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Science Conference Proceedings (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
Spectroscopic Studies of the Several Isomers of UO3
Science Conference Proceedings (OSTI)
Uranium trioxide is known to adopt seven different structural forms. While these structural forms have been well characterized using x-ray or neutron diffraction techniques, little work has been done to characterize their spectroscopic properties, particularly of the pure phases. Since the structural isomers of UO3 all have similar thermodynamic stabilities and most tend to hydrolyze under open atmospheric conditions, mixtures of UO3 phases and the hydrolysis products are common. Much effort went into isolating pure phases of UO3. Utilizing x-ray diffraction as a sample identification check, UV/Vis/NIR spectroscopic signatures of ?-UO3, ?-UO3, ?-UO3 and UO2(OH)2 products were obtained. The spectra of the pure phases can now be used to characterize typical samples of UO3, which are often mixtures of isomers.
Sweet, Lucas E.; Reilly, Dallas D.; Abrecht, David G.; Buck, Edgar C.; Meier, David E.; Su, Yin-Fong; Brauer, Carolyn S.; Schwantes, Jon M.; Tonkyn, Russell G.; Szecsody, James E.; Blake, Thomas A.; Johnson, Timothy J.
The density and purity of fused UO/sub 2/ from several suppliers was evaluated. Densities of large pieces varied widely, but variations in pycnometer and bulk densities of crushed UO/sub 2/ powder were small. Average oxygen- touranium ratios ranged from 1.94 to 2.14. Impurities visible as microscopic inclusions were U/sub 4/O/sub 9/, metallic uranium, UC, and UN/sub 2/. The chief trace metallic contaminants were aluminum, iron, and silicon. Hydrogen, nitrogen, carbon monoxide, and carbon dioxide were evolved during vacuum extraction. (auth)
Cole, G.R.
Damage Structure Evolution in Ion Irradiated UO2
Science Conference Proceedings (OSTI)
Symposium, Radiation Effects in Oxide Ceramics and Novel LWR Fuels ... To better understand low dose irradiation effects on defect creation in UO2, helium�...
Effect of Additives on Diffusion Processes in UO2
Science Conference Proceedings (OSTI)
The oxygen and uranium Frenkel pairs and the uranium-oxygen Schottky defects regulate the O/U ratio, which in turn influence diffusion processes in UO2.
PREPARATION OF HIGH DENSITY UO$sub 2$
A method is presented for the preparation of highdensity UO/sub 2/ from UF/sub 6/. In accordance with the invention, UF/sub 6/ is reacted with water and concentrated ammonium hydroxide is added to the resulting aqueous solution of UO/ sub 2/F/sub 2/. The resulting precipitate is calcined to U/sub 3/O/sub 8/ an d the U/sub 3/O/sub 8/ is reduced to UO/sub 2/ with a gaseous mixture comprised of carbon monoxide and carbon dioxide at a temperature of from 1600 to 1900 deg C.
Googin, J.M.
Simple but Stronger UO, Double but Weaker UNMe Bonds: The Tale Told by Cp2UO and Cp2UNR
Science Conference Proceedings (OSTI)
The free energies of reaction and the activation energies are calculated, with DFT (B3PW91) and small RECP (relativistic core potential) for uranium, for the reaction of Cp2UNMe and Cp2UO with MeCCMe and H3Si-Cl that yields the corresponding addition products. CAS(2,7) and DFT calculations on Cp2UO and Cp2UNMe give similar results, which validates the use of DFT calculations in these cases. The calculated results mirror the experimental reaction of [1,2,4-(CMe3)3C5H2]2UNMe with dimethylacetylene and [1,2,4-(CMe3)3C5H2]2UO with Me3SiCl. The net reactions are controlled by the change in free energy between the products and reactants, not by the activation energies, and therefore by the nature of the UO and UNMe bonds in the initial and final states. A NBO analysis indicates that the U-O interaction in Cp2UO is composed of a single U-O bond with three lone pairs of electrons localized on oxygen, leading to a polarized U-O fragment. In contrast, the U-NMe interaction in Cp2UNMe is composed of a and component and a lone pairof electrons localized on the nitrogen, resulting in a less polarized UNMe fragment, in accord with the lower electronegativity of NMe relative to O. The strongly polarized U(+)-O(-) bond is calculated to be about 70 kcal mol-1 stronger than the less polarized U=NMe bond.
LPCNO, CNRS-UPS-INSA, INSA Toulouse; Institut Charles Gerhardt, CNRS, Universite Montpellier; Laboratoire de Chimie et Physique Quantiques, CNRS, IRSAMC, Universite Paul Sabatier; Andersen, Richard; Barros, Noemi; Maynau, Daniel; Maron, Laurent; Eisenstein, Odile; Zi, Guofu; Andersen, Richard
Modifying Ceramic Fuel Pellets to Improve UO2 Properties
Science Conference Proceedings (OSTI)
... UO2 fuel will provide manufacturers with tools to optimize fuel performance. ... Electronic Structure Calculations of Structure and Chemistry of the Y2O3/Fe Interface ... Impacts of Hydrogen in Unirradiated Zircaloy Nuclear Cladding under Dry�...
PUREX/UO{sub 3} deactivation project management plan
Science Conference Proceedings (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.
Sorption of 237Np by UO2 under Repository Conditions
NLE Websites -- All DOE Office Websites (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: email@example.com e-mail: firstname.lastname@example.com e-mail: firstname.lastname@example.com 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
Description of results from single pass flowthrough tests showing the effect of dissolved oxygen and temperature on the dissolution of pure UO2 and UO2 with 8 wt% Gd2O3 doping.
Casella, Amanda J.; Hanson, Brady D.; Miller, William H.
Density Functional Theory Calculations of Mass Transport in UO2
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
Migration Mechanisms of Oxygen Interstitial Clusters in UO2
Science Conference Proceedings (OSTI)
Understanding the migration kinetics of radiation-induced point defects and defect clusters is a key to predicting the microstructural evolution and mass transport in nuclear fuels. Although the diffusion kinetics of point defects in UO2 is well explored both experimentally and theoretically, the kinetics of defect clusters is not well understood. In this work the migration mechanisms of oxygen interstitial clusters of size one to five atoms (1Oi � 5Oi) in UO2 are investigated by temperature-accelerated dynamics simulations without any a priori assumptions of migration mechanisms. It is found that the migration paths of oxygen interstitial clusters are complex and non-intuitive and that multiple migration paths and barriers exist for some clusters. It is also found that the cluster migration barrier does not increase with increasing cluster size and its magnitude has the following order: 2Oi < 3Oi < 1Oi < 5Oi < 4Oi. Possible finite-size effects are checked with three different sized systems. The results show good agreement with other available experimental and theoretical data. In particular, the relatively large migration barriers of cuboctahedral clusters (4Oi and 5Oi) are in good agreement with the experimentally measured oxygen diffusion activation energy in U4O9, which is thought to contain many such clusters. The cluster migration sequence may explain the interesting relationship between the oxygen diffusivity and stoichiometry in UO2+x.
Xian-Ming Bai; Anter El-Azab; Jianguo Yu; Todd R. Allen
Multi-Scale Modeling of Fission Gas Evolution in UO2
Science Conference Proceedings (OSTI)
Fission gases in uranium dioxide (UO2) nuclear fuels, of which Xe is one of the most prominent, influence fuel performance during reactor operation and have�...
UO 2 fission gas release rates from atomistic calculations of intrinsic ...
Science Conference Proceedings (OSTI)
Based on DFT and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of �...
Doping d-UO 2 fuel pellets for improved hardness and fracture ...
Science Conference Proceedings (OSTI)
This study investigates the effects of these oxide dopants on the fracture toughness of depleted uranium oxide (d-UO2) at temperatures between�...
PROCESS FOR THE PRODUCTION OF AN ACTIVATED FORM OF UO$sub 2$
A process for producing a highly active form of UO/sub 2/ characterized both by rapid oxidation in air and by rapid chlorination with CCl/sub 4/ vapor at an elevated temperature is reported. In accordance with the process, commercial UO/sub 2/, is subjected to a series of oxidation-reduction operations to produce a form of UC/sub 2/ of enhanced reactivity. By treatimg commercial UO/sub 2/ at a temperature between 335 and 485 deg C with methane, then briefly with an oxygen containing gas and followimg this by a second treatment with a methane containing gas, the original relatively stable charge of UO/sub 2/ will be transformed into an active form of UO/sub 2/.
Polissar, M.J.
Effect of Edge Dislocation on Thermal Transport in UO2
Molecular-dynamics simulations are used to characterize the effects of dislocations on the thermal transport properties of UO{sub 2}. Microstructures with various dislocation densities of the order of 10{sup 16} m{sup ?2} are simulated at temperatures between 800 and 1600 K. The effects of dislocations on the thermal-transport properties are found to be independent on temperature, consistent with the classic Klemens�Callaway analysis. The effect of dislocation density is also quantified. The simulation results are also fit to the pertinent part of the empirical formula for the thermal conductivity used in the FRAPCON fuel-performance code, which gives the overall effects of temperature and dislocation effects on thermal conductivity. The fitted results can be well-described within this formalism, indicating that the results of molecular-dynamics simulations can be used as a reliable source of parameters for models at longer length scales.
Deng, B; Chernatynskiy, Aleksandr; Shukla, P; Sinnott, Susan B; Phillpot, Simon R.
PUREX/UO{sub 3} facilities deactivation lessons learned history
The Plutonium-Uranium Extraction (PUREX) Facility operated from 1956-1972, from 1983-1988, and briefly during 1989-1990 to produce for national defense at the Hanford Site in Washington State. The Uranium Trioxide (UO{sub 3}) Facility operated at the Hanford Site from 1952-1972, 1984-1988, and briefly in 1993. Both plants were ordered to permanent shutdown by the U.S. Department of Energy (DOE) in December 1992, thus initiating their deactivation phase. Deactivation is that portion of a facility`s life cycle that occurs between operations and final decontamination and decommissioning (D&D). This document details the history of events, and the lessons learned, from the time of the PUREX Stabilization Campaign in 1989-1990, through the end of the first full fiscal year (FY) of the deactivation project (September 30, 1994).
Hamrick, D.G.; Gerber, M.S.
Preparation of UO/sub 2/ fragments for fuel-debris experiments
A unique process was developed for preparing multi-kilogram quantities of > 90% dense fragments of enriched and depleted UO/sub 2/ sized 20 mm to 0.038 mm for fuel debris experiments. Precipitates of UO/sub 4/ . xH/sub 2/O were treated to obtain UO/sub 2/ powders that would yield large cohesive green pieces when isostatically pressed to 206 MPa. The pressed pieces were crushed into fragments that were about 30% oversized, and heated to 1800/sup 0/C for 24 h in H/sub 2/. Oversizing compensates for shrinkage during densification. Effort was dramatically reduced by working on a large scale and by presizing the green UO/sub 2/ instead of directly crushing densified pellets.
Tinkle, M.C.; Kircher, J.A.; Zinn, R.M.; Eash, D.T.
Benchmarking of Graphite Reflected Critical Assemblies of UO2
Science Conference Proceedings (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
Molecular dynamics simulation of UO2 nanocrystals melting
In this article we study melting of uranium dioxide (UO2) nanocrystals (NC) isolated in vacuum (i.e. non-periodic boundary conditions) using molecular dynamics (MD) in the approximation of pair potentials and rigid ions. We calculate the size dependence of the temperature and heat of melting, the density jump for crystals of cubic shape and volumes up to 1000 nm^3 (50000 particles). Linear and parabolic extrapolations of these dependences to macroscopic (infinite) size are considered, the parabolic is found to be better suited for the analysis of data on the temperature and the heat of melting. The closest to the modern experimental data estimates of the melting temperature of macrocrystals are obtained using the interaction potentials Goel-08 (2969K), Yakub-09 (3105K) and MOX-07 (3291K). The density jump at melting is well reproduced by Yakub-09 (8.66%) and MOX-07 (7.97%). The heat of fusion for all sets of the potentials considered is found to be underestimated by 50-75%, possibly because of the excluded he...
Boyarchenkov, A S; Nekrasov, K A; Kupryazhkin, A Ya
Thermal diffusivity and thermal conductivity of sintered UO2 and UO2-Gd2O3. Technical report
The thermal diffusivity was measured using the laser flash method on sintered uranium dioxide (O/U=2.003, density=10.48X10 kg/m, from 300 to 2773 K), and urania and gadolinia mixed fuel (2,4 and 6 Wt% Gd2O3 content, from 600 to 1850 K). An equation was suggested for near-stoichiometric uranium dioxide over the temperature range 500-3100 K: K=(1-aP)(1/(A+BT)+DTxexp(-E/kT)x(1+H(E/kT+2)(sup 2))), where K in W/(m)(K), P is the fraction of porosity, a=2.74-5.8X10(sup 4-)T, A=3.68X10(sup 2-)(m)(K)/W, B=2.25X10(sup 4-)m/W, D=5.31X10(sup 3-)W/mXK2, H=0.264, E=1.15 ev, k is the Boltzmann constant. The thermal conductivity of UO2-Gd2O3 samples as a function of temperature and Gd2O3 content, X, could be expressed by phonon conduction; K=1/(A+BT) in the temperature range from 600 to 1700 K, where A=2.50 X+0.044(m)(K)/W.
Ying, S.; Ji, Z.
A nuclear fuel element consisting of metal coated UO/sub 2/ particles dispersed in a matrix of Zircalloy and having a cladding of Zircalloy is presented. (AEC)
Cain, F.M. Jr.; Eck, J.E.
PUREX/UO3 Facilities deactivation lessons learned history
Science Conference Proceedings (OSTI)
Disconnecting the criticality alarm permanently in June 1996 signified that the hazards in the PUREX (plutonium-uranium extraction) plant had been so removed and reduced that criticality was no longer a credible event. Turning off the PUREX criticality alarm also marked a salient point in a historic deactivation project, 1 year before its anticipated conclusion. The PUREX/UO3 Deactivation Project began in October 1993 as a 5-year, $222.5- million project. As a result of innovations implemented during 1994 and 1995, the project schedule was shortened by over a year, with concomitant savings. In 1994, the innovations included arranging to send contaminated nitric acid from the PUREX Plant to British Nuclear Fuels, Limited (BNFL) for reuse and sending metal solutions containing plutonium and uranium from PUREX to the Hanford Site tank farms. These two steps saved the project $36.9- million. In 1995, reductions in overhead rate, work scope, and budget, along with curtailed capital equipment expenditures, reduced the cost another $25.6 million. These savings were achieved by using activity-based cost estimating and applying technical schedule enhancements. In 1996, a series of changes brought about under the general concept of ``reengineering`` reduced the cost approximately another $15 million, and moved the completion date to May 1997. With the total savings projected at about $75 million, or 33.7 percent of the originally projected cost, understanding how the changes came about, what decisions were made, and why they were made becomes important. At the same time sweeping changes in the cultural of the Hanford Site were taking place. These changes included shifting employee relations and work structures, introducing new philosophies and methods in maintaining safety and complying with regulations, using electronic technology to manage information, and, adopting new methods and bases for evaluating progress. Because these changes helped generate cost savings and were accompanied by and were an integral part of sweeping ``culture changes,`` the story of the lessons learned during the PUREX Deactivation Project are worth recounting. Foremost among the lessons is recognizing the benefits of ``right to left`` project planning. A deactivation project must start by identifying its end points, then make every task, budget, and organizational decision based on reaching those end points. Along with this key lesson is the knowledge that project planning and scheduling should be tied directly to costing, and the project status should be checked often (more often than needed to meet mandated reporting requirements) to reflect real-time work. People working on a successful project should never be guessing about its schedule or living with a paper schedule that does not represent the actual state of work. Other salient lessons were learned in the PUREX/UO3 Deactivation Project that support these guiding principles. They include recognizing the value of independent review, teamwork, and reengineering concepts; the need and value of cooperation between the DOE, its contractors, regulators, and stakeholders; and the essential nature of early and ongoing communication. Managing a successful project also requires being willing to take a fresh look at safety requirements and to apply them in a streamlined and sensible manner to deactivating facilities; draw on the enormous value of resident knowledge acquired by people over years and sometimes decades of working in old plants; and recognize the value of bringing in outside expertise for certain specialized tasks.This approach makes possible discovering the savings that can come when many creative options are pursued persistently and the wisdom of leaving some decisions to the future. The essential job of a deactivation project is to place a facility in a safe, stable, low-maintenance mode, for an interim period. Specific end points are identified to recognize and document this state. Keeping the limited objectives of the project in mind can guide decisions that reduce risks with minimal manipul
Gerber, M.S.
THE EMANATION OF RADON 220 FROM SINTERED UO$sub 2$ POWDERS AND PLATES
The emanation of thoron (Rn/sup 220/) from sintered UO/sub 2/ powders and plates was measured as a function of temperature. The uranium oxide samples were indexed with radiothorium by coprecipitation and coevaporation techniques. The emanation measurements were performed in a flow system, using an alpha scintillation detector and a helium-hydrogen carrier gas mixture. Both the radiothorium concentration (5 to 100 mu C/g UO/sub 2/) and the uranium oxide density (71 to 99 percent TD) were varied. The surface areas and densities of the UO/sub 2/ plates were measured by krypton gas adsorption and liquid immersion techniques, respectively. Assuming a diffusion mechanism, diffusion coefficients for thoron in sintered UO/sub 2/ were calculated as a function of temperature. The data were represented by an equation of the form, D = D/sub O/ exp (-Q/RT). An apparent increase in both activation energy (Q) and D/sub O/ with density was observed for the 1100 to 1450 deg C temperature range. For some samples the thoron emanating power could be measured at temperatures as low as 400 deg C. Log D versus 1/T plots from 400 to 1450 deg C gave intersecttng straight lines with different activation energies. One intersection occurred near the Tammann temperature. The emanation of Rn/sup 220/ from UO/sub 2/ apparently involves several types of diffusion processes. (auth)
Clayton, J.C.; Aronson, S.
PREPARATION OF UO$sub 2$ FOR NUCLEAR REACTOR FUEL PELLETS
A method is given for preparing high-density UO/sub 2/ compacts. An aqueous uranyl fluoride solution is contacted with an aqueous ammonium hydroxide solution at an ammonium to-uranium ratio of 25: 1 to 30:1 to form a precipitate. The precipitate is separated from the- mother liquor, dried, and contacted with steam at a uniform temperature within the range of 400 to 650 deg C to produce U/ sub 3/O/sub 8/. The U/sub 3/O/sub 8/ is red uced to UO/sub 2/ with hydrogen at a uniform temperature within the range of 550 to 600 deg C. The UO/sub 2/ is then compressed into compacts and sintered. High-density compacts are fabricated to close tolerances without use of a binder and without machining or grinding. (AEC)
Googin, J.M.
Oxygen diffusion in UO2+x and (U,Pu)O2+-x
In the first part of this report we revisit an earlier study of oxygen diffusion in UO{sub 2+x}, in which we used density functional theory (DFT) calculations to parameterize a kinetic Monte Carlo (kMC) model. The results from these earlier kMC simulations are reproduced in Fig. 1 and they indicate fairly good agreement with available experiments. This work was later expanded to include a larger temperature range. However, since the publication of this study there have been a number of advancements in DFT methodology for UO{sub 2} and UO{sub 2+x} providing increased accuracy. We have also gained better understanding of the oxygen clustering phenomena occurring in UO{sub 2+x}. For these two reasons, the DFT calculations of the migration barriers of single oxygen interstitials and di-interstitial clusters have been repeated using the LDA+U and GGA+U methodologies. The earlier study used regular GGA and, even though this method captures similar trends as the more advanced LDA+U and GGA+U techniques, it does not fulfill the quantitative requirements set by some applications. Additionally, we have identified a mechanism for the most stable quad-interstitial clusters to migrate and here we calculate the corresponding barriers within both the LDA+U and GGA+U methodologies. The new LDA+U and GGA+U data sets are analyzed in terms of available experiments. In the second part of this report we present initial results for the impact of Pu on oxygen diffusion in UO{sub 2}. The first step in understanding this process is to calculate the binding energies of oxygen vacancies and interstitials to a Pu ion in the UO{sub 2} matrix. Possible diffusion mechanisms are discussed for (U,Pu)O{sub 2-x}, (U,Pu)O{sub 2} and (U,Pu)O{sub 2+x}.
Andersson, Anders D. [Los Alamos National Laboratory; Liu, Xiang-Yang [Los Alamos National Laboratory
Thermal conductivity is one of the most important metrics of nuclear fuel performance. Therefore, it is crucial to understand the impact of microstructure features on thermal conductivity, especially since the microstructure evolves with burn-up or time in the reactor. For example, UO{sub 2} fuels are polycrystalline and for high-burnup fuels the outer parts of the pellet experience grain sub-division leading to a very fine grain structure. This is known to impact important physical properties such as thermal conductivity as fission gas release. In a previous study, we calculated the effect of different types of {Sigma}5 grain boundaries on UO{sub 2} thermal conductivity and predicted the corresponding Kapitza resistances, i.e. the resistance of the grain boundary in relation to the bulk thermal resistance. There have been reports of pseudoanisotropic effects for the thermal conductivity in cubic polycrystalline materials, as obtained from molecular dynamics simulations, which means that the conductivity appears to be a function of the crystallographic direction of the temperature gradient. However, materials with cubic symmetry should have isotropic thermal conductivity. For this reason it is necessary to determine the cause of this apparent anisotropy and in this report we investigate this effect in context of our earlier simulations of UO{sub 2} Kapitza resistances. Another source of thermal resistance comes from fission products and fission gases. Xe is the main fission gas and when generated in sufficient quantity it dissolves from the lattice and forms gas bubbles inside the crystalline structure. We have performed studies of how Xe atoms dissolved in the UO{sub 2} matrix or precipitated as bubbles impact thermal conductivity, both in bulk UO{sub 2} and in the presence of grain boundaries.
Du, Shiyu [Los Alamos National Laboratory; Andersson, Anders D. [Los Alamos National Laboratory; Germann, Timothy C. [Los Alamos National Laboratory; Stanek, Christopher R. [Los Alamos National Laboratory
Acoustic emission from thermal-gradient cracks in UO$sub 2$
A feasibility study has been conducted to evaluate the potential use of acoustic emission to monitor thermal-shock damage in direct electrical heating of UO$sub 2$ pellets. In the apparatus used for the present tests, two acoustic- emission sensors were placed on extensions of the upper and lower electrical feedthroughs. Commercially available equipment was used to accumulate acoustic- emission data. The accumulation of events displayed on a cathode-ray-tube screen indicates the total number of acoustic-emission events at a particular location within the pellet stack. These tests have indicated that acoustic emission can be used to monitor thermal-shock damage in UO$sub 2$ pellets subjected to direct- electrical heating. 8 references. (auth)
Kennedy, C.R.; Kupperman, D.S.; Wrona, B.J.
MECHANICAL PROPERTIES OF STAINLESS STEEL-UO$sub 2$ DISPERSION FUEL ELEMENTS
Mechanical properties of stainless steel-- UO/sub 2/ dispersion fuel elements were determined on specimens fabricated (a) by the cold-binder extrusion and hot swaging technique, (b) by the single hot-coextrusion method, and (c) by the hotcoextrusion method followed by a second hot-extrusion, hot-rolling, swaging, or drawing. Tensile test results show that cold-binder material has very good tensile properties with the exception of ductility. Bend tests show thai coarse oxide material has better ductility than the fine oxide both before and after irradiation. Although the fuel element material is low in ductility, test results indicate that completed fuel elements composed of a dispersion of UO/ sub 2/ clad in stainless steel have fair mechanical properties for reactor use even after high burnups. (auth)
Valovage, W.D.; Siergiej, R.A.
DEVELOPMENT OF FERRITIC STAINLESS STEEL-UO$sub 2$ DISPERSION FUEL ELEMENTS
A preliminary definition of a fabrication process for ferritic stainless steel-UO/sub 2/ fuel clad with Type 430 stainiess steel has been achieved. The procedure is an adaptation of an earlier process for preparing austenitic fuel elements, and consists of the cold-binder extrusion of a plastic powder dispersion which is fired in dry hydrogen, fitted with welded end caps, and electroplated with nickel. This core is inserted into a length of type 430 stainless steel cladding and the assembly is bonded by hot-swaging at 800 C. The irradiation performance of such a fuel element was examined by means of an MTR test. The results of this test strongly indicated that the desired lowering of UO/sub 2/ particle operating temperature was realized and that ferritic stainless steel is not unduly susceptible to irradiation dannage. (auth)
Barney, W.K.; Ray, W.E.; Sowman, H.G.
Science Conference Proceedings (OSTI)
Thermal conductivity of irradiated fuel materials, which can be derived from measured thermal diffusivity (TD), is a key consideration in thermal performance and design of a fuel rod. However, without interpretation, the measured TD data cannot be used directly to calculate fuel temperatures during irradiation. This report provides such interpretation and presents an empirical model for the degradation of UO2 thermal conductivity with burn-up.
Sensitivity of UO2 Stability in a Reducing Environment on Radiolysis Model Parameters
Science Conference Proceedings (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.
Physics calculations for mixed PuO{sub 2}-UO{sub 2} NPR loadings
At the request of NRD (NPR Physics Subsection) a study was initiated to determine the physics characteristics of various plutonium-uranium composites as fuel for the NPR. From this study, the PuO{sub 2}-UO{sub 2} fuel system was selected to receive major attention. The effect of adding a burnable poison, B-10, in intimate contact with the mixed oxide fuel was also is to be considered. The present report summarizes the results of these investigations.
Bennett, C.L.
The determination of UO/sub 2/ and UF/sub 4/ in fused fluoride salts
Science Conference Proceedings (OSTI)
The determination of uranium oxide solubilities in fused fluoride salts is important in the electrolytic preparation of uranium metal. This project was initiated to develop a method for the determination of UO/sub 2/ separately from UF/sub 4/ in UF/sub 4/-CaF/sub 2/-LiF fused salts. Previous methods used for the determination of UO/sub 2/ in fused fluoride salts involved inert gas fusions where oxygen was liberated as CO/sub 2/, and hydrofluorination where oxygen was released as H/sub 2/O; but the special equipment used for these procedures was no longer available. These methods assumed that all of the oxygen liberated was due to UO/sub 2/ and does not consider impurities from reagents and other oxygen sources that amount to a bias of approximately 0.3 wt %. This titrimetric method eliminates the bias by selectively extracting the UF/sub 4/ with a Na/sub 2/EDTA-H/sub 3/BO/sub 3/ solution. The remaining uranium oxide residue is treated and titrated gravimetrically to a potentiometric endpoint with NBS standard K/sub 2/Cr/sub 2/O/sub 7/. An aliquot of the Na/sub 2/EDTA-H/sub 3/BO/sub 3/ extract is also titrated gravimetrically to a potentiometric endpoint, this uranium component is determined and calculated as UF/sub 4/. 4 refs., 2 figs., 2 tabs.
Batiste, D.J.; Lee, D.A.
Recovery of UO.sub.2 /Pu O.sub.2 in IFR electrorefining process
A process for converting PuO.sub.2 and UO.sub.2 present in an electrorefiner to the chlorides, by contacting the PuO.sub.2 and UO.sub.2 with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the UO.sub.2 and PuO.sub.2 to metals while converting Li metal to Li.sub.2 O. Li.sub.2 O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting O.sub.2 out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li.sub.2 O to disassociate to O.sub.2 and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl.sub.2.
Tomczuk, Zygmunt (Lockport, IL); Miller, William E. (Naperville, IL)
Recovery of UO[sub 2]/PuO[sub 2] in IFR electrorefining process
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.
The viability of depositing a thin porous coating of zirconia on the outer surface of an annular UO[subscript 2] pellet
Feinroth, H.
tert-butylcyclopentadienyl)uranium Oxide, Cp� 2 UO Guofu Zi,Abstract Reduction of the uranium metallocene, [ ? 5 -group is ubiquitous in uranium chemistry as shown by the
Zi, Guofu; Werkema, Evan L.; Walter, Marc D.; Gottfriedsen, Jochen P.; Andersen, Richard A.
Nuclear data uncertainties by the PWR MOX/UO{sub 2} core rod ejection benchmark
Science Conference Proceedings (OSTI)
Rod ejection transient of the OECD/NEA and U.S. NRC PWR MOX/UO{sub 2} core benchmark is considered under the influence of nuclear data uncertainties. Using the GRS uncertainty and sensitivity software package XSUSA the propagation of the uncertainties in nuclear data up to the transient calculations are considered. A statistically representative set of transient calculations is analyzed and both integral as well as local output quantities are compared with the benchmark results of different participants. It is shown that the uncertainties in nuclear data play a crucial role in the interpretation of the results of the simulation. (authors)
Pasichnyk, I.; Klein, M.; Velkov, K.; Zwermann, W.; Pautz, A. [Boltzmannstr. 14, D-85748 Garching b. Muenchen (Germany)
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
The UO SRML is a regional solar radiation data center whose goal is to provide sound solar resource data for planning, design, deployment, and operation of solar electric facilities in the Pacific Northwest. The laboratory has been in operation since 1975. Solar data includes solar resource maps, cumulative summary data, daily totals, monthly averages, single element profile data, parsed TMY2 data, and select multifilter radiometer data. A data plotting program and other software tools are also provided. Shade analysis information and contour plots showing the effect of tilt and orientation on annual solar electric system perfomance make up a large part of the photovoltaics data.(Specialized Interface)
Science Conference Proceedings (OSTI)
The primary focus of this work was to compare MgO with UO2 as target matrix material options for burning minor actinides in a transmutation target within a sodium fast reactor. This analysis compared the transmutation performance of target assemblies having UO2 matrix to those having specifically MgO inert matrix.
M. Pope; S. Bays; R. Ferrer
POST-IRRADIATION EVALUATION OF A PLATE-TYPE UO$sub 2$ FUEL ELEMENT
The premature failure of fuel Element M22, which had six compartments of 0.100-in.-thick, 96% TD UO/sub 2/ + 6 wt% ZrO/sub 2/ fuel, was attributed to the large irradiation-induced solid volume swelling of the UO/sub 2/ fuels. This volume swelling was the result of incomplete homogenization during fabrication of the mixed and sintered U/sup E/O/sub 2/ and U/sup N/O/sub 2/ fuel s in Element M22. In addition, heavy hydriding of the Ni-free Zircaloy-2 cladding occurred in the relatively hot areas adjacent to the fuel and to a lesser extent at the external cladding surfaces. By postulation, H/sub 2/ was apparently formed by the radiolytic decomposition of water entrapped between fuel and cladding after formation of the initial cladding defect, and was absorbed by the cladding so rapidly that it could not diffuse adequately down the thermal gradient to the cold side of the cladding. The corrosion behavior of the cladding was as expected and did not contribute to the hydriding. Analysis of the CR-X-3 loop operating history indicates that no abnormal conditions external to Element M22 existed in the loop other than U from inpile intentionally defected fuel elements. (auth)
Meieran, H.B.
Talk to Language and Cognition Seminar, School of Psychology, UoB. 6 Nov 2009 Why the "hard" problem
Talk to Language and Cognition Seminar, School of Psychology, UoB. 6 Nov 2009 Why the "hard" problem of consciousness is easy and the "easy" problem hard. (And how to make progress) Aaron Sloman Sem Birmingham 2009 Slide 1 Last revised: January 27, 2010 #12;Why? Because: 1. The "hard" problem can
Sloman, Aaron
Evaluation of sintering effects on SiC incorporated UO2 kernels under Ar and Ar-4%H2 environments
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
FLUIDIZED-BED COATING OF UO$sub 2$ POWDER WITH NIOBIUM AND OTHER ELEMENTS
The chemical vapor deposition of niobium, molybdenum, tungsten, chnomium, carbcn, and niobium--vanadium alloys in a fluidized bed of UO/sub 2/ powder particles wss used to provide uniform, dense, nonporous coatings on the individual particles. in the case of niobium, which received major attention, hydrogen reduction of niobium pentachloride vapor was used as the vapor- deposition reaction. The most serious problem was that of maintaining bed fluidity aad avoiding agglomeration. This problem was overcome to permit routine operation of the coating equipment. In the entire program of 68 experimental runs, only 1.1 per cent of the product was lost by agglomeration. In routine operation, this loss should be even lower. (auth)
Blocher, J.M. Jr.; Veigel, N.D.; Oxley, J.H.; Secrest, V.M.; Rose, E.E.
High-precision molecular dynamics simulation of UO2-PuO2: pair potentials comparison
Our series of articles is devoted to high-precision molecular dynamics simulation of mixed actinide-oxide (MOX) fuel in the rigid ions approximation using high-performance graphics processors (GPU). In the first article we assess 10 most relevant interatomic sets of pair potentials (SPP) by reproduction of solid phase properties of uranium dioxide (UO2) - temperature dependences of the lattice constant, bulk modulus, enthalpy and heat capacity. Measurements were performed with 1K accuracy in a wide temperature range from 300K up to melting point. The best results are demonstrated by two recent SPPs MOX-07 and Yakub-09, which both had been fitted to the recommended thermal expansion in the range of temperatures 300-3100K. Compared with them, the widely used SPPs Basak-03 and Morelon-03 reproduce the experimental data noticeably worse at temperatures above 2500K.
Potashnikov, S I; Nekrasov, K A; Kupryazhkin, A Ya
Interim results from UO/sub 2/ fuel oxidation tests in air
Science Conference Proceedings (OSTI)
An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO/sub 2/, fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO/sub 2/ pellets in the temperature range of 135 to 250/sup 0/C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10/sup 5/ R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10/sup 5/ R/h gamma field. 33 refs., 51 figs., 6 tabs.
Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j
The Effect Of Ion Implantation Of Selected Dopants On Some Of The Electrical Properties Of UO2
The United States Department of Energy (DOE) has {approx}1 billion pounds of surplus depleted uranium (i.e., uranium tails) from uranium gaseous diffusion enrichment facilities. Rather than treating this material as waste, DOE is investigating potential beneficial uses for this uranium. Of the many possible uses, uranium dioxide (UO2) has properties that make it an equal to or better than conventional photovoltaic (e.g., solar cell) materials. For example, the electronic bandgap of UO2 occurs at an efficiency equal to that of GaS and Si, and it has five radiation adsorption peaks instead of one. This paper describes the experimental work being conducted to develop urania photovoltaic devices.
Haire, M. J.; von Roedern, R. J.; Meek, T. T.; Tesmer, J.; Wetteland, C.
High temperature properties of reactor type UO/sub 2/ pellets were measured using a Direct Electrical Heating (DEH) Facility. Modifications to the experimental apparatus have been made so that successful and reproducible DEH runs may be carried out while protecting the pellets from oxidation at high temperature. X-ray diffraction measurements on the UO/sub 2/ pellets have been made before and after runs to assure that sample oxidation has not occurred. A computer code has been developed that will model the experiment using equations that describe physical properties of the material. This code allows these equations to be checked by comparing the model results to collected data. The thermal conductivity equation for UO/sub 2/ proposed by Weilbacher has been used for this analysis. By adjusting the empirical parameters in Weilbacher's equation, experimental data can be matched by the code. From the several runs analyzed, the resulting thermal conductivity equation is lambda = 1/4.79 + 0.0247T/ + 1.06 x 10/sup -3/ exp(-1.62/kT/) - 4410. exp(-3.71/kT/) where lambda is in w/cm K, k is the Boltzman constant, and T is the temperature in Kelvin.
Bassett, B
Graphite and Beryllium Reflector Critical Assemblies of UO2 (Benchmark Experiments 2 and 3)
INTRODUCTION A series of experiments was carried out in 1962-65 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2 wt% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 252 tightly-packed fuel rods (1.27-cm triangular pitch) with graphite reflectors [1], the second part used 252 graphite-reflected fuel rods organized in a 1.506-cm triangular-pitch array [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods in a 1.506-cm-triangular-pitch configuration and in a 7-tube-cluster configuration [3]. Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. All three experiments in the series have been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5]. The evaluation of the first experiment in the series was discussed at the 2011 ANS Winter meeting [6]. The evaluations of the second and third experiments are discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power systems [7].
Margaret A. Marshall; John D. Bess
WIMS/PANTHER analysis of UO{sub 2}/MOX cores using embedded super-cells
Science Conference Proceedings (OSTI)
This paper describes a method of analysing PWR UO{sub 2}MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)
Knight, M.; Bryce, P. [EDF Energy, Barnett Way, Barnwood, Gloucester (United Kingdom); Hall, S. [Advanced Modelling and Computation Group, Imperial College, London (United Kingdom)
Experiments show that inter-granular and intra-granular gas bubbles have different growth kinetics which results in heterogeneous gas bubble microstructures in irradiated nuclear fuels. A science-based model predicting the heterogeneous microstructure evolution kinetics is desired, which enables one to study the effect of thermodynamic and kinetic properties of the system on gas bubble microstructure evolution kinetics and morphology, improve the understanding of the formation mechanisms of heterogeneous gas bubble microstructure, and provide the microstructure to macroscale approaches to study their impact on thermo-mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking. In our previous report 'Mesoscale Benchmark Demonstration, Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing', we developed a phase-field model to simulate the intra-granular gas bubble evolution in a single crystal during post-irradiation thermal annealing. In this work, we enhanced the model by incorporating thermodynamic and kinetic properties at grain boundaries, which can be obtained from atomistic simulations, to simulate fission gas bubble growth kinetics in polycrystalline UO2 fuels. The model takes into account of gas atom and vacancy diffusion, vacancy trapping and emission at defects, gas atom absorption and resolution at gas bubbles, internal pressure in gas bubbles, elastic interaction between defects and gas bubbles, and the difference of thermodynamic and kinetic properties in matrix and grain boundaries. We applied the model to simulate gas atom segregation at grain boundaries and the effect of interfacial energy and gas mobility on gas bubble morphology and growth kinetics in a bi-crystal UO2 during post-irradiation thermal annealing. The preliminary results demonstrate that the model can produce the equilibrium thermodynamic properties and the morphology of gas bubbles at grain boundaries for given grain boundary properties. More validation of the model capability in polycrystalline is underway.
Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert O.; Gao, Fei; Sun, Xin
The purpose of this report is to document reference calculations performed using the SCALE-4.0 code system to determine the critical parameters of UO{sub 2}F{sub 2}-H{sub 2}O spheres. The calculations are an extension of those documented in ORNL/CSD/TM-284. Specifically, the data for low-enriched UO{sub 2}F{sub 2}-H{sub 2}O spheres have been extended to highly enriched uranium. These calculations, together with those reported in ORNL/CSD/TM-284, provide a consistent set of critical parameters (k{sub {infinity}}, volume, mass, mass of water) for UO{sub 2}F{sub 2} and water over the full range of enrichment and moderation ratio.
Jordan, W.C.
Uranium vacancy mobility at the sigma 5 symmetric tilt grain boundary in UO2
Science Conference Proceedings (OSTI)
An important consequence of the fissioning process occurring during burnup is the formation of fission products. These fission products alter the thermo-mechanical properties of the fuel. They also lead to macroscopic changes in the fuel structure, including the formation of bubbles that are connected to swelling of the fuel. Subsequent release of fission gases increase the pressure in the plenum and can cause changes in the properties of the fuel pin itself. It is thus imperative to understand how fission products, and fission gases in particular, behave within the fuel in order to predict the performance of the fuel under operating conditions. Fission gas redistribution within the fuel is governed by mass transport and the presence of sinks such as impurities, dislocations, and grain boundaries. Thus, to understand how the distribution of fission gases evolves in the fuel, we must understand the underlying transport mechanisms, tied to the concentrations and mobilities of defects within the material, and how these gases interact with microstructural features that might act as sinks. Both of these issues have been addressed in previous work under NEAMS. However, once a fission product has reached a sink, such as a grain boundary, its mobility may be different there than in the grain interior and predicting how, for example, bubbles nucleate within grain boundaries necessitates an understanding of how fission gases diffuse within boundaries. That is the goal of the present work. In this report, we describe atomic level simulations of uranium vacancy diffusion in the pressence of a {Sigma}5 symmetric tilt boundary in urania (UO{sub 2}). This boundary was chosen as it is the simplest of the boundaries we considered in previous work on segregation and serves as a starting point for understanding defect mobility at boundaries. We use a combination of molecular statics calculations and kinetic Monte Carlo (kMC) to determine how the mobility of uranium vacancies is altered at this particular grain boundary. Given that the diffusion of fission gases such as Xe are tied to the mobility of uranium vacancies, these results given insight into how fission gas mobility differs at grain boundaries compared to bulk urania.
Uberuaga, Blas P. [Los Alamos National Laboratory
RELAP5/MOD3.2 analysis of a VVER-1000 reactor with UO[2] fuel and MOX fuel
A RELAP5/MOD3.2 model of a VVER-1000/MODEL V320 nuclear power plant, Balakovo Unit 4, was updated, improved and validated on the basis of an input deck prepared by the Kurchatov Institute of Moscow. The RELAP5 model includes both the primary and the secondary systems. The Emergency Core Cooling System (ECCS) is modeled according to the plant configuration. The feedwater system, along with the emergency feedwater system, is included in the model. The point reactor kinetics model, in which the decay heat is calculated with ANS decay heat data, enables the model to be used for analysis of a large spectrum of transients and accidents. The plant model is used for analysis and prediction of a cold leg Large Break Loss-of-Coolant Accident (LBLOCA). The RELAP5/MOD3.2 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 than that of UO? assembly. An uncertainty analysis has been performed for the peak cladding temperature, in which Monte Carlo calculations have been performed using the response surface built up from fifteen sets of RELAP5 calculations. The result shows that the ECCS would be sufficient to keep the cladding temperature during the scenario of a LBLOCA well below the required licensing limit.
Fu, Chun
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
HEAVY ELEMENT ISOTOPIC ANALYSIS OF UO$sub 2$ FUEL IRRADIATED IN THE VBWR. Report No. 1
Slightly enriched UO/sub 2/ fuel, irradiated in the Vallecitos Boiling Water Reactor (VBWR), with exposures ranging from 100 Mwd/t to 3200 Mwd/t was analyzed for heavy element isotopic composition and compared with computed data. The primary objective of this program is to obtain improved data on the changes in nuclear characteristics with burnup of UO/sub 2/ fuel in a boiling water reactor. This information is important in both evaluating the economics of a given reactor design and also in providing a sounder physics basis for improving reactor designs to minimize the resuiting fuel costs. Uranium oxide pellets, with an enrichment of 2.8 atom percent, were analyzed at several axial positions along the fuel rod, spanning the void (steam fraction) range of 0 to 30%. The isotopic composition for each pellet was computed, utilizing a general fuel cycle depletion code. Results of the analysis of the comparison of the measured and computed data indicate that the total amount of Pu computed is consistently lower than that implied from the measurement by approximately 10%, and the percentage difference between the measured and computed data increases slightly with exposure. One rod was irradiated near a control rod which was approximately 25% inserted. As expected, since no control rod effects were included in the calcuiation, the measured data in that region of the rod shows a greater Pu production per Mwd/t than computed. Physical effects which might explain the small, but apparentiy consistent, differences between the measured and computed data were postulated. It is concluded that the observed differences are the result of a substantial underestimate of void fraction and small uncertainties in fuel exposure and cross sections. (auth)
Hackney, M.R.; Ruiz, C.P.
The economics of mechanical shear-leach and Sulfex decladding-core dissolution head end treatments for processing typical tubular bundles of stainless steel-jacketed UO/sub 2/ nuclear fuels were compared. A 2.66 metric ton U/day head end portion of a plant was designed for each process and capital and operating costs were developed. For plants of this size and larger, mechanical shear-leach processing has the advantage of ~20% lower capital cost and 50% lower operating cost. Processing costs of stainless steel-jacketed UO/ sub 2/ by the Sulfex and shear-leach methods, including amortization and waste disposal but excluding solvent extraction, were .78 and 7l/kg U, respectively. Storage of stainless steel waste produced by the shear-leach method is less costly by a factor of 20 than for Sulfex. (auth)
Adams, J.B.; Benis, A.M.; Watson, C.D.
Science Conference Proceedings (OSTI)
The local structure and chemical speciation of the mixed valence, fluorite-based oxides UO{sub 2+x} (0.00=
Conradson, Steven D. [Los Alamos National Laboratory, Chemistry Division, Materials Science and Technology Division and Nuclear Materials Technology Division, Los Alamos, NM 87545 (United States)]. E-mail: firstname.lastname@example.com; Begg, Bruce D. [Australian Nuclear Science and Technolgy Organisation, Menai, NSW 2234 (Australia); Clark, David L. [Los Alamos National Laboratory, Chemistry Division, Materials Science and Technology Division and Nuclear Materials Technology Division, Los Alamos, NM 87545 (United States)] [and others
The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO/sub 2/ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr. (auth)
Cottrell, W.B.; Copenhaver, C.M.; Culver, H.N.; Fontana, M.H.; Kelleghan, V.J.; Samuels, G.
Science Conference Proceedings (OSTI)
A series of critical experiments were performed with 4.31 wt % /sup 235/U enriched UO/sub 2/ fuel rods immersed in water containing various concentrations of boron ranging up to 2.55 g/l. The boron was added in the form of boric acid (H/sub 3/BO/sub 3/). Critical experimental data were obtained for two different lattice pitches wherein the water-to-uranium oxide volume ratios were 1.59 and 1.09. The experiments provide benchmarks on heavily borated systems for use in validating calculational techniques employed in analyzing fuel shipping casks and spent fuel storage systems that may utilize boron for criticality control.
Durst, B.M.; Bierman, S.R.; Clayton, E.D.
In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The Framatome-ANP production of its French low enriched (5 w/o) UO{sub 2} fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO{sub 2} blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach Framatome-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (authors)
Doucet, M.; Zheng, S. [Framatome-ANP Fuel Technology Service (France); Mouton, J.; Porte, R. [Framatome-ANP Fuel Fabrication Plant - FBFC (France)
Science Conference Proceedings (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)
A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling methods used in this study.
Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David
In-pile loop experiments L-2-15 and L-4-16 were designed to test the radiation corrosion of Zircaloy-2 and other possible reactor construction materials in UO/sub 2/SO/sub 4/ solutions. The solutions employed were 0.17 m UO/ sub 2/SO/sub 4/, 0.015 m CuSO/sub 4/, and 0.03 m H/sub 2/SO/sub 4/ in H/sub 2/O for experiment L-2-15, and 0.17 m UO/sub 2/SO/sub 4/, 0.015 m CuSO/sub 4/, and 0.025 m H/sub 2/SO/sub 4/ in H/sub 2/O for experiment L-4-16. The mainstream temperature in the experiments ranged from 278 to 280 deg C. Construction material for the loops was type 347 stainless steel. Specimens of types 347 and 309SCb stainless steels titanium-55A and -110AT, platinum, Zircaloy-2, crystalbar zirconium, and a variety of other zirconium alloys were tested. The power density at core specimens ranged from 19.8 to 4.6 w/ml in L-2-15 and from 5.7 to 1.3 w/ml in L-4-16. For loop L-2-15, the total time of hightemperature operation with UO/sub 2/SO/sub 4/ was 792 hr, during in-pile exposure, and the reactor energy was 1632 Mwh; for loop L-4-16, 1032 hr and 2325 Mwh. During both experiments most of the reactor energy was accumulated at 3-Mw power level. In general, stainless steel corrosion results from these experiments were comparable to those observed in other in-pile loop experiments. Corrosion was confined primarily to the core areas and was power-density dependent. Some variations in attack, both positive and negative, with velocity of solution flow past specimens have been observed in other experiments, but there was no apparent effect of varying velocities in the range 10 to 40 fps on either the core-channel or in- line channel specimens in the present experiments. The coreannulus steel specimens in L-2-15 corroded at rates very much greater than those in the channel. This difference may have resulted, in part, from the differences in velocities, however, it may have also been a result of galvanic actton between the steel annulus specimens and adjacent platinum specimens. In previous 250 deg C experiments the occurrence of a change in the stainless steel corrosion rate was correlated with a decrease in acidity and/or increase in the nickel concentration. The results for the oxygen consumption rates on steel during radiation exposure in the present experiments varied with radiation time in a manner qualitatively similar to that observed at the lower temperature. However, the concentration of excess acid in the present experiments probably remained fairly constant throughout the radiation exposures, and correlations similar to those obtained at the lower temperature could not be established. The acid concentration in the 280 deg C experiments was greater than the concentrations prevailing when corrosion rate changes occurred in the 250 deg C experiments. The difference in acid tolerance is probably a result of the increased temperature, since a similar beneficial effect of temperature occurs out-ofpile No overall correlation has been established for the various factors found to have influenced steel corrosion in previous experiments. Results of the present experiments provide additional evidence in support of previous findings but do not further their interpretation. Zircaloy-2 corrosion results from both loops have been discussed and correlated elsewhere in terms of the 280 deg C relationship between the corrosion rate R (mils per year, mpy), power density P (w/ml), and uranium sorption factor alpha : 1/R = 2.23/P alpha + 1/40. The data from these experiments obey this relationship. (This is only a portion of the Author abstract.)
Jenks, G.H.; Baker, J.E.
APS Long Range Schedule FY1999
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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
APS Long Range Schedule FY1998
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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
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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
UO 2 Mixed Oxide System Using Atomic Level Simulations
Science Conference Proceedings (OSTI)
Thorium-based nuclear materials offer the promise of increased proliferation resistance, longer fuel cycles, higher burnup and improved wasteform�...
Electrochemistry of Defects in Irradiated UO 2 - Programmaster.org
Science Conference Proceedings (OSTI)
This research was supported as a part of the EFRC on Materials Science of Nuclear Fuel funded by the U.S. DOE, BES under subcontract #00091538 from INL�...
Atomistic Studies of Defect Cluster Migration mechanisms in UO2
Science Conference Proceedings (OSTI)
... phase field modeling. This work is supported by the EFRC program funded by DOE BES under Award Number FWP 1356. Proceedings Inclusion? Planned:�...
Crack tip plasticity in single crystal UO2: Atomistic simulations
The fracture behavior of single crystal uranium dioxide is studied using molecular dynamics simulations at room temperature. Initially, an elliptical notch is created on either {111} or {110} planes, and tensile loading is applied normal to the crack planes. For cracks on both planes, shielding of crack tips by plastic deformation is observed, and crack extension occurs for crack on {111} planes only. Two plastic processes, dislocation emission and phase transformation are identified at crack tips. The dislocations have a Burgers vector of ?110?/2, and glide on {100} planes. Two metastable phases, the so-called Rutile and Scrutinyite phases, are identified during the phase transformation, and their relative stability is confirmed by separate density- functional-theory calculations. Examination of stress concentration near crack tips reveals that dislocation emission is not an effective shielding mechanism. The formation of new phases may effectively shield the crack provided all phase interfaces formed near the crack tips are coherent, as in the case of cracks residing on {110} planes.
Yongfeng Zhang; Paul C. Millett; Michael Tonks; Bulent Biner; Xiang-Yang Liu; David A. Andersson
CHARTER RESERVATION / BILLING FORM UOS Transportation Services -Passenger Transport
= _________________________ R/V Nucella (daily rate) # of days x $390.50/day = _________________________ Nucella fuel surcharge = _________________________ Ira C. fuel surcharge* _________________________ * At current fuel prices plan on $50 per day within offshore trips we can help you estimate fuel costs. Budget Darling Marine Center Visiting Graduate Student
Needleman, Daniel
Computer Simulation of Radiation Effects and Defects in UO 2
Science Conference Proceedings (OSTI)
Abstract Scope, There is renewed interest in nuclear power due to rising global energy demand, concerns about energy security and the adverse environmental
PUREX/UO{sub 3} facilities deactivation lessons learned: History
Science Conference Proceedings (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.
UoS PhD Studentship! A University of Sheffield PhD studentship within the framework of Project
Sunshine is available, held jointly at the Solar Physics and Space Plasma Research Centre (SP2 RC), School
Dixon, Peter
APS Long Range Schedule FY1997
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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
Random-Walk Monte Carlo Simulation of Intergranular Gas Bubble Nucleation in UO2 Fuel
Science Conference Proceedings (OSTI)
Using a random-walk particle algorithm, we investigate the clustering of fission gas atoms on grain bound- aries in oxide fuels. The computational algorithm implemented in this work considers a planar surface representing a grain boundary on which particles appear at a rate dictated by the Booth flux, migrate two dimensionally according to their grain boundary diffusivity, and coalesce by random encounters. Specifically, the intergranular bubble nucleation density is the key variable we investigate using a parametric study in which the temperature, grain boundary gas diffusivity, and grain boundary segregation energy are varied. The results reveal that the grain boundary bubble nucleation density can vary widely due to these three parameters, which may be an important factor in the observed variability in intergranular bubble percolation among grain boundaries in oxide fuel during fission gas release.
Yongfeng Zhang; Michael R. Tonks; S. B. Biner; D.A. Andersson
Comparison of Point-Defect Evolution in Irradiated UO2 and Ceo2 ...
Science Conference Proceedings (OSTI)
... Corrosion Inhibition for Hydrochloric Acid Pickling � Using Resistance Heating to Create Full-Scale API RP2Z CTOD Samples�...
Evaluation of UF6 to UO2 Conversion Capability at Commercial...
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. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 2.2.7 Korea . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....
A brief history of the PUREX and UO{sub 3} facilities
The Plutonium-Uranium Extraction (PUREX) Plant, conceived during the early Cold War years, was a vehicle to increase significantly US nuclear weapons production capacity. The original PUREX Plant was a concrete rectangle 1,005 feet long and 61.5 feet wide. The shielding capacity of the concrete was designed so that personnel in non-regulated service areas would not receive radiation in excess of 0.1 millirem per hour. This report discusses the design of the PUREX Plant, the production chronology, projects and equipment changes, equipment decontamination and reuse, waste management, and contamination events that have occurred during the operation of the plant. Additionally, the development and history of the Uranium Trioxide Plant are also covered.
Gerber, M.S.
Modeling of grain growth in UO2 under a temperature gradient
Science Conference Proceedings (OSTI)
Characterization of MOX fuel pellets by Photothermal microscopy � Correlation Between Thermal Conductivity and Microstructural Evolutions in CeO2 Upon�...
Recovery of UO{sub 2}/PuO{sub 2} in IFR electrorefining process
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.
ZeptoOS: Operating Systems for Petascale | Argonne National Laboratory
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Kamil Iskra Kazutomo Yoshii Other Contributors Ivan Beschastnikh (UW) Susan Coghlan (ALCF) Cameron Cooper (OSU) Aroon Nataraj (UO) Al Malony (UO) Sameer Shende (UO) Suravee...
An inductive double-contingency analysis (DCA) method developed by the criticality safety function at the Savannah River Site, was applied in Criticality Safety Evaluations (CSEs) of five major plant process systems at the Westinghouse Electric Corporation`s Commercial Nuclear Fuel Manufacturing Plant in Columbia, South Carolina (WEC-Cola.). The method emphasizes a thorough evaluation of the controls intended to provide barriers against criticality for postulated initiating events, and has been demonstrated effective at identifying common mode failure potential and interdependence among multiple controls. A description of the method and an example of its application is provided.
Skiles, S. K. [Westinghouse Savannah River Company, Aiken, SC (United States)
Science Conference Proceedings (OSTI)
Technical Paper / Advances in Nuclear Fuel Management - Increased Enrichment/High Burnup and Light Water Reactor Fuel Cycle Optimization
Robert Gregg; Andrew Worrall
?A program is being conducted to develop a process for cladding tungsten and tungsten cermet fuels with tungsten deposited from the vapor state by the hydrogen reduction of tungsten hexafluoride. Early work was performed using recrystallized, high purity, commercial tungsten as the substrate material. Temperatures in the range 660 to 12950F (350 to 1700�C) and pressures from 10 to 350 mm Hg were investigated. Hydrogen to WF 6 ratios of 10: 1 to 150: 1 were utilized. Efforts were directed toward optimizing deposition process parameters to attain control of qualities such as coating thickness, uniformity, density, impurity content, and surface quality. Substrate penetration methods have been investigated in the interest of completely eliminating the interface between the fueled substrate and cladding. In addition, the effects of process parameters and post-cladding heat treatments on the fuel retention properties of clad composites at 4500 degrees F (2480 degrees C) in hydrogen for 2 hours have been evaluated. As a result of work performed during the first phase of the program it has been shown that the rate of deposition of tungsten from WF 6 and the uniformity of the deposit can be varied in a predictable and reproducible manner by exercising control over the temperature, pressure, and gas flow rates at which the deposits are produced. A significant result of the study is the discovery that substrate nucleation and epitaxial growth in deposits made on both unfueled tungsten and fueled substrates may be effected by pretreating the substrates in hydrogen. High temperature fuel retention testing of tungsten clad W-U02 at 45000F (2480 degrees C) in hydrogen for 2 hours has demonstrated that the vapor deposited layer effectively and consistently restricts fuel loss.
Lamartine, J.T.; Hoppe, A.W.
E&nr Ph. S. W.. Wahhgt~n. D.C. 200242174, TIkpbnc (202) 48a60uo
Office of Legacy Management (LM)
under consideration; and six institutions recently iden- tq .o-OS tified during a search of Hanford records. p1,6-01 I '1. 2 8 The attached was prepared to document the...
A QM/MM Study on the Aqueous Solvation of the Tetrahydroxouranylate [UO?(OH)?]�? Complex Ion
We report a QM augmented QM/MM study on the coordination of the tetrahydroxouranylate ion in aqueous solution. QM/MM geometry optimizations followed by full QM single-point calculations on the optimized structures show that a hexa-coordinated structure is more stable than the hepta-coordinated structure by 43 kJ/mol. Charge transfer of the tetrahydroxouranylate to the solvating water molecules is relatively modest, and can be modeled by including a solvation layer consisting of 12 explicit water molecules.
Infante, Ivan A.; van Stralen, Bas; Visscher, Lucas
The radiological and toxicological consequences of a hypothetical release of U03 powder from the T-hopper storage pad adjacent the 2714-U building were calculated.
High-precision molecular dynamics simulation of UO2-PuO2: superionic transition in uranium dioxide
Our series of articles is devoted to high-precision molecular dynamics simulation of mixed actinide-oxide (MOX) fuel in the rigid ions approximation using high-performance graphics processors (GPU). In this article we assess the 10 most relevant interatomic sets of pair potential (SPP) by reproduction of the Bredig superionic phase transition (anion sublattice premelting) in uranium dioxide. The measurements carried out in a wide temperature range from 300K up to melting point with 1K accuracy allowed reliable detection of this phase transition with each SPP. The {\\lambda}-peaks obtained are smoother and wider than it was assumed previously. In addition, for the first time a pressure dependence of the {\\lambda}-peak characteristics was measured, in a range from -5 GPa to 5 GPa its amplitudes had parabolic plot and temperatures had linear (that is similar to the Clausius-Clapeyron equation for melting temperature).
Potashnikov, S I; Nekrasov, K A; Kupryazhkin, A Ya
A literature review on the chemical and physical properties of uranyl fluoride (UO sub 2 F sub 2 )
Science Conference Proceedings (OSTI)
This report reviews the preparation and properties of uranyl fluoride. Data are given on the crystal structure, solubility in water, specific gravity, density, specific heat, enthalpy, entropy, acidity, corrosion properties, and refractive indices. Empirical formulas are given to calculate specific gravity, density of aqueous solutions, molal volume, and refractive indices. 13 refs., 3 figs., 11 tabs.
Myers, W.L. (Los Alamos National Lab., NM (USA) Illinois Univ., Urbana, IL (USA). Dept. of Nuclear Engineering)
Office of Legacy Management (LM)
UO2 Dissolution ... 5-a 3: High Density UO2 Dissolution ... 6-a 4. Allowable Interactions With U02 Feed ... 6-a ., Table 1' .:...
Our current state-of-the-art X-ray diffraction experiments are primarily sensitive to the position of the uranium atom. While the uranium - low-Z element bond (such as U-H or U-F) changes under pressure and temperature the X-ray diffraction investigations do not reveal information about the bonding or the stoichiometry. Questions that can be answered by Raman spectroscopy are (i) whether the bonding strength changes under pressure, as observed by either blue- or red-shifted peaks of the Raman active bands in the spectrum and (ii) whether the low-Z element will eventually be liberated and leave the host lattice, i.e. do the fluorine, oxygen, or hydrogen atoms form dimers after breaking the bond to the uranium atom. Therefore Raman spectra were also collected in the range where those decomposition products would appear. Raman is particularly well suited to these types of investigations due to its sensitivity to trace amounts of materials. One challenge for Raman investigations of the uranium compounds is that they are opaque to visible light. They absorb the incoming radiation and quickly heat up to the point of decomposition. This has been dealt with in the past by keeping the incoming laser power to very low levels on the tens of milliWatt range consequently affecting signal to noise. Recent modern investigations also used very small laser spot sizes (micrometer range) but ran again into the problem of heating and chemical sensitivity to the environment. In the studies presented here (in contrast to all other studies that were performed at ambient conditions only) we employ micro-Raman spectroscopy of samples situated in a diamond anvil cell. This increases the trustworthiness of the obtained data in several key-aspects: (a) We surrounded the samples in the DAC with neon as a pressure transmitting medium, a noble gas that is absolutely chemically inert. (b) Through the medium the sample is thermally heat sunk to the diamond anvils, diamond of course possessing the very best heat conductivity of any material. Therefore local heating and decomposition are avoided, a big challenge with other approaches casting doubts on their results. (c) This in turn benefits the signal/noise ratio tremendously since the Raman features of uranium-compounds are very small. The placement of the samples in DACs allows for higher laser powers to impinge on the sample spot while keeping the spot-size larger than in previous studies and keep the samples from heating up. Raman spectroscopy is a very sensitive non-invasive technique and we will show that it is even possible to distinguish the materials by their origin / manufacturer as we have studied samples from Cameco (Canada) and IBI-Labs (US-Florida) and can compare with ambient literature data for samples from Strem (US-MA) and Areva (Pierrelatte, France).
Lipp, M J; Jenei, Z; Park-Klepeis, J; Evans, W J
Science Conference Proceedings (OSTI)
The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.
Carbajo, J.J.
Properties of Uranium Compounds
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Triuranium Octaoxide (U3O8) Uranium Dioxide (UO2) Uranium Tetrafluoride (U4) Uranyl Fluoride (UO2F2) The physical properties of the pertinent chemical forms of uranium are...
Fuel performance models have been developed to assess the performance of ThO[subscript 2]-UO[subscript 2]
Long, Y.
Calculations of Threshold Displacement Energies in Y2Ti2O7 and ...
Science Conference Proceedings (OSTI)
Symposium, Radiation Effects in Oxide Ceramics and Novel LWR Fuels ... Understanding Nuclear Fuel Thermal Conductivity from Phonons in UO2.
Shale gas extraction in the UK: What the people think? Applications are invited for a three -year/EU rate) and a maintenance grant (£13,590 in 12/13). The emergence of shale gas on the energy landscape in the transition to a low carbon economy. In the US, for example, the speed at which shale gas has been developed
Nottingham, University of
Nuclear materials research progress reports for 1977
Research is reported concerning radiation enhancement of stress corrosion cracking of Zircaloy, surface chemistry of epitaxial Si deposited by thermal cracking of silane, thermal gradient migration of metallic inclusions in UO/sub 2/, molecular beam studies of atomic H and reduction of oxides, mass transfer and reduction of UO/sub 2/, kinetics of laser pulse vaporization of UO/sub 2/, retention and release of water by UO/sub 2/ pellets, and solubility of H in UO/sub 2/. (FS)
Olander, D.R.
Investigation of Uranium Polymorphs
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.
Core Designs and Economic Analyses of Homogeneous Thoria-Urania Fuel in Light Water Reactors
The objective is to develop equilibrium fuel cycle designs for a typical pressurized water reactor (PWR) loaded with homogeneously mixed uranium-thorium dioxide (ThO{sub 2}-UO{sub 2}) fuel and compare those designs with more conventional UO{sub 2} designs.The fuel cycle analyses indicate that ThO{sub 2}-UO{sub 2} fuel cycles are technically feasible in modern PWRs. Both power peaking and soluble boron concentrations tend to be lower than in conventional UO{sub 2} fuel cycles, and the burnable poison requirements are less.However, the additional costs associated with the use of homogeneous ThO{sub 2}-UO{sub 2} fuel in a PWR are significant, and extrapolation of the results gives no indication that further increases in burnup will make thoria-urania fuel economically competitive with the current UO{sub 2} fuel used in light water reactors.
Saglam, Mehmet; Sapyta, Joe J.; Spetz, Stewart W.; Hassler, Lawrence A. [Framatome ANP, Inc. (France)
ABS>A method of separating a mixture of UO/sub 2/ and PuO/sub 2/ is given which comprises immersing the mixture in a fused NaCl-KCl bath, chlorinating with chlorine or phosgene, and preferentially electrolytically or chemically reducing the UO/sub 2/Cl/sub 2/ so produced to UO/sub 2/ and filtering it out. (AEC)
Benedict, G.E.; Lyon, W.L.
Accumulation of uranium at low concentration by the green alga ...
Science Conference Proceedings (OSTI)
by Scenedesmus obliquus 34 was rapid and energy-independent and the biosorption of. UO2+ ... by the green algaScenedesmus obliquus34 is described here.
Pyrolitic Uranium Compound (PYRUC)
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Pyrolitic Uranium Compound Pyrolitic Uranium Compound (PYRUC) PYRolitic Uranium Compound (PYRUC) is a shielding material consisting of depleted uranium UO2 or UC in either pellet...
Production and Handling Slide 14: Conversion of Yellow Cake to...
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Uranium dioxide UO2, called "brown oxide," is formed by reducing ammonium diuranate (NH4)2U2O7 by the addition of hydrogen. Slide 14...
Production and Handling Slide 15: Yellow Cake, Uranyl Nitrate...
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Skip Presentation Navigation First Slide Previous Slide Next Slide Last Presentation Table of Contents Yellow Cake, Uranyl Nitrate, ADU, UO2 Refer to caption below for image...
Production and Handling Slide 16: Conversion of Yellow Cake to...
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Hydrofluoric acid HF is added to uranium dioxide UO2 to form uranium tetrafluoride UF4, often called "green salt." Slide 16...
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For example, uranium dioxide (UO2) is the primary nuclear fuel in light-water reactors. ... He graduated from Ohio Wesleyan University in 1978 (BA in Physics �...
Metallic Fuel: Modeling and Simulation
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Oct 9, 2012 ... Thermophyical Properties of Thoria and ThO2-UO2 Mixed Oxide Fuels ... Thorium-based nuclear fuel cycles are promising for their intrinsic�...
184 Synthesis, Structure and Characterization of Silver Doped ...
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
098 Room Temperature Multiferroism in Nanocapacitors
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
080 Enhanced Magnetization and Magnetoelectric Coupling in ...
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
049 The Effect of Current Type on Morphology, Chemical ...
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
150 Evolution of Internal Strain with Temperature in Depleted ...
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
095 Oxyfluoride Based Low Dielectric Constant LTCC Materials
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
070 Characteristics of Superconducting YBCO Phase Formation ...
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
192 Fabrication and Application of Ag/Al(OH)3 Nanocomposite Film ...
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129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
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Jul 31, 2011 ... Lifetime extension of reactors - Nuclear materials aging, ... First-Principles Theory of Magnetism, Crystal Field and Phonon Spectrum of UO2.
3D Microstructural Characterization of Oxide Nuclear Fuel Surrogates
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... were developed to obtain Electron Backscatter Diffraction (EBSD) data for depleted UO2 pellets. ... Segregation of Ru to Edge Dislocations in Uranium Dioxide.
P1-04: 3D Microstructural Characterization of Uranium Oxide as a ...
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Presentation Title, P1-04: 3D Microstructural Characterization of Uranium ... to obtain Electron Backscatter Diffraction (EBSD) data for depleted UO2 pellets that �...
Colleagues and Friends Remember Milton Wadsworth, 1983 TMS ...
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Feb 18, 2013 ... He also held a number of administrative positions at UoU, including two separate terms as Department of Metallurgy chair, Utah Mining and�...
A Potential Approach to Address Materials Aging Issues in Nuclear ...
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Potential applications of this technology for nuclear reactor aystems will be ... First-Principles Theory of Magnetism, Crystal Field and Phonon Spectrum of UO2 .
Light Water Reactor Materials for Commercial Nuclear Power ...
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Presentation Title, Light Water Reactor Materials for Commercial Nuclear ... First- Principles Theory of Magnetism, Crystal Field and Phonon Spectrum of UO2.
Environmentally Assisted Cracking of Materials
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In-Situ Repairs of Oil Industry Pipelines, Tanks and Vessels by Welding Using Metal Arc Welding Under Oil (MAW-UO) � Interpretation of Crack Initiation and�...
Pair Distribution Function Analysis of Irradiated Cladding and Duct ...
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First-Principles Theory of Magnetism, Crystal Field and Phonon Spectrum of UO2 ... Light Water Reactor Materials for Commercial Nuclear Power Applications.
Practical estimation of veering effects on high-rise structures: a ...
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... Teshigawara, M. (2001), Structural design principles (chapter 6 ... RH and Lappe, UO (1964), �Wind and temperature ... on a 1400 ft tower�, J. Appl. ...
University of Pune Inde | India
, Anthropologie, Science Politique, Arts, Common law, Psychologie IMPORTANT NOTE: This document contains: History, Anthropology, Political Science, Arts, Common Law, Psychology Faculté | Faculty Titre de cours | Course Title Code �?quivalence UO UO Equivalency Sciences sociales Social Sciences International Political
Petriu, Emil M.
Multi-scale molecular simulations of biological systems: Parallelization of RAPTOR for Blue Gene/Q
/Q Adrian W. Lange MiraCon ALCF 03/06/13 #12;Multi-state empirical valence bond Goal: - To simulate (ALCF) - Each subcommunicator creates own instance of LAMMPS - Replica Exchange between subcomms LAMMPSC) Christopher Knight (ANL) Gard Nelson (UoC) Yuxing Peng (UoC) Jeff Hammond (ALCF) Luke Westby (Notre Dame
Kemner, Ken
KTH ReseaRcH assessmenT exeRcise 2012 KTH RESEARCH ASSESSMENT
, although some UoAs now belong to a different RF, and one new RF has been created since 2008 in the RF, as well as by the Vice President for Research. The RFC has overall responsibility for planning into 13 research fields and 47 UoAs, as follows: Rf1 mathematics Research field coordinator: Professor
Lagergren, Jens
A process is presented for separating colloidally dissed uranium oxides from the heavy water medium in upwhich they are contained. The method consists in treating such dispersions with hydrogen peroxide, thereby converting the uranium to non-colloidal UO/sub 4/, and separating the UO/sub 4/ sfter its rapid settling.
Wigner, E.P.; McAdams, W.A.
In-situ reductive biotransformation of subsurface U(VI) to U(IV) (as ?UO2?) has been proposed as a bioremediation method to immobilize uranium at contaminated DOE sites. The chemical stability of bacteriogenic ?UO2? is the seminal issue governing its success as an in-situ waste form in the subsurface. The structure and properties of chemically synthesized UO2+x have been investigated in great detail. It has been found to exhibit complex structural disorder, with nonstoichiometry being common, hence the designation ?UO2+x?, where 0 < x < 0.25. Little is known about the structures and properties of the important bacteriogenic analogs, which are believed to occur as nanoparticles in the environment. Chemically synthesized UO2+x exhibits an open fluorite structure and is known to accommodate significant doping of divalent cations. The extent to which bacteriogenic UO2+x incorporates common ground water cations (e.g., Ca2+) has not been investigated, and little is known about nonstoichiometry and structure defects in the bacteriogenic material. Particle size, nonstoichiometry, and doping may significantly alter the reactivity, and hence stability, of bacteriogenic UO2+x in the subsurface. The presence of associated sulfide minerals, and solid phase oxidants such as bacteriogenic Mn oxides may also affect the longevity of bacteriogenic UO2 in the subsurface.
Bargar, John R.
Defects and Related Phenomena I - Programmaster.org
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Metal oxides added to UO2 to improve material performance during irradiation, ... The oxygen and uranium Frenkel pairs and the uranium-oxygen Schottky ... Using the law of mass action to the Frenkel and Schottky defects in doped UO2, ... migration in the fluorite-structured oxide CeO2 is investigated at the atomistic level.
Chemical reduction of refractory oxides by atomic hydrogen
The chemical reduction of UO/sub 2/ and Al/sub 2/O/sub 3/ by atomic hydrogen was studied. Results of the UO/sub 2//H investigation indicates that reduction of UO/sub 2/ by atomic hydrogen proceeds by the production of water vapor and hypostoichiometric urania. Water vapor and aluminum metal are formed in the Al/sub 2/O/sub 3//H system. The relative ease which UO/sub 2/ is reduced by atomic hydrogen compared with Al/sub 2/O/sub 3/ is due to two factors. The first is related to the thermochemistry of the reactions. The second factor which favors efficient reduction of UO/sub 2/ but not of Al/sub 2/O/sub 3/ is the oxygen diffusivity. (LK)
Dooley, D.; Balooch, M.; Olander, D.R.
Reaction of uranium oxides with chlorine and carbon or carbon monoxide to prepare uranium chlorides
The preferred preparation concept of uranium metal for feed to an AVLIS uranium enrichment process requires preparation of uranium tetrachloride (UCI{sub 4}) by reacting uranium oxides (UO{sub 2}/UO{sub 3}) and chlorine (Cl{sub 2}) in a molten chloride salt medium. UO{sub 2} is a very stable metal oxide; thus, the chemical conversion requires both a chlorinating agent and a reducing agent that gives an oxide product which is much more stable than the corresponding chloride. Experimental studies in a quartz reactor of 4-cm ID have demonstrated the practically of some chemical flow sheets. Experimentation has illustrated a sequence of results concerning the chemical flow sheets. Tests with a graphite block at 850{degrees}C demonstrated rapid reactions of Cl{sub 2} and evolution of carbon dioxide (CO{sub 2}) as a product. Use of carbon monoxide (CO) as the reducing agent also gave rapid reactions of Cl{sub 2} and formation of CO{sub 2} at lower temperatures, but the reduction reactions were slower than the chlorinations. Carbon powder in the molten salt melt gave higher rates of reduction and better steady state utilization of Cl{sub 2}. Addition of UO{sub 2} feed while chlorination was in progress greatly improved the operation by avoiding the plugging effects from high UO{sub 2} concentrations and the poor Cl{sub 2} utilizations from low UO{sub 2} concentrations. An UO{sub 3} feed gave undesirable effects while a feed of UO{sub 2}-C spheres was excellent. The UO{sub 2}-C spheres also gave good rates of reaction as a fixed bed without any molten chloride salt. Results with a larger reactor and a bottom condenser for volatilized uranium show collection of condensed uranium chlorides as a loose powder and chlorine utilizations of 95--98% at high feed rates. 14 refs., 7 figs., 14 tabs.
Haas, P.A.; Lee, D.D.; Mailen, J.C.
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 similar conventional UO? fuel. Weapons-grade MOX behavior. However, MOX fuel rods feature higher fuel centerline temperatures due to a lower thermal conductivity. Moreover, higher diffusion in MOX fuel results in a slightly higher fission gas release. Finally, MOX fuel shows better mechanical behavior than UO? fuel due to lower pellet-cladding mechanical interaction and rod deformation. These results indicate that the MOX fuel meets all potential licensing requirements.
Bellanger, Philippe
Science Conference Proceedings (OSTI)
This is the final progress report for the FastOS (Phase 2) (FastOS-2) project with Argonne National Laboratory and the University of Oregon (UO). The project started at UO on July 1, 2008 and ran until April 30, 2010, at which time a six-month no-cost extension began. The FastOS-2 work at UO delivered excellent results in all research work areas: * scalable parallel monitoring * kernel-level performance measurement * parallel I/0 system measurement * large-scale and hybrid application performance measurement * onlne scalable performance data reduction and analysis * binary instrumentation
Allen D. Malony; Sameer Shende
Activities in a ptogram concerned with development of plasma-jet spray- coating techniques suitable for production of clad ceramic fuel plates are described. Experiments on application of zirconia coatings are also described. A survey of UO/sub 2/ powder was conducted to determine its suitability for plasma spraying. Also conditions were established for spraying fused and milled UO/sub 2/. The effects of process variables on coating and deposition characteristics were found to correlate. Densities of UO/sub 2/ coatings of 75 to 80% were achieved. (J.R.D.)
Weare, N.E.
Electrolytic process for preparing uranium metal
An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.
Haas, Paul A. (Knoxville, TN)
Production and Handling Slide 17: Yellow Cake, Uranyl Nitrate...
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, UF4 Skip Presentation Navigation First Slide Previous Slide Next Slide Last Presentation Table of Contents Yellow Cake, Uranyl Nitrate, ADU, UO2, UF4 Refer to caption below for...
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(breach) occurs and the depleted UF6 is exposed to water vapor in the air, uranyl fluoride (UO2F2) and hydrogen fluoride (HF) are formed. The uranyl fluoride is a solid that...
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such as water vapor in the air, the UF6 and water react, forming corrosive hydrogen fluoride (HF) and a uranium-fluoride compound called uranyl fluoride (UO2F2). For more...
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For inhalation or ingestion of soluble or moderately soluble compounds such as uranyl fluoride (UO2F2) or uranium tetrafluoride (UF4), the uranium enters the bloodstream and...
FAQ 28-What are the potential health effects from uranium exposure...
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For inhalation or ingestion of soluble or moderately soluble compounds such as uranyl fluoride (UO2F2) or uranium tetrafluoride (UF4), the uranium enters the bloodstream and...
FAQ 21-What happens if a cylinder of uranium hexafluoride leaks...
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(breach) occurs and the depleted UF6 is exposed to water vapor in the air, uranyl fluoride (UO2F2) and hydrogen fluoride (HF) are formed. The uranyl fluoride is a solid that...
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such as water vapor in the air, the UF6 and water react, forming corrosive hydrogen fluoride (HF) and a uranium-fluoride compound called uranyl fluoride (UO2F2). For this reason,...
Materials Models and Simulation for Nuclear Fuels Workshop
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DIF, France), Andersson, D. (LANL, USA), Stanek, C. (LANL, USA) 9:40-10:00 Monte Carlo simulation of phonon transport in UO2 crystals with defects El-Azab, Anter (Purdue, USA),...
Science Conference Proceedings (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
shrinkage cracking, (b) fission product attack of SiC, and (in a pebble and pebbles in a bed in a hybrid LIFE fissionDiffusion coefficients of fission product species in UO 2 as
Powers, Jeffrey
DEVELOPMENT OF CLAD CERAMIC FUEL PLATES BY SPRAY-COATING TECHNIQUES. Final Report, Phase I
Activities in a program to develop techniques of plasma spraying clad plate-type UO/sub 2/ fuel elements are reported. The investigation was also directed toward determining the limitations of the process as applied to fuel element fabrication. UO/sub 2/ powder coatings having densities of 90% theoretical were produced. At conditions required for spraying plates, densities of 86% appear to be practical. The rate and efficiency of UO/sub 2/ coating deposition were also determined for various spraying conditions. Gritblasting was found to provide the best surface for coating adherence. The O/U ratio of the UO/sub 2/ was maintained by spraying in an Ar atmosphere. Zircaloy-2 was found to be the most desirable cladding material. Cladding thicknesses of 0.035 in. are required in distortion-free 2-in.-wide plates. (J.R.D.)
Weare, N.E.
Science Conference Proceedings (OSTI)
The characteristics of the two-dimensional steady state flow of unbounded stratified Boussinesq fluid over an isolated obstacle of finite height are analyzed for the simplqst case in which the incident flow speed, UO, and Brunt-Vaisala frequency, ...
R. Laprise; W. R. Peltier
OF CARBON I N PHOTOSYNTHESIS, X U , KINETIC REIATIORSEIPS OFof steady. state photosynthesis in cUO2 provides informationThe path of carbon i n photosynthesis begins with a small
Benson, A.A.; Kawaguchi, S.; Hayes, P.; Calvin, M.
Electrochemistry of LiCl-Li2O-H2O Molten Salt Systems
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Abstract Scope, Uranium can be recovered from uranium oxide (UO2) spent fuel through the ... Cathodic Behavior of Silicon (?) in BaF2-CaF2 �SiO2 Melts.
Fabrication of Tungsten, Tungsten-Rhenium and Tungsten-CeO2 ...
Science Conference Proceedings (OSTI)
... kernels such as uranium dioxide (UO2) are encapsulated within a tungsten matrix. ... in the pre-exponential frequency factor for gas diffusion through such materials. ... Functional Composites: Fluorescent Carbon Nanotubes in Silica Aerogel.
Peer Review of Strategy for Characterizing Contamination in DUF6...
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plant (GDP) sites. Therefore, the assumption in Smith 1984 that 25percent (4.6 kilograms) of the neptunium received (18.4 kilograms) in the UO 3 will enter the cascade,...
Rational Ligand Design for U(VI) and Pu(IV)
1.6 Ligand Design for Pu(IV) �. ��..���������. ���������ligands used in UO 22+ and Pu(IV) structural studies ��.. 23Raymond group ligands for Pu(IV) decorporation �� 208
Szigethy, Geza
Evaluation of alternative fuel cycle strategies for nuclear power generation in the 21st century
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
The Properties of Spent Nuclear Fuel under Waste Disposal ...
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Symposium, Materials Issues in Nuclear Waste Management in the 21st Century ... UO2 in the form of a ceramic pellet with a density close to theoretical. ... On discharge fro reactor the pellets have undergone a number of physical and�...
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to Prevent Uraninite (UO 2 ) Oxidation Uranium (U) is one of the most prevalent radionuclide contaminants in soils and groundwater across the world as a result of nuclear fuel...
Nanoparticulate FeS as an Effective Redox Buffer to Prevent Uraninite...
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(UO2) Oxidation Saturday, August 31, 2013 Uranium (U) is one of the most prevalent radionuclide contaminants in soils and groundwater across the world as a result of nuclear fuel...
In-Situ Repairs of Oil Industry Pipelines, Tanks and Vessels by ...
Science Conference Proceedings (OSTI)
Abstract Scope, Metal arc welding under oil (MAW-UO) is a new, revolutionary process to repair a pipeline, tank or vessel by welding in case of flaws and�...
of the thermodynamic quantities of uranium(VI)�carboxylateComplexation of Uranium(VI) by Gluconate Thermodynamic Studyacid (H A ) Hexavalent uranium as the UO 22+ ion was studied
Zhang, Zhicheng
A Compendium of Radiocarbon Dates for Southern Idaho Archaeological Sites
Archaeology of the Shoup Rockshelters in East Central Idaho.PocateUo: Idaho State UniversUy Museum Occasional Papers No.Snake River Region of Idaho ca. 4150 B.P. - 1300 B.P.
Plew, Mark G; Pavesic, Max G
Materials for Nuclear Power: Digital Resource Center - REPORT ...
Science Conference Proceedings (OSTI)
Jun 27, 2007... Fate of the Epsilon Phase in UO2 of the Oklo Natural Fisson Reactors ... In spent nuclear fuel (SNF), the micron- to nano-sized epsilon phase�...
SLAC National Accelerator Laboratory: SLAC Science Focus Area...
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Clark, S D. Conradson, and J.R. Bargar (2008) Structure of biogenic UO2 produced by Shewanella Oneidensis, strain MR-1. Environ. Sci. Technol., 42, 7898-7904. K.-U. Ulrich, D....
Energy.gov (U.S. Department of Energy (DOE))
To develop mechanistic models for fuel thermal conductivity, the Fuel team used supercells up to 55 nm long to determine the thermal conductivity of UO2 with Xe incorporated.
SLAC National Accelerator Laboratory: SLAC Science Focus Area...
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program. The inner portion of the particles is well ordered and similar to stoichiometric or near-stoichiometric UO2.0, and the material consequently exhibits a solubility...
Radiation Effects in Oxide Ceramics and Novel LWR Fuels
Science Conference Proceedings (OSTI)
Nuclear fuels, such as uranium dioxide (UO2) and Mixed Oxide (MOX) fuels, have been used in current light water reactors (LWRs) to produce about 15% of the ... of oxide ceramics for nuclear applications through experiment, theory and�...
This patent relates to the preparation of high-density, spherical UO/sub 2/ particles 80 to 150 microns in diameter. Sinterable UO/sub 2/ powder is wetted with 3 to 5 weight per cent water and tumbled for at least 48 hours. The resulting spherical particles are then sintered. The sintered particles are useful in dispersion-type fuel elements for nuclear reactors. (AEC)
Levey, R.P. Jr.; Smith, A.E.
Nuclear materials research progress reports for 1979
Research is presented concerning iodide stress corrosion cracking of zircaloy, self-diffusion of oxygen in hypostoichiometric urania, surface chemistry of epitaxial silicon deposition by thermal cracking of silane, kinetics of laser pulse vaporization of UO/sub 2/, gas laser model for laser induced evaporation, solubility of hydrogen in uranium dioxide, thermal gradient migration of metallic inclusions in UO/sub 2/, molecular beam studies of atomic hydrogen reduction of oxides, and thermal gradient brine-inclusion migration in salt. (FS)
Olander, D.R.
The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: � Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, � Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, � Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, � Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and � Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.
Mac Donald, Philip Elsworth
Science Conference Proceedings (OSTI)
From density functional theory (DFT) based ab initio (Car-Parrinello) metadynamics, we compute the activation energies and mechanisms of water exchange between the first and second hydration shells of aqueous Uranyl (UO{sub 2}{sup 2+}) using the primary hydration number of U as the reaction coordinate. The free energy and activation barrier of the water dissociation reaction [UO{sub 2}(OH{sub 2}){sub 5}]{sup 2+}(aq) {yields} [UO{sub 2}(OH{sub 2})4]{sup 2+}(aq) + H{sub 2}O are 0.7 kcal and 4.7 kcal/mol respectively. The free energy is in good agreement with previous theoretical (-2.7 to +1.2 kcal/mol) and experimental (0.5 to 2.2 kcal/mol) data. The associative reaction [UO{sub 2}(OH{sub 2}){sub 5}]{sup 2+}(aq) + H{sub 2}O {yields} [UO{sub 2}(OH{sub 2})6]{sup 2+}(aq) is short-lived with a free energy and activation barrier of +7.9 kcal/mol and +8.9 kca/mol respectively; it is therefore classified as associative-interchange. On the basis of the free energy differences and activation barriers, we predict that the dominant exchange mechanism between [UO{sub 2}(OH{sub 2}){sub 5}]{sup 2+}(aq) and bulk water is dissociative.
Atta-Fynn, Raymond; Bylaska, Eric J.; De Jong, Wibe A.
Hydrogen retention and release from uranium dioxide
The ceramic samples (UO/sub 2/) are exposed to high pressure hydrogen gas at a fixed temperature for a time sufficient to achieve equilibrium. After rapid quenching, the hydrogen-saturated sample is transferred to a vacuum-outgassing furnace. The sample is outgassed in a linear temperature ramp and the released hydrogen is detected by an in-situ mass spectrometer. This technique measures the rate of release of hydrogen with a sensitivity level of about 2 ng of hydrogen (as D/sub 2/) per hour. In this study, experiments were conducted on both polycrystalline and single-crystal UO/sub 2/. Experimental variables included temperature (1000 to 1600/sup 0/C) and infusion pressure (5 to 32 atm D/sub 2/), and for the polycrystalline specimen, stoichiometry. Dissolution of H/sub 2/ in both single-crystal and polycrystalline UO/sub 2/ was found to obey Seivert's law. The Sievert's law constant of deuterium in single-crystal UO/sub 2/ was determined to be: 3.0 x 10/sup 7/exp(-235 kJ/RT) ppM atomic/..sqrt..atm and for polycrystalline UO/sub 2/: 5.5 x 10/sup 4/exp(-100 kJ/RT) ppM atomic/..sqrt..atm. The solubility of hydrogen in hypostoichiometric urania was found to be up to three orders of magnitude greater than in stoichiometric UO/sub 2/ depending on the O/U ratios, implying the anion vacancy is the primary solution site in the UO/sub 2/ lattice. The release-rate curves for the single crystal and polycrystalline UO/sub 2/ specimens exhibited multiple peaks, with most of the deuterium released between 600 and 1200/sup 0/C for the polycrystalline samples, and between 700 and 1800/sup 0/C in the single-crystal specimens. This release of hydrogen from UO/sub 2/ could not be adequately modeled as diffusion or diffusion with trapping and resolution. It was determined that release was governed by release from traps in both the polycrystalline and single crystal UO/sub 2/ specimens. 40 refs., 72 figs., 6 tabs.
Sherman, D.F.
Roasting and dissolution studies on nonirradiated thorium dioxide/uranium dioxide pellets
Science Conference Proceedings (OSTI)
Bench scale roasting and dissolution of ThO/sub 2/ and ThO/sub 2//UO/sub 2/ ractor-grade ceramic pellets were studied at the Savannah River Laboratory to define the key parameters affecting dissolution. Pellet breakup, and subsequent dissolution rates, were determined for ThO/sub 2/ and ThO/sub 2//UO/sub 2/ pellets roasted in air or in oxygen. Roasting ThO/sub 2//UO/sub 2/ pellets in air at temperatures from 400 to 900/sup 0/C caused the pellets to crack but not fragment. Roasting whole pellets or fine powdered materials decreased the rate of dissolution in a nitric acid solution containing a fluoride catalyst. Roasting 100% ThO/sub 2/ pellets did not cause cracking or affect the subsequent dissolution rate. Mixed ThO/sub 2//UO/sub 2/ ceramic pellets dissolved at a faster rate than the 100% ThO/sub 2/ pellets. The effect of MgO and CaO on dissolution rate was determined. MgO (approx. 1.0 wt %) increased the dissolution rate of ThO/sub 2/ pellets, an effect which was similar to that obtained by the addition of 20% UO/sub 2/ to the ThO/sub 2/ pellets. The combination of 1% MgO and 20% UO/sub 2/ did not result in an additional increase in dissolution rate. However, the addition of 0.25 to 0.50 wt % CaO did increase the dissolution rate of 80% ThO/sub 2//UO/sub 2/ ceramic pellets. High temperatures (and pressure) were ineffective in dissolving thoria-based fuels in HNO/sub 3/ in the absence of a fluoride catalyst. A process flowsheet outlining the required head end steps for the reprocessing of thoria-based fuels was developed.
Pickett, J.B.; Fowler, J.R.; Mosley, W.C. Jr.
The burnup dependence of light water reactor spent fuel oxidation
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.
1.1 This specification covers blended uranium trioxide (UO3), U3O8, or mixtures of the two, powders that are intended for conversion into a sinterable uranium dioxide (UO2) powder by means of a direct reduction process. The UO2 powder product of the reduction process must meet the requirements of Specification C 753 and be suitable for subsequent UO2 pellet fabrication by pressing and sintering methods. This specification applies to uranium oxides with a 235U enrichment less than 5 %. 1.2 This specification includes chemical, physical, and test method requirements for uranium oxide powders as they relate to the suitability of the powder for storage, transportation, and direct reduction to UO2 powder. This specification is applicable to uranium oxide powders for such use from any source. 1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of th...
American Society for Testing and Materials. Philadelphia
Surface Decontamination of System Components in Uranium Conversion Plant at KAERI
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.
Thermophysical properties of uranium dioxide - Version 0 for peer review
Science Conference Proceedings (OSTI)
Data on thermophysical properties of solid and liquid UO{sub 2} have been reviewed and critically assessed to obtain consistent thermophysical property recommendations for inclusion in the International Nuclear Safety Center Database on the World Wide Web (http://www.insc.anl.gov.). Thermodynamic properties that have been assessed are enthalpy, heat capacity, melting point, enthalpy of fusion, thermal expansion, density, surface tension, and vapor pressure. Transport properties that have been assessed are thermal conductivity, thermal diffusivity, viscosity, and emissivity. Summaries of the recommendations with uncertainties and detailed assessments for each property are included in this report and in the International Nuclear Safety Center Database for peer review. The assessments includes a review of the experiments and data, an examination of previous recommendations, the basis for selecting recommendations, a determination of uncertainties, and a comparison of recommendations with data and with previous recommendations. New data and research that have led to new recommendations include thermal expansion and density measurements of solid and liquid UO{sub 2}, derivation of physically-based equations for the thermal conductivity of solid UO{sub 2}, measurements of the heat capacity of liquid UO{sub 2}, and measurements and analysis of the thermal conductivity of liquid UO{sub 2}.
Fink, J.K.; Petri, M.C.
Uranium oxides in a molten eutectic mixture of NaClKCl were chlorinated by bubbling chlorine gas through the mixture. The reaction product, uranyl chloride. was soluble in the molten salt. Although UO/sub 2/ was the most common oxide used, the reaction was similar in the other oxides. Phosgene and aluminum chloride were also used as chlorinating agents. A dense, crystalline precipitate of pure UO/sub 2/ was prepared by the reduction of the uranyl chloride contained in the molten salt solution. The reduction was accomplished by contacting the salt solution with any of several metals, by reaction with hydrogen or dry ammonia gas, or by electrolysis. Several kilograms of UO/sub 2/ were prepared by electrolysis using graphite electrodes. The physical properties of the material made it potentially useful as a ceramic fuel material. The initial high particle density of the "as-produced" UO/sub 2/ was considered of great potential advantage for adapting this process to the refabrication of irradiated UO/sub 2/ into recycle fuel elements. (M.C.G.)
Lyon, W.L.; Voiland, E.E.
Molecular uranates - laser synthesis of uranium oxide anions in the gas phase
Science Conference Proceedings (OSTI)
Laser ablation of solid UO{sub 3} or (NH{sub 4}){sub 2}U{sub 2}O{sub 7} yielded in the gas phase molecular uranium oxide anions with compositions ranging from [UO{sub n}]{sup -} (n = 2-4) to [U{sub 14}O{sub n}]{sup -} (n = 32-35), as detected by Fourier transform ion cyclotron resonance mass spectrometry. The cluster series [U{sub x}O{sub 3x}]{sup -} for x {le} 6 and various [U{sub x}O{sub 3x-y}]{sup -}, in which y increased with increasing x, could be identified. A few anions with H atoms were also present, and their abundance increased when hydrated UO{sub 3} was used in place of anhydrous UO{sub 3}. Collision-induced dissociation experiments with some of the lower m/z cluster anions supported extended structures in which neutral UO{sub 3} constitutes the building block. Cationic uranium oxide clusters [U{sub x}O{sub n}]{sup +} (x = 2-9; n = 3-24) could also be produced and are briefly discussed. Common trends in the O/U ratios for both negative and positive clusters could be unveiled.
Marcalo, Joaquim; Santos, Marta; Pires de Matos, Antonio; Gibson, John K
1.1 This specification covers blended uranium trioxide (UO3), U3O8, or mixtures of the two, powders that are intended for conversion into a sinterable uranium dioxide (UO2) powder by means of a direct reduction process. The UO2 powder product of the reduction process must meet the requirements of Specification C 753 and be suitable for subsequent UO2 pellet fabrication by pressing and sintering methods. This specification applies to uranium oxides with a 235U enrichment less than 5 %. 1.2 This specification includes chemical, physical, and test method requirements for uranium oxide powders as they relate to the suitability of the powder for storage, transportation, and direct reduction to UO2 powder. This specification is applicable to uranium oxide powders for such use from any source. 1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of th...
American Society for Testing and Materials. Philadelphia
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
The activities of the Technical Division, including the HNO/sub 3/ recovery process, a study of UO/sub 3/ factors as related to the production of metal-grade UF/sub 4/, a thermobalance investigation of starch as a reducing agent for UO/sub 3/, reduction of UF/sub 4/ to uranium by a thermite type reaction, melting and casting of Nb-U alloy, uranium recovery from scrap materials, preparation of uranium shot, cal cium reduction of ThO/sub 2/ production of thorium ingots, "wet chemical" and spectrochemical development, ammonia precipitation and filtration studies from uranyl nitrate solutions and preparation of active UO/sub 2/ from UF/sub 6/ are reviewed. (W.L.H.)
Simmons, J.W. ed.
DEVELOPMENT OF PLUTONIUM-BEARING FUEL MATERIALS. Monthly Progress Letter for Month of January 1962
Development work leading to comparison of co-precipitated and mechanically blended UO/sub 2/-- PuO/sub 2? powders is reported in which special emphasis was placed on blending trials, pellet sintering studies, and subsequent evaluation of pellets made with blended material. Homogeneity studies indicate that currently used procedures are unsatisfactory because particle buildup occurs during blending. Powder preparation via the oxalate route was continued along with PuO/sub 2/ moisture pickup studies. Homogeneous precipitation studies on UO/ sub 2/ were continued to determine feasibility of direct preparation of dense UO/ sub 2/-- PuO/sub 2/ feed materials. Plasma-torch-produced PuO/sub 2/ spheres are being evaluated. (J.R.D.)
Puechl, K.H.
Kinetics of laser pulse vaporization of uranium dioxide by mass spectrometry
Science Conference Proceedings (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.
Greening academia: Developing sustainable waste management at Higher Education Institutions
Science Conference Proceedings (OSTI)
Higher Education Institutions (HEIs) are often the size of small municipalities. Worldwide, the higher education (HE) sector has expanded phenomenally; for example, since the 1960s, the United Kingdom (UK) HE system has expanded sixfold to >2.4 million students. As a consequence, the overall production of waste at HEIs throughout the world is very large and presents significant challenges as the associated legislative, economic and environmental pressures can be difficult to control and manage. This paper critically reviews why sustainable waste management has become a key issue for the worldwide HE sector to address and describes some of the benefits, barriers, practical and logistical problems. As a practical illustration of some of the issues and problems, the four-phase waste management strategy developed over 15 years by one of the largest universities in Southern England - the University of Southampton (UoS) - is outlined as a case study. The UoS is committed to protecting the environment by developing practices that are safe, sustainable and environmentally friendly and has developed a practical, staged approach to manage waste in an increasingly sustainable fashion. At each stage, the approach taken to the development of infrastructure (I), service provision (S) and behavior change (B) is explained, taking into account the Political, Economic, Social, Technological, Legal and Environmental (PESTLE) factors. Signposts to lessons learned, good practice and useful resources that other institutions - both nationally and internationally - can access are provided. As a result of the strategy developed at the UoS, from 2004 to 2008 waste costs fell by around Pounds 125k and a recycling rate of 72% was achieved. The holistic approach taken - recognizing the PESTLE factors and the importance of a concerted ISB approach - provides a realistic, successful and practical example for other institutions wishing to effectively and sustainably manage their waste.
Zhang, N. [School of Civil Engineering and the Environment, University of Southampton, University Rd., Highfield, Southampton, Hampshire SO17 1BJ (United Kingdom); Williams, I.D., E-mail: email@example.com [School of Civil Engineering and the Environment, University of Southampton, University Rd., Highfield, Southampton, Hampshire SO17 1BJ (United Kingdom); Kemp, S. [School of Civil Engineering and the Environment, University of Southampton, University Rd., Highfield, Southampton, Hampshire SO17 1BJ (United Kingdom); Smith, N.F. [Estates and Facilities Management, University of Southampton, University Rd., Highfield, Southampton, Hampshire SO17 1BJ (United Kingdom)
An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO{sub 2}-Al dispersion fuel, and with LEU (19.75%)UO{sub 2}-Al dispersion fuel that has a uranium density of about 2.5 g/cm{sup 3}. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.
Pond, R. B.; Hanan, N. A.; Matos, J. E.; Maraczy, C.
The design features and the performance of UO/sub 2/SO/sub 4/ blanket processing Loop P-1 are evaluated from an engineering viewpoint. This unique experiment development loop was operated with pump heating to study the behavior of plutonium in 1.4 M UO/sub 2/SO/sub 4/at 250 deg C and was designed for mixed O/ sub 2/-H/sub 2/ gas pressurization The canned loop and the feed and sampling systems in glove boxes completely contained the plutonium throughout the experimental program. (auth)
Snider, J.W.; Clinton, S.D.
Variation of stability constants of thorium and uranium oxalate complexes with ionic strength
Extraction of Th(IV) and UO{sub 2}{sup 2+} by a solution of TTA and HDEHP, respectively, in toluene was used to obtain stability constants of their oxalate complexes in 1, 3, 5, 7 and 9 M ionic strength (NaClO{sub 2}) solutions. The complexes formed were the MOx, MHOx, MOx{sub 2} and M(HOx){sub 2} (M = Th, UO{sub 2}) species. The values were analyzed by the Specific Interaction Theory and agreed to I {le} 3 M but required an additional term for fitting at I > 3 M.
Erten, H.N; Mohammed, A.K.; Choppin, G.R. [Florida State Univ., Tallahassee, FL (United States). Dept. of Chemistry
Preparation of thorium-uranium gel spheres
Ceramic oxide spheres with diameters of 15 to 1500 ..mu..m are being evaluated for fabrication of power reactor fuel rods. (Th,U)O/sub 2/ spheres can be prepared by internal or external chemical gelation of nitrate solutions or oxide sols. Two established external gelation techniques were tested but proved to be unsatisfactory for the intended application. Established internal gelation techniques for UO/sub 2/ spheres were applied with minor modifications to make 75% ThO/sub 2/-25% UO/sub 2/ spheres that sinter to diameters of 200 to 1400 ..mu..m (99% T.D.).
Spence, R.D.; Haas, P.A.
The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied. Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)
Progress in the design, development, and construction of PWR power plant systems and components and PWR core and auxiliaries is summarized. The blanket assembly design is described and illustrated. Results of MTR evaluation of fuel element failure instrumentation are reported. Development of fabrication and testing tochaiques for clad fuel elements, fuel rods, plates, and assemblies is described. Investigations of fuel and cladding alloys include crystal structure and thermal stability determinations on U--Mo alloys, studies on the nature of the hydride phase formed during corrosion of gamma -phase alloys in high- temperature water, and specific heat, resistivity, and phase diagram studies of U- -Mo and U--Nb alloys. The equilibrium and kinetics in the system UO/sub 2/--O/ sub 2/ are being studied to gain information on the structure and stability of UO/ sub 2/ under various conditions. Results of irradiation tests on UO/sub 2/ samples and of thermal cycling tests of Zircaloy-2 clad UO/sub 2/ rods are reported. Corrosion test resuIts are summarized for unclad and Zircaloy-2 clad U- - Mo and U--Nb samples. The radiation induced volume change of prototype fuel reds has been investigated. Studies of the fabrication cladding, tensile properties, and corrosion of U-- Si systems are described. Corrosion tests are continuing on Zircaloy-2 clad U-- Zr fuel elements and on various experimental Al alloys for cladding applications. Work was continued on the preparation, corrosion and sinterability of pure UO/sub 2/ and UO/sub 2/ containing additives. Operation and chemical analysis of in-pile loop experiments are described. Results are reported from studies of the erosion of UO/sub 2/ in high-velocity coolant, decontamination of water by ion exchange resins, sorption of radioisotopes on stainless steel, and decontamination of corrosion loops. Work in reactor physics has included PWR control calculations using a 2-dimensional UNIVAC code, calculation of fission product activity in the primary coolant, and criticaiity studies on the Flexibie Critical Experiment and on a lattice of UO/ sub 2/ fuel reds in the TRX. Current PWR plant parameters are recapitulated. (D.E.B.)
Effects of Time, Heat, and Oxygen on K Basin Sludge Agglomeration, Strength, and Solids Volume
Sludge disposition will be managed in two phases under the K Basin Sludge Treatment Project. The first phase is to retrieve the sludge that currently resides in engineered containers in the K West (KW) Basin pool at ~10 to 18�C. The second phase is to retrieve the sludge from interim storage in the sludge transport and storage containers (STSCs) and treat and package it in preparation for eventual shipment to the Waste Isolation Pilot Plant. The work described in this report was conducted to gain insight into how sludge may change during long-term containerized storage in the STSCs. To accelerate potential physical and chemical changes, the tests were performed at temperatures and oxygen partial pressures significantly greater than those expected in the T Plant canyon cells where the STSCs will be stored. Tests were conducted to determine the effects of 50�C oxygenated water exposure on settled quiescent uraninite (UO2) slurry and a full simulant of KW containerized sludge to determine the effects of oxygen and heat on the composition and mechanical properties of sludge. Shear-strength measurements by vane rheometry also were conducted for UO2 slurry, mixtures of UO2 and metaschoepite (UO3�2H2O), and for simulated KW containerized sludge. The results from these tests and related previous tests are compared to determine whether the settled solids in the K Basin sludge materials change in volume because of oxidation of UO2 by dissolved atmospheric oxygen to form metaschoepite. The test results also are compared to determine if heating or other factors alter sludge volumes and to determine the effects of sludge composition and settling times on sludge shear strength. It has been estimated that the sludge volume will increase with time because of a uranium metal ? uraninite ? metaschoepite oxidation sequence. This increase could increase the number of containers required for storage and increase overall costs of sludge management activities. However, the volume might decrease because of decreases in the water-volume fraction caused by sludge solid reactions, compaction, or intergrowth and recrystallization of metaschoepite. In that case, fewer STSCs may be needed, but the shear strength would increase, and this could challenge recovery by water jet erosion and require more aggressive retrieval methods. Overall, the tests described herein indicate that the settled solids volume remains the same or decreases with time. The only case for which the sludge solids volumes increase with time is for the expansion factor attendant upon the anoxic corrosion of uranium metal to produce UO2 and subsequent reaction with oxygen to form equimolar UO2.25 and UO3�2H2O.
Delegard, Calvin H.; Sinkov, Sergey I.; Schmidt, Andrew J.; Daniel, Richard C.; Burns, Carolyn A.
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Pyrolytic carbon-coated nuclear fuel
An improved nuclear fuel kernel having at least one pyrolytic carbon coating and a silicon carbon layer is provided in which extensive interaction of fission product lanthanides with the silicon carbon layer is avoided by providing sufficient UO.sub.2 to maintain the lanthanides as oxides during in-reactor use of said fuel.
Lindemer, Terrence B. (Oak Ridge, TN); Long, Jr., Ernest L. (Oak Ridge, TN); Beatty, Ronald L. (Wurlingen, CH)
Water-Moderated and -Reflected Slabs of Uranium Oxyfluoride
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
are Pressurized Heavy Water Reactors (PHWR) otherwise known as CANDU2 , a similar Russian design called Reaktor reactors are commonly called CANDU reactors. #12;28 A BWR is a simpler design than a PWR, but it exposes, Japan, Sweden 94 86.4 enriched UO2 water water Pressurised Heavy Water Reactor 'CANDU' (PHWR) Canada 43
Haykin, Simon
Table of Contents Page 1 NERC 2007 Long-Term Reliability Assessment
are Pressurized Heavy Water Reactors (PHWR) otherwise known as CANDU2 , a similar Russian design called Reaktor reactors are commonly called CANDU reactors. #12;28 A BWR is a simpler design than a PWR, but it exposes, Japan, Sweden 94 86.4 enriched UO2 water water Pressurised Heavy Water Reactor 'CANDU' (PHWR) Canada 43
PL2ESIDCNI The District Engineer, U. S. Engineer Office,
Office of Legacy Management (LM)
lb., f.o.b. East Liverpool, Ohio. 500 lb. at 2.15 Fer lb., f.o.b. Los iingel.es, California. SCgIttd UR:lAT?7 (Approxi.utely 84" U,Os) : 1485 lb. at 1.55 per lb., f.o.b....
Studies were made on the effects of spray-coating variables on coating characteristics in the development of plasma-jet spraying techniques for making clad UO/sub 2/ fuel plates. UO/sub 2/ coatings of up to 90% theoretical density and - O/U ratios of nearly 2.00 were deposited at efficiencies of 40%. Adherent UO/sub 2/ coatings up to 0.100 inch thick can be deposited on 0.030-inch thick stainless steel and Zircaloy-2 substrates. Studies of coated composite bends and coating adherence at room temperature indicate that, for best results, the coating temperature should be maintained below 870 deg C and the substrate below 450 deg C during deposition. A plasma spray torch was tested for spraying UO/sub 2/ at 40 kw and found to be equivalent to operation at 25 kw. A preliminary cost analysis indicated considerably lower fabrication costs using plasma jet sprayingn ~ 0/kg U as compared to ~ 0/kg U for oxide pellet-in-tube elements. (D.L.C.)
Weare, N.E.; Buchanan, E.; Marchandise, H.
Kheshbn No. 147 - Spring 2006 - Journal
DK" - .pNT ^KD pra tsu TO iy Diy^a - ." pnt^yDi IOO^KII ,i^ONT ^yoa'^ayao'ra ]x iy 0"Diy oxra iy DIO VN� s D'o .p' ? npun piraya piKn pK .f?" rnKO Diy TO -]KA f? yn TO uoKnya /
The Development of Models to Optimize Selection of Nuclear Fuels through Atomic-Level Simulation
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
Nuclear Energy: Where do we go from here? Keith Bradley
Alamos National Laboratories, studying the physics of nuclear weapons, technology development for nuclear Energy Summit 9 "In comparison with the standard UO2+Zircaloy system, develop a fuel/clad system Modular Reactor Ltd. (PBMR)] #12;Where can experimentation help? § Development of new
Levi, Anthony F. J.
process efficiency (UoK, GA) · Estimate the size and cost of the process equipment (All) #12;s NERI H2 6 cycle analysis (SNL) · Develop detailed chemical flowsheet for selected process and determine projected UT-3 process is conceptually simple. . . l Invented at Univ. of Tokyo, being pursued in Japan, SI
VII. SOLAR RADIATION DATA COMPARISONS In this section some of the solar radiation data
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
Process design information concerning equipment and operation of a plant to produce UF/sub 4/ from UO/sub 3/ is presented. Included are process flow diagrams, drawings of ventilation and dust control systems, and vent gas systenas. Equipment lists and estimated utilities are also included as well as a description of the process. (J.R.D.)
Holby, G.V.; James, F.
Microsoft PowerPoint - 2012IPRC_BDD_Park
NLE Websites -- All DOE Office Websites (Extended Search)
Cathode: 4Li + + 4e - 4Li UO 2 + 4Li U + 2Li 2 O 650 o C, 1 wt% Li 2 O-LiCl molten salt 4 O 2 at anode Stable in LiCl CO 2 at anode Stable in LiCl Replacement of...
Bioremediation of Uranium Plumes with Nano-scale
Bioremediation 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(IV) (UO2[s], uraninite) Anthropogenic · Release of mill tailings during uranium mining - Mobilization
Cushing, Jim. M.
214 Synthesis of ZnO Nanostructures and Their Influence on ...
Science Conference Proceedings (OSTI)
Our results suggest that synthesis method can be used to produce desired .... 107 Alignment of BN Nanosheets Using DC and Nano Pulse-Width Electric Field ... 129 Minimization of Diametric Tolerance of U3O8-added UO2 Annular Pellet .... 220 Microwave Processing of Polymer Coatings on Instrument-Grade Wood.
a full simulated HLW stream based upon 4:1 ratio of high burn up UO2/mixed oxide (HBU/MOX) fuel. EXPERIMENTAL A series of simplified simulated HLW glasses (based on the 4:1 HBU/MOX composition) were melted
Sheffield, University of
Some Electrical Properties of Ion-Implanted Urania--Part II: Preprint
As part of the U.S. Department of Energy's effort to evaluate the use of UO2 as a material for photovoltaic (e.g., solar cell) applications, single-crystal UO2 samples were characterized as to their electrical and electro-optical properties. Samples of UO2 were ion implanted with boron and sulfur dopants, as well as with boron and sulfur co-dopants, at the Ion Beam Materials Laboratory facility at Los Alamos National Laboratory. Activation energies for electrical conduction were measured to be from 0.13 to 0.26 eV, when temperatures varied from 180 to 450 K. Dark current was measured followed by light current under 1-sun illumination. In general, the dark and light currents were about an order of magnitude greater than those reported earlier for polycrystalline UO2. Optical and infrared absorption and transmission data were also obtained and are reported. Transmission data on the single-crystal samples revealed a complex structure that made it difficult to resolve a single optical bandgap.
von Roedern, B.; Meek, T. T.; Haire, M. J.
Optimal scaling of the ADMM algorithm for distributed quadratic ...
for Strategic Research (SSF), the Swedish Research Council (VR), a UoM Early Career .... ? > 0, there exists an R that does not change the constraint set in (2) and ..... a future work, we plan to extend the results to a broader class of distributed�...
The separation of uranium from an aqueous solution containing a water soluble uranyl salt is described. The process involves adding an alkali thiocyanate to the aqueous solution, contacting the resulting solution with methyl isobutyl ketons and separating the resulting aqueous and organic phase. The uranium is extracted in the organic phase as UO/sub 2/(SCN)/sub/.
McVey, W.H.; Reas, W.H.
Science Conference Proceedings (OSTI)
Uraninite (UO2) and metaschoepite (UO3�2H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21�C and 50�C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.004�0.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21�C than the particles prepared at 50�C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.
Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.
NLE Websites -- All DOE Office Websites (Extended Search)
ple c opy a lgorithms i n N AMD & f orce fi eld d evelopment Yun (Lyna) L uo PI: B enoit R oux Project G oal NAMD --- T he E ngine f or L arge---Scale C lassical M olecular...
Application of the active well coincidence counter to the measurement of uranium
Science Conference Proceedings (OSTI)
An Active Well Coincidence Counter has been developed to assay uranium fuel material in field inspection applications. The unit is used to measure bulk UO/sub 2/ samples, high enrichment uranium metals, LWR fuel pellets, and /sup 233/U-Th fuel materials which have very high gamma-ray backgrounds.
Menlove, H.O.; Foley, J.E.; Bosler, G.E.
Bi-annual report 2007-2008 Chemical Science and Engineering
nuclear fuel. A. Beyond the saturation: Co adsorbed on iron oxide nanoparticles During the last decade in simulated irradiated UO2 nuclear fuel. One of the main goals of research in nuclear energy is to improve the economic and safety performance of nuclear fuels. One solution is to extend its life time in reactor
Lagergren, Jens
Triangular Atomic Patterns as Key to the Chemistry of Nuclear Fuels (Poster 12) 16 Resolving Shock #12;16 Triangular Atomic Patterns as Key to the Chemistry of Nuclear Fuels (Poster 12) David Andersson dioxide, UO2, is typically the major constituent of current nuclear fuel. Understanding of its properties
Varela, Carlos
While the product of microbial uranium reduction is often reported to be�UO2�, a comprehensive characterization including stoichiometry and unit cell determination is available for only one Shewanella species. Here, we compare the products of batch uranyl reduction by a collection of dissimilatory metal- and sulfate-reducing bacteria of the genera Shewanella, Geobacter, Anaeromyxobacter, and Desulfovibrio under similar laboratory conditions. Our results demonstrate that U(VI) bioreduction by this assortment of commonly studied, environmentally relevant bacteria leads to the precipitation of uraninite with a composition between UO2.00 and UO2.075, regardless of phylogenetic or metabolic diversity. Coupled analyses, including electron microscopy, X-ray absorption spectroscopy, and powder diffraction, confirm that structurally and chemically analogous uraninite solids are produced. These biogenic uraninites have particle diameters of about 2-3 nm and lattice constants consistent with UO2.0 and exhibit a high degree of intermediate-range order. Results indicate that phylogenetic and metabolic variability within delta- and gamma-proteobacteria has little effect on nascent biouraninite structure or crystal size under the investigated conditions.
Sharp, Jonathan; Schofield, Eleanor J.; Veeramani, Harish; Suvorova, Elena; Kennedy, David W.; Marshall, Matthew J.; Mehta, Apurva; Bargar, John R.; Bernier-Latmani, Rizlan
Functional blockade of impulse trains caused by acute nerve compression
4UOJOB! P jo ia(wnu P. o4 anp aq ur,a jr,t4ua4od uopar,uo jaojq ains -said P. o4 anp st ssot aqa 8 4ug4 aztsag4od6q
Jewett, Don L
AEC PHOTOVOLTAIC TEST FACILITY FIRST YEAR TEST DATA James Krumsick
of Oregon Solar Radiation Lab 1274 University of Oregon Eugene, OR 97403-1274 e-mail: fev, the University of Oregon Solar Radiation Monitoring Lab (UO SRML) under a contract with the Energy Trust environmental conditions. The test facility consists of a 25 KW rooftop array separated into eight systems. Each
Oregon, University of
This patent describes a method of making uniform, spherical, nonpyrophoric UC. UO/sub 2/ and carbon are mixed in stoichiometric proportions and passed through a plasma flame of inert gas at 10,000 to 13,000 deg C. (AEC)
White, G.D.; O' Rourke, D.C.
Homogeneous Reactor Experiment. Final data were obtained on the effectiveness of Cu/sup ++/ as an internal recombiner catalyst for radiolytic gas. Homogeneous Reactor Test. Criticality calculations have been completed for three blanket configurations using D/sub 2/O. ThO/sub 2/ slurry, and unenriched UO/sub 2/SO/sub 4/ solutions. Investigations on the temperature coefficient of reactivity and kinetic studies with respect to core pressure rise were also undertaken. Startup and shutdown procedures will involve the addition and removal of fuel concentrate. Revised flow sheet and design data sheets are presented, including the reactor vessel assembly, fuel pressurizer, recombiner- condenser, outer dump tank condenser, and fuel solution evaporator. The physical properties of HRT fuel and blanket solution at 2000 psia are given. Present evidence indicates that the Xe poison level can be maintained below 1% by continuous stripping with only 2% of the decomposition gases which would be produced if no Cu/sup ++/ catalyst were present for recombination. Revised inlet and outlet valve designs for the HRT pulsafeeder system have resulted in improved performance. General Homogeneous Reactor Studies. Principal activities in engineering development consist of the design of several representative heat exchanger layouts and recombiner loops. An extensive body of data on the corrosion of stainless steel by UO/sub 2/SO/sub 4/ solutions has been accumulated. The corrosive effects of boiling HNO/sub 3/ on stainless steel, and the UO/sub 2/ SO/sub 4/ corrosion of Zircaloy were also investigated. Stress relief annealing at 1000 deg F improved the dynamic corrosion resistance of austenitic stainless steel welds corroded by UO/sub 2/SO/sub 4/. The dynamic corrosion of Zircaloy-2 and Ti by UO/sub 2/SO/sub 4/ showed no marked effect on the impact behavior of these materials. The yield of N/sub 2/ from irradiated Th(NO/sub 3/)/sub 4/ is considerably less than previous values reported for UO/sub 2/(NO/sub 3/)/sub 2/ solutions. A revised phase diagram of the UO/sub 2/F/sub 2/-H/sub 2/O system is presented. The preparation and corrosive properties of ThO/sub 2/ slurries were investigated, together with the abrasion characteristics on stainless steel and Ti. Considerations associated with chemical processing of fuel and blanket solutions, such as the solubility of rare-earth sulfates, are discussed. (T.R.H.)
McDuffie, H.F. comp.
Inhibited Release of Mobile Contaminants from Hanford Tank Residual Waste
Investigations of contaminant release from Hanford Site tank residual waste have indicated that in some cases certain contaminants of interest (Tc and Cr) exhibit inhibited release. The percentage of Tc that dissolved from residual waste from tanks 241-C-103, 241-C-106, 241-C-202, and 241-C-203 ranged from approximately 6% to 10%. The percent leachable Cr from residual waste from tanks C-103, C 202, and C-203 ranged from approximately 1.1% to 44%. Solid phase characterization results indicate that the recalcitrant forms of these contaminants are associated with iron oxides. X-ray absorption near edge structure analysis of Tc and Cr in residual waste indicates that these contaminants occur in Fe oxide particles as their lower, less soluble oxidation states [Tc(IV) and Cr(III)]. The form of these contaminants is likely as oxides or hydroxides incorporated within the structure of the Fe oxide. Leaching behavior of U from tank residual waste was studied using deionized water, and CaCO3 and Ca(OH)2 saturated solutions as leachants. The release behavior of U from tank residual waste is complex. Initial U concentrations in water and CaCO3 leachants are high due to residual amounts of the highly soluble U mineral cejkaite. As leaching and dilution occur NaUO2PO4 {center_dot} xH2O, Na2U2O7(am) and schoepite (or a similar phase) become the solubility controlling phases for U. In the case of the Ca(OH)2 leachant, U release from tank residual waste is dramatically reduced. Thermodynamic modeling indicates that the solubility of CaUO4(c) controls release of U from residual waste in the Ca(OH)2 leachants. It is assumed the solubility controlling phase is actually a hydrated version of CaUO4 with a variable water content ranging from CaUO4 to CaUO4 {center_dot} (H2O). The critically reviewed value for CaUO4(c) (log KSP0 = 15.94) produced good agreement with our experimental data for the Ca(OH)2 leachates.
Cantrell, Kirk J.; Heald, Steve M.; Arey, Bruce W.; Lindberg, Michael J.
MARMOT Enhanced | Department of Energy
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
u.s. DEPARTMENT OF ENERGY EERE PROJECT MANAG EMENT CENTER NEPA DETERl.VIINATION
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Homogeneous Reactor Test. Experiments and tests conducted preparatory to operation with unenriched UO//sub 2/SO/sub 4/ are reviewed. Remote maintenance practlces and operation at reduced pressures and temperatures were analyzed. A simplified freeze jackct design for remote operation wlil be used in the HRT system. A differential-temperature flowmeter was designed for use on feed and purge pumps. The preliminary design of the replacement pressure vessel is shown. Fuel and blanket feed and purge pump test using UO/sub 2/SO/sub 4/ solutions were continued. Performance and corrosion results are given. HRT controls and instrumentation status is reviewed. Flowsheets for the fuel and blanket chemical processing systems are shown. The construction and engineering test status of the fuel processing plant are presented. Reactor Design and Analysis. Prellminary design parameters indicate the feasibility of a 500 Mw homogeneous research reactor using UO/sub 2/SO/sub 4/. The fuel costs for homogeneous reactors whose fuel is processed on a batch rather than a continuous basis were analyzed. Tables of data were prepared indicating the effecta of fuel isotopic concentration, the use of Li//sub 2/SO/sub 4/ additives, fission product removal, reactor operating periods, and Fu recovery on per krv power cost. The effects of Xe/sup 135/ on reactor conditions following shutdown were determined for the case of a U/sup 235/O/sub 2/SO/sub 4/, D/sub 2/O, spheri cal reactor operating at 280 deg C, assuming no fission product removal. Engineering Development. Developmental studies of ThO/sub 2/ blanket slurries were continued. Experience with ThO/sub 2/ deposits in circulation loops is tabulated; and shear diagrams, friction factors, and heat transfer characteristics are plotted. Slurry blanket system operational tests indicate satisfactory operation up to 300 deg C. Corrosion and Matertals. Studies of the corrosive effects of UO/sub 2/SO/ sub 4/ on Zircaloy, stainless steel, and Al//sub 2/O/sub were continued. The appearance of liquid phases as a function of temperature in UO//sub 2/SO/sub 4/-- Li/sub 2/SO/sub 4/, UO/sub 2/SO /sub 4/-BeSO/sub 4/, and BeSO/sub 4/-- UO/sub 3/ systems is plotted. Additional observations of pump corrosion and performance in reactor blanket loops contalning a range of ThO/sub 2/ concentrations are reported. Further attempts were made to establish the effects of slurry particle size on corrosive attack rates. Results of metallographic examinations of in- pile corrosion specimens of Zr and Ti alloys and stainless steel are tabulated and discussed. The effects of high temperatures and welding on crystalline phase changes in Zr alloys were investigated. The crystailine phase changes in H pickup in Ti, Zr, Al-Ti-V alloy, and Zircaloy upon exposure in the recombiner loop were determined. The effects of aging and temperature on Zircaloy impact strength are plotted. Chemical Engineering Revelopment. Chemical and engineering studies associated wlth HRT fuel processing are reported. In the study of Pu-producer blanket chemistry, adsorption of Pu on metals, Pu behavior in UO/sub 2/SO/sub 4/ solutions at 250 deg C, and dissolutlon of corrosion product oxides were considered. Slurries of ThO/sub 2/-U0//sub 3/-MoO/sub 3/-H/ sub 2/O were prepared and irradiated. Radiation effects and gas recombination rates for this type slurry are repcrted. Methods of particle size control in Th and U oxide preparation, and the effects of additives on oxide sedimentation rates in slurries were investigated. Supporting Chemical Research. Studies of slurry particle preparation and suspension are reported. The methods used in separating Pa/sup 231/ from Mallinckrodt waste are reviewed. (D.E.B.)
Aeorsol Characterization from a Simulated HCDA
Environmental conditions simulating the HCDA on a reduced scale provided the following information: Aerosols resulting from the condensation of gaseous constituents without sodium generally comprise small, spherical particles (diameter 0.01 to 0.25 um) and branched chain-like structures. Aerosols resulting from the condensation of gaseous constituents with sodium generally comprise spherical, small (diameter 0.01 to 0.50 um) particles, with some branched chain-like structures and some agglomerating particles. Electron diffraction analyses identified actinide dioxides, the constituents of stainless steel, an oxide of sodium (Na2O), sodium uranates (Na3UO4 and Na4UO5), and a sodium plutonate compound (Na4PuO5). Initial solubility studies indicated that 12.1% of the plutonium-239 dissolved in distilled water when a mixed-oxide (Pu, U) stainless steel pellet was vaporized with sodium. Reaction products are controlled kinetically during cooling rather than by equilibrium thermodynamics.
Zanotelli, W. A.; Miller, G. D.; Johnson, E. W.
In a recent criticality analysis for an array of water-moderated packages containing highly enriched uranium, the 123-group cross-section library in the SCALE system was observed to have a nonconservative discrepancy of approximately 3 to 3.5% when compared with more recently developed libraries. A simple representative system of UO{sub 2}F{sub 2}-H{sub 2}O was used to identify that the problem results from a lack of resonance data for {sup 235}U. Only a single set of self-shielded cross sections, most likely corresponding to a water-moderated infinite dilute system, was provided with the original data. The UO{sub 2}F{sub 2}-H{sub 2}O study indicates that this limitation may cause nonconservative discrepancies as high as 5.5% for some water-moderated, highly enriched uranium systems. Characteristics of the systems where the discrepancy is evident are identified and discussed.
Parks, C.V.; Wright, R.Q.; Jordan, W.C. [Oak Ridge National Lab., TN (United States)
Fuel System Compatibility Issues for Prometheus-1
Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO{sub 2} as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO{sub 2}-based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined.
DC Noe; KB Gibbard; MH Krohn
The carbonate and organic leaching processes for the recovery of U from its ores are outlined. The Excer prccess (ion-exchange conversion and electrolytic reduction) and the Fluorox process (starch-- HF reaction) for the production of UF/sub 4/ from ore concentrate and depleted reactor fuels are described. The fluidized-bed process for UF/sub 4/ production from UO/sub 2/(NO/ sub 3/)/sub 2/ is also described. Methods for improving the reactivity of UO/sub 3/ and mechanical and thermal processes for increasing the density of UF/sub 4/ were investigated. Applications of fluoride volatility prccesses to feed materials are discussed. (C.W.H.)
The plant arrangement, component design, and the functions of various systems are described and illustrated. Relative estimated costs of the systems and major components are indicated. The reactor core is designed around requiremouts for 254 thermal megawatts, 950 deg F maximum sodium temperature, stainless steel clad graphite moderator blocks, and low enrichment (0.015 to 0.04 U/sup 235/) uranium fuel elements. The fuel cycle is described for the possible fuel elements. The fuel cost factors are discussed. Burn-up limitations encountered for metallic fuel in the SGR temperature range indicate UO/sub 2/ the more desirable choice. The estimated cost of electrical energy associated with the UO/sub 2/ fuel is given. (auth)
Olson, R.L.; Gerber, R.C.; Gordon, R.B.; Ross-Clunis, H.A.; Stolz, J.F.
Method for fluorination of uranium oxide
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)
U.S. Department of Energy Categorical Exclusion Determination Form
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
An Insulating Breakthrough | Advanced Photon Source
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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,
CX-008356: Categorical Exclusion Determination | Department of Energy
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Nuclear criticality safety evaluation of SRS 9971 shipping package
This evaluation is requested to revise the criticality evaluation used to generate Chapter 6 (Criticality Evaluation) of the Safety Analysis Report for Packaging (SARP) for shipment Of UO{sub 3} product from the Uranium Solidification Facility (USF) in the SRS 9971 shipping package. The pertinent document requesting this evaluation is included as Attachment I. The results of the evaluation are given in Attachment II which is written as Chapter 6 of a NRC format SARP.
Vescovi, P.J.
Use of Savannah River Site facilities for blend down of highly enriched uranium
Westinghouse Savannah River Company was asked to assess the use of existing Savannah River Site (SRS) facilities for the conversion of highly enriched uranium (HEU) to low enriched uranium (LEU). The purpose was to eliminate the weapons potential for such material. Blending HEU with existing supplies of depleted uranium (DU) would produce material with less than 5% U-235 content for use in commercial nuclear reactors. The request indicated that as much as 500 to 1,000 MT of HEU would be available for conversion over a 20-year period. Existing facilities at the SRS are capable of producing LEU in the form of uranium trioxide (UO{sub 3}) powder, uranyl nitrate [UO{sub 2}(NO{sub 3}){sub 2}] solution, or metal. Additional processing, and additional facilities, would be required to convert the LEU to uranium dioxide (UO{sub 2}) or uranium hexafluoride (UF{sub 3}), the normal inputs for commercial fuel fabrication. This study`s scope does not include the cost for new conversion facilities. However, the low estimated cost per kilogram of blending HEU to LEU in SRS facilities indicates that even with fees for any additional conversion to UO{sub 2} or UF{sub 6}, blend-down would still provide a product significantly below the spot market price for LEU from traditional enrichment services. The body of the report develops a number of possible facility/process combinations for SRS. The primary conclusion of this study is that SRS has facilities available that are capable of satisfying the goals of a national program to blend HEU to below 5% U-235. This preliminary assessment concludes that several facility/process options appear cost-effective. Finally, SRS is a secure DOE site with all requisite security and safeguard programs, personnel skills, nuclear criticality safety controls, accountability programs, and supporting infrastructure to handle large quantities of special nuclear materials (SNM).
Bickford, W.E.; McKibben, J.M.
Draft report on melt point as a function of composition for urania-based systems
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
High Burn-Up Properties of the Fuel Variants Irradiated in IFA-649
Science Conference Proceedings (OSTI)
The "standard product" uranium dioxide (UO2) fuel pellet has remained unchanged for many years and provides excellent performance in all but the most extreme reactor operation. The requirement to prolong fuel residence in commercial reactors, thus increasing discharge levels of burn-up, has led to a need for detailed measurements of high burn-up properties under a variety of normal and off-normal conditions. The changes in fuel material properties, such as density and swelling, ...
NFIR-IV Disc Irradiation Project
Science Conference Proceedings (OSTI)
Fuel restructuring observed in uranium dioxide (UO2) based fuel at high burnup coincides with observations of enhanced fission gas release and reduced thermal conductivity of the fuel material. The transformation of fuel microstructure to so-called high burnup structure (HBS) is thus perceived as a potential limitation to fuel performance. To pursue future research and development (RD) in these and other fuel-pellet-related areas, the Nuclear Fuel Industry Research (NFIR) program commissioned an in-pile ...
Science Conference Proceedings (OSTI)
In pursuit of higher burnups at nuclear plants, fuel designers have introduced the use of 'advanced' fuel types, including doped fuels. Completing a systematic program to acquire data on the basic properties of these fuels, this project measured the specific heat and density of high burn-up UO2 and (U, Gd)O2 using irradiated materials of the same origin as those on which thermal diffusivity measurements had previously been made and thermal recovery phenomena investigated.
An economic analysis is presented for the fabrication of replacement cores for SM and PM type reactors, including analysis of various core types and core fabrication technologies. The analysis indicates that major savings are possible by utilizing Type 3 cores (40-mil plates, 25 wt% UO/sub 2/, welded assembly) in all SM and PM type reactors, and that significant savings are possible by multiple core procurement and reprocessing, and relaxation of cobalt and tantalum requirements in Type 347 stainless steel. (auth)
Wilder, A.S.
Process and product requirements to be met in the fabrication of Type 3 fuel elements are presented. The fuel elements specified consist of thin plates of a dispersion of highly enriched UO/sub 2/ and ZrB/sub 2/ in a stainless steel matrix which is clad with stainless steel on all surfaces. Quality assurance provisions are discussed. Process and material specifications and packaging and packing for shipment are described. Sample calculations and drawings are included. (M.C.G.)
Edgar, E.C.; Clayton, H.R.
High-harmonic XUV source for time- and angle-resolved photoemission spectroscopy
Science Conference Proceedings (OSTI)
We present a laser-based apparatus for visible pump/XUV probe time- and angle-resolved photoemission spectroscopy (TRARPES) utilizing high-harmonic generation from a noble gas. Femtosecond temporal resolution for each selected harmonic is achieved by using a time-delay-compensated monochromator (TCM). The source has been used to obtain photoemission spectra from insulators (UO{sub 2}) and ultrafast pump/probe processes in semiconductors (GaAs).
Dakovski, Georgi L [Los Alamos National Laboratory; Li, Yinwan [Los Alamos National Laboratory; Durakiewicz, Tomasz [Los Alamos National Laboratory; Rodriguez, George [Los Alamos National Laboratory
Accommodation of Uranium into the Garnet Structure Sergey V.Yudintsev1
Accommodation of Uranium into the Garnet Structure Sergey V.Yudintsev1 , Marya I. Lapina1 for uranium, the CaO Fe2O3 Al2O3 SiO2 ZrO2 Gd2O3 UO2 system was studied. Experiments were- corporation of U was found to be greatly dependent on the phase composition. Uranium content decreased from 18
Utsunomiya, Satoshi
Sur la radioactivit des solutions de sels d'uranium Par L. MICHIELS
432 Sur la radioactivité des solutions de sels d'uranium Par L. MICHIELS [Laboratoire de'une substanceradioactive ne produisant pas d'émanation, telle que l'uranium. D'une première série d'expériences effectuées au moyen de solutions de sulfate uranico-potassique K(UO)SO4+H2O dont la teneur, exprimée en uranium
Paris-Sud XI, Université de
Diffusion of uranium in H-451 graphite at 900 to 1400/sup 0/C
In this study, the diffusion of uranium (as a stand-in for plutonium) was investigated under conditions approximating those of the primary coolant loop in a High Temperature Gas-Cooled Reactor (HTGR). Profiles were obtained for uranium penetration in H-451 graphite (from the Great Lakes Carbon Company) at temperatures ranging from 900 to 1400/sup 0/C. Diffusion coefficients are established for UO/sub 2/ and UC/sub 2/.
Tallent, O.K.; Wichner, R.P.; Towns, R.L.
Magnetic resonance as a structural probe of a uranium (VI) sol-gel process
NMR investigations on the ORNL process for sol-gel synthesis of microspherical nuclear fuel (UO{sub 2}), has been useful in sorting out the chemical mechanism in the sol-gel steps. {sup 13}C, {sup 15}N, and {sup 1}H NMR studies on the HMTA gelation agent (Hexamethylene tetramine, C{sub 6}H{sub l2}N{sub 4}) has revealed near quantitative stability of this adamantane-like compound in the sol-Gel process, contrary to its historical role as an ammonia source for gelation from the worldwide technical literature. {sub 17}0 NMR of uranyl (UO{sub 2}{sup ++}) hydrolysis fragments produced in colloidal sols has revealed the selective formation of a uranyl trimer, ((UO{sub 2}){sub 3}({mu}{sub 3}-O)({mu}{sub 2}-OH){sub 3}){sup +}, induced by basic hydrolysis with the HMTA gelation agent. Spectroscopic results show that trimer condensation occurs during sol-gel processing leading to layered polyanionic hydrous uranium oxides in which HMTAH{sup +} is occluded as an intercalation'' cation. Subsequent sol-gel processing of microspheres by ammonia washing results in in-situ ion exchange and formation of a layered hydrous ammonium uranate with a proposed structural formula of (NH{sub 4}){sub 2}((UO{sub 2}){sub 8}O{sub 4}(OH){sub 10}) {center dot} 8H{sub 2}0. This compound is the precursor to sintered U0{sub 2} ceramic fuel.
King, C.M.; Thompson, M.C.; Buchanan, B.R. (Westinghouse Savannah River Co., Aiken, SC (United States)); King, R.B. (Georgia Univ., Athens, GA (United States). Dept. of Chemistry); Garber, A.R. (South Carolina Univ., Columbia, SC (United States). Dept. of Chemistry)
Magnetic resonance as a structural probe of a uranium (VI) sol-gel process
NMR investigations on the ORNL process for sol-gel synthesis of microspherical nuclear fuel (UO{sub 2}), has been useful in sorting out the chemical mechanism in the sol-gel steps. {sup 13}C, {sup 15}N, and {sup 1}H NMR studies on the HMTA gelation agent (Hexamethylene tetramine, C{sub 6}H{sub l2}N{sub 4}) has revealed near quantitative stability of this adamantane-like compound in the sol-Gel process, contrary to its historical role as an ammonia source for gelation from the worldwide technical literature. {sub 17}0 NMR of uranyl (UO{sub 2}{sup ++}) hydrolysis fragments produced in colloidal sols has revealed the selective formation of a uranyl trimer, [(UO{sub 2}){sub 3}({mu}{sub 3}-O)({mu}{sub 2}-OH){sub 3}]{sup +}, induced by basic hydrolysis with the HMTA gelation agent. Spectroscopic results show that trimer condensation occurs during sol-gel processing leading to layered polyanionic hydrous uranium oxides in which HMTAH{sup +} is occluded as an ``intercalation`` cation. Subsequent sol-gel processing of microspheres by ammonia washing results in in-situ ion exchange and formation of a layered hydrous ammonium uranate with a proposed structural formula of (NH{sub 4}){sub 2}[(UO{sub 2}){sub 8}O{sub 4}(OH){sub 10}] {center_dot} 8H{sub 2}0. This compound is the precursor to sintered U0{sub 2} ceramic fuel.
King, C.M.; Thompson, M.C.; Buchanan, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); King, R.B. [Georgia Univ., Athens, GA (United States). Dept. of Chemistry; Garber, A.R. [South Carolina Univ., Columbia, SC (United States). Dept. of Chemistry
Nuclear carrier business volume projections, 1980-2000
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.
Determination of ultratrace levels of uranium by selective laser excitation of precipitates
Selective excitation of probe ion luminescence (SEPIL) is applied to the analysis of uranium by coprecipitation into calcium fluoride. Subsequent ignition of the precipitate in air yields intense fluorescence transitions from UO/sub 6//sup 6 -/ species which can be selectively excited with a narrow band dye laser system. A detection limit of 0.4 pg/mL is obtained. The interference effects of several ions are also presented.
Johnston, M.V.; Wright, J.C.
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
FAQ 3-What are the common forms of uranium?
NLE Websites -- All DOE Office Websites (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.
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
Multiscale Simulation of Thermo-mechancial Processes in Irradiated Fission-reactor Materials.
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
Experimental Results for SimFuels
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.
The reaction of water with stoichiometric and O-defective UO{sub 2} thin film surfaces is studied by high-resolution photoelectron spectroscopy using synchrotron X-rays radiation. The decomposition of D{sub 2}O molecules and the oxidative healing of defects on the reduced surfaces was observed and quantified. D{sub 2}O adsorption on the stoichiometric UO{sub 2} surface at 300 K showed small amounts of OD species (ca. 532 eV) probably formed on trace amounts of surface defects, while at 95 K D2O ice (533.5 eV) was the main surface species. On the contrary, a large signal of OD species was seen on the 300 K-Ar{sup +}-sputtered (reduced) surface, UO{sub 2-x}. This was concomitant with a rapid healing of surface defects as monitored by their U4f signal. Quantitative analysis of the OD signal with increasing temperature showed their disappearance by 550 K. The disappearance of these species while hydrogen molecules are still desorbing from the surface as monitored by TPD [S.D. Senanayake, H. Idriss, Surf. Sci. 563 (1-3) (2004) 135; S.D. Senanayake, R. Rousseau, D. Colegrave, H. Idriss, J. Nucl. Mater. 342 (2005) 179] is shedding light on the re-combinative desorption mechanism from dissociatively adsorbed water molecules on the surfaces of this defective metal oxide.
Senanayake,S.; Waterhouse, G.; Chan, A.; Madey, T.; Mullins, D.; Idriss, H.
In situ treatment of VOCs by recirculation technologies
Science Conference Proceedings (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)
Science Conference Proceedings (OSTI)
A method was developed and successfully applied to characterize large uranyl fluoride (UO{sub 2}F{sub 2}) deposits at the former Oak Ridge Gaseous Diffusion Plant. These deposits were formed by a wet air in-leakage into the UF{sub 6} process gas lines over a period of years. The resulting UO{sub 2}F{sub 2} is hygroscopic, readily absorbing moisture from the air to form hydrates as UO{sub 2}F{sub 2}-nH{sub 2}O. The ratio of hydrogen to uranium can vary from 0--16, and has significant nuclear criticality safety impacts for large deposits. In order to properly formulate the required course of action, a non-intrusive characterization of the distribution of the fissile material within the pipe, its total mass, and amount of hydration was necessary. The Nuclear Weapons Identification System (NWIS) previously developed at the Oak Ridge Y-12 Plant for identification of uranium weapons components in storage containers was used to successfully characterize these deposits.
Wyatt, M.S. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Uckan, T.; Mihalczo, J.T.; Valentine, T.E. [Oak Ridge National Lab., TN (United States); Hannon, T.F. [East Tennessee Technology Park, Oak Ridge, TN (United States)
Science Conference Proceedings (OSTI)
A method was developed and successfully applied to characterize large uranyl fluoride (UO{sub 2}F{sub 21}) deposits at the former Oak Ridge Gaseous Diffusion Plant. These deposits were formed by a wet air in-leakage into the UF{sub 6} process gas lines over a period of years. The resulting UO{sub 2}F{sub 2} is hygroscopic, readily absorbing moisture from the air to form hydrates as UO{sub 2}F{sub 2}-nH{sub 2}O. The ratio of hydrogen to uranium, denoted H/U, can vary from 0--16, and has significant nuclear criticality safety impacts for large deposits. In order to properly formulate the required course of action, a non-intrusive characterization of the distribution of the fissile material within the pipe, its total mass, and amount of hydration was needed. The Nuclear Weapons Identification System (NWIS) previously developed at the Oak Ridge Y-12 Plant for identification of uranium weapons components in storage containers was used to successfully characterize the distribution, hydration, and total mass of these deposits.
Wyatt, M.S. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Uckan, T.; Mihalczo, J.T.; Valentine, T.E. [Oak Ridge National Lab., TN (United States); Hannon, T.F. [East Tennessee Technology Park, Oak Ridge, TN (United States)
Final assessment of MOX fuel performance experiment with Japanese PWR specification fuel in the HBWR
Science Conference Proceedings (OSTI)
In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pu-fissile enrichment of about 6 wt% had been irradiated in the HBWR from 1995 to 2006. The peak burn-up of MOX pellet achieved 72 GWd/tM. In this test, fuel centerline temperature, rod internal pressure, stack length and cladding length were measured for MOX fuel and UO{sub 2} fuel as reference. MOX fuel temperature is confirmed to have no significant difference in comparison with UO{sub 2}, taking into account of adequate thermal conductivity degradation due to PuO{sub 2} addition and burn-up development. And the measured fuel temperature agrees well with FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly greater than UO{sub 2} based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behavior. MOX fuel swelling rate agrees well with solid swelling rate in the literature. Cladding elongation data shows onset of PCMI in high power region. (authors)
Fujii, Hajime; Teshima, Hideyuki; Kanasugi, Katsumasa [Mitsubishi Heavy Industries, Ltd., 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan); Kosaka, Yuji [Nuclear Development Corporation, 622-12 Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Arakawa, Yasushi [The Kansai Electric Power Co., Inc., 8 Yokota, 13 Goichi, Mihama-cho, Mikata-gun, Fukui, 919-1141 (Japan)
Helium Behavior in Oxide Nuclear Fuels: First Principles Modeling
Science Conference Proceedings (OSTI)
UO2 and (U, Pu)O2 solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein. We present density functional theory (DFT) calculations for UO2, PuO2 and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO2 remains semiconducting with He in the octahedral position while UO2 requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.
Gryaznov, D.; Rashkeev, Sergey N.; Kotomin, E. A.; Heifets, Eugene; Zhukovskii, Yuri F.
A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method, the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ converted to UF/sub 6/ by reaction with a fluorinating agent, such as CoF/sub 3/. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reac tion chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. After nitrogen gas is used to sweep out the hydrogen and the water vapor formed, and while continuing to inaintain the temperature between 400 deg C and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion of UO/sub 2/ to UF/sub 4/ the temperature of the reaction chamber is lowered to about 400 deg C or less, the UF/sub 4/ is mixed with the requisite quantity of CoF/sub 3/, and after evacuating the chamber, the mixture is heated to 300 to 400 deg C, and the resulting UF/sub 6/ is led off and delivered to a condenser.
Fowler, R.D.
Urania vapor composition at very high temperatures
Due to the chemically unstable nature of uranium dioxide its vapor composition at very high temperatures is, presently, not sufficiently studied though more experimental knowledge is needed for risk assessment of nuclear reactors. We used laser vaporization coupled to mass spectrometry of the produced vapor to study urania vapor composition at temperatures in the vicinity of its melting point and higher. The very good agreement between measured melting and freezing temperatures and between partial pressures measured on the temperature increase and decrease indicated that the change in stoichiometry during laser heating was very limited. The evolutions with temperature (in the range 2800-3400 K) of the partial pressures of the main vapor species (UO{sub 2}, UO{sub 3}, and UO{sub 2}{sup +}) were compared with theoretically predicted evolutions for equilibrium noncongruent gas-liquid and gas-solid phase coexistences and showed very good agreement. The measured main relative partial pressure ratios around 3300 K all agree with calculated values for total equilibrium between condensed and vapor phases. It is the first time the three main partial pressure ratios above stoichiometric liquid urania have been measured at the same temperature under conditions close to equilibrium noncongruent gas-liquid phase coexistence.
Pflieger, Rachel [Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, 76125 Karlsruhe (Germany); Marcoule Institute for Separation Chemistry (ICSM), UMR 5257, CEA-CNRS-UMII-ENSCM, Bagnols sur Ceze Cedex (France); Colle, Jean-Yves [Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, 76125 Karlsruhe (Germany); Iosilevskiy, Igor [Joint Institute for High Temperature, Russian Academy of Science, 125412 Moscow (Russian Federation); Moscow Institute of Physics and Technology, State University, 141700 Moscow (Russian Federation); Extreme Matter Institute (EMMI), 64291 Darmstadt (Germany); Sheindlin, Michael [Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, 76125 Karlsruhe (Germany); Joint Institute for High Temperature, Russian Academy of Science, 125412 Moscow (Russian Federation)
Irradiation of SiC Clad Fuel Rods in the HFIR
Science Conference Proceedings (OSTI)
During 2009 and- 2010, new test capability for the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was developed that allows testing of advanced nuclear fuels and cladding under prototypic light-water-reactor (LWR) operating conditions (i.e., cladding and fuel temperatures, fuel average linear heat generation rates, and cladding fluence). For the initial experiments for this test program, ORNL teamed with commercial fuel/cladding vendors who have developed an advanced composite-wound SiC cladding material for possible use in LWRs. The first experiment, containing SiC-clad UN fuel, was inserted in HFIR in June 2010, and the second experiment, containing SiC-clad UO2 fuel, was inserted in October 2010. Two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in November 2011 at an estimated fuel burnup of approximately 10 GWd/MTHM; and two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in February 2013 at an estimated fuel burnup of approximately 20 GWd/MTHM. These capsules are currently awaiting PIE. This paper will describe the experiment, as-run operating conditions for these capsules, and current PIE plans and status.
Ott, Larry J [ORNL; Bell, Gary L [ORNL; Ellis, Ronald James [ORNL; McDuffee, Joel Lee [ORNL; Morris, Robert Noel [ORNL
Electrochemistry of LiCl-Li2O-H2O Molten Salt Systems
Uranium can be recovered from uranium oxide (UO2) spent fuel through the combination of the oxide reduction and electrorefining processes. During oxide reduction, the spent fuel is introduced to molten LiCl-Li2O salt at 650 degrees C and the UO2 is reduced to uranium metal via two routes: (1) electrochemically, and (2) chemically by lithium metal (Li0) that is produced electrochemically. However, the hygroscopic nature of both LiCl and Li2O leads to the formation of LiOH, contributing hydroxyl anions (OH-), the reduction of which interferes with the Li0 generation required for the chemical reduction of UO2. In order for the oxide reduction process to be an effective method for the treatment of uranium oxide fuel, the role of moisture in the LiCl-Li2O system must be understood. The behavior of moisture in the LiCl-Li2O molten salt system was studied using cyclic voltammetry, chronopotentiometry and chronoamperometry, while reduction to hydrogen was confirmed with gas chromatography.
Natalie J. Gese; Batric Pesic
Photoelectron Spectroscopy of U Oxide at LLNL
Science Conference Proceedings (OSTI)
In our laboratory at LLNL, an effort is underway to investigate the underlying complexity of 5f electronic structure with spin-resolved photoelectron spectroscopy using chiral photonic excitation, i.e. Fano Spectroscopy. Our previous Fano measurements with Ce indicate the efficacy of this approach and theoretical calculations and spectral simulations suggest that Fano Spectroscopy may resolve the controversy concerning Pu electronic structure and electron correlation. To this end, we have constructed and commissioned a new Fano Spectrometer, testing it with the relativistic 5d system Pt. Here, our preliminary photoelectron spectra of the UO{sub 2} system are presented. X-ray photoelectron spectroscopy has been used to characterize a sample of UO{sub 2} grown on an underlying substrate of Uranium. Both AlK{alpha} (1487 eV) and MgK{alpha} (1254 eV) emission were utilized as the excitation. Using XPS and comparing to reference spectra, it has been shown that our sample is clearly UO{sub 2}.
Tobin, J G; Yu, S; Chung, B W; Waddill, G D
General-Purpose Heat Source: Research and development program: Cold-Process Verification Test Series
The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of {sup 238}Pu decay to an array of thermoelectric elements. Because any space mission could experience a launch abort or return from orbit, the heat source must be designed and constructed to survive credible accident environments. Previous testing conducted in support of the Galileo and Ulysses missions documented the response of GPHSs and individual GPHS capsules fueled with {sup 238}UO{sub 2} ({sup 235}U-depleted) to a variety of explosive overpressure and impact events. In the early 1990s, Los Alamos National Laboratory (LANL) resumed fabrication of {sup 238}UO{sub 2} GPHS pellets. The Cold-Process Verification (CPV) Test Series was designed to compare the response of GPHS heat sources loaded with recently fabricated hot- and cold-pressed {sup 238}UO{sub 2} pellets to the response of urania pellets used in the Galileo and Ulysses performance tests. This report documents eleven bare-capsule impacts and one impact of a fully loaded GPHS module. All of the failures observed in the bare-clad impact tests were similar to failures observed in previous safety tests. No failures occurred in the module impact test.
Reimus, M.A.H.; George, T.G.
Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core
Science Conference Proceedings (OSTI)
The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO/sub 2/ and UO/sub 2//metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO/sub 2/ crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process.
Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.
Conceptual Design of a CERMET NTR Fission Core Using Multiphysics Modeling Techniques
An initial pre-conceptual CERMET Nuclear Thermal Propulsion reactor system is investigated within this paper. Reactor configurations are investigated where the fuel consists of 60 vol.% UO2 and 40 vol.% W where the UO2 consists of Gd2O3 concentrations of 5 and 10 mol.%.Gd2O3. The fuel configuration consisting of 5 mol.% UO2 was found to have a total mass of 2761 kg and a thrust to weight ratio of 4.10 and required a coolant channel surface area to fueled volume ratio of approximately 15.0 in order to keep the centerline temperature below 3000 K. The configuration consisting of 10 mol.% Gd2O3 required a surface area to volume ratio of approximately 12.2 to cool the reactor to a peak temperature of 3000 K and had a total mass of 3200 kg and a thrust to weight ratio of 3.54. It is not known yet what concentration of Gd2O3 is required to maintain fuel stability at 3000 K; however, both reactors offer the potential for operations at 25,000 lb, and at a specific impulse which may range from 900 to 950 seconds.
Jonathan A. Webb; Brian J. Gross; William T. Taitano
Polyethylene Encapsulated Depleted Uranium
NLE Websites -- All DOE Office Websites (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.
The commercial nuclear power industry is investing heavily in advanced fuels that can produce higher power levels with a higher safety margin and be produced at low cost. Although chemically stable and inexpensive to manufacture, the in-core performance of UO{sub 2} fuel is limited by its low thermal conductivity. There will be enormous financial benefits to any utility that can exploit a new type of fuel that is chemically stable, has a high thermal conductivity, and is inexpensive to manufacture. At reactor operating temperatures, UO{sub 2} has a very low thermal conductivity (<5 W/m {center_dot}K), which decreases with temperature and fuel burnup. This low thermal conductivity limits the rate at which energy can be removed from the fuel, thus limiting the total integrated reactor power. If the fuel thermal conductivity could be increased, nuclear reactors would be able to operate at higher powers and larger safety margins thus decreasing the overall cost of electricity by increasing the power output from existing reactors and decreasing the number of new electrical generating plants needed to meet base load demand. The objective of the work defined herein is to produce an advanced nuclear fuel based on the current UO{sub 2} fuel with superior thermal conductivity and structural integrity that is suitable for current and future nuclear reactors, using the existing fuel fabrication infrastructure with minimal modifications. There are two separate components to the research: (1) Enhanced Thermal Conductivity (ETC) - adding high conductivity fibers to the UO{sub 2} prior to sintering, which act as conduits for moving the heat energy generated within the pellet to the outer surface, (2) Microstructural Engineering (ME) - adding second phase particulates to UO{sub 2} bodies to retard grain growth and to increase thermal conductivity, as well as improve fracture and creep resistance. Different groups will perform the laboratory work for each of these research components with some overlap in personnel. The overlapping areas primarily involve computer simulations and final testing of the fuel in a reactor. The estimated cost and duration of this project is $5,000,000 over three years.
Hollenbach, Daniel F [ORNL; Ott, Larry J [ORNL; Besmann, Theodore M [ORNL; Armstrong, Beth L [ORNL; Wereszczak, Andrew A [ORNL; Lin, Hua-Tay [ORNL; Ellis, Ronald James [ORNL; Becher, Paul F [ORNL; Jubin, Robert Thomas [ORNL; Voit, Stewart L [ORNL
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.
During this reporting period, particular effort was of aced on powder blending and pellet sintering studies prior to irradiation sample fabrication, and, subsequently, the production and characterization of the pellets slated for irradiation. Also, PuO/sub 2/ and UO/sub 2/-PuO/sub 2/ characterization studies were continued, and new techniques are being developed. Specifically, dynamic moisture pickup determinations on PuO/sub 2/ were made in moist air, N, and CO/ sub 2/ atmospheres using a recording thermogravimetric balance; the Sharples Micromerograph was committed to Pu, and powder particle size distributions were measured and compared with previous determinations made with air-permeability equipment; and the suitability and reliability of analytical chemistry assaying procedures such as x-ray-fluorescence and gamma spectrometry are being evaluated. Prototype work on UO/sub 2/ for the direct precipitation of PuO/sub 2/ and PuO/ sub 2/-UO/sub 2/ feed materials for swaging, vibratory compaction, and dispersion fabrication was also continued. In addition, investigation of PuO/sub 2/ spherical particle formation by mechanical buildup and by plasma torch fusion was extended. Associated reactor physics studies were concentrated on the further comparison of Pu and U/sup 235/ in near-thermal converter reactors. In preparation for the fabrication of irradiation test specimens to be prepared by the mechanical blending of individuaI PuO/sub 2/ and UO/sub 2/ powders, bIending studies were initiated to develop methods required for the attainment of desired homogeneity. Sintering studies were carried out on PuOs/sub 2/ to study the effects of compaction pressure, firing temperature, firing time, and firing atmosphere. It was determined that 1400 to 1500 deg C is the best firing temperature to obtain maximum pellet density, and that sintering in air yields higher densities than sintering in a N/sub 2/--H/sub 2/ atmosphere. Further, it was noted that the degree of Pu/sub 2/O/sub 3/ formation while sintering in an N/ sub 2/--H/sub 2/ atm osphere is inversely proportional to compaction pressure, indicating that the degree of formation is determined by the exposed surface area. Two additional Iots of UO/sub 2/-5 wt% PuO/sub 2/ powder were precipitated during this period. Powder characterization data for these and two previously produced lots were obtained. Also, powder characteristics were remeasured following hammermilling in order to allow assessment of the effects of this treatment. In preparation for work with PuO/sub 2/ and UO/sub 2/--PuO/sub 2/, prototype studies are being carried out with UO/sub 2/ to assess the possibility of producing directly high density granular feed for swaging, vibratory compaction, and dispersion fuel fabrication. Effort was continued on the fabrication of spherical PuO/sub 2/ particles by mechanical buildup and by plasma torch fusion. Reactor physics studies were continued to allow assessment of Pu relative to U/ sup 235/ in near-thermal reactor sys tems. Under cost assumptions used previously, it was shown that optimum fuel cycle costs from Pu-natural U fueled systems are well below those attainable with slightly enriched U iueled systems even if it is assumed that radiation damage is not limiting and that an ideal burnable poison (or solution poison) exists to limit the reactivity. (auth)
Bulk and surface controlled diffusion of fission gas atoms
Science Conference Proceedings (OSTI)
Fission gas retention and release impact nuclear fuel performance by, e.g., causing fuel swelling leading to mechanical interaction with the clad, increasing the plenum pressure and reducing the gap thermal conductivity. All of these processes are important to understand in order to optimize operating conditions of nuclear reactors and to simulate accident scenarios. Most fission gases have low solubility in the fuel matrix, which is especially pronounced for large fission gas atoms such as Xe and Kr, and as a result there is a significant driving force for segregation of gas atoms to extended defects such as grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. Several empirical or semi-empirical models have been developed for fission gas release in nuclear fuels, e.g. [1-6]. One of the most commonly used models in fuel performance codes was published by Massih and Forsberg [3,4,6]. This model is similar to the early Booth model [1] in that it applies an equivalent sphere to separate bulk UO{sub 2} from grain boundaries represented by the sphere circumference. Compared to the Booth model, it also captures trapping at grain boundaries, fission gas resolution and it describes release from the boundary by applying timedependent boundary conditions to the circumference. In this work we focus on the step where fission gas atoms diffuse from the grain interior to the grain boundaries. The original Massih-Forsberg model describes this process by applying an effective diffusivity divided into three temperature regimes. In this report we present results from density functional theory calculations (DFT) that are relevant for the high (D{sub 3}) and intermediate (D{sub 2}) temperature diffusivities of fission gases. The results are validated by making a quantitative comparison to Turnbull's [8-10] and Matzke's data [12]. For the intrinsic or high temperature regime we report activation energies for both Xe and Kr diffusion in UO{sub 2{+-}x}, which compare favorably to available experiments. This is an extension of previous work [13]. In particular, it applies improved chemistry models for the UO{sub 2{+-}x} nonstoichiometry and its impact on the fission gas activation energies. The derivation of these models follows the approach that used in our recent study of uranium vacancy diffusion in UO{sub 2} [14]. Also, based on the calculated DFT data we analyze vacancy enhanced diffusion mechanisms in the intermediate temperature regime. In addition to vacancy enhanced diffusion we investigate species transport on the (111) UO{sub 2} surface. This is motivated by the formation of small voids partially filled with fission gas atoms (bubbles) in UO{sub 2} under irradiation, for which surface diffusion could be the rate-limiting transport step. Diffusion of such bubbles constitutes an alternative mechanism for mass transport in these materials.
Andersson, Anders D. [Los Alamos National Laboratory
Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels
The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center�s investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center�s research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.
Todd R. Allen, Director
A Multi-Modular Neutronically Coupled Power Generation System
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 self-contained pressurized modules, each containing fuel elements in pressurized channels surrounded by a graphite moderator, and Brayton cycle turbo-machinery. Each module is subcritical by itself, and when several modules are brought into proximity of one another, a single critical core is formed. The multi-modular approach and use of BeO-UO2 fuel with graphite moderator and supercritical CO2 coolant leads to an inherently safe system capable of high efficiency operation. The pressure channel design and multi-modular approach eliminates engineering challenges associated with large pressure vessels. The subcriticality of the modules ensures inherent safety during construction, transportation, and after decommissioning. Serpent, a continuous-energy Monte-Carlo reactor physics burnup calculation code, was used to develop a critical configuration of the subcritical modules using UO2 fuel enriched with 5 wt% 235U with a 5 wt% BeO additive. The core lifetime was found to be 14.6 years operating at 10 MWth, though the U enrichment and power can be altered to achieve desired core lifetimes. Negative fuel and moderator temperature coefficients of reactivity were found that could maintain safety during operation. The multi-modular design was found to be beneficial compared to a core with all fuel elements in one module. Batch battery type refueling was found to be beneficial and the feasibility of controlling the reactor was demonstrated through the use of control shells that surround each module. The HT-IMMTR design is an inherently safe, highly efficient, economically competitive, and most important, feasible reactor design that takes advantage of proven technologies to facilitate the demonstration of a successful commercial deployment.
Patel, Vishal
Science Conference Proceedings (OSTI)
High Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resources thanks to a higher breeding ratio than standard LWR. Their uses together with the current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and the US today. However, one of the issues related to HCLWR is their void reactivity coefficient (VRC), which can be positive. Accurate predictions of void reactivity coefficient in HCLWR conditions and their comparisons with representative experiments are therefore required. In this paper an inter comparison of modern codes and cross-section libraries is performed for a former Benchmark on Void Reactivity Effect in PWRs conducted by the OECD/NEA. It shows an overview of the k-inf values and their associated VRC obtained for infinite lattice calculations with UO{sub 2} and highly enriched MOX fuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with the libraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spread of results for voided conditions is found for the high content MOX fuel. The spread of eigenvalues for the moderated and voided UO{sub 2} fuel are about 200 pcm and 700 pcm, respectively. The standard deviation for the VRCs for the UO{sub 2} fuel is about 0.7% while the one for the MOX fuel is about 13%. This work shows that an appropriate treatment of the unresolved resonance energy range is an important issue for the accurate determination of the void reactivity effect for HCLWR. A comparison to experimental results is needed to resolve the presented discrepancies. (authors)
Hursin, M.; Koeberl, O.; Perret, G. [Paul Scherrer Institut PSI, 5232 Villigen (Switzerland)
Validation of the WATEQ4 geochemical model for uranium
As part of the Geochemical Modeling and Nuclide/Rock/Groundwater Interactions Studies Program, a study was conducted to partially validate the WATEQ4 aqueous speciation-solubility geochemical model for uranium. The solubility controls determined with the WATEQ4 geochemical model were in excellent agreement with those laboratory studies in which the solids schoepite (UO/sub 2/(OH)/sub 2/ . H/sub 2/O), UO/sub 2/(OH)/sub 2/, and rutherfordine ((UO/sub 2/CO/sub 3/) were identified as actual solubility controls for uranium. The results of modeling solution analyses from laboratory studies of uranyl phosphate solids, however, identified possible errors in the characterization of solids in the original solubility experiments. As part of this study, significant deficiencies in the WATEQ4 thermodynamic data base for uranium solutes and solids were corrected. Revisions included recalculation of selected uranium reactions. Additionally, thermodynamic data for the hydroxyl complexes of U(VI), including anionic (VI) species, were evaluated (to the extent permitted by the available data). Vanadium reactions were also added to the thermodynamic data base because uranium-vanadium solids can exist in natural ground-water systems. This study is only a partial validation of the WATEQ4 geochemical model because the available laboratory solubility studies do not cover the range of solid phases, alkaline pH values, and concentrations of inorganic complexing ligands needed to evaluate the potential solubility of uranium in ground waters associated with various proposed nuclear waste repositories. Further validation of this or other geochemical models for uranium will require careful determinations of uraninite solubility over the pH range of 7 to 10 under highly reducing conditions and of uranyl hydroxide and phosphate solubilities over the pH range of 7 to 10 under oxygenated conditions.
Krupka, K.M.; Jenne, E.A.; Deutsch, W.J.
Calculations were performed using MOCUP, which includes the use of MCNP for neutron transport and ORIGEN for depletion. The MOCUP calculations were done using a unit cell (pin cell) model, where the ThO2 varied from 65-75wt% and the UO2 varied from 25-35wt%. The fission products and actinides being tracked in the calculations account for >97% of the parasitic captures in the fuel. The fuel pin was surrounded by four reflecting planes, where typical parameters were used for a 17x17 PWR assembly. The hydrogen to heavy metal ratio (H/HM) was varied by increasing or decreasing the water density in the cell. The results show that the drier lattices have insufficient reactivity due to the limited enrichment of the uranium. However, a slightly wetter lattice will increase the reactivity-limited burnup by 26% for the 25% UO2 � 75% ThO2, and 19% for the 35% UO2 � 65% ThO2 as compared to the standard coolant density. This is appears to be consistent with similar studies done with all-uranium lattices, where advantages are gained by hardening or further softening the neutron spectrum. Although some advantage is gained by softening the spectrum, the same can be said of all-uranium fueled cores. The spectral changes and, to a lesser extent, competing resonances between the 238U and bred-in 233U appear to hamper advantages in the conversion of thorium in homogeneous fuel that might otherwise be gained by shifting the neutron spectrum. Physically separating the uranium and thorium (e.g., in micro-heterogeneous and seed-and-blanket arrangements) have been shown alleviate this problem. A change in moderation may further enhance the reactivity-limited burnup of these lattices, and will be the focus of future work.
Weaver, Kevan Dean; Herring, James Stephen
Science Conference Proceedings (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
Liquid Metal Bond for Improved Heat Transfer in LWR Fuel Rods
Science Conference Proceedings (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
Ex-vessel core catcher materials interactions. Annual progress report. [LMFBR
A twelve-month program to investigate ex-vessel core catcher materials interactions has been completed. The investigations, involving depleted uranium dioxide, magnesia brick, stainless steel, and low-carbon steel, were conducted in furnaces and associated facilities existing at Aerospace, which were modified to process molten and solidified radioactive samples. In addition to developing efficient methods for the melting, pouring, and sustained heating of UO/sub 2/, extensive sample characterizations and microanalyses were performed. Theoretical analyses were also made in data interpretation for the purpose of understanding the interaction kinetics.
Swanson, D.G.; Zehms, E.H.; Ang, C.Y.; McClelland, J.D.; Meyer, R.A.; vanPaassen, H.L.L.
Progress is reported on the gross solubility of U in digestions of Mallinokrodt feed materials, studies of variables affecting U purity in a TBP hexane extraction cycle, low-acid flowsheet for TBP--hexane extraction process based on a 440 g U/liter in lM HNO/sub 3/ digest liquor, hacking studies in the pilot plant pumperdecanter system, recovery of U from residues from the dingot process, lowering the H level in dingot metal, forging of dingot bar stock, dingot extrusion, fubrication of UO/sub 2/ fuel elements, and the determination of H content of derby and ingot metal. (W.L.H.)
Kuhlman, N. ed.
Actinide speciation in glass leach-layers: An EXAFS study
Uranium L{sub 3} X-ray absorption data were obtained from two borosilicate glasses, which are considered as models for radioactive wasteforms, both before and after leaching. Surface sensitivity to uranium speciation was attained by a novel application of simultaneous fluorescence and electron-yield detection. Changes in speciation are clearly discernible, from U(VI) in the bulk to (UO{sub 2}){sup 2+}-uranyl in the leach layer. The leach-layer uranium concentration variations with leaching times are also determined from the data.
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)
Developments in chemical engineering are described. Pyrometallurgical development, fuel-processing facilities for EBR-II, pyrometallurgical research, laboratory and engineering-scale investigations of fluoride volatility processes, conversion of UF/sup 6/ to UO/sup 2/, calcination studies in small-diamenter columns, metal oxidation and ignition kinetics, metal-water reactions, determination of nuclear constants, thermally regenerative emf cell, thermoelectricity research, reactor decontamination, waste processing, and the high-level gamma-irradiation facility are discussed. (M.C.G.)
The conceptual design of a 100 Mw(e) nuclear power plant is described. The plant utilized a sodium-cooled graphite-moderated reactor with stainless- steel clad. slightiy enriched UO/sub 2/ fuel. The reactor is provided with three main coolant circuits, and the steam cycle has three stages of regenerative heating. The plant control system allows automatic operation over the range of 20 to 100% load, or manual operation at all loads. The site, reactor, sodium systems, reactor auxiliaries, fuel handling, instrumentation, turbine-generator, buildings. and safety measures are described. Engineering drawings are included. (W.D.M.)
Stainless steel-base fuel components of thin plate-typs construction and containing a dispersion of enriched UO/sub 2/ have been successfully employed in powering the Army package Power Reactor. The stationary fuel compcnent proposed for operation in the second core loading of the reactor is discussed. The component is designed for radioactive service in pressurized water at 4504DEF and consists of eighteen composite fuel plates joined into an Integral unit or assembly by brazing. Design specifications covering the material and dimensional requirements as well as the operating conditions are discussed. Step-by-step procedures developed and utilized in manufacturing the component are presented in detail. (auth)
Cunningham, J.E.; Beaver, R.J.
Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications
Science Conference Proceedings (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
A process is presented for making a ceramic-like material suitable for use as a nuclear fuel. The material consists of a solid solution of plutonium dioxide in uranium dioxide and is produced from a uranyl nitrate -plutonium nitrate solution containing uraniunm and plutonium in the desired ratio. The uranium and plutonium are first precipitated from the solution by addition of NH/ sub 4/OH and the dried precipitate is then calcined at 600 C in a hydrogen atmosphere to yield the desired solid solution of PuO/sub 2/ in UO/sub 2/.
Alter, H.W.; Davidson, J.K.; Miller, R.S.; Mewherter, J.L.
A method of hardening U metal involves the forming of a fine dispersion of UO/sub 2/. This method consists of first hydriding the U to form a finely divided powder and then exposing the powder to a very dilute O gas in an inert atmosphere under such pressure and temperature conditions as to cause a thin oxide film to coat each particle of the U hydride, The oxide skin prevents agglomeration of the particles as the remaining H is removed, thus preserving the small particle size. The oxide skin coatings remain as an oxide dispersion. The resulting product may be workhardened to improve its physical characteristics. (AEC)
Arbiter, W.
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.
Separation of uranium from technetium in recovery of spent nuclear fuel
A method for decontaminating uranium product from the Purex process comprises addition of hydrazine to the product uranyl nitrate stream from the Purex process, which contains hexavalent (UO.sub.2.sup.2+) uranium and heptavalent technetium (TcO.sub.4 -). Technetium in the product stream is reduced and then complexed by the addition of oxalic acid (H.sub.2 C.sub.2 O.sub.4), and the Tc-oxalate complex is readily separated from the uranium by solvent extraction with 30 vol. % tributyl phosphate in n-dodecane.
Friedman, Horace A. (Oak Ridge, TN)
Separation of uranium from technetium in recovery of spent nuclear fuel
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.
Corrosion of Spent Nuclear Fuel: The Long-Term Assessment
Science Conference Proceedings (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
Science Conference Proceedings (OSTI)
This report presents results from a parametric study of equilibrium fuel cycle costs for a closed fuel cycle with multi-recycling of plutonium in fast reactors (FRs) compared to an open, once-through fuel cycle using PWRs. The study examines the impact on fuel cycle costs from changes in the unit costs of uranium, advanced PUREX reprocessing of discharged uranium dioxide (UO2) fuel and fast-reactor mixed-oxide (FR-MOX) fuel, and FR-MOX fuel fabrication. In addition, the impact associated with changes in ...
Heat pipe reactors for space power applications
A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kW(e) and operate in the temperature range 1200 to 1700/sup 0/K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO/sub 2/. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor.
Koenig, D.R.; Ranken, W.A.; Salmi, E.W.
Status of ANL out-of-pile investigations of severe accident phenomena for liquid metal reactors
Research addressing LMFBR whole core accidents has been terminated, and there is now emphasis on quantifying reactivity feedbacks, and in particular enhancing negative feedback, so that advanced LMR designs will provide inherently safe operation. The status of recent HCDA-related laboratory research performed at ANL, up to the time that such activities were no longer needed to support CRBR licensing, is described. Included are descriptions of programs addressing sodium channel voiding, fuel sweepout, fuel dispersal and plugging, boiled-up pool, UO/sub 2//sodium FCI, and debris coolability. Descriptions of recent investigations involving the metal fuel/sodium system are also included.
Spencer, B.W.; Marchaterre, J.F.; Anderson, R.P.; Armstrong, D.R.; Baker, L.; Cho, D.H.; Gabor, J.D.; Pedersen, D.R.; Sienicki, J.J.; Stein, R.P.
A method is given for making an electrode for use in the electrolytic reduction of uranium oxides to uranium metal in a fused salt electrolyte. Uranlum oxide such as UO/sub 2/ is mixed with somewhat less than the stoichiometric amount of carbon needed for the reduction, and the mixture is baked and crushed to make a nonspherical material. The latter is then mixed with a carbon binder sufficient to satisfy stoichiometry, pressed into a shape such as a cylinder, and baked. (AEC)
Piper, R.D.; Leifield, R.F.
NEAMS Update Quarterly Highlights
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
A study was made of the distribution of Ag, Au, Bt, and Pd in the residues from an extraction process in which purified UO/sub 2/(NO/sub 3/)/sub 2/ is prepared from a pitchblende ore. The residues studied were the gangue Pb cake, the BaSO/sub 4/ cake, the filtrate boildown cake, the raffinate cake, and the purified UO/sub 2/(NO/sub 3/)/sub 2/. Samples representing three different batches of ore were used. The following conclusions were reached: (1) neither Ag nor Pd are present in the ore in significant quantities; (2) approximately 30 ppm of Au (based on U content) is present, of which 90% was found in the residue (gangue Pb cake) from the initial co-precipitation step; this residue contains gangue, PbSO/sub 4/, RaSO/sub 4/, and other insoluble sulfates; (3) the ore contains 15 ppm of Pt, which is distributed in an extremely varying manner in the residues leaving the process. (W. L.H.)
Shearer, R.W.
Over the last decades, many analytical experiments (in-pile and out-of-pile) have underlined the active role of the inter-granular gases on the global fuel transient behavior under accidental conditions such as RIA and/or LOCA. In parallel, the improvement of fission gas release modeling in nuclear fuel performance codes needs direct experimental determination/validation regarding the local gas distribution inside the fuel sample. In this context, an experimental program, called 'ADAGIO' (French acronym for Discriminating Analysis of Accumulation of Inter-granular and Occluded Gas), has been initiated through a joint action of CEA, EDF and AREVA NP in order to develop a new device/technique for quantitative and direct measurement of local fission gas distribution within an irradiated fuel pellet. ADAGIO technique is based on the fact that fission gas inventory (intra and inter-granular parts) can be distinguished by controlled fuel oxidation, since grain boundaries oxidize faster than the bulk. The purpose of the current paper is to present both the methodology and the associated results of the ADAGIO program performed at CEA. It has been divided into two main parts: (i) feasibility (UO{sub 2} and MOX fuels), (ii) application on high burn up UO{sub 2} fuel. (authors)
Pontillon, Y.; Noirot, J.; Caillot, L. [Commissariat a l'Energie Atomique, DEN/DEC/SA3C, Centre d'Etudes de Cadarache, BP1, 13108 Saint Paul Les Durance (France); Muller, E. [Commissariat a l'Energie Atomique, DEN/DEC/SESC, Centre d'Etudes de Cadarache, BP1, 13108 Saint Paul Les Durance (France)
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
Infrared absorption data and their correlation with laboratory reactivity studies of various types of uranium UO/sub 3/ are presented. For certain materials, increased absorption at 11.5 microns can be correlated with increased reactivity of the UO/sub 2/ produced. A mathematical study of the sedimentation method for particle size analysis indicates that no correction is necessary for the length of the plummet. The fine precipitate formed in dingot uranium contalning small alloying additions of iron and silicon was shown to grow in size or dissolve by thermal treatments. Changes in the size and number of the particles have produced pronounced changes in the beta-treated grain size. The successful adaptation of a strain-gage pressure transducer to the vacuum fusion gas analysis of uranium metal is described. The transducer provides continuous indication of the pressure in the system as an aid in establishing blank rates and in following the course of reactions. Modification of the Consolidated Model 21-611-1 mass spectrometer to perform hydrogen determinations in uranium is described. Results and precision were comparable to those of the vacuum extraction equipment. (auth)
A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ convented to UF/sub 6/ by reaction with a fluorinating agent. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reaction chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. The oven is then swept clean of hydrogen and the water vapor formed by means of nitrogen and then while continuing to maintain the temperature between 400 and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion to uranium tetrafluoride, the temperature of the reaction chamber is lowered to ahout 400 deg C, and elemental fluorine is used as the fluorinating agent for the conversion of UF/sub 4/ into UF/sub 6/. The fluorine gas is passed into the chamber, and the UF/sub 6/ formed passes out and is delivered to a condenser.
Fowler, R.D.
Stationary and protable instruments for assay of HEU (highly enriched uranium) solids holdup
Two NaI(Tl)-based instruments, one stationary and one portable, designed for automated assay of highly enriched uranium (HEU) solids holdup, are being evaluated at the scrap recovery facility of the Oak Ridge Y-12 Plant. The stationary instrument, a continuous monitor of HEU within the filters of the chip burner exhaust system, measures the HEU deposits that accumulate erratically and rapidly during chip burner operation. The portable system was built to assay HEU in over 100 m of elevated piping used to transfer UO/sub 3/, UO/sub 2/, and UF/sub 4/ powder to, from, and between the fluid bed conversion furnances and the powder storage hoods. Both instruments use two detector heads. Both provide immediate automatic readout of accumulated HEU mass. The 186-keV /sup 235/U gamma ray is the assay signature, and the 60-keV gamma ray from an /sup 241/Am source attached to each detector is used to normalize the 186-keV rate. The measurement geometries were selected for compatibility with simple calibration models. The assay calibrations were calculated from these models and were verified and normalized with measurements of HEU standards built to match geometries of uniform accumulations on the surfaces of the process equipment. This instrumentation effort demonstrates that simple calibration models can often be applied to unique measurement geometries, minimizing the otherwise unreasonable requirements for calibration standards and allowing extension of the measurements to other process locations.
Russo, P.A.; Sprinkle, J.K. Jr.; Stephens, M.M.; Brumfield, T.L.; Gunn, C.S.; Watson, D.R.
The development of plasma-jet spray-coating techniques for producing clad ceramic fuel plates is discussed. Conditions for spraying fused UO/sub 2/ powder were established by depositing cones on stationary substrates. It was found that the arc-gas flow range within which deposition occurs is very narrow. Coatings were made from --200 +325, --270 + 325, and de-slimed -325 mesh fused UO/ sub 2/ powders. To provide data regarding the economics of the process, deposition rates and efficiencies were determined under various conditions. The effects of powder size, power input, arcgas flow rate, spray distance, traverse rate, power feed rate, powder-gas flow rate, and cover-gas flow rate on deposition efficiency are discussed. Oxygen-to-uranium ratios of coatings made for evaluation of density were determined by gravimetric and volumetric methods. Preparation of the surface without distortion for plasma spraying is discussed. Fixturing and instrumentation methods were designed for measuring substrate and coating temperatures during spraying of typical fuel-element-cladding thickesses of stainless steel and Zircaloy-2. (M.C.G.)
ABS>The stoicbiometric HNO/sub 3/ concentrations required to digest feed materials to 200g U/I, 3N excess acid are tabulated. Stoichiometrically the minimum HNO/sub 3/ concentration required to dissolve feed materials to flowsheet conditions flowsheet tests is presented. The sensitivity of the Weldon Spring TBP-extraction system to process flow rates is discussed. A series of pilot plant experiments are deserined for removing TBP from raffinate. Steam distillstion of the raffinate slurries proved to be more effective than washing with hexone in a pulse column. The flame fusion melting of UO/sub 2/ to form dense rods of relatively constant cross section continued. The rods and tubes which were extruded from micronized UO/sub 2/ had excellent surface quality and sintered densities greater than 98% of theoretical. Bomb center temperatures of 500 deg F and higher prior to fining, obtained by increasing the firing time, tended to produce low hydrogen U. A method for producing dingot U with an acceptable H/sub 2/ contact was developed. A number of solid additions to the UF/ sub 4/-Mg bomb charge were investigated. The tonnage of forged bar stock produced during this period is given. The variables affeeting diagot extrusion are discussed. An investigation of the dingot process is given. The determination of plant decontamination faetors for Ce/sup 3+/, Mg, and Al are presented. (W.L.H.)
Elliott, B. ed.
Science Conference Proceedings (OSTI)
IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% � 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.
Ayman I. Hawari; Mohamed A. Bourham
This note describes: the rationale for the test pressure of the inner ECN cryostat vessel, the equipment to be used in this test, the test procedure, the status of the vessel prior to the test, the actual test results, and a schematic diagram of the testing set up and the pressure testing permit. The test, performed in the evening of July 17, 1991, was a major success. Based on a neglible pressure drop indicated on the pressure gages (1/4 psi), the vessel appeared to be structurally sound throughout the duration of the test (approx. 1.5 hrs.). No pressure increases were observed on the indicators looking at the beam tube bellows volumes. There was no indication of bubbles form the soap test on the welds and most of the fittings that were checked. There were some slight deviations in the actual procedure used. The UO filter was removed after the vessel had bled down to about 18 psig in order to speed up that aspect of the test. The rationale was that the higher velocity gas had already passed through at the higher pressures and there was no visible traces of the black uo particles. The rate of 4 psi/10 minutes seemed incredibly slow and often that time was reduced to just over half that rate. The testing personnel was allowed to stay in the pit throughout the duration of the test; this was a slight relaxation of the rules.
Dixon, K.; /Fermilab
Structural determination of fluorite-type oxygen excess uranium oxides using EXAFS spectroscopy
Science Conference Proceedings (OSTI)
Extended x-ray absorption fine structure (EXAFS) spectroscopy has been carried out at 77 K at the uranium L/sub III/ edge for UO/sub 2/, ..beta..-U/sub 3/O/sub 7/, and U/sub 4/O/sub 9/ with the aim of determining the structure of these highly defective (oxygen excess) uranium oxide phases, which are of industrial importance. Use has been made of a difference Fourier technique for U/sub 3/O/sub 7/, in which the EXAFS of a perfect lattice model is subtracted. U--O bond lengths calculated from the remaining EXAFS signal, assumed to result only from interstitial oxygens, have been used to determine the atomic coordinates of these interstitials. The analysis of EXAFS data in terms of coordination number has allowed an insight into the defect aggregate arrangement of oxygens in U/sub 3/O/sub 7/ and U/sub 4/O/sub 9/. Furthermore, EXAFS data indicate that the uranium sublattice is perturbed by the incorporation of additional oxygen atoms.
Jones, D.J.; Roziere, J.; Allen, G.C.; Tempest, P.A.
Nuclear criticality safety modeling of an LEU deposit
The construction of the Oak Ridge Gaseous Diffusion Plant (now known as the K-25 Site) began during World War H and eventually consisted of five major process buildings: K-25, K-27, K-29, K-31, and K-33. The plant took natural (0.711% {sup 231}U) uranium as feed and processed it into both low-enriched uranium (LEU) and high-enriched uranium (HEU) with concentrations up to {approximately}93% {sup 231}U. The K-25 and K-27 buildings were shut down in 1964, but the rest of the plant produced LEU until 1985. During operation, inleakage of humid air into process piping and equipment caused reactions with gaseous uranium hexafluoride (UF{sub 6}) that produced nonvolatile uranyl fluoride (UO{sub 2}F{sub 2}) deposits. As part of shutdown, most of the uranium was evacuated as volatile UF{sub 6}. The UO{sub 2}F{sub 2} deposits remained. The U.S. Department of Energy has mitiated a program to unprove nuclear criticality safety by removing the larger enriched uranium deposits.
Haire, M.J.; Elam, K.R.; Jordan, W.C.; Dahl, T.L.
Structure of Biogenic Uraninite Produced By Shewanella Oneidensis Strain MR-1
Science Conference Proceedings (OSTI)
The stability of biogenic uraninite with respect to oxidation is seminal to the success of in situ bioreduction strategies for remediation of subsurface U(VI) contamination. The properties and hence stability of uraninite are dependent on its size, structure, and composition. In this study, the local-, intermediate-, and long-range molecular-scale structure of nanoscale uraninite produced by Shewanella oneidensis strain MR-1 was investigated using EXAFS, SR-based powder diffraction and TEM. The uraninite products were found to be structurally homologous with stoichiometric UO{sub 2} under all conditions considered. Significantly, there was no evidence for lattice strain of the biogenic uraninite nanoparticles. The fresh nanoparticles were found to exhibit a well-ordered interior core of diameter ca. 1.3 nm and an outer region of thickness ca {approx}0.6 nm in which the structure is locally distorted. The lack of nanoparticle strain and structural homology with stoichiometric UO{sub 2} suggests that established thermodynamic parameters for the latter material are an appropriate starting point to model the behavior of nanobiogenic uraninite. The detailed structural analysis in this study provides an essential foundation for subsequent investigations of environmental samples.
Schofield, E.J.; Veeramani, H.; Sharp, J.O.; Suvorova, E.; Bernier-Latmani, R.; Mehta, A.; Stahlman, J.; Webb, S.M.; Clark, D.L.; Conradson, S.D.; Ilton, E.S.; Bargar, J.R.
Experience making mixed oxide fuel with plutonium from dismantled weapons
Science Conference Proceedings (OSTI)
Mixed depleted UO{sub 2} and PuO{sub 2} (MOX) pellets prototypic of fuel proposed for use in commercial power reactors were made with plutonium recovered from dismantled weapons. We characterized plutonium dioxide powders that were produced at the Los Alamos and Lawrence Livermore National Laboratories (LANL and LLNL) using various methods to recover the plutonium from weapons parts and to convert It to oxide. The gallium content of the PUO{sub 2} prepared at LANL was the same as in the weapon alloy while the content of that prepared at LLNL was less. The MOX was prepared with a five weight percent plutonium content. We tested various MOX powders milling methods to improve homogeneity and found vibratory milling superior to ball milling. The sintering behavior of pellets made with the PuO{sub 2} from the two laboratories was similar. We evaluated the effects of gallium and of erbium and gadolinium, that are added to the MOX fuel as deplorable neutron absorbers, on the pellet fabrication process and an the sintered pellets. The gallium content of the sintered pellets was <10 ppm, suggesting that the gallium will not be an issue in the reactor, but that it will be an Issue in the operation of the fuel fabrication processing equipment unless it is removed from the PuO{sub 2} before it is blended with the UO{sub 2}.
Blair, H.T.; Ramsey, K.B.
Process for continuous production of metallic uranium and uranium alloys
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.
Process for continuous production of metallic uranium and uranium alloys
A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.
Hayden, Jr., Howard W. (Oakridge, TN); Horton, James A. (Livermore, CA); Elliott, Guy R. B. (Los Alamos, NM)
Modeling and design of a reload PWR core for a 48-month fuel cycle
Science Conference Proceedings (OSTI)
The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.
McMahon, M.V.; Driscoll, M.J.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States)
The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.
Michael A. Pope; R. Sonat Sen; Brian Boer; Abderrafi M. Ougouag; Gilles Youinou
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
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
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel
Science Conference Proceedings (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
Science Conference Proceedings (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.
Science Conference Proceedings (OSTI)
The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.
R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu
NLE Websites -- All DOE Office Websites (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-
Microsoft Word - ICEM05_DCURETE.doc
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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
BISON Enhanced | Department of Energy
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
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.
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
Distribution Category: Atomic, Molecular, and Chemical Physics
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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
Coolant Sub-Channel and Smeared-Cracking Models in BISON | Department of
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Studies of the effect of UO/sub 2/ dispersions in uranium metal upon properties which exhibit resistance to radiation damage were continued. Procedures were developed for preparing uranium powders of particle size less than 5 mu by hydride decomposition, and methods were developed for controlled oxidation of the powders obtained. Equipment for vacuum hot pressing and/or extrusion of powders was designed and fabricated. Samples of dispersion-hardened uranium, containing 13 to 33 vol.% uranium oxide, were prepared by extrusion in the gamma uranium temperature range. These samples were subjected to thermal cycling tests through the alpha - beta transformation temperature using a total cycle time of 15 to 20 min. Dimensional stability was observed to be superior to thai of wrought, unalloyed uranium. Transverse bending tests revealed the hightemperature strength of the dispersion-hardened compositions to be substantially greater than that of wrought, unalloyed uranium. (For preceding period see NDA-21121.) (C.J.G.)
Arbiter, W.
United abominations: Density functional studies of heavy metal chemistry
Science Conference Proceedings (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
Solid-state diffusion bonding of tungsten to tungsten was investigated at temperatures ranging from 1700 to 2600 nif- C, under surface contact pressures up to 3000 psi, while under high vacuum or hydrogen atmosphere. Various interface coatings were employed to promote diffusion, including graphite, oxide, metal slurries, electroplates, direct surface oxidation, and Mo - -W deposits from carbonyl decompositions. Thorough metallurgical bonding was achieved, particularly with the latter two surface coatings, after 2 hours at 2350 nif- C in H/sub 2/ under 1400 psi. Corresponding tensile strengths of 30,000 psi were obtained. Powder-compacted tungsten sheet containing 50 vo1% UO/sub 2/, spray-coated with an outer layer of tungsten, was effectively bonded to itself and to tungsten metal under 2 hour diffusion treatments at 2000 nif- C and moderate pressures of the order of 1000 psi. (auth)
Batista, R.I.; Hanks, G.S.; Murphy, D.J.
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
Corium Spreading Over Concrete: The Vulcano VE-U7 and VE-U8 Tests
Science Conference Proceedings (OSTI)
Two experiments have been performed in the VULCANO facility in which prototypic corium has been spread over concrete. In the VE-U7 test, a mixture representative of what can be expected at the opening of EPR reactor-pit gate has been spread on siliceous concrete and on a reference channel in inert refractory ceramic. The spreading progression was not much affected by the presence of concrete and sparging gases. In the VE-U8 test, a UO{sub 2}-ZrO{sub 2} mixture, prototypic of in-vessel corium, has been spread over a lime-siliceous concrete. Although residual power was not simulated in this experiment, up to 2 cm of concrete have been eroded during the test. Results in terms of spreading behaviour, effects of gases, concrete erosion and thermal attack are presented and discussed. (authors)
Journeau, Christophe; Boccaccio, Eric; Fouquart, Pascal; Jegou, Claude; Piluso, Pascal [CEA Cadarache, F-13108 St Paul lez Durance cedex (France)
Transient Testing of Nuclear Fuels and Materials in United States
Science Conference Proceedings (OSTI)
The US Department of Energy (DOE) has been engaged in an effort to develop and qualify next generation LWR fuel with enhanced performance and safety and reduced waste generation since 2010. This program, which has emphasized collaboration between the DOE, U.S. national laboratories and nuclear industry, was refocused from enhanced performance to enhanced accident tolerance following the events at Fukushima in 2011. Accident tolerant fuels have been specifically described as fuels that, in comparison with standard UO2-Zircaloy, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events. The program maintains an ambitious goal to insert a lead test assembly (LTA) of the new design into a commercial power reactor by 2022 .
Daniel M. Wachs
The first phase of a program aimed toward the development of materials and a core-subassembly design for the second core of the Fermi Reactor is outlined. The ground rules established by APDA for the study were based upon the performance requirements of the reactor plant with some modification to permit hlgher power generation and upon a value of Pu produced of about per g. Consideratlons of various factors led to the selection of the dispersion or cermet fuel as havlng the most potential for the Core II application. The fuel selected was a 30 vol. % dispersion of UO/sub 2/ in U--10 wt. % Mo or gamma- phasetype alloy, zirconium clad, in a flat-plate-subassembly design. The plant economics for subassembly designs of the fuel systems were evaluated at a core power level of 616 Mw. (W.D.M.)
Fawcett, S.L. ed.
Development and processing of LEU targets for {sup 99}Mo production
Substituting LEU for HEU in targets for producing fission-product {sup 99}Mo requires changes in target design and chemical processing. We have made significant progress in developing targets and chemical processes for this purpose. Target development was concentrated on a U- metal foil target as a replacement for the coated-UO{sub 2} Cintichem- type target. Although the first designs were not successful because of ion mixing-induced bonding of the U foil to the target tubes, recent irradiations of modified targets have proven successful. It was shown that only minor modifications of the Cintichem chemical process are required for the U-metal foil targets. A demonstration using prototypically irradiated targets is anticipated by the end of 1996. Progress was also made in basic dissolution of both U-metal foil and Al-clad U{sub 3}Si{sub 2} dispersion fuel targets, and work in this area is also continuing.
Snelgrove, J.L.; Vandergrift, G.F.; Hofman, G.L.
A survey was made of methods potentially useful for the determination of the oxygen to metal ratio in mixed oxides of uranium and plutonium. A gravimetric method was selected as being the most promising for adaptation in a short period of time. Development of the technique resulted in a reliable method which meets the requirements for unirradiated mixed oxide fuel samples. The method, based upon an equilibrium weight at 700 deg C in dry hydrogen, was shown to be capable of measurement of O/(Pu + U) ratios in 20% PuO/sub 2/--80% UO/sub 2/ pellets with a standard deviation of plus or minus 0.001. (auth)
Lyon, W.L.
A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)
Daniels, F.
Energy loss spectra from fluorite-structured ZrO2, CeO2, and UO2 compounds are compared with theoretical calculations based on density functional theory (DFT) and its extensions, including the use of Hubbard-U corrections (DFT + U) and hybrid functionals. Electron energy loss spectra (EELS) were obtained from each oxide using a scanning transmission electron microscope (STEM). The same spectra were computed within the framework of the full-potential linear augmented plane-wave (FLAPW) method. The theoretical and experimental EEL spectra are compared quantitatively using non-linear least squares peak fitting and a cross-correlation approach, with the best level of agreement between experiment and theory being obtained using the DFT + U and hybrid computational approaches.
Aguiar, Jeff; Ramasse, Q. M.; Asta, Mark D.; Browning, Nigel D.
Shielding considerations for advanced space nuclear reactor systems
To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.
Angelo, J.P. Jr.; Buden, D.
The core design and the fuel element concept for the high-flux Argonne Advanced Research Reactor are presented. The core is cooled and moderated by light water and utilizes beryllium as a reflector. The fuel element assembly is rhomboidal in cross section and consists of 27 plates fastened together at their edges by dovetailed locking keys, and at each end by end fittings. Each fuel plate is 40 mils thick and contains a uniform dispersion of highly enriched UO/ sub 2/ particles, up to a maximum of 37 wt%, in a matrix of sintered stainless steel powder. A 5 mil thick stainless steel cladding is metallurgically bonded to each side of the fueled matrix. (N.W.R.)
Adolph, N.R.; Silberstein, M.S.; Weinstein, A.
Science Conference Proceedings (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 (UO{sub 2(s)}) deployed in groundwater at Rifle, characterization of a zone of natural bioreduction exhibiting relevant reduced mineral phases, and laboratory studies of the oxidative capacity of Fe(III) and reductive capacity of Fe(II) with regard to U(IV) and U(VI), respectively.
Campbell, K. M. [USGS, Menlo Park, CA (United States); Davis, J. A. [USGS, Menlo Park, CA (United States) and Lawrence Berkeley National Lab., Berkeley, CA (United States); Bargar, J. [Stanford Synchrotron Radiation Lightsource, Menlo Park, CA (United States); Giammar, D. [Washington Univ., St. Louis, MO (United States); Bernier-Latmani, R. [Ecole Polytechnique Federale de Lausanne (Switzerland). Environmental Microbiology Lab.; Kukkadapu, R. [Pacific Northwest National Lab., Richland, WA (United States); Williams, K. H. [Lawrence Berkeley National Lab., Berkeley, CA (United States); Veramani, H. [Washington Univ., St. Louis, MO (United States); Ulrich, K. U. [Washington Univ., St. Louis, MO (United States) and BGD Boden- und Grundwasserlabor GmbH Dresden (Germany); Stubbs, J. [Stanford Synchrotron Radiation Lightsource, Menlo Park, CA (United States); Yabusaki, S. [Pacific Northwest National Lab., Richland, WA (United States); Figueroa, L. [Colorado School of Mines, Golden, CO (United States); Lesher, E. [Colorado School of Mines, Golden, CO (United States); Wilkins, M. J. [Pacific Northwest National Lab., Richland, WA (United States); Peacock, A. [Haley and Aldrich, Oak Ridge, TN (United States); Long, P. E. [Pacific Northwest National Lab., Richland, WA (United States)
Science Conference Proceedings (OSTI)
Five actual Savannah River Site tank waste samples and three chemically-modified samples were tested to determine solubility limits for uranium and plutonium over a one year time period. Observed final uranium concentrations ranged from 7 mg U/L to 4.5 g U/L. Final plutonium concentrations ranged from 4 {micro}g Pu/L to 12 mg Pu/L. Actinide carbonate complexation is believed to result in the dramatic solubility increases observed for one sample over long time periods. Clarkeite, NaUO{sub 2}(O)OH {center_dot} H{sub 2}O, was found to be the dominant uranium solid phase in equilibrium with the waste supernate in most cases.
King, W.; Hobbs, D.; Wilmarth, B.; Edwards, T.
Fully Ceramic Microencapsulated Fuel Development for LWR Applications
The concept, fabrication, and key feasibility issues of a new fuel form based on the microencapsulated (TRISO-type) fuel which has been specifically engineered for LWR application and compacted within a SiC matrix will be presented. This fuel, the so-called fully ceramic microencapsulated fuel is currently undergoing development as an accident tolerant fuel for potential UO2 replacement in commercial LWRs. While the ability of this fuel to facilitate normal LWR cycle performance is an ongoing effort within the program, this will not be a focus of this paper. Rather, key feasibility and performance aspects of the fuel will be presented including the ability to fabricate a LWR-specific TRISO, the need for and route to a high thermal conductivity and fully dense matrix that contains neutron poisons, and the performance of that matrix under irradiation and the interaction of the fuel with commercial zircaloy clad.
Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Terrani, Kurt A [ORNL; Voit, Stewart L [ORNL
A process amd associated apparatus for producing UF/sub 4/ from U/sub 3/ O/sub 8/ by a fluidized'' technique are reported. The U/sub 3/O/sub 8/ is first reduced to UO/sub 2/ by reaction with hydrogen, and the lower oxide of uranium is then reacted with gaseous HF to produce UF/sub 4/. In each case the reactant gas is used, alone or in combination with inert gases, to fluidize'' the finely divided reactant solid. The complete setup of the plant equipment including bins, reactor and the associated piping and valving, is described. An auxiliary fluorination reactor allows for the direct production of UF/sub 6/ from UF/sub 4/ and fluorine gas, or if desired, UF/sub 4/ may be collected as the product.
Murphree, E.V.
Apparatus and method for the direct, nondestructive evaluation of the /sup 235/U nuclide content of samples containing UF/sub 6/, UF/sub 4/, or UO/sub 2/ utilizing the passive neutron self-interrogation of the sample resulting from the intrinsic production of neutrons therein. The ratio of the emitted neutron coincidence count rate to the total emitted neutron count rate is determined and yields a measure of the bulk fissile mass. The accuracy of the method is 6.8% (1sigma) for cylinders containing UF/sub 6/ with enrichments ranging from 6% to 98% with measurement times varying from 3-6 min. The samples contained from below 1 kg to greater than 16 kg. Since the subject invention relies on fast neutron self-interrogation, complete sampling of the UF/sub 6/ takes place, reducing difficulties arising from inhomogeneity of the sample which adversely affects other assay procedures. 4 figs., 1 tab.
Menlove, H.O.; Stewart, J.E.
FAST OXIDE BREEDER-REACTOR. PART I. PARAMETRIC STUDY OF 300(e) MW REACTOR CORE
Physics scoping studies of a 300-Mw(e) PuO/sub 2/-UO/sub 2/-fueled fast- breeder reactor are reported. Physics design parameters that effect fuel costs, full conservation, and reactor safety were evaluated for use in the selection of parameters for a reference design. The total breeding ratio varied from 1.1 to 1.5 in the range of parameters corsidered. Plutonium core loading ranged from 500 to 1500 kg. Doubling time was found to be reduced by high-density fuel and low steel content. A compromise figure on fuel-rod range of sizes (about 100 mils) yields a 5 operating reactivity and a small, negative sodium temperature coefficient. (J.R.D.)
Greebler, P.; Aline, P.; Sueoka, J.
Conceptual design and economic studies of a steamcooled fast breeder reactor that can also be used as a source of power are presented. Two reactor plant sizes were considered: a 300-Mw(e) central power station plant and a 40 Mw(e) plant. It was concluded that attractive economics and good breeding characteristics breeding ratios from 1.27 to 1.42) can be achieved in steam- cooled PuO/sub 2/UO/sub 2/ fueled fast reactors. Low capital costs can be obtained by a compact reactor core and the absence of large heat exchangers and complicated process systems. Reactor design data are discussed. Analysis showed that these reactors can be prevented from going prompt critical, when fully flooded, by incorporating a tolerable amount of high resonance absorption materials such as hafnium or indium. An increase in reactivity on loss of coolant was indicated by preliminary calculations. (M.C.G.)
Sofer, G.; Hankel, R.; Goldstein, L.; Birman, G.
Nuclear and thermal analyses were performed to determine the characteristics of the Type 5 core in the SM-1 and SM-1A reactor plants as a function of geometry and composition. The following nuclear properties were investigated: core energy release, maximum midlife reactivity, average fuel burnup fraction, B-10 reactivity coefficient, and power distribution. Thermal parameter surveys determined the effects of channel thickness and power distribution upon the DNBR, nominal and hot channel thermal performance, and fuel plate thermal stress. From the nuclear and thermal analyses, a Type 5 core reference design was selected with fuel plates of 70-mil plate thick ness, 7-mil clad thickness, and 38 wt % UO/sub 2/ in the matrix, having initial core loading o4 108 Kg U/syup 235 and 260 gm B/sup 10/. (auth)
Davidson, S.L.; Paluszkiewicz, S.
Methodology for Developing the REScheckTM Software through Version 4.2
This report explains the methodology used to develop Version 4.2 of the REScheck software developed for the 1992, 1993, and 1995 editions of the MEC, and the 1998, 2000, 2003, and 2006 editions of the IECC, and the 2006 edition of the International Residential Code (IRC). Although some requirements contained in these codes have changed, the methodology used to develop the REScheck software for these five editions is similar. REScheck assists builders in meeting the most complicated part of the code?the building envelope Uo-, U-, and R-value requirements in Section 502 of the code. This document details the calculations and assumptions underlying the treatment of the code requirements in REScheck, with a major emphasis on the building envelope requirements.
Bartlett, Rosemarie; Connell, Linda M.; Gowri, Krishnan; Lucas, R. G.; Schultz, Robert W.; Taylor, Zachary T.; Wiberg, John D.
A set of 55 experiments hss been outiined to provide a minimum irradiation program for selection of UO/sub 2/, pellet geometry and fabricntion techniques, and canning technology. These experiments fall into three catagories: prototype: untts in which radial dimension and heat fluxes sre close to proposed design values, but irradiation times are long; reduced-size prototype for accelerated tests in which most variables will be studied; and miniaurized pellet irradiation to obtain high burnup for fission gas release studies. Reactor space has been found generally available and several installations are now examining their capabilities to participate in the program. A tentative schedule has been drawn to illustrate the feasibility of the program. (auth)
Heavy Ion Beam in Resolution of the Critical Point Problem for Uranium and Uranium Dioxide
Important advantages of heavy ion beam (HIB) irradiation of matter are discussed in comparison with traditional sources - laser heating, electron beam, electrical discharge etc. High penetration length (~ 10 mm) is of primary importance for investigation of dense matter properties. This gives an extraordinary chance to reach the uniform heating regime when HIB irradiation is being used for thermophysical property measurements. Advantages of HIB heating of highly-dispersive samples are claimed for providing free and relatively slow quasi-isobaric heating without fast hydrodynamic expansion of heated sample. Perspective of such HIB application are revised for resolution of long-time thermophysical problems for uranium and uranium-bearing compounds (UO2). The priorities in such HIB development are stressed: preferable energy levels, beam-time duration, beam focusing, deposition of the sample etc.
Igor Iosilevskiy; Victor Gryaznov
Research Programme for the 660 Mev Proton Accelerator Driven MOX-Plutonium Subcritical Assembly
The paper presents a research programme of the Experimental Acclerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton acceletator operating at the Laboratory of Nuclear Problems of the JINR, Dubna. MOX fuel (25% PuO_2 + 75% UO_2) designed for the BN-600 reactor use will be adopted for the core of the assembly. The present conceptual design of the experimental subcritical assembly is based on a core of a nominal unit capacity of 15 kW (thermal). This corresponds to the multiplication coefficient k_eff = 0.945, energetic gain G = 30 and the accelerator beam power 0.5 kW.
Barashenkov, V S; Buttseva, G L; Dudarev, S Yu; Polanski, A; Puzynin, I V; Sissakian, A N
MOX Reprocessing at Tokai Reprocessing Plant
Science Conference Proceedings (OSTI)
In March 2007, the first reprocessing of the 'Type B' MOX spent fuels of the Prototype Advanced Thermal Reactor FUGEN was initiated at Tokai Reprocessing Plant as a plant-scale demonstration of MOX fuel reprocessing. The operation was advanced satisfactorily and it has been confirmed that the MOX fuels as well as UO{sub 2} fuels can be reprocessed safely. Some characteristics of MOX fuels on reprocessing, such as properties of undissolved residue affecting the clarification process, are becoming visible. Reprocessing of the 'Type B' MOX fuels will be continued for several more years from now on, further investigations on solubility of fuels, characteristics of undissolved residues, progress of solvent degradation and so on will be continued. (authors)
Taguchi, Katsuya; Nagaoka, Shinichi; Yamanaka, Atsushi; Nakamura, Yoshinobu; Omori, Eiichi [Tokai Reprocessing Technology Development Center, Japan Atomic Energy Agency 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan); SATO, Takehiko; MIURA, Nobuyuki [Nuclear Fuel Cycle Technology Development Directorate, Japan Atomic Energy Agency 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)
The PACSAT Communications Experiment (PCE)
While VITA (Volunteers in Technical Assistance) is the recognized world leader in low earth orbiting (LEO) satellite technology (below 1 GHz), its involvement in communications technologies is to facilitate renewable energy technology transfer to developing countries. A communications payload was incorporated into the UoSat 2 satellite (Surrey Univ., UK), launched in 1984; a prototype satellite (PCE) was also launched Jan 1990. US DOE awarded a second grant to VITA to design and test the prototype ground stations (command and field), install field ground stations in several developing country sites, pursue the operational licensing process, and transfer the evaluation results to the design of an operating system. This report covers the principal tasks of this grant.
The PACSAT Communications Experiment (PCE). Final report, August 13, 1990--February 12, 1992
While VITA (Volunteers in Technical Assistance) is the recognized world leader in low earth orbiting (LEO) satellite technology (below 1 GHz), its involvement in communications technologies is to facilitate renewable energy technology transfer to developing countries. A communications payload was incorporated into the UoSat 2 satellite (Surrey Univ., UK), launched in 1984; a prototype satellite (PCE) was also launched Jan 1990. US DOE awarded a second grant to VITA to design and test the prototype ground stations (command and field), install field ground stations in several developing country sites, pursue the operational licensing process, and transfer the evaluation results to the design of an operating system. This report covers the principal tasks of this grant.
Progress is reported on the following investigations: pilot-plant evaluation of U ore concentrates; low-acid extraction of U; scrub-column operation, corrosion of reactor materials in sparge tanks containing UNH; effect of UO/sub 2/F/sub 2/ content on UF/sub 4/ reduction; thermal densification of UF/ sub 4/,; operation of the moving bed reactor; reduction of UF/sub 4/ with Mg; development of a mold insulator; preparation of Th (C/sub 2/O/sub 4/)/sub 2/ and ThCI/sub 4/; production of Th metal; ore resistant t transform ation of U; effect of H/sub 2/ content of slug canning behavior; centrifugal casting of slugs; determination of Zr and Mo in U alloys; and analysis of U and Th ores for rare earths. (W.L.H.)
Simmons, J.W. ed.
Selection of LEU/Th reference fuel for the HTGR-SC/C lead plant
Science Conference Proceedings (OSTI)
This paper describes the reference fuel materials for the high-temperature gas-cooled reactor (HTGR) plant for steam cycle/cogeneration (SC/C). A development and testing program carried out in 1978 through 1982 led to the selection of coated fuel particles of uranium-oxycarbide (UCO) for fissile materials and thorium oxide (ThO/sub 2/) for fertiel materials. Low-enriched uranium (LEU) is the enrichment basis for the HTGR-SC/C application. While UC/sub 2/ and UO/sub 2/ would also meet the essential criteria for fissile fuel, the UCO, alternative was selected on the basis of improved performance, economics, and process conditions.
Turner, R.F.; Neylan, A.J.; Baxter, A.M.; McEachern, D.W.; Stansfield, O.M.
A process for preparing UF/sub 6/ by reacting UF/sub 4/ and oxygen is described. The UF/sub 4/ and oxygen are continuously introduced into a fluidized bed of UO/sub 2/F/sub 2/ at a temperature of 600 to 900 deg C. The concentration of UF/sub 4/ in the bed is maintained below 25 weight per cent in order to avoid sintering and intermediate compound formation. By-product U0/sub 2/F/sub 2/ is continuously removed from the top of the bed recycled. In an alternative embodiment heat is supplied to the reaction bed by burning carbon monoxide in the bed. The product UF/sub 6/ is filtered to remove entrained particles and is recovered in cold traps and chemical traps. (AEC)
Adams, J.B.; Bresee, J.C.; Ferris, L.M.
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)
Material balance in atmospheric-pressure water and steam corrosion of uranium have been studied by examination of the phase composition and valence state of the corrosion product and by hydrogen-evolution measurements. The corrosion rates in atmospheric-pressure steam above 100 deg C are lower than those obtained in tests carried out in water with a hydrogen overpressure. The atmospheric-pressure-water corrosion product was found to be two phase: an oxygen- rich oxide, UO/sub 2.2/, and uncorroded metal particles. No hydride phase was detected, in contrast to previously reported evidence for hydride in uranium corrosion. The differences are explained on the basis of hydrogen pressure in the reaction vessel. (auth)
Stewart, O.M.; Berry, W.E.; Miller, P.D.; Vaughan, D.A.; Schroeder, J.B.; Fink, F.W.; Schwartz, C.M.
VARIABLE MODERATOR REACTOR DEVELOPMENT PROGRAM. Quarterly Progress Report No. 1
Development of the boiling water UO/sub 2/ fueled Variable Moderator Reactor (VMR) is conducted under contract for the USAEC. The initiation and progress of work under Phase I of the contract, Physics and Kinetic Analysis and Initial Evaluation,'' and the preparation for Phase II, Critical Experiment and Analysis of Results,'' are reported. A hydrodynamic flow sheet representing the sequence of calculations for the BOCH program was prepared. A preliminary block diagram of the kinetics model of the VMR was prepared. Work is reported on the PUREE code which is designed to give an accurate representation of the physics of the VMR core. A fuel element fabrication speciftcation was prepared and released for quotations. A study was made to select the most appropriate material for void simulation throughout the range of interest in the VMR. (W.D.M.)
Explosive stimulation of a geothermal well: GEOFRAC
This paper describes the first known explosive stimulation successfully conducted in a geothermal well. Two tests were performed in a 2690-meter-(8826-ft.) deep Union Oil well at the Geysers field in Northern California in December 1981. The heat-resistant process, called GEOFRAC, uses a new unique, explosive HITEX 2, which is a nondetonable solid at room temperature. Upon melting at a temperature of 177[degrees]C (350[degrees]F), the HITEX 2 liquid becomes an explosive that can be safely heated to temperatures greater than 260[degrees]C (500[degrees]F). These unique properties of the explosive were exploited in the GEOFRAC process through the cooperative efforts of Physics International Company (PI), Rocket Research Company (RRC), Union oil Company (UO), and the university of California Los Alamos National Laboratories (LANL).
Mumma, D.M. (Physics International Co., San Leandro, CA (United States))
Science Conference Proceedings (OSTI)
In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.
Ade, Brian J [ORNL; Gauld, Ian C [ORNL
Method for cleaning solution used in nuclear-fuel reprocessing. [DOE patent application
A nuclear fuel processing solution containing: (1) hydrocarbon diluent; (2) tri-n-butyl phosphate or tri-2-ethylhexyl phosphate; and (3) monobutyl phosphate, dibutyl phosphate, mono-2-ethylhexyl phosphate, di-2-ethylhexyl phosphate, or a complex formed by plutonium, uranium, or a fission product thereof with monobutyl phosphate, dibutyl phosphate, mono-2-ethylhexyl phosphate, or di-2-ethylhexyl phosphate is contacted with silica gel having alkali ions absorbed thereon to remove any degradation products from said solution. The principal impurities removed from TBP solvent by the process of this invention are monobutyl phosphate, dibutyl phosphate, UO/sub 2//sup 2 +/, Pu/sup 4 +/, and fission products of plutonium and uranium complexed with monobutyl phosphate or dibutyl phosphate. Nitric acid is also removed from the TBP solution by the treated silica gel. Conventional adsorption column techniques are applicable for the process of the invention.
Tallent, O.K.; Dodson, K.E.; Mailen, J.C.
Science Conference Proceedings (OSTI)
Using spectroscopic data produced in the experimental investigations of bulk systems, including X-Ray Absorption Spectroscopy (XAS), Photoelectron Spectroscopy (PES) and Bremstrahlung Isochromat Spectroscopy (BIS), the theoretical results within for UO{sub 2}{sup 6}, PuO{sub 2}{sup 6} and Pu{sup 7} clusters have been evaluated. The calculations of the electronic structure of the clusters have been performed within the framework of the Relativistic Discrete-Variational Method (RDV). The comparisons between the LLNL experimental data and the Russian calculations are quite favorable. The cluster calculations may represent a new and useful avenue to address unresolved questions within the field of actinide electron structure, particularly that of Pu. Observation of the changes in the Pu electronic structure as a function of size suggests interesting implications for bulk Pu electronic structure.
Tobin, J G; Yu, S W; Chung, B W; Ryzhkov, M V; Mirmelstein, A
Heat pipe nuclear reactor for space power
A heat-pipe cooled nuclear reactor has been designed to provide 3.2 MW(t) to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat pipe temperature of 1675/sup 0/K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum, lithium vapor, heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO/sub 2/ pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber and a BeO reflector containing boron loaded control drums.
Koenig, D.R.
Microorganisms present in the natural environment play a significant role in the mobilization and immobilization of uranium. Fundamental understanding of the mechanisms of microbiological transformations of various chemical forms of uranium present in wastes and contaminated soils and water has led to the development of novel bioremediation processes. One process uses anaerobic bacteria to stabilize the radionuclides and toxic metals from the waste, with a concurrent reduction in volume due to the dissolution and removal of nontoxic elements from the waste matrix. In an another process, uranium and other toxic metals are removed from contaminated soils and wastes by extracting with the chelating agent citric acid. Uranium is recovered from the citric acid extract after biodegradation/photodegradation in a concentrated form as UO{sub 3} {center_dot} 2H{sub 2}O for recycling or appropriate disposal.
Science Conference Proceedings (OSTI)
Los Alamos National Laboratory is engaged in producing mixed actinide (i.e., U, Np, Pu, and Am) oxides to study candidates for nuclear fuels. Correlation of composition and processing technique with initial morphology and crystallographic structure is critical to understanding and predicting the performance of these fuels. In this presentation, I will communicate the results of characterization of fuels ranging in actinide composition from UO{sub 2}, U{sub 0.8}Pu{sub 0.2} to U{sub 0.75}Np{sub 0.02}Pu{sub 0.2}Am{sub 0.03} via Resonant Ultrasound Spectroscopy (RUS) for recently fabricated fuel candidates.
Saleh, Tarik A [Los Alamos National Laboratory; Luther, Erik P [Los Alamos National Laboratory; Safarik, Douglas J [Los Alamos National Laboratory; Ulrich, Timothy J [Los Alamos National Laboratory; Byler, D D [Los Alamos National Laboratory; Freibert, F J [Los Alamos National Laboratory; Willson, S P [Los Alamos National Laboratory
Uranium-contaminated soils: Ultramicrotomy and electron beam analysis
Science Conference Proceedings (OSTI)
Uranium-contaminated soils from the U.S. Department of Energy (DOE) Fernald Site, Ohio, have been examined by a combination of scanning electron microscopy with backscattered electron imaging (SEM/BSE) and analytical electron microscopy (AEM). The inhomogeneous distribution of particulate uranium phases in the soil required the development of a method for using ultramicrotomy to prepare transmission electron microscopy (TEM) thin sections of the SEM mounts. A water-miscible resin was selected that allowed comparison between SEM and TEM images, permitting representative sampling of the soil. Uranium was found in iron oxides, silicates (soddyite), phosphates (autunites), and fluorite (UO{sub 2}). No uranium was detected in association with phyllosilicates in the soil.
Buck, E.C.; Dietz, N.L.; Bates, J.K.; Cunnane, J.C.
Preliminary results of the TRAN experiment B-1 (annular channel)
The current safety case for CRBR assumes that molten fuel will penetrate long distances through steel structures (such as the gaps between subassembly can walls in the axial and radial blankets), thus removing fuel and mitigating or terminating the accident. This process involves freezing of molten core materials on flat surfaces and curved surfaces of positive and negative curvature. Phenomenological data have been obtained on the freezing of thermite simulant in pin-bundle geometries and in tubes, and on the freezing of pure UO/sub 2/ in tubes. There is reason to believe that fuel crusts will be more stable in tubes because the tube wall holds the crust in compression. On the other hand, in flat-plate geometry or on the outside of fuel pins, the crust may not be stable since it is not in compression.
McArthur, D.A.; Mast, P.K.
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
PRIVACY IMPACT ASSESSMENT: INL Education Programs PIA Template
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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,
Microsoft Word - ICEM05_Np_Sorption_paper.doc
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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
CX-009241: Categorical Exclusion Determination | Department of Energy
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Microsoft Word - 1aDOE-ID-12-047 Westinghouse EC B3-6 NRC.doc
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
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
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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
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study |
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
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
Microsoft Word - DOE-ID-13-071 Hunter College EC B3-6.doc
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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
Presentations | MMSNF 2013 Chicago
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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
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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
CMSNF | U.S. DOE Office of Science (SC)
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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
BISON Enhanced | Department of Energy
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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
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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.
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Pierluigi Mancarella Pierluigi Mancarella Lecturer Sustainable Energy Systems, School of Electrical and Electronic Engineering University of Manchester, UK firstname.lastname@example.com 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
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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
Solubility limits of importance to leaching
This project developed from the Oklo natural fission reactor studies. It had been determined in the Oklo studies that many fission products and actinides remained in the reactor site during the periods of their radioactive decay following formation in the reactor zone two billion years ago. An explanation for this retention of fission products and actinides uses the extreme insolubility of uraninite (UO/sub 2/) in very reducing water environments. One can estimate from available thermodynamic data that the concentration of uranium in equilibrium with uraninite in pH 7 water that is free of dissolved oxygen is approx. 7 x 10/sup -6/ ppM. This low value suggested that the reducing conditions that can occur in deep geologic burial would result in a very slow leaching of spent fuel elements in contact with water since spent fuel elements are largely sintered UO/sub 2/. Studies on the leaching of spent fuel elements were conducted to verify this phenomenon. Results of the studies show that the solubilities of some radionuclides, especially rare earths and actinides, may be an important and controlling factor in leaching of waste forms. These solubilities should be measured accurately as a function of pH and not as a part of a multicomponent system. Although the amount of data is small it is interesting to postulate that a negative temperature coefficient of solubility is being exhibited by the actinides and rare earths. Individual solubilities should be measured as a function of temperature to determine if a kinetic effect is being observed in the data. A negative temperature coefficient of solubility for actinides and rare earths in water would have important consequences for nuclear reactor safety and for the management of nuclear wastes.
Ogard, A.; Bentley, G.; Bryant, E.; Duffy, C.; Grisham, J.; Norris, E.; Orth, C.; Thomas, K.
A 218 neutron group master cross section library for criticality safety studies
The AMPX system was used to generate a P$sub 3$ 218 neutron group master cross-section library from ENDF/B-IV data for the fuel, structural, and neutron- absorbing materials tabulated. The library is the data base for the generation of broad-group cross sections for shipping cask calculations and other criticality safety analyses using codes such as KENO and ANISN. Selection of the fine-group energy structure for the 3-eV to 20-MeV energy range included consideration of the resonance structure of prominent nuclei, the thresholds of important reactions, and the fission spectra. For 10$sup -5$ less than or equal to E/sub n/ less than 3 eV, 78 closely spaced thermal groups were chosen to examine the effects of low-energy resonances and thermal-neutron upscatter. Distribution of the 218 groups within the Hansen-Roach 16-group boundaries is shown. Adequacy of the group structure and validity of selected data sets from the library were tested by P$sub 3$S$sub 8$ XSDRNPM calculations of k-eff for two benchmark critical experiments; first, the 4.9 percent-enriched UO$sub 2$F$sub 2$- H$sub 2$O solution critical sphere experiment of Johnson and Cronin was analyzed, and, second, the 93.2 percent-enriched UO$sub 2$F$sub 2$-H$sub 2$O solution critical sphere experiment of Fox was analyzed. (auth)
Ford, W.E. III; Westfall, R.M.; Webster, C.C.
Fabrication of Uranium Oxycarbide Kernels for HTR Fuel
Science Conference Proceedings (OSTI)
Babcock and Wilcox (B&W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-�m, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B&W produced 425-�m, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B&W also produced 500-�m, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B&W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.
Charles Barnes; CLay Richardson; Scott Nagley; John Hunn; Eric Shaber
Science Conference Proceedings (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} 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 safetime,'' 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. 2 refs., 5 figs., 1 tab.
Newvahner, R.L. (Portsmouth Gaseous Diffusion Plant, OH (United States)); Pryor, W.A. (PAI Corp., Oak Ridge, TN (United States))
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
Nuclear fuels technologies fiscal year 1998 research and development test plan
Science Conference Proceedings (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
Surface studies of UFe2 and evaluation of its catalytic properties with a 2H2:CO mixture
The reactivity of UFe2 with O2, CO and CO2 were studied using x-ray photoelectron spectroscopy (XPS). Adsorption of O2 on clean UFe2 surfaces (Fe/U approx. = 2.0), produced by argon-ion sputtering, leads to the formation of UO2 and depletion of Fe from the surface layer probed by XPS (Fe/U approx. = 0.8). The oxidation state of Fe in this layer, as determined by XPS (Fe 2p/sub 3/2/ = 710.4 eV), is between Fe S and Fe T of pure Fe oxides. Exposure of sputtered-clean UFe2 to CO and CO2 results in a slight broadening of the U 4f peaks, indicating U oxidation, and some Fe depletion in the analyzed layer (Fe/U approx. = 1.7). The O ls (530.2 and 530.4 eV for CO and CO2, respectively) and C ls (282.7 and 282.6 eV for CO and CO2, respectively) indicate that dissociative chemisorption to O and C atoms occurs. UFe2 ground into a fine powder was tested as a catalyst in a differential high-pressure flow reactor with a 2H2:CO gas mixture. A significant amount of methanol and hydrocarbons are produced at 577K; while hydrocarbons are the main products (>99%) at 739K. XPS analysis of the used catalyst indicates that U is present as UO/sub 2+x/ and Fe as Fe2O3.
Schultz, J.; Naegele, J.; Spirlet, J.C.; Colmenares, C.
Design of Mega-Voltage X-ray Digital Radiography and Computed Tomography Performance Phantoms
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
Simulations of the Thermodynamic and Diffusion Properties of Actinide Oxide Fuel Materials
Spent nuclear fuel from commercial reactors is comprised of 95-99 percent UO{sub 2} and 1-5 percent fission products and transuranic elements. Certain actinides and fission products are of particular interest in terms of fuel stability, which affects reprocessing and waste materials. The transuranics found in spent nuclear fuels are Np, Pu, Am, and Cm, some of which have long half- lives (e.g., 2.1 million years for {sup 237}Np). These actinides can be separated and recycled into new fuel matrices, thereby reducing the nuclear waste inventory. Oxides of these actinides are isostructural with UO{sub 2}, and are expected to form solid solutions. This project will use computational techniques to conduct a comprehensive study on thermodynamic properties of actinide-oxide solid solutions. The goals of this project are to: Determine the temperature-dependent mixing properties of actinide-oxide fuels; Validate computational methods by comparing results with experimental results; Expand research scope to complex (ternary and quaternary) mixed actinide oxide fuels. After deriving phase diagrams and the stability of solid solutions as a function of temperature and pressure, the project team will determine whether potential phase separations or ordered phases can actually occur by studying diffusion of cations and the kinetics of potential phase separations or ordered phases. In addition, the team will investigate the diffusion of fission product gases that can also have a significant influence on fuel stability. Once the system has been established for binary solid solutions of Th, U, Np, and Pu oxides, the methodology can be quickly applied to new compositions that apply to ternaries and quaternaries, higher actinides (Am, Cm), burnable poisons (B, Gd, Hf), and fission products (Cs, Sr, Tc) to improve reactivity.
Becker, Udo [Univ. of Michigan (United States)
Two Trends of Composition Variation of Zircons and Their Significance in Origin Discrimination
Zircons can crystallize in a wide range of physical and chemical conditions. At the same time, they have high stability and durability. Therefore zircons can grow and survive in a variety of geological processes. In addition, the diffusivity of chemical compositions in their crystals is very low. Consequently,we can trace back the evolution history of the planetary materials containing zircon by zircon U-Th-Pb geochronology and geochemistry studies. However, this depends on our ability to decipher its genesis,namely magmatic or metamorphic origins. In this paper, we have found that there are obvious differences between magmatic and metamorphic zircons in their chemical composition zonations. The magmatic zircons exhibit composition zonation of increasing HfO2, and (UO2 + ThO2) content and decreasing ZrO2/HfO2 ratio and ZrO2 content from inner to outer parts within each growth zone or from core to rim of a crysta1. The metamorphic zircons exhibit compositional variation trend opposite to that of magmatic (igneous) zircons,tending to decrease in HfO2, (UO2+ ThO2)and increase in ZrO2/HfO2 ratio and ZrO2 from core to rim of a crystal. These chemical composition variation trends are thought to be controlled by the crystal chemical features of ions themselves and the evolution trends of magmatism and metamorphism respectively, and can be used to identify the genesis of zircons. Their morphological features are also discussed.
Xuezhao Bao
Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor
Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.
Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer
Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) � only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO2 and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO2 and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior is dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.
Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer
Interaction of hot solid core debris with concrete
The Hot Solid program is intended to measure, model, and assess the thermal, gas evolution, and fission product source terms produced as a consequence of hot, solid, core debris-concrete interactions. Two preliminary experiments, HSS-1 and HSS-3, were performed in order to compare hot solid UO/sub 2/-concrete and hot solid steel-concrete interactions. The HSS-1 experiment ablated 6 cm of limestone-common sand concrete in a little more than three hours using a 9 kg slug of 304 stainless steel at an average debris temperature of 1350/sup 0/C. The HSS-3 experiment ablated 6.5 cm of limestone-common sand concrete in four hours using a 10 kg slug of 80% UO/sub 2/-20% ZrO/sub 2/ at an average debris temperature of 1650/sup 0/C. Both experiments were inductively heated and contained in a 22 cm alumina sleeve to simulate one-dimensional axial erosion. The HOTROX computer code model was evaluated using the results from the HSS tests. HOTROX is a 1-D concrete ablation model that calculates transient conduction and gas release in the concrete as well as heatup of the hot solid slug. Using the HSS-1 power input history and geometry, HOTROX calculates 6.2 cm of concrete erosion in 200 minutes. Using the HSS-3 input conditions, HOTROX predicts 6.8 cm of erosion in 190 minutes. These results compare favorably with the experimental erosion rates. The calculated thermal response of the concrete is also close to experimentally measured values. The information from the Hot Solid Program will be used both to expand the post-accident phenomena data base and to extend the range of applicability of current accident analysis computer models such as CORCON and CONTAIN.
Copus, E.R.; Bradley, D.R.
ORNL fission product release tests VI-6
The ORNL fission product release tests investigate release and transport of the major fission products from high-burnup fuel under LWR accident conditions. The two most recent tests (VI-4 and VI-5) were conducted in hydrogen. In three previous tests in this series (VI-1, VI-2, and VI-3), which had been conducted in steam, the oxidized Zircaloy cladding remained largely intact and acted as a barrier to steam reaction with the UO{sub 2}. Test VI-6 was designed to insure significant oxidation of the UO{sub 2} fuel, which has been shown to enhance release of certain fission products, especially molybdenum and ruthenium. The BR3 fuel specimen used in test VI-6 will be heated in hydrogen to 2300 K; the Zircaloy cladding is expected to melt and runoff at {approximately}2150 K. Upon reaching the 2300 K test temperature, the test atmosphere will be changed to steam, and that temperature will be maintained for 60 min, with the three collection trains being operated for 2-, 18-, and 40-min periods. The releases of {sup 85}Kr and {sup 137}Cs will be monitored continuously throughout the test. Posttest analyses of the material collected on the three trains will provide results on the release and transport of Mo, Ru, Sb, Te, Ba, Ce, and Eu as a function of time at 2300 K. Continuous monitoring of the hydrogen produced during the steam atmosphere period at high temperature will provide a measure of the oxidation rate of the cladding and fuel. Following delays in approval of the safety documentation and in decontamination of the hot cell and test apparatus, test VI-6 will be conducted in late May.
Osborne, M.F.; Lorenz, R.A.; Collins, J.L.; Lee, C.S.
Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR
Science Conference Proceedings (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
The US Department of Energy`s former uranium production facility located at Fernald, OH (18 mi NW of Cincinnati) is the host site for an Integrated Demonstration for remediation of uranium-contaminated soils. A wide variety of source terms for uranium contamination have been identified reflecting the diversity of operations at the facility. Most of the uranium contamination is contained in the top {approximately}1/2 m of soil, but uranium has been found in perched waters indicating substantial migration. In support of the development of remediation technologies and risk assessment, we are conducting uranium speciation studies on untreated and treated soils using molecular spectroscopies. Untreated soils from five discrete sites have been analyzed. We have found that {approximately}80--90% of the uranium exists as hexavalent UO{sub 2}{sup 2+} species even though many source terms consisted of tetravalent uranium species such as UO{sub 2}. Much of the uranium exists as microcrystalline precipitates (secondary minerals). There is also clear evidence for variations in uranium species from the microscopic to the macroscopic scale. However, similarities in speciation at sites having different source terms suggest that soil and groundwater chemistry may be as important as source term in defining the uranium speciation in these soils. Characterization of treated soils has focused on materials from two sites that have undergone leaching using conventional extractants (e.g., carbonate, citrate) or novel chelators such as Tiron. Redox reagents have also been used to facilitate the leaching process. Three different classes of treated soils have been identified based on the speciation of uranium remaining in the soils. In general, the effective treatments decrease the total uranium while increasing the ratio of U(IV) to U(VI) species.
Allen, P.G.; Berg, J.M.; Chisholm-Brause, C.J.; Conradson, S.D.; Donohoe, R.J.; Morris, D.E.; Musgrave, J.A.; Tait, C.D.
Science Conference Proceedings (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
Science Conference Proceedings (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 accident-tolerant 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 rate 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. (authors)
Sen, R. S.; Pope, M. A.; Ougouag, A. M.; Pasamehmetoglu, K. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Venneri, F. [UltraSafe Nuclear (United States)
Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor
Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)
Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Boer, B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); SCK-CEN, Boertang 200, BE-2400 Mol (Belgium)
Closed ThUOX Fuel Cycle for LWRs with ADTT (ATW) Backend for the 21st Century
Science Conference Proceedings (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.
Unsteady-state material balance model for a continuous rotary dissolver
Science Conference Proceedings (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.
Isostatic-pressing techniques using reversible gels as the pressing medium were studied for improving the uniformity in density and structure of UC-- ZrC bodies. Control of powder-size fraction was studied as a means for controlling the pore distribution. Control of the carbon content by thermal treatment in a mixture of H/sub 2/ and hydrocarbon was also studied. Near stoichiometric 30 UC--70 ZrC powder was prepared by gas-metal reaction. Assembly of an apparatus for study of the thermochemical vapor-deposition of tungsten is near completion. The vaporization and fission product release rates of a hot- pressed high-density 30 UC -70 ZrC sample were measured from 1800 to 2000 deg C. A low-pressure gas adsorption apparatus was set up for measuring the true surface area of UC-ZrC samples. The cell used for the study of fission product diffusion through tungsten was fabricated. Samples are being prepared for fuel-clad compatibility and refractory-metal interdiffusion studies. Diffusion-emission studies were made on a rhenium-clad UC sample at 1800 deg C. The molybdenum pedestal of the loading device of a high-temperature mechanical testing furnace was modified. Thermionic emission microscopy showed that areas of high work function can co-exist with betteremitting UO/sub 2/ dispersions on the surface of a W-UO/sub 2/ cermet at 1650 deg C. Life-testing cells for fueled cesium converters using planar, as well as cylindrical fueled emitters were designed. The cylindrical configuration was favored, since a reliable cylindrical cell envelope was developed. (auth)
Reactor Materials and Components. The stresses required to produce creep rates of 0.001, 0.01, and O.l%/hr at 650, 800, and 1000 deg C were measured for Nb--2.37 wt.% Cr, Nb--3.34 wt.% Zr, snd Nb--5.21 wt.% V; anneals of these three alloys showed recrystallization only for heat treatment at 1260 deg C. Fabrication of miniature heating elements and thermocouples for a power-balancing thermal neutron fiux sensor is described. Fuels. Rolling propenties of Nb-10 wt.% U and Nb--20 wt.% U at 1000 deg F are described. Air oxidation tests were made at 1100 deg F for the following compounds: (Th/sub 9/U)C, ThC - lOUC - 5NbC, ThC - 10UC - 5SiC, ThC - 10UC - 2.5Mo/sub 2/C, ThC - 10UC - 5ZrC, (Th/sub 9/U)C/ sub 2/, ThC/sub 2/ - 10UC/sub 2/ - 5NbC, ThC/sub 2/ - 10UC/sub 2/ - 5SiC, ThC/sub 2/ - 10UC/sub 2/ 2.5Mo/sub 2/C, and ThC/sub 2/ - 10UC/sub 2/ - 5ZrC. Hardness measurements on Th--5 wt.% Pu snd Th--10 wt.% Pu show that complete recrystallization occurs within 1 min at 700 deg C. X-ray diffraction sthdies indicate No--25.6 wt.% Pu-6.1 wt.% Si to consist of free Nb and Nb/sub 5/Si/sub 3/ . Sintering and melting expeniments on PuO/sub 2/ in Ar and H/sub 2/ show that melting is caused by the existence of Pu/sub 2/O/sub 3/ in Ar or by reduction of PuO/sub 2/ to Pu/sub 2/O/sub 3/ in dry H/sub 2/. Compatibility tests with PuO/ sub 2/ pellets encased in types 304, 316, 318, and 347 stainless steel and Inconel, run for 2 hr at 1900 or 2000 deg F in H/sub 2/, gave no evidence of failure. An equation is presented which describes the in-pile steady-state release rate of fission products from dense plates. Fuel Element Development. A reaction layer was found at the interface between type 304 stainless steel cladding and UN-type 347 stainless steel cerments gas-pressure bonded at 2300 deg F. Fabrication of UN cerments is described. Experiments conducted with compacted mixtures of plastic binder, stainless steel powder, and UO/sub 2/ powder are described. Developmext of Uranium Carbide. Cores of UC and UC --10 wt.% Mo/sub 2/C were hot pressed at 1480 deg C, and density measurements before and after sintering are reported along with particle-size data. Uranium carbide melts were cast; analyses of castings gave 5.1 plus or minus 0.3 wt.% C. Phase studies of as-cast U--5.9 wt.% C and U--6.5 wt.% C were conducted which show UC --UC/sub 2/ mixtures and indicate a eutectlc in the UC--UC/sub 2/ system. Diffusion studies gave a preliminary value of 5 x 10/sup -//sup 8/ cm/sup 2//sec for the self- diffusion coefficient of C in UC at 1600 deg C. The research program on radiation effects on UC is described briefly. Growth of UO/sub 2/ Single Crystals. Experiments using a high-temperature tungsten reslstance furnace to grow single UO/sub 2/ crystals from melt are reported. It appears that tungsten is compatible with molten UO/sub 2/ and that dissociation of UO/sub 2/ does not occur up to its melting point. Other topics discussed in the abstract include: Radioisotope and Radiation Applications; Materials Evaluation; Coated-Particle Fuel Materials; Recovery of Spent Reactor Fuel Elements; Fueled Graphite Elements for Pebble-Bed Reactor; Cold Bonding of Zircaloy-2 to Type 410 Stainless Steel; Gas Cooled Reactor Program; Corrosion of Thorium and Uranium; and Gas-Pressure Bonding of Be-Clad Elements.
Dayton, R.W.; Tipton, C.R. Jr.
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
MDD Status Letter Report (AFCI CETE Milestone)
Current flow sheets for processing used nuclear fuels do not produce separated streams of all of the actinides. These aqueous processing streams must be converted into solid forms suitable for recycle (fuel/target fabrication), storage, or disposal, necessitating co-conversion. A process developed at ORNL in the 1980s to make UO{sub 3} suitable as fuel feedstock was studied for preparation of mixed actinide oxides with similarly favorable ceramic properties. The process, Modified Direct Denitration (MDD), uses ammonium nitrate to alter the thermal decomposition behavior of metal nitrates and improve the ceramic properties of the resulting solid oxide. Since plutonium (IV) and neptunium(IV) form compounds similar to uranium with the ammonium ion [(NH{sub 4}){sub 2}Pu(NO{sub 3}){sub 6}, (NH{sub 4}){sub 2}Np(NO{sub 3}){sub 6}], MDD-conversion of these metals was considered to be applicable. Co-conversion has advantages for making mixed oxides over individual element conversions that are followed by dry mixing of the oxide powders. Issues associated with preparing a mixture from individual oxides include use of additional equipment, dusting associated with feeding and milling, time requirements for milling, blending to obtain a uniform mixture, and inhomogenity at higher plutonium concentrations. These issues can be partially or wholly avoided by using MDD coconversion in which the mixing of the individual metals occurs in liquid solution; thus, adjusting relative metal concentrations is simpler and the resulting mixed oxide is more uniform than that produced by blending the individual oxides. Utilizing MDD also eliminates the need for mechanical treatment of the powder to obtain the desired ceramic properties, such as surface area and particle size distribution, since these characteristics are acceptable as-produced. The original MDD development work established that uranium oxide with good ceramic properties could be made. Following the discovery, a more fundamental understanding of the chemistry of the uranium-ammonium double nitrate salt was developed. Later pilot-scale studies produced kilogram quantities of UO{sub 3} using engineering-scale (1 kg/hour), continuously-operated equipment, while establishing the reliability of the process and equipment. The current work was performed in support of the Advanced Fuel Cycle Initiative (AFCI), utilizing glove-box-contained equipment (100 g/hour) to produce UO{sub 3}, PuO{sub 2}, and mixed oxides of uranium, plutonium, neptunium, and americium from a nitrate solution of those actinides. Then the MDD glove-box system was utilized in the Coupled-End-To-End (CETE) project to convert the U-Pu-Np and uranium product solutions into oxide powders. As part of the CETE project, a powder characterization laboratory was established in gloveboxes with instruments required for the determination of: (1) surface area by the BET methodology; (2) tap density by using a Quantachrome AutoTap; (3) flow properties by using a Freeman technology powder rheometer; (4) material composition and crystalline structure by using a powder X-ray diffractometer; (5) particle size distribution by using a laser light-scattering analyzer; and (6) imaging of the powders with a stereomicroscope. These instruments can be used to characterize the products and to determine the effects of MDD operating parameters on product powder morphology. Ultimately, the powder characteristics necessary to produce high-density, sintered MOX pellets can be determined.
Vedder, Raymond James [ORNL; Jubin, Robert Thomas [ORNL
Tentative creep data are reported for annealed Zircaloy-2 sheet. A program directed toward the development of corrosion-resistant welding alloys for use with vacuum-melted low-carbon Hastelloy F to contain HAPO spent-fuel-element decladding solutions was initiatBe. Research to develop more satisfactory fuels from the Al--U system is reported. The development of a radiometric method for the determination of CaO in Port land cement was completed. In the study of radiation-induced nitration of hydrocarbons, a series of thermal and irradiation runs was completed in the liquid phase of the HNO/sub 3/-- cyclohexane determine the effects of ultra-high preasure and high temperature on uranium oxides and on the reactions of uranium oxides with mixed oxides. The irradiationBurveillance program was continued on type 347 stainless steel. Tensile data are reported for Nb-base alloys. A summary is reported of corrosion results obtained on Nb alloys exposed in high-temperature water and steam. The creep properties of Zircaloy-2 during irradiation at elevated temperatures are being investigated. Corrosion data are reported for Nb--U alloys exposed to high-temperature water and NaK. A program devoted to the determination of causes of fission-gas release in UO/sub 2/ is reported. Cermet and ceramic-type fuels are being clad with Mo and Nb by the gas-pressure-bonding technique. Data are reported on the densification of UO/sub 2/ by various pressure -bending conditions. Methods of producing dense UC pellets by powdermetallurgy methods are being investigated. Techniques for the production of high-quality cast shapes of UC are being developed. The rates of interdiffusion of U and C in the U-monocarbide-dicarbide system and the rates of selfdiffusion of U and C- in UC are being investigated. Hydrogen migration in Kr under the influence of a thermal gradient is being studied. Neutron-activation and in-pile experiments are being conducted to determine fission-gas retention and the effect on radiation on fueled-graphite spheres. Chemical analysis of cold-rolled binary Ta alloys is reported. Data are presented on the fission-gas release from UC -- graphite, UC/sub 2/-- graphite, and UO/sub 2/-- BeO specimens during postirradiation heat treatment in vacuum at 1800 deg F for 24 hours. Techniques are being developed for the fabrication of fuel elements, suppressor components, and control rods for the SM-2 reactor. Studies directed toward the development of compact gas--cooled reactors are reported. Research on core materials in support of the MGCR program is in progress. (For preceding period see BMI-1377.) (W.L.H.)
Dayton, R.W.; Tipton, C.R. Jr.
This report presents the estimation of required ATR irradiation levels for the DB-FCM fuel design (fueled with Pu and MAs). The fuel and assembly designs are those considered in a companion report [R. S. Sen et al., FCR&D-2011- 00037 or INL/EXT-11-23269]. These results, pertaining to the DB-FCM fuel, are definitive in as much as the design of said fuel is definitive. In addition to the work performed, as required, for DB-FCM fuel, work has started in a preliminary fashion on single-cell UO2 and UN fuels. These latter activities go beyond the original charter of this project and although the corresponding work is incomplete, significant progress has been achieved. However, in this context, all that has been achieved is only preliminary because the corresponding fuel designs are neither finalized nor optimized. In particular, the UO2 case is unlikely to result in a viable fuel design if limited to enrichment at or under 20 weight % in U-235. The UN fuel allows reasonable length cycles and is likely to make an optimal design possible. Despite being limited to preliminary designs and offering only preliminary conclusions, the irradiation planning tasks for UO2 and UN fuels that are summarized in this report are useful to the overall goal of devising and deploying FCM-LWR fuel since the methods acquired and tested in this project and the overall procedure for planning will be available for planning tests for the finalized fuel design. Indeed, once the fuel design is finalized and the expected burnup level is determined, the methodology that has been assembled will allow the prompt finalization of the neutronic planning of the irradiation experiment and would provide guidance on the expected experimental performance of the fuel. Deviations from the expected behavior will then have to be analyzed and the outcome of the analysis may be corrections or modifications for the assessment models as well as, possibly, fuel design modifications, and perhaps even variation of experimental control for future experimental phases. Besides the prediction of irradiation times, preliminary work was carried out on other aspects of irradiation planning. In particular, a method for evaluating the interplay of depletion, material performance modeling and irradiation is identified by reference to a companion report. Another area that was addressed in a preliminary fashion is the identification and selection of a strategy for the physical and mechanical design of the irradiation experiments. The principal conclusion is that the similarity between the FCM fuel and the fuel compacts of the Next Generation Nuclear Plant prismatic design are strong enough to warrant using irradiation hardware designs and instrumentation adapted from the AGR irradiation tests. Modifications, if found necessary, will probably be few and small, except as pertains to the water environment and its implications on the use of SiC cladding or SiC matrix with no additional cladding.
Abderrafi M. Ougouag; R. Sonat Sen; Michael A. Pope; Brian Boer
Effects of deposition conditions on the properties of pyrolytic carbon deposited in a fluidized bed
Science Conference Proceedings (OSTI)
The high-density, isotropic pyrolytic carbon layer beneath the silicon carbide (IPyC) plays a key role in the irradiation performance of coated particle fuel. The IPyC layer protects the kernel from reactions with chlorine during deposition of the SiC layer, provides structural support for the SiC layer, and protects the SiC from fission products and carbon monoxide. The process conditions used by the Germans to deposit the IPyC coating produced a highly isotropic, but somewhat permeable IPyC coating. The permeability of the IPyC coating was acceptable for use with the dense German UO{sub 2} kernels, but may not be suitable when coating UCO kernels. The UCO kernels are typically more porous and thus have a larger surface area than UO{sub 2} kernels. The lower density and the higher surface area of UCO kernels could make them more susceptible to attack by HCl gas during the silicon carbide (SiC) coating process, which could result in heavy metal dispersion into the buffer and IPyC coatings and a higher level of as-manufactured SiC defects. The relationship between IPyC deposition conditions, permeability, and anisotropy must be understood and the appropriate combination of anisotropy and permeability for particle fuel containing UCO kernels selected. A reference set of processing conditions have been determined from review of historical information and results of earlier coating experiments employing 350 and 500 {micro}m UO{sub 2} kernels. It was decided that a limited study would be conducted, in which only coating gas fraction (CGF) and temperature would be varied. Coatings would be deposited at different rates and with a range of microstructures. Thickness, density, porosity and anisotropy would be measured and permeability evaluated using a chlorine leach test. The results would be used to select the best IPyC coating conditions for use with the available natural enrichment uranium carbide/uranium oxide (NUCO) kernels. The response plots from the investigation of the deposition of pyrolytic carbon in a fluidized bed graphically depict the relationships between processing parameters and coating properties. The additional figures present optical, scanning electron microscopy, and other images to highlight microstructural details. For the study, only two parameters (factors), coating gas fraction and deposition temperature, were varied. The plots reveal obvious trends and links between factors and responses. The dominant relationships determined by this study for this range of coating conditions are: (1) rate is dependent upon coating gas fraction or in other terms, reactant concentration; (2) density is controlled by deposition temperature; (3) efficiency is influenced by both CGF and temperature; (4) anisotropy is affect by CGF and temperature, however, the relationship is more complex than for other properties; (5) permeability is dependent upon deposition temperature (thus density); and (6) open porosity is affect by CGF thus is influenced by coating rate. The response plots can be used as 'maps' for the deposition process and are thus valuable for selecting coating conditions necessary to produce desired combinations of properties. The information is useful in predicting the effects of changes to processing on properties and is beneficial in optimizing the process and product properties. Although the study was limited to only two parameters, the information provides a foundation from which other aspects of the coating process can be more easily investigated.
Lowden, Richard Andrew [ORNL; Hunn, John D [ORNL; Nunn, Stephen D [ORNL; Kercher, Andrew K [ORNL; Price, Jeffery R [ORNL; Jellison Jr, Gerald Earle [ORNL
Drying results of K-Basin fuel element 3128W (run 2)
Science Conference Proceedings (OSTI)
An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-East Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of N-Reactor spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from an open K-East canister (3128W) during the first fuel selection campaign conducted in 1995, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. Although it was judged to be breached during in-basin (i.e., K-Basin) examinations, visual inspection of this fuel element in the hot cell indicated that it was likely intact. Some scratches on the coating covering the cladding were identified before the furnace test. The drying test was conducted in the Whole Element Furnace Testing System located in G-Cell within the PTL. This test system is composed of three basic systems: the in-cell furnace equipment, the system gas loop, and the analytical instrument package. Element 3128W was subjected to the drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step. Results of the Pressure Rise and Gas Evolution Tests suggest that most of the free water in the system was released during the extended CVD cycle (68 hr versus 8 hr for the first run). An additional {approximately}0.34 g of water was released during the subsequent HVD phase, characterized by multiple water release peaks, with a principle peak at {approximately}180 C. This additional water is attributed to decomposition of a uranium hydrate (UO{sub 4}{center_dot}4H{sub 2}O/UO{sub 4}{center_dot}2H{sub 2}O) coating that was observed to be covering the surface of the fuel element to a thickness of {approximately}1.6 mg/cm{sup 2}. A limited quantity of hydrogen ({approximately}9 mg) was also released during HVD, mainly at temperatures above 300 C, likely from hydride decomposition.
Abrefah, J.; Klinger, G.S.; Oliver, B.M.; Marshman, S.C.; MacFarlan, P.J.; Ritter, G.A. [Pacific Northwest National Lab., Richland, WA (United States); Flament, T.A. [Numatec Hanford Corp., Richland, WA (United States)
Alpha Radiolysis of Sorbed Water on Uranium Oxides and Uranium Oxyfluorides
The radiolysis of sorbed water and other impurities contained in actinide oxides has been the focus of a number of studies related to the establishment of criteria for the safe storage and transport of these materials. Gamma radiolysis studies have previously been performed on uranium oxides and oxyfluorides (UO{sub 3}, U{sub 3}O{sub 8}, and UO{sub 2}F{sub 2}) to evaluate the long-term storage characteristics of {sup 233}U. This report describes a similar study for alpha radiolysis. Uranium oxides and oxyfluorides (with {sup 238}U as the surrogate for {sup 233}U) were subjected to relatively high alpha radiation doses (235 to 634 MGy) by doping with {sup 244}Cm. The typical irradiation time for these samples was about 1.5 years, which would be equivalent to more than 50 years irradiation by a {sup 233}U sample. Both dry and wet (up to 10 wt % water) samples were examined in an effort to identify the gas pressure and composition changes that occurred as a result of radiolysis. This study shows that several competing reactions occur during radiolysis, with the net effect that only very low pressures of hydrogen, nitrogen, and carbon dioxide are generated from the water, nitrate, and carbon impurities, respectively, associated with the oxides. In the absence of nitrate impurities, no pressures greater than 1000 torr are generated. Usually, however, the oxygen in the air atmosphere over the oxides is consumed with the corresponding oxidation of the uranium oxide. In the presence of up to 10 wt % water, the oxides first show a small pressure rise followed by a net decrease due to the oxygen consumption and the attainment of a steady-state pressure where the rate of generation of gaseous components is balanced by their recombination and/or consumption in the oxide phase. These results clearly demonstrate that alpha radiolysis of either wet or dry {sup 233}U oxides will not produce deleterious pressures or gaseous components that could compromise the long-term storage of these materials.
Icenhour, A.S.
Atomistic Simulations of Mass and Thermal Transport in Oxide Nuclear Fuels
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
Thermal-conductivity and electrical-resistivity measurements were made on U-1.5 wt,% Zr, encapsulated in Zircaloy-2 with a NaK thermal bond. Data are presented on thermal conductivity measurements made on UO/sub2/ specimens. A method of sink-float measurement to identify factors affecting irradiation- induced volume changes in graphite was continued. The stress-rupture and creep properties of 15% cold-worked Zircaloy-2 are being investigated in the 290 to 400 deg C temperature range in vacuum. High-purity single crystals of Mo are being prepared for irradiation-damage studies. A feasibility study of the neutron- activation method for the elemental analysis of portland cement was completed and preliminary experiments were initiated. The chemical analyses of irradiated and nonirradiated cetane, stearic acid, and the corresponding urea complexes were continued. A study of the mechanisms by which U solidifies in a cylindrical mold is being made. Coatings of Mo compounds on Croloy 2 1/4 were studied as potential protective coatings against liquid Bi. Work on the study of valence effects of oxide additions to UO/sub 2/ indicates that additions of Y/sub 6/O/sub 3/ are better stabilizers than La/sub 2/O/sub 3/ in 3000 deg F air. Measurements of hydrogen-absorption isotherms for the Zr-25 wt.% U alloy were made. An investigation was made of the effect of irradiation damage on the mechanical properties of type 347 stainless steel. Niobium uranium alloys containing 20 to 70 wt.% U were prepared in small heats. A program aimed at improving the irradiation stability and corrosion resistance of Th-U alloys is being conducted. Fatigue studies of Inconel and INOR-8 are presented. A study of the kinetics and mechanism of the reaction of N/sub 2/ with Nb is presented. A study of the U-Nb constitutional diagram is being performed. Data are reported on the fabrication of fueled graphite spheres. Materials of construction for use in the various proposed processes for the recovery of spent reactor fuel elements are being evaluated. Titanium is being investigated for the Darex Process dissolver, Carpenter 20Cb, Ni-o-nel, and Illium 98 for Thorex and Sulfex solutions, and INOR- 1 and INOR8 for the fluoride-volatility processes. Properties of UC are being determined. The preparation of an initial series of Ta and Ta-W alloys for Pu- alloy compatibility evaluation is in progress. A study to determine the effect of neutron irradiation on the mechanical properties of Ta is reported. (For preceding period see BMI-1294.) (W.L.H.)
Dayton, R. W.; Tipton, Jr., C. R.
Progress in fuels and materials development is summarized. Major areas of investigation include a materials study by means of sample fuel plates containing uranium alloys or cermets, burnable poisons, non-uniform fuel and poison distributions and clad with various aluminum alloys; and an engineering study of fuel element geometries optimized in heat transfer, hydraulics, and materials strength. Up to 45 wt% U-Al alloys, 6 to 65 wt% UO/-Al and U3O6-Al dispersions, including enrichments ranging from 20% to 93%, were tested to 70% burnup in de-ionized water at 200 deg F in the MTR. Their performance at higher temperature is still being investigated. Test results for the MTR conditions indicate that all of the compositions investigated to date will successfully withstand even the longest irradiation at these conditions if properly fabricated. Some high strength aluminum alloy claddings, not yet fully tested, show some peculiar surface effects which may be related to corrosion. Metallographic studies of irradiated cermets reveal a reaction'' (diffusion) zone produced around UO/sub 2/ particles in contact with aluminum. These zones are being studied by means of x-ray diffraction, electron microscopy, and electron microprobe analysis. From engineering studies has come promise of improved heat removal and lower pumping requlrements for reactors through artificial roughening of fuel plates. Computer optimizatlon studies and hydraulic tests indicated 80% improvement in heat transfer or 60% less flow for the same heat load are obtainable for MTR conditions. Heat transfer test results from 0.110 x 2.624 ' electrically-heated channels using heat fluxes up to 2.88 x 10/sup 6/ Btu/hr-ft/ sup 2/, sgree better with correlations based on bulk temperatures than with the more widely used modified Colburn equation. In this range, a modifled Colburn equation with a 20% safety factor, as is presently used, seems adequate. However, an equation based on the bulk coolant temperature could be used employing a smaller safety factor because of its greater accuracy. ( auth)
Gibson, G.W.; Shupe, O.K.
Science Conference Proceedings (OSTI)
A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle costs are included in the analysis, with the fast reactors having a higher $/kw(e) capital cost than the LWRs, the overall busbar generation cost ($/MWh) for the closed cycles is approximately 12% higher than for the all-LWR once-through fuel cycle case, again based on the expected values from an uncertainty analysis. It should be noted that such a percentage increase in the cost of nuclear power is much smaller than that expected for fossil fuel electricity generation if CO2 is costed via a carbon tax, cap and trade regimes, or carbon capture and sequestration (CCS).
Williams, Kent Alan [ORNL; Shropshire, David E. [Idaho National Laboratory (INL)
Student Progress Report: Summer 2012
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
Advanced Fuel Cycle Scenarios with AP1000 PWRs and VHTRs and Fission Spectrum Uncertainties
Minimization of HLW inventories and U consumption are key elements guaranteeing nuclear energy expansion. The integration of complex nuclear systems into a viable cycle yet constitutes a challenging multi-parametric optimization problem. The reactors and fuel cycle performance parameters may be strongly dependent on minor variations in the system's input data. Proven discrepancies in nuclear data evaluations could affect the validity of the system optimization metrics. This study first analyzes various advanced AP1000-VHTR fuel cycle scenarios by assessing their TRU destruction and their U consumption minimization capabilities, and by computing reactor performance parameters such as the time evolution of the effective multiplication factor keff, the reactors' energy spectrum or the isotopic composition/activity at EOL. The performance metrics dependence to prompt neutron fission spectrum discrepancies is then quantified to assess the viability of one strategy. Fission spectrum evaluations are indeed intensively used in reactors' calculations. Discrepancies higher than 10% have been computed among nuclear data libraries for energies above 8MeV for 235U. TRU arising from a 3wt% 235U-enriched UO2-fueled AP1000 were incinerated in a VHTR. Fuels consisting of 20%, 40% and 100% of TRU completed by UO2 were examined. MCNPX results indicate that up to 88.9% of the TRU initially present in a VHTR fueled with 20% of TRU and 80% of ThO2 were transmuted. Additionally, the use of WgPu instead of RgPu should reduce the daily consumption of 235U by 1.3 and augment core lifetime. To estimate the system metrics dependence to fission spectrum discrepancies and validate optimization studies outputs, the VTHR 235U fission spectrum distribution was altered successively in three manners. keff is at worst lowered by 1.7% of the reference value and the energy spectrum by 5% between 50meV and 2MeV when a significantly distorted fission spectrum tail is used. 233U, 236Pu and 237Pu inventories and activities are multiplied by 263, 523 and 34 but are still negligible compared to 239Pu mass or the total activity. The AP1000-VHTR system is in conclusion not dependent on the selected fission spectrum variations. TRU elimination optimization studies in AP1000-VHTR systems will be facilitated by freeing performance metrics dependency from 1 input parameter.
Cuvelier, Marie-Hermine
Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows for surface alloying well in excess of the thermodynamically dictated solubility limits, an effect that is particularly relevant to this research due to the negligible solubility of boron and gadolinium in zirconium. University of Wisconsin is performing the near surface materials characterization and analysis, aiding Sandia in process optimization, and promoting educational activities. Westinghouse is performing process manufacturability and scale-up analysis and is performing autoclave testing of the surface treated samples. The duration of this NERI project is 2 years, from 9/2002 to 9/2004.
Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L
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.
Science Conference Proceedings (OSTI)
The Deposit Removal Project was undertaken with the support of the U. S. Department of Energy at the East Tennessee Technology Park (ETTP) formerly the Oak Ridge K-25 Site. The project team performed the safe removal of the hydrated uranyl fluoride (UO{sub 2}F{sub 2}) deposits from the K-29 Building of the former Oak Ridge Gaseous Diffusion Plant. The deposits had developed as a result of air leakage into UF{sub 6} gas process pipes; UO{sub 2}F{sub 2} became hydrated by moisture from the air and deposited inside the pipes. The mass, its distribution, and the hydrogen content [that is, the ratio of H to U (H/U)], were the key parameters that controlled the nuclear criticality safety of the deposits. Earlier gamma-ray spectrometry measurements in K-29 had identified the largest deposits in the building. The first and third largest deposits in the building were measured in this program. The first deposit, found in the Unit 2, Cell 7, B-Line Outlet process pipe (called the ''Hockey Stick'') was about 1,300 kg ({+-} 50% uncertainty) at 3.34 wt% {sup 235}U enrichment ({+-}50% uncertainty) and according to the gamma-ray spectroscopy was uniformly distributed. The second deposit (the third-largest deposit in the building), found in the Unit 2, Cell 6, A-Line Outlet process pipe (called the ''Tee-Pipe''), had a uranium deposit estimated to be about 240 kg ({+-} 50% uncertainty) at 3.4 wt % {sup 235}U enrichment ({+-} 20% uncertainty). Before deposit removal activities began, the Deposit Removal Project team needed to survey the inside of the pipes intrusively to assess the nuclear criticality safety of the deposits. Therefore, the spatial distribution of the deposits, the total uranium deposit mass, and the moderation level resulting from hydration of the deposits, all of which affect nuclear criticality safety were required. To perform the task safely and effectively, the Deposit Removal Project team requested that Oak Ridge National Laboratory (ORNL) characterize the two largest deposits with the {sup 252}Cf-source-driven transmission (CFSDT) technique, an active neutron interrogation method developed for use at the Oak Ridge Y-12 Plant to identify nuclear weapons components in containers. The active CFSDT measurement technique uses CFSDT time-of-flight measurements of prompt neutrons and gamma rays from an externally introduced {sup 252}Cf source.
Hannon, T.F.; Mihalczo, J.T.; Mullens, J.A.; Uckan, T.; Valentine, T.E.; Wyatt, M.S.
EDF Nuclear Power Plants Operating Experience with MOX fuel
Science Conference Proceedings (OSTI)
EDF started Plutonium recycling in PWR in 1987 and progressively all the 20 reactors, licensed in using MOX fuel, have been loaded with MOX assemblies. At the origin of MOX introduction, these plants operated at full power in base load and the core management limited the irradiation time of MOX fuel assemblies to 3 annual cycles. Since 1995 all these reactors can operate in load follow mode. Since that time, a large amount of experience has been accumulated. This experience is very positive considering: - Receipt, handling, in core behaviour, pool storage and shipment of MOX fuel; - Operation of the various systems of the plant; - Environment impact; - Radioprotection; - Safety file requirements; - Availability for the grid. In order to reduce the fuel cost and to reach a better adequacy between UO{sub 2} fuel reprocessing flow and plutonium consumption, EDF had decided to improve the core management of MOX plants. This new core management call 'MOX Parity' achieves parity for MOX and UO{sub 2} assemblies in term of discharge burn-up. Compared to the current MOX assembly the Plutonium content is increased from 7,08% to 8,65% (equivalent to natural uranium enriched to respectively 3,25% and 3,7%) and the maximum MOX assembly burn-up moves from 42 to 52 GWd/t. This amount of burn-up is obtained from loading MOX assemblies for one additional annual cycle. Some, but limited, adaptations of the plant are necessary. In addition a new MOX fuel assembly has been designed to comply with the safety criteria taking into account the core management performances. These design improvements are based on the results of an important R and D program including numerous experimental tests and post-irradiated fuel examinations. In particular, envelope conditions compared to MOX Parity neutronic solicitations has been extensively investigated in order to get a full knowledge of the in reactor fuel behavior. Moreover, the operating conditions of the plant have been evaluated in many details and finally no important impact is anticipated. The industrial maturity of plutonium recycling activities is fully demonstrated and a new progress can be done with a complete confidence. The licensing process of 'MOX Parity' core management is in progress and its implementation on the 20 PWR is now expected at mid 2007. (author)
Thibault, Xavier [EDF Generation, Tour EDF Part Dieu - 9 rue des Cuirassiers B.P.3181 - 69402 Lyon Cedex 03 (France)
Deep Burn Fuel Cycle Integration: Evaluation of Two-Tier Scenarios
Science Conference Proceedings (OSTI)
The use of a deep burn strategy using VHTRs (or DB-MHR), as a means of burning transuranics produced by LWRs, was compared to performing this task with LWR MOX. The spent DB-MHR fuel was recycled for ultimate final recycle in fast reactors (ARRs). This report summarizes the preliminary findings of the support ratio (in terms of MWth installed) between LWRs, DB-MHRs and ARRs in an equilibrium �two-tier� fuel cycle scenario. Values from literature were used to represent the LWR and DB-MHR isotopic compositions. A reactor physics simulation of the ARR was analyzed to determine the effect that the DB-MHR spent fuel cooling time on the ARR transuranic consumption rate. These results suggest that the cooling time has some but not a significant impact on the ARRs conversion ratio and transuranic consumption rate. This is attributed to fissile worth being derived from non-fissile or �threshold-fissioning� isotopes in the ARR�s fast spectrum. The fraction of installed thermal capacity of each reactor in the DB-MHR 2-tier fuel cycle was compared with that of an equivalent MOX 2-tier fuel cycle, assuming fuel supply and demand are in equilibrium. The use of DB-MHRs in the 1st-tier allows for a 10% increase in the fraction of fleet installed capacity of UO2-fueled LWRs compared to using a MOX 1st-tier. Also, it was found that because the DB-MHR derives more power per unit mass of transuranics charged to the fresh fuel, the �front-end� reprocessing demand is less than MOX. Therefore, more fleet installed capacity of DB-MHR would be required to support a given fleet of UO2 LWRs than would be required of MOX plants. However, the transuranic deep burn achieved by DB-MHRs reduces the number of fast reactors in the 2nd-tier to support the DB-MHRs �back-end� transuranic output than if MOX plants were used. Further analysis of the relative costs of these various types of reactors is required before a comparative study of these options could be considered complete.
S. Bays; H. Zhang; M. Pope
A concentration cell electrode was installed in a fritted glass surface and was used to measure the surface KCl concentration while water was being transpired through the surface into a mixed tank of 0.1 M KCl. The results from the first Fluorox run made with crude UF/sub 4/ showed that 85.3% of the theoretical amount of UF/sub 6/ was accounted for, with 17.9% being collected in cold traps and the remainder being consumed in various side reactions. Installation was completed of apparatus to study the electrolytic removal of nickel on an engineering scale from concentrated HRT fuel solution. An extremely low yield stress of 0.02 lb/sq ft was measured for a thoria-urania slurry containing 2.5 wt.% Al/sub 2/O/sub 3/ and spherical particles of 1.7 micron mean diameter. Preliminary data on the rate of uranium sorption on Dowex 21K from sulfate solutions were correlated with a simple spherical diffusion model. Laboratory scale studies to determine the effect of reflux time, HNO/sub 3/ concentration, and air sparge on chloride removal from APPR-type Darex dissolver product showed that an air sparge is definitely beneficial in chloride removal. Preliminary cyclic cleaning tests showed that boiling 25% caustic solution was adequate for cleaning type F (20 micron) porous metal filter elements fouled in the absence of filter aid with siliceous sludge from Darex solvent extraction feed solutions. Sections of unirradiated stainless steel-clad UO/sub 2/, sheared into lengths of 1/2 in., 1 in., 2 in., and 2 3/4 in., were leached free of UO/sub 2/ with 10 M HNO/sub 3/ in 30, 45, 60, and 75 min, respectively. The abrasive blade cost for the removal of inert end adapters from selected reactor assemblies ranges from 12 for the Gas-Cooled Reactor to 19 for Consolidated Edison. A zirconium dissolution was carried to completion at 600 ction prod- C in a NaF-LiF salt mixture containing initially 23 mol per cent ZrF/sub 4/ at an average rate of 1.3 mg/(cm/sup 2/)(min) with an HF feed rate of 2 lb/hr. The thermal conductivity and thermal diffusivity at the center of an 8-in. diameter cake of Darex waste calcined at about 900 ction prod- C were determined as a function of temperature. (For preceding period see CF-59-3-61.) (auth)
Bresee, J.C.; Haas, P.A.; Horton, R.W.; Watson, C.D.; Whatley, M.E.
Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, January-March 1979
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
The concentration gradients of uranyl ion in aqueous and organic solutions were analyzed by taking a macro photograph of the desired gradient by monochromatic (436 m mu ) light transmitted by the solution normal to the gradient in an appropriate diffusion cell. Two Druhm runs were terminated due to malfunction of the sodium metering system and the third run was terminated when the UF/sub 6/ nozzle ruptured. Calculations of particle temperature versus time relations for the flame denitration-calcination method of preparing metallic oxide from nitrate solutions indicate that the times required for heat transfer are controlled by the rate of radiant heat transfer to particle surfaces instead of by conductive heat transfer within the particles. A completed experimental study indicated that electrolysis in a cell with a mercury cathode and a platinum anode is a practical process for removing nickel from HRT fuel solution. The apparent diffusion coefficient of uranium loading on Dowex 21K was shown to be directly related to the resin size. An explosion of sufficient violence to blow apart the Pyrex pipe dissolver occurred during the fifth Darex dissolution of simulated SRE fuel probably from a rapid gas phase reaction between hydrogen and oxidizing gases such as NO/sub 2/. Materials handling flowsheets were completed for (A) decladding, washing, recanning and storing spent SRE uranium fuel slugs and (B) the shearing and leaching of stainless steel clad UO/sub 2/ and UO/sub 2/- ThO/sub 2/ fuels. A literature survey is being conducted dealing with reactor coolant and coolant loop contamination and decontamination. During run R-17 for calcination of evaporated Darex waste, the same as run R-16 which deformed the bottom calcination vessel except that one of the three added pressure probes was vibrated to keep it unplugged, the bottom of the calcination vessel did not deform, and there was no pressure indicated on the pressure probes during the test. (For preceding period see CF-59-7-58.) (auth)
Bresee, J.C.; Haas, P.A.; Horton, R.W.; Watson, C.D.; Whatley, M.E.
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.
Reaction of U-VI with titanium-substituted magnetite: Influence of Ti on U-IV speciation
Reduction of hexavalent uranium (UVI) to less soluble tetravalent uranium (UIV) through enzymatic or abiotic redox reactions has the potential to alter U mobility in subsurface environments. As a ubiquitous natural mineral, magnetite (Fe3O4) is of interest because of its ability to act as a rechargeable reductant for UVI. Natural magnetites are often impure with titanium, and structural Fe3+ replacement by TiIV yields a proportional increase in the relative Fe2+ content in the metal sublattice to maintain bulk charge neutrality. In the absence of oxidation, the Ti content sets the initial bulk Fe2+/Fe3+ ratio (R). Here, we demonstrate that Ti-doped magnetites (Fe3 xTixO4) reduce UVI to UIV. The UVI-Fe2+ redox reactivity was found to be controlled directly by R, but was otherwise independent of Ti content (xTi). However, in contrast to previous studies with pure magnetite where UVI was reduced to nanocrystalline uraninite (UO2), the presence of structural Ti (xTi = 0.25 0.53) results in the formation of UIV species that lack the bidentate U-O2-U bridges of uraninite. Extended x-ray absorption fine structure spectroscopic analysis indicated that the titanomagnetite-bound UIV phase has a novel UIV-Ti binding geometry, different from the coordination of UIV in the mineral brannerite (UIVTi2O6). The observed UIV-Ti coordination at a distance of 3.43 � suggests a binuclear corner-sharing adsorption/incorporation UIV complex with the solid phase. Furthermore, we explored the effect of oxidation (decreasing R) and solids-to-solution ratio on the reduced UIV phase. The formation of the non-uraninite UIV-Ti phase appears to be controlled by availability of surface Ti sites, rather than R. Our work highlights a previously unrecognized role of Ti in the environmental chemistry of UIV and suggests that further work to characterize the long-term stability of UIV phases formed in the presence of Ti is warranted.
Latta, Drew; Pearce, Carolyn I.; Rosso, Kevin M.; Kemner, Kenneth M.; Boyanov, Maxim I.
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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
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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
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Residential Energy Consumption Survey:
Gasoline and Diesel Fuel Update (EIA)
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Analysis of gamma-ray dosimetry experiments in the zero power MINERVE facility
Science Conference Proceedings (OSTI)
The objective of this study is to develop nuclear heating measurement methods in zero power experimental reactors. These developments contribute to the qualification of photonics calculation schemes for the assessment of gamma heating in the future Jules Horowitz Material Testing Reactor. This paper presents the analysis of thermoluminescent detector (TLD) experiments in the UO{sub 2} core of the MINERVE Research Reactor at the French Alternative Energies and Atomic Energy Commission center in Cadarache. The experimental sources of uncertainty in the gamma dose have been reduced by improving the measurement conditions and the repeatability of the calibration step for each individual TLD. The interpretation of these measurements needs to take into account the calculation of cavity correction factors related to calibration and irradiation configurations, as well as neutron correction calculations. These calculations are based on Monte Carlo simulations of neutron-gamma and gamma-electron transport coupled particles. The comparison between calculated and measured integral gamma-ray absorbed doses in the aluminum material surrounding the TLD shows that calculations slightly overestimate the measurement, with a calculated versus experimental ratio equal