of defects and phase transition in UO2. The electronic properties, structure and phase transformation of UO2 have...

88

Preparation and Reactions of Base-Free Bis(1,2,4-tri-tert-butylcyclopentadienyl)uranium Oxide, Cp'2UO  

E-Print Network [OSTI]

4. Barriers to Me 3 C-Group Site Exchange in Cp 2 U(X)(Y) acompound T c ? G Cp 2 UCl2 ( 1 ) Cp 2 UF 2 ( 4 ) Cp 2 U(O)(py) Cp 2 U(O)(dmap) (

Zi, Guofu; Werkema, Evan L.; Walter, Marc D.; Gottfriedsen, Jochen P.; Andersen, Richard A.

2005-01-01T23:59:59.000Z

89

The role of pe, pH, and carbonate on the solubility of UO{sub 2} and uraninite under nominally reducing conditions  

SciTech Connect (OSTI)

Experimental data obtained from uranium dioxide solubility studies as a function of pH and under nominally reducing conditions in a 0.008 mol/dm{sup 3} perchlorate medium and in a 1 mol/dm{sup 3} chloride solution are presented. The solubility of extensively characterized uraninite samples from Cigar Lake (Canada), Jachymov (Czech Republic), and Oklo (Gabon) was determined in a solution matching the composition of a groundwater associated with granitic terrain. The redox potential of the test solution was monitored throughout the experimental period. The results obtained were modeled using aqueous formation constants compiled by the NEA, using stability constants corrected to appropriate ionic strengths. A lower value of the solubility product of the uranium dioxide phase defined as fuel in the SKB uranium database provides reasonable solubilities for a wide span of experimental results at near to neutral pH. Differences in solubility between natural and synthetic samples are attributed to the presence of carbonate in the experiments performed with uraninites, while differences in solubility observed among the natural samples can be correlated to radiation effects at atomic scale.

Casas, I.; Pablo, J. de; Gimenez, J.; Torrero, M.E. [Polytechnic Univ. of Catalunya, Barcelona (Spain). Dept. of Chemical Engineering] [Polytechnic Univ. of Catalunya, Barcelona (Spain). Dept. of Chemical Engineering; Bruno, J.; Cera, E. [QuantiSci SL, Cerdanyola (Spain)] [QuantiSci SL, Cerdanyola (Spain); Finch, R.J.; Ewing, R.C. [Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Earth and Planetary Sciences] [Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Earth and Planetary Sciences

1998-07-01T23:59:59.000Z

90

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

91

Embrittlement of Zircaloy-2 on exposure to ThO2UO2 and simulated high burn-up fuel (SIMFUEL) powders at 1200C  

Science Journals Connector (OSTI)

In order to assess the changes in mechanical properties of Zircaloy cladding at extreme conditions, tensile testing and hardness measurements were carried out on Zircaloy-2 specimens that were annealed with ThO2...

R. K. Bhagat; T. R. G. Kutty; Arun Kumar; Srikumar Banerjee

2012-04-01T23:59:59.000Z

92

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

Office of Legacy Management (LM)

No. P.O. Box 158 Mt.He&lly Qq99 Q' ' - Ciacian& 31. 01 Sample Nos. 992' I HtAL I H ANU SAFt H-S 1-H J MATI LEID-WAIJY. OF OIUO station rio Type of S&h d' r dQsr CYS ..- . . -.....

93

(front end fuel cycle) 2.1 (CANDU  

E-Print Network [OSTI]

.2.1. , , , , . UNH(Uranyl Nitrate Hexahydrate) . UNH TBP(Tri-Butyl Phosphate) TBP . TBP UNH . UNH ADU(Ammonium Diuranate) AUC(Ammonium Uranyl Carbonate) UO2 . #12; UO2 3.3% U

Hong, Deog Ki

94

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

95

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

96

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]

cells. Nowadays this technology has stirred new interest in prGducing high efficiency ~lIS-SSF solar+pp+ bifacial SSF solar cells are used to experimentally analyse the interphase in a similar way a 5i layer. These conclusions are checked with ex-, perimental results of other workers. Recommendations for BSF solar cell

del Alamo, Jesús A.

97

Effects of pH, temperature, and aqueous organic material on the dissolution kinetics of meta-autunite minerals, (Na, Ca)2?1[(UO2)(PO4)]2 3H2O  

Science Journals Connector (OSTI)

...Laboratory, Argonne, Illinois. Buck, E.C., Dietz, N.L., Fortner, J.A., Bates, J.K., and Brown, N.R. (1995) Characterization...single-pass flow-through apparatus. McGrail, P.B., Ebert, W.L., Bakel, A.J., and Peeler, D.K. (1997...

Dawn M. Wellman; Jonathan P. Icenhower; Amy P. Gamerdinger; Steven W. Forrester

98

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

99

Synthesis and characterization of mono- and bis-(tetraalkylmalonamide)uranium(VI) complexes  

Science Journals Connector (OSTI)

The complex [UO2(NO3)2(TMMA)] (TMMA=N,N,N?,N?-tetramethylmalonamide) was structurally characterized by single-crystal X-ray diffraction. The complex consists of two bidentate nitrate ions and one bidentate TMMA ligand coordinated to the UO22+ ion. The complex [UO2(THMA)2]2+ (THMA=N,N,N?,N?-tetrahexylmalonamide) was prepared as the BF4? salt; this material tended to form an oil. However, [UO2(TMMA)2](OTf)2 (OTf=triflate) was isolated as a crystalline solid. Comparison of the Fourier transform infrared spectra of these complexes to the spectra of complexes formed in liquidliquid extraction systems supports the hypothesis that complexes of the type [UO2(NO3)2L] and [UO2L2](NO3)2 (L=diamide extractant) form in the extraction systems.

Gregg J Lumetta; Bruce K McNamara; Brian M Rapko; Richard L Sell; Robin D Rogers; Grant Broker; James E Hutchison

2000-01-01T23:59:59.000Z

100

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

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

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,...

102

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...

103

the toxicity of gonyaulax monilata howell to mugil cephalus  

Science Journals Connector (OSTI)

U. S. Fish and Wildlife Scrvicc, Galveston, Texas. ABSTRACT. Laboratory expcrimcnts wcrc conducted to dctcrmine the effects of in u&o cultures of the mnrinc...

104

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...

105

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

106

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...

107

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.

108

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

109

An Efficient Enthalpy-type Method for the Stefan Problem  

Science Journals Connector (OSTI)

......t) ((x,t)erT), i (E) u = uo(x) and v = vo(x) (xeQ,t = 0),J where a(u) = ctu (u>0), a(u) = c2u (u<0), if we set voed(uo) (x e Q). The function t; represents the concentration of the liquid phase, 'liquid fraction......

UWE STREIT

1989-07-01T23:59:59.000Z

110

Miscibility gap in the U-Nd-O phase diagram: a new approach of nuclear oxides in the environment?  

SciTech Connect (OSTI)

To some extent, rare-earth-doped UO{sub 2} is representative of an irradiated nuclear fuel. The two phases we observed previously in neodymium-doped UO{sub 2} are now interpreted as the existence of a miscibility gap in the U-Nd-O phase diagram using new results obtained with Raman spectroscopy. Extrapolating the miscibility gap in the U-Nd-O phase diagram to irradiated UO{sub 2} opens the path to a new understanding of nuclear oxides in the environment. (authors)

Desgranges, L.; Pontillon, Y.; Matheron, P.; Marcet, M. [CEA DEN DEC, F-13108 St Paul Les Durance, (France); Simon, P.; Guimbretiere, G. [CEMHTI, CNRS UPR 3079, F-45071 Orleans 2, (France); Porcher, F. [Ctr Etud Saclay, CEA DSM IRAMIS, F-91191 Gif Sur Yvette, (France)

2012-09-15T23:59:59.000Z

111

Structure-Property Relationships in Lithium, Silver, and Cesium Uranyl Borates  

SciTech Connect (OSTI)

Four new uranyl borates, Li[UO{sub 2}B{sub 5}O{sub 9}]H{sub 2}O (LiUBO-1), Ag[(UO{sub 2})B{sub 5}O{sub 8}(OH){sub 2}] (AgUBO-1), ?-Cs[(UO{sub 2}){sub 2}B{sub 11}O{sub 16}(OH){sub 6}] (CsUBO-1), and ?-Cs[(UO{sub 2}){sub 2}B{sub 11}O{sub 16}(OH){sub 6}] (CsUBO-2) were synthesized via the reaction of uranyl nitrate with a large excess of molten boric acid in the presence of lithium, silver, or cesium nitrate. These compounds share a common structural motif consisting of a linear uranyl, UO{sub 2}{sup 2+}, cation surrounded by BO{sub 3} triangles and BO{sub 4} tetrahedra to create an UO{sub 8} hexagonal bipyramidal environment around uranium. The borate anions bridge between uranyl units to create sheets. Additional BO{sub 3} triangles extend from the polyborate layers, and are directed approximately perpendicular to the sheets. In Li[(UO{sub 2})B{sub 5}O{sub 9}]H{sub 2}O, the additional BO{sub 3} triangles connect these sheets together to form a three-dimensional framework structure. Li[UO{sub 2})B{sub 5}O{sub 9}]H{sub 2}O and ?-Cs[(UO{sub 2}){sub 2}B{sub 11}O{sub 16}(OH){sub 6}] adopt noncentrosymmetric structures, while Ag[(UO{sub 2})B{sub 5}O{sub 8}(OH){sub 2}] and ?-Cs[(UO{sub 2}){sub 2}B{sub 11}O{sub 16}(OH){sub 6}] are centrosymmetric. Li[(UO{sub 2})B{sub 5}O{sub 9}]H{sub 2}O, which can be obtained as pure phase, displays second-harmonic generation of 532 nm light from 1064 nm light. Topological relationships of all actinyl borates are developed.

Wang, Shuao; Alekseev, Evgeny V.; Stritzinger, Jared T.; Liu, Guokui; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.

2010-01-01T23:59:59.000Z

112

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

113

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 2þ , is consid- ered a contaminant introduced into the environment near mining, processing and production

Illinois at Chicago, University of

114

Characteristics of solid hold up and circulation rate in the CFB reactor with 3-loops  

Science Journals Connector (OSTI)

The effects of the Uo..., PA/[PA+SA] ratio, total solid inventory and fluidizing velocity of loopseal on the axial solid holdup and the solid circulation rate have been determined with different particle sizes (1...

Jong-Min Lee; Jae-Sung Kim; Jong-Jin Kim

2001-11-01T23:59:59.000Z

115

Effect of catechins and tannins on hydroxyl radical formation in depleted uranium-hydrogen peroxide systems  

Science Journals Connector (OSTI)

The effects of catechins and tannins on the uranyl ion (UO2 2+)-hydrogen peroxide (H2O2) system were examined using the spin-trapping method. Epigallocatechin (EGC), having low OH-scavenging abil...

Akira Nakajima; Emiko Matsuda

2010-01-01T23:59:59.000Z

116

Effect of the militarily-relevant heavy metals, depleted uranium and heavy metal tungsten-alloy on gene expression in human liver carcinoma cells (HepG2)  

Science Journals Connector (OSTI)

Depleted uranium (DU) and heavy-metal tungsten alloys ... in military applications. Chemically similar to natural uranium, but depleted of the higher activity 235U and 234U...in vitro. Using insoluble DU-UO2 and ...

Alexandra C. Miller; Kia Brooks; Jan Smith

2004-01-01T23:59:59.000Z

117

Sizing particles of natural uranium and nuclear fuels using poly-allyl-diglycol carbonate autoradiography  

Science Journals Connector (OSTI)

......particles of natural uranium and nuclear fuels...low enriched, depleted and natural uranium and also aged...committed doses and cancer risks(4...Bristol, UK, sized uranium fragments found...nuclear fuels of depleted uranium (depUO2......

G. Hegyi; R. B. Richardson

2008-07-01T23:59:59.000Z

118

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

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

119

E-Print Network 3.0 - approach part ii Sample Search Results  

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

Summary: information available in Appendix II-A of Part II of the Chemical Hygiene Plan (CHP) at www.uos.harvard.eduehsihCHPPartII... Instructions for use of the Chemical...

120

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

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

An improved, polarographic dissolved oxygen probe.  

Science Journals Connector (OSTI)

Feb 22, 1973 ... ceed 2.2 NA for three units (<0.0220/o of full scale). Scale factor does not change by more than 3.5% over the UOC range. References.

2000-01-02T23:59:59.000Z

122

Uranium Extraction From Laboratory Synthesized, Uranium-Doped...  

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

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 ...

123

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.

124

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

E-Print Network [OSTI]

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

125

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

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

the NEAMS tool for the computer model MARMOT. Surrogate oxide (CeO2) and depleted uranium oxide (UO2) will be used for this study and will be formed into pellets of various...

126

Neutron field characterisation at mixed oxide fuel plant  

Science Journals Connector (OSTI)

......plutonium oxide (PuO2) and 70 % depleted uranium oxide (UO2) are blended together...and typical field conditions. Health Phys. (1990) 58(6):691-704...Power Plants Quality Assurance, Health Care Radiation Dosage Radiation......

C. Passmore; M. Million; M. Kirr; J. Bartz; M. S. Akselrod; A. Devita; J. Berard

2012-06-01T23:59:59.000Z

127

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

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

(2007) 10291032 www.elsevier.comlocateradmeas Summary: similar Abbreviations: DU, Depleted uranium; V79, Chinese hamster lung cells; DU-UO2NO3 depleted uranium... - tial for DU...

128

Polyethylene Encapsulation of Depleted Uranium Trioxide  

Science Journals Connector (OSTI)

Depleted uranium, in the form of uranium trioxide (UO3) powder, was encapsulated in molten polyethylene forming a stable, dense composite henceforth known as DUPoly (patent pending). Materials were fed by calibra...

J. W. Adams; P. R. Lageraaen; P. D. Kalb

2002-01-01T23:59:59.000Z

129

Stanford Synchrotron Radiation Lightsource  

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

such as solid uraninite, UO2, and its low solubility reduces the environmental risk. Naturally-occurring iron sulfide (FeS) is known to be an important electron source for the...

130

Transformations of the Micro-Domain Structure of Polyimide Films during Thermally Induced Chemical Conversion:? Characterization via Thermodynamics of Irreversible Processes  

Science Journals Connector (OSTI)

To find ?ys? and ??uo uniquely interrelated is a hallmark of the model developed here. ... Rows obsd. in the process are attributed to surface diffusion processes conducted by local stresses in oriented surface layers. ...

Hanns-Georg Kilian; Sergei Bronnikov; Tatiana Sukhanova

2003-11-14T23:59:59.000Z

131

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

132

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

133

Ris Report No. Danish Atomic Energy Commission  

E-Print Network [OSTI]

. J. A. Leth, Reactor Dept 19 Development of Nuclear Heat Calorimeters and Dose Separation of Nuclear UO.-Zr Fuel Pins. F. List, Reactor Dept. and P. Knudsen, Metallurgy Dept 45 Dispersion

134

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...

135

Summer Program for Undergraduate Research Alaska Oregon Research Training Alliance  

E-Print Network [OSTI]

Summer Program for Undergraduate Research Alaska Oregon Research Training Alliance NSF REU Site Research Training Alliance (AORTA) aorta.uoregon.edu NSF REU Site Program in Molecular Biosciences (UO

Oregon, University of

136

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

137

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

138

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

139

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 granulométrique finale. MOTS-CLES : poudre, dioxyde d'uranium, évolution morphologique, granulométrie, four tournant INTRODUCTION La poudre de dioxyde d'uranium UO2 utilisée pour la fabrication de pastilles de

Paris-Sud XI, Université de

140

Complexation of uranyl ion by tetrahexylmalonamides: an equilibrium modeling and infrared spectroscopic study  

Science Journals Connector (OSTI)

We investigated the extraction of uranyl nitrate from aqueous sodium nitrate with a series of tetrahexylmalonamides. The tetrahexylmalonamides considered were N,N,N?,N?-tetrahexylmalonamide (THMA), N,N,N?,N?-tetrahexyl-2-methylmalonamide (MeTHMA), and N,N,N?,N?-tetrahexyl-2,2-dimethylmalonamide (DiMeTHMA). This series allowed for a systematic determination of the effects of alkyl substitution of the methylene carbon. Equilibrium modeling of the extraction data indicates that at 1 M NaNO3, two extracted species are formed: UO2(NO3)2L2 and UO2(NO3)2L3. The relative abundance of these two species depends on the nature of the tetrahexylmalonamide ligand. The UO2(NO3)2L2 species is dominant in the DiMeTHMA system, with very little formation of the UO2(NO3)2L3 species. In contrast, the UO2(NO3)2L3 species is more predominant in the MeTHMA case. The case of THMA lies in between. The greater propensity of MeTHMA versus THMA to bind in a 3:1 fashion to uranyl ion might reflect the greater basicity of the carbonyl oxygens in MeTHMA. The fact that DiMeTHMA binds primarily in 2:1 fashion suggests that steric constraints are more important in that ligand. As the nitrate concentration is increased, the ligand-to-metal ratios tend to decrease, i.e. the UO2(NO3)2L2 species tends to predominate, while the UO2(NO3)2L3 species becomes less important. In the case of THMA and MeTHMA, equilibrium modeling suggests the existence of a UO2(NO3)2L species at higher nitrate concentrations. FTIR spectral studies confirm that at least two uranylTHMA complexes formed, one of which has been identified as UO2(NO3)2(THMA) by thermogravimetric analysis (TGA). The identity of the second species has not been definitively determined, but is most likely UO2(NO3)2(THMA)2.

Gregg J. Lumetta; Bruce K. McNamara; Brian M. Rapko; James E. Hutchison

1999-01-01T23:59:59.000Z

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.


141

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

142

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

143

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 Centers research focused on five major areas: (i) The fundamental aspects of anharmonicity in UO2 crystals and its impact on thermal transport; (ii) The effects of radiation microstructure on thermal transport in UO2; (iii) The mechanisms of defect clustering in UO2 under irradiation; (iv) The effect of temperature and oxygen environment on the stoichiometry of UO2; and (v) The mechanisms of growth of dislocation loops and voids under irradiation. The Center has made important progress in each of these areas, as summarized below.

D. Hurley; Todd R. Allen

2014-09-01T23:59:59.000Z

144

J.G. Tobin and S.-W. Yu Lawrence Livermore National Laboratory, Livermore, CA, USA  

National Nuclear Security Administration (NNSA)

Differentiation of 5f and 6d Components Differentiation of 5f and 6d Components in the Unoccupied Electronic Structure of UO 2 J.G. Tobin and S.-W. Yu Lawrence Livermore National Laboratory, Livermore, CA, USA Summary: 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, two break-though experiments will be discussed, which have allowed the direct determination of the U5f and U6d contributions to the unoccupied density of states (UDOS) in Uranium Dioxide (UO 2 ). [1] First, a combined soft X-ray Absorption and Bremstrahlung Isochromat Spectroscopy (XAS and BIS, respectively) study of UO 2 will be discussed. [2] Second, a novel Resonant Inverse Photoelectron and X-ray Emission Spectroscopy (RIPES and

145

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

146

Current status and future development of coated fuel particles for high temperature gas-cooled reactors  

Science Journals Connector (OSTI)

The coated particles were first invented by Roy Huddle in Harwell 1957. Through five decades of development, the German UO2 coated particle and US LEU UCO coated particle represent the highly successful coated particle designs up to now. In this paper, current status as well as the failure mechanisms of coated particle so far is reviewed and discussed. The challenges associated with high temperatures for coated particles applied in future VHTR are evaluated. And future development prospects of advanced coated particle suited for higher temperatures are presented. According to the past coated fuel particle development experience, it is unwise to make multiple simultaneous changes in the coated particle design. Two advanced designs which are modifications of standard German UO2 coated particle (UO2? herein) and US UCO coated particle (TRIZO) are promising and feasible under the world-wide cooperations and efforts.

X.W. Zhou; C.H. Tang

2011-01-01T23:59:59.000Z

147

Kinetics of laser pulse vaporization of uranium dioxide by mass spectrometry  

SciTech Connect (OSTI)

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

Tsai, C.

1981-11-01T23:59:59.000Z

148

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

149

Wasser- und wasserstoffgehalt in oxidischen LWR-kernbrennstoffen und in borcarbid  

Science Journals Connector (OSTI)

Analyzing hydrogen and water released from UO2, UO2-Gd2O3 and UO2-PuO2 pellets by different hot extraction methods led to the result that nearly all the water is adsorbed on the surface and in open pores. The released gases are approximately 80% water and up to 20% hydrogen. Only 10% of the water content of B4C is adsorbed to the surface and in open pores. 90% of the water is chemicaly bonded as boric acids adhered to the B4C. During more than five years of special measuring experience and quality control equivalent water contents in the range of 1 3 ppm were found in UO2 and mixed oxide fuel rods. These data fulfil the LWR-specification requirement of 10 ppm with high confidence. Fuel rod failures by cladding hydriding are now excluded. Zusammenfassung Die Bindung des Wassers an das UO2 ist nach den vorliegenden Untersuchungen praktisch nur eine Sorption an der zugnglichen Oberflche. Vergleichende Untersuchungen wie hier an B4C durchgefhrt und Modellbetrachtungen ber die Oberflchenbelegung (2 3 ppm H2O fllen mit einer monomolekularen Schicht die experimentell bestimmte BET-Oberflche aus) untersttzen diese Annahme. Aus den intensiven Untersuchungen an sehr vielen oxidischen Kernbrennstoffpellets ber einen mehr als fnfjhrigen Fertigungszeitraum ergibt sich, da? quivalentwassergehalte von 1 3 ppm an Brennstben mit UO2-Pellets und Mischoxidpellets sicher realisierbar sind. Diese niedrigen Gehalte erfllen die Spezifikationsforderung mit hohem Sicherheitsabstand. Seit der Beherrschung des Feuchteproblems in der Brennstoff- und Brennstabfertigung sind Sunburst -Defekte an LWR-Brennstben nicht mehr aufgetreten.

E. Brandau; L. Khler

1979-01-01T23:59:59.000Z

150

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

151

NEAMS Update  

Broader source: Energy.gov (indexed) [DOE]

April - June 2013 Published September 2013 April - June 2013 Published September 2013 Nuclear Energy ANL/NEAMS-13/3 Quarterly Highlights } } The BISON team is refining and validating the new friction model for fuel-cladding interactions (pages 2 and 3). } } Gas bubble equilibrium configurations in UO 2 were simulated, an important step toward modeling fission gas movement in oxide fuels (page 2). } } Benchmark calculations for the thermal conductivity of UO 2 have been prepared as part of the effort to predict fuel

152

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 18C. 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 50C 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 (UO32H2O), 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 UO32H2O.

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

2011-01-04T23:59:59.000Z

153

United States Government  

Broader source: Energy.gov (indexed) [DOE]

uq/Uu.3/uo U-L:i ' rAA OuL a uo oUu. 0tri.l± i m,.i,*, u". run.' r.yrcir V e.u uq/Uu.3/uo U-L:i ' rAA OuL a uo oUu. 0tri.l± i m,.i,*, u". run.' r.yrcir V e.u O000DOE F 1325.8 (08-93) Department of Energy United States Government Department of Energy Memorandum OFFICE OF INSPECTOR GENERAL DATE: March 31,2006 REPLY TO ATTN OF: IG-34 (A05TG028) Audit Report No.: OAS-L-06-10 SUBJECT: Report on Audit of "The Department's Information Technology Capital Planning and Investment Control Process" TO: Chief Information Officer, IM-1 INTRODUCTION AND OBJECTIVE Federal guidance requires that Agencies develop and implement capital planning and investment control (CPIC) processes to help ensure that their major information technology investments achieve intended outcomes, represent the best allocation of resources, and reach strategic goals and objectives. The Department of Energy

154

Business Continuity Plan for Critical Function This form should only be completed once the contents of Form 2 business impact analysis has  

E-Print Network [OSTI]

substation is on Hospital property off Myrtle Road. Three HV cables feed UoB intake sub-station, along Myrtle park. 3. Two 11 kv HV ring mains circuits University owned. 4. 23 No. Distribution Sub-Stations. 5) RESIDENCES ­ STOKE BISHOP 1. Intake Sub-Station at Badock 2. 1 HV ring main University owned 3. 4

Bristol, University of

155

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

156

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 1,377C, 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 1,300C:

Ronald baney; James Tulenko

2012-11-20T23:59:59.000Z

157

Large scale processing of seismic data in search of regional and global stress patterns  

Science Journals Connector (OSTI)

...d~qsm.c~a~d~Ioo~qa!qa~ ~aoc~avjoqc~iivaoj'ouo qouo'c~aaaaoapuvoaanosoq3o...96I'lv~amoqsua}v-ua~t)aoc~asjopncli.idmsosuodsoa isasnaogo~ssHoq...pspus',~aiSus&.iss qpam.m -aoc~ops~.qa~.qaxjorJot.c~s~uo...

A. Ben-Menahem; H. Jarosch; M. Rosenman

158

UNIVERSITY SERVICES RISK REGISTER Risk Impact Likelihood Risk  

E-Print Network [OSTI]

12 · Alignment of planning and budgeting · Regular budget review · Participation in UoG-wide planning Review Programme 6 The risk that key projects are not managed effectively and that standard business/management tool 1 The risk that US is unable to deliver its plan due to insufficient availability of resource 4 3

Glasgow, University of

159

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

160

Lim and van Oudenaarden Supplementary Methods  

E-Print Network [OSTI]

states is relatively small and the reactions are in equilibrium and therefore, dt dUN = dt dUO = dt dH=0.12 (that is, 1/35th of kM). The unmethylated loss is 2x10-3 cells/generation1 . Substitution

van Oudenaarden, Alexander

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

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

162

Molten uranium dioxide structure and dynamics  

Science Journals Connector (OSTI)

...prevented ~64 10 12 kg of CO 2 -equivalent emissions since 1971, corresponding to a saving of 1.84 million air pollutionrelated deaths (1). Because the majority of currently operating nuclear reactors use either UO 2 or mixed oxide...

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

2014-11-21T23:59:59.000Z

163

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

164

Gordon Research Conferences  

Science Journals Connector (OSTI)

...re-versal devices: D. Baker, "The ZT-40 experiment with a metal liner"; M. Ya-mada, "Plasma formation in proto S-l spheromak"; M. Schaffer, "OHTE." 24 June. Stellarators: K. Uo, "Neutral beam injection in the high-shear helio-tron...

Alexander M. Cruickshank

1981-03-13T23:59:59.000Z

165

Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels  

E-Print Network [OSTI]

?.????????????????????............................. 102 APPENDIX C?.???????????????????????????. 104 VITA????????????????????????????????.111 viii LIST OF FIGURES FIGURE Page 1 Flow Chart of the CSAS6 Control Module????????????.? 12 2 Fuel Graphite Block... for the Annular Core Configuration with UO2 and MA fuel loadings???.??. 75 x LIST OF TABLES TABLE Page I Design Specifications for the HTTR????????..???????. 17 II Fuel Graphite Block Properties??????????????...??? 22 III...

Ames, David E, II

2006-10-30T23:59:59.000Z

166

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

167

Radiation safety assessment of a system of small reactors for distributed energy  

Science Journals Connector (OSTI)

......inventory (t) 19.8 Fuel Outer diameterpitch...in UO2 No. of fuel assembles 69 Control...3014.0 Reactor vessel Inner diameterheight...of the energy consumption area, aiming...For design of a fuel exchange facility...of the energy consumption area because no...Development of in-vessel type control rod......

N. Odano; T. Ishida

2005-12-20T23:59:59.000Z

168

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

169

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

170

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

171

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

E-Print Network [OSTI]

pellets was carried out in LiF­CaF2 (+2 mass.% Li2O) at 850 °C. An inert gold anode was used instead of the usual reactive sacrificial carbon anode. In this case, oxidation of oxide ions present in the melt yields O2 gas evolution on the anode. Electrochemical charac- terisations of UO2 pellets were performed

Mailhes, Corinne

172

Lina Mareike Jansen Influence on 3D Natural Stimuli on Eye Movements  

E-Print Network [OSTI]

Lina Mareike Jansen Influence on 3D Natural Stimuli on Eye Movements and the Selection of Fixation Lina Mareike Jansen lijansen@uos.de April 3, 2008 Supervisors: Peter König Neurobiopsychology of Cognitive Science University of Osnabrück Germany #12;Master's Thesis Lina Jansen 2 Influence of 3D Natural

Kallenrode, May-Britt

173

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

174

A TIME-OF-FLIGHT SPECTROMETER FOR ELASTIC NEUTRON SCATTERING ON POWDERED SAMPLES  

E-Print Network [OSTI]

on test samples with simple structure (UO2 give a reliability factor R of 1.3 % which com- pares fairly of the Melusine 8 MW light-water reactor of the Centre d'Etudes Nucléaires de Grenoble. The thermal flux pattern, a matrix formalism has been deve- loped [6,] [7], [8], allowing the analysis of a real

Paris-Sud XI, Université de

175

Caulobacter crescentus as a Whole-Cell Uranium Biosensor  

Science Journals Connector (OSTI)

...results, we constructed a uranium reporter that places...strongly upregulated under uranium stress conditions. MATERIALS...Pb(NO3)2], and depleted uranyl nitrate [UO2...and by Damon Runyon Cancer Research Foundation fellowship...specificity for chelated uranium(VI): isolation and...

Nathan J. Hillson; Ping Hu; Gary L. Andersen; Lucy Shapiro

2007-09-28T23:59:59.000Z

176

Disposal Options for Depleted Uranium Trioxide (DU03) Study  

SciTech Connect (OSTI)

There exists at SRS 50 million pounds of depleted UO3 (DUO) stored in 55-gallon drums stacked three high in several buildings. This storage configuration does not allow access to the individual drums for monitoring drum integrity and material accountability.

Jones, T.M.

2002-08-02T23:59:59.000Z

177

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

178

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

179

Proceedings of ICRC 2001: 3273 c Copernicus Gesellschaft 2001 The influence of magnetic clouds on the propagation of energetic  

E-Print Network [OSTI]

with the solar wind and adiabatic deceleration (Ruf- folo, 1995). Focusing always is considered for simple ge.-B. Kallenrode (mkallenr@uos.de) 2 The Model Since we are concerned with particles with energies in the MeV and tens of MeV range, solar wind effects such as con- vection and adiabatic deceleration are of minor

Steinhoff, Heinz-Jürgen

180

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

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

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

182

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

183

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

184

Interaction of uranium dioxide with molten zircaloy  

SciTech Connect (OSTI)

Laboratory experiments in which gram quantities of molten Zircaloy were held in contact with UO/sub 2/ for known times (20-600 s) and temperatures (1900-2200/sup 0/C) were conducted. Following each experiment, polished sections of the specimen were examined by optical microscopy, electron microprobe, scanning Auger microscopy, and x-ray fluorescence spectroscopy. Three closely-related experiments were conducted. In the first, the molten metal was contained in a UO/sub 2/ crucible. The dissolution rate in this system was found to be dominated by natural convection in the melt driven by density gradients established by the dissolving uranium. The mechanism of the interaction also was observed to involve penetration and detachment of the grains of the oxide by the molten metal. Similar tests with single-crystal UO/sub 2/ specimens showed similar dissolution behavior. Less severe attack occurred because of the absence of grain boundaries, although subgrain boundaries or dislocations provided high-diffusivity pathways for preferential oxygen removal. In the third type of test, a disk of UO/sub 2/ was placed at the bottom of a ThO/sub 2/ crucible. This arrangement prevented establishment of unstable density gradients in the liquid phase, resulting in a purely diffusion-controlled interaction.

Kim, K.T.

1987-01-01T23:59:59.000Z

185

Estimation of surface precipitation constants for sorption of divalent metals onto hydrous ferric oxide and calcite  

E-Print Network [OSTI]

+ , Mg2+ , Ca2+ , Mn2+ , Co2+ , Ni2+ , Sr2+ , Sn2+ , Ba2+ , Eu2+ , Ra2+ , Pb2+ , Hg2+ , Cu2+ , and UO2 2 estimation of free energies and, hence, equilibrium constants of the surface precipitation reactions for Be2

Polly, David

186

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

187

U6+ MINERALS AND INORGANIC COMPOUNDS: INSIGHTS INTO AN EXPANDED STRUCTURAL HIERARCHY OF CRYSTAL STRUCTURES  

Science Journals Connector (OSTI)

...Crystal chemistry of uranyl molybdates. XI. Crystal structures of Cs2[(UO2...L.E., Bonthrone, K.M., Yong, Ping Goddard, D.T. (2000): Enzymically...33 , 1091-1101. Wang, Xiqu, Huang, Jin, Liu, Lumei Jacobson, A.J. (2002...

Peter C. Burns

188

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

189

AUSTRALIAN NATIONAL UNIVERSITY OEPARTf'.iENT OF NUCLEAR PHYSICS  

E-Print Network [OSTI]

AUSTRALIAN NATIONAL UNIVERSITY OEPARTf'.iENT OF NUCLEAR PHYSICS 14UO TANK OPENING REPORT NO. 25 found when we had to go into the tank. We feel that anyone closely associated with accelerators mlmbers wherever possible. We often leave the shafts running for vacuum reasons when the machine

Chen, Ying

190

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

191

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...

192

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

193

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

194

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

195

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

196

U.S. Department of Energy Categorical Exclusion Determination Form  

Broader source: Energy.gov (indexed) [DOE]

Heating Actinide Materials in a 2.9 vol % Hydrogen Atmosphere Using a Laboratory Furnace in the C-155 Glovebox Heating Actinide Materials in a 2.9 vol % Hydrogen Atmosphere Using a Laboratory Furnace in the C-155 Glovebox Savannah River Site Aiken/Aiken/South Carolina The preparation of uranium (IV) oxide (UO2) and mixed actinide oxides containing UO2 can be accomplished by heating actinide compounds in the presence of H2. Hydrogen is required as a reducing agent to prevent the oxidation of U(IV) to a higher oxidation state. The capability to heat actinide materials in a reducing environment using 2.9 vol % H2 in Ar was installed in glovebox 46 in lab C-155. Multiple R&D programs will utilize the furnace capability to prepare actinide oxides. B3.6 - Small-scale research and development, laboratory operations, and pilot projects Andrew R. Grainger

197

MARMOT Enhanced | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

MARMOT Enhanced MARMOT Enhanced MARMOT Enhanced January 29, 2013 - 10:23am Addthis Lower-length-scale Model Development 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. Atomistic simulations were used to determine thermal resistance values for four different types of grain boundaries, and these values have been used in meso-scale simulations of heat transport through representative fuel microstructures. [LANL] Density functional theory techniques, previously applied to diffusion of Xe in UO2, have now been extended to Kr. Thus, both major gaseous fission products are now included in the simulations, which have identified the transport mechanism as being vacancy mediated. Activation energies have

198

u.s. DEPARTMENT OF ENERGY EERE PROJECT MANAG EMENT CENTER NEPA DETERl.VIINATION  

Broader source: Energy.gov (indexed) [DOE]

U.O!) U.O!) u.s. DEPARTMENT OF ENERGY EERE PROJECT MANAG EMENT CENTER NEPA DETERl.VIINATION RECIPIENT:Commonwealth of the Northem Mari ana Islands Energy Division PROJECT TITLE : State Energy Program Formula Grant Page 1 of2 STATE: M P Funding Opportunity Announcement Number DE-FOA-OOOO507 Procurement Instrument Number DE-EE0004510 NEPA Control Number em Number GF0-0004510-OO1 GO Based on my review or the information concerning the proposed action, as NEPA Compliance Officer (authorized under DOE Order 451. IA), I have made the following determination: ex, EA, EIS APPENDIX AND NUMBER: Description : A91nfonnation gatheri ng, analysis, and disseminatlon A11 Technical advice and assistance t o organizations Informatton gathering (induding, but not limited to, literature surveys, inventories, site ViSits, and audits), data

199

An Insulating Breakthrough | Advanced Photon Source  

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

Science Highlights Archives: 2013 | 2012 | 2011 | 2010 Science Highlights Archives: 2013 | 2012 | 2011 | 2010 2009 | 2008 | 2007 | 2006 2005 | 2004 | 2003 | 2002 2001 | 2000 | 1998 | Subscribe to APS Science Highlights rss feed An Insulating Breakthrough JANUARY 8, 2007 Bookmark and Share Tungsten Diselenide A new insulating material with the lowest thermal conductivity ever measured for a fully dense solid has been created at the University of Oregon (UO) and tested at the XOR/UNI 33-BM beamline at the U.S. Department of Energy's Advanced Photon Source (APS) at Argonne. The research was carried out by collaborators from the UO, the University of Illinois at Urbana-Champaign, the Rensselaer Polytechnic Institute, and Argonne. While far from having immediate application, the principles involved, once understood, could lead to improved insulation for a wide variety of uses,

200

CX-008356: Categorical Exclusion Determination | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

6: Categorical Exclusion Determination 6: Categorical Exclusion Determination CX-008356: Categorical Exclusion Determination Heating Actinide Materials in a 2.9 volume % Hydrogen Atmosphere Using a Laboratory Furnace in the C-155 Glovebox CX(s) Applied: B3.6 Date: 04/18/2012 Location(s): South Carolina Offices(s): Savannah River Operations Office The preparation of uranium (IV) oxide (UO2) and mixed actinide oxides containing UO2 can be accomplished by heating actinide compounds in the presence of hydrogen. Hydrogen is required as a reducing agent to prevent the oxidation of U(IV) to a higher oxidation state. The capability to heat actinide materials in a reducing environment using 2.9 volume % hydrogen in Ar was installed in glovebox 46 in lab C-155. Multiple research and demonstration programs will utilize the furnace capability to prepare

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201

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

202

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

203

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

204

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

205

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

206

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

207

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. (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 3.471025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.531025 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 987C in Capsule 6 to 1296C in Capsule 2 for UCO, and from 996 to 1062C 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 210-6. 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

2014-08-01T23:59:59.000Z

208

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 3.471025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.531025 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 987C in Capsule 6 to 1296C in Capsule 2 for UCO, and from 996 to 1062C 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 210-6. 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

2014-07-01T23:59:59.000Z

209

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 3.471025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.531025 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 987C in Capsule 6 to 1296C in Capsule 2 for UCO, and from 996 to 1062C 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 210-6. 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

210

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

211

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

212

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

213

Uranium mononitride as a potential commercial LWR fuel  

SciTech Connect (OSTI)

This paper evaluated uranium mononitride (UN) as a potential replacement for 5% enriched UO{sub 2} fuel in Generation III and III+ commercial light water reactors (LWRs). Significant improvement in LWR performance depends on developing and implementing changes in the nuclear fuel used in these reactors. Compared to UO{sub 2}, UN offers several advantages such as higher uranium loading and better thermal conductivity. In this paper, the thermal safety margin of UN was evaluated at both normal and accident conditions using a readily available coupled CFD model developed for the US DOE CASL program. One of the prime technical challenges in utilization of UN as LWR fuel is the water compatibility because pure phase UN is not stable in water at 350 deg. C. The water corrosion resistance of UN and the corrosion mechanism were reviewed and mitigation methods were proposed. (authors)

Xu, P.; Yan, J.; Lahoda, E. J.; Ray, S. [Westinghouse Electric Company, LLC, 5801 Bluff Rd, Columbia, SC 29209 (United States)

2012-07-01T23:59:59.000Z

214

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

215

Nuclear carrier business volume projections, 1980-2000  

SciTech Connect (OSTI)

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

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

1980-05-01T23:59:59.000Z

216

SYNTHESIS, CHARACTERIZATION, AND STRUCTURE OF A URANYL COMPLEX WITH A DISULFIDE LIGAND, BIS(DI-n-PROPYLAMMONIUM) DISULFIDOBIS (DI-n-PROPYLMONOTHIOCARBAMATO) DIOXOURANATE(VI)  

SciTech Connect (OSTI)

Olive-green crystals of the title compound, [({underline n}-C{sub 3}H{sub 7}){sub 2}NH{sub 2}{sup +}]{sub 2} [UO(({underline n}-C{sub 3}H{sub 7}){sub 2}NCOS){sub 2}(S{sub 2}){sup -2}, are orthorhombic, space group Pcan, with {underline a}= 15.326(6) {Angstrom}, {underline b} = 17.474(6) {Angstrom}, {underline C} = 14.728(6) {Angstrom}, and Z = 4, (d{sub X} = 1.45 g/cm{sup 3}). For 1833 data, I >{sigma}, R = 0.052, and R{sub w} = 0.069. The structure was revealed by single-crystal x-ray diffraction studies to consist of [(n-C{sub 3}H{sub 7}){sub 2}NH{sub 2}]+ cations and [UO{sub 2}(({underline n|-C{sub 3}H{sub 7}){sub 2}NCOS){sub 2}(S{sub 2}){sup -2} anions with the uranium atom at the center of an irregular hexagonal bipyramid. The uranyl oxygen atoms occupy the axial positions. The equatorial coordination plane contains the disulfide (S{sub 2}{sup -2}) group bonded in a "side-on" fashion, and two oxygen and two sulfur donor atoms from the monothiocarbamate ligands. Interatomic distances are S-S = 2.05(1) {Angstrom}, U-S= 2.714(3) {Angstrom} (disulfide); U-S= 2.871(3) {Angstrom} and U-O = 2.46(1) {Angstrom} (thiocarbamate); U-O = 1.81(1) {Angstrom} (uranyl), The nitrogen atom in the dipropylammonium cation is hydrogen bonded to the uranyl oxgyen atoms,

Perry, Dale L.; Zalkin, Allan; Ruben, Helena; Templeton, David H.

1981-05-01T23:59:59.000Z

217

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

218

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

219

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

220

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

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

OPTICAL DIFFERENCE FREQUENCY GENERATION OF FAR INFRARED RADIATION  

E-Print Network [OSTI]

$N=-l $~=AA $8=b8 $BMA=B-A $BPA=B+A VJ=0. ,*aPA $UO=0.5*TEM*FUO=FJl $FhO=F~l $BMA=B-A $BPA=B+A $GU TO 70 ~=N-l $lFI~.SB12'~1 $A=SA12(NI BMA=B-A $BPA=B+A $\\IO=SV1ZINI $UC=SUIZtNI

Morris, J.R.

2010-01-01T23:59:59.000Z

222

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

223

Raman spectroscopy study of selected uranophanes  

Science Journals Connector (OSTI)

Raman spectra at 298 and 77K of three uranophane samples from different localities are described and interpreted. The spectra are sample dependent. UO bond lengths in uranyls are calculated from the spectra and compared with the published data of single crystal structure and EXAFS spectroscopy. Hydrogen-bonding of water molecules and silanols is discussed and the proton mobility in uranophane sheet crystal structure is assumed.

Ray L. Frost; Ji? ?ejka; Matt L. Weier; Wayde N. Martens

2006-01-01T23:59:59.000Z

224

Onions and Bunch Crops at Beeville.  

E-Print Network [OSTI]

to market. IRRIGATION AND YIELD TEST. The purpose of this experiment was to determine the re la ti^ cost and yields of irrigated and unirrigated onion crops, with espc cia1 reference to the quantity, cost and value of water require( The plats employed... ....................................... Plant Seed Co ......... ........ Turnip Non Plus Ultra ............... Plant Seed Uo Early Short Top Long Scarlet ... Plant Seed Do ......... Long Black Spanish .................... Plant Seed Cc ......... Turnip Triumph...

Robertson, J. K.; Green, Edward C.

1904-01-01T23:59:59.000Z

225

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology  

SciTech Connect (OSTI)

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO{sub 2}) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO{sub 2} is separately collected for {approx}60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which has the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO{sub 2} spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to {approx}10 wt% of the total spent fuel owing to the prior UO{sub 2} recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.

Ohta, Hirokazu; Inoue, Tadashi; Sakamura, Yoshiharu; Kinoshita, Kensuke

2005-05-15T23:59:59.000Z

226

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

227

DOE/EIA-0321/HRIf Residential Energy Consumption Survey. Consumption  

Gasoline and Diesel Fuel Update (EIA)

/HRIf /HRIf Residential Energy Consumption Survey. Consumption and Expenditures, April 1981 Through March 1982 an Part I: National Data Energy Information Administration Washington, D.C. (202) 20fr02 'O'Q 'uoifkjjUSBM ujiuud juaoiujeAog 'S'n siuawnooQ jo luapuaiuuadns - 0088-292 (202) 98S02 '0'Q 8f 0-d I 6ujp|ing uoiieflSjUjiup v UOIIBUJJOJU | ABjau 3 02-13 'jaiuao UOIJBUJJOJUI XBjaug IBUO!;BN noA pasopua s; uujoi japjo uy 'MO|aq jeadde sjaqoinu auoydajaj PUB sassajppv 'OI3N 9>4i oi papajip aq pinoqs X6jaue uo suotjsenQ '(OIBN) J9»ueo aqjeiMJO^ui ASjaug (BUOIJEN s,vi3 QMi JO OdO 941 UUGJJ peuiBiqo eq ABOI suoijBonqnd (vi3) UO!JBJ;S!UILUPV UOIIBUUJO|U| XBjeug jaiflo PUB SJMJ p ssBiiojnd PUB UOIIBLUJO^JI 6uuepjQ (Od9) 90IWO Bujjuud luetuujaAOQ -g'n 'sjuaiunooa p juapuaiuuedng aqt LUOJI aiqB||BAB si uoHBOjiqnd sjt|i

228

FAQ 3-What are the common forms of uranium?  

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

are the common forms of uranium? are the common forms of uranium? What are the common forms of uranium? Uranium can take many chemical forms. In nature, uranium is generally found as an oxide, such as in the olive-green-colored mineral pitchblende. Uranium oxide is also the chemical form most often used for nuclear fuel. Uranium-fluorine compounds are also common in uranium processing, with uranium hexafluoride (UF6) and uranium tetrafluoride (UF4) being the two most common. In its pure form, uranium is a silver-colored metal. The most common forms of uranium oxide are U3O8 and UO2. Both oxide forms have low solubility in water and are relatively stable over a wide range of environmental conditions. Triuranium octaoxide (U3O8) is the most stable form of uranium and is the form most commonly found in nature. Uranium dioxide (UO2) is the form in which uranium is most commonly used as a nuclear reactor fuel. At ambient temperatures, UO2 will gradually convert to U3O8. Because of their stability, uranium oxides are generally considered the preferred chemical form for storage or disposal.

229

Polyethylene Encapsulated Depleted Uranium  

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

Poly DU Poly DU Polyethylene Encapsulated Depleted Uranium Technology Description: Brookhaven National Laboratory (BNL) has completed preliminary work to investigate the feasibility of encapsulating DU in low density polyethylene to form a stable, dense product. DU loadings as high as 90 wt% were achieved. A maximum product density of 4.2 g/cm3 was achieved using UO3, but increased product density using UO2 is estimated at 6.1 g/cm3. Additional product density improvements up to about 7.2 g/cm3 were projected using DU aggregate in a hybrid technique known as micro/macroencapsulation.[1] A U.S. patent for this process has been received.[2] Figure 1 Figure 1: DU Encapsulated in polyethylene samples produced at BNL containing 80 wt % depleted UO3 A recent DU market study by Kapline Enterprises, Inc. for DOE thoroughly identified and rated potential applications and markets for DU metal and oxide materials.[3] Because of its workability and high DU loading capability, the polyethylene encapsulated DU could readily be fabricated as counterweights/ballast (for use in airplanes, helicopters, ships and missiles), flywheels, armor, and projectiles. Also, polyethylene encapsulated DU is an effective shielding material for both gamma and neutron radiation, with potential application for shielding high activity waste (e.g., ion exchange resins, glass gems), spent fuel dry storage casks, and high energy experimental facilities (e.g., accelerator targets) to reduce radiation exposures to workers and the public.

230

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

231

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

232

Microbial dissolution and reduction of uranyl crystals by Shewanella oneidensis MR-1  

Science Journals Connector (OSTI)

Abstract Dissimilatory metal-reducing bacteria (DMRB) can harvest energy for growth and activities by respiring metals, but it is so far unknown whether DMRB can acquire crystalline-phase actinides. In the present study, we used Shewanella oneidensis MR-1 to investigate microbially-mediated dissolution and reduction of U(VI) in three uranyl(VI) borate and boronate crystals (i.e., UO2(CH3BO2)(H2O) (UCBO); UO2B2O4 (UBO); and Na[(UO2)B6O10(OH)]2H2O (NaUBO)). Comparison of the dissolved U(VI) concentrations between samples with and without bacteria indicates that MR-1 facilitated dissolution of UCBO and UBO. Based on the assumption that only dissolved U(VI) was reduced, U(VI) reduction was substantially underestimated for UCBO and NaUBO, indicating that MR-1 directly reduced crystalline U(VI) in these two compounds. Laser ablation-inductively coupled plasma mass spectrometry (LA-ICP-MS) analysis implied that interactions occurred between microbial ligands and the residual particles of uranyl compounds. We found that S. oneidensis MR-1 can mediate the dissolution and reduction of crystalline U(VI) through facilitated dissolution and consequent reduction of crystals or direct reduction of U(VI) in crystals. These results help evaluate the environmental fate of solid-phase U(VI), critical for predicting U transport and remediating U-contaminated sites.

Yu Yang; Shuao Wang; Thomas E. Albrecht-Schmitt

2014-01-01T23:59:59.000Z

233

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

234

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

235

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

236

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

237

Parametric acoustic arrays: A Bergen view.  

Science Journals Connector (OSTI)

At the University of Bergen (UoB) Norway research activity in physical acoustics started in the mid?1960s with investigations on the parametric acoustic array (PAA). The newly appointed professor in applied mathematics Sigve Tjo/tta had some years earlier been at Brown University and was inspired by the concept at a fundamental level but also wanted experimental confirmation. No previous acoustical activity existed at UoB. The PAA project was started as a master project at Department of Physics where the main activity was in nuclear high?energy and ionospheric physics. Bellin and Beyers experiment served as a model. The results provided new information on the axial and directional properties of the difference frequency wave field. Inspired by this theoretical modeling continued along with further measurements. Other nonlinear effects like acoustic streaming (boundary layer density gradient) were also investigated. In 1975 a project together with SIMRAD and Norwegian Technical University resulted in a bottom penetrating PAA later commercialized as TOPAS. Numerical modeling based on the KZK equation resulted in the Bergen Code still in use for computing nonlinear acoustic propagation problems. In later years activity at UoB has expanded to encompass linear physical acoustics of various sorts occasionally using PAA as a tool.

2009-01-01T23:59:59.000Z

238

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

239

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

240

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

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241

X-ray spectroscopic studies of microbial transformations of uranium  

SciTech Connect (OSTI)

Several uranium compounds U-metal ({alpha}-phase), UO{sub 2}, U{sub 3}O{sub 8}, {gamma}-UO{sub 3}, uranyl acetate, uranyl nitrate, uranyl sulfate, aqueous and solid forms of 1:1 U:citric acid and 1:1:2 U:Fe:citric acid mixed-metal complexes, and a precipitate obtained by photodegradation of the U-citrate complex were characterized by X-ray spectroscopy using XPS, XANES, and EXAFS. XPS and XANES were used to determine U oxidation states. Spectral shifts were obtained at the U 4f{sub 7/2} and U 4f{sub 5/2} binding energies using XPS, and at the uranium M{sub V} absorption edge using XANES. The magnitude of the energy shift with oxidation state, and the ability to detect mixed-valent forms make these ideal techniques for determining uranium speciation in wastes subjected to bacterial action. The structure of 1:1 U:citric acid complex in both the aqueous and solid state was determined by EXAFS analysis of hexavalent uranium at the L{sub M} absorption edge and suggests the presence of a binuclear complex with a (UO{sub 2}){sub 2}({mu},{eta}{sup 2} {minus}citrato){sub 2} core with a U-U distance of 5.2 {angstrom}. The influence of Fe on the structure of U-citrate complex was determined by EXAFS and the presence of a binuclear mixed-metal citrate complex with a U-Fe distance of 4.8 {angstrom} was confirmed. The precipitate resulting from photodegradation of U-citrate complex was identified as an amorphous form of uranium trioxide by XPS and EXAS.

Dodge, C.J.; Francis, A.J. [Brookhaven National Lab., Upton, NY (United States); Clayton, C.R. [SUNY at Stony Brook, Stony Brook, NY (United States). Dept. of Materials Science and Engineering

1995-10-01T23:59:59.000Z

242

Actinide cation-cation complexes  

SciTech Connect (OSTI)

The +5 oxidation state of U, Np, Pu, and Am is a linear dioxo cation (AnO{sub 2}{sup +}) with a formal charge of +1. These cations form complexes with a variety of other cations, including actinide cations. Other oxidation states of actinides do not form these cation-cation complexes with any cation other than AnO{sub 2}{sup +}; therefore, cation-cation complexes indicate something unique about AnO{sub 2}{sup +} cations compared to actinide cations in general. The first cation-cation complex, NpO{sub 2}{sup +}{center_dot}UO{sub 2}{sup 2+}, was reported by Sullivan, Hindman, and Zielen in 1961. Of the four actinides that form AnO{sub 2}{sup +} species, the cation-cation complexes of NpO{sub 2}{sup +} have been studied most extensively while the other actinides have not. The only PuO{sub 2}{sup +} cation-cation complexes that have been studied are with Fe{sup 3+} and Cr{sup 3+} and neither one has had its equilibrium constant measured. Actinides have small molar absorptivities and cation-cation complexes have small equilibrium constants; therefore, to overcome these obstacles a sensitive technique is required. Spectroscopic techniques are used most often to study cation-cation complexes. Laser-Induced Photacoustic Spectroscopy equilibrium constants for the complexes NpO{sub 2}{sup +}{center_dot}UO{sub 2}{sup 2+}, NpO{sub 2}{sup +}{center_dot}Th{sup 4+}, PuO{sub 2}{sup +}{center_dot}UO{sub 2}{sup 2+}, and PuO{sub 2}{sup +}{center_dot}Th{sup 4+} at an ionic strength of 6 M using LIPAS are 2.4 {plus_minus} 0.2, 1.8 {plus_minus} 0.9, 2.2 {plus_minus} 1.5, and {approx}0.8 M{sup {minus}1}.

Stoyer, N.J. [Lawrence Berkeley Lab., CA (United States); Seaborg, G.T. [Lawrence Livermore National Lab., CA (United States)

1994-12-01T23:59:59.000Z

243

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 centers 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 centers 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

244

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

245

Polarized x-ray-absorption spectroscopy of the uranyl ion: Comparison of experiment and theory  

SciTech Connect (OSTI)

The x-ray linear dichroism of the uranyl ion (UO{sub 2}{sup 2+}) in uranium {ital L}{sub 3}-edge extended x-ray-absorption fine structure (EXAFS), and {ital L}{sub 1}- and {ital L}{sub 3}-edge x-ray-absorption near-edge structure (XANES), has been investigated both by experiment and theory. A striking polarization dependence is observed in the experimental XANES and EXAFS for an oriented single crystal of uranyl acetate dihydrate [UO{sub 2}(CH{sub 3}CO{sub 2}){sub 2}{center_dot}2H{sub 2}O], with the x-ray polarization vector aligned either parallel or perpendicular to the bond axis of the linear uranyl cation (O-U-O). Single-crystal results are compared to experimental spectra for a polycrystalline uranyl acetate sample and to calculations using the {ital ab} {ital initio} multiple-scattering (MS) code FEFF 6. Theoretical XANES spectra for uranyl fluoride (UO{sub 2}F{sub 2}) reproduce all the features of the measured uranyl acetate spectra. By identifying scattering paths which contribute to individual features in the calculated spectrum, a detailed understanding of the {ital L}{sub 1}-edge XANES is obtained. MS paths within the uranyl cation have a notable influence upon the XANES. The measured {ital L}{sub 3}-edge EXAFS is also influenced by MS, especially when the x-ray polarization is parallel to the uranyl species. These MS contributions are extracted from the total EXAFS and compared to calculations. The best agreement with the isolated MS signal is obtained by using nonoverlapped muffin-tin spheres in the FEFF 6 calculation. This contrasts the {ital L}{sub 1}-edge XANES calculations, in which overlapping was required for the best agreement with experiment. {copyright} {ital 1996 The American Physical Society.}

Hudson, E.A.; Allen, P.G.; Terminello, L.J. [Glenn T. Seaborg Institute for Transactinium Science, Lawrence Livermore National Laboratory, University of California, Livermore, California 94551 (United States)] [Glenn T. Seaborg Institute for Transactinium Science, Lawrence Livermore National Laboratory, University of California, Livermore, California 94551 (United States); Denecke, M.A.; Reich, T. [Institut fuer Radiochemie, Forschungszentrum Rossendorf, Postfach 510119, D-01314 Dresden (Germany)] [Institut fuer Radiochemie, Forschungszentrum Rossendorf, Postfach 510119, D-01314 Dresden (Germany)

1996-07-01T23:59:59.000Z

246

Characteristics of a Mixed Thorium-Uranium Dioxide High-Burnup Fuel  

SciTech Connect (OSTI)

Future nuclear fuels must satisfy three sets of requirements: longer times between refueling; concerns for weapons proliferation; and development of a spent fuel form more suitable for direct geologic disposal. This project has investigated a fuel consisting of mixed thorium and uranium dioxide to satisfy these requirements. Results using the SCALE 4.3 code system indicated that the mixed Th-U fuel could be burned to 72 MWD/kg or 100 MWD/kg using 25% of 35% UO2 respectively. The uranium remained below 20% total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MWD was a factor of 4.5 less in the Th-U fuel than in the conventional fuel; Pu-239 production per MWD was a factor of 6.5 less; and the plutonium produced was high in Pu-238, leading to a decay heat 5 times greater than that from plutonium derived from conventional fuel and 40 times greater than weapons grade plutonium. High decay heat would require active cooling of any crude weapon, lest the components surrounding the plutonium be melted. Spontaneous neutron production for plutonium from Th-U fuel was 2.3 times greater than that from conventional fuel and 15 times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO2 is the highest oxide of thorium, while UO2 can be oxidized further to U3O8, ThO2-UO2 fuel may be a superior wasteform if the spent fuel is ever to be exposed to oxygenated water. Even if the cost of fabricating the mixed Th-U fuel is $100/kg greater, the cost of the Th-U fuel is 13% to 15% less than that of the fuels using uranium only.

J. S. Herring; P. E. MacDonald

1999-06-01T23:59:59.000Z

247

Characteristics of a Mixed Thorium - Uranium Dioxide High-Burnup Fuel  

SciTech Connect (OSTI)

Future nuclear fuel must satisfy three sets of requirements: longer times between refueling; concerns for weapons proliferation; and development of a spent fuel form more suitable for direct geologic disposal. This project has investigated a fuel consisting of mixed thorium and uranium dioxide to satisfy these requirements. Results using the SCALE 4.3 code system indicated that the mixed Th-U fuel could be burned to 72 MWD/kg or 100 MWD/kg using 25% and 35% UO2 respectively. The uranium remained below 20 % total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MWD was a factor of 4.5 less in the Th-U fuel than in the conventional fuel; Pu-239 production per MWD was a factor of 6.5 less; and the plutonium produced was high in Pu-238, leading to a decay heat 5 times greater than that from plutonium derived from conventional fuel and 40 times greater than weapons grade plutonium. High decay heat would require active cooling of any crude weapon, lest the components surrounding the plutonium be melted. Spontaneous neutron production for plutonium from Th-U fuel was 2.3 times greater than that from conventional fuel and 15 times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO2 is the highest oxide of thorium, while UO2 can be oxidized further to U3O8, ThO2- UO2 fuel may be a superior wasteform if the spent fuel is ever to be exposed to oxygenated water. Even if the cost of fabricating the mixed Th-U fuel is $100/kg greater, the cost of the Th-U fuel is 13% to 25% less than that of the fuels using uranium only.

Herring, James Stephen; Mac Donald, Philip Elsworth

1999-06-01T23:59:59.000Z

248

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

249

NEAMS Update Quarterly Highlights  

Broader source: Energy.gov (indexed) [DOE]

 The interface of AMP was changed to prepare it for  The interface of AMP was changed to prepare it for integration with Sharp (p. 2).  Bison was enhanced with improved models for cladding and coolant channels (p. 2).  FRAPCON and OECD-NEA databases are being used to evaluate Bison (pp. 2, 6, and 8).  The validation of Bison is being implemented with the recently developed discovery, accumulation, and assessment process (p. 7).  A study of microstructure and fission gas bubbles in UO 2 fuels showed how these characteristics affect fuel thermal

250

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

251

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

252

Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel  

SciTech Connect (OSTI)

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

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

2007-07-01T23:59:59.000Z

253

Retention of neptunium in uranyl alteration phases formed during spent fuel corrosion  

SciTech Connect (OSTI)

Uranyl oxide hydrate phases are known to form during contact of oxide spent nuclear fuel with water under oxidizing conditions; however, less is known about the fate of fission and neutron capture products during this alteration. We describe, the first time, evidence that neptunium can become incorporated into the uranyl secondary phase, dehydrated schoepite (UO{sub 3}{lg_bullet}0.8H{sub 2}O). Based on the long-term durability of natural schoepite, the retention of neptunium in this alteration phase may be significant during spent fuel corrosion in an unsaturated geologic repository.

Buck, E.C.; Finch, R.J.; Finn, P.A.; Bates, J.K.

1997-10-01T23:59:59.000Z

254

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

SciTech Connect (OSTI)

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

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

255

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

256

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

257

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

258

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

259

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

260

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

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.


261

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

SciTech Connect (OSTI)

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

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

2010-09-30T23:59:59.000Z

262

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

263

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.

Dennise Templeton

264

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

E-Print Network [OSTI]

Figure 1. Basic land Resource Areas LEGEND ; ? L B la ck la nd Pra i r ies CO Cocs t Prai r ie FC Forested Coastal Plain FC -C Forested Coastal Plain (Flatwoods) BO Bo t t om lands OT.'TSIDE HEAVY SOLID LI2IE - Boundary of the San Jacinto... 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...

Buckley, Frank A.

1951-01-01T23:59:59.000Z

265

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

266

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.

Dennise Templeton

2013-10-01T23:59:59.000Z

267

Uranium speciation in glass corrosion layers: An XAFS study  

SciTech Connect (OSTI)

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

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

1997-12-31T23:59:59.000Z

268

Final analysis of the GCFR radial blanket and shield integral experiment  

SciTech Connect (OSTI)

An integral experiment has been performed for verification of radiation transport methods and nuclear data used in the design of the radial shield for the proposed gas-cooled fast breeder reactor demonstration plant. The experiment was conducted at the ORNL Tower Shielding Facility and consisted of integral and spectral measurements of the neutron and gamma-ray flux transmitted through slabs of materials which modeled a GCFR-type radial blanket and radial shield. Both UO/sub 2/ and ThO/sub 2/ blankets were investigated as well as several shield designs comprising stainless steel, graphite, and boronated graphite.

Ingersoll, D.T.; Williams, L.R.

1981-04-01T23:59:59.000Z

269

Feather mites (Acarina: Analgesoidea) of some Texan and Mexican falconiform and strigiform birds  

E-Print Network [OSTI]

species found on eacl!. Harris haw!i-Parabuteo unicinc. tu harris! (Audubon) P eudal)biotin&rs rniilvu'in is (Trouessart) Caracara-Caracars c?&eriway audubonii (Cassir&) Ga?&ucinia hyalothrix Gaud and iouchet Red-tailed Hawk-B& tco jama...&censis borealis (Gmelin) Gabucinia tistata sp. n. Swainson's I. awk ? Bi tco swa&'neon! Btnnp:. rte ztveous tex i. . us sp. n. Long horned o I- u!&o virgin!anus G . el! n Protaiecs "ttenuatus (Buchholz) Barred o. rl-Sirix var!a Bangs Dernonoton serve!. us...

Elbihari, Sabir

2012-06-07T23:59:59.000Z

270

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

271

Harvesting and Storing Soybeans.  

E-Print Network [OSTI]

)OC TA245 .7 :73 ), 1543 exas Agricultural Extension Service t/ 8-1543 _~._~----r-" ~'pbJ H~~ P~'pbJ -------------- Harvesting & Storing Soybeans LI BRA RY DEC 11 1986 Texas Agricultural Extension Service. Zerle L. Carpenter, 0 rector... ? ThEA T6XiI 1 eXlS AQlIVl ---- IV"'SU&UoV."'''" ? College Station, Texas HARVESTING AND STORING SOYBEANS Henry O'Neal and Richard E. Withers Extension Agricultural Engineers Travis D. Miller and Arlen D. Klosterboer Extension Agronomists...

O'Neal, Henry; Withers, Richard E.; Miller, Travis D.; Klosterboer, Arlen D.

1986-01-01T23:59:59.000Z

272

Design considerations for inverters in fluid flow control  

E-Print Network [OSTI]

~ Load Curves Bed uo St" tc dead X Lo. S. at, c Head F joe ate Fig. 2. 5 System load curves with difl'erent static head. C. SYSTEM REQUIREMENTS Cost of a variable frequency drive is so moderate and the energy saving potential so great that many... Inverter This type of drive rectifies AC input power and delivers variable DC voltage to a section of the power conversion unit called the inverter section. The inverter section then inverts the variable voltage DC to variable voltage and frequency AC...

Guggari, Mallappa Ishwarappa

2012-06-07T23:59:59.000Z

273

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

274

Residential Energy Consumption Survey: Housing Characteristics,  

Gasoline and Diesel Fuel Update (EIA)

tni tni Residential Energy Consumption Survey: Housing Characteristics, 1981 Energy Information Administration Washington. D.C August 1983 T86T -UJ9AO9 aiji uuojj pasenojnd uaaq (OdO) i|oii)/v\ suoijdijosqns o; Ajdde jou saop aoiiou :e|ON asBa|d 'pjBo^sod at|j noA j| 3Sj| Suiije'Lu vi3 3M1 uo ;u!Buuaj o^sn o} }i ujnja> isnoi nox 'pJBOisod iuB»jodoi! UB aABL) pjnons hoA '}s\\ BujUBUJ VI3 9L|} uo ajB noA|| 'MaiAaj jsij SUJMBUJ suouBOjiqnd |BnuuBS}j BUJ -jonpuoo Sj (vi3) uoijej^siujuupv UOIJBLUJOIUI Afijau^ agj 'uoiieinBaj iuaoiujaAOQ Aq pajmbaj sv 30HON 02-13 maoj aapao ay 05. pa^oajjp aq pus siuamnooa jo 0088-353 (303) S8SOZ "D'Q 'uoiSu-pqsBtt T rao°H 50 UOT^BOLIOJUI

275

Energy Information Administration  

Gasoline and Diesel Fuel Update (EIA)

Washington, 0 C Washington, 0 C Housing Characteristics 1984 i if I ^^^PVrjuV 9861 wo suoiidu.)sqns ot ,< iou Xq sn oj it ujnpj jsnui no^ - via ^Mi uo 3-ic no^ JI ')si -uoo si (VI3) uoiiBJisiuiuipv uoiieuuojui 3DI1ON meuoduii UB noX Suipuas sir jo -986! ' J '9861 uoos [((.w a Xq pwmbw sy (202) jo 0098-2SZ (202) S8S02 0 0 'uoi8u!M«eM 6uip|ing J0| soi aq XSLU si jepjo uy «0|eq jesdde sjaqainu auot|de|a] ptie sessaippv 'QI3N ^Ml oi uo suoqsano '(OI3N) J9iueo uoijeiujojui ASjeug IBUOIIBN S.VI3 aiJi JO Od9 (VI3) uoiiejisiunupy uot;6tux>|ui Xfijaug jat»o pue snji jo aseqajnd pue uorieauofui lueuiuWAOQ 5 Tl 'sjuauunooQ jo luepueiuiJ&dng &LJJ 0104 8iqet!*AB si uoiieoitqnd DOE/EiA-0314(84) Distribution Category UC-98 Residential Energy Consumption v^-^s--. Survey: Housing Characteristics 1984

276

United States Goverment  

Broader source: Energy.gov (indexed) [DOE]

UO/J±0ou4 TcdJ ± O:S'. Aa. ou* o *.I. I 01j ' . UO/J±0ou4 TcdJ ± O:S'. Aa. ou* o *.I. I 01j ' . - - 00E F 1325,8 (08-93) United States Goverment Department of Energy memorandum DATE: August 13, 2007 Audit Report Number: OAS-L-07-19 REPLY TO ATTN OF: IG-32 (A07PR059) SUBJECT: Audit of Executive Compensation at Selected Office of Science Sites TO: Chief Operating, Officer, Office of Science INTRODUCTION AND OBJECTIVE As part of a Department of Energy-wide audit of executive compensation, we reviewed seven Office of Science sites. Specifically, we reviewed executive compensation costs incurred ~,r claim~.- fr- F".*l*- Y. rs 2003, 2 , and 2005 at - Argonne National Laboratory (Argonne), Brookhaven National Laboratory (Brookhaven), Lawrence Berkeley National Laboratory (LBNL), Oak Ridge Institute for Science and Education, Oak Ridge National Laboratory, Princeton Plasma Physics

277

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

278

Proliferation resistance of small modular reactors fuels  

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

279

Additional Information for E-Area Vault Performance Assessment, Appendix I `Suspect Soil Performance` - Results of Modeling the Effects of Organic Matter on the Mobility of Radionuclides as it Relates to the Disposal of Wood Products in E-Area Slit Trenches  

SciTech Connect (OSTI)

Numerous laboratory and field studies have shown that the chemical form (i.e., speciation) of many metals and radionuclides is affected by the presence of naturally occurring organic matter (OM) and its degradation products. The effects of OM (e.g., wood products) on the speciation and, therefore, the mobility of Am, Bk, Cf, Cm, Cs, Ni, NpO{sub 2}, Rb, Sr. UO{sub 2}, and Zr were estimated through use of geochemical and groundwater flow modeling. Due to the complex mixture nature of naturally occurring OM, the OM system was simplified through use of surrogate compounds (citric acid and ethylenedinitrilotetraacetic acid (EDTA)) to estimate effects of OM on radionuclide mobility. Using this approach, OM was found to have no effect on the inventory limits for Cs, NpO{sub 2}, Rb and Zr. The inventory limits for the isotopes of Am, Bk, Cf, Cm, Ni, Pd, PuO{sub 2}, Sr, and UO{sub 2} calculated in the presence of OM decreased over a range of 26 percent for U-233 to 48 percent for Pu-240. The information in this report will be included in the next revision of the E-Area Vaults Performance Assessment.

Serkiz, S.M. [Westinghouse Savannah River Company, AIKEN, SC (United States); Myers, J.L.

1996-04-24T23:59:59.000Z

280

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

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

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

282

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

283

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

284

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

285

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

286

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 AArsine Generation-Atomic Absorption (Sections 5-10), and Test Method BGraphite 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

287

Mechanical Properties of K Basin Sludge Constituents and Their Surrogates  

SciTech Connect (OSTI)

A survey of the technical literature was performed to summarize the mechanical properties of inorganic components in K Basins sludge. The components included gibbsite, ferrihydrite, lepidocrocite and goethite, hematite, quartz, anorthite, calcite, basalt, Zircaloy, aluminum, and, in particular, irradiated uranium metal and uranium dioxide. Review of the technical literature showed that information on the hardness of uranium metal at irradiation exposures similar to those experienced by the N Reactor fuel present in the K Basins (typically up to 3000 MWd/t) were not available. Measurements therefore were performed to determine the hardness of coupons taken from three irradiated N Reactor uranium metal fuel elements taken from K Basins. Hardness values averaged 30 {+-} 8 Rockwell C units, similar to values previously reported for uranium irradiated to {approx}1200 MWd/t. The physical properties of candidate uranium metal and uranium dioxide surrogates were gathered and compared. Surrogates having properties closest to those of irradiated uranium metal appear to be alloys of tungsten. The surrogate for uranium dioxide, present both as particles and agglomerates in actual K Basin sludge, likely requires two materials. Cerium oxide, CeO2, was identified as a surrogate of the smaller UO2 particles while steel grit was identified for the UO2 agglomerates.

Delegard, Calvin H.; Schmidt, Andrew J.; Chenault, Jeffrey W.

2004-12-06T23:59:59.000Z

288

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

289

PROTS-RF: A Robust Model for Predicting Mutation-Induced Protein Stability Changes  

E-Print Network [OSTI]

TS fe at ur es FB oc c 0. 01 34 2 0. 02 86 0. 01 2 2 0. 02 71 2. 26 10 2 22 Th e po te nt ia ld iff er en ce fr om th e oc cu rr en ce of co nt in uo us te tr a- pe pt id e fr ag m en ts FB he l 0. 00 43 0 0. 00 10 0 0. 00 29 2 0. 00 21 0 0. 08 5 Th e... po te nt ia ld iff er en ce fr om th e oc cu rr en ce of co nt in uo us te tr a- pe pt id e fr ag m en ts w hi ch in he lix ,s he et ,c oi l, bu rie d, ex po se d or in te rm ed ia te st at us . FB sh e 0. 00 25 0 2 0. 01 52 0. 00 29 6 0. 00 14 7 0...

Li, Yunqi; Fang, Jianwen

2012-10-15T23:59:59.000Z

290

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

291

Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel  

SciTech Connect (OSTI)

Spent uranium oxide nuclear fuel hosts a variety of trace chemical constituents, many of which must be sequestered from the biosphere during fuel storage and disposal. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the plutonium and neptunium in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO{sub 2} matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl U(VI)O{sub 2}{sup 2+} mineral assemblage that is depleted in plutonium and neptunium relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin ({approx}20 micrometer) layer that is enriched in plutonium and neptunium within a predominantly U{sup 4+} environment. Available data for the standard reduction potentials for NpO{sup 2+}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. Neptunium is an important radionuclide in dose contribution according to performance assessment models of the proposed U. S. repository at Yucca Mountain, Nevada. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of neptunium is needed to predict its behavior at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel [1]. In the immediate vicinity of the spent fuel's surface the redox and nucleation behavior is likely to promote/enhance nucleation of NpO{sub 2} and Np{sub 2}O{sub 5}. Alternatively, Np may be incorporated into uranyl (UO{sub 2}{sup 2+}) alteration phases [2]. In some cases, less-soluble elements such as plutonium will be enriched near the surface of the corroding fuel [3]. We have used focused synchrotron x-rays from the MRCAT beam line at the Advanced Photon Source (APS) at Argonne National Lab to examine a specimen of spent nuclear fuel that had been subject to 10 years of corrosion testing in an environment of humid air and dripping groundwater at 90 C [4]. We find evidence of a region, approximately 20 microns in thickness, enriched in plutonium and neptunium at the corrosion front that exists between the uranyl silicate alteration mineral rind and the unaltered uranium oxide fuel (Figures 1 and 2). The uranyl silicate is itself found to be depleted in these transuranic elements relative to their abundance relative to uranium in the parent fuel. This suggests a low mobility of these components owing to a resistance to oxidize further in the presence of a UO{sub 2}{sup 2+}/U{sup 4+} couple [5].

J.A> Fortner; A.J. Kropf; R.J. Finch; J.C. Cunnane

2006-06-20T23:59:59.000Z

292

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

293

Infrared Spectroscopy of Dioxouranium (V) Complexes with Solvent Molecules: Effect of Reduction  

SciTech Connect (OSTI)

UO2+-solvent complexes having the general formula [UO? (ROH)]+ (R = H, CH?, C?H?, and n-C?H?) were formed using electrospray ionization and stored in a Fourier transform ion cyclotron resonance mass spectrometer, where they were isolated by mass-to-charge ratio, and then photofragmented using a free electron laser scanning through the 10 ?m region of the infrared spectrum. Antisymmetric O=U=O stretching frequencies (?3) were measured for all four complexes, which ranged from ~ 953 cm for H?O to ~ 944 cm for n- PrOH, with the value for the EtOH-containing complex intermediate, systematically decreasing with increasing nucleophilicity of the solvent. The value for the MeOH-containing did not follow the trend, and had a measured ?3 value equal to that of the n-PrOH-containing complex. The ?3 frequency values for these U(V) complexes are comparable to those for the anionic [UO? (NO?)?]- complex, and lower than previously reported values for ligated uranyl (VI) dication complexes by 40 70 cm, and cationic uranyl (VI) ion-pair complexes by 10 40 cm. The lower frequency is attributed to weakening of the O=U=O bonds by repulsion related to reduction of the U metal center, which increases electron density in the antibonding ?* orbitals of the uranyl moiety. Computational modelling of the ?3 frequencies of these species using PBE, B3LYP and LDA functionals showed good agreement with the IRMPD measurements. In general, expected trend in ?3 frequencies expected for the H?O MeOH EtOH n-PrOH series was produced by all three computational methods, however the three alcohols produced very similar values. The inverted order of MeOH and EtOH was not directly accounted for by the models, but is probably the result of overlapping C-H wagging modes that shift the apparent maxima of the O=U=O ?3 absorptions in the MeOH and EtOH complexes.

Groenewold, G. S.; Van Stipdonk, Michael J.; De Jong, Wibe A.; Oomens, Jos; Gresham, Garold L.; McIIwain, Michael E.; Gao, Da; Siboulet, Bertrand; Visscher, Lucas; Kullman, Michael; Polfer, Nick

2008-05-13T23:59:59.000Z

294

Infrared Spectroscopy of Dioxouranium (V) Complexes with Solvent Molecules: Effect of Reduction  

SciTech Connect (OSTI)

UO2+-solvent complexes having the general formula [UO2(ROH)]+ (R = H, CH3, C2H5, and n-C3H7) were formed using electrospray ionization and stored in a Fourier transform ion cyclotron resonance mass spectrometer, where they were isolated by mass-to-charge ratio, and then photofragmented using a free electron laser scanning through the 10 ?m region of the infrared spectrum. Antisymmetric O=U=O stretching frequencies (?3) were measured for all four complexes, which ranged from ~ 953 cm-1 for H2O to ~ 944 cm-1 for n-PrOH, with the value for the EtOH-containing complex intermediate, systematically decreasing with increasing nucleophilicity of the solvent. The value for the MeOH-containing did not follow the trend, and had a measured ?3 value equal to that of the n-PrOH-containing complex. The ?3 frequency values for these U(V) complexes are comparable to those for the anionic [UO2(NO3)3]- complex, and lower than previously reported values for ligated uranyl (VI) dication complexes by 40 70 cm-1, and cationic uranyl (VI) ion-pair complexes by 10 40 cm-1. The lower frequency is attributed to weakening of the O=U=O bonds by repulsion related to reduction of the U metal center, which increases electron density in the antibonding ?* orbitals of the uranyl moiety. Computational modelling of the ?3 frequencies of these species using PBE, B3LYP and LDA functionals showed good agreement with the IRMPD measurements. In general, expected trend in ?3 frequencies expected for the H2O MeOH EtOH n-PrOH series was produced by all three computational methods, however the three alcohols produced very similar values. The inverted order of MeOH and EtOH was not directly accounted for by the models, but is probably the result of overlapping C-H wagging modes that shift the apparent maxima of the O=U=O ?3 absorptions in the MeOH and EtOH complexes.

Gary S. Groenewold; Michael J. Van Stipdonk; Wibe A. de Jong; Jos Oomens; Garold L. Gresham; Michael E. McIlwain; Da Gao; Bertrand Siboulet; Lucas Visscher; Michael Kullman; Nick Polfer; Ivan Infante

2008-06-01T23:59:59.000Z

295

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

296

Microsoft Word - 1aDOE-ID-12-047 Westinghouse EC B3-6 NRC.doc  

Broader source: Energy.gov (indexed) [DOE]

7 7 SECTION A. Project Title: Development of LWR Fuels Enhanced Accident Tolerance - Westinghouse Electric Company LLC SECTION B. Project Description The Westinghouse team, which includes General Atomics, Idaho National Laboratory (INL), Massachusetts Institute of Technology (MIT), Texas A&M University, Edison Welding Institute, Los Alamos National Laboratory, and Southern Nuclear Operating Company, will work to develop fuel and cladding concepts with strong potential to replace the currently used Zr + UO2 fuel system with and enhanced accident tolerant fuel. This will be done by investigating a new fuel system comprised of a cladding capable of surviving high temperatures and significantly reducing any in-core reactions with steam and a high density fuel of increased U-235

297

TO  

Office of Legacy Management (LM)

J. fhith, Chief, J. fhith, Chief, TO : Hew Pork Operations Branoh, D*TE: July 10, 1951 ! I FROM : 'Russell H. BdcL, Chief, Research Service Branch, Berkeley &~'?A - wp+=q - 'I s , This Br'ea desires 500 pounds of U03 for research work on the TTA /: axtraotion process where the subjeot materisl will be spiked with Pu to make a synthetio solution for pilot plant runs. , It is desired that the material be obtained from the Msllinckrodt Chemical Works.similar to that prooured for this Area in 1950 on SR-1649 under the code number: 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. desired. In any event it is the oxid,? which is presently

298

kiedron_RSSoverh_09.ppt  

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

RSS Overhaul Status RSS Overhaul Status March 23, 2009 P. Kiedron and J. Berndt RSS at NOAA RSS Optical Layout New design: CCD chamber New design: CCD chamber * Vacuum chamber CAD design completed * Vacuum chamber manufacturing by A&N Corp. to be completed in 3rd week of April * Vacuum fused-silica window flange purchased * New CCD purchased * New CCD holder design in progress * Vacuum gauge on order SECTION A-A SECTION B-B 1 1 2 2 3 3 4 4 A A B B A&N CORPORATION WILLISTON, FLORIDA (800)FLANGE1 WWW.ANCORP.COM SIZE DRAWING FILE DRAWING NO. REV. SCALE INVENTOR 2009 SHEET OF APPROVALS DATE DRAWN CHECKED MATERIAL FINISH THIRD ANGLE PROJECTION UNLESS OTHERWISE SPECIFIED, DIMENSIONS ARE IN INCHES UNLESS OTHERWISE SPECIFIED, BREAK ALL EDGES .015 X .015 TOLERANCES: .X ±.025 .XX ±.010 .XXX ±.005 ANGLES ±.5 DEG FINISH: 32 µin. MAX, UOS DO

299

EERE PROJECT MAN AGEMENT CENTER NEPA DETFIU.TINATION  

Broader source: Energy.gov (indexed) [DOE]

RTl\IENT OF ENERGY RTl\IENT OF ENERGY EERE PROJECT MAN AGEMENT CENTER NEPA DETFIU.TINATION Page 1 of2 RECIPIENT:CT Department of Energy and Environmental Protection STATE: CT PROJECf TITLE: CONNECTICUT SEP ANNUAL PY12 Funding Opportunity Announcement Number PrCK':urement Instrument Number NEPA Control Number CID Number DE-FOA.Q000643 DE-EEOOO5301 GF0-0005301-OO 1 Based on my review of the informlltioD concerning the proposed action, as NEPA Compliance Officer (authorized under DOE Order 451.1A),1 have made the following determination: ex, EA, EIS APPENDIX AND NUMBER: De~ription : All Technical advice and assistance to org an izations AS InformaUo n gat herin g, analysis, and dissemi nation Rational for determination: Technical advice and planning assistance to Intemational, national

300

Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study |  

Broader source: Energy.gov (indexed) [DOE]

LWR Nuclear Fuel Cladding System Development Trade-off LWR Nuclear Fuel Cladding System Development Trade-off Study Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study The LWR Sustainability (LWRS) Program activities must support the timeline dictated by utility life extension decisions to demonstrate a lead test rod in a commercial reactor within 10 years. In order to maintain the demanding development schedule that must accompany this aggressive timeline, the LWRS Program focuses on advanced fuel cladding systems that retain standard UO2 fuel pellets for deployment in currently operating LWR power plants. The LWRS work scope focuses on fuel system components outside of the fuel pellet, allowing for alteration of the existing zirconium-based clad system through coatings, addition of ceramic sleeves, or complete replacement

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301

Microsoft Word - DOE-ID-13-071 Hunter College EC B3-6.doc  

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

1 1 SECTION A. Project Title: Recovery of Uranium from Seawater: Polymer-Supported Aminophosphinates as Selective - Hunter College of the City University of New York SECTION B. Project Description Hunter College proposes to develop a polymer-supported extractant for the recovery of uranium from seawater. Work includes evaluating polymers with the highest capacities with authentic seawater at 3 ppb UO 2 2+ . SECTION C. Environmental Aspects / Potential Sources of Impact Chemical Use/Storage / Chemical Waste Disposal - Per week use includes 1 L each of dioxane, toluene, pyrrolidone, THF, and much smaller quantities of chemicals such as chlorodiethylphosphate, vinybenzyl chloride, triethylphophite, pentaerythritol, dilute acids and bases. Wastes will be managed according to the Hunter College Hazardous Waste Management Plan with waste collection performed

302

Presentations | MMSNF 2013 Chicago  

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

Privacy and Security Notice Presentations Available Presentations from the Materials Modeling and Simulation of Nuclear Fuels (MMSNF) 2013 workshop. Presented on Presentation Title Authors Session Oct. 14, 2013 Welcome and announcements Ewing, Tom (ANL, USA) and Rosner, Robert (UC, USA) Opening Oct. 14, 2013 First-principles DFT+U modeling of paramagnetic UO2 and (U,Pu) mixed oxides [366KB, posted: Oct. 13, 2013 ] Dorado, Boris (CEA, DAM, DIF, France), Garcia, Philippe (CEA, DEN, DEC, France) Atomistic Models and Simulations Oct. 14, 2013 Computational study of energetics and defect-ordering tendencies for rare earth elements in uranium dioxide [1.5MB, posted: Oct. 28, 2013 ] Solomon, Jonathan M. (UC Berkeley, USA), Alexandrov, Vitaly (UC Berkeley, UC Davis, USA), Sadigh, Babak (LLNL, USA), Navrotsky, Alexandra (UC Davis, USA), Asta, Mark (UC Berkeley, UC Davis, USA) Atomistic Models and Simulations

303

untitled  

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

tuo weeks. tuo weeks. For a personal retention copy, call Tech. Info. Division, Ext. 5545 -- - TJNIVERSITY O F CALIFORNIA Radiation Laboratory C ont rae t No, W-74.05-eng-48 THE PATH OF CARBON I N PHOTOSYNTHESIS, X U , KINETIC REIATIORSEIPS OF THE I N T m ~ ~ I A T E S IN sTum STATE PHOTOSYT\JTHESIS A , A. Benson, S . Icawaguchi, F, Hayes and M, Calvfr, Berkeley, Gallfomlh KlMETIC RELATIONSHIPS OF THE INTEiQBDIATLS 3 3 STEADH STATE E'HOTOSY NTHES IS A, A, Benson, So hawaguchf, Po Hayes and M, Calvin Ibadiation Laboratory and liegwtment 0% Chemistry University of California, Berkeley 1 A kinetic study of the accumulation of cL4 in the intermediates of steady. state photosynthesis in cUO2 provides information regarding the sequence of reactfona involvedo The work described applied the rpdfo-

304

Microsoft Word - ICEM05_DCURETE.doc  

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

6 6 OPTIMIZATION OF COMPOSITION AND PRODUCTION TECHNOLOGY OF HIGH-DENSITY CONCRETE WITH CERAMIC AGGREGATE BASED ON DEPLETED URANIUM DIOXIDE S.G. Ermichev, V.I. Shapovalov, N. V. Sviridov (RFNC-VNIIEF, Sarov, Russia) V.K. Orlov, V.M. Sergeev, A. G. Semyenov, A.M. Visik, A.A. Maslov, A. V. Demin, D.D. Petrov, V.V. Noskov, V. I. Sorokin, O. I. Yuferov (VNIINM, Moscow, Russia) L. Dole (ORNL, Oak Ridge, USA) ABSTRACT Russian is researching the production and testing of concretes with ceramic aggregate based on depleted uranium dioxide (UO 2 ). These DU concretes (DUCRETE) are to be used as structural and radiation-shielded material for casks for A-plant spent nuclear fuel transportation and storage. This paper presents the results of studies aimed at selection of ceramics

305

BISON Enhanced | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

BISON Enhanced BISON Enhanced BISON Enhanced January 29, 2013 - 10:42am Addthis Pin-scale Code Development A mechanistic, smeared fuel cracking model for UO2 has been implemented in BISON and tested with simulations of IFA-432 Rod 1, an experiment conducted in the Halden reactor. ("Smeared" refers to the fact that cracks are represented in aggregate, rather than as discrete, individual cracks.) Failure to account for fuel cracking can result in temperature predictions that are off by as much as 200°C at beginning-of-life. Excellent agreement between prediction and measurement is obtained when an empirical correlation for fuel relocation is used (as in the NRC's FRAPCON), which was expected, since the empirical correlation was fit to data that included this experiment; this result is important, however, in

306

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

Office of Legacy Management (LM)

MO 0-01 MO 0-01 FUSRAP Considered Sites Site: SPENCER CHEMICAL CO. (MO.0-01) Eliminated from further consideration under FUSRAP - an AEC licensed operation Designated Name: Not Designated Alternate Name: Jayhawk Works MO.0-01-1 Location: Joplin , Missouri MO.0-01-1 Evaluation Year: 1985 MO.0-01-2 Site Operations: Processed enriched uranium (UF-6) and scrap to produce primarily uranium dioxide (UO-2) under AEC licenses. MO.0-01-3 MO.0-01-4 Site Disposition: Eliminated - No Authority MO.0-01-2 Radioactive Materials Handled: Yes Primary Radioactive Materials Handled: Normal and Enriched Uranium, Thorium MO.0-01-6 Radiological Survey(s): Yes MO.0-01-5 Site Status: Eliminated from further consideration under FUSRAP - an AEC licensed operation Also see Documents Related to SPENCER CHEMICAL CO.

307

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

Office of Legacy Management (LM)

KS 0-01 KS 0-01 FUSRAP Considered Sites Site: SPENCER CHEMICAL CO. (KS.0-01 ) Eliminated from further consideration under FUSRAP - an AEC licensed operation Designated Name: Not Designated Alternate Name: Jayhawk Works KS.0-01-1 Location: Pittsburg , Kansas KS.0-01-1 Evaluation Year: 1985 KS.0-01-2 Site Operations: Processed enriched uranium (UF-6) and scrap to produce primarily uranium dioxide (UO-2) under AEC licenses. KS.0-01-3 KS.0-01-4 Site Disposition: Eliminated - No Authority - AEC licensed KS.0-01-2 Radioactive Materials Handled: Yes Primary Radioactive Materials Handled: Normal and Enriched Uranium; Thorium KS.0-01-6 Radiological Survey(s): Yes KS.0-01-5 Site Status: Eliminated from further consideration under FUSRAP - an AEC licensed operation

308

Distribution Category: Atomic, Molecular, and Chemical Physics  

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

Atomic, Atomic, Molecular, and Chemical Physics (UC-411) ARGONNE NATIONAL LABORATORY 9700 South Cass Avenue Argonne, TIlinois 60439 ANLI APSILS-151 RESULTS OF DESIGN CALCULATIONS FOR THE MODULATOR OF THE CROSSED FIELD UNDULATOR DEVICE by Roland S8:voy Advanced Photon Source August 1990 Work sponsored by ~--~,P:a7te~n7t~C~le-.a-re-d--b\-!------ Pen"" .... + D - CII, epartrnent, AND R':-lr-!, ("'1:' ' "'"",,, l... ,r:.. ,'\')k. . f\UTHOF?IZED BY 1l;J6r1l11Cal Publications Ser " O(;ite~ ~ 'vjces Technicallnf ~avld R .* ·i;;~rln - ormatIon Services, ANL Uo So DEPARTMENT OF ENERGY Office of Energy Research 1 Abstract: The modulator in the crossed field undulator device is used to shift the

309

Coolant Sub-Channel and Smeared-Cracking Models in BISON | Department of  

Broader source: Energy.gov (indexed) [DOE]

Coolant Sub-Channel and Smeared-Cracking Models in BISON Coolant Sub-Channel and Smeared-Cracking Models in BISON Coolant Sub-Channel and Smeared-Cracking Models in BISON January 29, 2013 - 10:45am Addthis Coolant Sub-Channel and Smeared-Cracking Models in BISON A single-pin coolant sub-channel model was implemented in BISON, the pin-scale simulation code. This enables BISON to compute the heat transfer coefficient and coolant temperature as a function of axial position along the fuel pin (rather than requiring this information to be supplied by the user). At present, the model is only applicable to pressurized water reactor coolant conditions, but modifications to include boiling water reactor (BWR) coolant conditions are in progress. A preliminary UO2 thermal and irradiation creep model has been implemented in BISON and is

310

Posters | MMSNF 2013 Chicago  

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

Privacy and Security Notice Privacy and Security Notice Posters Available Posters from the Materials Modeling and Simulation of Nuclear Fuels (MMSNF) 2013 workshop. Presented on Poster Title Author(s) ID Session Oct. 14, 2013 Atomistic simulation of radiation damage of metallic and oxide fuels by swift heavy ion irradiation Starikov, Sergey (JIHT RAS, Russia), Pisarev, Vasily (JIHT RAS, Russia), Kuksin, Alexey (JIHT RAS, Russia), Stegailov, Vladimir (JIHT RAS, Russia) PA4 A Oct. 14, 2013 Density functional theory for fission products transport in UO2 [2.5MB, posted: Oct. 28, 2013 ] Ducher, Roland (IRSN, France), Dubourg, Roland (IRSN, France) PA5 A Oct. 14, 2013 Kinetic Monte Carlo study of oxygen defect migration in urania fuel Hoffman III, Richard T. (GA Tech, USA), Bahera, Rakesh (GA Tech, USA), Deo, Chaitanya S. (GA Tech, USA) PA7 A

311

CMSNF | U.S. DOE Office of Science (SC)  

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

CMSNF CMSNF Energy Frontier Research Centers (EFRCs) EFRCs Home Centers Research Science Highlights News & Events Publications Contact BES Home Centers CMSNF Print Text Size: A A A RSS Feeds FeedbackShare Page Center for Materials Science of Nuclear Fuel Director(s): Todd Allen Lead Institution: Idaho National Laboratory Mission: 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. Research Topics: phonons, thermal conductivity, nuclear (including radiation effects), defects, materials and chemistry by design Materials Studied: MATERIALS: actinide INTERFACES: solid/solid NANOSTRUCTURED MATERIALS: 3D

312

BISON Enhanced | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Enhanced Enhanced BISON Enhanced January 29, 2013 - 10:42am Addthis Pin-scale Code Development A mechanistic, smeared fuel cracking model for UO2 has been implemented in BISON and tested with simulations of IFA-432 Rod 1, an experiment conducted in the Halden reactor. ("Smeared" refers to the fact that cracks are represented in aggregate, rather than as discrete, individual cracks.) Failure to account for fuel cracking can result in temperature predictions that are off by as much as 200°C at beginning-of-life. Excellent agreement between prediction and measurement is obtained when an empirical correlation for fuel relocation is used (as in the NRC's FRAPCON), which was expected, since the empirical correlation was fit to data that included this experiment; this result is important, however, in

313

HIGH-DENSITY CONCRETE WITH CERAMIC AGGREGATE BASED ON DEPLETED URANIUM DIOXIDE  

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

DENSITY CONCRETE WITH CERAMIC AGGREGATE BASED ON DEPLETED URANIUM DENSITY CONCRETE WITH CERAMIC AGGREGATE BASED ON DEPLETED URANIUM DIOXIDE S.G. Ermichev, V.I. Shapovalov, N.V.Sviridov (RFNC-VNIIEF, Sarov, Russia) V.K. Orlov, V.M. Sergeev, A. G. Semyenov, A.M. Visik, A.A. Maslov, A. V. Demin, D.D. Petrov, V.V. Noskov, V. I. Sorokin, O. I. Uferov (VNIINM, Moscow, Russia) L. Dole (ORNL, Oak Ridge, USA) Abstract - Russia is researching the production and testing of concretes with ceramic aggregate based on depleted uranium dioxide (UO 2 ). These DU concretes are to be used as structural and radiation-shielded material for casks for A-plant spent nuclear fuel transportation and storage. This paper presents the results of studies aimed at selection of ceramics and concrete composition, justification of their production technology, investigation of mechanical properties, and chemical stability.

314

Pierluigi Mancarella  

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

Pierluigi Mancarella Pierluigi Mancarella Lecturer Sustainable Energy Systems, School of Electrical and Electronic Engineering University of Manchester, UK p.mancarella@manchester.ac.uk This speaker was a visiting speaker who delivered a talk or talks on the date(s) shown at the links below. This speaker is not otherwise associated with Lawrence Berkeley National Laboratory, unless specifically identified as a Berkeley Lab staff member. Dr. Pierluigi Mancarella is a Lecturer in Sustainable Energy Systems in the School of Electrical and Electronic Engineering, University of Manchester (UoM), UK. He is part of the Electrical Energy and Power Systems (EEPS) group and teaches "Power systems operation and economics" and "Smart Grid and sustainable electricity systems" in the Electrical Power Systems

315

Photostat Price S /  

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

Photostat Price S / Photostat Price S / . p d Microfilm Price $ /- 80 Available from the Office of Technical Services Department of Commerce Washington 25, D. C. A. ifetallurgi c a l Pro.1 ect PHYSICS rnSEARR u E. Fermi, Division Director; G a l e Young, Section Chief * * * . - 1 I - t khCALC'ULATIOM OF TEIE CRITICAL SIZE AND MULTIPUCATIQ! , . - . - L C O N S T A N T OF A H@dOGENBOUS UO2 - DZO MIXTURFS E . P. Nigner, A. M. Ileinberg, J, Stephenson February 11, 1944 The roultiplication constant w d optimal concentra- tion of a slurry p i l e is recalculated on the basis of Uitchell's re'cmt experiments on resonance absorption. -\ The smallest chain reacting unit contains &S t o 55 m3 of D~O. DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the

316

Effect of environmental factors on the persistence of fluorodifen on soils  

E-Print Network [OSTI]

in fluorodifen content occurred after 1 week. A combination of 2l~ C with short wave ultraviolet li ht (wave lenqtb 253 mu~ intensity 200 uM/cm ) resulted in a siUnificant 35v loss of fluorodifen. Samples subjected to long wave ultraviolet light (wave length...' pl m m w g ID m? c DP 0 m C c m Pl ~ O ID IP tp I IP m lp m cp DO m' m&' +7 g g rl IP ID oog 4 m lp m& o, D p p p p p lPi cued Tpo LOOTS. uoTq. ueq. ea queo ze 21 as or2anic matter present in those soils havin~ a hi...

Wiemers, Garland Ray

2012-06-07T23:59:59.000Z

317

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

318

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

319

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

320

The use of oxidation-reduction dyes in the determination of the shelf-life of meats  

E-Print Network [OSTI]

!. hvle?o l, lu . re inc t r'on th ie 21 0 2 4 6 8 IO l2 REDUCTION TIN(E (HOURS) I'~Coot C. Rolaticn botuo o now'&o o o. ' rt(it oo, " ~i;m- io r'sic' o;. t ttoz 1. it ni . 't. tio . t it. 22 ~t 7 C & 0 3- I 0 2 4 6 8 10 12 l4 REDUCTIOI'J Tll... 0 0 4 4 0 0 0 0 4 4 0 0 00 0 4 0 y = 7 8131 ?. 3069X e k kkk 0 2 4 6 8 10 12 14 REDUCTION TIME (HOURS) Figure 19. Relation b tw en number of o: idase-uos itive bacterra in all products and rcsasurin red iction tim e @ 0 Z, 6...

Bush, Janis Carolyn

2012-06-07T23:59:59.000Z

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321

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

322

Preliminary Criticality Analysis of Degraded SNF Accumulations to a Waste Package (SCPB: N/A)  

SciTech Connect (OSTI)

This study is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide input to a separate evaluation on the probability of criticality in the far-field environment. These calculations are performed in sufficient detail to provide conservatively bounding configurations to support separate probabilistic analyses. The objective of this evaluation is to provide input to a risk analysis which will show that criticalities involving commercial spent nuclear fuel (SNF) are not credible, or indicate additional measures that are required for the Engineered Barrier Segment (EBS) to make such events incredible. Minimum critical volumes and masses of UO{sub 2}/H{sub 2}O/tuff mixtures are determined without application of regulatory safety limits. This study does not address or demonstrate compliance with regulatory limits.

J.W. Davis

2005-12-15T23:59:59.000Z

323

Russian-US collaboration on implementation of the active well coincidence counter (AWCC)  

SciTech Connect (OSTI)

The feasibility of using a standard AWCC at the Obninsk IPPE has been demonstrated through active measurements of single UO{sub 2} (36% enriched) disks and through passive measurements of plutonium metal disks used for simulating reactor cores. The role of the measurements is to verify passport values assigned to the disks by the facility, and thereby facilitate the mass accountability procedures developed for the very large inventory of fuel disks at the facility. The AWCC is a very flexible instrument for verification measurements of the large variety of nuclear material items at the Obninsk IPPE and other Russian facilities. Future work at the IPPE will include calibration and verification measurements for other materials, both in individual disks and in multi-disk storage tubes; it will also include training in the use of the AWCC.

Mozhajev, V.; Pshakin, G. [Gosudarstvennyj Komitet po Ispol`zovaniyu Atomnoj Ehnergii SSSR, Obninsk (Russian Federation). Fiziko-Ehnergeticheskij Inst.; Stewart, J.; Hatcher, C.; Krick, M.; Menlove, H.; Siebelist, R. [Los Alamos National Lab., NM (United States)

1996-09-01T23:59:59.000Z

324

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

325

Effect of Gadolinium Doping on the Air Oxidation of Uranium Dioxide  

SciTech Connect (OSTI)

Researchers at the Pacific Northwest National Laboratory (PNNL) investigated the effects of gadolinia concentration on the air oxidization of gadolinia-doped uranium dioxide using thermogravimetry and differential scanning calorimetry to determine if such doping could improve uranium dioxide's stability as a nuclear fuel during potential accident scenarios in a nuclear reactor or during long-term disposal. We undertook this study to determine whether the resistance of the uranium dioxide to oxidation to the orthorhombic U3O8 with its attendant crystal expansion could be prevented by addition of gadolinia. Our studies found that gadolinium has little effect on the thermal initiation of the first step of the reported two-step air oxidation of UO2; however, increasing gadolinia content does stabilize the initial tetragonal or cubic product allowing significant oxidation before the second expansive step to U3O8 begins.

Scheele, Randall D.; Hanson, Brady D.; Cumblidge, Stephen E.; Jenson, Evan D.; Kozelisky, Anne E.; Sell, Rachel L.; MacFarlan, Paul J.; Snow, Lanee A.

2004-12-04T23:59:59.000Z

326

X-ray emissions in 3d, 4d, and 5d ranges for uranium ions  

Science Journals Connector (OSTI)

Radiative decay of nd-15fm+1 excited states in UO2 induced by electron collisions is studied theoretically and experimentally. Energies, transition probabilities, and photoexcitation cross sections for the relevant configurations of U4+ are calculated by using the multiconfiguration Dirac-Fock method. Experimental observations are made in the 4d range. Direct recombination of the excited 5f electron to the 4d hole and 4d-6p emission in the presence of the spectator excited 5f electron are observed. From the theoretical results, the spectra are simulated and compared to the observed spectra in the three nd regions. The agreement is correct and describes the evolution of the coupling scheme in the nd-15f3 excited states from n=3 to n=5.

C. Bonnelle; P. Jonnard; C. Barr; G. Giorgi; J. Bruneau

1997-05-01T23:59:59.000Z

327

Physicochemical characterisation of depleted uranium (DU) particles at a UK firing test range  

Science Journals Connector (OSTI)

Depleted uranium (DU) particles were isolated from soils at Eskmeals, UK, where DU munitions have been tested against hard targets and unfired DU buried in soils for corrosion studies. Using electron microscopy and X-ray analyses, three classes of particles were identified: (1) DU aerosols and fragments, typically 120?m diameter, composed mainly of uranium as UO2 and U3O8, (2) solidified molten particles, typically 200500?m diameter, composed of U, mixed with Fe from target materials and (3) deposits and coatings, often of metaschoepite on sand grains up to 500?m diameter. The first two particle types are derived from firing impacts, the last from corrosion of buried uranium metal. Alpha and mass spectrometry allowed quantitative elemental and isotopic characterisation of DU-containing particulate environmental samples.

Mustafa Sajih; Francis R. Livens; Rebeca Alvarez; Mathew Morgan

2010-01-01T23:59:59.000Z

328

Action Sheet 36 Final Report  

SciTech Connect (OSTI)

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

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

2012-02-24T23:59:59.000Z

329

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

330

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

E-Print Network [OSTI]

zo 'Ogg'Ot", eg By sqdeg~otld etlS yo ayevs aqua, 6uyddsm tran pyB uz so pesn mes gyp' 'gp xeq6aaog pa~ap 8y-H88 B&g&&s po 908 GGT Pue 89-H88 Se'F&es ? 001 K6 pu- '9L 89 'l'Y 9 stlde&6O&+M ~ d". m ayboyoa6 ycuts e'. l ea&emT ~ pass. oBTB OSCst...i) a6pfxa pun 'sauxea 'pnoXO ~q (TX6T)?fed 9o uoygeezog ufegunog dey yeuy6yxo a~ uzoxg pauygapax semi Cpu&N 0%8*-FI II QV li N X 9 BIIZ 'uu7%$ usg 6 xacgi18'g Quogsalufg u'f Qgunog der 'uoT6ax ouegg -oxd pue sxagy~~ e poo6 Exalt axe spues Ezoqo...

Chauvin, Aaron Lawrence

2012-06-07T23:59:59.000Z

331

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

332

Polarization Dependent High Energy Resolution X-ray Absorption Study of Dicesium Uranyl Tetrachloride  

Science Journals Connector (OSTI)

Tonya Vitova *, Jennifer C. Green , Robert G. Denning , Matthias Lble , Kristina Kvashnina , Joshua J. Kas ?, Kevin Jorissen ?, John J. Rehr ?, Thomas Malcherek ?, and Melissa A. Denecke ... This study confirms theoretically predicted electronic transitions to U 6dxy (6d?) orbital and measures relative energies of U 5f?, 5f?, 5f?, and 5f? orbitals of oriented uranyl (UVIO22+) in the same spectrum with remarkable energy resolution. ... We have obtained angle-resolved electronic structure information for oriented Cs2UO2Cl4 crystal, specifically relative energies of 5f and 6d valence orbitals probed with extraordinary energy resolution by polarization dependent high energy resolution X-ray absorption near edge structure (PD-HR-XANES) and compare these with predictions from quantum chemical Amsterdam density functional theory (ADF) and ab initio real space multiple-scattering Greens function based FEFF codes. ...

Tonya Vitova; Jennifer C. Green; Robert G. Denning; Matthias Lble; Kristina Kvashnina; Joshua J. Kas; Kevin Jorissen; John J. Rehr; Thomas Malcherek; Melissa A. Denecke

2014-12-08T23:59:59.000Z

333

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

334

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

335

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

336

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

337

Bridging the Gap in the Chemical Thermodynamic Database for Nuclear Waste Repository: Studies of the Effect of Temperature on Actinide Complexation  

SciTech Connect (OSTI)

Recent results of thermodynamic studies on the complexation of actinides (UO{sub 2}{sup 2+}, NpO{sub 2}{sup +} and Pu{sup 4+}) with F{sup -}, SO{sub 4}{sup 2-} and H{sub 2}PO{sub 4}{sup -}/HPO{sub 4}{sup 2-} at elevated temperatures are reviewed. The data indicate that, for all systems except the 1:1 complexation of Np(V) with HPO{sub 4}{sup 2-}, the complexation of actinides is enhanced by the increase in temperature. The enhancement is primarily due to the increase in the entropy term (T{Delta}S) that exceeds the increase in the enthalpy ({Delta}H) as the temperature is increased. These data bridge the gaps in the chemical thermodynamic database for nuclear waste repository where the temperature could remain significantly higher than 25 C for a long time after the closure of the repository.

Rao, Linfeng; Tian, Guoxin; Xia, Yuanxian; Friese, Judah I.; Zanonato, PierLuigi; Di Bernardo, Plinio

2009-12-21T23:59:59.000Z

338

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

339

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

340

Uranium removal by chitosan impregnated with magnetite nanoparticles: adsorption and desorption  

Science Journals Connector (OSTI)

A magnetic biosorbent composed of nanoparticles of magnetite covered with chitosan, denominated magnetic chitosan, was prepared. The magnetic chitosan showed a magnetic response of intense attraction in the presence of a magnetic field without becoming magnet, a typical behaviour of superparamagnetic material. Its adsorption performance was evaluated by the adsorption isotherm models of Langmuir and Freundlich for uranium ions, and the desorption behaviour using carbonate and oxalate ions was investigated. The adsorption equilibrium data fitted well to the Langmuir model, being the maximum adsorption capacity equal 42 mg g?1. In the desorption studies, 94% of recovered UO2+2 with carbonate ion was verified under the conditions studied. The chitosan, available as a byproduct of marine food processing, is environmentally safe and can be a low cost adsorbent for U removal from wastewater. The magnetic chitosan as adsorbent of U to treat radioactive wastewater is a sustainable technology.

Luiz Claudio Barbosa Stopa; Mitiko Yamaura

2010-01-01T23:59:59.000Z

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,
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We encourage you to perform a real-time search of NLEBeta
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341

Equilibrium adsorption isotherm of U(VI) at pH 4 and pH 5 onto synthetic magnetite nanoparticles  

Science Journals Connector (OSTI)

The adsorption features of synthetic magnetite nanoparticles have been investigated for removal of uranyl ions from nitric solutions. These magnetic nanoparticles can be used to adsorb contaminants in wastewater and can subsequently be removed from the medium by a magnetic process because of their superparamagnetic behaviour. The magnetite nanoparticles were synthesised by simultaneous precipitating of Fe2+ and Fe3+ ions in a NaOH solution. Batch experiments were carried out to investigate the adsorption of UO22+ ions from nitric solution, pH 4 and 5, onto magnetic nanoparticles. Two minutes of application of a magnetic field were sufficient for removing the magnetite nanoparticles from the liquid medium. Equilibrium adsorption isotherm was evaluated using Freundlich and Langmuir models. The adsorption was between 40% and 80% under conditions studied.

Roberto Leal; Mitiko Yamaura

2011-01-01T23:59:59.000Z

342

Analysis of a proposed fuel freezing mechanism in a rod bundle  

E-Print Network [OSTI]

void thermite fuel simulant (80 w/o UO2, 20 w/o Mo at about 3470 K, approximately 330 K superheat) and stainless steel walls. The main system variables in these tests were the initial wall temperatures and the fuel downward injection pressures... 215. 4 NPa cr 4. 0 90 3. 0 F4 85 K 1 P = 2. 0 MPa 80 75 t 70 0 10 20 30 40 50 60 x (cm) Fig. 7. Total stress in the crust at the steel melting front for T = 373 K. wo 35 100 5. 0 cT = 202. 3 MPa cr 95 4. 0 90 3. 0 e t4 85 P 2. 0...

Nguyen-Wayne, David Loc

1983-01-01T23:59:59.000Z

343

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

344

Joint operation of wind farm, photovoltaic, pump-storage and energy storage devices in energy and reserve markets  

Science Journals Connector (OSTI)

Abstract Renewable resources generation scheduling is one of the newest problems of the power markets. In this paper, joint operation (JO) of wind farms (WF), pump-storage units (PSU), photo-voltaic (PV) resources, and energy storage devices (ESD) is studied in the energy and ancillary service markets. There are uncertainties in wind power generation (WPG), photovoltaic power generation (PVPG) and the market prices. To model these uncertainties, the WPG is forecasted by using ARMA model and its scenarios are generated using Weibull distribution function. Moreover, other uncertain parameters are forecasted first, and their uncertainties are modeled by using scenario generation and scenario reduction method. The proposed JO method is used to determine the optimal bidding strategy of the PSU, PV, ESD and WF of IEEE 118-bus standard system. The results for these renewable energy resources confirm that the JO of these resources increases the profit and decreases the risk of the resources in comparison with their uncoordinated operation (UO).

Moein Parastegari; Rahmat-Allah Hooshmand; Amin Khodabakhshian; Amir-Hossein Zare

2015-01-01T23:59:59.000Z

345

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

346

Thermal Controls for the On-Site Transfer of Mixed Oxide Scrap  

SciTech Connect (OSTI)

Mixed oxide scrap consisting primarily of PuO{sub 2} and UO{sub 2} is stored in crimp-sealed product cans at Savannah River Site (SRS). The product cans are to be transported onsite to a processing facility for dissolution using an earlier version of the 9975 (prior to the redesigned drum closure) package called DDF-1. This paper compares the maximum plutonium temperatures inside the DDF-1 and the maximum temperatures when the product can is in a storage vault. The comparison shows that the maximum Pu temperature for low wattage cans are marginally higher during transport provided the drum packages are kept out of sunlight. At higher wattage the differences become significant. The application of this work is to provide guidance and an estimate of temperature sensitive chemical reactions during transport compared with storage.

Gupta, N.K.

2001-05-08T23:59:59.000Z

347

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

348

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 1980s. 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 1250C 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 1150C, 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 1250C, 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 1150C 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

349

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

SciTech Connect (OSTI)

Reductive biostimulation is currently being explored as a possible remediation strategy for uranium (U) contaminated groundwater, and is currently being investigated at a field site in Rifle, CO, USA. The long-term stability of the resulting U(IV) phases is a key component of the overall performance and depends upon a variety of factors, including rate and mechanism of reduction, mineral associations in the subsurface, and propensity for oxidation. To address these factors, several approaches were used to evaluate the redox sensitivity of U: measurement of the rate of oxidative dissolution of biogenic uraninite (UO2(s)) deployed in groundwater at Rifle, characterization of a zone of natural bioreduction exhibiting relevant reduced mineral phases, and laboratory studies of the oxidative capacity of Fe(III) and reductive capacity of Fe(II) with regard to U(IV) and U(VI), respectively.

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

2011-03-26T23:59:59.000Z

350

Interaction of noble-metal fission products with pyrolytic silicon carbide  

SciTech Connect (OSTI)

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain layers of pyrolytic carbon and silicon carbide, which act as a miniature pressure vessel and form the primary fission product barrier. Of the many fission products formed during irradiation, the noble metals are of particular interest because they interact significantly with the SiC layer and their concentrations are somewhat higher in the low-enriched uranium fuels currently under consideration. To study fission product-SiC interactions, particles of UO/sub 2/ or UC/sub 2/ are doped with fission product elements before coating and are then held in a thermal gradient up to several thousand hours. Examination of the SiC coatings by TEM-AEM after annealing shows that silver behaves differently from the palladium group.

Lauf, R.J.; Braski, D.N.

1982-01-01T23:59:59.000Z

351

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

352

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

353

Redox Energetics and Kinetics of Uranyl Coordination Complexes in Aqueous Solution  

Science Journals Connector (OSTI)

The redox chemistry of uranium is remarkably rich and diverse as a consequence of the number of readily accessible oxidation states (III?VI) and the sensitivity of the redox potentials to the coordination environment around the metal. ... 1-3 In aqueous environments, the most stable oxidation state for uranium is the hexavalent state for which the dominant species is the uranyl ion, UO22+.1-3 The redox chemistry of uranyl has been extensively investigated and manipulated, particularly for separations and process chemistry relating to the nuclear fuel cycle and other defense-related applications. ... The aquo, carbonato, acetato, and chloro complexes of uranyl are all known or potentially relevant species in environmental groundwaters, and the hydroxo and carbonato complexes are relevant species in alkaline high-level radioactive waste storage tanks associated with nuclear fuel processing. ...

David E. Morris

2002-05-31T23:59:59.000Z

354

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

355

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

356

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

357

Repellents to prevent cattle browsing of pine seedlings  

E-Print Network [OSTI]

Tm ZzaA ssA Zqyaywm. c". pqZ gso~qv ~ 'ZBTTsq=om qsmqsu 'BuTTfmzq zus Buys?ozq syqqev usqq usqqo eaeuss moTg sBomsTr Buy~aaB syqqso Zq pseuss sar~ cq. oqd: &~ssqo uo Zqyqs~~crz s~? ~o fiqTuopem sqq. puc 'eBuyqTsae Bssp go aBsquoozacT q. eaqDyu sctq.... go s~q a~ q- zoBTA ~s ~cxtB uV SBuT~aae To~rum Oq. Ze~e ZxaA Sq Oq gauearMS qua eqnnqe Aau qun ucrjTPG0', ' '((0 GTigv&T) Bgv. 'Q'v. J9- i Ortoo Teq90 0+9, "0 . . (90 UT. 0 uTTjiaos (xvtjx 'Teoj(((aq" oqG. go 090e~m OIrcogoqXqiI 0~, enp Xjgveppxo c...

Duncan, Don Arlen

2012-06-07T23:59:59.000Z

358

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

359

PRIVACY IMPACT ASSESSMENT: INL Education Programs PIA Template  

Broader source: Energy.gov (indexed) [DOE]

Education Education Programs PIA Template Version 3 - May, 2009 Department of Energy Privacy Impact Assessment (PIA) Guidance is provided in the template. See DOE Order 206.1, Department of Energy Privacy Program, Appendix A, Privacy Impact Assessments, for requirements and additional guidance for conducting a PIA: http://www.directives.doe.gov/pdfs/doe/doetextlneword/206/o2061.pdf Please complete electronically: no hand-written submissions will be accepted. This template may not be modified. MODULE I - PRIVACY NEEDS ASSESSMENT Date Departmental Elernent'& (Site 24/Jun/09 Idaho National Laboratory Information Operations and Research Center (IORC) Nameofll,f..,rrnatlon INL Education Programs System or IfPi'()ject Business Enclave Exhibit Proj.ctlUO NA NewPIA D Update [~] DOE PIA - INL Education Program Finallxw.doc N T "tl I Contact Information arne,

360

Microsoft Word - ICEM05_Np_Sorption_paper.doc  

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

5 5 SORPTION OF LONG-LIVED RADIONUCLIDES FROM GEOLOGIC REPOSITORY UNDERGROUND WATERS BY URANIUM OXIDES T.V.Kazakovskaya RFNC-VNIIEF,Russia, V.I. Shapovalov, RFNC-VNIIEF, Russia E.V. Zakharova IPC RAS, Russia S.N.Kalmykov, IPC RAS, Russia M.J.Haire, ORNL, USA 2 Copyright © 2005 by AS ABSTRACT Uranium dioxide (UO 2 ) from unburned nuclear fuel is present in large quantities in spent nuclear fuel geologic repositories. Furthermore, depleted uranium dioxide (DUO 2 ) can be used as a component of the geologic repository waste package as an absorbent for migrating radionuclides.. A potentially important use of DU oxides is to provide an additional engineered chemical barrier in the Yucca Mountain repository. If the DU oxides can be shown to substantially inhibit transport of important actinide elements and fission

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

CX-009241: Categorical Exclusion Determination | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

241: Categorical Exclusion Determination 241: Categorical Exclusion Determination CX-009241: Categorical Exclusion Determination Development of Light Water Reactor Fuels Enhanced Accident Tolerance - Westinghouse Electric Company LLC CX(s) Applied: B3.6 Date: 09/25/2012 Location(s): New Mexico Offices(s): Nuclear Energy The Westinghouse team, which includes General Atomics, Idaho National Laboratory, Massachusetts Institute of Technology, Texas A&M University, Edison Welding Institute, Los Alamos National Laboratory, and Southern Nuclear Operating Company, will work to develop fuel and cladding concepts with strong potential to replace the currently used zirconium uranium oxide (Zr+UO2) fuel system with an enhanced accident tolerant fuel. This will be done by investigating a new fuel system comprised of a cladding capable of

362

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

363

Development of Molten Corium Using An Exothermic Chemical Reaction for the Molten- Fuel Moderator-Interaction Studies at Chalk River Laboratories  

SciTech Connect (OSTI)

Atomic Energy of Canada Limited (AECL) has partnered with Argonne National Laboratory to develop a corium thermite prototypical of Candu material and test the concept of ejecting {approx}25 kg of the molten material from a pressure tube with a driving pressure of 10 MPa. This development program has been completed and the technology transferred to AECL. Preparation for the molten-fuel moderator-interaction tests at AECL's Chalk River Laboratories is well underway. A mixture of 0.582 U/0.077 U{sub 3}O{sub 8}/0.151 Zr/0.19 CrO{sub 3} (wt%) as reactant chemicals has been demonstrated to produce a corium consisting of 0.73 UO{sub 2}/0.11 Zr/0.06 ZrO{sub 2}/0.10 Cr (wt%) at {approx}2400 deg. C. This is comparable to the target Candu specific corium of 0.9 UO{sub 2}/0.1 Zr (wt%), with limited oxidation. The peak melt temperature was confirmed from small-scale thermitic reaction tests. Several small-scale tests were completed to qualify the thermite to ensure operational safety and a quantifiable experimental outcome. The proposed molten-fuel moderator-interaction experiments at Chalk River Laboratories will consist of heating the thermite mixture inside a 1.14-m long insulated pressure tube. Once the molten material has reached the desired temperature of {approx}2400 deg. C, the pressure inside the tube will be raised to about 10 MPa, and the pressure tube will fail at a pre-machined flaw, ejecting the molten material into the surrounding tank of water. The test apparatus, instrumentation, data acquisition and control systems have been assembled, and a series of successful commissioning tests have been completed. (authors)

Nitheanandan, T.; Sanderson, D.B.; Kyle, G. [Chalk River Laboratories, Atomic Energy of Canada Limited, Chalk River, Ontario, K0J 1J0 (Canada); Farmer, M. [Argonne National Laboratory, 9700, S. Cass Avenue, Argonne, IL 60439 (United States)

2004-07-01T23:59:59.000Z

364

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

365

Purification Testing for HEU Blend Program  

SciTech Connect (OSTI)

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

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

1998-06-01T23:59:59.000Z

366

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

367

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

368

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

369

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

370

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

371

Natural fission reactors in the Franceville basin, Gabon: A review of the conditions and results of a {open_quotes}critical event{close_quotes} in a geologic system  

SciTech Connect (OSTI)

Natural nuclear fission reactors are only known in two uranium deposits in the world, the Oklo and Bangombe deposits of the Franceville basin: Gabon. Since 1982, five new reactor zones have been discovered in these deposits and studied since 1989 in a cooperative European program. New geological, mineralogical, and geochemical studies have been carried out in order to understand the behavior of the actinides and fission products which have been stored in a geological environment for more than 2.0 Ga years. The Franceville basin and the uranium deposits remained geologically stable over a long period of time. Therefore, the sites of Oklo and Bangombe are well preserved. For the reactors, two main periods of actinide and radionuclides migration have been observed: during the criticality, under P-T conditions of 300 bars and 400-500{degrees}C, respectively, and during a distention event which affected the Franceville basin 800 to 900 Ma ago and which was responsible for the intrusion of dolerite dikes close to the reactors. New isotopic analyses on uranium dioxides, clays, and phosphates allow us to determine their respective importance for the retention of fission products. The UO{sub 2} matrix appears to be efficient at retaining most actinides and fission products such as REEs, Y, and Zr but not the volatile fission products (Cd, Cs, Xe, and Kr) nor Rb, Sr, and Ba. Some fissiogenic elements such as Mo, Tc, Ru, Rh, Pd, and Te could have formed metallic and oxide inclusion in the UO{sub 2} matrix which are similar to those observed in artificial spent fuel. Clays and phosphate minerals also appear to have played a role in the retention of fissiogenic REEs and also of Pu. 82 refs., 21 figs., 12 tabs.

Gauthier-Lafaye, F. [CNRS, Strasbourg (France)] [CNRS, Strasbourg (France); Holliger, P. [CEA-Cadarache, Saint-Paul-les-Durance (France)] [CEA-Cadarache, Saint-Paul-les-Durance (France); Blanc, P.L. [Institut de Protection et de Surete Nucleaire, Fontenay-aux-Roses (France)] [Institut de Protection et de Surete Nucleaire, Fontenay-aux-Roses (France)

1996-12-01T23:59:59.000Z

372

Multiple-scattering calculations of the uranium {ital L}{sub 3}-edge x-ray-absorption near-edge structure  

SciTech Connect (OSTI)

A theoretical study of the uranium {ital L}{sub 3}-edge x-ray absorption near-edge structure (XANES) is presented for several uranium compounds, including oxides, intermetallics, uranyl fluoride, and {alpha}-uranium. Calculations were performed using FEFF6, an {ital ab} {ital initio} multiple-scattering (MS) code that includes the most important features of current theories. The results, which account for both the fine structure {chi} and the atomiclike background {mu}{sub 0} of the absorption coefficient {mu}, are compared to new and previously measured experimental spectra, reavealing very good agreement for most systems. For several compounds, a more detailed theoretical analysis determined the influence of cluster size and scattering order upon the calculated spectra. Results indicate that MS paths and scattering paths that include rather distant atoms make significant contributions for UO{sub 2}, whereas XANES for crystals with lower symmetry and density can be modeled using only shorter single-scattering paths. In most cases, assumption of a screened final state in the calculation gives better agreement with experiment than use of an unscreened final state. The successful modeling of spectra for a variety of different uranium compounds, with differing spectral features, indicates that the semirelativistic treatment of XANES used here is adequate even for heavy elements. The well-known resonance, observed experimentally for uranyl (UO{sub 2}{sup 2+}) compounds {approx}15 eV above the white line, is successfully modeled here for the first time, using multiple-scattering paths within the O-U-O axial bonds. Overlapping muffin-tin spheres were required in the calculation, probably as a result of the short uranyl axial bonds.

Hudson, E.A. [Glenn T. Seaborg Institute for Transactinium Science, Lawrence Livermore National Laboratory, University of California, Livermore, California 94551 (United States)] [Glenn T. Seaborg Institute for Transactinium Science, Lawrence Livermore National Laboratory, University of California, Livermore, California 94551 (United States); Rehr, J.J. [Department of Physics, University of Washington, Seattle, Washington 98195 (United States)] [Department of Physics, University of Washington, Seattle, Washington 98195 (United States); Bucher, J.J. [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States)] [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States)

1995-11-15T23:59:59.000Z

373

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

374

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

375

High-temperature reactor fuel fission product release and distribution at 1600 to 1800 degrees C  

SciTech Connect (OSTI)

The essential feature of small, modular high-temperature reactors (HTRs) is the inherent limitation in maximum accident temperature to below 1600{degrees} C combined with the ability of coated particle fuel to retain all safety-relevant fission products under these conditions. To demonstrate this ability, spherical fuel elements with modern TRISO particles are irradiated and subjected to heating tests. Even after extended heating times at 1600{degrees} C, fission product release does not exceed the already low values projected for normal operating conditions. In this paper details of fission product distribution within spherical fuel elements heated at constant temperatures of 1600, 1700, and 1800{degrees} C are presented. The measurements confirm the silicon carbide (SiC) coating layer as the most important fission product barrier up to 1800{degrees} C. If the SiC fails (or is defective), the following transport properties at 1600 to 1800{degrees} C can be observed; cesium shows the fastest release from the UO{sub 2} kernel but is highly sorbed in the buffer layer of the particle and in the matrix graphite of the sphere; strontium is retained strongly both in the UO{sub 2} kernels and in matrix graphite, but can penetrate SiC in some cases where cesium is still completely retained; only if all coating layers are breached can iodine and noble gases be released. For the first 100 h at 1600{degrees} C (enveloping all possible accident scenarios of small HTRs), these fission products are almost completely retained in the coated particles.

Schenk, W.; Nabielek, H. (Forschungszentrum Juelich, Postfach 1913, W-5170 Juelich (DE))

1991-12-01T23:59:59.000Z

376

Nuclear fuels technologies fiscal year 1998 research and development test plan  

SciTech Connect (OSTI)

A number of research and development (R and D) activities are planned at Los Alamos National Laboratory (LANL) in FY98 in support of the Department of Energy Office of Fissile Materials Disposition (DOE-MD). During the past few years, the ability to fabricate mixed oxide (MOX) nuclear fuel using surplus-weapons plutonium has been researched, and various experiments have been performed. This research effort will be continued in FY98 to support further development of the technology required for MOX fuel fabrication for reactor-based plutonium disposition. R and D activities for FY98 have been divided into four major areas: (1) feed qualification/supply, (2) fuel fabrication development, (3) analytical methods development, and (4) gallium removal. Feed qualification and supply activities encompass those associated with the production of both PuO{sub 2} and UO{sub 2} feed materials. Fuel fabrication development efforts include studies with a new UO{sub 2} feed material, alternate sources of PuO{sub 2}, and determining the effects of gallium on the sintering process. The intent of analytical methods development is to upgrade and improve several analytical measurement techniques in support of other R and D and test fuel fabrication tasks. Finally, the purpose of the gallium removal system activity is to develop and integrate a gallium removal system into the Pit Disassembly and Conversion Facility (PDCF) design and the Phase 2 Advanced Recovery and Integrated Extraction System (ARIES) demonstration line. These four activities will be coordinated and integrated appropriately so that they benefit the Fissile Materials Disposition Program. This plan describes the activities that will occur in FY98 and presents the schedule and milestones for these activities.

Alberstein, D.; Blair, H.T.; Buksa, J.J. [and others

1998-06-01T23:59:59.000Z

377

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

378

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

379

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

380

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

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381

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

382

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

383

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

SciTech Connect (OSTI)

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

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

2010-03-15T23:59:59.000Z

384

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

385

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

386

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

387

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

388

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

SciTech Connect (OSTI)

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

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

2008-10-15T23:59:59.000Z

389

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

SciTech Connect (OSTI)

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

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

1993-09-15T23:59:59.000Z

390

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

391

O  

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

Voorbeeld C + Voorbeeld C + O 2 1 n + 235 U D ( 2 H) + T ( 3 H) ⇒ CO 2 ⇒ 143 Ba + 91 Kr + 2 1 n ⇒ 4 He+ 1 n Typische Kolen,Olie UO 2 (3% 235 U Deuterium Grondstoffen en Lucht + 97% 238 U) en Lithium Typ. Temperatuur (K) 1000 1000 100 000 000 Energie die vrijkomt per kg brandstof (J/kg) 3,3 x 10 7 2,1 x 10 12 3,4 x 10 14 <------ Grootte:10 -1 m ------> Plasma van sterren die geboren worden Plasma van sterren die geboren worden Zwaartekracht F u s i e Fysica van een Fundamentele Energiebron TWEE BELANGRIJKE FUSIEPROCESSEN REALISATIE VAN DE VOORWAARDEN VOOR FUSIE Opsluitingskwaliteit n i τ(m -3 s) 1970-75 sinds 1990 1975-80 1980-90 Ionentemperatuur (K) 10 21 10 20 10 19 10 18 10 17 10 6 10 7 10 8 10 9 Inertiële fusie Magnetische fusie Reactorregime Reactorregime EXPERIMENTELE RESULTATEN UIT HET FUSIEONDERZOEK Aanzienlijke hoeveelheden

392

Microsoft PowerPoint - IPRC presentation 2012 - R Watson [Compatibility Mode]  

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

Trials Trials of Inert Anodes for the Electrochemical Reduction of Metal Oxide Powders Robert Watson, Arfon Jones & Tim Paget AWE, Aldermaston Robert.Watson@awe.co.uk www.awe.co.uk © British Crown Owned Copyright 2012 /AWE. Published with the permission of the Controller of Her Britannic Majesty's Stationery Office. 2 Objectives  Investigate reduction of metal oxide powders by FFC type process  Investigate non-consumable anodes 3 Reduction of powders  Cavity electrodes - mechanistic studies  UO 2 reduction in LiCl at 923K  Potentiostatic, E ~3.2V  Pt anode, Steel cathode container  Li generation ?  CeO 2 reduction in CaCl 2 -KCl at 1163K  Galvanostatic, E to ~3.5V, Ca generation  C anode, MgO cathode container  Some Ce and CeOCl formed Jeong et al Electrochimica Acta 54 (2009) 6335; Claux et al Electrochimica Acta 56(2011) 2771 4 TiO 2

393

Esempio  

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

Esempio Esempio C + O 2 1 n + 235 U D ( 2 H) + T ( 3 H) ⇒ CO 2 ⇒ 143 Ba + 91 Kr + 2 1 n ⇒ 4 He+ 1 n Materie prime (ingresso Carbone UO 2 (3% 235 U Deuterio alla centrale) e Aria + 97% 238 U) e Litio Temperature tipiche (K) 1000 1000 100 000 000 Energia liberata per kg di combustibile (J/kg) 3,3 x 10 7 2,1 x 10 12 3,4 x 10 14 <------ Dimensioni: 10 -1 m ------> Durata del plasma : 10 -9 - 10 -7 s Plasma di formazione stellare Plasma di formazione stellare Gravità La Fusione Fisica di una fonte energetica fondamentale DUE IMPORTANTI REAZIONI DI FUSIONE CREARE LE CONDIZIONI PER LA FUSIONE Parametro du confinamento n i τ(m -3 s) 1970-75 da 1990 1975-80 1980-90 Temperatura ionica (K) 10 21 10 20 10 19 10 18 10 17 10 6 10 7 10 8 10 9 Fusione inerziale Fusione magnetica Regime del reattore Regime del reattore Massa atomica

394

La Fusión  

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

Fusión Fusión Física de una fuente de energía fundamental La Fusión Física de una fuente de energía fundamental Ejemplos C + O 2 1 n + 235 U D ( 2 H) + T ( 3 H) ⇒ CO 2 ⇒ 143 Ba + 91 Kr + 2 1 n ⇒ 4 He+ 1 n Combustibles Carbón UO 2 (3% 235 U Deuterio (para una central) y Aire + 97% 238 U) y Litio Temperatura típica (K) 1000 1000 100 000 000 Energía generada por kg de combustible (J/kg) 3,3 x 10 7 2,1 x 10 12 3,4 x 10 14 <------ Dimensiones: 10 -1 m ------> Tiempos de vida del plasma: 10 -9 - 10 -7 s Gravedad DOS REACCIONES DE FUSIÓN IMPORTANTES CREANDO LAS CONDICIONES DE LA FUSIÓN Calidad del confinamiento n i τ(m -3 s) 1970-75 desde 1990 1975-80 1980-90 Temperatura iónica (K) 10 21 10 20 10 19 10 18 10 17 10 6 10 7 10 8 10 9 Fusión inercial Fusión magnética Régimen de un reactor Régimen de un reactor

395

I  

Office of Legacy Management (LM)

?am-3 . ,' .*. . - yp: -.* : .- ., ._ ' Yi * <. ? :+". thfa prcbputir. 80,UUU lb. of tmmiuu, J.m,cDu lb. of 3wukdlw crper' tiwu 5.8 t&i8 l atr:irur ral u&d i.Wttd&?# Bir;n8 i;orammant end rUl rid nrtrlcial by uo&utboFlwd putqlm. ). The ~&&a, ' 8m ;altielJ 79 p-rmlt arrgora ted and ttw tap t.ha aikalini~, . L pokotlal brlf)r, bU88M 8-i .ii.i co# sat8 awtaet wltb the mBtmtl8a. aada q*iast fb a8v0-*..u @ow +.ta p-?Y h&al. . .; . ' 6 G.. ..*... . ,,z.. ,. ..*,:: I c,; i ; ' . Total oont of= oc.rLpl.cu~ed Jo); 8 War Ilsp3~-Zz,-i !- . 2.7 -,I -,:,JI:' :' ---- - 2OCCI.3 y- . . i_;: ,.+-a,., ;: s;,!z ; . 5;1 :. ,' ' ;. I &. , I "1 .~ : .. . '. :, :1' 5 .j. j' . ,*,' jF,-,',' , .: . '5 i- t 3 1 . _? ., \ i : ' .

396

Residential Energy Consumption Survey:  

Gasoline and Diesel Fuel Update (EIA)

E/EIA-0262/2 E/EIA-0262/2 Residential Energy Consumption Survey: 1978-1980 Consumption and Expenditures Part II: Regional Data May 1981 U.S. Department of Energy Energy Information Administration Assistant Administrator for Program Development Office of the Consumption Data System Residential and Commercial Data Systems Division -T8-aa * N uojssaooy 'SOS^-m (£03) ao£ 5925 'uofSfAfQ s^onpojj aa^ndmoo - aojAaag T BU T3gN am rcoj? aig^IT^^ '(adBx Q-naugBH) TOO/T8-JQ/30Q 30^703 OQ ' d jo :moaj ajqBfT^A^ 3J^ sjaodaa aAoqe aqa jo 's-TZTOO-eoo-Tgo 'ON ^ois odo 'g^zo-via/aoQ 'TBST Sujpjjng rXaAang uojidmnsuoo XSaaug sSu-ppjprig ON ^oo^s OdO '^/ZOZO-Via/aOQ *086T aunr '6L6I ?sn§ny og aunf ' jo suja^Bd uoj^dmnsuoo :XaAjng uo^^dmnsuoQ XSaaug OS '9$ '6-ieTOO- 00-T90 OdD 'S/ZOZO-Via/aOa C

397

Exemplo  

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

Exemplo Exemplo C + O 2 1 n + 235 U D ( 2 H) + T ( 3 H) ⇒ CO 2 ⇒ 143 Ba + 91 Kr + 2 1 n ⇒ 4 He+ 1 n Matérias primas Carvão UO 2 (3% 235 U Deutério (da central) e Ar + 97% 238 U) e Lítio Temperatura tipica (K) 1000 1000 100 000 000 Energia libertada por kg de combustivel (J/kg) 3,3 x 10 7 2,1 x 10 12 3,4 x 10 14 <------ Dimensões: 10 -1 m ------> Tempo de vida do plasma: 10 -9 - 10 -7 s Gravidade A Fusão Física de uma fonte de energia fundamental DUAS REACÇÕES DE FUSÃO IMPORTANTES OBTENÇÃO DAS CONDIÇÕES DE FUSÃO Qualidade do confinamento n i τ(m -3 s) 1970-75 desde 1990 1975-80 1980-90 Temperatura iónica (K) 10 21 10 20 10 19 10 18 10 17 10 6 10 7 10 8 10 9 Fusão inercial Fusão magnética Regime de um reactor Regime de um reactor Massas nucleares (A massa do electrão é de 0.000549 u.a.) Simbolo Partícula

398

M186  

National Nuclear Security Administration (NNSA)

U U . S . Department of E n m National NucIear Security A d ~ t i o n P.O. Box M50 Oak Ridge, TN 37831 PAGE ! of 3 PAGES Ahf~NDMENT OF SOLICITATION/MODWJCATZON OF CONTaACT --- I - I 8. NAMEAND ADDRESS OF CONTMCn,R (Uo., me#, &my, Zm We) I 9 k AMENDMENT OF SOLKITATION NO. 1. CONTRACT I D CODE A C 2. kMENDMENTMODFICATION NO. MI86 B a h d & W ~ X T a d Y - 1 2 , LLC P.0, Box 2009 MS 8014 Oak Ridge, 'JX 37?33143014 3. EFlBIWE DATE Sct Block l k . Offasmust~~h&edgtroaeiptdthis t prior to Ihc l m m and dak qxa5d in &e wlicitatim as by m e o f t h following mthk (n)ByeompktiagItu~rj8 and 1 5 , d z - wpics ofhe m m h m r ; (b) By achowldging rexipt ofthis amendment on each cqy of tbe offer ~ u b m i w or (c) By separate letter w telegnm which includg a rcfcmm to the solicitdon and mcndumt numbas. FAILURE OF YOUR ACKNOWLEIXMENT TO 8E

399

NOI1VU1SININQV NOLLVINUOdNI A9H3N3 AO^HNH  

Gasoline and Diesel Fuel Update (EIA)

-661 J9QUJ9AON -661 J9QUJ9AON NOI1VU1SININQV NOLLVINUOdNI A9H3N3 AO^HNH 1661 '98902 00 'uoi6mi|SEM 'MS 'anuaAV aouapuadspui QOOl 'IG2-I3 'UOJIBJISIUILUPV UOIJBIUJOJUI 'yoo/jno Xfi/eu3 uuaj_-no^s o; sa6ueiJO ssejppe puas :yaiSVlNlSOd 'saojjjo BUIUBUI iBuomppe pus '8666-99002 OO 'uoiBingsEM JB p|Bd sBeisod ssep-puooag -(soiiou aouBApe >noni!/« a6uBi|o oj joafqns soud) jseA jed 00>L$ J J snas pus 'gggos 00 'uoj6u!MSBM 'MS 'enuoAV eouepuedepui oOOI-'uoiJBJisiuiijupv UOUBUUOJUI ABjaug agj Aq A|jaiJBnb pagsnqnd S] (t?090-Ct'/0 NSSl)^ooWO A6JSU3 UUB±-)JOLJS aqj. 1-661 '9 JOqiiieAON :6u!iuud JQJ p8SB9|9u -j-|AJ '8UIII UJ8JSB8 "LU'd S - 'WE 9 18U-989 (202) ^ '8lUI}UJaiSE8 "Urd g- 'WE 8 3MJ J 0) 2LS-202-LXVd 0088-989(202) 8E2E-E8/ (202) 98902 DO 'uo}6u|L|SBM

400

Release and transport of fission product cesium in the TMI-2 accident  

SciTech Connect (OSTI)

Approximately 50% of the fission product cesium was released from the overheated UO/sub 2/ fuel in the TMI-2 accident. Steam that boiled away from a water pool in the bottom of the reactor vessel transported the released fission products throughout the reactor coolant system (RCS). Some fission products passed directly through a leaking valve with steam and water into the containment structure, but most deposited on dry surfaces inside of the RCS before being dissolved or resuspended when the RCS was refilled with water. A cesium transport model was developed that extended measured cesium in the RCS back to the first day of the accident. The model revealed that approx.62% of the released /sup 137/Cs deposited on dry surfaces inside of the RCS before being slowly leached and transported out of the RCS in leaked or letdown water. The leach rates from the model agreed reasonably well with those measured in the laboratory. The chemical behavior of cesium in the TMI-2 accident agreed with that observed in fission product release tests at Oak Ridge National Laboratory (ORNL).

Lorenz, R.A.; Collins, J.L.

1986-01-01T23:59:59.000Z

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.


401

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

SciTech Connect (OSTI)

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

Michael L. Abbott; Keith A. Daum

2011-08-01T23:59:59.000Z

402

Criticality Benchmark Analysis of Water-Reflected Uranium Oxyfluoride Slabs  

SciTech Connect (OSTI)

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

Margaret A. Marshall; John D. Bess

2009-11-01T23:59:59.000Z

403

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

SciTech Connect (OSTI)

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

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

2014-07-01T23:59:59.000Z

404

High temperature chemistry of advanced heavy water reactor fuel  

Science Journals Connector (OSTI)

The Department of Atomic Energy envisages the use of thoria based fuel in the third phase of nuclear power generation. The fuel will consist of solid solution of thorium-uranium and thorium-plutonium in the form of their oxides. The former will contain 2.5 mole % UO2 while the latter about 4 mole % PuO2. Since no other country in the world has used such fuel, no data is available on its behavior under long-term irradiation. The high temperature chemistry of fuel can however provide some insight into the behavior of such fuel during irradiation and could be of considerable help in the assessment of its long-term integrity. The high temperature chemistry of the fuel essentially involves the measurement of thermodynamic properties of the compounds formed in the multi-component systems comprising the fuel matrix, the fission products and the clad. The physical integrity of the fuel under long-term irradiation can be predicted with the help of basic thermodynamic data such as the Gibbs energy of formation of various compounds and their thermophysical properties such as thermal conductivity and coefficient of thermal expansion derived from experimental measurements. The paper highlights the measurements made on some typical systems relevant to the prediction of thoria based fuel behaviour during long-term irradiation. The experimental problems faced in such measurements are also discussed.

S.R. Dharwadkar

2002-01-01T23:59:59.000Z

405

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

406

A Combined Neutronic-Thermal Hydraulic Model of CERMET NTR Reactor  

SciTech Connect (OSTI)

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

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

2011-02-01T23:59:59.000Z

407

Development of ultrasonic thermometry for high-temperature high-resolution temperature profiling applications in LMFBR safety research  

SciTech Connect (OSTI)

Ultrasonic thermometry has been developed as a high temperature profiling diagnostic for use in the LMFBR Debris Coolability Program at Sandia National Laboratories. These instruments have been used successfully in the dc series experiments and the D10 experiment. Temperatures approaching 3000/sup 0/C with spatial resolution of 10 mm and indicated temperature gradients of 700/sup 0/C/cm have been measured. Instruments have operated in molten sodium, molten steel, and molten UO/sub 2/ environments. Up to 14 measurement zones on a single instrument in molten sodium have been used with 12 mm and 15 mm spatial resolution. Hermetically sealed units operated at elevated temperatures have been used. Posttest examination has revealed very little systematic calibration drifts (<10/sup 0/C) with random drifts occurring with less than 40/sup 0/C standard deviation in a 10 to 12 mm measurement zone. The stability of the system varies from +-1/sup 0/C to +-15/sup 0/C depending on the sensor design constraints for a particular application. Doped tungsten sensors have been developed to permit operation of total measurement zone lengths of 30 cm at temperatures above 2500/sup 0/C. 33 refs., 13 figs.

Field, M.E.

1986-05-01T23:59:59.000Z

408

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content  

SciTech Connect (OSTI)

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N. [Westinghouse Electric Co. LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01T23:59:59.000Z

409

Materials prepared at the Institute for Reference Materials and Measurements (IRMM) for reactor dosimetry applications  

SciTech Connect (OSTI)

Thirteen reference materials prepared under the Euratom Working Group on Reactor Dosimetry (EWGRD) monitor materials program are now available. The determination of trace impurities has recently been completed for rhodium metal (iridium 26.0 {+-} 0.6 mg{center_dot}kg{sup {minus}1}, platinum < 5 mg{center_dot}kg{sup {minus}1}) and these values will be proposed for certification. Certification analyses of titanium metal, which are continuing, indicate scandium < 0.1 mg{center_dot}kg{sup {minus}1}. The preparation of alloys for neutron fluence rate measurements are discussed. Satisfactory homogeneity in aluminum alloys can be difficult to achieve on a milligram scale due to the formation of coarse intermetallic particles, and suitable processing with a high solidification rate, cold working and annealing is essential. The example is given of the alloys Al-15 wt% U and Al-14 wt% Pu, where the inhomogeneity (in terms of relative standard deviation of alloying element) was determined by isotope dilution mass spectrometry as < 0.2% in each case for 20 mg samples. Some recent examples of dosimeter encapsulation are given including NpO{sub 2} and UO{sub 2} microspheres in high purity vanadium and Al-Pu and Al-U fission foils in cold-pressed aluminum containers.

Ingelbrecht, C.D.; Peetermans, F.W.; Palmeri, S.; Robouch, P.B. [Inst. for Reference Materials and Measurements, Geel (Belgium). Sample Preparation Group

1994-12-31T23:59:59.000Z

410

Application of debris-bed dryout data to CONACS-1 modeling  

SciTech Connect (OSTI)

CONACS-1 (the first version of the DOE Containment Analysis Code System) which is currently under development requires extension of existing debris-bed dryout data to the wide range of conditions possible in an accident sequence. While there has been much effort on modeling of debris-bed dryout from first principles, there is no definitive model for beds of irregularly shaped particles of wide size distribution and for beds of varying heights. Existing debris-bed heat-transfer data with real materials are for the most part limited to dryout as a function of bed depth for adiabatic conditions of atmospheric pressure. These data must be extrapolated with the use of appropriate theoretical models based on experimentation with simulant materials to the ocnditions under consideration in the containment analysis. The data base selected for CONACS-1 is from measurements of dryout for beds of 100 to 1000 ..mu..m UO/sub 2/ with the sodium phase Joule heated. This particle size range which is typical for debris from fuel-coolant interactions is the most reasonable to use.

Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.

1982-01-01T23:59:59.000Z

411

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

SciTech Connect (OSTI)

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

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

2006-07-01T23:59:59.000Z

412

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

413

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

414

An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors  

SciTech Connect (OSTI)

An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

Gentry, Cole A [ORNL] [ORNL; Godfrey, Andrew T [ORNL] [ORNL; Terrani, Kurt A [ORNL] [ORNL; Gehin, Jess C [ORNL] [ORNL; Powers, Jeffrey J [ORNL] [ORNL; Maldonado, G Ivan [ORNL] [ORNL

2014-01-01T23:59:59.000Z

415

Coyote-prey interactions on an intensively managed south Texas ranch  

E-Print Network [OSTI]

saoua1oS saVzaqs33 pue agVIpIVN :qoaOqnS uoCeg 8$6L 3DN3IDS HO H3ISVN go aauSap aqg cog squamaaynbas aqq go quamIIygIng IeVqzed ug XqrszawVuO gpV sexag go aSaIIoD aqenpea5 aqq oq paqqVmqnS Z3Ha Z. ZOOS XHV5 sTsaqJ, HDNVH SVX31 HLllOS G35VNVW... ?13hISN3LNI NV NO SNOIIDVH3INI 7. 3HH-3IOJOD ()()6i gsnNnV (quamqzedag go peag) EIPImqoS ( Pgneg nqS . st Su Hoe(' quez5 3 meTIIIM (aaq ImmoO go uleqO) a eg 'H Ialued :Eq quaquoo pue aIXqs oq se panouddV M3HQ 110OS LHV5 s'tsaq1 NONVH SVX3J...

Drew, Gary Scott

2012-06-07T23:59:59.000Z

416

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

SciTech Connect (OSTI)

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

Shannon Bragg-Sitton

2014-02-01T23:59:59.000Z

417

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

418

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

419

Nuclear fuel corrosion over millennia interpreted using geologic data  

SciTech Connect (OSTI)

Corrosion of nuclear fuel over the 10,000 year regulatory period in a geologic repository will be a function of physical characteristics (e.g., crystallinity, crystal sizes, crystal forms) and chemical characteristics (e.g., crystal composition, compositional variability, accessory phases). Natural uraninite (nominally UO{sub 2+x}) which has undergone long-term corrosion can be studied to infer the long-term behavior of nuclear fuel. Previously, uraninite from the Nopal I deposit, Pena Blanca district, Chihuahua, Mexico, has been shown to constitute an outstanding analog material for comparison with nuclear fuel. Similarities between Nopal I uraninite and nuclear fuel have been shown to include bulk composition, general crystal structure, and total trace element content. Data presented here suggest that, as a bulk material, Nopal I uraninite compares favorably with irradiated nuclear fuel. Nevertheless, some fine-scale differences are noted between Nopal I uraninite and irradiated nuclear fuel with respect to both internal structures and compositions. These observations suggest that whereas the long-term responses of the two materials to oxidative alteration in a geologic repository may be similar, the detailed mechanisms of initial oxidant penetration and the short-term oxidative alternation of Nopal I uraninite and irradiated nuclear fuel are likely to be different.

Pearcy, E.C.; Manaktala, H.K. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

1994-12-31T23:59:59.000Z

420

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

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

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

422

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 m radius (calculated) sediment pores were drained, followed by draining pores to 15 m radius. U release in the first two weeks was similar between sediments and pore sizes with a range of 4.4 to 5.6 M U in the 14 day sample. The 7.0 m bgs sediment U declined in the larger pores to 0.22 M at day 84, whereas the small pores released U to 6.7 M at day 84. The 7.6 m bgs sediment released 1.4 M 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

423

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

424

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

425

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

426

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

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

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

429

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

430

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

431

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 Mssbauer 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

432

Use of ion conductors in the pyrochemical reduction of oxides  

DOE Patents [OSTI]

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

Miller, W.E.; Tomczuk, Z.

1994-02-01T23:59:59.000Z

433

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

434

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

SciTech Connect (OSTI)

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

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

2014-04-01T23:59:59.000Z

435

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

SciTech Connect (OSTI)

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

Jason Hales; Various

2014-06-01T23:59:59.000Z

436

Conversion of depleted uranium hexafluoride to a solid uranium compound  

DOE Patents [OSTI]

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

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

2001-01-01T23:59:59.000Z

437

IMGA [Irradiated Microsphere Gamma Analyzer] examination of the Set No. 4 fuel under project work statement FD-20  

SciTech Connect (OSTI)

Results of an examination of over 10,800 unbonded fuel particles from three irradiated spherical fuel elements by the Irradiated Microsphere Gamma Analyzer system are reported. The investigation was initiated to assess fission product behavior in LEU UO{sub 2} TRISO-coated fuel particles at elevated temperatures. Of the three spheres considered, one was reserved as a control and the other two were subjected to simulated accident-condition temperatures of 1600{degree}C and 1800{degree}C, respectively. For the control sphere and the sphere tested at 1600{degree}C, no statistical evidence of fission product release (cesium) from individual particles was observed. At fuel temperatures of 1800{degree}C, however, fission product release (cesium) from individual particles was significant and there was large particles-to-particle variation. At 1800{degree}C, individual particle release (cesium) was on average ten times the Kernforschungsanlage-measured integral spherical fuel element release value. Particle release data from the sphere tested at 1800{degree}C indicate that there may be two distinct modes of failure at fuel temperatures of 1800{degree}C and above. 5 refs., 9 figs., 9 tabs.

Baldwin, C.A.; Kania, M.J.

1990-03-01T23:59:59.000Z

438

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

SciTech Connect (OSTI)

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

B. Boer; A. M. Ougouag

2010-05-01T23:59:59.000Z

439

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

SciTech Connect (OSTI)

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

Xu Wu; Piyush Sabharwall; Jason Hales

2014-07-01T23:59:59.000Z

440

Westinghouse Hanford Company FY 1995 Materials Management Plan (MMP)  

SciTech Connect (OSTI)

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

Higginson, M.C.

1994-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "uo uo uo" from the National Library of EnergyBeta (NLEBeta).
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We encourage you to perform a real-time search of NLEBeta
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441

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

442

Pre- and postirradiation evaluation of fuel in capsule HRB-14. [HTGR  

SciTech Connect (OSTI)

Capsule HRB-14 was irradiated jointly by Oak Ridge National Laboratory and General Atomic Company (GA). This report covers the pre- and postirradiation characterization and evaluation of the GA fuel. The experiment was primarily to characterize the irradiation performance of TRISO-coated low-enriched (Th,U)O/sub 2/, UC/sub 0/ /sub 7/O/sub 0/ /sub 5/, and UO/sub 2/ particles and TRISO ThO/sub 2/ particles. Twenty cured-in-place fuel rods containing the fuel particles were irradiated at 960/sup 0/ to 1130/sup 0/C. The fluence was 4.5 to 8.3 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/. The fertile burnup was 4.1% to 8.5% fissions per initial metal atom (FIMA); the mixed oxide burnup was 14.5% to 19.3% FIMA; the uranium fissile burnup was 27.9% to 29.5% FIMA.

Young, C.A.

1980-09-01T23:59:59.000Z

443

Estimates of Zircaloy integrity during dry storage of spent nuclear fuel: Final report  

SciTech Connect (OSTI)

The analytical and experimental work described in this report is intended to predict the integrity of light-water reactor (LWR) fuel rods when the fuel rods are stored dry. The analytical portion considered all failure mechanisms that could be expected to operate under dry storage conditions, including creep rupture, external oxidation stress-corrosion cracking (SCC), fatigue, and clad splitting by UO/sub 2/ oxidation. Existing physically based models were used to predict the probability that LWR fuel rod cladding will fail in 100 years, as a function of the temperature at which the rods are stored. In the experimental portion, SCC tests were conducted on irradiated Zircaloy cladding to determine characteristics under conditions relevant to dry storage. ''Precracked'' and ''smooth'' (with only small naturally occurring flaws) specimens of irradiated cladding were subjected to ''split ring'' tests at initial stresses on the order of the yield stress in a variety of atmospheres containing iodine or cesium/cadmium. Most precracked specimens failed by SCC, and about one-third of smooth specimens irradiated to fluence above 2.5 /times/ 10/sup 24/ n/m/sup 2/ also failed. However, the stresses present in these tests were much higher than those expected in stored fuel cladding; therefore, the experimental results do not necessarily indicate likely SCC problems in dry-storage fuel. 68 refs., 54 figs., 35 tabs.

Miller, A.K.; Brooks, M.; Cheung, T.Y.; Tasooji, A.; Wood, J.C.; Kelm, J.R.; Surette, B.A.; Frost, C.R.

1989-05-01T23:59:59.000Z

444

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

SciTech Connect (OSTI)

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

Ono, M.; Wada, K.; Kitada, T. [Osaka Univ., 2-1, yamadaoka, Suita-shi, Osaka, 565-0871 (Japan)

2012-07-01T23:59:59.000Z

445

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

SciTech Connect (OSTI)

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

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

2014-10-01T23:59:59.000Z

446

Coordination structures of lanthanide(III) and uranyl(VI) nitrato complexes with N,N?-dimethyl-N,N?-dibutylmalonamide. Part II  

Science Journals Connector (OSTI)

The coordination structures of lanthanide(III) and uranyl(VI) nitrato complexes with N,N?-dimethyl-N,N?-dibutylmalonamide (DMDBMA) were investigated in terms of 1H, 13C and 14N NMR measurements, infrared spectrum, molar conductivity measurement, absorption spectrum, fluorescence spectrum, magnetic susceptibility measurement and thermal analysis. The chemical formulae of the isolated lanthanide(III) and uranyl(VI) nitrato-DMDBMA complexes are Ln(NO3)3 2DMDBMA and UO2(NO3)2 DMDBMA, respectively. The DMDBMA coordinates to the lanthanide(III) and uranyl(VI) ions, with the oxygen atoms of carbonyl group and nitrate ions coordinate to the central metal ions in a bidentate manner. For the lanthanide(III) nitrato complexes, the coordination number is ten, and a change of the coordination structure occurs between the lighter and heavier complexes. The uranyl(VI) nitrato-DMDBMA complex has an eight-coordinated structure with the uranyl group surrounded by six oxygen atoms, four from the bidentate nitrate groups and two from the bidentate DMDBMA, lying in the plane perpendicular to the axial uranyl group.

Takashi Nakamura; Chie Miyake

1996-01-01T23:59:59.000Z

447

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

448

Novel Accident-Tolerant Fuel Meat and Cladding  

SciTech Connect (OSTI)

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

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

2013-09-01T23:59:59.000Z

449