High temperature radiance spectroscopy measurements of solid and liquid uranium and plutonium carbides

NASA Astrophysics Data System (ADS)

Manara, D.; De Bruycker, F.; Boboridis, K.; Tougait, O.; Eloirdi, R.; Malki, M.

2012-07-01

In this work, an experimental study of the radiance of liquid and solid uranium and plutonium carbides at wavelengths 550 nm ⩽ λ ⩽ 920 nm is reported. A fast multi-channel spectro-pyrometer has been employed for the radiance measurements of samples heated up to and beyond their melting point by laser irradiation. The melting temperature of uranium monocarbide, soundly established at 2780 K, has been taken as a radiance reference. Based on it, a wavelength-dependence has been obtained for the high-temperature spectral emissivity of some uranium carbides (1 ⩽ C/U ⩽ 2). Similarly, the peritectic temperature of plutonium monocarbide (1900 K) has been used as a reference for plutonium monocarbide and sesquicarbide. The present spectral emissivities of solid uranium and plutonium carbides are close to 0.5 at 650 nm, in agreement with previous literature values. However, their high temperature behaviour, values in the liquid, and carbon-content and wavelength dependencies in the visible-near infrared range have been determined here for the first time. Liquid uranium carbide seems to interact with electromagnetic radiation in a more metallic way than does the solid, whereas a similar effect has not been observed for plutonium carbides. The current emissivity values have also been used to convert the measured radiance spectra into real temperature, and thus perform a thermal analysis of the laser heated samples. Some high-temperature phase boundaries in the systems U-C and Pu-C are shortly discussed on the basis of the current results.

Electron-beam-driven RI separator for SCRIT (ERIS) at RIKEN RI beam factory

NASA Astrophysics Data System (ADS)

Ohnishi, T.; Ichikawa, S.; Koizumi, K.; Kurita, K.; Miyashita, Y.; Ogawara, R.; Tamaki, S.; Togasaki, M.; Wakasugi, M.

2013-12-01

We constructed a radioactive isotope (RI) separator named ERIS (electron-beam-driven RI separator for SCRIT) for the SCRIT (Self-Confinement RI Target) electron scattering facility at RIKEN RI Beam Factory (RIBF). In ERIS, production rate of fission products in the photofission of uranium is estimated to be 2.2 ×1011 fissions/s with 30 g of uranium and a 1-kW electron beam. During the commissioning of ERIS, the mass resolution and overall efficiency, including ionization, extraction, and transmission, were found to be 1660 and 21%, respectively, using natural xenon gas. The preparation of uranium carbide (UC2) RI production targets is described from which a 132Sn beam was successfully separated in our first attempt at RI production.

Studies of uranium carbide targets of a high density

NASA Astrophysics Data System (ADS)

Panteleev, V. N.; Alyakrinskiy, O.; Barbui, M.; Barzakh, A. E.; Dubois, M.; Eleon, C.; Essabaa, S.; Fedorov, D. V.; Gaubert, G.; Ionan, A. M.; Ivanov, V. S.; Jardin, P.; Lau, C.; Leroy, R.; Lhersonneau, G.; Mezilev, K. A.; Mhamed, C.; Molkanov, P. L.; Moroz, F. V.; Orlov, S. Yu.; Saint Laurent, M. G.; Stroe, L.; Tecchio, L. B.; Tonezzer, M.; Villari, A. C. C.; Volkov, Yu. M.

2008-10-01

Production of Cs and Fr isotopes from uranium carbide targets of a high density has been investigated at IRIS (Investigation Radioactive Isotopes at Synchrocyclotron), Gatchina. The UC target material with a density of 12 g/cm3 was prepared in a form of pellets. Two targets were tested on-line under the same temperature conditions: (a) a reference small target with a thickness of 4.5 g/cm2; (b) a heavier (so called intermediate) target with a thickness of 91 g/cm2. Yields and release efficiencies of nuclides with half-lives from some minutes to some milliseconds produced by 1 GeV protons in these targets are presented. It is remarkable that yields, even those of very short-lived isotopes such as 214Fr (T1/2 = 5 ms) and 219Fr (T1/2 = 20 ms), increase proportionally to the target thickness. A one month off-line heating test of the 91 g/cm2 target at a temperature of 2000 °C has been carried out successfully. The yields and release efficiencies of Cs and Fr measured on-line before and after the heating test coincided within the limits of measurement errors, thereby demonstrating the conservation of the target unit parameters. Based on these very promising results, a heavier target with a mass about 0.7 kg is prepared presently at IRIS.

Methods for making a porous nuclear fuel element

DOEpatents

Youchison, Dennis L; Williams, Brian E; Benander, Robert E

2014-12-30

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

2012-01-01

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

Processing of solid solution, mixed uranium/refractory metal carbides for advanced space nuclear power and propulsion systems

NASA Astrophysics Data System (ADS)

Knight, Travis Warren

Nuclear thermal propulsion (NTP) and space nuclear power are two enabling technologies for the manned exploration of space and the development of research outposts in space and on other planets such as Mars. Advanced carbide nuclear fuels have been proposed for application in space nuclear power and propulsion systems. This study examined the processing technologies and optimal parameters necessary to fabricate samples of single phase, solid solution, mixed uranium/refractory metal carbides. In particular, the pseudo-ternary carbide, UC-ZrC-NbC, system was examined with uranium metal mole fractions of 5% and 10% and corresponding uranium densities of 0.8 to 1.8 gU/cc. Efforts were directed to those methods that could produce simple geometry fuel elements or wafers such as those used to fabricate a Square Lattice Honeycomb (SLHC) fuel element and reactor core. Methods of cold uniaxial pressing, sintering by induction heating, and hot pressing by self-resistance heating were investigated. Solid solution, high density (low porosity) samples greater than 95% TD were processed by cold pressing at 150 MPa and sintering above 2600 K for times longer than 90 min. Some impurity oxide phases were noted in some samples attributed to residual gases in the furnace during processing. Also, some samples noted secondary phases of carbon and UC2 due to some hyperstoichiometric powder mixtures having carbon-to-metal ratios greater than one. In all, 33 mixed carbide samples were processed and analyzed with half bearing uranium as ternary carbides of UC-ZrC-NbC. Scanning electron microscopy, x-ray diffraction, and density measurements were used to characterize samples. Samples were processed from powders of the refractory mono-carbides and UC/UC 2 or from powders of uranium hydride (UH3), graphite, and refractory metal carbides to produce hypostoichiometric mixed carbides. Samples processed from the constituent carbide powders and sintered at temperatures above the melting point of UC showed signs of liquid phase sintering and were shown to be largely solid solutions. Pre-compaction of mixed carbide powders prior to sintering was shown to be necessary to achieve high densities. Hypostoichiometric, samples processed at 2500 K exhibited only the initial stage of sintering and solid solution formation. Based on these findings, a suggested processing methodology is proposed for producing high density, solid solution, mixed carbide fuels. Pseudo-binary, refractory carbide samples hot pressed at 3100 K and 6 MPa showed comparable densities (approximately 85% of the theoretical value) to samples processed by cold pressing and sintering at temperatures of 2800 K.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

2010-02-23

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Composition and method for brazing graphite to graphite

DOEpatents

Taylor, Albert J.; Dykes, Norman L.

1984-01-01

The present invention is directed to a brazing material for joining graphite structures that can be used at temperatures up to about 2800.degree. C. The brazing material formed of a paste-like composition of hafnium carbide and uranium oxide with a thermosetting resin. The uranium oxide is converted to uranium dicarbide during the brazing operation and then the hafnium carbide and uranium dicarbide form a liquid phase at a temperature about 2600.degree. C. with the uranium diffusing and vaporizing from the joint area as the temperature is increased to about 2800.degree. C. so as to provide a brazed joint consisting essentially of hafnium carbide. This brazing temperature for hafnium carbide is considerably less than the eutectic temperature of hafnium carbide of about 3150.degree. C. The brazing composition also incorporates the thermosetting resin so that during the brazing operation the graphite structures may be temporarily bonded together by thermosetting the resin so that machining of the structures to final dimensions may be completed prior to the completion of the brazing operation. The resulting brazed joint is chemically and thermally compatible with the graphite structures joined thereby and also provides a joint of sufficient integrity so as to at least correspond with the strength and other properties of the graphite.

High temperature fuel/emitter system for advanced thermionic fuel elements

NASA Astrophysics Data System (ADS)

Moeller, Helen H.; Bremser, Albert H.; Gontar, Alexander; Fiviesky, Evgeny

1997-01-01

Specialists in space applications are currently focusing on bimodal power systems designed to provide both electric power and thermal propulsion (Kennedy, 1994 and Houts, 1995). Our work showed that thermionics is a viable technology for nuclear bimodal power systems. We demonstrated that materials for a thermionic fuel-emitter combination capable of performing at operating temperatures of 2473 K are not only possible but available. The objective of this work, funded by the US Department of Energy, Office of Space and Defense Power Systems, was to evaluate the compatibility of fuel material consisting of an uranium carbide/tantalum carbide solid solution with an emitter material consisting of a monocrystalline tungsten-niobium alloy. The uranium loading of the fuel material was 70 mole% uranium carbide. The program was successfully accomplished by a B&W/SIA LUTCH team. Its workscope was integrated with tasks being performed at both Babcock & Wilcox, Lynchburg Research Center, Lynchburg, Virginia, and SIA LUTCH, Podolsk, Russia. Samples were fabricated by LUTCH and seven thermal tests were performed in a hydrogen atmosphere. The first preliminary test was performed at 2273 K by LUTCH, and the remaining six tests were performed At B&W. Three tests were performed at 2273 K, two at 2373 K, and the final test at 2473 K. The results showed that the fuel and emitter materials were compatible in the presence of hydrogen. No evidence of liquid formation, dissolution of the uranium carbide from the uranium carbide/tantalum carbide solid solution, or diffusion of the uranium into the monocrystalline tungsten alloy was observed. Among the highlights of the program was the successful export of the fuel samples from Russia and their import into the US by commercial transport. This paper will discuss the technical aspects of this work.

Some observations on uranium carbide alloy/tungsten compatibility

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Chemical compatibility between both pure and thoriated tungsten and uranium carbide alloys was studied at 1800 C for up to 3300 hours. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, dependent upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. The presence of a thermal gradient had no effect on the reactions observed nor did the presence of thoria in the tungsten clad.

Some observations on uranium carbide alloy/tungsten compatibility.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Results of chemical compatibility tests between both pure tungsten and thoriated tungsten run at 1800 C for up to 3300 hours with uranium carbide alloys. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, depending upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. Neither the presence of a thermal gradient nor the presence of thoria in the tungsten clad affect the reactions observed.

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

NASA Astrophysics Data System (ADS)

Knight, Travis W.; Anghaie, Samim

2002-11-01

Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels - namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.

Irradiation test of tungsten clad uranium carbide-zirconium carbide ((U,Zr)C) specimens for thermionic reactor application at conditions conductive to long-term performance

NASA Technical Reports Server (NTRS)

Creagh, J. W. R.; Smith, J. R.

1973-01-01

Uranium carbide fueled, thermionic emitter configurations were encapsulated and irradiated. One capsule contained a specimen clad with fluoride derived chemically vapor deposited (CVD) tungsten. The other capsule used a duplex clad specimen consisting of chloride derived on floride derived CVD tungsten. Both fuel pins were 16 millimeters in diameter and contained a 45.7-millimeter length of fuel.

A two-dimensional, finite-difference model of the oxidation of a uranium carbide fuel pellet

NASA Astrophysics Data System (ADS)

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.; Heggs, Peter J.

2015-12-01

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

The oxidative corrosion of carbide inclusions at the surface of uranium metal during exposure to water vapour.

PubMed

Scott, T B; Petherbridge, J R; Harker, N J; Ball, R J; Heard, P J; Glascott, J; Allen, G C

2011-11-15

The reaction between uranium and water vapour has been well investigated, however discrepancies exist between the described kinetic laws, pressure dependence of the reaction rate constant and activation energies. Here this problem is looked at by examining the influence of impurities in the form of carbide inclusions on the reaction. Samples of uranium containing 600 ppm carbon were analysed during and after exposure to water vapour at 19 mbar pressure, in an environmental scanning electron microscope (ESEM) system. After water exposure, samples were analysed using secondary ion mass spectrometry (SIMS), focused ion beam (FIB) imaging and sectioning and transmission electron microscopy (TEM) with X-ray diffraction (micro-XRD). The results of the current study indicate that carbide particles on the surface of uranium readily react with water vapour to form voluminous UO(3) · xH(2)O growths at rates significantly faster than that of the metal. The observation may also have implications for previous experimental studies of uranium-water interactions, where the presence of differing levels of undetected carbide may partly account for the discrepancies observed between datasets. Crown Copyright © 2011. Published by Elsevier B.V. All rights reserved.

Creep behavior of uranium carbide-based alloys

NASA Technical Reports Server (NTRS)

Seltzer, M. S.; Wright, T. R.; Moak, D. P.

1975-01-01

The present work gives the results of experiments on the influence of zirconium carbide and tungsten on the creep properties of uranium carbide. The creep behavior of high-density UC samples follows the classical time-dependence pattern of (1) an instantaneous deformation, (2) a primary creep region, and (3) a period of steady-state creep. Creep rates for unalloyed UC-1.01 and UC-1.05 are several orders of magnitude greater than those measured for carbide alloys containing a Zr-C and/or W dispersoid. The difference in creep strength between alloyed and unalloyed materials varies with temperature and applied stress.

Influence of target thickness on the release of radioactive atoms

NASA Astrophysics Data System (ADS)

Guillot, Julien; Roussière, Brigitte; Tusseau-Nenez, Sandrine; Barré-Boscher, Nicole; Borg, Elie; Martin, Julien

2017-03-01

Nowadays, intense exotic beams are needed in order to study nuclei with very short half-life. To increase the release efficiency of the fission products, all the target characteristics involved must be improved (e.g. chemical composition, dimensions, physicochemical properties such as grain size, porosity, density…). In this article, we study the impact of the target thickness. Released fractions measured from graphite and uranium carbide pellets are presented as well as Monte-Carlo simulations of the Brownian motion.

Developing uranium dicarbide-graphite porous materials for the SPES project

NASA Astrophysics Data System (ADS)

Biasetto, L.; Zanonato, P.; Carturan, S.; Di Bernardo, P.; Colombo, P.; Andrighetto, A.; Prete, G.

2010-09-01

Uranium carbide dispersed in graphite was produced under vacuum by means of carbothermic reduction of different uranium oxides (UO 2, U 3O 8 and UO 3), using graphite as the source of carbon. The thermal process was monitored by mass spectrometry and the gas evolution confirmed the reduction of the U 3O 8 and UO 3 oxides to UO 2 before the carbothermic reaction, that started to occur at T > 1000 °C. XRD analysis confirmed the formation of α-UC 2 and of a minor amount of UC. The morphology of the produced uranium carbide was not affected by the oxides employed as the source of uranium.

Studies for aluminum photoionization in hot cavity for the selective production of exotic species projecta)

NASA Astrophysics Data System (ADS)

Scarpa, D.; Vasquez, J.; Tomaselli, A.; Grassi, D.; Biasetto, L.; Cavazza, A.; Corradetti, S.; Manzolaro, M.; Montano, J.; Andrighetto, A.; Prete, G.

2012-02-01

Selective production of exotic species (SPES) is an ISOL-based accelerator facility that will be built in the Legnaro INFN Laboratory (Italy), intended to provide an intense neutron-rich radioactive ion beams obtained by proton induced fission of an uranium carbide target. Beside this main target, a silicon carbide (SiC) target will the first to be used to deliver some p-rich beams. This target will validate also the functionality of the SPES facility with aluminum beam as result of hitting SiC target with protons. In the past off-line studies on laser photoionization of aluminum have performed in Pavia Spectroscopy Laboratory and in Laboratori Nazionali di Legnaro where, recently, a XeCl excimer laser was installed in order to test the laser ionization in the SPES hot cavity. Results are promising to justify further studies with this technique, aiming a better characterization of the SPES ion extraction capability under laser photoionization.

Fabrication of thorium bearing carbide fuels

DOEpatents

Gutierrez, Rueben L.; Herbst, Richard J.; Johnson, Karl W. R.

1981-01-01

Thorium-uranium carbide and thorium-plutonium carbide fuel pellets have been fabricated by the carbothermic reduction process. Temperatures of 1750.degree. C. and 2000.degree. C. were used during the reduction cycle. Sintering temperatures of 1800.degree. C. and 2000.degree. C. were used to prepare fuel pellet densities of 87% and >94% of theoretical, respectively. The process allows the fabrication of kilogram quantities of fuel with good reproducibility of chemicals and phase composition. Methods employing liquid techniques that form carbide microspheres or alloying-techniques which form alloys of thorium-uranium or thorium-plutonium suffer from limitation on the quantities processed of because of criticality concerns and lack of precise control of process conditions, respectively.

THERMAL FISSION REACTOR COMPOSITIONS AND METHOD OF FABRICATING SAME

DOEpatents

Blainey, A.

1959-10-01

A body is presented for use in a thermal fission reactor comprising a sintered compressed mass of a substance of the group consisting of uranium, thorium, and oxides and carbides of uranium and thorium, enclosed in an envelope of a sintered, compacted, heat-conductive material of the group consisting of beryllium, zirconium, and oxides and carbides of beryllium and zirconium.

Processing of uranium oxide and silicon carbide based fuel using polymer infiltration and pyrolysis

NASA Astrophysics Data System (ADS)

Singh, Abhishek K.; Zunjarrao, Suraj C.; Singh, Raman P.

2008-09-01

Ceramic composite pellets consisting of uranium oxide, UO 2, contained within a silicon carbide matrix, were fabricated using a novel processing technique based on polymer infiltration and pyrolysis (PIP). In this process, particles of depleted uranium oxide, in the form of U 3O 8, were dispersed in liquid allylhydridopolycarbosilane (AHPCS), and subjected to pyrolysis up to 900 °C under a continuous flow of ultra high purity argon. The pyrolysis of AHPCS, at these temperatures, produced near-stoichiometric amorphous silicon carbide ( a-SiC). Multiple polymer infiltration and pyrolysis (PIP) cycles were performed to minimize open porosity and densify the silicon carbide matrix. Analytical characterization was conducted to investigate chemical interaction between U 3O 8 and SiC. It was observed that U 3O 8 reacted with AHPCS during the very first pyrolysis cycle, and was converted to UO 2. As a result, final composition of the material consisted of UO 2 particles contained in an a-SiC matrix. The physical and mechanical properties were also quantified. It is shown that this processing scheme promotes uniform distribution of uranium fuel source along with a high ceramic yield of the parent matrix.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Metz, Lori A.; Friese, Judah I.; Finn, Erin C.

Critical assemblies provide one method of achieving a fast neutron spectrum that is close to a 235U fission-energy neutron spectrum for nuclear data measurements. Previous work has demonstrated the use of a natural boron carbide capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method for performing fission-energy neutron experiments. Previous fission products measurements showed that the neutron spectrum achievable with natural boron carbide was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a boron carbide capsule 96% enriched in 10B formore » irradiations resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. The current work describes an experiment involving a highly-enriched uranium target irradiated under the new 10B4C capsule. Fission product yields were measured following radiochemical separations and are presented here. Reactor dosimetry measurements for characterizing neutron spectra and fluence for the enriched boron carbide capsule and critical assemblies are also discussed.« less

Radiation shielding materials and containers incorporating same

DOEpatents

Mirsky, Steven M.; Krill, Stephen J.; Murray, Alexander P.

2005-11-01

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound ("PYRUC") shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Radiation Shielding Materials and Containers Incorporating Same

DOEpatents

Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

2005-11-01

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Photo-ionization of aluminum in a hot cavity for the selective production of exotic species project

NASA Astrophysics Data System (ADS)

Scarpa, D.; Makhathini, L.; Tomaselli, A.; Grassi, D.; Corradetti, S.; Manzolaro, M.; Vasquez, J.; Calderolla, M.; Rossignoli, M.; Monetti, A.; Andrighetto, A.; Prete, G.

2014-02-01

SPES (Selective Production of Exotic Species) is an Isotope Separation On-Line (ISOL) based accelerator facility that will be built in the Legnaro-Istituto Nazionale di Fisica Nucleare (INFN) Laboratory (Italy), intended to provide intense neutron-rich radioactive ion beams obtained by proton-induced fission of a uranium carbide (UCx) target. Besides this main target material, silicon carbide (SiC) will be the first to be used to deliver p-rich beams. This target will also validate the functionality of the SPES facility with aluminum beam as result of impinging SiC target with proton beam. In the past, off line studies on laser photoionization of aluminum have been performed in Pavia Spectroscopy Laboratory and in Laboratori Nazionali di Legnaro; a XeCl excimer laser was installed in order to test the laser ionization in the SPES hot cavity. With the new Wien filter installed a better characterization of the ionization process in terms of efficiency was performed and results are discussed.

PROCESS OF PREPARING URANIUM CARBIDE

DOEpatents

Miller, W.E.; Stethers, H.L.; Johnson, T.R.

1964-03-24

A process of preparing uranium monocarbide is de scribed. Uranium metal is dissolved in cadmium, zinc, cadmium-- zinc, or magnesium-- zinc alloy and a small quantity of alkali metal is added. Addition of stoichiometric amounts of carbon at 500 to 820 deg C then precipitates uranium monocarbide. (AEC)

Composite uranium carbide targets at TRIUMF: Development and characterization with SEM, XRD, XRF and L-edge densitometry

NASA Astrophysics Data System (ADS)

Kunz, Peter; Bricault, Pierre; Dombsky, Marik; Erdmann, Nicole; Hanemaayer, Vicky; Wong, John; Lützenkirchen, Klaus

2013-09-01

The production of radioactive ion beams (RIB) from spallation targets by irradiation with a continuous 500 MeV proton beam, has been routine at TRIUMF for several years. Based on the experience with composite refractory carbide targets a procedure for the fabrication of UC2/C targets was developed. It includes the preparation of UC2 by carbothermal reduction of UO2, the slip-casting of fine-grained UC2/C slurry on graphite foil under inert gas atmosphere and the cutting of composite target discs which are stacked up to a lamellar structure. The thermal properties of such an arrangement are adequate to withstand the high power deposition of an intense, continuous proton beam and also beneficial for the fast release of short-lived radioactive isotopes. Molecular structure, particle size and the impact of sintering of the target discs were investigated via XRD and SEM. Thickness and mass distribution were measured with position-sensitive LIII-edge densitometry. The results confirm that the properties of the UC2/C target material are well suited for RIB production at TRIUMF while there is still room for improvement with regard to uniformity of mass distribution in target disc thickness.

XPS and SIMS study of the surface and interface of aged C + implanted uranium

DOE PAGES

Donald, Scott B.; Siekhaus, Wigbert J.; Nelson, Art J.

2016-09-08

X-ray photoelectron spectroscopy in combination with secondary ion mass spectrometry depth profiling were used to investigate the surface and interfacial chemistry of C + ion implanted polycrystalline uranium subsequently oxidized in air for over 10 years at ambient temperature. The original implantation of 33 keV C + ions into U 238 with a dose of 4.3 × 10 17 cm –3 produced a physically and chemically modified surface layer that was characterized and shown to initially prevent air oxidation and corrosion of the uranium after 1 year in air at ambient temperature. The aging of the surface and interfacial layersmore » were examined by using the chemical shift of the U 4f, C 1s, and O 1s photoelectron lines. In addition, valence band spectra were used to explore the electronic structure of the aged carbide surface and interface layer. Moreover, the time-of-flight secondary ion mass spectrometry depth profiling results for the aged sample confirmed an oxidized uranium carbide layer over the carbide layer/U metal interface.« less

Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4

NASA Astrophysics Data System (ADS)

Hallman, Luther, Jr.

Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism. One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are an increased corrosion resistance, high mechanical strength, and irradiation stability. However, with increased temperatures, SiC exhibits an extremely brittle nature. The brittle behavior of SiC is not fully understood and thus it is unknown how SiC would respond to the added stress of a swelling UC fuel. To better understand the interaction between these advanced materials, each has been implemented into FRAPCON, the preferred fuel performance code of the Nuclear Regulatory Commission (NRC); additionally, the material properties for a helium coolant have been incorporated. The implementation of UC within FRAPCON required the development of material models that described not only the thermophysical properties of UC, such as thermal conductivity and thermal expansion, but also models for the swelling, densification, and fission gas release associated with the fuel's irradiation behavior. This research is intended to supplement ongoing analysis of the performance and behavior of uranium carbide and silicon carbide in a helium-cooled reactor.

A wet chemical method for the estimation of carbon in uranium carbides.

PubMed

Chandramouli, V; Yadav, R B; Rao, P R

1987-09-01

A wet chemical method for the estimation of carbon in uranium carbides has been developed, based on oxidation with a saturated solution of sodium dichromate in 9M sulphuric acid, absorption of the evolved carbon dioxide in a known excess of barium hydroxide solution, and titration of the excess of barium hydroxide with standard potassium hydrogen phthalate solution. The carbon content obtained is in good agreement with that obtained by combustion and titration.

Experimental investigation of the ionization mechanisms of uranium in thermal ionization mass spectrometry in the presence of carbon

NASA Astrophysics Data System (ADS)

Kraiem, M.; Mayer, K.; Gouder, T.; Seibert, A.; Wiss, T.; Thiele, H.; Hiernaut, J.-P.

2010-01-01

Thermal ionization mass spectrometry (TIMS) is a well established instrumental technique for providing accurate and precise isotope ratio measurements of elements with reasonably low first ionization potential. In nuclear safeguards and in environmental research, it is often required to measure the isotope ratios in small samples of uranium. Empirical studies had shown that the ionization yield of uranium and plutonium in a TIMS ion source can be significantly increased in the presence of a carbon source. But, even though carbon appeared crucial in providing high ionization yields, processes taking place on the ionization surface were still not well understood. This paper describes the experimental results obtained from an extended study on the evaporation and ionization mechanisms of uranium occurring on a rhenium mass spectrometry filament in the presence of carbon. Solid state reactions were investigated using X-ray photoelectron spectroscopy (XPS) and scanning electron microscopy (SEM). Additionally, vaporization measurements were performed with a modified-Knudsen cell mass spectrometer for providing information on the neutral uranium species in the vapor phase. Upon heating, under vacuum, the uranyl nitrate sample was found to turn into a uranium carbide compound, independent of the type of carbon used as ionization enhancer. With further heating, uranium carbide leads to formation of single charged uranium metal ions and a small amount of uranium carbide ions. The results are relevant for a thorough understanding of the ion source chemistry of a uranyl nitrate sample under reducing conditions. The significant increase in ionization yield described by many authors on the basis of empirical results can be now fully explained and understood.

Preparation of uranium fuel kernels with silicon carbide nanoparticles using the internal gelation process

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Silva, G. W. C. M.; Lindemer, T. B.; Anderson, K. K.; Collins, J. L.

2012-08-01

The US Department of Energy continues to use the internal gelation process in its preparation of tristructural isotropic coated fuel particles. The focus of this work is to develop uranium fuel kernels with adequately dispersed silicon carbide (SiC) nanoparticles, high crush strengths, uniform particle diameter, and good sphericity. During irradiation to high burnup, the SiC in the uranium kernels will serve as getters for excess oxygen and help control the oxygen potential in order to minimize the potential for kernel migration. The hardness of SiC required modifications to the gelation system that was used to make uranium kernels. Suitable processing conditions and potential equipment changes were identified so that the SiC could be homogeneously dispersed in gel spheres. Finally, dilute hydrogen rather than argon should be used to sinter the uranium kernels with SiC.

Cobalt-Free Permanent Magnet Alloys.

DTIC Science & Technology

1984-10-01

carbide co- UC CbC lumbium carbide M003 Uranium carbide - tho- UC 2 25ThC rium carbide ZrO2 MgO WOs Use of this Process for MnAlC As indicated in the...cobalt. Free World Cobal Consumption Estimated Breakdown by End Uses Magnetic alloys 20% Cemented carbides - 5% 30 SuPerolloy _ 15% Other steels and...would normally result in the formation of binary alloy of TbFe 2 and preventing the formation of amorphous alloy (Fe-B) contain- ing Tb. The

Beam heating of thick targets for on-line mass separators

NASA Astrophysics Data System (ADS)

Eaton, T. W.; Ravn, H. L.; Isolde Collaboration

1987-05-01

Energy deposition computations have been made on a variety of target materials utilized for the production of radioisotopes by means of 600-MeV protons. Results have shown that, when a proton current of 100 μA is assumed, dispersed target materials, such as uranium carbide powder and magnesium oxide, are best able to withstand the energy absorption and consequent beam heating without the need of additional cooling. Modified foil targets of titanium, zirconium and tantalum also appear capable of withstanding a full beam current, whilst liquid metal targets in their present form appear to have limitations in terms of the maximum allowable beam current. A redesign of the target container is proposed which allows higher proton currents to be used with these targets also.

Kinetic study of the carbothermic synthesis of uranium monocarbide microspheres

NASA Astrophysics Data System (ADS)

Mukerjee, S. K.; Dehadraya, J. V.; Vaidya, V. N.; Sood, D. D.

1990-06-01

Uranium monocarbide microspheres were synthesized by carbothermic reduction of porous uranium oxide microspheres with uniformly dispersed carbon black. Kinetics of the reduction was studied under vacuum and flowing inert gas from 1250 to 1550° C. The carbon monoxide gas concentration in the effluent stream during reduction was used to determine the rate of carbide formation. Under vacuum, reduction was found to be controlled by reaction at the reactant-product interface whereas under flowing gas conditions, the diffusion of carbon monoxide gas through the carbide layer was the rate controlling process. The activation energy was 335.1 ± 8.6 and 363.7 ± 7.6 kJ/mol for reduction under vacuum and flowing gas, respectively.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donald, Scott B.; Siekhaus, Wigbert J.; Nelson, Art J.

X-ray photoelectron spectroscopy in combination with secondary ion mass spectrometry depth profiling were used to investigate the surface and interfacial chemistry of C + ion implanted polycrystalline uranium subsequently oxidized in air for over 10 years at ambient temperature. The original implantation of 33 keV C + ions into U 238 with a dose of 4.3 × 10 17 cm –3 produced a physically and chemically modified surface layer that was characterized and shown to initially prevent air oxidation and corrosion of the uranium after 1 year in air at ambient temperature. The aging of the surface and interfacial layersmore » were examined by using the chemical shift of the U 4f, C 1s, and O 1s photoelectron lines. In addition, valence band spectra were used to explore the electronic structure of the aged carbide surface and interface layer. Moreover, the time-of-flight secondary ion mass spectrometry depth profiling results for the aged sample confirmed an oxidized uranium carbide layer over the carbide layer/U metal interface.« less

Characterization of uranium carbide target materials to produce neutron-rich radioactive beams

NASA Astrophysics Data System (ADS)

Tusseau-Nenez, Sandrine; Roussière, Brigitte; Barré-Boscher, Nicole; Gottberg, Alexander; Corradetti, Stefano; Andrighetto, Alberto; Cheikh Mhamed, Maher; Essabaa, Saïd; Franberg-Delahaye, Hanna; Grinyer, Joanna; Joanny, Loïc; Lau, Christophe; Le Lannic, Joseph; Raynaud, Marc; Saïd, Abdelhakim; Stora, Thierry; Tougait, Olivier

2016-03-01

In the framework of a R&D program aiming to develop uranium carbide (UCx) targets for radioactive nuclear beams, the Institut de Physique Nucléaire d'Orsay (IPNO) has developed an experimental setup to characterize the release of various fission fragments from UCx samples at high temperature. The results obtained in a previous study have demonstrated the feasibility of the method and started to correlate the structural properties of the samples and their behavior in terms of nuclear reaction product release. In the present study, seven UCx samples have been systematically characterized in order to better understand the correlation between their physicochemical characteristics and release properties. Two very different samples, the first one composed of dense UC and the second one of highly porous UCx made of multi-wall carbon nanotubes, were provided by the ActILab (ENSAR) collaboration. The others were synthesized at IPNO. The systems for irradiation and heating necessary for the release studies have been improved with respect to those used in previous studies. The results show that the open porosity is hardly the limiting factor for the fission product release. The homogeneity of the microstructure and the pore size distribution contributes significantly to the increase of the release. The use of carbon nanotubes in place of traditional micrometric graphite particles appears to be promising, even if the homogeneity of the microstructure can still be enhanced.

Infrared spectra and quantum chemical calculations of the uranium-carbon molecules UC, CUC, UCH, and U(CC)2

NASA Astrophysics Data System (ADS)

Wang, Xuefeng; Andrews, Lester; Ma, Dongxia; Gagliardi, Laura; Gonçalves, António P.; Pereira, Cláudia C. L.; Marçalo, Joaquim; Godart, Claude; Villeroy, Benjamin

2011-06-01

Laser evaporation of carbon rich uranium/carbon alloy targets into condensing argon or neon matrix samples gives weak infrared absorptions that increase on annealing, which can be assigned to new uranium carbon bearing species. New bands at 827.6 cm-1 in solid argon or 871.7 cm-1 in neon become doublets with mixed carbon 12 and 13 isotopes and exhibit the 1.0381 carbon isotopic frequency ratio for the UC diatomic molecule. Another new band at 891.4 cm-1 in argon gives a three-band mixed isotopic spectrum with the 1.0366 carbon isotopic frequency ratio, which is characteristic of the anti-symmetric stretching vibration of a linear CUC molecule. No evidence was found for the lower energy cyclic U(CC) isomer. Other bands at 798.6 and 544.0 cm-1 are identified as UCH, which has a uranium-carbon triple bond similar to that in UC. Evidence is found for bicyclic U(CC)2 and tricyclic U(CC)3. This work shows that U and C atoms react spontaneously to form the uranium carbide U≡C and C≡U≡C molecules with uranium-carbon triple bonds.

Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

NASA Astrophysics Data System (ADS)

Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

2001-02-01

Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented. .

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

Carbothermic synthesis of 820 μm uranium nitride kernels: Literature review, thermodynamics, analysis, and related experiments

NASA Astrophysics Data System (ADS)

Lindemer, T. B.; Voit, S. L.; Silva, C. M.; Besmann, T. M.; Hunt, R. D.

2014-05-01

The US Department of Energy is developing a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with uranium nitride (UN) kernels with diameters near 825 μm. This effort explores factors involved in the conversion of uranium oxide-carbon microspheres into UN kernels. An analysis of previous studies with sufficient experimental details is provided. Thermodynamic calculations were made to predict pressures of carbon monoxide and other relevant gases for several reactions that can be involved in the conversion of uranium oxides and carbides into UN. Uranium oxide-carbon microspheres were heated in a microbalance with an attached mass spectrometer to determine details of calcining and carbothermic conversion in argon, nitrogen, and vacuum. A model was derived from experiments on the vacuum conversion to uranium oxide-carbide kernels. UN-containing kernels were fabricated using this vacuum conversion as part of the overall process. Carbonitride kernels of ∼89% of theoretical density were produced along with several observations concerning the different stages of the process.

Development of Spacecraft Materials and Structures Fundamentals.

DTIC Science & Technology

1985-08-01

900. This is comparable to the dihedral angle observed in uranium dioxide’ ° and silicon carbide ,’ 2 which...necesjary and identify by bigich numberp FIELD GROUP I suB. GR. Boron carbide , sintering, grain growth, microstructure, microcracking, mechanical...Compacts of boron carbide powders with specific surface area >, 8 m2 / were sintered in argon at temperatures near 2200*C. Several of these powders were

Carbothermic Synthesis of 820 m UN Kernels: Literature Review, Thermodynamics, Analysis, and Related Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindemer, Terrence; Voit, Stewart L; Silva, Chinthaka M

2014-01-01

The U.S. Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with large, dense uranium nitride (UN) kernels. This effort explores many factors involved in using gel-derived uranium oxide-carbon microspheres to make large UN kernels. Analysis of recent studies with sufficient experimental details is provided. Extensive thermodynamic calculations are used to predict carbon monoxide and other pressures for several different reactions that may be involved in conversion of uranium oxides and carbides to UN. Experimentally, the method for making themore » gel-derived microspheres is described. These were used in a microbalance with an attached mass spectrometer to determine details of carbothermic conversion in argon, nitrogen, or vacuum. A quantitative model is derived from experiments for vacuum conversion to an uranium oxide-carbide kernel.« less

Neutron production by stopping 55 MeV deuterons in carbon and heavy water

NASA Astrophysics Data System (ADS)

Lhersonneau, G.; Malkiewicz, T.; Jones, P.; Ketelhut, S.; Trzaska, W. H.

2012-09-01

Neutron production by stopping 55 MeV deuterons in thick carbon and heavy-water targets has been measured by the activation method. The geometry was close to the one defined for the SPIRAL2 uranium-carbide target in the initial phase. A comparative method for obtaining the neutron flux has been used and is presented in detail. The neutron flux generated by 55 MeV deuterons on carbon is 2.3 times the flux at the deuteron energy of 40 MeV. The flux further increases by a factor 1.4 when using a heavy-water target. These results are discussed in the context of an energy upgrade of the SPIRAL2 driver accelerator.

A physical model for evaluating uranium nitride specific heat

NASA Astrophysics Data System (ADS)

Baranov, V. G.; Devyatko, Yu. N.; Tenishev, A. V.; Khlunov, A. V.; Khomyakov, O. V.

2013-03-01

Nitride fuel is one of perspective materials for the nuclear industry. But unlike the oxide and carbide uranium and mixed uranium-plutonium fuel, the nitride fuel is less studied. The present article is devoted to the development of a model for calculating UN specific heat on the basis of phonon spectrum data within the solid state theory.

METHOD OF OPERATING A CALUTRON

DOEpatents

Davidson, P.H.

1960-01-12

A method of operating an electromagnetic isotope separator of the calutron class is reported whereby uranium tetrachloride is produced at a controlled rate within the source rather than betng introduced therein as was formerly practiced. This is accomplished by placing a uranium-bearing material, such as uranium metal, uranium trichloride, or uranium carbide in the charge receptacle of the calutron, heating this material to about to produce uranium tetrachloride vapor at a rate controlled by the chlorine gas flow into the source. The vapor is subsequently ionized by an electric arc and mass separated by conventional calutron methods.

Navy Aegis Ballistic Missile Defense (BMD) Program: Background and Issues for Congress

DTIC Science & Technology

2010-09-28

co-development with Japan, for chemical vapor composite silicon carbide and silicon carbide corrugated mirror processes for the SM–3 Block IIA...all the more urgent in light of Iran’s continued uranium enrichment program. Iran continues to defy international obligations, and there continues to

Fabrication of thorium bearing carbide fuels

DOEpatents

Gutierrez, R.L.; Herbst, R.J.; Johnson, K.W.R.

Thorium-uranium carbide and thorium-plutonium carbide fuel pellets have been fabricated by the carbothermic reduction process. Temperatures of 1750/sup 0/C and 2000/sup 0/C were used during the reduction cycle. Sintering temperatures of 1800/sup 0/C and 2000/sup 0/C were used to prepare fuel pellet densities of 87% and > 94% of theoretical, respectively. The process allows the fabrication of kilogram quantities of fuel with good reproductibility of chemical and phase composition.

Navy Aegis Ballistic Missile Defense (BMD) Program: Background and Issues for Congress

DTIC Science & Technology

2011-04-19

for SM-3 Block IIA co-development with Japan, for chemical vapor composite silicon carbide and silicon carbide corrugated mirror processes for the SM...to say, this concern is all the more urgent in light of Iran’s continued uranium enrichment program. Iran continues to defy international obligations

Development of target ion source systems for radioactive beams at GANIL

NASA Astrophysics Data System (ADS)

Bajeat, O.; Delahaye, P.; Couratin, C.; Dubois, M.; Franberg-Delahaye, H.; Henares, J. L.; Huguet, Y.; Jardin, P.; Lecesne, N.; Lecomte, P.; Leroy, R.; Maunoury, L.; Osmond, B.; Sjodin, M.

2013-12-01

The GANIL facility (Caen, France) is dedicated to the acceleration of heavy ion beams including radioactive beams produced by the Isotope Separation On-Line (ISOL) method at the SPIRAL1 facility. To extend the range of radioactive ion beams available at GANIL, using the ISOL method two projects are underway: SPIRAL1 upgrade and the construction of SPIRAL2. For SPIRAL1, a new target ion source system (TISS) using the VADIS FEBIAD ion source coupled to the SPIRAL1 carbon target will be tested on-line by the end of 2013 and installed in the cave of SPIRAL1 for operation in 2015. The SPIRAL2 project is under construction and is being design for using different production methods as fission, fusion or spallation reactions to cover a large area of the chart of nuclei. It will produce among others neutron rich beams obtained by the fission of uranium induced by fast neutrons. The production target made from uranium carbide and heated at 2000 °C will be associated with several types of ion sources. Developments currently in progress at GANIL for each of these projects are presented.

First-principles study of uranium carbide: Accommodation of point defects and of helium, xenon, and oxygen impurities

NASA Astrophysics Data System (ADS)

Freyss, Michel

2010-01-01

Point defects and volatile impurities (helium, xenon, oxygen) in uranium monocarbide UC are studied by first-principles calculations. Preliminarily, bulk properties of UC and of two other uranium carbide phases, UC2 and U2C3 , are calculated in order to compare them to experimental data and to get confidence in the use of the generalized gradient approximation for this class of compounds. The subsequent study of different types of point defects shows that the carbon sublattice best accommodates the defects. The perturbation of the crystal structure induced by the defects is weak and the interaction between defects is found short range. Interstitial carbon dumbbells possibly play an important role in the diffusion of carbon atoms. The most favorable location of diluted helium, xenon, and oxygen impurities in the UC crystal lattice is then determined. The rare-gas atoms occupy preferably a uranium substitution site or a uranium site in a U-C bivacancy. But their incorporation in UC is, however, not energetically favorable, especially for xenon, suggesting their propensity to diffuse in the material and/or form bubbles. On the other hand, oxygen atoms are very favorably incorporated as diluted atoms in the UC lattice, confirming the easy oxidation of UC. The oxygen atoms preferably occupy a carbon substitution site or the carbon site of a U-C bivacancy. Our results are compared to available experimental data on UC and to similar studies by first-principles calculations for other carbides and nitrides with the rock-salt structure.

New production systems at ISOLDE

NASA Astrophysics Data System (ADS)

Hagebø, E.; Hoff, P.; Jonsson, O. C.; Kugler, E.; Omtvedt, J. P.; Ravn, H. L.; Steffensen, K.

1992-08-01

New target systems for the ISOLDE on-line mass separator facility are presented. Targets of carbides, metal/graphite mixtures, foils of refractory metals, molten metals and oxides have been tested. Beams of high intensity of neutron-rich isotopes of a large number of elements are obtained from a uranium carbide target with a hot plasma-discharge ion source. A target of ZrO 2 has been shown to provide high intensity beams of neutron-deficient isotopes of Mn, Cu, Zn, Ga, Ge, As, Se, Br, Kr and Rb, while a SiC target with a hot plasma ion source gives intense beams of radioactive isotopes of a number of light elements. All these systems are rather chemically unselective. Chemically selective performance has been obtained for several systems, i.e.: the production of neutron-deficient Au from ( 3He, pχn) reactions on a Pt/graphite target with a hot plasma ion source; the production of neutron-deficient Lu and LuF + and Hf and HfF 3+ from a Ta-foil target with a hot plasma ion source under CF 4 addition; the production of neutron-deficient Sr as SrF + and Y as YF 2+ form a Nb-foil target with a W surface ionizer under CF 4 addition; the production of neutron-deficient Se as COSe + from a ZrO 2 target with a hot plasma ion source under O 2 addition; and the production of radioactive F from a SiC target with a hot plasma ion source operating in Al vapour.

Morphological and functional effects of graphene on the synthesis of uranium carbide for isotopes production targets.

PubMed

Biasetto, L; Corradetti, S; Carturan, S; Eloirdi, R; Amador-Celdran, P; Staicu, D; Blanco, O Dieste; Andrighetto, A

2018-05-29

The development of tailored targets for the production of radioactive isotopes represents an active field in nuclear research. Radioactive beams find applications in nuclear medicine, in astrophysics, matter physics and materials science. In this work, we study the use of graphene both as carbon source for UO 2 carbothermal reduction to produce UC x targets, and also as functional properties booster. At fixed composition, the UC x target grain size, porosity and thermal conductivity represent the three main points that affect the target production efficiency. UC x was synthesized using both graphite and graphene as the source of carbon and the target properties in terms of composition, grain size, porosity, thermal diffusivity and thermal conductivity were studied. The main output of this work is related to the remarkable enhancement achieved in thermal conductivity, which can profitably improve thermal dissipation during operational stages of UC x targets.

Evaluation of Non-Oxide Fuel for Fission-based Nuclear Reactors on Spacecraft

DTIC Science & Technology

smaller and potentially lighter core, whichis a significant advantage. The results of this study indicate that use of both UC and UN may result in significant weight savings due tohigher uranium loading density....The goal of this project was to study the performance of atypical uranium-based fuels in a nuclear reactor capable of producing 1 megawattof thermal...UN), or uranium carbide (UC) and compared their performance to uranium oxide (UO2) which is thefuel form used in the vast majority of commercial

Plasma Processing Systems for the Manufacture of Refractory Metals and their Alloys for Military Needs

DTIC Science & Technology

1978-10-09

melting point is around 4000*K. An exceedingly interesting feature of these solidification composites is the formation of fibrous MC type carbide ...the matrix could be refractory metal binary alloys with copper or uranium and the eutectic phase could be carbide of tungsten, * molybdenum, tantalum or...42 Accs -n or - *DTTI Tf Avn ! -7ll ’ i CrDi t , l’’*i,;. LIST OF FIGURES FIG. 1 Flow Diagram of Cemented Carbide Manufacture

The radioactive beam facility ALTO

NASA Astrophysics Data System (ADS)

Essabaa, Saïd; Barré-Boscher, Nicole; Cheikh Mhamed, Maher; Cottereau, Evelyne; Franchoo, Serge; Ibrahim, Fadi; Lau, Christophe; Roussière, Brigitte; Saïd, Abdelhakim; Tusseau-Nenez, Sandrine; Verney, David

2013-12-01

The Transnational Access facility ALTO (TNA07-ENSAR/FP7) has been commissioned and received from the French safety authorities, the operation license. It is allowed to run at nominal intensity to produce 1011 fissions/s in a thick uranium carbide target by photo-fission using a 10 μA, 50 MeV electron beam. In addition the recent success in operating the selective laser ion source broadens the physics program with neutron-rich nuclear beams possible at this facility installed at IPN Orsay. The facility also aims at being a test bench for the SPIRAL2 project. In that framework an ambitious R&D program on the target ion source system is being developed.

Phase formation during the carbothermic reduction of eudialyte concentrate

NASA Astrophysics Data System (ADS)

Krasikov, S. A.; Upolovnikova, A. G.; Sitnikova, O. A.; Ponomarenko, A. A.; Agafonov, S. N.; Zhidovinova, S. V.; Maiorov, D. V.

2013-07-01

The phase transformations of eudialyte concentrate during the carbothermic reduction in the temperature range 25-2000°C are studied by thermodynamic simulation, differential thermal analysis, and X-ray diffraction. As the temperature increases to 1500°C, the following phases are found to form sequentially: iron and manganese carbides, free iron, niobium carbide, iron silicides, silicon and titanium carbides, and free silicon. Strontium, yttrium, and uranium in the temperature range under study are not reduced and are retained in an oxide form, and insignificant reduction of zirconium oxides with the formation of carbide ZrC is possible only at temperatures above 1500°C.

As-cast uranium-molybdenum based metallic fuel candidates and the effects of carbon addition

NASA Astrophysics Data System (ADS)

Blackwood, Van Stephen

The objective of this research was to develop and recommend a metallic nuclear fuel candidate that lowered the onset temperature of gamma phase formation comparable or better than the uranium-10 wt. pct. molybdenum alloy, offered a solidus temperature as high or higher than uranium-10 wt. pct. zirconium (1250°C), and stabilized the fuel phase against interaction with iron and steel at least as much as uranium-10 wt. pct. zirconium stabilized the fuel phase. Two new as-cast alloy compositions were characterized to assess thermal equilibrium boundaries of the gamma phase field and the effect of carbon addition up to 0.22 wt. pct. The first system investigated was uranium- x wt. pct. M where x ranged between 5-20 wt. pct. M was held at a constant ratio of 50 wt. pct. molybdenum, 43 wt. pct. titanium, and 7 wt. pct. zirconium. The second system investigated was the uranium-molybdenum-tungsten system in the range 90 wt. pct. uranium - 10 wt. pct. molybdenum - 0 wt. pct. tungsten to 80 wt. pct. uranium - 10 wt. pct. molybdenum - 10 wt. pct. tungsten. The results showed that the solidus temperature increased with increased addition of M up to 12.5 wt. pct. for the uranium-M system. Alloy additions of titanium and zirconium were removed from uranium-molybdenum solid solution by carbide formation and segregation. The uranium-molybdenum-tungsten system solidus temperature increased to 1218°C at 2.5 wt. pct. with no significant change in temperature up to 5 wt. pct. tungsten suggesting the solubility limit of tungsten had been reached. Carbides were observed with surrounding areas enriched in both molybdenum and tungsten. The peak solidus temperatures for the alloy systems were roughly the same at 1226°C for the uranium-M system and 1218°C for the uranium-molybdenum-tungsten system. The uranium-molybdenum-tungsten system required less alloy addition to achieve similar solidus temperatures as the uranium-M system.

Some Thermodynamic Features of Uranium-Plutonium Nitride Fuel in the Course of Burnup

NASA Astrophysics Data System (ADS)

Rusinkevich, A. A.; Ivanov, A. S.; Belov, G. V.; Skupov, M. V.

2017-12-01

Calculation studies on the effect of carbon and oxygen impurities on the chemical and phase compositions of nitride uranium-plutonium fuel in the course of burnup are performed using the IVTANTHERMO code. It is shown that the number of moles of UN decreases with increasing burnup level, whereas UN1.466, UN1.54, and UN1.73 exhibit a considerable increase. The presence of oxygen and carbon impurities causes an increase in the content of the UN1.466, UN1.54 and UN1.73 phases in the initial fuel by several orders of magnitude, in particular, at a relatively low temperature. At the same time, the presence of impurities abruptly reduces the content of free uranium in unburned fuel. Plutonium in the considered system is contained in form of Pu, PuC, PuC2, Pu2C3, and PuN. Plutonium carbides, as well as uranium carbides, are formed in small amounts. Most of the plutonium remains in the form of nitride PuN, whereas unbound Pu is present only in the areas with a low burnup level and high temperatures.

The South African isotope facility project

NASA Astrophysics Data System (ADS)

Bark, R. A.; Barnard, A. H.; Conradie, J. L.; de Villiers, J. G.; van Schalkwyk, P. A.

2018-05-01

The South African Isotope Facility (SAIF) is a project in which iThemba LABS plans to build a radioactive-ion beam (RIB) facility. The project is divided into the Accelerator Centre of Exotic Isotopes (ACE Isotopes) and the Accelerator Centre for Exotic Beams (ACE Beams). For ACE Isotopes, a high-current, 70 MeV cyclotron will be acquired to take radionuclide production off the existing Separated Sector Cyclotron (SSC). A freed up SSC will then be available for an increased tempo of nuclear physics research and to serve as a driver accelerator for the ACE Beams project, in which protons will be used for the direct fission of Uranium, producing beams of fission fragments. The ACE Beams project has begun with "LeRIB" - a Low Energy RIB facility, now under construction. In a collaboration with INFN Legnaro, the target/ion-source "front-end" will be a copy of the front-end developed for the SPES project. A variety of targets may be inserted into the SPES front-end; a uranium-carbide target has been designed to produce up to 2 × 1013 fission/s using a 70 MeV proton beam of 150 µA intensity.

Uranium passivation by C + implantation: A photoemission and secondary ion mass spectrometry study

NASA Astrophysics Data System (ADS)

Nelson, A. J.; Felter, T. E.; Wu, K. J.; Evans, C.; Ferreira, J. L.; Siekhaus, W. J.; McLean, W.

2006-03-01

Implantation of 33 keV C + ions into polycrystalline U 238 with a dose of 4.3 × 10 17 cm -2 produces a physically and chemically modified surface layer that prevents further air oxidation and corrosion. X-ray photoelectron spectroscopy and secondary ion mass spectrometry were used to investigate the surface chemistry and electronic structure of this C + ion implanted polycrystalline uranium and a non-implanted region of the sample, both regions exposed to air for more than a year. In addition, scanning electron microscopy was used to examine and compare the surface morphology of the two regions. The U 4f, O 1s and C 1s core-level and valence band spectra clearly indicate carbide formation in the modified surface layer. The time-of-flight secondary ion mass spectrometry depth profiling results reveal an oxy-carbide surface layer over an approximately 200 nm thick UC layer with little or no residual oxidation at the carbide layer/U metal transitional interface.

Overview of the ISOL facility for the RISP

NASA Astrophysics Data System (ADS)

Woo, H. J.; Kang, B. H.; Tshoo, K.; Seo, C. S.; Hwang, W.; Park, Y.-H.; Yoon, J. W.; Yoo, S. H.; Kim, Y. K.; Jang, D. Y.

2015-02-01

The key feature of the Isotope Separation On-Line (ISOL) facility is its ability to provide high-intensity and high-quality beams of neutron-rich isotopes with masses in the range of 80-160 by means of a 70-MeV proton beam directly impinging on uranium-carbide thin-disc targets to perform forefront research in nuclear structure, nuclear astrophysics, reaction dynamics and interdisciplinary fields like medical, biological and material sciences. The technical design of the 10-kW and the 35-kW direct fission targets with in-target fission rates of up to 1014 fissions/s has been finished, and for the development of the ISOL fission-target chemistry an initial effort has been made to produce porous lanthanum-carbide (LaCx) discs as a benchmark for the final production of porous UCx discs. For the production of various beams, three classes of ion sources are under development at RISP (Rare Isotope Science Project), the surface ion source, the plasma ion source (FEBIAD), the laser ion source, and the engineering design of the FEBIAD is in progress for prototype fabrication. The engineering design of the ISOL target/ion source front-end system is also in progress, and a prototype will be used for an off-line test facility in front of the pre-separator. The technical designs of other basic elements at the ISOL facility, such as the RF-cooler, the high-resolution mass separator, and the A/q separator, have been finished, and the results, along with the future plans, are introduced.

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

DOE PAGES

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; ...

2017-03-10

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

The SPES High Power ISOL production target

NASA Astrophysics Data System (ADS)

Andrighetto, A.; Corradetti, S.; Ballan, M.; Borgna, F.; Manzolaro, M.; Scarpa, D.; Monetti, A.; Rossignoli, M.; Silingardi, R.; Mozzi, A.; Vivian, G.; Boratto, E.; De Ruvo, L.; Sattin, N.; Meneghetti, G.; Oboe, R.; Guerzoni, M.; Margotti, A.; Ferrari, M.; Zenoni, A.; Prete, G.

2016-11-01

SPES (Selective Production of Exotic Species) is a facility under construction at INFN-LNL (Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro), aimed to produce intense neutron-rich radioactive ion beams (RIBs). These will be obtained using the ISOL (Isotope Separation On-Line) method, bombarding a uranium carbide target with a proton beam of 40MeV energy and currents up to 200μA. The target configuration was designed to obtain a high number of fissions, up to 1013 per second, low power deposition and fast release of the produced isotopes. The exotic isotopes generated in the target are ionized, mass separated and re-accelerated by the ALPI superconducting LINAC at energies of 10AMeV and higher, for masses in the region of A = 130 amu , with an expected rate on the secondary target up to 109 particles per second. In this work, recent results on the R&D activities regarding the SPES RIB production target-ion source system are reported.

Targets used in the production of radioactive ion beams at the HRIBF

NASA Astrophysics Data System (ADS)

Stracener, D. W.; Alton, G. D.; Auble, R. L.; Beene, J. R.; Mueller, P. E.; Bilheux, J. C.

2004-03-01

Radioactive ion beams are produced at the Holifield Radioactive Ion Beam Facility using the Isotope Separation On-Line (ISOL) technique where the atoms are produced in a thick target, transported to an ion source, ionized, and extracted from the ion source to form an ion beam. These radioactive ion beams are then accelerated to energies of a few MeV per nucleon and delivered to experimental stations for use in nuclear physics and nuclear astrophysics experiments. At the heart of this facility is the RIB production target, where the radioactive nuclei are produced using beams of light ions (p, d, 3He, α) to induce nuclear reactions in the target nuclei. Several target materials have been developed and used successfully, including Al 2O 3, HfO 2, SiC, CeS, liquid Ge, liquid Ni, and a low-density matrix of uranium carbide. The details of these targets and some of the target developments that led to the delivery of high-quality radioactive ion beams are discussed in this paper.

Surface spectroscopy studies of the oxidation behavior of uranium

NASA Astrophysics Data System (ADS)

Bloch, J.; Atzmony, U.; Dariel, M. P.; Mintz, M. H.; Shamir, N.

1982-02-01

Auger electron spectroscopy (AES) and X-ray photoelectron spectroscopy (XPS) techniques were utilized to study the oxidation behavior of clean uranium surfaces, at very low pressures of various atmospheres (UHV, H 2, O 2, and CO 2), at room temperature. Both for O 2 and CO 2, a precursor chemisorbed oxygen species has been identified at the very initial stage of the oxidation reaction. This chemisorbed oxygen transforms to the oxide form at a rate which depends on the pressure of the oxidizing atmosphere. Residual gaseous carbon compounds which are present even under UHV conditions result in the simultaneous formation of surface carbide which accompanies the initial stage of oxidation. This carbide however decomposes later as oxidation proceeds. Adventitious hydrocarbon adsorption occurs on the formed oxide layer.

Passivation of uranium towards air corrosion by N 2+ and C + ion implantation

NASA Astrophysics Data System (ADS)

Arkush, R.; Mintz, M. H.; Shamir, N.

2000-10-01

The passivation of uranium surfaces against air corrosion, by ion implantation processes was studied, using surface analysis methods. Implanting 45 keV N +2 and C + ions produces thin modified surface layers with gradual gradients of the corresponding compounds (i.e., nitrides and carbides, respectively), which avoid the formation of discontinuous interfaces typical to coatings. Such gradual interfaces impart excellent mechanical stability and adhesion to the modified layers, in spite of the large misfit between the metal substrate and the implantation on induced compounds. It turns out that these layers provide an almost absolute protection against air corrosion. A rapid initial stage of oxidation of the modified surface layers takes place, forming very thin protective oxidation zones (1-4 nm thick), which practically stop further air oxidation for years. The mechanism of the initial oxidation stage of the modified layers seems to vary with the type of surface (i.e., either nitrides or carbides). However, in any case the protection ability of the formed oxidation products is excellent, probably due to the close match between these compounds and the underlying nitrides or carbides.

The SPES surface ionization source

NASA Astrophysics Data System (ADS)

Manzolaro, M.; D'Agostini, F.; Monetti, A.; Andrighetto, A.

2017-09-01

Ion sources and target systems play a crucial role in isotope separation on line facilities, determining the main characteristics of the radioactive ion beams available for experiments. In the context of the selective production of exotic species (SPES) facility, a 40 MeV, 200 μA proton beam directly impinges a uranium carbide target, generating approximately 1013 fissions per second. The radioactive isotopes produced by the 238U fissions are delivered to the 1+ ion source by means of a tubular transfer line. Here they can be ionized and subsequently accelerated toward the experimental areas. In this work, the characterization of the surface ionization source currently adopted for the SPES facility is presented, taking as a reference ionization efficiency and transversal emittance measurements. The effects of long term operation at high temperature are also illustrated and discussed.

Thermal-electric numerical simulation of a surface ion source for the production of radioactive ion beams

NASA Astrophysics Data System (ADS)

Manzolaro, Mattia; Meneghetti, Giovanni; Andrighetto, Alberto

2010-11-01

In a facility for the production of radioactive ion beams (RIBs), the target system and the ion source are the most critical objects. In the context of the Selective Production of Exotic Species (SPES) project, a proton beam directly impinges a Uranium Carbide production target, generating approximately 10 13 fissions per second. The radioactive isotopes produced by the 238U fissions are then directed to the ion source to acquire a charge state. After that, the radioactive ions obtained are transported electrostatically to the subsequent areas of the facility. In this work the surface ion source at present adopted for the SPES project is studied by means of both analytical and numerical thermal-electric models. The theoretical results are compared with temperature and electric potential difference measurements.

Simulations for the future converter of the e-linac for the TRIUMF ARIEL facility

NASA Astrophysics Data System (ADS)

Lebois, M.; Bricault, P.

2011-09-01

In the next years, TRIUMF activity will be focused on building a new facility to produce very intense neutron rich radioactive ion beams. Unlike others ISOL facilities, the e-linac primary beam, that will induce the fission, is an intense electron beam (50 MeV energy and 10 mA intensity). This challenging choice, which make this installation unique, despite the ALTO facility, makes an average fission rate of 1013-14fissions/s in the target.This beam is sent on an uranium carbide target (UCx), but due to its power, it is essential to insert a "converter" on the beam path to avoid a target overheating. The purpose of this converter is to convert electrons into Bremsstralhung radiation. The γ rays produce excite the dipole resonance of 23892U (15 MeV) inducing fission. Energy deposition, fission rate and thermal behavior were simulated using Monte Carlo techniques are presented in this paper

Secondary neutrons as the main source of neutron-rich fission products in the bombardment of a thick U target by 1 GeV protons

NASA Astrophysics Data System (ADS)

Barzakh, A. E.; Lhersonneau, G.; Batist, L. Kh.; Fedorov, D. V.; Ivanov, V. S.; Mezilev, K. A.; Molkanov, P. L.; Moroz, F. V.; Orlov, S. Yu.; Panteleev, V. N.; Volkov, Yu. M.; Alyakrinskiy, O.; Barbui, M.; Stroe, L.; Tecchio, L. B.

2011-05-01

The diffusion-effusion model has been used to analyse the release and yields of Fr and Cs isotopes from uranium carbide targets of very different thicknesses (6.3 and 148 g/cm2) bombarded by a 1 GeV proton beam. Release curves of several isotopes of the same element and production efficiency versus decay half-life are well fitted with the same set of parameters. Comparison of efficiencies for neutron-rich and neutron-deficient Cs isotopes enables separation of the contributions from the primary ( p + 238U) and secondary (n + 238U) reactions to the production of neutron-rich Cs isotopes. A rather simple calculation of the neutron contribution describes these data fairly well. The FLUKA code describes the primary and secondary-reaction contributions to the Cs isotopes production efficiencies for different targets quite well.

Electrical-thermal-structural finite element simulation and experimental study of a plasma ion source for the production of radioactive ion beams

NASA Astrophysics Data System (ADS)

Manzolaro, M.; Meneghetti, G.; Andrighetto, A.; Vivian, G.

2016-03-01

The production target and the ion source constitute the core of the selective production of exotic species (SPES) facility. In this complex experimental apparatus for the production of radioactive ion beams, a 40 MeV, 200 μA proton beam directly impinges a uranium carbide target, generating approximately 1013 fissions per second. The transfer line enables the unstable isotopes generated by the 238U fissions in the target to reach the ion source, where they can be ionized and finally accelerated to the subsequent areas of the facility. In this work, the plasma ion source currently adopted for the SPES facility is analyzed in detail by means of electrical, thermal, and structural numerical models. Next, theoretical results are compared with the electric potential difference, temperature, and displacement measurements. Experimental tests with stable ion beams are also presented and discussed.

The adsorption of methyl iodide on uranium and uranium dioxide: Surface characterization using X-ray photoelectron spectroscopy (XPS) and Auger electron spectroscopy (AES)

NASA Astrophysics Data System (ADS)

Dillard, J. G.; Moers, H.; Klewe-Nebenius, H.; Kirch, G.; Pfennig, G.; Ache, H. J.

1984-09-01

The adsorption of methyl iodide on uranium and on uranium dioxide has been studied at 25 °C. Surfaces of the substrates were characterized before and after adsorption by X-ray photoelectron spectroscopy (XPS) and Auger electron spectroscopy (AES). The XPS binding energy results indicate that CH 3I adsorption on uranium yields a carbide-type carbon, UC, and uranium iodide, UI 3. On uranium dioxide the carbon electron binding energy measurements are consistent with the formation of a hydrocarbon, —CH 3-type moiety. The interpretation of XPS and AES spectral features for CH 3I adsorption on uranium suggest that a complex dissociative adsorption reaction takes place. Adsorption of CH 3I on UO 2 occurs via a dissociative process. Saturation coverage occurs on uranium at approximately two langmuir (1 L = 10 -6 Torr s) exposure whereas saturation coverage on uranium dioxide is found at about five langmuir.

Characterization and fabrication of target materials for RIB generation

NASA Astrophysics Data System (ADS)

Welton, R. F.; Janney, M. A.; Mueller, P. E.; Ortman, W. K.; Rauniyar, R.; Stracener, D. W.; Williams, C. L.

2001-07-01

This report discusses two techniques developed at the Oak Ridge National Laboratory (ORNL) that are employed for the fabrication and characterization of targets used in the production of Radioactive Ion Beams (RIBs). First, our method of in-house fabrication of uranium carbide targets is discussed. We have found that remarkably uniform coatings of UC2 can be formed on the microstructure of porous C matrices. The technique has been used to form UC2 layers on highly thermally conductive graphitic foams. Targets fabricated in this fashion have been tested under low-intensity proton bombardment and yields of selected radioactive species are reported. This report also describes an off-line test stand for the investigation of effusive and diffusive transport in RIB target/ion sources. Permeation rates of gases and vapors passing through a high temperature membrane or through an effusive channel constructed from the material under investigation are recorded. Diffusion coefficients and adsorption enthalpies, which characterize the interaction of RIB species with materials of the target/ion source, are extracted from the time profile of the recorded data. Examples of diffusion, effusion, and conductance measurements are provided.

The Alto Tandem and Isol Facility at IPN Orsay

NASA Astrophysics Data System (ADS)

Franchoo, Serge

Alto is an infrastructure for experimental nuclear physics in France that comprises both an on-line isotope-separation facility based on the photofission of uranium and a stable-ion beam facility based on a 14.5-MV tandem accelerator. The isotope-separation on-line section of Alto is dedicated to the production of neutron-rich radioactive ion beams (RIB) from the interaction of the γ-flux induced by a 50-MeV 10-µA electron beam in a uranium-carbide target. It is dimensioned for 1011 fissions per second. The RIB facility is exploited in alternating mode with the tandem-based section of Alto, capable of accelerating both light ions for nuclear astrophysics and heavy ions for γ-spectroscopy. The facility thereby offers the opportunity to deliver beams to a large range of physics programmes from nuclear to interdisciplinary physics. In this article, we present the Alto facility as well as some of the highlights and prospects of the experimental programme.

Understanding Release from Actinide Targets -- Recent Results from RIB Development

NASA Astrophysics Data System (ADS)

Kronenberg, Andreas; Carter, H. K.; Spejewski, E. H.; Stracener, D. W.

2006-10-01

Development of ion beams of short-lived isotopes is crucial for modern nuclear structure and nuclear astrophysics. The Holifield Radioactive Ion Beam Facility at Oak Ridge National Laboratory uses the ISOL (Isotope Separation Online) technique to provide radioactive ion beams. So far, uranium carbide has been used as a target to produce neutron-rich fission fragments. Thermodynamic calculations indicate the possibility of in-situ chemical side band formations of volatile species of refractory metals, such as V and Re. These elements release out of oxide targets after production in a nuclear reaction, and can occur only through in-situ formation of their volatile oxide. These have been confirmed experimentally. The results from recent, more detailed investigations of ThO2, UB4 and other actinide targets as well as conclusions from systematic studies will be presented. This research was sponsored by the NNSA under Stewardship Science Academic Alliance program through DOE Cooperative Agreement # DE-FC03-3NA00143.

Target materials for exotic ISOL beams

NASA Astrophysics Data System (ADS)

Gottberg, A.

2016-06-01

The demand for intensity, purity, reliability and availability of short-lived isotopes far from stability is steadily high, and considerably exceeding the supply. In many cases the ISOL (Isotope Separation On-Line) method can provide beams of high intensity and purity. Limitations in terms of accessible chemical species and minimum half-life are driven mainly by chemical reactions and physical processes inside of the thick target. A wide range of materials are in use, ranging from thin metallic foils and liquids to refractory ceramics, while poly-phasic mixed uranium carbides have become the reference target material for most ISOL facilities world-wide. Target material research and development is often complex and especially important post-irradiation analyses are hindered by the high intrinsic radiotoxicity of these materials. However, recent achievements have proven that these investigations are possible if the effort of different facilities is combined, leading to the development of new material matrices that can supply new beams of unprecedented intensity and beam current stability.

Identification and distribution of inclusions in derby and ingot uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schwartz, C.M.; Vaughan, D.A.

1953-08-31

Inclusions in derby and ingot uranium have been identified by X-ray diffraction methods. Metallographic and microradiographic examinations have shown that inclusions of MgF2, UO2, UN, and UO are concentrated at the top of derby and ingot metal by gravity separation. UC inclusions are distributed throughout the ingot metal. The amount of the carbide phase in the ingot varies with the temperature maintained during remelt of derby metal.

Quantitative ion beam analysis of M-C-O systems: application to an oxidized uranium carbide sample

NASA Astrophysics Data System (ADS)

Martin, G.; Raveu, G.; Garcia, P.; Carlot, G.; Khodja, H.; Vickridge, I.; Barthe, M. F.; Sauvage, T.

2014-04-01

A large variety of materials contain both carbon and oxygen atoms, in particular oxidized carbides, carbon alloys (as ZrC, UC, steels, etc.), and oxycarbide compounds (SiCO glasses, TiCO, etc.). Here a new ion beam analysis methodology is described which enables quantification of elemental composition and oxygen concentration profile over a few microns. It is based on two procedures. The first, relative to the experimental configuration relies on a specific detection setup which is original in that it enables the separation of the carbon and oxygen NRA signals. The second concerns the data analysis procedure i.e. the method for deriving the elemental composition from the particle energy spectrum. It is a generic algorithm and is here successfully applied to characterize an oxidized uranium carbide sample, developed as a potential fuel for generation IV nuclear reactors. Furthermore, a micro-beam was used to simultaneously determine the local elemental composition and oxygen concentration profiles over the first microns below the sample surface. This method is adapted to the determination of the composition of M?C?O? compounds with a sensitivity on elemental atomic concentrations around 1000 ppm.

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

NASA Astrophysics Data System (ADS)

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

2015-05-01

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

Equation of state for shock compression of distended solids

NASA Astrophysics Data System (ADS)

Grady, Dennis; Fenton, Gregg; Vogler, Tracy

2014-05-01

Shock Hugoniot data for full-density and porous compounds of boron carbide, silicon dioxide, tantalum pentoxide, uranium dioxide and playa alluvium are investigated for the purpose of equation-of-state representation of intense shock compression. Complications of multivalued Hugoniot behavior characteristic of highly distended solids are addressed through the application of enthalpy-based equations of state of the form originally proposed by Rice and Walsh in the late 1950's. Additive measures of cold and thermal pressure intrinsic to the Mie-Gruneisen EOS framework is replaced by isobaric additive functions of the cold and thermal specific volume components in the enthalpy-based formulation. Additionally, experimental evidence reveals enhancement of shock-induced phase transformation on the Hugoniot with increasing levels of initial distension for silicon dioxide, uranium dioxide and possibly boron carbide. Methods for addressing this experimentally observed feature of the shock compression are incorporated into the EOS model.

Equation of State for Shock Compression of High Distension Solids

NASA Astrophysics Data System (ADS)

Grady, Dennis

2013-06-01

Shock Hugoniot data for full-density and porous compounds of boron carbide, silicon dioxide, tantalum pentoxide, uranium dioxide and playa alluvium are investigated for the purpose of equation-of-state representation of intense shock compression. Complications of multivalued Hugoniot behavior characteristic of highly distended solids are addressed through the application of enthalpy-based equations of state of the form originally proposed by Rice and Walsh in the late 1950's. Additivity of cold and thermal pressure intrinsic to the Mie-Gruneisen EOS framework is replaced by isobaric additive functions of the cold and thermal specific volume components in the enthalpy-based formulation. Additionally, experimental evidence supports acceleration of shock-induced phase transformation on the Hugoniot with increasing levels of initial distention for silicon dioxide, uranium dioxide and possibly boron carbide. Methods for addressing this experimentally observed facet of the shock compression are introduced into the EOS model.

UCx target preparations and characterizations

NASA Astrophysics Data System (ADS)

Andrighetto, Alberto; Corradetti, Stefano; Manzolaro, Mattia; Scarpa, Daniele; Monetti, Alberto; Rossignoli, Massimo; Borgna, Francesca; Ballan, Michele; Agostini, Mattia; D'Agostini, Fabio; Ferrari, Matteo; Zenoni, Aldo

2018-05-01

The Target-Ion Source unit is the core of an ISOL-RIB facility. Many international ISOL facilities have chosen different layouts of this unit. Many research groups are involved in research and development of targets capable of dissipating high power and, at the same time, be able to have a fast isotope release. This is mandatory in order to produce beams of short half-life isotopes. The research of new materials with advanced microstructural features is crucial in this field. The design of a proper target is indeed strictly related to the obtainment of porous refractory materials, which are capable to work under extreme conditions (temperatures up to 2000 °C in high vacuum) with a high release efficiency. For SPES, the second generation Italian ISOL-RIB Facility, the target will be made of uranium carbide (UCx) in which, by fission induced by a proton beam of 40 MeV of energy (8 kW of power), isotopes in the 60-160 amu mass region are produced. The current technological developments are also crucial in the study of third generation ISOL facilities.

Carbide fuels for nuclear thermal propulsion

NASA Astrophysics Data System (ADS)

Matthews, R. B.; Blair, H. T.; Chidester, K. M.; Davidson, K. V.; Stark, W. E.; Storms, E. K.

1991-09-01

A renewed interest in manned exploration of space has revitalized interest in the potential for advancing nuclear rocket technology developed during the 1960's. Carbide fuel performance, melting point, stability, fabricability and compatibility are key technology issues for advanced Nuclear Thermal Propulsion reactors. The Rover fuels development ended with proven carbide fuel forms with demonstrated operating temperatures up to 2700 K for over 100 minutes. The next generation of nuclear rockets will start where the Rover technology ended, but with a more rigorous set of operating requirements including operating lifetime to 10 hours, operating temperatures greater that 3000 K, low fission product release, and compatibility. A brief overview of Rover/NERVA carbide fuel development is presented. A new fuel form with the highest potential combination of operating temperature and lifetime is proposed that consists of a coated uranium carbide fuel sphere with built-in porosity to contain fission products. The particles are dispersed in a fiber reinforced ZrC matrix to increase thermal shock resistance.

Thermionic System Evaluation Test: Ya-21U System Topaz International Program

DTIC Science & Technology

1996-07-01

by enriched uranium dioxide (U02) fuel pellets, as illustrated by Figure 5. The work section of the converter contained 34 TFEs that provided power...power system. This feature permitted transportation of the highly enriched uranium oxide fuel in separate containers from the space power system and...by Figure 8. The radial reflector contained three safety and nine control drums. Each drum contained a section of boron carbide (B4C) neutron poison

Thermodynamic and experimental study of UC powders ignition

NASA Astrophysics Data System (ADS)

Le Guyadec, F.; Rado, C.; Joffre, S.; Coullomb, S.; Chatillon, C.; Blanquet, E.

2009-09-01

Mixed plutonium and uranium carbide (UPuC) is considered as a possible fuel material for future nuclear reactors. However, UPuC is pyrophoric and fine powders of UPuC are subject to temperature increase due to oxidation with air and possible ignition during conditioning and handling. In a first approach and to allow easier experimental conditions, this study was undertaken on uranium monocarbide (UC) with the aim to determine safe handling conditions for the production and reprocessing of uranium carbide fuels. The reactivity of uranium monocarbide in oxidizing atmosphere was studied in order to analyze the ignition process. Experimental thermogravimetric analysis (TGA) and differential thermal analysis (DTA) revealed that UC powder obtained by arc melting and milling is highly reactive in air at about 200 °C. The phases formed at the various observed stages of the oxidation process were analyzed by X-ray diffraction. At the same time, ignition was analyzed thermodynamically along isothermal sections of the U-C-O ternary diagram and the pressure of the gas produced by the UC + O 2 reaction was calculated. Two possible oxidation schemes were identified on the U-C-O phase diagram and assumptions are proposed concerning the overall oxidation and ignition paths. It is particularly important to understand the mechanisms involved since temperatures as high as 2500 °C could be reached, leading to CO(g) production and possibly to a blast effect.

Dimensional Analysis and Extended Hydrodynamic Theory Applied to Long-Rod Penetration of Ceramics

DTIC Science & Technology

2016-07-01

thick ceramic targets by tungsten long rod projectiles. The ceramics are AD-995 alumina, aluminum nitride, silicon carbide, and boron carbide. Test...of confined thick ceramic targets by tungsten long rod projectiles. The ceramics are AD-995 alumina, aluminum nitride, silicon carbide, and boron ...since the mid 20th century. Popular candidate ceramics for such systems include alumina, aluminum nitride, boron carbide, silicon carbide, and titanium

Status of the ion sources developments for the Spiral2 project at GANILa)

NASA Astrophysics Data System (ADS)

Lehérissier, P.; Bajeat, O.; Barué, C.; Canet, C.; Dubois, M.; Dupuis, M.; Flambard, J. L.; Frigot, R.; Jardin, P.; Leboucher, C.; Lemagnen, F.; Maunoury, L.; Osmond, B.; Pacquet, J. Y.; Pichard, A.; Thuillier, T.; Peaucelle, C.

2012-02-01

The SPIRAL 2 facility is now under construction and will deliver either stable or radioactive ion beams. First tests of nickel beam production have been performed at GANIL with a new version of the large capacity oven, and a calcium beam has been produced on the heavy ion low energy beam transport line of SPIRAL 2, installed at LPSC Grenoble. For the production of radioactive beams, several target/ion-source systems (TISSs) are under development at GANIL as the 2.45 GHz electron cyclotron resonance ion source, the surface ionization source, and the oven prototype for heating the uranium carbide target up to 2000 °C. The existing test bench has been upgraded for these developments and a new one, dedicated for the validation of the TISS before mounting in the production module, is under design. Results and current status of these activities are presented.

Thermal-electric coupled-field finite element modeling and experimental testing of high-temperature ion sources for the production of radioactive ion beams

NASA Astrophysics Data System (ADS)

Manzolaro, M.; Meneghetti, G.; Andrighetto, A.; Vivian, G.; D'Agostini, F.

2016-02-01

In isotope separation on line facilities the target system and the related ion source are two of the most critical components. In the context of the selective production of exotic species (SPES) project, a 40 MeV 200 μA proton beam directly impinges a uranium carbide target, generating approximately 1013 fissions per second. The radioactive isotopes produced in this way are then directed to the ion source, where they can be ionized and finally accelerated to the subsequent areas of the facility. In this work both the surface ion source and the plasma ion source adopted for the SPES facility are presented and studied by means of numerical thermal-electric models. Then, numerical results are compared with temperature and electric potential difference measurements, and finally the main advantages of the proposed simulation approach are discussed.

NUCLEAR REACTOR ELEMENT

DOEpatents

Sanz, M.C.; Scully, C.N.

1961-06-27

The patented fuel element is a hexagonal graphite body having an axial channel therethrough. The graphite is impregnated with uranium which is concentrated near the axial channel. Layers of tantalum nitride and tantalum carbide are disposed on the surface of the body confronting the channel.

Rational control of the interlayer space inside two-dimensional titanium carbides for highly efficient uranium removal and imprisonment.

PubMed

Wang, Lin; Tao, Wuqing; Yuan, Liyong; Liu, Zhirong; Huang, Qing; Chai, Zhifang; Gibson, John K; Shi, Weiqun

2017-11-07

Though two-dimensional early transition metal carbides and carbonitrides (MXenes) have attracted extensive interest recently, their superb abilities in various scientific applications always suffer from the very narrow interlayer space inside the multilayered structure. Here we demonstrate an unprecedented large adsorption capacity enhancement of Ti 3 C 2 T x toward radionuclide removal via a hydrated intercalation strategy. By rational control of the interlayer space, the potential for imprisoning the representative actinide U(vi) inside multilayered Ti 3 C 2 T x was also confirmed.

Constitutional studies on carbide nuclear fuels for fast breeder reactors. III. The uranium-barium-carbon ternary system at 1400°C, with observations on the uranium-strontium-carbon system

NASA Astrophysics Data System (ADS)

Peatfield, M.; Brett, N. H.; Potter, P. E.

1980-03-01

The ternary system U-Ba-C has been examined at 1400°C and the solid-state compatibility lines established. No compound formation was found to occur and solubility effects were found to be minimal. A tentative examination of compositions in the U-Sr-C system indicates that it is of a similar form to that of the U-Ba-C system.

Application of an X-ray Fluorescence Instrument to Helicopter Wear Debris Analysis

DTIC Science & Technology

2008-04-01

from magnesium (Mg) to uranium (U) using two X-ray detection sensors: a FOCUS 5+ detector AlX-ray tube X-ray Detector 1. Incident X-ray...zinc (Zn), whilst the PIN detector is used to detect elements from calcium (Ca) to uranium (U) [4]. Elements between calcium (Ca) to zinc (Zn) can be... carbide paper, however polishing is not a normal sample preparation requirement for the Twin-X (see Figure 16). The samples were placed polished side

Chemical compatibility between UO2 fuel and SiC cladding for LWRs. Application to ATF (Accident-Tolerant Fuels)

NASA Astrophysics Data System (ADS)

Braun, James; Guéneau, Christine; Alpettaz, Thierry; Sauder, Cédric; Brackx, Emmanuelle; Domenger, Renaud; Gossé, Stéphane; Balbaud-Célérier, Fanny

2017-04-01

Silicon carbide-silicon carbide (SiC/SiC) composites are considered to replace the current zirconium-based cladding materials thanks to their good behavior under irradiation and their resistance under oxidative environments at high temperature. In the present work, a thermodynamic analysis of the UO2±x/SiC system is performed. Moreover, using two different experimental methods, the chemical compatibility of SiC towards uranium dioxide, with various oxygen contents (UO2±x) is investigated in the 1500-1970 K temperature range. The reaction leads to the formation of mainly uranium silicides and carbides phases along with CO and SiO gas release. Knudsen Cell Mass Spectrometry is used to measure the gas release occurring during the reaction between UO2+x and SiC powders as function of time and temperature. These experimental conditions are representative of an open system. Diffusion couple experiments with pellets are also performed to study the reaction kinetics in closed system conditions. In both cases, a limited chemical reaction is observed below 1700 K, whereas the reaction is enhanced at higher temperature due to the decomposition of SiC leading to Si vaporization. The temperature of formation of the liquid phase is found to lie between 1850 < T < 1950 K.

SOLID SOLUTION CARBIDES ARE THE KEY FUELS FOR FUTURE NUCLEAR THERMAL PROPULSION

NASA Technical Reports Server (NTRS)

Panda, Binayak; Hickman, Robert R.; Shah, Sandeep

2005-01-01

Nuclear thermal propulsion uses nuclear energy to directly heat a propellant (such as liquid hydrogen) to generate thrust for space transportation. In the 1960 s, the early Rover/Nuclear Engine for Rocket Propulsion Application (NERVA) program showed very encouraging test results for space nuclear propulsion but, in recent years, fuel research has been dismal. With NASA s renewed interest in long-term space exploration, fuel researchers are now revisiting the RoverMERVA findings, which indicated several problems with such fuels (such as erosion, chemical reaction of the fuel with propellant, fuel cracking, and cladding issues) that must be addressed. It is also well known that the higher the temperature reached by a propellant, the larger the thrust generated from the same weight of propellant. Better use of fuel and propellant requires development of fuels capable of reaching very high temperatures. Carbides have the highest melting points of any known material. Efforts are underway to develop carbide mixtures and solid solutions that contain uranium carbide, in order to achieve very high fuel temperatures. Binary solid solution carbides (U, Zr)C have proven to be very effective in this regard. Ternary carbides such as (U, Zr, X) carbides (where X represents Nb, Ta, W, and Hf) also hold great promise as fuel material, since the carbide mixtures in solid solution generate a very hard and tough compact material. This paper highlights past experience with early fuel materials and bi-carbides, technical problems associated with consolidation of the ingredients, and current techniques being developed to consolidate ternary carbides as fuel materials.

Thermal Stability of Milled Nanocrystalline Tungsten Powders

DTIC Science & Technology

2011-05-01

v Acknowledgments This research is supported by mission funding from the Depleted Uranium Replacement Program and is the result of collaborative...L.; Reddy, R. G. Synthesis of Titanium Carbide Nano-powders by Thermal Plasma. Scripta Materialia 2005, 52 (12), 1253–1258. 4. Karagedov, G. R

Chemical reactivity of CVC and CVD SiC with UO 2 at high temperatures

DOE PAGES

Silva, Chinthaka M.; Katoh, Yutai; Voit, Stewart L.; ...

2015-02-11

Two types of silicon carbide (SiC) synthesized using two different vapor deposition processes were embedded in UO 2 pellets and evaluated for their potential chemical reaction with UO 2. While minor reactivity between chemical-vapor-composited (CVC) SiC and UO 2 was observed at comparatively low temperatures of 1100 and 1300 C, chemical-vapor-deposited (CVD) SiC did not show any such reactivity, according to microstructural investigations. But, both CVD and CVC SiCs showed some reaction with UO 2 at a higher temperature (1500 C). Elemental maps supported by phase maps obtained using electron backscatter diffraction indicated that CVC SiC was more reactive thanmore » CVD SiC at 1500 C. Moreover, this investigation indicated the formation of uranium carbides and uranium silicide chemical phases such as UC, USi 2, and U 3Si 2 as a result of SiC reaction with UO 2.« less

In situ synthesis and characterization of uranium carbide using high temperature neutron diffraction

NASA Astrophysics Data System (ADS)

Reiche, H. Matthias; Vogel, Sven C.; Tang, Ming

2016-04-01

We investigated the formation of UCx from UO2+x and graphite in situ using neutron diffraction at high temperatures with particular focus on resolving the conflicting reports on the crystal structure of non-quenchable cubic UC2. The agents were UO2 nanopowder, which closely imitates nano grains observed in spent reactor fuels, and graphite powder. In situ neutron diffraction revealed the onset of the UO2 + 2C → UC + CO2 reaction at 1440 °C, with its completion at 1500 °C. Upon further heating, carbon diffuses into the uranium carbide forming C2 groups at the octahedral sites. This resulting high temperature cubic UC2 phase is similar to the NaCl-type structure as proposed by Bowman et al. Our novel experimental data provide insights into the mechanism and kinetics of formation of UC as well as characteristics of the high temperature cubic UC2 phase which agree with proposed rotational rehybridization found from simulations by Wen et al.

Uranium carbide dissolution in nitric solution: Sonication vs. silent conditions

NASA Astrophysics Data System (ADS)

Virot, Matthieu; Szenknect, Stéphanie; Chave, Tony; Dacheux, Nicolas; Moisy, Philippe; Nikitenko, Sergey I.

2013-10-01

The dissolution of uranium carbide (UC) in nitric acid media is considered by means of power ultrasound (sonication) or magnetic stirring. The induction period required to initiate UC dissolution was found to be dramatically shortened when sonicating a 3 M nitric solution (Ar, 20 kHz, 18 W cm-2, 20 °C). At higher acidity, magnetic stirring offers faster dissolution kinetics compared to sonication. Ultrasound-assisted UC dissolution is found to be passivated after ∼60% dissolution and remains incomplete whatever the acidity which is confirmed by ICP-AES, LECO and SEM-EDX analyses. In general, the kinetics of UC dissolution is linked to the in situ generation of nitrous acid in agreement with the general mechanism of UC dissolution; the nitrous acid formation is reported to be faster under ultrasound at low acidity due to the nitric acid sonolysis. The carbon balance shared between the gaseous, liquid, and solid phases is strongly influenced by the applied dissolution procedure and HNO3 concentration.

Modeling Early-Stage Processes of U-10 Wt.%Mo Alloy Using Integrated Computational Materials Engineering Concepts

NASA Astrophysics Data System (ADS)

Wang, Xiaowo; Xu, Zhijie; Soulami, Ayoub; Hu, Xiaohua; Lavender, Curt; Joshi, Vineet

2017-12-01

Low-enriched uranium alloyed with 10 wt.% molybdenum (U-10Mo) has been identified as a promising alternative to high-enriched uranium. Manufacturing U-10Mo alloy involves multiple complex thermomechanical processes that pose challenges for computational modeling. This paper describes the application of integrated computational materials engineering (ICME) concepts to integrate three individual modeling components, viz. homogenization, microstructure-based finite element method for hot rolling, and carbide particle distribution, to simulate the early-stage processes of U-10Mo alloy manufacture. The resulting integrated model enables information to be passed between different model components and leads to improved understanding of the evolution of the microstructure. This ICME approach is then used to predict the variation in the thickness of the Zircaloy-2 barrier as a function of the degree of homogenization and to analyze the carbide distribution, which can affect the recrystallization, hardness, and fracture properties of U-10Mo in subsequent processes.

Feasibility study of a small, thorium-based fission power system for space and terrestrial applications

NASA Astrophysics Data System (ADS)

Worrall, Michael Jason

One of the current challenges facing space exploration is the creation of a power source capable of providing useful energy for the entire duration of a mission. Historically, radioisotope batteries have been used to provide load power, but this conventional system may not be capable of sustaining continuous power for longer duration missions. To remedy this, many forays into nuclear powered spacecraft have been investigated, but no robust system for long-term power generation has been found. In this study, a novel spin on the traditional fission power system that represents a potential optimum solution is presented. By utilizing mature High Temperature Gas Reactor (HTGR) technology in conjunction with the capabilities of the thorium fuel cycle, we have created a light-weight, long-term power source capable of a continuous electric power output of up to 70kW for over 15 years. This system relies upon a combination of fissile, highly-enriched uranium dioxide and fertile thorium carbide Tri-Structural Isotropic (TRISO) fuel particles embedded in a hexagonal beryllium oxide matrix. As the primary fissile material is consumed, the fertile material breeds new fissile material leading to more steady fuel loading over the lifetime of the core. Reactor control is achieved through an innovative approach to the conventional boron carbide neutron absorber by utilizing sections of borated aluminum placed in rotating control drums within the reflector. Borated aluminum allows for much smaller boron concentrations, thus eliminating the potential for 10B(n,alpha)6Li heating issues that are common in boron carbide systems. A wide range of other reactivity control systems are also investigated, such as a radially-split rotating reflector. Lastly, an extension of the design to a terrestrial based system is investigated. In this system, uranium enrichment is dropped to 20 percent in order to meet current regulations, a solid uranium-zirconium hydride fissile driver replaces the uranium dioxide TRISO particles, and the moderating material is changed from beryllium oxide to graphite. These changes result in an increased core size, but the same long-term power generation potential is achieved. Additionally, small amounts of erbium are added to the hydride matrix to further extend core lifetime.

A thermodynamic study of the gaseous thorium carbides, ThC, ThC2, ThC3, ThC4, ThC5, and ThC6

NASA Astrophysics Data System (ADS)

Gupta, Satish K.; Gingerich, Karl A.

1980-02-01

Six gaseous carbides of thorium, ThCn(n=1-6), have been identified in a Knudsen effusion mass spectrometric investigation of the vapor phase above a thorium-uranium-rhodium-graphite system at high temperatures. The partial pressures of the thorium containing species were measured as a function of temperature in the 2300-2700 °K range. Third law enthalpies for the reactions Th(g)+nC(graphite) =ThCn, n=1 to 6, and of various other homogeneous and heterogeneous reactions were evaluated. By combining the experimental enthalpies with appropriate thermodynamic data taken from literature, the following values for the atomization energies ΔH °at,298, and standard heats of formation ΔH °f,298 of thorium carbides have been derived:

Out-of-pile creep behavior of uranium carbide

NASA Technical Reports Server (NTRS)

Wright, T. R.; Seltzer, M. S.

1974-01-01

Compression creep tests were investigated on various UC-based fuel materials having a variation in both density and composition. Specimens were prepared by casting and by hot pressing. Steady-state creep rates were measured under vacuum at 1400 to 1800 C in the stress range 500-4000 psi.

The RIB production target for the SPES project

NASA Astrophysics Data System (ADS)

Monetti, Alberto; Andrighetto, Alberto; Petrovich, Carlo; Manzolaro, Mattia; Corradetti, Stefano; Scarpa, Daniele; Rossetto, Francesco; Martinez Dominguez, Fernando; Vasquez, Jesus; Rossignoli, Massimo; Calderolla, Michele; Silingardi, Roberto; Mozzi, Aldo; Borgna, Francesca; Vivian, Gianluca; Boratto, Enrico; Ballan, Michele; Prete, Gianfranco; Meneghetti, Giovanni

2015-10-01

Facilities making use of the Isotope Separator On-Line (ISOL) method for the production of Radioactive Ion Beams (RIB) attract interest because they can be used for nuclear structure and reaction studies, astrophysics research and interdisciplinary applications. The ISOL technique is based on the fast release of the nuclear reaction products from the chosen target material together with their ionization into short-lived nuclei beams. Within this context, the SPES (Selective Production of Exotic Species) facility is now under construction in Italy at INFN-LNL (Istituto Nazionale di Fisica Nucleare — Laboratori Nazionali di Legnaro). The SPES facility will produce RIBs mainly from n-rich isotopes obtained by a 40 MeV cyclotron proton beam (200 μA) directly impinging on a uranium carbide multi-foil fission target. The aim of this work is to describe and update, from a comprehensive point of view, the most important results obtained by the analysis of the on-line behavior of the SPES production target assembly. In particular an improved target configuration has been studied by comparing different codes and physics models: the thermal analyses and the isotope production are re-evaluated. Then some consequent radioprotection aspects, which are essential for the installation and operation of the facility, are presented.

Formation of uranium and cerium nitrides by the reaction of carbides with NH 3 and N 2/H 2 stream

NASA Astrophysics Data System (ADS)

Nakagawa, Takashi; Matsuoka, Hirotaka; Sawa, Masaji; Hirota, Masayuki; Miyake, Masanobu; Katsura, Masahiro

1997-08-01

UC or CeC 2 were converted into U 2N 3 or CeN by the use of NH 3 or an N 2/H 2 gas mixture. A stream of NH 3 works not only as a nitriding agent but also as a carbon clearing agent due to its high nitriding and hydriding activities. When the carbide is converted into nitride, carbon is liberated. Some experiments were performed in order to examine the role of the carbon activity of carbon materials (amorphous carbon or graphite) in the formation of CH 4.

Coupling XRD, EXAFS, and 13C NMR to study the effect of the carbon stoichiometry on the local structure of UC(1±x).

PubMed

Carvajal Nuñez, U; Martel, L; Prieur, D; Lopez Honorato, E; Eloirdi, R; Farnan, I; Vitova, T; Somers, J

2013-10-07

A series of uranium carbide samples, prepared by arc melting with a C/U ratio ranging from 0.96 to 1.04, has been studied by X-ray diffraction (XRD), (13)C nuclear magnetic resonance (NMR), and extended X-ray absorption fine structure (EXAFS). XRD determines phase uniqueness and the increase of the lattice parameter versus the carbon content. In contrast, (13)C NMR detects the different carbon environments in the lattice and in this study, clearly identifies the presence of discrete peaks for carbon in the octahedral lattice site in UC and an additional peak associated with excess carbon in hyperstoichiometric samples. Two peaks associated with different levels of carbon deficiency are detected for all hypostoichiometric compositions. More than one carbon environment is always detected by (13)C NMR. This exemplifies the difficulty in obtaining a perfect stoichiometric uranium monocarbide UC(1.00). The (13)C MAS spectra of uranium carbides exhibit the effects resulting from the carbon content on both the broadening of the peaks and on the Knight shift. An abrupt spectral change occurs between hypo- and hyperstoichiometric samples. The results obtained by EXAFS highlight subtle differences between the different stoichiometries, and in the hyperstoichiometric samples, the EXAFS results are consistent with the excess carbon atoms being in the tetrahedral interstitial position.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerczak, Tyler J.; Smith, Kurt R.; Petrie, Christian M.

Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept consisting of a spherical fuel kernel made of uranium oxide and uranium carbide, surrounded by a porous carbonaceous buffer layer and successive layers of dense inner pyrolytic carbon (IPyC), silicon carbide (SiC) deposited by chemical vapor , and dense outer pyrolytic carbon (OPyC). This fuel concept is being considered for advanced reactor applications such as high temperature gas-cooled reactors (HTGRs) and molten salt reactors (MSRs), as well as for accident-tolerant fuel for light water reactors (LWRs). Development and implementation of TRISO fuel for these reactor concepts support the US Departmentmore » of Energy (DOE) Office of Nuclear Energy mission to promote safe, reliable nuclear energy that is sustainable and environmentally friendly. During operation, the SiC layer serves as the primary barrier to metallic fission products and actinides not retained in the kernel. It has been observed that certain fission products are released from TRISO fuel during operation, notably, Ag, Eu, and Sr [1]. Release of these radioisotopes causes safety and maintenance concerns.« less

Evaluation of phases in Pu-C-O and (U, Pu)-C-O systems by X-ray diffractometry and chemical analysis

NASA Astrophysics Data System (ADS)

Jain, G. C.; Ganguly, C.

1993-12-01

Preparation and characterisation of the carbides of uranium, plutonium and mixed uranium and plutonium form a part of advanced fuel development programs for fast breeder reactors. In the present study, the compositions of the phases of Pu-C-O and (U.Pu)-C-O systems have been determined by chemical analysis and lattice parameter measurement. The carbide samples have been prepared by vacuum carbothermic synthesis of tabletted oxide-graphite powder mixture. Dependence of stoichiometry of Pu 2C 3 phase on the oxygen content of Pu(C,O) phase in Pu(C,O) + Pu 2C 3 phase mixture has been evaluated. Stoichiometry and oxygen solubility of (U 0.3Pu 0.7)(C,O) phase in multiple phase mixture have been determined. Segregation of plutonium in (U,Pu) 2C 3 phase of (U,Pu)(C,O) + (U,Pu) 2C 3 phase mixture and its dependence on the oxygen content of (U,Pu)(C,O) phase have also been determined from the measurement of the lattice parameter of (U,Pu) 2C 3 phase.

Thermodynamic study of the gaseous thorium carbides, ThC, ThC/sub 2/, ThC/sub 3/, ThC/sub 4/, ThC/sub 5/, and ThC/sub 6/

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gupta, S.K.; Gingerich, K.A.

Six gaseous carbides of thorium, ThC/sub n/(n=1--6), have been identified in a Knudsen effusion mass spectrometric investigation of the vapor phase above a thorium--uranium--rhodium--graphite system at high temperatures. The partial pressures of the thorium containing species were measured as a function of temperature in the 2300--2700 /sup 0/K range. Third law enthalpies for the reactions Th(g)+nC(graphite) =ThC/sub n/, n=1 to 6, and of various other homogeneous and heterogeneous reactions were evaluated. By combining the experimental enthalpies with appropriate thermodynamic data taken from literature, the following values for the atomization energies ..delta..H /sup 0//sub at,298/, and standard heats of formation ..delta..H/supmore » tsdegree//sub f/,298 of thorium carbides have been derived:« less

Powder Processing of High Temperature Cermets and Carbides at Marshall Space Flight Center

NASA Technical Reports Server (NTRS)

Salvail, Pat; Panda, Binayak; Hickman, Robert R.

2007-01-01

The Materials and Processing Laboratory at NASA Marshall Space Flight Center is developing Powder Metallurgy (PM) processing techniques for high temperature cermet and carbide material consolidation. These new group of materials would be utilized in the nuclear core for Nuclear Thermal Rockets (NTR). Cermet materials offer several advantages for NTR such as retention of fission products and fuels, better thermal shock resistance, hydrogen compatibility, high thermal conductivity, and high strength. Carbide materials offer the highest operating temperatures but are sensitive to thermal stresses and are difficult to process. To support the effort, a new facility has been setup to process refractory metal, ceramic, carbides and depleted uranium-based powders. The facility inciudes inert atmosphere glove boxes for the handling of reactive powders, a high temperature furnace, and powder processing equipment used for blending, milling, and sieving. The effort is focused on basic research to identify the most promising compositions and processing techniques. Several PM processing methods including Cold and Hot Isostatic Pressing are being evaluated to fabricate samples for characterization and hot hydrogen testing.

Molecular structure of uranium carbides: Isomers of UC3

NASA Astrophysics Data System (ADS)

Zalazar, M. Fernanda; Rayón, Víctor M.; Largo, Antonio

2013-03-01

In this article, the most relevant isomers of uranium tricarbide are studied through quantum chemical methods. It is found that the most stable isomer has a fan geometry in which the uranium atom is bonded to a quasilinear C3 unit. Both, a rhombic and a ring CU(C2) structures are found about 104-125 kJ/mol higher in energy. Other possible isomers including linear geometries are located even higher. For each structure, we provide predictions for those molecular properties (vibrational frequencies, IR intensities, dipole moments) that could eventually help in their experimental detection. We also discuss the possible routes for the formation of the different UC3 isomers as well as the bonding situation by means of a topological analysis of the electron density.

Molecular structure of uranium carbides: isomers of UC3.

PubMed

Zalazar, M Fernanda; Rayón, Víctor M; Largo, Antonio

2013-03-21

In this article, the most relevant isomers of uranium tricarbide are studied through quantum chemical methods. It is found that the most stable isomer has a fan geometry in which the uranium atom is bonded to a quasilinear C3 unit. Both, a rhombic and a ring CU(C2) structures are found about 104-125 kJ/mol higher in energy. Other possible isomers including linear geometries are located even higher. For each structure, we provide predictions for those molecular properties (vibrational frequencies, IR intensities, dipole moments) that could eventually help in their experimental detection. We also discuss the possible routes for the formation of the different UC3 isomers as well as the bonding situation by means of a topological analysis of the electron density.

Decommissioning ALARA programs Cintichem decommissioning experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adler, J.J.; LaGuardia, T.S.

1995-03-01

The Cintichem facility, originally the Union Carbide Nuclear Company (UCNC) Research Center, consisted primarily of a 5MW pool type reactor linked via a four-foot-wide by twelve-foot-deep water-filled canal to a bank of five adjacent hot cells. Shortly after going into operations in the early 1960s, the facility`s operations expanded to provide various reactor-based products and services to a multitude of research, production, medical, and education groups. From 1968 through 1972, the facility developed a process of separating isotopes from mixed fission products generated by irradiating enriched Uranium target capsules. By the late 1970s, 20 to 30 capsules were being processedmore » weekly, with about 200,000 curies being produced per week. Several isotopes such as Mo{sup 99}, I{sup 131}, and Xe{sup 133} were being extracted for medical use.« less

METHOD OF MAKING FUEL BODIES

DOEpatents

Goeddel, W.V.; Simnad, M.T.

1963-04-30

This patent relates to a method of making a fuel compact having a matrix of carbon or graphite which carries the carbides of fissile material. A nuclear fuel material selected from the group including uranium and thorium carbides, silicides, and oxides is first mixed both with sufficient finely divided carbon to constitute a matrix in the final product and with a diffusional bonding material selected from the class consisting of zirconium, niobium, molybdenum, titanium, nickel, chromium, and silicon. The mixture is then heated at a temperature of 1500 to 1800 nif- C while maintaining it under a pressure of over about 2,000 pounds per square inch. Preferably, heating is accomplished by the electrical resistance of the compact itself. (AEC)

Electron-spectroscopy studies of clean thorium and uranium surfaces. Chemisorption and initial stages of reaction with O2, CO, and CO2

NASA Astrophysics Data System (ADS)

McLean, W.; Colmenares, C. A.; Smith, R. L.; Somorjai, G. A.

1982-01-01

The adsorption of O2, CO, and CO2 on the thorium (111) crystal face and on polycrystalline α-uranium has been investigated by x-ray photoelectron spectroscopy, Auger electron spectroscopy (AES), and secondary-ion mass spectroscopy (SIMS) at 300 K. Oxygen adsorption on both metals resulted in the formation of the metal dioxide. CO and CO2 adsorption on Th(111) produced species derived from atomic carbon and oxygen; the presence of molecular CO was also detected. Only atomic carbon and oxygen were observed on uranium. Elemental depth profiles by AES and SIMS indicated that the carbon produced by the dissociation of CO or CO2 diffused into the bulk of the metals to form a carbide, while the oxygen remained on their surfaces as an oxide.

DYNAMIC PROPERTIES OF SHOCK LOADED THIN URANIUM FOILS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robbins, D. L.; Kelly, A. M.; Alexander, D. J.

A series of spall experiments has been completed with thin depleted uranium targets, nominally 0.1 mm thick. The first set of uranium spall targets was cut and ground to final thickness from electro-refined, high-purity, cast uranium. The second set was rolled to final thickness from low purity uranium. The impactors for these experiments were laser-launched 0.05-mm thick copper flyers, 3 mm in diameter. Laser energies were varied to yield a range of flyer impact velocities. This resulted in varying degrees of damage to the uranium spall targets, from deformation to complete spall or separation at the higher velocities. Dynamic measurementsmore » of the uranium target free surface velocities were obtained with dual velocity interferometers. Uranium targets were recovered and sectioned after testing. Free surface velocity profiles were similar for the two types of uranium, but spall strengths (estimated from the magnitude of the pull-back signal) are higher for the high-purity cast uranium. Velocity profiles and microstructural evidence of spall from the sectioned uranium targets are presented.« less

Enhanced thermal conductivity of uranium dioxide-silicon carbide composite fuel pellets prepared by Spark Plasma Sintering (SPS)

NASA Astrophysics Data System (ADS)

Yeo, S.; Mckenna, E.; Baney, R.; Subhash, G.; Tulenko, J.

2013-02-01

Uranium dioxide (UO2)-10 vol% silicon carbide (SiC) composite fuel pellets were produced by oxidative sintering and Spark Plasma Sintering (SPS) at a range of temperatures from 1400 to 1600 °C. Both SiC whiskers and SiC powder particles were utilized. Oxidative sintering was employed over 4 h and the SPS sintering was employed only for 5 min at the highest hold temperature. It was noted that composite pellets sintered by SPS process revealed smaller grain size, reduced formation of chemical products, higher density, and enhanced interfacial contact compared to the pellets made by oxidative sintering. For given volume of SiC, the pellets with powder particles yielded a smaller grain size than pellets with SiC whiskers. Finally thermal conductivity measurements at 100 °C, 500 °C, and 900 °C revealed that SPS sintered UO2-SiC composites exhibited an increase of up to 62% in thermal conductivity compared to UO2 pellets, while the oxidative sintered composite pellets revealed significantly inferior thermal conductivity values. The current study points to the improved processing capabilities of SPS compared to oxidative sintering of UO2-SiC composites.

Experimental studies of laser-ablated zirconium carbide plasma plumes: Fuel corrosion diagnostic development

NASA Astrophysics Data System (ADS)

Wantuck, P. J.; Butt, D. P.; Sappey, A. D.

Understanding the corrosion behavior of nuclear fuel materials, such as refractory carbides, in a high temperature hydrogen environment is critical for several proposed nuclear thermal propulsion (NTP) concepts. Monitoring the fuel corrosion products is important not only for understanding corrosion characteristics, but to assess the performance of an actual, operating nuclear propulsion system as well. In this paper, we describe an experimental study initiated to develop, test, and subsequently utilize non-intrusive, laser-based diagnostics to characterize the gaseous product species which are expected to evolve during the exposure of representative fuel samples to hydrogen. Laser ablation is used to produce high temperature, vapor plumes from solid solution, uranium-free, zirconium carbide (ZrC) forms for probing by other laser diagnostic methods, predominantly laser-induced fluorescence (LIF). We discuss the laser ablation technique, results of plume emission measurements, as well as the use of planar LIF to image both the ZrC plumes and actual NTP fuel corrosion constituents.

Spes: An intense source of Neutron-Rich Radioactive Beams at Legnaro

NASA Astrophysics Data System (ADS)

Andrighetto, A.; Manzolaro, M.; Corradetti, S.; Scarpa, D.; Monetti, A.; Rossignoli, M.; Ballan, M.; Borgna, F.; D'Agostini, F.; Gramegna, F.; Prete, G.; Meneghetti, G.; Ferrari, M.; Zenoni, A.

2018-02-01

The Isotope Separation On-Line (ISOL) method for the production of Radioactive Ion Beams (RIB) is attracting significant interest in the worldwide nuclear physics community. Within this context the SPES (Selective Production of Exotic Species) RIB facility is now under construction at INFN LNL (Istituto Nazionale di Fisica Nucleare Laboratori Nazionali di Legnaro). This technique is established as one of the main techniques for high intensity and high quality beams production. The SPES facility will produce n-rich isotopes by means of a 40 MeV proton beam, emitted by a cyclotron, impinging on a uranium carbide multi-foil fission target. The aim of this work is to describe the most important results obtained by the study of the on-line behavior of the SPES production target assembly. This target system will produce RIBs at a rate of about 1013 fissions per second, it will be able to dissipate a total power of up to 10 kW, and it is planned to work continuously for 2 week-runs of irradiation. ISOL beams of 24 different elements will be produced, therefore a target and ion source development is ongoing to ensure a great variety of produced isotopes and to improve the beam intensity and purity.

Method of depositing a high-emissivity layer

DOEpatents

Wickersham, Charles E.; Foster, Ellis L.

1983-01-01

A method of depositing a high-emissivity layer on a substrate comprising RF sputter deposition of a carbide-containing target in an atmosphere of a hydrocarbon gas and a noble gas. As the carbide is deposited on the substrate the hydrocarbon gas decomposes to hydrogen and carbon. The carbon deposits on the target and substrate causing a carbide/carbon composition gradient to form on the substrate. At a sufficiently high partial pressure of hydrocarbon gas, a film of high-emissivity pure carbon will eventually form over the substrate.

Theoretical Model for Volume Fraction of UC, 235U Enrichment, and Effective Density of Final U 10Mo Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Devaraj, Arun; Prabhakaran, Ramprashad; Joshi, Vineet V.

2016-04-12

The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective 235U enrichment in the U-10Mo matrix after accounting for loss of 235U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reilly, Sean Douglas; May, Iain; Copping, Roy

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted tomore » concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.« less

FUEL COMPOSITION FOR NUCLEAR REACTORS

DOEpatents

Andersen, J.C.

1963-08-01

A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

Simultaneous determination of uranium carbide dissolution products by capillary zone electrophoresis.

PubMed

Sladkov, Vladimir; Fourest, Blandine

2009-03-20

Separation and simultaneous determination of a number of organic acid anions (oxalate, mellitate, trimellitate and benzoate) and U(VI) with direct UV detection is developed for analysis of uranium carbide (UC) dissolution products by capillary zone electrophoresis (CZE). Reverse polarity mode is used. It is found that complex formation of U(VI) with carbonate, used as a carrier electrolyte, allows U(VI) to be determined, as negatively charged species, in a single run with organic acid anions. Some parameters such as pH value, composition of electrolyte and detection wavelength are optimized. Under the chosen conditions (carbonate buffer (ionic strength of 100 mM), pH 9.8, 0.15 mM tetradecyltrimethylammonium bromide (TTAB)) a complete separation is achieved. Calibration plots are linear in two ranges of concentration for U(VI) ( approximately 1 x 10(-5) to 1 x 10(-3)), mellitate and trimellitate ( approximately 5 x 10(-6) to 5 x 10(-4)), and about one range ( approximately 1 x 10(-4) to 5 x 10(-3)) for oxalate and benzoate. Accuracy of the procedure is checked by the "added-found" method in standard mixture solutions. Relative standard deviation is within the range of 2-10% and the recovery is in the range of 90-110%. This method is applied for the analysis of real UC dissolution samples.

Method for fracturing silicon-carbide coatings on nuclear-fuel particles

DOEpatents

Turner, Lloyd J.; Willey, Melvin G.; Tiegs, Sue M.; Van Cleve, Jr., John E.

1982-01-01

This invention is a device for fracturing particles. It is designed especially for use in "hot cells" designed for the handling of radioactive materials. In a typical application, the device is used to fracture a hard silicon-carbide coating present on carbon-matrix microspheres containing nuclear-fuel material, such as uranium or thorium compounds. To promote remote control and facilitate maintenance, the particle breaker is pneumatically operated and contains no moving parts. It includes means for serially entraining the entrained particles on an anvil housed in a leak-tight chamber. The flow rate of the gas is at a value effecting fracture of the particles; preferably, it is at a value fracturing them into product particulates of fluidizable size. The chamber is provided with an outlet passage whose cross-sectional area decreases in the direction away from the chamber. The outlet is connected tangentially to a vertically oriented vortex-flow separator for recovering the product particulates entrained in the gas outflow from the chamber. The invention can be used on a batch or continuous basis to fracture the silicon-carbide coatings on virtually all of the particles fed thereto.

Device for fracturing silicon-carbide coatings on nuclear-fuel particles

DOEpatents

Turner, L.J.; Willey, M.G.; Tiegs, S.M.; Van Cleve, J.E. Jr.

This invention is a device for fracturing particles. It is designed especially for use in hot cells designed for the handling of radioactive materials. In a typical application, the device is used to fracture a hard silicon-carbide coating present on carbon-matrix microspheres containing nuclear-fuel materials, such as uranium or thorium compounds. To promote remote control and facilitate maintenance, the particle breaker is pneumatically operated and contains no moving parts. It includes means for serially entraining the entrained particles on an anvil housed in a leak-tight chamber. The flow rate of the gas is at a value effecting fracture of the particles; preferably, it is at a value fracturing them into product particulates of fluidizable size. The chamber is provided with an outlet passage whose cross-sectional area decreases in the direction away from the chamber. The outlet is connected tangentially to a vertically oriented vortex-flow separator for recovering the product particulates entrained in the gas outflow from the chamber. The invention can be used on a batch or continuous basis to fracture the silicon-carbide coatings on virtually all of the particles fed thereto.

Distribution of Pd, Ag & U in the SiC Layer of an Irradiated TRISO Fuel Particle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomas M. Lillo; Isabella J. van Rooyen

2014-08-01

The distribution of silver, uranium and palladium in the silicon carbide (SiC) layer of an irradiated TRISO fuel particle was studied using samples extracted from the SiC layer using focused ion beam (FIB) techniques. Transmission electron microscopy in conjunction with energy dispersive x-ray spectroscopy was used to identify the presence of the specific elements of interest at grain boundaries, triple junctions and precipitates in the interior of SiC grains. Details on sample fabrication, errors associated with measurements of elemental migration distances and the distances migrated by silver, palladium and uranium in the SiC layer of an irradiated TRISO particle frommore » the AGR-1 program are reported.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, Igor; Ronningen, Reginald Martin

The research studied one-step and two-step Isotope Separation on Line (ISOL) targets for future radioactive beam facilities with high driver-beam power through advanced computer simulations. As a target material uranium carbide in the form of foils was used because of increasing demand for actinide targets in rare-isotope beam facilities and because such material was under development in ISAC at TRIUMF when this project started. Simulations of effusion were performed for one-step and two step targets and the effects of target dimensions and foil matrix were studied. Diffusion simulations were limited by availability of diffusion parameters for UC x material atmore » reduced density; however, the viability of the combined diffusion?effusion simulation methodology was demonstrated and could be used to extract physical parameters such as diffusion coefficients and effusion delay times from experimental isotope release curves. Dissipation of the heat from the isotope-producing targets is the limiting factor for high-power beam operation both for the direct and two-step targets. Detailed target models were used to simulate proton beam interactions with the targets to obtain the fission rates and power deposition distributions, which were then applied in the heat transfer calculations to study the performance of the targets. Results indicate that a direct target, with specification matching ISAC TRIUMF target, could operate in 500-MeV proton beam at beam powers up to ~40 kW, producing ~8 10 13 fission/s with maximum temperature in UCx below 2200 C. Targets with larger radius allow higher beam powers and fission rates. For the target radius in the range 9-mm to 30-mm the achievable fission rate increases almost linearly with target radius, however, the effusion delay time also increases linearly with target radius.« less

Design of an EBIS charge breeder system for rare-isotope beams

NASA Astrophysics Data System (ADS)

Park, Young-Ho; Son, Hyock-Jun; Kim, Jongwon

2016-09-01

Rare-isotope beams will be produced by using the isotope separation on-line (ISOL) system at the Rare Isotope Science Project (RISP). A proton cyclotron is the driver accelerator for ISOL targets, and uranium carbide (UCx) will be a major target material. An isotope beam of interest extracted from the target will be ionized and selected by using a mass separator. The beam emittance will then be reduced by using a radio-frequency quadrupole (RFQ) cooler before the beam is injected into the electron-beam ion-source (EBIS) charge breeder (CB). The maximum electron beam current of the EBIS is 3 A from a cathode made of IrCe in an applied magnetic field of 0.2 T. The size of the electron beam is compressed by magnetic fields of up to 6 T caused in the charge-breeding region by a superconducting solenoid. The design of EBIS-CB was performed by using mechanics as well as beam optics. A test stand for the electron gun and its collector, which can take an electron-beam power of 20 kW, are under construction. The gun assembly was first tested by using a high-voltage pulse so as to measure its perveance. The design of the EBIS, along with its test stand, is described.

Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, J.; Yuan, B.; Jin, M.

2012-07-01

Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate themore » demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)« less

The nucleation and growth of uranium on the basal plane of graphite studied by scanning tunneling microscopy

NASA Astrophysics Data System (ADS)

Tench, R. J.

1992-11-01

For the first time, nanometer scale uranium clusters were created on the basal plane of highly oriented pyrolytic graphite by laser ablation under ultra-high vacuum conditions. The physical and chemical properties of these clusters were investigated by scanning tunneling microscopy (STM) as well as standard surface science techniques. Auger electron and X-ray photoelectron spectroscopies found the uranium deposit to be free of contamination and showed that no carbide had formed with the underlying graphite. Clusters with sizes ranging from 42 to 630 sq A were observed upon initial room temperature deposition. Surface diffusion of uranium was observed after annealing the substrate above 800 K, as evidenced by the decreased number density and the increased size of the clusters. Preferential depletion of clusters on terraces near step edges as a result of annealing was observed. The activation energy for diffusion deduced from these measurements was found to be 15 Kcal/mole. Novel formation of ordered uranium thin films was observed for coverages greater than two monolayers after annealing above 900 K. These ordered films displayed islands with hexagonally faceted edges rising in uniform step heights characteristic of the unit cell of the P-phase of uranium. In addition, atomic resolution STM images of these ordered films indicated the formation of the (beta)-phase of uranium. The chemical properties of these surfaces were investigated and it was shown that these uranium films had a reduced oxidation rate in air as compared to bulk metal and that STM imaging in air induced a polarity-dependent enhancement of the oxidation rate.

Recovery and recycling of uranium from rejected coated particles for compact high temperature reactors

NASA Astrophysics Data System (ADS)

Pai, Rajesh V.; Mollick, P. K.; Kumar, Ashok; Banerjee, J.; Radhakrishna, J.; Chakravartty, J. K.

2016-05-01

UO2 microspheres prepared by internal gelation technique were coated with pyrolytic carbon and silicon carbide using CVD technique. The particles which were not meeting the specifications were rejected. The rejected/failed UO2 based coated particles prepared by CVD technique was used for oxidation and recovery and recycling. The oxidation behaviour of sintered UO2 microspheres coated with different layers of carbon and SiC was studied by thermal techniques to develop a method for recycling and recovery of uranium from the failed/rejected coated particles. It was observed that the complete removal of outer carbon from the spheres is difficult. The crushing of microspheres enabled easier accessibility of oxygen and oxidation of carbon and uranium at 800-1000 °C. With the optimized process of multiple crushing using die & plunger and sieving the broken coated layers, we could recycle around fifty percent of the UO2 microspheres which could be directly recoated. The rest of the particles were recycled using a wet recycling method.

Atomization from a tantalum surface in graphite furnace atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Gregoire, D. C.; Chakrabarti, C. L.

The mechanism of atom formation of U, V, Mo, Ni, Mn, Cu and Mg atomized from pyrolytic graphite and tantalum metal surfaces has been studied. The mechanism of atom formation for U from a graphite tube atomizer is reported for the first time. The peak absorbance for U and Cu is increased by factors of 59.7 and 2.0, respectively, whereas that of V, Mo and Ni is reduced by several orders of magnitude when they are atomized from a tantalum metal surface. The peak absorbance of Mn and Mg is not appreciably affected by the material of the atomization surface. Interaction of Mn and Mg with the graphite surface and formation of their refractory carbides was found to be negligible. Uranium forms a refractory carbide when heated from a graphite surface.

Cryochemical and CVD processing of shperical carbide fuels for propulsion reactors

NASA Astrophysics Data System (ADS)

Blair, H. Thomas; Carroll, David W.; Matthews, R. Bruce

1991-01-01

Many of the nuclear propulsion reactor concepts proposed for a manned mission to Mars use a coated spherical particle fuel form similar to that used in the Rover and NERVA propulsion reactors. The formation of uranium dicarbide microspheres using a cryochemical process and the coating of the UC2 spheres with zirconium carbide using chemical vapor deposition are being developed at Los Alamos National Laboratory. The cryochemical process is described with a discussion of the variables affecting the sphere formation and carbothermic reduction to produce UC2 spheres from UO2. Emphasis is placed on minimizing the wastes produced by the process. The ability to coat particles with ZrC was recaptured, and improvements in the process and equipment were developed. Volatile organometallic precursors were investigated as alternatives to the original ZrCl4 precursor.

Fission product palladium-silicon carbide interaction in htgr fuel particles

NASA Astrophysics Data System (ADS)

Minato, Kazuo; Ogawa, Toru; Kashimura, Satoru; Fukuda, Kousaku; Shimizu, Michio; Tayama, Yoshinobu; Takahashi, Ishio

1990-07-01

Interaction of fission product palladium (Pd) with the silicon carbide (SiC) layer was observed in irradiated Triso-coated uranium dioxide particles for high temperature gas-cooled reactors (HTGR) with an optical microscope and electron probe microanalyzers. The SiC layers were attacked locally or the reaction product formed nodules at the attack site. Although the main element concerned with the reaction was palladium, rhodium and ruthenium were also detected at the corroded areas in some particles. Palladium was detected on both the hot and cold sides of the particles, but the corroded areas and the palladium accumulations were distributed particularly on the cold side of the particles. The observed Pd-SiC reaction depths were analyzed on the assumption that the release of palladium from the fuel kernel controls the whole Pd-SiC reaction.

Development of a polarized 31Mg+ beam as a spin-1/2 probe for BNMR

NASA Astrophysics Data System (ADS)

Levy, C. D. P.; Pearson, M. R.; Dehn, M. H.; Karner, V. L.; Kiefl, R. F.; Lassen, J.; Li, R.; MacFarlane, W. A.; McFadden, R. M. L.; Morris, G. D.; Stachura, M.; Teigelhöfer, A.; Voss, A.

2016-12-01

A 28 keV beam of 31Mg+ ions was extracted from a uranium carbide, proton-beam-irradiated target coupled to a laser ion source. The ion beam was nuclear-spin polarized by collinear optical pumping on the 2it {S}_{1/2}-2it {P}_{1/2} transition at 280 nm. The polarization was preserved by an extended 1 mT guide field as the beam was transported via electrostatic bends into a 2.5 T longitudinal magnetic field. There the beam was implanted into a single crystal MgO target and the beta decay asymmetry was measured. Both hyperfine ground states were optically pumped with a single frequency light source, using segmentation of the beam energy, which boosted the polarization by approximately 50 % compared to pumping a single ground state. The total decay asymmetry of 0.06 and beam intensity were sufficient to provide a useful spin-1/2 beam for future BNMR experiments. A variant of the method was used previously to optically pump the full Doppler-broadened absorption profile of a beam of 11Be+ with a single-frequency light source.

Nuclear reactor target assemblies, nuclear reactor configurations, and methods for producing isotopes, modifying materials within target material, and/or characterizing material within a target material

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toth, James J.; Wall, Donald; Wittman, Richard S.

Target assemblies are provided that can include a uranium-comprising annulus. The assemblies can include target material consisting essentially of non-uranium material within the volume of the annulus. Reactors are disclosed that can include one or more discrete zones configured to receive target material. At least one uranium-comprising annulus can be within one or more of the zones. Methods for producing isotopes within target material are also disclosed, with the methods including providing neutrons to target material within a uranium-comprising annulus. Methods for modifying materials within target material are disclosed as well as are methods for characterizing material within a targetmore » material.« less

Gamma-ray spectroscopy of 131Sn81 via the (9Be, 8Be γ) reaction

NASA Astrophysics Data System (ADS)

Burcher, Sean; Bey, A.; Jones, K.; Ahn, S. H.; Ayres, A.; Schmitt, K. T.; Allmond, J.; Galindo-Urribari, A.; Radford, D. C.; Liang, J. F.; Neseraja, C. D.; Pain, S. D.; Pittman, S. T.; Smith, M. S.; Stracener, D. W.; Varner, R. L.; Bardayan, D. W.; O'Malley, P. D.; Cizewski, J. A.; Howard, M. E.; Manning, B. M.; Garcia Ruiz, R. F.; Kozub, R. L.; Matos, M.; Padilla-Rodal, E.

2016-09-01

Nuclear data in the region of the doubly-magic nucleus 132Sn82 is useful for benchmarking nuclear structure theories due to the clean single-particle nature of the nuclear wavefunction near the closed shells. At the Holifield Radioactive Ion Beam Facility (HRIBF) neutron-rich beams in the 132Sn82 region were produced via proton-induced fission of a Uranium-Carbide target. The CLARION array of HPGe detectors was coupled with the HyBall array of CsI detectors to allow for particle-gamma coincidence measurements. The gamma-ray de-excitation of the four lowest lying single-neutron states has been observed for the first time via the (9Be,8Be γ) reaction. The excitation energy of these states have been measured to higher precision than was possible with the previous charged particle measurement. This work was supported in part by the U.S. Department of Energy and the National Science Foundation.

Significance of breeding in fast nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raza, S.M.; Abidi, S.B.M.

1983-12-01

Only breeder reactors--nuclear power plants that produce more fuel than they consume--are capable in principle of extracting the maximum amount of fission energy contained in uranium ore, thus offering a practical long-term solution to uranium supply problems. Uranium would then constitute a virtually inexhaustible fuel reserve for the world's future energy needs. The ultimate argument for breeding is to conserve the energy resources available to mankind. A long-term role for nuclear power with fast reactors is proven to be economically viable, environmentally acceptable and capable of wide scale exploitation in many countries. In this paper, various suggestions pertaining to themore » fuel fabrication route, fuel cycle economics, studies of the physics of fast nuclear reactors and of engineering design simplifications are presented. Fast reactors contain no moderator and inherently require enriched fuel. In general, the main aim is to suggest an improvement in the understanding of the safety and control characteristics of fast breeder power reactors. Development work is also being devoted to new carbide and nitride fuels, which are likely to exhibit breeding characteristics superior to those of the oxides of plutonium and uranium.« less

Uranium oxidation kinetics monitored by in-situ X-ray diffraction

NASA Astrophysics Data System (ADS)

Zalkind, S.; Rafailov, G.; Halevy, I.; Livneh, T.; Rubin, A.; Maimon, H.; Schweke, D.

2017-03-01

The oxidation kinetics of U-0.1 wt%Cr at oxygen pressures of 150 Torr and the temperature range of 90-150 °C was studied by means of in-situ X-ray diffraction (XRD). A "breakaway" in the oxidation kinetics is found at ∼0.25 μm, turning from a parabolic to a linear rate law. At the initial stage of oxidation the growth plane of UO2(111) is the prominent one. As the oxide thickens, the growth rate of UO2(220) plane increases and both planes grow concurrently. The activation energies obtained for the oxide growth are Qparabolic = 17.5 kcal/mol and Qlinear = 19 kcal/mol. Enhanced oxidation around uranium carbide (UC) inclusions is clearly observed by scanning electron microscopy (SEM).

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, M.H.

1981-01-09

Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

CERAMIC FUEL ELEMENT MATERIAL FOR A NEUTRONIC REACTOR AND METHOD OF FABRICATING SAME

DOEpatents

Duckworth, W.H.

1957-12-01

This patent relates to ceramic composition, and to neutronic reactor fuel elements formed therefrom. These ceramic elements have high density and excellent strength characteristics and are formed by conventional ceramic casting and sintering at a temperature of about 2700 deg F in a nitrogen atmosphere. The composition consists of silicon carbide, silicon, uranium oxide and a very small percentage of molybdenum. Compositions containing molybdenum are markedly stronger than those lacking this ingredient.

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, Milton H.

1983-01-01

Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Acceptance Test Data for BWXT Coated Particle Batch 93164A Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-02-01

Coated particle fuel batch J52O-16-93164 was produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used as demonstration production-scale coated particle fuel for other experiments. The tristructural-isotropic (TRISO) coatings were deposited in a 150-mm-diameter production-scale fluidizedbed chemical vapor deposition (CVD) furnace onto 425-μm-nominal-diameter spherical kernels from BWXT lot J52L-16-69316. Each kernel contained a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO) and was coated with four consecutive CVD layers:more » a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (i.e., 93164A).« less

Edge on Impact Simulations and Experiments

DTIC Science & Technology

2013-09-01

silicon carbide ( SiC ) and aluminum oxynitride (AlON) ceramics are predicted using the Kayenta macroscopic constitutive model. Aspects regarding...damage propagation. 2.1. Silicon Carbide SiC is an opaque ceramic explored by the armor community. It is perhaps the most extensively characterized...the Weibull modulus for SiC . 4.1. Silicon Carbide Figures 3 and 4 compare experimental images with model predictions of EOI of SiC targets at respective

Measurement of fuel corrosion products using planar laser-induced fluorescence

NASA Astrophysics Data System (ADS)

Wantuck, Paul J.; Sappey, Andrew D.; Butt, Darryl P.

1993-01-01

Characterizing the corrosion behavior of nuclear fuel material in a high-temperature hydrogen environment is critical for ascertaining the operational performance of proposed nuclear thermal propulsion (NTP) concepts. In this paper, we describe an experimental study undertaken to develop and test non-intrusive, laser-based diagnostics for ultimately measuring the distribution of key gas-phase corrosion products expected to evolve during the exposure of NTP fuel to hydrogen. A laser ablation technique is used to produce high temperature, vapor plumes from uranium-free zirconium carbide (ZrC) and niobium carbide (NbC) forms for probing by various optical diagnostics including planar laser-induced fluorescence (PLIF). We discuss the laser ablation technique, results of plume emission measurements, and we describe both the actual and proposed planar LIF schemes for imaging constituents of the ablated ZrC and NbC plumes. Envisioned testing of the laser technique in rf-heated, high temperature gas streams is also discussed.

Carbon monoxide formation in UO2 kerneled HTR fuel particles containing oxygen getters

NASA Astrophysics Data System (ADS)

Proksch, E.; Strigl, A.; Nabielek, H.

1986-01-01

Mass spectrometric measurements of CO in irradiated UO2 fuel particles containing oxygen getters are summarized. Uranium carbide addition in the 3% to 15% range reduces the CO release by factors between 25 and 80, up to burn-up levels as high as 70% FIMA. Unintentional gettering by SiC in TRISO coated particles with failed inner pyrocarbon layers results in CO reduction factors between 15 and 110. For ZrC, ambiguous results are obtained; ZrC probably results in CO reduction by a factor of 40; Ce2O3 and La2O3 seem less effective than the carbides; for Ce2O3, reduction factors between 3 and 15 are found. However, the results are possibly incorrect due to premature oxidation of the getter already during fabrication. Addition of SiO2 + Al2O3 has no influence on CO release.

Metallic impurities-silicon carbide interaction in HTGR fuel particles

NASA Astrophysics Data System (ADS)

Minato, Kazuo; Ogawa, Toru; Kashimura, Satoru; Fukuda, Kousaku; Shimizu, Michio; Tayama, Yoshinobu; Takahashi, Ishio

1990-12-01

Corrosion of the coating layers of silicon carbide (SiC) by metallic impurities was observed in irradiated Triso-coated uranium dioxide particles for high temperature gas-cooled reactors with an optical microscope and an electron probe micro-analyzer. The SiC layers were attacked from the outside of the particles. The main element observed in the corroded areas was iron, but sometimes iron and nickel were found. These elements must have been contained as impurities in the graphite matrix in which the coated particles were dispersed. Since these elements are more stable thermodynamically in the presence of SiC than in the presence of graphite at irradiation temperatures, they were transferred to the SiC layer to form more stable silicides. During fuel manufacturing processes, intensive care should be taken to prevent the fuel from being contaminated with those elements which react with SiC.

Wet-chemical dissolution of TRISO-coated simulated high-temperature-reactor fuel particles

NASA Astrophysics Data System (ADS)

Skolo, K. P.; Jacobs, P.; Venter, J. H.; Klopper, W.; Crouse, P. L.

2012-01-01

Chemical etching with different mixtures of acidic solutions has been investigated to disintegrate the two outermost coatings from tri-structural isotropic coated particles containing zirconia kernels, which are used in simulated particles instead of uranium dioxide. A scanning electron microscope (SEM) was used to study the morphology of the particles after the first etching step as well as at different stages of the second etching step. SEM examination shows that the outer carbon layer can be readily removed with a CrO 3-HNO 3/H 2SO 4 solution. This finding was verified by energy dispersive spectroscopy (EDS) analysis. Etching of the silicon carbide layer in a hydrofluoric-nitric solution yielded partial removal of the coating and localized attack of the underlying coating layers. The SEM results provide evidence that the etching of the silicon carbide layer is strongly influenced by its microstructure.

Growth of High Purity Oxygen-Free Silicon by Cold Crucible Techniques.

DTIC Science & Technology

1982-06-01

Liquid Metals (A Review). High Temp.-High Pressures 2(6), 583-586, 1970. 1971 Knights, C.F. and Perkins, R. Levitation Melting of Uranium Mono- Carbide . J...content - typically I PPM or less. c) The crystals grown exhibited a high level of carbon contamination (2-30 PPM ) which we believe, is caused by the...grown from melts confined in the cold crucible exhibit an unusually low oxygen content - typically 1 PPM or less. c.) The crystals grown exhibited a

Vaporization and atomization of uranium in a graphite tube electrothermal vaporizer: a mechanistic study using electrothermal vaporization inductively coupled plasma mass spectrometry and graphite furnace atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Goltz, D. M.; Grégoire, D. C.; Byrne, J. P.; Chakrabarti, C. L.

1995-07-01

The mechanism of vaporization and atomization of U in a graphite tube electrothermal vaporizer was studied using graphite furnace atomic absorption spectrometry (GFAAS) and electrothermal vaporization inductively coupled plasma mass spectrometry (ETV-ICP-MS). Graphite furnace AAS studies indicate U atoms are formed at temperatures above 2400°C. Using ETV-ICP-MS, an appearance temperature of 1100°C was obtained indicating that some U vaporizes as U oxide. Although U carbides form at temperatures above 2000°C, ETV-ICP-MS studies show that they do not vaporize until 2600°C. In the temperature range between 2200°C and 2600°C, U atoms in GFAAS are likely formed by thermal dissociation of U oxide, whereas at higher temperatures, U atoms are formed via thermal dissociation of U carbide. The origin of U signal suppression in ETV-ICP-MS by NaCl was also investigated. At temperatures above 2000°C, signal suppression may be caused by the accelerated rate of formation of carbide species while at temperatures below 2000°C, the presence of NaCl may cause intercalation of the U in the graphite layers resulting in partial retention of U during the vaporization step. The use of 0.3% freon-23 (CHF 3) mixed with the argon carrier gas was effective in preventing the intercalation of U in graphite and U carbide formation at 2700°C.

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

Effect of spin-orbit and on-site Coulomb interactions on the electronic structure and lattice dynamics of uranium monocarbide

NASA Astrophysics Data System (ADS)

Wdowik, U. D.; Piekarz, P.; Legut, D.; Jagło, G.

2016-08-01

Uranium monocarbide, a potential fuel material for the generation IV reactors, is investigated within density functional theory. Its electronic, magnetic, elastic, and phonon properties are analyzed and discussed in terms of spin-orbit interaction and localized versus itinerant behavior of the 5 f electrons. The localization of the 5 f states is tuned by varying the local Coulomb repulsion interaction parameter. We demonstrate that the theoretical electronic structure, elastic constants, phonon dispersions, and their densities of states can reproduce accurately the results of x-ray photoemission and bremsstrahlung isochromat measurements as well as inelastic neutron scattering experiments only when the 5 f states experience the spin-orbit interaction and simultaneously remain partially localized. The partial localization of the 5 f electrons could be represented by a moderate value of the on-site Coulomb interaction parameter of about 2 eV. The results of the present studies indicate that both strong electron correlations and spin-orbit effects are crucial for realistic theoretical description of the ground-state properties of uranium carbide.

High temperature surface effects of He + implantation in ICF fusion first wall materials

NASA Astrophysics Data System (ADS)

Zenobia, Samuel J.; Radel, R. F.; Cipiti, B. B.; Kulcinski, Gerald L.

2009-06-01

The first wall armor of the inertial confinement fusion reactor chambers must withstand high temperatures and significant radiation damage from target debris and neutrons. The resilience of multiple materials to one component of the target debris has been investigated using energetic (20-40 keV) helium ions generated in the inertial electrostatic confinement device at the University of Wisconsin. The materials studied include: single-crystalline, and polycrystalline tungsten, tungsten-coated tantalum-carbide 'foams', tungsten-rhenium alloy, silicon carbide, carbon-carbon velvet, and tungsten-coated carbon-carbon velvet. Steady-state irradiation temperatures ranged from 750 to 1250 °C with helium fluences between 5 × 10 17 and 1 × 10 20 He +/cm 2. The crystalline, rhenium alloyed, carbide foam, and powder metallurgical tungsten specimens each experienced extensive pore formation after He + irradiation. Flaking and pore formation occurred on silicon carbide samples. Individual fibers of carbon-carbon velvet specimens sustained erosion and corrugation, in addition to the roughening and rupturing of tungsten coatings after helium ion implantation.

Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-03-01

Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide andmore » uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. Thesemore » data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These datamore » were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Porous silicon carbide and aluminum oxide with unidirectional open porosity as model target materials for radioisotope beam production

NASA Astrophysics Data System (ADS)

Czapski, M.; Stora, T.; Tardivat, C.; Deville, S.; Santos Augusto, R.; Leloup, J.; Bouville, F.; Fernandes Luis, R.

2013-12-01

New silicon carbide (SiC) and aluminum oxide (Al2O3) of a tailor-made microstructure were produced using the ice-templating technique, which permits controlled pore formation conditions within the material. These prototypes will serve to verify aging of the new advanced target materials under irradiation with proton beams. Before this, the evaluation of their mechanical integrity was made based on the energy deposition spectra produced by FLUKA codes.

Revised Point of Departure Design Options for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Fittje, James E.; Borowski, Stanley K.; Schnitzler, Bruce

2015-01-01

In an effort to further refine potential point of departure nuclear thermal rocket engine designs, four proposed engine designs representing two thrust classes and utilizing two different fuel matrix types are designed and analyzed from both a neutronics and thermodynamic cycle perspective. Two of these nuclear rocket engine designs employ a tungsten and uranium dioxide cermet (ceramic-metal) fuel with a prismatic geometry based on the ANL-200 and the GE-710, while the other two designs utilize uranium-zirconium-carbide in a graphite composite fuel and a prismatic fuel element geometry developed during the Rover/NERVA Programs. Two engines are analyzed for each fuel type, a small criticality limited design and a 111 kN (25 klbf) thrust class engine design, which has been the focus of numerous manned mission studies, including NASA's Design Reference Architecture 5.0. slightly higher T/W ratios, but they required substantially more 235U.

Nuclear fuels for very high temperature applications

NASA Astrophysics Data System (ADS)

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

The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cols, H.; Cristini, P.; Marques, R.

1997-08-01

The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing highmore » enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.« less

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

NASA Astrophysics Data System (ADS)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

Genesis Silicon Carbide Concentrator Target 60003 Preliminary Ellipsometry Mapping Results

NASA Technical Reports Server (NTRS)

Calaway, M. J.; Rodriquez, M. C.; Stansbery, E. K.

2007-01-01

The Genesis concentrator was custom designed to focus solar wind ions primarily for terrestrial isotopic analysis of O-17/O-16 and O-18/O-16 to +/-1%, N-15/N-14 to +/-1%, and secondarily to conduct elemental and isotopic analysis of Li, Be, and B. The circular 6.2 cm diameter concentrator target holder was comprised of four quadrants of highly pure semiconductor materials that included one amorphous diamond-like carbon, one C-13 diamond, and two silicon carbide (SiC). The amorphous diamond-like carbon quadrant was fractured upon impact at Utah Test and Training Range (UTTR), but the remaining three quadrants survived fully intact and all four quadrants hold an important collection of solar wind. The quadrants were removed from the target holder at NASA Johnso n Space Center Genesis Curation Laboratory in April 2005, and have been housed in stainless steel containers under continual nitrogen purge since time of disintegration. In preparation for allocation of a silicon carbide target for oxygen isotope analyses at UCLA, the two SiC targets were photographed for preliminary inspection of macro particle contamination from the hard non-nominal landing as well as characterized by spectroscopic ellipsometry to evaluate thin film contamination. This report is focused on Genesis SiC target sample number 60003.

A combined experimental and numerical approach for the control and monitoring of the SPES target during operation at high temperature

NASA Astrophysics Data System (ADS)

Ballan, Michele; Manzolaro, Mattia; Meneghetti, Giovanni; Andrighetto, Alberto; Monetti, Alberto; Bisoffi, Giovanni; Prete, Gianfranco

2016-06-01

The SPES project at INFN-LNL aims at the production of neutron-rich Radioactive Ion Beams (RIBs) using the ISOL (Isotope Separation On Line) technique. A 40 MeV 200 μA proton beam will directly impinge a uranium carbide target, generating approximately 1013 fissions per second. The target system is installed under vacuum inside a water-cooled chamber, and have to maintain high working temperatures, close to 2000 °C. During operation the proton beam provides the heating power required to keep the target at the desired temperature level. As a consequence, its characteristics have to be strictly controlled in order to avoid undesired overheating. According to the original design of the control system, the proton beam can be suddenly interrupted in case of out of range vacuum or cooling water flow levels. With the aim to improve the reliability of the control system a set of temperature sensors has been installed close to the target. Their types and installation positions were defined taking into consideration the detailed information coming from a dedicated thermal-electric model that allowed to investigate the most critical and inaccessible target hot-spots. This work is focused on the definition and experimental validation of the aforementioned numerical model. Its results were used to appropriately install two type C thermocouples, a PT100 thermo-resistance and a residual primary beam current detector. In addition the numerical model will be used for the definition of appropriate thresholds for each installed temperature sensor, since it allows to define a relationship between the locally measured values with the overall calculated temperature field. In case of over temperatures the monitoring system will send warning signals or in case interrupt the proton beam.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reiche, Helmut Matthias; Vogel, Sven C.

New in situ data for the U-C system are presented, with the goal of improving knowledge of the phase diagram to enable production of new ceramic fuels. The none quenchable, cubic, δ-phase, which in turn is fundamental to computational methods, was identified. Rich datasets of the formation synthesis of uranium carbide yield kinetics data which allow the benchmarking of modeling, thermodynamic parameters etc. The order-disorder transition (carbon sublattice melting) was observed due to equal sensitivity of neutrons to both elements. This dynamic has not been accurately described in some recent simulation-based publications.

Electron Correlation and Tranport Properties in Nuclear Fuel Materials

NASA Astrophysics Data System (ADS)

Yin, Quan; Haule, Kristjan; Kotliar, Gabriel; Savrasov, Sergey; Pickett, Warren

2011-03-01

Using first principle LDA+DMFT method, we conduct a systematic study on the correlated electronic structures and transport properties of select actinide carbides, nitrides, and oxides, many of which are nuclear fuel materials. Our results capture the metal--insulator Mott transition within the studied systems, and the appearance of the Zhang-Rice state in uranium dioxide. More importantly, by understanding the physics underlying their transport properties, we suggest ways to improve the efficiency of currently used fuels. This work is supported by the DOE Nuclear Energy University Program, contract No. 00088708.

Carbon monoxide formation in UO 2 kerneled HTR fuel particles containing oxygen getters

NASA Astrophysics Data System (ADS)

Proksch, E.; Strigl, A.; Nabielek, H.

1986-06-01

Mass spectrometric measurements of CO in irradiated UO 2 kerneled HTR fuel particles containing various oxygen getters are summarized and evaluated. Uranium carbide addition in the 3 to 15% range reduces the CO release by factors between 25 and 80, up to burn-up levels as high as 70% FIMA. Unintentional gettering by SiC in TRISO coated particles with failed inner pyrocarbon layers results in CO reduction factors between 15 and 110. For ZrC, only somewhat ambiguous results have been obtained; most likely, ZrC results in CO reduction by a factor of about 40. Ce 2O 3 and La 2O 3 seem to be somewhat less effective than the three carbides; for Ce 2O 3, reduction factors between 3 and 15 have been found. However, these results are possibly incorrect due to premature oxidation of the getter already during fabrication. Addition of SiO 2 + Al 2O 3 has no influence on CO release at all.

Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel

NASA Astrophysics Data System (ADS)

Lillo, T. M.; van Rooyen, I. J.

2016-05-01

In this study, the fission product precipitates at silicon carbide grain boundaries from an irradiated TRISO particle were identified and correlated with the associated grain boundary characteristics. Precession electron diffraction in the transmission electron microscope provided the crystallographic information needed to identify grain boundary misorientation and boundary type (i.e., low angle, random high angle or coincident site lattice (CSL)-related). The silicon carbide layer was found to be composed mainly of twin boundaries and small fractions of random high angle and low angle grain boundaries. Most fission products were found at random, high-angle grain boundaries, with small fractions at low-angle and CSL-related grain boundaries. Palladium (Pd) was found at all types of grain boundaries while Pd-uranium and Pd-silver precipitates were only associated with CSL-related and random, high-angle grain boundaries. Precipitates containing only Ag were found only at random, high-angle grain boundaries, but not at low angle or CSL-related grain boundaries.

Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

NASA Astrophysics Data System (ADS)

Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

2018-03-01

The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, Francine Joyce; Stempien, John Dennis

2016-09-01

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within amore » specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.« less

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

NASA Astrophysics Data System (ADS)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Low-enriched uranium high-density target project. Compendium report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vandegrift, George; Brown, M. Alex; Jerden, James L.

2016-09-01

At present, most 99Mo is produced in research, test, or isotope production reactors by irradiation of highly enriched uranium targets. To achieve the denser form of uranium needed for switching from high to low enriched uranium (LEU), targets in the form of a metal foil (~125-150 µm thick) are being developed. The LEU High Density Target Project successfully demonstrated several iterations of an LEU-fission-based Mo-99 technology that has the potential to provide the world’s supply of Mo-99, should major producers choose to utilize the technology. Over 50 annular high density targets have been successfully tested, and the assembly and disassemblymore » of targets have been improved and optimized. Two target front-end processes (acidic and electrochemical) have been scaled up and demonstrated to allow for the high-density target technology to mate up to the existing producer technology for target processing. In the event that a new target processing line is started, the chemical processing of the targets is greatly simplified. Extensive modeling and safety analysis has been conducted, and the target has been qualified to be inserted into the High Flux Isotope Reactor, which is considered above and beyond the requirements for the typical use of this target due to high fluence and irradiation duration.« less

The brain is a target organ after acute exposure to depleted uranium.

PubMed

Lestaevel, P; Houpert, P; Bussy, C; Dhieux, B; Gourmelon, P; Paquet, F

2005-09-01

The health effects of depleted uranium (DU) are mainly caused by its chemical toxicity. Although the kidneys are the main target organs for uranium toxicity, uranium can also reach the brain. In this paper, the central effects of acute exposure to DU were studied in relation to health parameters and the sleep-wake cycle of adult rats. Animals were injected intraperitoneally with 144+/-10 microg DU kg-1 as nitrate. Three days after injection, the amounts of uranium in the kidneys represented 2.6 microg of DU g-1 of tissue, considered as a sub-nephrotoxic dosage. The central effect of uranium could be seen through a decrease in food intake as early as the first day after exposure and shorter paradoxical sleep 3 days after acute DU exposure (-18% of controls). With a lower dosage of DU (70+/-8 microg DU kg-1), no significant effect was observed on the sleep-wake cycle. The present study intends to illustrate the fact that the brain is a target organ, as are the kidneys, after acute exposure to a moderate dosage of DU. The mechanisms by which uranium causes these early neurophysiological perturbations shall be discussed.

Phase Contrast Imaging of Damage Initiation During Ballistic Impact of Boron Carbide

NASA Astrophysics Data System (ADS)

Schuster, Brian; Tonge, Andrew; Ramos, Kyle; Rigg, Paulo; Iverson, Adam; Schuman, Adam; Lorenzo, Nicholas

2017-06-01

For several decades, flash X-ray imaging has been used to perform time-resolved investigations of the response of ceramics under ballistic impact. Traditional absorption based contrast offers little insight into the early initiation of inelastic deformation mechanisms and instead typically only shows the gross deformation and fracture behavior. In the present work, we employed phase contrast imaging (PCI) at the Dynamic Compression Sector (DCS) at the Advanced Photon Source, Argonne National Laboratory, to investigate crack initiation and propagation following the impact of copper penetrators into boron carbide targets. These experiments employed a single-stage propellant gun to launch small-scale (0.6 mm diameter by 3 mm long) pure copper impactors at velocities ranging from 0.9 to 1.9 km/s into commercially available boron carbide targets that were 8 mm on a side. At the lowest striking velocities the penetrator undergoes dwell or interface defeat and the target response is consistent with the cone crack formation at the impact site. At higher striking velocities there is a distinct transition to massive fragmentation leading to the onset of penetration.

Uranium XAFS analysis of kidney from rats exposed to uranium

PubMed Central

Kitahara, Keisuke; Numako, Chiya; Terada, Yasuko; Nitta, Kiyohumi; Homma-Takeda, Shino

2017-01-01

The kidney is the critical target of uranium exposure because uranium accumulates in the proximal tubules and causes tubular damage, but the chemical nature of uranium in kidney, such as its chemical status in the toxic target site, is poorly understood. Micro-X-ray absorption fine-structure (µXAFS) analysis was used to examine renal thin sections of rats exposed to uranyl acetate. The U L III-edge X-ray absorption near-edge structure spectra of bulk renal specimens obtained at various toxicological phases were similar to that of uranyl acetate: their edge position did not shift compared with that of uranyl acetate (17.175 keV) although the peak widths for some kidney specimens were slightly narrowed. µXAFS measurements of spots of concentrated uranium in the micro-regions of the proximal tubules showed that the edge jump slightly shifted to lower energy. The results suggest that most uranium accumulated in kidney was uranium (VI) but a portion might have been biotransformed in rats exposed to uranyl acetate. PMID:28244440

Uranium XAFS analysis of kidney from rats exposed to uranium.

PubMed

Kitahara, Keisuke; Numako, Chiya; Terada, Yasuko; Nitta, Kiyohumi; Shimada, Yoshiya; Homma-Takeda, Shino

2017-03-01

The kidney is the critical target of uranium exposure because uranium accumulates in the proximal tubules and causes tubular damage, but the chemical nature of uranium in kidney, such as its chemical status in the toxic target site, is poorly understood. Micro-X-ray absorption fine-structure (µXAFS) analysis was used to examine renal thin sections of rats exposed to uranyl acetate. The U L III -edge X-ray absorption near-edge structure spectra of bulk renal specimens obtained at various toxicological phases were similar to that of uranyl acetate: their edge position did not shift compared with that of uranyl acetate (17.175 keV) although the peak widths for some kidney specimens were slightly narrowed. µXAFS measurements of spots of concentrated uranium in the micro-regions of the proximal tubules showed that the edge jump slightly shifted to lower energy. The results suggest that most uranium accumulated in kidney was uranium (VI) but a portion might have been biotransformed in rats exposed to uranyl acetate.

Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium-Plutonium Carbide based fuel and 300 MWt Power

NASA Astrophysics Data System (ADS)

Clief Pattipawaej, Sandro; Su'ud, Zaki

2017-01-01

A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

DOE Office of Scientific and Technical Information (OSTI.GOV)

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

Reactivity of formic acid (HCOOD and DCOOH) at uranium and UO 2.0 surfaces

NASA Astrophysics Data System (ADS)

Manner, William L.; Lloyd, Jane A.; Paffett, Mark T.

1999-10-01

Interactions of DCOOH and HCOOD with uranium and UO 2.0 surfaces have been examined using surface-specific techniques of thermal desorption mass spectroscopy (TDMS), X-ray photoelectron spectroscopy (XPS), and static secondary ion mass spectroscopy (SSIMS). On the clean uranium surface, formate is the predominant product following formic acid adsorption at 100 K. A wide range of products is observed after annealing to 200 K, including formate, hydroxyl, O ads, and H ads (D ads) groups. Adsorbed formate decomposes by 300 K increasing the concentration of the remaining surface products. Surface-adsorbed carbon following TDMS measurements remains as the carbide, as indicated from XPS and SSIMS measurements. The only gaseous species created in high yields from the clean surface upon annealing are H 2, HD, and D 2. On the oxide surface (UO 2.0), adsorbed formate groups are more stable toward dissociation in comparison with the clean uranium surface. Between 100 and 300 K the predominant species on the UO 2.0 surface are surface formate and hydroxyl groups. Hydroxyl groups react between 300 and 350 K to release water from the surface. Adsorbed formate groups decompose between 400 and 500 K to release CO and H 2CO (D 2CO) groups from the oxide surface. Carbon was not detected on the oxide surface by XPS or SSIMS after annealing to 500 K, indicating that all carbon-containing species either desorb in the form of CO-containing products or migrate into the surface.

The Use of Thorium within the Nuclear Power Industry - 13472

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Keith

2013-07-01

Thorium is 3 to 4 times more abundant than uranium and is widely distributed in nature as an easily exploitable resource in many countries. Unlike natural uranium, which contains ∼0.7% fissile {sup 235}U isotope, natural thorium does not contain any fissile material and is made up of the fertile {sup 232}Th isotope only. Therefore thorium and thorium-based fuel as metal, oxide or carbide, has been utilized in combination with fissile {sup 235}U or {sup 239}Pu in nuclear research and power reactors for conversion to fissile {sup 233}U, thereby enlarging fissile material resources. During the pioneering years of nuclear energy, frommore » the mid 1950's to mid 1970's, there was considerable interest worldwide to develop thorium fuels and fuel cycles in order to supplement uranium reserves. Thorium fuels and fuel cycles are particularly relevant to countries having large thorium deposits but very limited uranium reserves for their long term nuclear power programme. The feasibility of thorium utilization in high temperature gas cooled reactors (HTGR), light water reactors (LWR), pressurized heavy water reactors (PHWRs), liquid metal cooled fast breeder reactors (LMFBR) and molten salt breeder reactors (MSBR) were demonstrated. The initial enthusiasm for thorium fuels and fuel cycles was not sustained among the developing countries later, due to new discovery of uranium deposits and their improved availability. However, in recent times, the need for proliferation-resistance, longer fuel cycles, higher burnup, and improved waste form characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material has led to renewed interest in thorium-based fuels and fuel cycles. (authors)« less

The effect of thermomechanical processing on second phase particle redistribution in U-10 wt%Mo

NASA Astrophysics Data System (ADS)

Hu, Xiaohua; Wang, Xiaowo; Joshi, Vineet V.; Lavender, Curt A.

2018-03-01

The multi-pass hot-rolling process of an annealed uranium-10 wt% molybdenum (U10Mo) coupon was studied by plane-strain compression finite element modeling. As-cast U10Mo typically contains second phase particles such as uranium carbides (UC) and silicides along the grain boundaries. The volume fraction of UC is typically large, while the other phases can be redissolved in the matrix by certain heat treatments. The UC particle distribution is important due to its influence on the recrystallization processes (particle stimulated nucleation) that occur during annealing between rolling passes. Unfavorable particle distribution and fracture after rolling can affect the grain size and also influence the fuel performance in the reactor. A statistical method, i.e., the two-point correlation function (2PCF), was used to analyze the carbide particle distribution after each rolling reduction. The hot rolling simulation results show that the alignment of UC particles along grain boundaries will rotate during rolling until it is parallel to the rolling direction, to form stringer-like distributions which are typically observed in rolled products that contain inclusions. 2PCF analysis shows that the interparticle spacing shrinks along the normal direction (ND) and increases along the rolling direction (RD). The simulated particle distribution is very similar to that measured experimentally for similar rolling reductions. The magnitudes of major peaks of 2PCF along the ND decrease after large reduction. The locations of major peaks indicate the inter-stringer distances. Many more small peaks appear for the 2PCF along the RD, and this is related to the neighboring particles within stringers, which are along the RD.

Dissolved radon and uranium in groundwater in a potential coal seam gas development region (Richmond River Catchment, Australia).

PubMed

Atkins, Marnie L; Santos, Isaac R; Perkins, Anita; Maher, Damien T

2016-04-01

The extraction of unconventional gas resources such as shale and coal seam gas (CSG) is rapidly expanding globally and often prevents the opportunity for comprehensive baseline groundwater investigations prior to drilling. Unconventional gas extraction often targets geological layers with high naturally occurring radioactive materials (NORM) and extraction practices may possibly mobilise radionuclides into regional and local drinking water resources. Here, we establish baseline groundwater radon and uranium levels in shallow aquifers overlying a potential CSG target formation in the Richmond River Catchment, Australia. A total of 91 groundwater samples from six different geological units showed highly variable radon activities (0.14-20.33 Bq/L) and uranium levels (0.001-2.77 μg/L) which were well below the Australian Drinking Water Guideline values (radon; 100 Bq/L and uranium; 17 μg/L). Therefore, from a radon and uranium perspective, the regional groundwater does not pose health risks to consumers. Uranium could not explain the distribution of radon in groundwater. Relatively high radon activities (7.88 ± 0.83 Bq/L) in the fractured Lismore Basalt aquifer coincided with very low uranium concentrations (0.04 ± 0.02 μg/L). In the Quaternary Sediments aquifers, a positive correlation between U and HCO3(-) (r(2) = 0.49, p < 0.01) implied the uranium was present as uranyl-carbonate complexes. Since NORM are often enriched in target geological formations containing unconventional gas, establishing radon and uranium concentrations in overlying aquifers comprises an important component of baseline groundwater investigations. Copyright © 2016 Elsevier Ltd. All rights reserved.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, George F.; Vissers, Donald R.; Marshall, Simon L.; Varma, Ravi

1989-01-01

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm.sup.2 of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99.

Progress on the chemical separation of fission fragments from 236Np produced by proton irradiation of natural uranium target

NASA Astrophysics Data System (ADS)

Larijani, C.; Jerome, S. M.; Lorusso, G.; Ivanov, P.; Russell, B.; Pearce, A. K.; Regan, P. H.

2017-11-01

The aim of the current work is to develop and validate a radiochemical separation scheme capable of separating both 236gNp and 236Pu from a uranium target of natural isotopic composition ( 1 g uranium) and 200 MBq of fission decay products. A target containing 1.2 g of UO2 was irradiated with a beam of 25 MeV protons with a typical beam current of 30 μA for 19 h in December 2013 at the University of Birmingham Cyclotron facility. Using literature values for the production cross-section for fusion of protons with uranium targets, we estimate that an upper limit of approximately 250 Bq of activity from the 236Np ground state was produced in this experiment. Using a radiochemical separation scheme, Np and Pu fractions were separated from the produced fission decay products, with analyses of the target-based final reaction products made using Inductively Couple Plasma Mass Spectrometry (ICP-MS) and high-resolution α particle and γ-ray spectrometry.

Penetration of tungsten-alloy rods into composite ceramic targets: Experiments and 2-D simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosenberg, Z.; Dekel, E.; Hohler, V.

1998-07-10

A series of terminal ballistics experiments, with scaled tungsten-alloy penetrators, was performed on composite targets consisting of ceramic tiles glued to thick steel backing plates. Tiles of silicon-carbide, aluminum nitride, titanium-dibroide and boron-carbide were 20-80 mm thick, and impact velocity was 1.7 km/s. 2-D numerical simulations, using the PISCES code, were performed in order to simulate these shots. It is shown that a simplified version of the Johnson-Holmquist failure model can account for the penetration depths of the rods but is not enough to capture the effect of lateral release waves on these penetrations.

Thin films deposited by femtosecond pulsed laser ablation of tungsten carbide

NASA Astrophysics Data System (ADS)

De Bonis, A.; Teghil, R.; Santagata, A.; Galasso, A.; Rau, J. V.

2012-09-01

Ultra-short Pulsed Laser Deposition has been applied to the production of thin films from a tungsten carbide target. The gaseous phase obtained by the laser ablation shows a very weak primary plume, in contrast with a very strong secondary one. The deposited films, investigated by Scanning Electron Microscopy, Atomic Force Microscopy, X-Ray Photoelectron Spectroscopy and X-Ray Diffraction, present a mixture of WC and other phases with lower carbon content. All films are amorphous, independently from the substrate temperature. The characteristics of the deposits have been explained in terms of thermal evaporation and cooling rate of molten particles ejected from the target.

Long-term ecological effects of exposure to uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanson, W.C.; Miera, F.R. Jr.

1976-03-01

The consequences of releasing natural and depleted uranium to terrestrial ecosystems during development and testing of depleted uranium munitions were investigated. At Eglin Air Force Base, Florida, soil at various distances from armor plate target butts struck by depleted uranium penetrators was sampled. The upper 5 cm of soil at the target bases contained an average of 800 ppM of depleted uranium, about 30 times as much as soil at 5- to 10-cm depth, indicating some vertical movement of depleted uranium. Samples collected beyond about 20 m from the targets showed near-background natural uranium levels, about 1.3 +- 0.3 ..mu..g/gmore » or ppM. Two explosives-testing areas at the Los Alamos Scientific Laboratory (LASL) were selected because of their use history. E-F Site soil averaged 2400 ppM of uranium in the upper 5 cm and 1600 ppM at 5-10 cm. Lower Slobovia Site soil from two subplots averaged about 2.5 and 0.6 percent of the E-F Site concentrations. Important uranium concentration differences with depth and distance from detonation points were ascribed to the different explosive tests conducted in each area. E-F Site vegetation samples contained about 320 ppM of uranium in November 1974 and about 125 ppM in June 1975. Small mammals trapped in the study areas in November contained a maximum of 210 ppM of uranium in the gastrointestinal tract contents, 24 ppM in the pelt, and 4 ppM in the remaining carcass. In June, maximum concentrations were 110, 50, and 2 ppM in similar samples and 6 ppM in lungs. These data emphasized the importance of resuspension of respirable particles in the upper few millimeters of soil as a contamination mechanism for several components of the LASL ecosystem.« less

Trace metal assay of U(3)O(8) powder by electrothermal AAS.

PubMed

Page, A G; Godbole, S V; Kulkarni, M J; Porwal, N K; Shelar, S S; Joshi, B D

1983-10-01

Methods have been developed for the direct determination of Ag, Ca, K., Li, Mg, Na, Pb, Sn and Zn in U(3)O(8) powder samples by electrothermal AAS. Nanogram and lower amounts of these elements have been determined with a relative standard deviation of 6-16% in mg amounts of sample (either alone or mixed with an equal weight of graphite). The results for NBL reference samples were in reasonable agreement with the certified values. X-Ray diffraction studies on the residues left from the graphite mixtures after the atomization cycle, confirmed the formation of uranium carbide (UC(2)).

Quasi-Static Compression and Tensile Stress-Strain Curves, Tantalum - 10% Tungsten and 300 Grade Maraging Steel

DTIC Science & Technology

1986-05-01

available high density metal. It was investigated as a practical substitute for ballistic tungsten or uranium alloys. It was also used as a core material... PPM ) H ɝ ɝ ɝ C ᝺ ណ ណ N 12 ɝ 7 0 41 អ 60 Si ម ម Ti ង V ង ង ង Cr ᝺ ង ង Fe ង ម ម Co ង ᝺ ᝺ Ni ង ង ង Cu...3M Tri-M- ite WETORDRY trademark silicon carbide paper, wet with a trichloroethane base coolant, to remove about 0.05 mm in the center. An 0.1

METHOD OF MAKING FUEL BODIES

DOEpatents

Goeddel, W.V.; Simnad, M.T.

1962-04-24

An improved method of making a fuel body containing carbon for reactors is described. Carbides of uranium and thorium having a particle size of from 100 to 500 microns are mixed with carbon having a particle size that will pass a 200 mesh screen but be retained by a 325 mesh screen, and 10 per cent by weight pitch. The mixture is heated to a temperature of about 700 to 900 deg C, at which point bonding is effected while maintaining it under mechanical pressure of over 3,000 pounds per square inch. The entire compact is heated to a uniform temperature during the process, preferably by electrical resistance of the compact itself. (AEC)

Depleted uranium instead of lead in munitions: the lesser evil.

PubMed

Jargin, Sergei V

2014-03-01

Uranium has many similarities to lead in its exposure mechanisms, metabolism and target organs. However, lead is more toxic, which is reflected in the threshold limit values. The main potential hazard associated with depleted uranium is inhalation of the aerosols created when a projectile hits an armoured target. A person can be exposed to lead in similar ways. Accidental dangerous exposures can result from contact with both substances. Encountering uranium fragments is of minor significance because of the low penetration depth of alpha particles emitted by uranium: they are unable to penetrate even the superficial keratin layer of human skin. An additional cancer risk attributable to the uranium exposure might be significant only in case of prolonged contact of the contaminant with susceptible tissues. Lead intoxication can be observed in the wounded, in workers manufacturing munitions etc; moreover, lead has been documented to have a negative impact on the intellectual function of children at very low blood concentrations. It is concluded on the basis of the literature overview that replacement of lead by depleted uranium in munitions would be environmentally beneficial or largely insignificant because both lead and uranium are present in the environment.

Processing of LEU targets for {sup 99}Mo production--testing and modification of the Cintichem process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wu, D.; Landsberger, S.; Buchholz, B.

1995-09-01

Recent experimental results on testing and modification of the Cintichem process to allow substitution of low enriched uranium (LEU) for high enriched uranium (HEU) targets are presented in this report. The main focus is on {sup 99}Mo recovery and purification by its precipitation with {alpha}-benzoin oxime. Parameters that were studied include concentrations of nitric and sulfuric acids, partial neutralization of the acids, molybdenum and uranium concentrations, and the ratio of {alpha}-benzoin oxime to molybdenum. Decontamination factors for uranium, neptunium, and various fission products were measured. Experiments with tracer levels of irradiated LEU were conducted for testing the {sup 99}Mo recoverymore » and purification during each step of the Cintichem process. Improving the process with additional processing steps was also attempted. The results indicate that the conversion of molybdenum chemical processing from HEU to LEU targets is possible.« less

Determination of Depleted Uranium in Environmental Bio-monitor Samples and Soil from Target sites in Western Balkan Region

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sahoo, Sarata K.; Enomoto, Hiroko; Tokonami, Shinji

2008-08-07

Lichen and Moss are widely used to assess the atmospheric pollution by heavy metals and radionuclides. In this paper, we report results of uranium and its isotope ratios using mass spectrometric measurements (followed by chemical separation procedure) for mosses, lichens and soil samples from a depleted uranium (DU) target site in western Balkan region. Samples were collected in 2003 from Han Pijesak (Republika Srpska in Bosnia and Hercegovina). Inductively coupled plasma mass spectrometry (ICP-MS) measurements show the presence of high concentration of uranium in some samples. Concentration of uranium in moss samples ranged from 5.2-755.43 Bq/Kg. We have determined {supmore » 235}U/{sup 238}U isotope ratio using thermal ionization mass spectrometry (TIMS) from the samples with high uranium content and the ratios are in the range of 0.002097-0.002380. TIMS measurement confirms presence of DU in some samples. However, we have not noticed any traces of DU in samples containing lesser amount of uranium or from any samples from the living environment of same area.« less

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, G.F.; Vissers, D.R.; Marshall, S.L.; Varma, R.

1987-10-26

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

Bone as a Possible Target of Chemical Toxicity of Natural Uranium in Drinking Water

PubMed Central

Kurttio, Päivi; Komulainen, Hannu; Leino, Aila; Salonen, Laina; Auvinen, Anssi; Saha, Heikki

2005-01-01

Uranium accumulates in bone, affects bone metabolism in laboratory animals, and when ingested in drinking water increases urinary excretion of calcium and phosphate, important components in the bone structure. However, little is known about bone effects of ingested natural uranium in humans. We studied 146 men and 142 women 26–83 years of age who for an average of 13 years had used drinking water originating from wells drilled in bedrock, in areas with naturally high uranium content. Biochemical indicators of bone formation were serum osteocalcin and amino-terminal propeptide of type I procollagen, and a marker for bone resorption was serum type I collagen carboxy-terminal telopeptide (CTx). The primary measure of uranium exposure was uranium concentration in drinking water, with additional information on uranium intake and uranium concentration in urine. The data were analyzed separately for men and women with robust regression (which suppresses contributions of potential influential observations) models with adjustment for age, smoking, and estrogen use. The median uranium concentration in drinking water was 27 μg/L (interquartile range, 6–116 μg/L). The median of daily uranium intake was 36 μg (7–207 μg) and of cumulative intake 0.12 g (0.02–0.66 g). There was some suggestion that elevation of CTx (p = 0.05) as well as osteocalcin (p = 0.19) could be associated with increased uranium exposure (uranium in water and intakes) in men, but no similar relationship was found in women. Accordingly, bone may be a target of chemical toxicity of uranium in humans, and more detailed evaluation of bone effects of natural uranium is warranted. PMID:15626650

The distribution of uranium and thorium in granitic rocks of the basin and range province, Western United States

USGS Publications Warehouse

McNeal, J.M.; Lee, D.E.; Millard, H.T.

1981-01-01

Some secondary uranium deposits are thought to have formed from uranium derived by the weathering of silicic igneous rocks such as granites, rhyolites, and tuffs. A regional geochemical survey was made to determine the distribution of uranium and thorium in granitic rocks of the Basin and Range province in order to evaluate the potential for secondary uranium occurrences in the area. The resulting geochemical maps of uranium, thorium, and the Th:U ratio may be useful in locating target areas for uranium exploration. The granites were sampled according to a five-level, nested, analysis-of-variance design, permitting estimates to be made of the variance due to differences between:(1) two-degree cells; (2) one-degree cells; (3) plutons; (4) samples; and (5) analyses. The cells are areas described in units of degrees of latitude and longitude. The results show that individual plutons tend to differ in uranium and thorium concentrations, but that each pluton tends to be relatively homogeneous. Only small amounts of variance occur at the two degree and the between-analyses levels. The three geochemical maps that were prepared are based on one-degree cell means. The reproducibility of the maps is U > Th ??? Th:U. These geochemical maps may be used in three methods of locating target areas for uranium exploration. The first method uses the concept that plutons containing the greatest amounts of uranium may supply the greatest amounts of uranium for the formation of secondary uranium occurrences. The second method is to examine areas with high thorium contents, because thorium and uranium are initially highly correlated but much uranium could be lost by weathering. The third method is to locate areas in which the plutons have particularly high Th:U ratios. Because uranium, but not thorium, is leached by chemical weathering, high Th:U ratios suggest a possible loss of uranium and possibly a greater potential for secondary uranium occurrences to be found in the area. ?? 1981.

Hydrologic and Temporal Influences of Evaporite Minerals on the Vertical Distribution, Storage, and Mobility of Uranium

NASA Astrophysics Data System (ADS)

Roycroft, S. J.; Noel, V.; Boye, K.; Besancon, C.; Weaver, K. L.; Johnson, R. H.; Dam, W. L.; Fendorf, S. E.; Bargar, J.

2016-12-01

Uranium contaminated groundwater in Riverton, Wyoming persists despite anticipated natural attenuation outside of a former uranium ore processing facility. The inability of natural flushing to dilute the uranium below the regulatory threshold indicates that sediments act as secondary sources likely (re)supplying uranium to groundwater. Throughout the contaminated floodplain, uranium rich-evaporites are readily abundant in the upper 2 m of sediments and are spatially coincident with the location of the plume, which suggests a likely link between evaporites and increased uranium levels. Knowledge of where and how uranium is stored within evaporite-associated sediments is required to understand processes controlling the mobility of uranium. We expect that flooding and seasonal changes in hydrologic conditions will affect U phase partitioning, and thus largely control U mobility. The primary questions we are addressing in this project are: What is the relative abundance of uranium incorporated in various mineral complexes throughout the evaporite sediments? How do the factors of depth, location, and seasonality influence the relative incorporation, mobility and speciation of uranium?We have systematically sampled from two soil columns over three dates in Riverton. The sampling dates span before and after a significant flooding event, providing insight into the flood's impact on local uranium mobility. Sequential chemical extractions are used to decipher the reactivity of uranium and approximate U operationally defined within reactants targeting carbonate, silicate, organic, and metal oxide bound or water and exchangeable phases. Extractions throughout the entirety of the sediment cores provide a high-resolution vertical profile of the distribution of uranium in various extracted phases. Throughout the profile, the majority (50-60%) of uranium is bound within carbonate-targeted extracts, a direct effect of the carbonate-rich evaporite sediments. The sum of our analyses provide a dynamic model of uranium incorporation within evaporite sediments holding implications for the fate of uranium throughout contaminated sites across the Colorado River Basin.

FY-15 Progress Report on Cleanup of irradiated SHINE Target Solutions Containing 140g-U/L Uranyl Sulfate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bennett, Megan E.; Bowers, Delbert L.; Vandegrift, George F.

2015-09-01

During FY 2012 and 2013, a process was developed to convert the SHINE Target Solution (STS) of irradiated uranyl sulfate (140 g U/L) to uranyl nitrate. This process is necessary so that the uranium solution can be processed by the UREX (Uranium Extraction) separation process, which will remove impurities from the uranium so that it can be recycled. The uranyl sulfate solution must contain <0.02 M SO 4 2- so that the uranium will be extractable into the UREXsolvent. In addition, it is desired that the barium content be below 0.0007 M, as this is the limit in the Resourcemore » Conservation and Recovery Act (RCRA).« less

Defeating Hard and Deeply Buried Targets in 2035

DTIC Science & Technology

2012-02-15

Hafnium Carbide 12.2 33 722 120 3000 Tantalum 16.4 24 680 113 3017 Tantalum Carbide 14.3 28 746 124 3880 Tungsten 19.3 21 550 92 3422 Tungsten ...lethality. Concepts for employing the weapon included “vertical delivery with the bomb detonated at or just outside portal, skip bomb with short fuse (first...or second contact), skip bomb with long fuse (penetrate door, maximize distance down adits [underground facility entrances or passages]), and

800-MeV proton irradiation of thorium and depleted uranium targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, G.J.; Brun, T.O.; Pitcher, E.J.

As part of the Los Alamos Fertile-to-Fissile-Conversion (FERFICON) program in the late 1980`s, thick targets of the fertile materials thorium and depleted uranium were bombarded by 800-MeV protons to produce the fissile materials {sup 233}U and {sup 239}Pu, respectively. The amount of {sup 233}U made was determined by measuring the {sup 233}Pa activity, and the yield of {sup 239}Pu was deduced by measuring the activity of {sup 239}Np. For the thorium target, 4 spallation products and 34 fission products were also measured. For the depleted uranium target, 3 spallation products and 16 fission products were also measured. The number ofmore » fissions in each target was deduced from fission product mass-yield curves. In actuality, axial distributions of the products were measured, and the distributions were then integrated over the target volume to obtain the total number of products for each reaction.« less

Inherently safe in situ uranium recovery

DOEpatents

Krumhansl, James L; Brady, Patrick V

2014-04-29

An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

Nevada Test and Training Range Depleted Uranium Target Disposal Environmental Assessment

DTIC Science & Technology

2005-03-01

to establish the probability and scope of such transport. Long-Term Fate of Depleted Uranium at Aberdeen and Yuma Proving Grounds Phase II: Human...1990. Long-Term Fate of Depleted Uranium at Aberdeen and Yuma Proving Grounds Final Report, Phase 1: Geochemical Transport and Modeling. Los...of Depleted Uranium at Aberdeen and Yuma Proving Grounds , Phase II: Human Health and Ecological Risk Assessments. Los Alamos National Laboratory

Dynamic Failure and Fragmentation of a Hot-Pressed Boron Carbide

NASA Astrophysics Data System (ADS)

Sano, Tomoko; Vargas-Gonzalez, Lionel; LaSalvia, Jerry; Hogan, James David

2017-12-01

This study investigates the failure and fragmentation of a hot-pressed boron carbide during high rate impact experiments. Four impact experiments are performed using a composite-backed target configuration at similar velocities, where two of the impact experiments resulted in complete target penetration and two resulted in partial penetration. This paper seeks to evaluate and understand the dynamic behavior of the ceramic that led to either the complete or partial penetration cases, focusing on: (1) surface and internal failure features of fragments using optical, scanning electron, and transmission electron microscopy, and (2) fragment size analysis using state-of-the-art particle-sizing technology that informs about the consequences of failure. Detailed characterization of the mechanical properties and the microstructure is also performed. Results indicate that transgranular fracture was the primary mode of failure in this boron carbide material, and no stress-induced amorphization features were observed. Analysis of the fragment sizes for the partial and completely penetrated experiments revealed a possible correlation between larger fragment sizes and impact performance. The results will add insight into designing improved advanced ceramics for impact protection applications.

Mechanical intermixing of components in (CoMoNi)-based systems and the formation of (CoMoNi)/WC nanocomposite layers on Ti sheets under ball collisions

NASA Astrophysics Data System (ADS)

Romankov, S.; Park, Y. C.; Shchetinin, I. V.

2017-11-01

Cobalt (Co), molybdenum (Mo), and nickel (Ni) components were simultaneously introduced onto titanium (Ti) surfaces from a composed target using ball collisions. Tungsten carbide (WC) balls were selected for processing as the source of a cemented carbide reinforcement phase. During processing, ball collisions continuously introduced components from the target and the grinding media onto the Ti surface and induced mechanical intermixing of the elements, resulting in formation of a complex nanocomposite structure onto the Ti surface. The as-fabricated microstructure consisted of uniformly dispersed WC particles embedded within an integrated metallic matrix composed of an amorphous phase with nanocrystalline grains. The phase composition of the alloyed layers, atomic reactions, and the matrix grain sizes depended on the combination of components introduced onto the Ti surface during milling. The as-fabricated layer exhibited a very high hardness compared to industrial metallic alloys and tool steel materials. This approach could be used for the manufacture of both cemented carbides and amorphous matrix composite layers.

Radiation protection considerations along a radioactive ion beam transport line

NASA Astrophysics Data System (ADS)

Sarchiapone, Lucia; Zafiropoulos, Demetre

2016-09-01

The goal of the SPES project is to produce accelerated radioactive ion beams for Physics studies at “Laboratori Nazionali di Legnaro” (INFN, Italy). This accelerator complex is scheduled to be built by 2016 for an effective operation in 2017. Radioactive species are produced in a uranium carbide target, by the interaction of 200 μA of protons at 40 MeV. All of the ionized species in the 1+ state come out of the target (ISOL method), and pass through a Wien filter for a first selection and an HMRS (high mass resolution spectrometer). Then they are transported by an electrostatic beam line toward a charge state breeder (where the 1+ to n+ multi-ionization takes place) before selection and reacceleration at the already existing superconducting linac. The work concerning dose evaluations, activation calculation, and radiation protection constraints related to the transport of the radioactive ion beam (RIB) from the target to the mass separator will be described in this paper. The FLUKA code has been used as tool for those calculations needing Monte Carlo simulations, in particular for the evaluation of the dose rate due to the presence of the radioactive beam in the selection/interaction points. The time evolution of a radionuclide inventory can be computed online with FLUKA for arbitrary irradiation profiles and decay times. The activity evolution is analytically evaluated through the implementation of the Bateman equations. Furthermore, the generation and transport of decay radiation (limited to gamma, beta- and beta+ emissions) is possible, referring to a dedicated database of decay emissions using mostly information obtained from NNDC, sometimes supplemented with other data and checked for consistency. When the use of Monte Carlo simulations was not feasible, the Bateman equations, or possible simplifications, have been used directly.

New developments of the in-source spectroscopy method at RILIS/ISOLDE

NASA Astrophysics Data System (ADS)

Marsh, B. A.; Andel, B.; Andreyev, A. N.; Antalic, S.; Atanasov, D.; Barzakh, A. E.; Bastin, B.; Borgmann, Ch.; Capponi, L.; Cocolios, T. E.; Day Goodacre, T.; Dehairs, M.; Derkx, X.; De Witte, H.; Fedorov, D. V.; Fedosseev, V. N.; Focker, G. J.; Fink, D. A.; Flanagan, K. T.; Franchoo, S.; Ghys, L.; Huyse, M.; Imai, N.; Kalaninova, Z.; Köster, U.; Kreim, S.; Kesteloot, N.; Kudryavtsev, Yu.; Lane, J.; Lecesne, N.; Liberati, V.; Lunney, D.; Lynch, K. M.; Manea, V.; Molkanov, P. L.; Nicol, T.; Pauwels, D.; Popescu, L.; Radulov, D.; Rapisarda, E.; Rosenbusch, M.; Rossel, R. E.; Rothe, S.; Schweikhard, L.; Seliverstov, M. D.; Sels, S.; Sjödin, A. M.; Truesdale, V.; Van Beveren, C.; Van Duppen, P.; Wendt, K.; Wienholtz, F.; Wolf, R. N.; Zemlyanoy, S. G.

2013-12-01

At the CERN ISOLDE facility, long isotope chains of many elements are produced by proton-induced reactions in target materials such as uranium carbide. The Resonance Ionization Laser Ion Source (RILIS) is an efficient and selective means of ionizing the reaction products to produce an ion beam of a chosen isotope. Coupling the RILIS with modern ion detection techniques enables highly sensitive studies of nuclear properties (spins, electromagnetic moments and charge radii) along an isotope chain, provided that the isotope shifts and hyperfine structure splitting of the atomic transitions can be resolved. At ISOLDE the campaign to measure the systematics of isotopes in the lead region (Pb, Bi, Tl and Po) has been extended to include the gold and astatine isotope chains. Several developments were specifically required for the feasibility of the most recent measurements: new ionization schemes (Po, At); a remote controlled narrow line-width mode of operation for the RILIS Ti:sapphire laser (At, Au, Po); isobar free ionization using the Laser Ion Source Trap, LIST (Po); isobar selective particle identification using the multi-reflection time-of-flight mass separator (MR-ToF MS) of ISOLTRAP (Au, At). These are summarized as part of an overview of the current status of the in-source resonance ionization spectroscopy setup at ISOLDE.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reilly, Sean Douglas; Smith, Paul Herrick; Jarvinen, Gordon D.

Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that themore » following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U 3O 8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl 3, and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially-available phosphate buffer would significantly reduce the solubility of PuCl 3 by the precipitation of PuPO 4.« less

Feasibility study on the use of uranium in photoneutron target and BSA optimization for Linac based BNCT

NASA Astrophysics Data System (ADS)

Rahmani, Faezeh; Shahriari, Majid; Minoochehr, Abdolhamid; Nedaie, Hasan

2011-06-01

A hybrid photoneutron target including natural uranium has been studied for a 20 MeV linear electron accelerator (Linac) based Boron Neutron Capture Therapy (BNCT) facility. In this study the possibility of using uranium to increase the neutron intensity has been investigated by focusing on the time dependence behavior of the build-up and decay of the delayed gamma rays from fission fragments and activation products through photo-fission reactions in the BSA (Beam Shaping Assembly) configuration design. Delayed components of neutrons and photons were calculated. The obtained BSA parameters are in agreement with the IAEA recommendation and compared to the hybrid photoneutron target without U. The epithermal flux in the suggested design is 2.67E9 (n/cm 2s/mA).

Acceptance Test Data for BWXT Coated Particle Batches 93172B and 93173B—Defective IPyC and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Helmreich, Grant W.; Dyer, John A.

Coated particle batches J52O-16-93172B and J52O-16-93173B were produced by Babcock and Wilcox Technologies (BWXT) as part of the production campaign for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), but were not used in the final fuel composite. However, these batches may be used as demonstration production-scale coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture ofmore » 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93172A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017b]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93172B).« less

The Toxicity of Depleted Uranium

PubMed Central

Briner, Wayne

2010-01-01

Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a clear and defined set of symptoms. Chronic low-dose, or subacute, exposure to depleted uranium alters the appearance of milestones in developing organisms. Adult animals that were exposed to depleted uranium during development display persistent alterations in behavior, even after cessation of depleted uranium exposure. Adult animals exposed to depleted uranium demonstrate altered behaviors and a variety of alterations to brain chemistry. Despite its reduced level of radioactivity evidence continues to accumulate that depleted uranium, if ingested, may pose a radiologic hazard. The current state of knowledge concerning DU is discussed. PMID:20195447

Progress in Chile in the development of the fission {sup 99}Mo production using modified CINTICHEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schrader, R.; Klein, J.; Medel, J.

2008-07-15

Fission {sup 99}Mo will be produced in Chile irradiating low-enriched uranium (LEU) foil in a MTR research reactor. For the purpose of developing the capability to fabricate the target, which is done of uranium foil enclosed in swaged concentric aluminum tubes, dummy targets are being fabricated using 130 {mu}m copper foil instead of the uranium foil, wrapped in a 14{mu}m nickel fission-recoil barrier. Dummy targets using several dimensions of copper foil have been assembled; however, the emphasis is being set in targets fabricated using the dimensions of the LEU foil that KAERI will provide, i.e. 50 mm x 100mm xmore » 0.130 mm. The assembling of target using the last dimensions has not been free of difficulties. Neutronic calculations and preliminary thermal and fluid analyses were performed to estimate the fission products activity and the heat removal capability for a 13 grams LEU-foil annular target, which will be irradiated in the RECH-1 research reactor at the level power of 5 MW during 48 hours. In a fume hood, Cintichem processing of natural uranium shavings with the addition of different carriers were performed, obtaining recovery over 90% of the added Mo carrier. Expertise has been gained in (a) foil dissolution process in a dissolver locally designed, (b) in Mo precipitation process, and (c) preparation of the purification columns with AgC, C and HZrO. Additionally, the irradiated target cutting machine with an innovative design was finally assembled. (author)« less

Environmental assessment of remedial action at the Slick Rock uranium mill tailings sites, Slick Rock, Colorado. Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1994-09-01

The Uranium Mill Tailings Radiation Control Act of 1978 (42 USC {section}7901 et seq.), hereafter referred to as the UMTRCA, authorized the US Department of Energy (DOE) to clean up two uranium mill tailings processing sites near Slick Rock, Colorado, in San Miquel County. Contaminated materials cover an estimated 63 acres of the Union Carbide (UC) processing site and 15 ac of the North Continent (NC) processing site. The sites are within 1 mile of each other and are adjacent to the Dolores River. The sites contain concrete foundations of mill buildings, tailings piles, and areas contaminated by windblown andmore » waterborne radioactive tailings materials. The total estimated volume of contaminated materials is approximately 621,300 cubic yards (yd{sup 3}). In addition to the contamination in the two processing site areas, four VPs were found to contain contamination. As a result of the tailings being exposed to the environment, contamination associated with the UC and NC sites has leached into shallow ground water. Surface water has not been affected. The closest residence is approximately 0.3 air mi from either site. The proposed action is to remediate the UC and NC sites by removing all contaminated materials within the designing site boundaries or otherwise associated with the sites, and relocating them to, and stabilizing them at, a location approximately 5 road mi northeast of the sites on land administered by the Bureau of Land Management (BLM).« less

The Proliferation Security Initiative: A Means to an End for the Operational Commander

DTIC Science & Technology

2009-05-04

The Reduced Enrichment for Research and Test Reactors ( RERTR ) Program develops technology necessary to enable the conversion of civilian...facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets. The RERTR Program was initiated by the U.S. Department of...processes have been developed for producing radioisotopes with LEU targets. The RERTR Program is managed by the Office of Nuclear Material Threat

Uranium and the Central Nervous System: What Should We Learn from Recent New Tools and Findings?

PubMed

Dinocourt, Céline

2017-01-01

Increasing industrial and military use of uranium has led to environmental pollution, which may result in uranium reaching the brain and causing cerebral dysfunction. A recent literature review has discussed data published over the last 10 years on uranium and its effects on brain function (Dinocourt C, Legrand M, Dublineau I, et al., Toxicology 337:58-71, 2015). New models of uranium exposure during neonatal brain development and the emergence of new technologies (omics and epigenetics) are of value in identifying new specific targets of uranium. Here we review the latest studies of neurogenesis, epigenetics, and metabolic dysfunctions and the identification of new biomarkers used to establish potential pathophysiological states of neurodevelopmental and neurodegenerative diseases.

Fertile-to-fissile and fission measurements for depleted uranium and thorium bombarded by 800-MeV protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilmore, J.S.; Russell, G.J.; Robinson, H.

Axial distributions of fissions and of fertile-to-fissile conversions in thick depleted uranium and thorium targets bombarded by 800-MeV protons have been measured. The amounts of /sup 239/Pu and /sup 233/U produced were determined by measuring the yields of /sup 239/Np and /sup 233/Pa, respectively. The number of fissions was deduced from fission product mass-yield curves. Integration of the axial distributions gave the total number of conversions and fissions occurring in the targets. For the uranium target, experimental results were 5.90 +- 0.25 fissions and 3.81 +- 0.01 atoms of /sup 239/Pu produced per incident portion. Corresponding calculated results were 6.14more » +- 0.04 and 3.88 +- 0.03. In the thorium target, 1.56 +- 0.25 fissions and 1.25 +- 0.01 atoms of /sup 233/U per incident proton were measured; the calculated values were 1.54 +- 0.01 fissions and 1.27 +- 0.01 atom/proton.« less

An experimental study of the reflection from spherical and flat ended cylindrical targets suitable for fetal Doppler performance assessment.

PubMed

Preston, R C; Bond, A D

1997-01-01

The performance of small-diameter targets suitable for use as oscillating targets for testing the sensitivity of Doppler fetal heartbeat detectors has been systematically studied. Experimental results are presented in the 1.6-3.0 MHz frequency range for the plane-wave reflection loss for a total of 16 targets: spherical balls made of stainless steel; hemispherical-ended rods made of PTFE; and flat-ended rods made of stainless steel, PTFE, polycarbonate, and tungsten carbide. Results show that the fine-structure variation of reflection loss with frequency is greatest in the case of spherical ball targets and least for flat-ended targets. It has been shown that, providing care is taken during manufacture, the reflection loss from a flat-ended target can be predicted using a simple theory based on a plane disc reflector. Tungsten carbide targets consisting of a long rod with a diameter of 1.6 mm tapered down to a cylindrical flat end with a diameter of 0.4, 0.5, and 0.6 mm have been shown to provide reflection losses of between 60 and 40 dB, and to have a smooth variation of reflection loss with frequency. They can also be manufactured in a form that allows no significant interference from the supporting structure and, therefore, are ideal targets to meet the requirements of International Electrotechnical Commission 1266:1995.

First principles study of the electronic and magnetic structures and bonding properties of UCoC2 ternary, characteristic of C-C units

NASA Astrophysics Data System (ADS)

Matar, Samir F.

2013-03-01

The electronic structure of UCoC2, a di-carbide with the C-C units is examined from ab initio with an assessment of the properties of chemical bonding. The energy-volume equation of state shows large anisotropy effects due to C-C alignment along tetragonal c-axis leading to high linear incompressibility. Relevant features of selective bonding of uranium and cobalt with carbon at two different Wyckoff sites and strong C-C interactions are remarkable. The vibrational frequencies for C⋯C stretching modes indicate closer behavior to aliphatic C-C rather than Cdbnd C double bond. A ferromagnetic ground state is proposed from the calculations.

Experimental study of UC polycrystals in the prospect of improving the as-fabricated sample purity

NASA Astrophysics Data System (ADS)

Raveu, Gaëlle; Martin, Guillaume; Fiquet, Olivier; Garcia, Philippe; Carlot, Gaëlle; Palancher, Hervé; Bonnin, Anne; Khodja, Hicham; Raepsaet, Caroline; Sauvage, Thierry; Barthe, Marie-France

2014-12-01

Uranium and plutonium carbides are candidate fuels for Generation IV nuclear reactors. This study is focused on the characterization of uranium monocarbide samples. The successive fabrication steps were carried out under atmospheres containing low oxygen and moisture concentrations (typically less than 100 ppm) but sample transfers occurred in air. Six samples were sliced from four pellets elaborated by carbothermic reaction under vacuum. Little presence of UC2 is expected in these samples. The α-UC2 phase was indeed detected within one of these UC samples during an XRD experiment performed with synchrotron radiation. Moreover, oxygen content at the surface of these samples was depth profiled using a recently developed nuclear reaction analysis method. Large oxygen concentrations were measured in the first micron below the sample surface and particularly in the first 100-150 nm. UC2 inclusions were found to be more oxidized than the surrounding matrix. This work points out to the fact that more care must be given at each step of UC fabrication since the material readily reacts with oxygen and moisture. A new glovebox facility using a highly purified atmosphere is currently being built in order to obtain single phase UC samples of better purity.

PLD deposition of tungsten carbide contact for diamond photodiodes. Influence of process conditions on electronic and chemical aspects

NASA Astrophysics Data System (ADS)

Cappelli, E.; Bellucci, A.; Orlando, S.; Trucchi, D. M.; Mezzi, A.; Valentini, V.

2013-08-01

Tungsten carbide, WC, contacts behave as very reliable Schottky contacts for opto-electronic diamond devices. Diamond is characterized by superior properties in high-power, high frequency and high-temperature applications, provided that thermally stable electrode contacts will be realized. Ohmic contacts can be easily achieved by using carbide-forming metals, while is difficult to get stable Schottky contacts at elevated temperatures, due to the interface reaction and/or inter-diffusion between metals and diamond. Novel type of contacts, made of tungsten carbide, WC, seem to be the best solution, for their excellent thermal stability, high melting point, oxidation and radiation resistance and good electrical conductivity. Our research was aimed at using pulsed laser deposition for WC thin film deposition, optimizing experimental parameters, to obtain a final device characterized by excellent electronic properties, as a detector for radiation in deep UV or as X-ray dosimeter. We deposited our films by laser ablation from a target of pure WC, using different reaction conditions (i.e., substrate heating, vacuum or reactive atmosphere (CH4/Ar), RF plasma activated), to optimize both the stoichiometry of the film and its structure. Trying to obtain a material with the best electronic response, we used also two sources of laser radiation for target ablation, i.e., nano-second pulsed excimer laser ArF, and ultra-short fs Ti:Sapphire laser. The structure and chemical aspects have been evaluated by Raman and X-ray photoelectron spectroscopy (XPS), while the dosimeter photodiode response has been tested by the I-V measurements, under soft X-ray irradiation.

Enriched but not depleted uranium affects central nervous system in long-term exposed rat.

PubMed

Houpert, Pascale; Lestaevel, Philippe; Bussy, Cyrill; Paquet, François; Gourmelon, Patrick

2005-12-01

Uranium is well known to induce chemical toxicity in kidneys, but several other target organs, such as central nervous system, could be also affected. Thus in the present study, the effects on sleep-wake cycle and behavior were studied after chronic oral exposure to enriched or depleted uranium. Rats exposed to 4% enriched uranium for 1.5 months through drinking water, accumulated twice as much uranium in some key areas such as the hippocampus, hypothalamus and adrenals than did control rats. This accumulation was correlated with an increase of about 38% of the amount of paradoxical sleep, a reduction of their spatial working memory capacities and an increase in their anxiety. Exposure to depleted uranium for 1.5 months did not induce these effects, suggesting that the radiological activity induces the primary events of these effects of uranium.

Separation and Depleted Uranium Fragments from Gun Test Catchment. Volume 2. Catchment System and Separations Methods

DTIC Science & Technology

1993-12-30

projectile fragments from target materials, principally sand. Phase I activities included (1) literature review of separations technology , (2) site visits, (3...the current operation, evaluation of alternative means for separation of DU from sand, a review of uranium mining technology for v possible...the current operation, evaluation of alternative means for separation of DU from sand, a review of uranium mining technology for possible

ANL progress on the cooperation with CNEA for the Mo-99 production : base-side digestion process.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gelis, A. V.; Quigley, K. J.; Aase, S. B.

2004-01-01

Conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) targets for the Mo-99 production requires certain modifications of the target design, the digestion and the purification processes. ANL is assisting the Argentine Comision Nacional de Energia Atomica (CNEA) to overcome all the concerns caused by the conversion to LEU foil targets. A new digester with stirring system has been successfully applied for the digestion of the low burn-up U foil targets in KMnO4 alkaline media. In this paper, we report the progress on the development of the digestion procedure with stirring focusing on the minimization of the liquid radioactive waste.

Scalable and Tunable Carbide-Phosphide Composite Catalyst System for the Thermochemical Conversion of Biomass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Regmi, Yagya; Rogers, Bridget; Labbe, Nicole

We have prepared composite materials of hexagonal nickel phosphide and molybdenum carbide (Mo2C) utilizing a simple and scalable two-stage synthesis method comprised of carbothermic reduction followed by hydrothermal incubation. We observe the monophasic hexagonal phosphide Ni2P in the composite at low phosphide-to-carbide (P:C) ratios. Upon increasing the proportion of P:C, the carbide surface becomes saturated, and we detect the emergence of a second hexagonal nickel phosphide phase (Ni5P4) upon annealing. We demonstrate that vapor-phase upgrading (VPU) of whole biomass via catalytic fast pyrolysis is achievable using the composite material as a catalyst, and we monitor the resulting product slates usingmore » pyrolysis gas chromatography/mass spectrometry. Our analysis of the product vapors indicates that variation of the P:C molar ratio in the composite material affords product slates of varying complexity and composition, which is indicated by the number of products and their relative proportions in the product slate. Our results demonstrate that targeted vapor product composition can be obtained, which can potentially be utilized to tune the composition of the bio-oil downstream.« less

Scalable and Tunable Carbide-Phosphide Composite Catalyst System for the Thermochemical Conversion of Biomass

DOE PAGES

Regmi, Yagya; Rogers, Bridget; Labbe, Nicole; ...

2017-07-13

We have prepared composite materials of hexagonal nickel phosphide and molybdenum carbide (Mo2C) utilizing a simple and scalable two-stage synthesis method comprised of carbothermic reduction followed by hydrothermal incubation. We observe the monophasic hexagonal phosphide Ni2P in the composite at low phosphide-to-carbide (P:C) ratios. Upon increasing the proportion of P:C, the carbide surface becomes saturated, and we detect the emergence of a second hexagonal nickel phosphide phase (Ni5P4) upon annealing. We demonstrate that vapor-phase upgrading (VPU) of whole biomass via catalytic fast pyrolysis is achievable using the composite material as a catalyst, and we monitor the resulting product slates usingmore » pyrolysis gas chromatography/mass spectrometry. Our analysis of the product vapors indicates that variation of the P:C molar ratio in the composite material affords product slates of varying complexity and composition, which is indicated by the number of products and their relative proportions in the product slate. Our results demonstrate that targeted vapor product composition can be obtained, which can potentially be utilized to tune the composition of the bio-oil downstream.« less

Sputtering of uranium

NASA Technical Reports Server (NTRS)

Gregg, R.; Tombrello, T. A.

1978-01-01

Results are presented for an experimental study of the sputtering of U-235 atoms from foil targets by hydrogen, helium, and argon ions, which was performed by observing tracks produced in mica by fission fragments following thermal-neutron-induced fission. The technique used allowed measurements of uranium sputtering yields of less than 0.0001 atom/ion as well as yields involving the removal of less than 0.01 monolayer of the uranium target surface. The results reported include measurements of the sputtering yields for 40-120-keV protons, 40-120-keV He-4(+) ions, and 40- and 80-keV Ar-40(+) ions, the mass distribution of chunks emitted during sputtering by the protons and 80-keV Ar-40(+) ions, the total chunk yield during He-4(+) sputtering, and some limited data on molecular sputtering by H2(+) and H3(+). The angular distribution of the sputtered uranium is discussed, and the yields obtained are compared with the predictions of collision cascade theory.

Highly Enriched Uranium Metal Cylinders Surrounded by Various Reflector Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernard Jones; J. Blair Briggs; Leland Monteirth

A series of experiments was performed at Los Alamos Scientific Laboratory in 1958 to determine critical masses of cylinders of Oralloy (Oy) reflected by a number of materials. The experiments were all performed on the Comet Universal Critical Assembly Machine, and consisted of discs of highly enriched uranium (93.3 wt.% 235U) reflected by half-inch and one-inch-thick cylindrical shells of various reflector materials. The experiments were performed by members of Group N-2, particularly K. W. Gallup, G. E. Hansen, H. C. Paxton, and R. H. White. This experiment was intended to ascertain critical masses for criticality safety purposes, as well asmore » to compare neutron transport cross sections to those obtained from danger coefficient measurements with the Topsy Oralloy-Tuballoy reflected and Godiva unreflected critical assemblies. The reflector materials examined in this series of experiments are as follows: magnesium, titanium, aluminum, graphite, mild steel, nickel, copper, cobalt, molybdenum, natural uranium, tungsten, beryllium, aluminum oxide, molybdenum carbide, and polythene (polyethylene). Also included are two special configurations of composite beryllium and iron reflectors. Analyses were performed in which uncertainty associated with six different parameters was evaluated; namely, extrapolation to the uranium critical mass, uranium density, 235U enrichment, reflector density, reflector thickness, and reflector impurities. In addition to the idealizations made by the experimenters (removal of the platen and diaphragm), two simplifications were also made to the benchmark models that resulted in a small bias and additional uncertainty. First of all, since impurities in core and reflector materials are only estimated, they are not included in the benchmark models. Secondly, the room, support structure, and other possible surrounding equipment were not included in the model. Bias values that result from these two simplifications were determined and associated uncertainty in the bias values were included in the overall uncertainty in benchmark keff values. Bias values were very small, ranging from 0.0004 ?k low to 0.0007 ?k low. Overall uncertainties range from ? 0.0018 to ? 0.0030. Major contributors to the overall uncertainty include uncertainty in the extrapolation to the uranium critical mass and the uranium density. Results are summarized in Figure 1. Figure 1. Experimental, Benchmark-Model, and MCNP/KENO Calculated Results The 32 configurations described and evaluated under ICSBEP Identifier HEU-MET-FAST-084 are judged to be acceptable for use as criticality safety benchmark experiments and should be valuable integral benchmarks for nuclear data testing of the various reflector materials. Details of the benchmark models, uncertainty analyses, and final results are given in this paper.« less

First principles investigations of the electronic structure and chemical bonding of U3Si2C2 - A uranium silicide-carbide with the rare [SiC] unit

NASA Astrophysics Data System (ADS)

Matar, S. F.; Pöttgen, R.

2012-10-01

The electronic structure of U3Si2C2, with the rare [SiC] unit is examined from ab initio with an assessment of the properties of chemical bonding. We show that plain GGA fails describing the experimental lattice parameters and the electronic structure. A better agreement with experiment (crystal determination and magnetic properties) is obtained with the GGA + U method and U = 4 eV. The energy-volume equation of state and the set of elastic constants are obtained showing incompressibility along the c-axis with U-C-Si alignment and a brittle material. Bonding of U1 and U2 selectively with Si and C and Si-C bonds are remarkable

The effect of thermomechanical processing on second phase particle redistribution in U-10 wt%Mo

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Xiaohua; Wang, Xiaowo; Joshi, Vineet V.

2018-03-01

The multi-pass hot-rolling process of an annealed uranium-10 wt% molybdenum coupon was studied by plane-strain compression finite element modeling. Two point correlation function (2PCF) was used to analyze the carbide particle distribution after each rolling reduction. The hot rolling simulation results show that the alignment of UC particles along grain boundaries will rotate during rolling until it is parallel to the rolling direction, to form stringer-like distributions which are typically observed in rolled products that contain inclusions. 2PCF analysis of simulation shows that the interparticle spacing shrinks along the normal direction. The number of major peaks of 2PCF along NDmore » decreases after large reduction. The locations of major peaks indicate the inter-stringer distances.« less

Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna

This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

DOE PAGES

Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; ...

2015-10-26

This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Validation of Monte Carlo simulation of neutron production in a spallation experiment

DOE PAGES

Zavorka, L.; Adam, J.; Artiushenko, M.; ...

2015-02-25

A renewed interest in experimental research on Accelerator-Driven Systems (ADS) has been initiated by the global attempt to produce energy from thorium as a safe(r), clean(er) and (more) proliferation-resistant alternative to the uranium-fuelled thermal nuclear reactors. The ADS research has been actively pursued at the Joint Institute for Nuclear Research (JINR), Dubna, since decades. Most recently, the emission of fast neutrons was experimentally investigated at the massive (m = 512 kg) natural uranium spallation target QUINTA. The target has been irradiated with the relativistic deuteron beams of energy from 0.5 AGeV up to 4 AGeV at the JINR Nuclotron acceleratormore » in numerous experiments since 2011. Neutron production inside the target was studied through the gamma-ray spectrometry measurement of natural uranium activation detectors. Experimental reaction rates for (n,γ), (n,f) and (n,2n) reactions in uranium have provided valuable information about the neutron distribution over a wide range of energies up to some GeV. The experimental data were compared to the predictions of Monte Carlo simulations using the MCNPX 2.7.0 code. In conclusion, the results are presented and potential sources of partial disagreement are discussed later in this work.« less

Isolation and characterization of a uranium(VI)-nitride triple bond

NASA Astrophysics Data System (ADS)

King, David M.; Tuna, Floriana; McInnes, Eric J. L.; McMaster, Jonathan; Lewis, William; Blake, Alexander J.; Liddle, Stephen T.

2013-06-01

The nature and extent of covalency in uranium bonding is still unclear compared with that of transition metals, and there is great interest in studying uranium-ligand multiple bonds. Although U=O and U=NR double bonds (where R is an alkyl group) are well-known analogues to transition-metal oxo and imido complexes, the uranium(VI)-nitride triple bond has long remained a synthetic target in actinide chemistry. Here, we report the preparation of a uranium(VI)-nitride triple bond. We highlight the importance of (1) ancillary ligand design, (2) employing mild redox reactions instead of harsh photochemical methods that decompose transiently formed uranium(VI) nitrides, (3) an electrostatically stabilizing sodium ion during nitride installation, (4) selecting the right sodium sequestering reagent, (5) inner versus outer sphere oxidation and (6) stability with respect to the uranium oxidation state. Computational analyses suggest covalent contributions to U≡N triple bonds that are surprisingly comparable to those of their group 6 transition-metal nitride counterparts.

Calculating Capstone Depleted Uranium Aerosol Concentrations from Beta Activity Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szrom, Fran; Falo, Gerald A.; Parkhurst, MaryAnn

2009-03-01

Beta activity measurements were used as surrogate measurements of uranium mass in aerosol samples collected during the field testing phase of the Capstone Depleted Uranium (DU) Aerosol Study. These aerosol samples generated by the perforation of armored combat vehicles were used to characterize the depleted uranium (DU) source term for the subsequent human health risk assessment (HHRA) of Capstone aerosols. Establishing a calibration curve between beta activity measurements and uranium mass measurements is straightforward if the uranium isotopes are in equilibrium with their immediate short-lived, beta-emitting progeny. For DU samples collected during the Capstone study, it was determined that themore » equilibrium between the uranium isotopes and their immediate short lived, beta-emitting progeny had been disrupted when penetrators had perforated target vehicles. Adjustments were made to account for the disrupted equilibrium and for wall losses in the aerosol samplers. Correction factors for the disrupted equilibrium ranged from 0.16 to 1, and the wall loss correction factors ranged from 1 to 1.92.« less

Toward a comprehensive UV laser ablation modeling of multicomponent materials—A non-equilibrium investigation on titanium carbide

NASA Astrophysics Data System (ADS)

Ait Oumeziane, Amina; Parisse, Jean-Denis

2018-05-01

Titanium carbide (TiC) coatings of great quality can be produced using nanosecond pulsed laser deposition (PLD). Because the deposition rate and the transfer of the target stoichiometry depend strongly on the laser-target/laser-plasma interaction as well as the composition of the laser induced plume, investigating the ruling fundamental mechanisms behind the material ablation and the plasma evolution in the background environment under PLD conditions is essential. This work, which extends previous investigations dedicated to the study of nanosecond laser ablation of pure target materials, is a first step toward a comprehensive non-equilibrium model of multicomponent ones. A laser-material interaction model coupled to a laser-plasma interaction one is presented. A UV 20 ns KrF (248 nm) laser pulse is considered. Ablation depths, plasma ignition thresholds, and shielding rates have been calculated for a wide range of laser beam fluences. A comparison of TiC behavior with pure titanium material under the same conditions is made. Plasma characteristics such as temperature and composition have been investigated. An overall correlation between the various results is presented.

Background and Source Term Identification in Active Neutron Interrogation Methods

DTIC Science & Technology

2011-03-24

interactions occurred to observe gamma ray peaks and not unduly increase simulation time. Not knowing the uranium enrichment modeled by Gozani, pure U...neutron interactions can occur. The uranium targets, though, should have increased neutron fluencies as the energy levels become below 2 MeV. This is...Assessment Monitor Site (TEAMS) at Kirtland AFB, NM. Iron (Fe-56), lead (Pb-207), polyethylene (C2H4 –– > C-12 & H-1), and uranium (U-235 and U-238) were

Production and Evaluation of 236gNp and Reference Materials for Naturally Occurring Radioactive Materials

NASA Astrophysics Data System (ADS)

Larijani, Cyrus Kouroush

This thesis is based on the development of a radiochemical separation scheme capable of separating both 236gNp and 236Pu from a uranium target of natural isotopic composition ( 1 g uranium) and 200 MBq of fission decay products. The isobaric distribution of fission residues produced following the bombardment of a natural uranium target with a beam of 25 MeV protons has been evaluated. Decay analysis of thirteen isobarically distinct fission residues were carried out using high-resolution gamma-ray spectrometry at the UK National Physical Laboratory. Stoichiometric abundances were calculated via the determination of absolute activity concentrations associated with the longest-lived members of each isobaric chain. This technique was validated by computational modelling of likely sequential decay processes through an isobaric decay chain. The results were largely in agreement with previously published values for neutron bombardments on natural uranium at energies of 14 MeV. Higher relative yields of products with mass numbers A 110-130 were found, consistent with the increasing yield of these radionuclides as the bombarding energy is increased. Using literature values for the production cross-section for fusion of protons with uranium targets, it is estimated that an upper limit of approximately 250 Bq of activity from the 236Np ground state was produced in this experiment. Using a radiochemical separation scheme, Np and Pu fractions were separated from the produced fission decay products, with analyses of the target-based final reaction products made using Inductively Couple Plasma Mass Spectrometry (ICP-MS) and high-resolution alpha and gamma-ray spectrometry. In a separate research theme, reliable measurement of Naturally Occurring Radioactive Materials is of significance in order to comply with environmental regulations and for radiological protection purposes. The thesis describes the standardisation of three reference materials, namely Sand, Tuff and TiO2 which can serve as quality control materials to achieve traceability, method validation and instrument calibration. The sample preparation, material characterization via gamma, alpha and Inductively Coupled Plasma Mass Spectrometry (ICP-MS) and the assignment of values for both the 4n Thorium and 4n + 2 Uranium decay series are presented.

Interactions between Impacting Particles and Target in Two-Phase Flow

NASA Astrophysics Data System (ADS)

Kang, Sang-Wook; Chow, Tze-Show

1996-11-01

The time-dependent interaction phenomena between a target and the incident solid particles borne by supersonic gas-jet stream have been numerically analyzed. In particular, the analysis dealt with particles such as aluminum, copper, and uranium ipinging on aluminum, copper, or uranium targets at various impact velocities ranging from 200 m/s to 1,000 m/s. Typical particle sizes were 50 to 100 micrometers. Results show considerable deformation of both the incident particles and the target when the velocity is greater than 500 m/s. Experiments performed on copper particles impacting an aluminum target demonstrate that under certain conditions (such as a supersonic gas jet issuing from a nozzle carrying solid particles) the impacts not only deform but also cause deposition of the particles on the surface. The present analysis shows the plausibility of such behavior when the particles impact the target at high velocities.

Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.; Demkowicz, Paul A.; Baldwin, Charles A.

2016-11-01

The PARFUME (PARticle FUel ModEl) code was used to predict silver release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO2) kernels. Post-irradiation examination (PIE) measurements were performedmore » to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO2 compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude.« less

Development of Nanomaterials for Nuclear Energetics

NASA Astrophysics Data System (ADS)

Petrunin, V. F.

Structure and properties peculiarities of the nanocrystalline powders give the opportunity to design new and to develop a modernization of nuclear energy industry materials. It was shown experimentally, that addition of 5-10% uranium dioxide nanocrystalline powder to traditional coarse powder allows to decrease the sintering temperature or to increase the fuel tablets size of grain. Similar perspectives for the technology of neutron absorbing tablets of control-rod modernization are shown by nanopowder of dysprosium hafnate changing instead now using boron carbide. It is powders in nanocrystalline state get an opportunity to sinter them and to receive compact tablet with 8,2-8,4 g/cm2 density for automatic defence system of nuclear reactor. Resource of dysprosium hafnate ceramics can be 18-20 years instead 4-5 years for boron carbide. To step up the radiation-damage stability of fuel element jacket material was suggested to strengthen a heat-resistant ferrite-martensite steel by Y2O3 nanocrystalline powder addition. Nanopowder with size of particles 560 nm and crystallite size 9 nm was prepeared by chemical coprecipitation method. To make lighter the container for transport and provisional disposal of exposed fuel from nuclear reactor a new boron-aluminium alloy called as boral was developed. This composite armed with nanopowders of boron-containing materials and heavy metals oxides can replace succesburnt-up corrosion-resistant steels.

X-ray Analysis of Defects and Anomalies in AGR-5/6/7 TRISO Particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-06-01

Coated particle fuel batches J52O-16-93164, 93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used for other tests. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.4%-enriched uranium carbide and uranium oxide (UCO), with the exception of Batchmore » 93164, which used similar kernels from BWXT lot J52L-16-69316. The TRISO-coatings consisted of a ~50% dense carbon buffer layer with 100-μmnominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. Each coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (e.g., 93164A). Secondary upgrading by sieving was performed on the upgraded batches to remove specific anomalies identified during analysis for Defective IPyC, and the upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B). Following this secondary upgrading, coated particle composite J52R-16-98005 was produced by BWXT as fuel for the AGR Program’s AGR-5/6/7 irradiation test in the INL ATR. This composite was comprised of coated particle fuel batches J52O-16-93165B, 93168B, 93169B, and 93170B.« less

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

NASA Astrophysics Data System (ADS)

Widargo, Reza; Anghaie, Samim

1999-01-01

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight ratio.

Acceptance Test Data for the AGR-5/6/7 Irradiation Test Fuel Composite Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

Coated particle composite J52R-16-98005 was produced by Babcock and Wilcox Technologies (BWXT) as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). This composite was comprised of four coated particle fuel batches J52O-16-93165B (26%), 93168B (26%), 93169B (24%), and 93170B (24%), chosen based on the Quality Control (QC) data acquired for each individual candidate AGR-5/6/7 batch. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT Lot J52R-16-69317more » containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B).« less

Controlling intake of uranium in the workplace: Applications of biokinetic modeling and occupational monitoring data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leggett, Richard Wayne; Eckerman, Keith F; McGinn, Wilson

2012-01-01

This report provides methods for interpreting and applying occupational uranium monitoring data. The methods are based on current international radiation protection guidance, current information on the chemical toxicity of uranium, and best available biokinetic models for uranium. Emphasis is on air monitoring data and three types of bioassay data: the concentration of uranium in urine; the concentration of uranium in feces; and the externally measured content of uranium in the chest. Primary Reference guidance levels for prevention of chemical effects and limitation of radiation effects are selected based on a review of current scientific data and regulatory principles for settingmore » standards. Generic investigation levels and immediate action levels are then defined in terms of these primary guidance levels. The generic investigation and immediate actions levels are stated in terms of radiation dose and concentration of uranium in the kidneys. These are not directly measurable quantities, but models can be used to relate the generic levels to the concentration of uranium in air, urine, or feces, or the total uranium activity in the chest. Default investigation and immediate action levels for uranium in air, urine, feces, and chest are recommended for situations in which there is little information on the form of uranium taken into the body. Methods are prescribed also for deriving case-specific investigation and immediate action levels for uranium in air, urine, feces, and chest when there is sufficient information on the form of uranium to narrow the range of predictions of accumulation of uranium in the main target organs for uranium: kidneys for chemical effects and lungs for radiological effects. In addition, methods for using the information herein for alternative guidance levels, different from the ones selected for this report, are described.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bakel, Allen J.; Conner, Cliff; Quigley, Kevin

One of the missions of the Reduced Enrichment for Research and Test Reactors (RERTR) program (and now the National Nuclear Security Administrations Material Management and Minimization program) is to facilitate the use of low enriched uranium (LEU) targets for 99Mo production. The conversion from highly enriched uranium (HEU) to LEU targets will require five to six times more uranium to produce an equivalent amount of 99Mo. The work discussed here addresses the technical challenges encountered in the treatment of uranyl nitrate hexahydrate (UNH)/nitric acid solutions remaining after the dissolution of LEU targets. Specifically, the focus of this work is themore » calcination of the uranium waste from 99Mo production using LEU foil targets and the Modified Cintichem Process. Work with our calciner system showed that high furnace temperature, a large vent tube, and a mechanical shield are beneficial for calciner operation. One- and two-step direct calcination processes were evaluated. The high-temperature one-step process led to contamination of the calciner system. The two-step direct calcination process operated stably and resulted in a relatively large amount of material in the calciner cup. Chemically assisted calcination using peroxide was rejected for further work due to the difficulty in handling the products. Chemically assisted calcination using formic acid was rejected due to unstable operation. Chemically assisted calcination using oxalic acid was recommended, although a better understanding of its chemistry is needed. Overall, this work showed that the two-step direct calcination and the in-cup oxalic acid processes are the best approaches for the treatment of the UNH/nitric acid waste solutions remaining from dissolution of LEU targets for 99Mo production.« less

Hydrogen sulfide (H2S) attenuates uranium-induced acute nephrotoxicity through oxidative stress and inflammatory response via Nrf2-NF-κB pathways.

PubMed

Zheng, Jifang; Zhao, Tingting; Yuan, Yan; Hu, Nan; Tang, Xiaoqing

2015-12-05

As an endogenous gaseous mediator, H2S exerts anti-oxidative, anti-inflammatory and cytoprotective effects in kidneys. This study was designed to investigate the protective effect of H2S against uranium-induced nephrotoxicity in adult SD male rats after in vivo effect of uranium on endogenous H2S formation was explored in kidneys. The levels of endogenous H2S and H2S-producing enzymes (CBS and CSE) were measured in renal homogenates from rats intoxicated by an intraperitoneally (i.p.) injection of uranyl acetate at a single dose of 2.5, 5 or 10 mg/kg. In rats injected i.p. with uranyl acetate (5 mg/kg) or NaHS (an H2S donor, 28 or 56 μmol/kg) alone or in combination, we determined biochemical parameters and histopathological alteration to assess kidney function, examined oxidative stress markers, and investigated Nrf2 and NF-κB pathways in kidney homogenates. The results suggest that uranium intoxication in rats decreased endogenous H2S generation as well as CBS and CSE protein expression. NaHS administration in uranium-intoxicated rats ameliorated the renal biochemical indices and histopathological effects, lowered MDA accumulation, and restored GSH level and anti-oxidative enzymes activities like SOD, CAT, GPx and GST. NaHS treatment in uranium-intoxicated rats activated uranium-inhibited protein expression and nuclear translocation of transcription factor Nrf2, which increased protein expression of downstream target-Nrf2 genes HO-1, NQO-1, GCLC, and TXNRD-1. NaHS administration in uranium-intoxicated rats inhibited uranium-induced nuclear translocation and phosphorylation of transcription factor κB/p65, which decreased protein expression of target-p65 inflammatory genes TNF-α, iNOS, and COX-2. Taken together, these data implicate that H2S can afford protection to rat kidneys against uranium-induced adverse effects through induction of antioxidant defense by activating Nrf2 pathway and reduction of inflammatory response by suppressing NF-κB pathway. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

Uranium Anodic Dissolution under Slightly Alkaline Conditions Progress Report Full-Scale Demonstration with DU Foil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gelis, A.; Brown, M. A.; Wiedmeyer, S.

2014-02-18

Argonne National Laboratory (Argonne) is developing an alternative method for digesting irradiated low enriched uranium (LEU) foil targets to produce 99Mo in neutral/alkaline media. This method consists of the electrolytic dissolution of irradiated uranium foil in sodium bicarbonate solution, followed by precipitation of base-insoluble fission and activation products, and uranyl-carbonate species with CaO. The addition of CaO is vital for the effective anion exchange separation of 99MoO 4 2- from the fission products, since most of the interfering anions (e.g., CO 3 2-) are removed from the solution, while molybdate remains in solution. An anion exchange is used to retainmore » and to purify the 99Mo from the filtrate. The electrochemical dissolver has been designed and fabricated in 304 stainless-steel (SS), and tested for the dissolution of a full-size depleted uranium (DU) target, wrapped in Al foil. Future work will include testing with low-burn-up DU foil at Argonne and later with high-burn-up LEU foils at Oak Ridge National Laboratory.« less

Physicochemical characterization of Capstone depleted uranium aerosols III: morphologic and chemical oxide analyses.

PubMed

Krupka, Kenneth M; Parkhurst, Mary Ann; Gold, Kenneth; Arey, Bruce W; Jenson, Evan D; Guilmette, Raymond A

2009-03-01

The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using x-ray diffraction (XRD), and particle morphologies were examined using scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS). The oxidation state of a DU aerosol is important as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles were spherical, occasionally with dendritic or lobed surface structures. Others appear to have fractures that perhaps resulted from abrasion and comminution, or shear bands that developed from plastic deformation of the DU material. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small bits of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of Health Physics to interpret the results of lung solubility studies and in selecting input parameters for dose assessments.

Physicochemical Characterization of Capstone Depleted Uranium Aerosols III: Morphologic and Chemical Oxide Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krupka, Kenneth M.; Parkhurst, MaryAnn; Gold, Kenneth

2009-03-01

The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using X-ray diffraction (XRD) and particle morphologies using scanning electron microscopy/energy dispersive spectrometry (SEM/EDS). The oxidation state of a DU aerosol is importantmore » as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles appear to have been fractured (perhaps as a result of abrasion and comminution); others were spherical, occasionally with dendritic or lobed surface structures. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small chunks of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of The Journal of Health Physics to interpret the results of lung solubility studies and in selecting input parameters for dose assessments.« less

Identification of bacteria synthesizing ribosomal RNA in response to uranium addition during biostimulation at the Rifle, CO Integrated Field Research site

DOE PAGES

McGuinness, Lora R.; Wilkins, Michael J.; Williams, Kenneth H.; ...

2015-09-18

Understanding which organisms are capable of reducing uranium at historically contaminated sites provides crucial information needed to evaluate treatment options and outcomes. One approach is determination of the bacteria which directly respond to uranium addition. In this research, uranium amendments were made to groundwater samples from a site of ongoing biostimulation with acetate. The active microbes in the planktonic phase were deduced by monitoring ribosomes production via RT-PCR. The results indicated several microorganisms were synthesizing ribosomes in proportion with uranium amendment up to 2 μM. Concentrations of U (VI) >2 μM were generally found to inhibit ribosome synthesis. Two activemore » bacteria responding to uranium addition in the field were close relatives of Desulfobacter postgateii and Geobacter bemidjiensis. Since RNA content often increases with growth rate, our findings suggest it is possible to rapidly elucidate active bacteria responding to the addition of uranium in field samples and provides a more targeted approach to stimulate specific populations to enhance radionuclide reduction in contaminated sites.« less

Identification of Bacteria Synthesizing Ribosomal RNA in Response to Uranium Addition During Biostimulation at the Rifle, CO Integrated Field Research Site

PubMed Central

McGuinness, Lora R.; Wilkins, Michael J.; Williams, Kenneth H.; Long, Philip E.; Kerkhof, Lee J.

2015-01-01

Understanding which organisms are capable of reducing uranium at historically contaminated sites provides crucial information needed to evaluate treatment options and outcomes. One approach is determination of the bacteria which directly respond to uranium addition. In this study, uranium amendments were made to groundwater samples from a site of ongoing biostimulation with acetate. The active microbes in the planktonic phase were deduced by monitoring ribosomes production via RT-PCR. The results indicated several microorganisms were synthesizing ribosomes in proportion with uranium amendment up to 2 μM. Concentrations of U (VI) >2 μM were generally found to inhibit ribosome synthesis. Two active bacteria responding to uranium addition in the field were close relatives of Desulfobacter postgateii and Geobacter bemidjiensis. Since RNA content often increases with growth rate, our findings suggest it is possible to rapidly elucidate active bacteria responding to the addition of uranium in field samples and provides a more targeted approach to stimulate specific populations to enhance radionuclide reduction in contaminated sites. PMID:26382047

Uranium: A Dentist's perspective

PubMed Central

Toor, R. S. S.; Brar, G. S.

2012-01-01

Uranium is a naturally occurring radionuclide found in granite and other mineral deposits. In its natural state, it consists of three isotopes (U-234, U-235 and U-238). On an average, 1% – 2% of ingested uranium is absorbed in the gastrointestinal tract in adults. The absorbed uranium rapidly enters the bloodstream and forms a diffusible ionic uranyl hydrogen carbonate complex (UO2HCO3+) which is in equilibrium with a nondiffusible uranyl albumin complex. In the skeleton, the uranyl ion replaces calcium in the hydroxyapatite complex of the bone crystal. Although in North India, there is a risk of radiological toxicity from orally ingested natural uranium, the principal health effects are chemical toxicity. The skeleton and kidney are the primary sites of uranium accumulation. Acute high dose of uranyl nitrate delays tooth eruption, and mandibular growth and development, probably due to its effect on target cells. Based on all previous research and recommendations, the role of a dentist is to educate the masses about the adverse effects of uranium on the overall as well as the dental health. The authors recommended that apart from the discontinuation of the addition of uranium to porcelain, the Public community water supplies must also comply with the Environmental Protection Agency (EPA) standards of uranium levels being not more than 30 ppb (parts per billion). PMID:24478959

Mathematics and statistics research department. Progress report, period ending June 30, 1981

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lever, W.E.; Kane, V.E.; Scott, D.S.

1981-09-01

This report is the twenty-fourth in the series of progress reports of the Mathematics and Statistics Research Department of the Computer Sciences Division, Union Carbide Corporation - Nuclear Division (UCC-ND). Part A records research progress in biometrics research, materials science applications, model evaluation, moving boundary problems, multivariate analysis, numerical linear algebra, risk analysis, and complementary areas. Collaboration and consulting with others throughout the UCC-ND complex are recorded in Part B. Included are sections on biology and health sciences, chemistry, energy, engineering, environmental sciences, health and safety research, materials sciences, safeguards, surveys, and uranium resource evaluation. Part C summarizes the variousmore » educational activities in which the staff was engaged. Part D lists the presentations of research results, and Part E records the staff's other professional activities during the report period.« less

ATOMIC POWER PLANT

DOEpatents

Daniels, F.

1957-11-01

This patent relates to neutronic reactor power plants and discloses a design of a reactor utilizing a mixture of discrete units of a fissionable material, such as uranium carbide, a neutron moderator material, such as graphite, to carry out the chain reaction. A liquid metal, such as bismuth, is used as the coolant and is placed in the reactor chamber with the fissionable and moderator material so that it is boiled by the heat of the reaction, the boiling liquid and vapors passing up through the interstices between the discrete units. The vapor and flue gases coming off the top of the chamber are passed through heat exchangers, to produce steam, for example, and thence through condensers, the condensed coolant being returned to the chamber by gravity and the non- condensible gases being carried off through a stack at the top of the structure.

Measurement of carbon distribution in nuclear fuel pin cladding specimens by means of a secondary ion mass spectrometer

NASA Astrophysics Data System (ADS)

Bart, Gerhard; Aerne, Ernst Tino; Burri, Martin; Zwicky, Hans-Urs

1986-11-01

Cladding carburization during irradiation of advanced mixed uranium plutonium carbide fast breeder reactor fuel is possibly a life limiting fuel pin factor. The quantitative assessment of such clad carbon embrittlement is difficult to perform by electron microprobe analysis because of sample surface contamination, and due to the very low energy of the carbon K α X-ray transition. The work presented here describes a method developed at the Swiss Federal Institute for Reactor Research (EIR) to use shielded secondary ion mass spectrometry (SIMS) as an accurate tool to determine radial distribution profiles of carbon in radioactive stainless steel fuel pin cladding. Compared with nuclear microprobe analysis (NMA) [1], which is also an accurate method for carbon analysis, the SIMS method distinguishes itself by its versatility for simultaneous determination of additional impurities.

Safety Testing of AGR-2 UCO Compacts 5-2-2, 2-2-2, and 5-4-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.

2016-08-01

Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign, and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO 2) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where somemore » of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to those for AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels, notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling.« less

Direct Compositional Characterization of (U,Th)O2 Powders, Microspheres, and Pellets Using TXRF.

PubMed

Dhara, Sangita; Prabhat, Parimal; Misra, N L

2015-10-20

A total reflection X-ray fluorescence (TXRF) analysis method for direct compositional characterization of sintered and green (U,Th)O2 samples in different forms (e.g., pellets, powders, and microspheres) without sample dissolution has been developed for the first time. The methodology involves transfer of only a few nanograms of the sample on the TXRF sample support by gently rubbing the samples on supports or taking their tiny uniform slurry in collodion on the sample support, drying them to make thin film, and measuring the TXRF spectra of the specimens thus prepared. This approach minimizes the matrix effects. Uranium determinations from the TXRF spectra of such specimens were made with respect to thorium, considering it as an internal standard. Samples having uranium atom percent (at%) from 0 to 100 in (U,Th)O2 were analyzed for uranium in comparison to thorium. The results showed an average precision of 2.6% (RSD, 2σ, n = 8). The TXRF-determined results deviated from expected values within 5%. The TXRF results were compared with those of biamperometry with good agreement. The lattice parameters of the solid solutions were calculated using their XRD patterns. A good correlation between lattice parameters and TXRF-determined U at% and between TXRF-determined U at% and expected U at%, calculated on the basis of preparation of (U,Th)O2 solid solutions, was obtained. The developed method is capable of analyzing (U,Th)O2 samples directly with almost negligible sample preparation and is well suited for radioactive samples. The present study suggests that this method can be extended for the determination of U,Th and Pu in other nuclear fuel materials (e.g., nitrides, carbides, etc.) in the form of pellets, powders, and microspheres after suitable modifications in sample handling procedure.

Remanent Activation in the Mini-SHINE Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Micklich, Bradley J.

2015-04-16

Argonne National Laboratory is assisting SHINE Medical Technologies in developing a domestic source of the medical isotope 99Mo through the fission of low-enrichment uranium in a uranyl sulfate solution. In Phase 2 of these experiments, electrons from a linear accelerator create neutrons by interacting in a depleted uranium target, and these neutrons are used to irradiate the solution. The resulting neutron and photon radiation activates the target, the solution vessels, and a shielded cell that surrounds the experimental apparatus. When the experimental campaign is complete, the target must be removed into a shielding cask, and the experimental components must bemore » disassembled. The radiation transport code MCNPX and the transmutation code CINDER were used to calculate the radionuclide inventories of the solution, the target assembly, and the shielded cell, and to determine the dose rates and shielding requirements for selected removal scenarios for the target assembly and the solution vessels.« less

A fast-neutron detection detector based on fission material and large sensitive 4H silicon carbide Schottky diode detector

NASA Astrophysics Data System (ADS)

Liu, Linyue; Liu, Jinliang; Zhang, Jianfu; Chen, Liang; Zhang, Xianpeng; Zhang, Zhongbing; Ruan, Jinlu; Jin, Peng; Bai, Song; Ouyang, Xiaoping

2017-12-01

Silicon carbide radiation detectors are attractive in the measurement of the total numbers of pulsed fast neutrons emitted from nuclear fusion and fission devices because of high neutron-gamma discrimination and good radiation resistance. A fast-neutron detection system was developed based on a large-area 4H-SiC Schottky diode detector and a 235U fission target. Excellent pulse-height spectra of fission fragments induced by mono-energy deuterium-tritium (D-T) fusion neutrons and continuous energy fission neutrons were obtained. The detector is proven to be a good candidate for pulsed fast neutron detection in a complex radiation field.

Recovering and recycling uranium used for production of molybdenum-99

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reilly, Sean Douglas; May, Iain; Copping, Roy

A processes for recycling uranium that has been used for the production of molybdenum-99 involves irradiating a solution of uranium suitable for forming fission products including molybdenum-99, conditioning the irradiated solution to one suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina. Another process involves irradiation of a solid target comprising uranium, forming an acidic solution from the irradiated targetmore » suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina.« less

Modeling spallation reactions in tungsten and uranium targets with the Geant4 toolkit

NASA Astrophysics Data System (ADS)

Malyshkin, Yury; Pshenichnov, Igor; Mishustin, Igor; Greiner, Walter

2012-02-01

We study primary and secondary reactions induced by 600 MeV proton beams in monolithic cylindrical targets made of natural tungsten and uranium by using Monte Carlo simulations with the Geant4 toolkit [1-3]. Bertini intranuclear cascade model, Binary cascade model and IntraNuclear Cascade Liège (INCL) with ABLA model [4] were used as calculational options to describe nuclear reactions. Fission cross sections, neutron multiplicity and mass distributions of fragments for 238U fission induced by 25.6 and 62.9 MeV protons are calculated and compared to recent experimental data [5]. Time distributions of neutron leakage from the targets and heat depositions are calculated. This project is supported by Siemens Corporate Technology.

Inhibition of poly(ADP-ribose)polymerase-1 and DNA repair by uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cooper, Karen L.; Dashner, Erica J.; Tsosie, Ranalda

Uranium has radiological and non-radiological effects within biological systems and there is increasing evidence for genotoxic and carcinogenic properties attributable to uranium through its heavy metal properties. In this study, we report that low concentrations of uranium (as uranyl acetate; < 10 μM) is not cytotoxic to human embryonic kidney cells or normal human keratinocytes; however, uranium exacerbates DNA damage and cytotoxicity induced by hydrogen peroxide, suggesting that uranium may inhibit DNA repair processes. Concentrations of uranyl acetate in the low micromolar range inhibited the zinc finger DNA repair protein poly(ADP-ribose) polymerase (PARP)-1 and caused zinc loss from PARP-1 protein.more » Uranyl acetate exposure also led to zinc loss from the zinc finger DNA repair proteins Xeroderma Pigmentosum, Complementation Group A (XPA) and aprataxin (APTX). In keeping with the observed inhibition of zinc finger function of DNA repair proteins, exposure to uranyl acetate enhanced retention of induced DNA damage. Co-incubation of uranyl acetate with zinc largely overcame the impact of uranium on PARP-1 activity and DNA damage. These findings present evidence that low concentrations of uranium can inhibit DNA repair through disruption of zinc finger domains of specific target DNA repair proteins. This may provide a mechanistic basis to account for the published observations that uranium exposure is associated with DNA repair deficiency in exposed human populations. - Highlights: • Low micromolar concentration of uranium inhibits polymerase-1 (PARP-1) activity. • Uranium causes zinc loss from multiple DNA repair proteins. • Uranium enhances retention of DNA damage caused by ultraviolet radiation. • Zinc reverses the effects of uranium on PARP activity and DNA damage repair.« less

a Plutonium Ceramic Target for Masha

NASA Astrophysics Data System (ADS)

Wilk, P. A.; Shaughnessy, D. A.; Moody, K. J.; Kenneally, J. M.; Wild, J. F.; Stoyer, M. A.; Patin, J. B.; Lougheed, R. W.; Ebbinghaus, B. B.; Landingham, R. L.; Oganessian, Yu. Ts.; Yeremin, A. V.; Dmitriev, S. N.

2005-09-01

We are currently developing a plutonium ceramic target for the MASHA mass separator. The MASHA separator will use a thick plutonium ceramic target capable of tolerating temperatures up to 2000 °C. Promising candidates for the target include oxides and carbides, although more research into their thermodynamic properties will be required. Reaction products will diffuse out of the target into an ion source, where they will then be transported through the separator to a position-sensitive focal-plane detector array. Experiments on MASHA will allow us to make measurements that will cement our identification of element 114 and provide for future experiments where the chemical properties of the heaviest elements are studied.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youker, Amanda J.; Krebs, John F.; Quigley, Kevin J.

With funding from the National Nuclear Security Administrations Material Management and Minimization Office, Argonne National Laboratory (Argonne) is providing technical assistance to help accelerate the U.S. production of Mo-99 using a non-highly enriched uranium (non-HEU) source. A potential Mo-99 production pathway is by accelerator-initiated fissioning in a subcritical uranyl sulfate solution containing low enriched uranium (LEU). As part of the Argonne development effort, we are undertaking the AMORE (Argonne Molybdenum Research Experiment) project, which is essentially a pilot facility for all phases of Mo-99 production, recovery, and purification. Production of Mo-99 and other fission products in the subcritical target solutionmore » is initiated by putting an electron beam on a depleted uranium (DU) target; the fast neutrons produced in the DU target are thermalized and lead to fissioning of U-235. At the end of irradiation, Mo is recovered from the target solution and separated from uranium and most of the fission products by using a titania column. The Mo is stripped from the column with an alkaline solution. After acidification of the Mo product solution from the recovery column, the Mo is concentrated (and further purified) in a second titania column. The strip solution from the concentration column is then purified with the LEU Modified Cintichem process. A full description of the process can be found elsewhere [1–3]. The initial commissioning steps for the AMORE project include performing a Mo-99 spike test with pH 1 sulfuric acid in the target vessel without a beam on the target to demonstrate the initial Mo separation-and-recovery process, followed by the concentration column process. All glovebox operations were tested with cold solutions prior to performing the Mo-99 spike tests. Two Mo-99 spike tests with pH 1 sulfuric acid have been performed to date. Figure 1 shows the flow diagram for the remotely operated Mo-recovery system for the AMORE project. There are two separate pumps and flow paths for the acid and base operations. The system contains three sample ladders with eight sample loops per ladder for target mixing; column loading, including acid and water washes; and column stripping, including the final water wash.« less

Extraction of reduced alteration information based on Aster data: a case study of the Bashibulake uranium ore district

NASA Astrophysics Data System (ADS)

Ye, Fa-wang; Liu, De-chang

2008-12-01

Practices of sandstone-type uranium exploration in recent years in China indicate that the uranium mineralization alteration information is of great importance for selecting a new uranium target or prospecting in outer area of the known uranium ore district. Taking a case study of BASHIBULAKE uranium ore district, this paper mainly presents the technical minds and methods of extracting the reduced alteration information by oil and gas in BASHIBULAKE ore district using ASTER data. First, the regional geological setting and study status in BASHIBULAKE uranium ore district are introduced in brief. Then, the spectral characteristics of altered sandstone and un-altered sandstone in BASHIBULAKE ore district are analyzed deeply. Based on the spectral analysis, two technical minds to extract the remote sensing reduced alteration information are proposed, and the un-mixing method is introduced to process ASTER data to extract the reduced alteration information in BASHIBULAKE ore district. From the enhanced images, three remote sensing anomaly zones are discovered, and their geological and prospecting significances are further made sure by taking the advantages of multi-bands in SWIR of ASTER data. Finally, the distribution and intensity of the reduced alteration information in Cretaceous system and its relationship with the genesis of uranium deposit are discussed, the specific suggestions for uranium prospecting orientation in outer of BASHIBULAKE ore district are also proposed.

Inhalation of uranium nanoparticles: respiratory tract deposition and translocation to secondary target organs in rats.

PubMed

Petitot, Fabrice; Lestaevel, Philippe; Tourlonias, Elie; Mazzucco, Charline; Jacquinot, Sébastien; Dhieux, Bernadette; Delissen, Olivia; Tournier, Benjamin B; Gensdarmes, François; Beaunier, Patricia; Dublineau, Isabelle

2013-03-13

Uranium nanoparticles (<100 nm) can be released into the atmosphere during industrial stages of the nuclear fuel cycle and during remediation and decommissioning of nuclear facilities. Explosions and fires in nuclear reactors and the use of ammunition containing depleted uranium can also produce such aerosols. The risk of accidental inhalation of uranium nanoparticles by nuclear workers, military personnel or civilian populations must therefore be taken into account. In order to address this issue, the absorption rate of inhaled uranium nanoparticles needs to be characterised experimentally. For this purpose, rats were exposed to an aerosol containing 10⁷ particles of uranium per cm³ (CMD=38 nm) for 1h in a nose-only inhalation exposure system. Uranium concentrations deposited in the respiratory tract, blood, brain, skeleton and kidneys were determined by ICP-MS. Twenty-seven percent of the inhaled mass of uranium nanoparticles was deposited in the respiratory tract. One-fifth of UO₂ nanoparticles were rapidly cleared from lung (T(½)=2.4 h) and translocated to extrathoracic organs. However, the majority of the particles were cleared slowly (T(½)=141.5 d). Future long-term experimental studies concerning uranium nanoparticles should focus on the potential lung toxicity of the large fraction of particles cleared slowly from the respiratory tract after inhalation exposure. Copyright © 2013 Elsevier Ireland Ltd. All rights reserved.

Inhibition of poly(ADP-ribose)polymerase-1 and DNA repair by uranium

PubMed Central

Cooper, Karen L.; Dashner, Erica J.; Tsosie, Ranalda; Cho, Young Mi; Lewis, Johnnye

2015-01-01

Uranium has radiological and non-radiological effects within biological systems and there is increasing evidence for genotoxic and carcinogenic properties attributable to uranium through its heavy metal properties. In this study, we report that low concentrations of uranium (as uranyl acetate; <10 μM) is not cytotoxic to human embryonic kidney cells or normal human keratinocytes; however, uranium exacerbates DNA damage and cytotoxicity induced by hydrogen peroxide, suggesting that uranium may inhibit DNA repair processes. Concentrations of uranyl acetate in the low micromolar range inhibited the zinc finger DNA repair protein poly(ADP-ribose) polymerase (PARP)-1 and caused zinc loss from PARP-1 protein. Uranyl acetate exposure also led to zinc loss from the zinc finger DNA repair proteins Xeroderma Pigmentosum, Complementation Group A (XPA) and aprataxin (APTX). In keeping with the observed inhibition of zinc finger function of DNA repair proteins, exposure to uranyl acetate enhanced retention of induced DNA damage. Co-incubation of uranyl acetate with zinc largely overcame the impact of uranium on PARP-1 activity and DNA damage. These findings present evidence that low concentrations of uranium can inhibit DNA repair through disruption of zinc finger domains of specific target DNA repair proteins. This may provide a mechanistic basis to account for the published observations that uranium exposure is associated with DNA repair deficiency in exposed human populations. PMID:26627003

Inhibition of poly(ADP-ribose)polymerase-1 and DNA repair by uranium.

PubMed

Cooper, Karen L; Dashner, Erica J; Tsosie, Ranalda; Cho, Young Mi; Lewis, Johnnye; Hudson, Laurie G

2016-01-15

Uranium has radiological and non-radiological effects within biological systems and there is increasing evidence for genotoxic and carcinogenic properties attributable to uranium through its heavy metal properties. In this study, we report that low concentrations of uranium (as uranyl acetate; <10 μM) is not cytotoxic to human embryonic kidney cells or normal human keratinocytes; however, uranium exacerbates DNA damage and cytotoxicity induced by hydrogen peroxide, suggesting that uranium may inhibit DNA repair processes. Concentrations of uranyl acetate in the low micromolar range inhibited the zinc finger DNA repair protein poly(ADP-ribose) polymerase (PARP)-1 and caused zinc loss from PARP-1 protein. Uranyl acetate exposure also led to zinc loss from the zinc finger DNA repair proteins Xeroderma Pigmentosum, Complementation Group A (XPA) and aprataxin (APTX). In keeping with the observed inhibition of zinc finger function of DNA repair proteins, exposure to uranyl acetate enhanced retention of induced DNA damage. Co-incubation of uranyl acetate with zinc largely overcame the impact of uranium on PARP-1 activity and DNA damage. These findings present evidence that low concentrations of uranium can inhibit DNA repair through disruption of zinc finger domains of specific target DNA repair proteins. This may provide a mechanistic basis to account for the published observations that uranium exposure is associated with DNA repair deficiency in exposed human populations. Copyright © 2015 Elsevier Inc. All rights reserved.

Investigations into Alternative Desorption Agents for Amidoxime-Based Polymeric Uranium Adsorbents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gill, Gary A.; Kuo, Li-Jung; Strivens, Jonathan E.

2015-06-01

Amidoxime-based polymeric braid adsorbents that can extract uranium (U) from seawater are being developed to provide a sustainable supply of fuel for nuclear reactors. A critical step in the development of the technology is to develop elution procedures to selectively remove U from the adsorbents and to do so in a manner that allows the adsorbent material to be reused. This study investigates use of high concentrations of bicarbonate along with targeted chelating agents as an alternative means to the mild acid elution procedures currently in use for selectively eluting uranium from amidoxime-based polymeric adsorbents.

Method for production of ceramic oxide and carbide bodies by polymer inclusion and decomposition

DOEpatents

Quinby, Thomas C.

1985-01-01

A method for the preparation of thin, free-standing metal oxide films which are useful as nuclear accelerator target materials. Cations of any metal except those of Group IA and precious metals, such as, U, Zr, Nd, Ce, Th, pr or Cr, are absorbed on a thin film of polymeric material, such as, carboxymethylcellulose, viscose rayon or cellophane. The cation impregnated polymeric material is dried. Then the impregnated film is heated in an inert atmosphere to form a carbonized membrane. The carbonized membrane is oxidized to yield a thin, self-supporting, metal oxide membrane. Or, the membrane can be heated in an inert atmosphere to yield a thin, self-supporting, metal carbide-containing membrane.

Method for production of ceramic oxide and carbide bodies by polymer inclusion and decomposition

DOEpatents

Quinby, T.C.

1984-08-30

A method for the preparation of thin, free-standing metal oxide films which are useful as nuclear accelerator target materials is described. Cations of any metal except those of Group IA and precious metals, such as, U, Zr, Nd, Ce, Th, Pr or Cr, are absorbed on a thin film of polymeric material, such as carboxymethylcellulose, viscose rayon or cellophane. The cation impregnated polymeric material is dried. Then the impregnated film is heated in an inert atmosphere to form a carbonized membrane. The carbonized membrane is oxidized to yield a thin, self-supporting, metal oxide membrane. Or, the membrane can be heated in an inert atmosphere to yield a thin, self-supporting, metal carbide-containing membrane.

Evolution of the atomic order and valence state of rare-earth atoms and uranium in a new carbon-metal composite—diphthalocyanine pyrolysate C64H32N16 Me ( Me = Y, La, Ce, Eu, and U)

NASA Astrophysics Data System (ADS)

Sovestnov, A. E.; Kapustin, V. K.; Tikhonov, V. I.; Fomin, E. V.; Chernenkov, Yu. P.

2014-08-01

The structure of a metal-carbon composite formed by the pyrolysis of diphthalocyanine of some rare-earth elements (Y, La, Ce, Eu) and uranium in the temperature range T ann = 800-1700°C has been investigated for the first time by the methods of X-ray diffraction analysis and X-ray line shift. It has been shown that, in the general case, the studied pyrolysates consist of three phases. One phase corresponds to the structure of graphite. The second phase corresponds to nitrides, carbides, and oxides of basic metal elements with a crystallite size ranging from 5 to 100 nm. The third phase is amorphous or consisting of crystallites with a size of ˜1 nm. It has been found that all the basic elements (Y, La, Ce, Eu, U) and incorporated iodine atoms in the third phase are in a chemically bound state. The previously unobserved electronic configurations have been revealed for europium. The possibility of including not only atoms of elements forming diphthalocyanine but also other elements (for example, iodine) in the composite structure is of interest, in particular, for the creation of a thermally, chemically, and radiation resistant metal-carbon matrix for the radioactive waste storage.

Process for microwave sintering boron carbide

DOEpatents

Holcombe, C.E.; Morrow, M.S.

1993-10-12

A method of microwave sintering boron carbide comprises leaching boron carbide powder with an aqueous solution of nitric acid to form a leached boron carbide powder. The leached boron carbide powder is coated with a glassy carbon precursor to form a coated boron carbide powder. The coated boron carbide powder is consolidated in an enclosure of boron nitride particles coated with a layer of glassy carbon within a container for microwave heating to form an enclosed coated boron carbide powder. The enclosed coated boron carbide powder is sintered within the container for microwave heating with microwave energy.

Process for microwave sintering boron carbide

DOEpatents

Holcombe, Cressie E.; Morrow, Marvin S.

1993-01-01

A method of microwave sintering boron carbide comprises leaching boron carbide powder with an aqueous solution of nitric acid to form a leached boron carbide powder. The leached boron carbide powder is coated with a glassy carbon precursor to form a coated boron carbide powder. The coated boron carbide powder is consolidated in an enclosure of boron nitride particles coated with a layer of glassy carbon within a container for microwave heating to form an enclosed coated boron carbide powder. The enclosed coated boron carbide powder is sintered within the container for microwave heating with microwave energy.

Low resistance barrier layer for isolating, adhering, and passivating copper metal in semiconductor fabrication

DOEpatents

Weihs, Timothy P.; Barbee, Jr., Troy W.

2002-01-01

Cubic or metastable cubic refractory metal carbides act as barrier layers to isolate, adhere, and passivate copper in semiconductor fabrication. One or more barrier layers of the metal carbide are deposited in conjunction with copper metallizations to form a multilayer characterized by a cubic crystal structure with a strong (100) texture. Suitable barrier layer materials include refractory transition metal carbides such as vanadium carbide (VC), niobium carbide (NbC), tantalum carbide (TaC), chromium carbide (Cr.sub.3 C.sub.2), tungsten carbide (WC), and molybdenum carbide (MoC).

HEU Holdup Measurements in 321-M B and Spare U-Al Casting Furnaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salaymeh, S.R.

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Decontamination Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. This report covers holdup measurements in two uranium aluminum alloy (U-Al) casting furnaces. Our results indicate an upper limit of 235U content for the B and Spare furnaces of 51 and 67 g respectively. This report discusses themore » methodology, non-destructive assay (NDA) measurements, and results of the uranium holdup on the two furnaces.« less

AGR-5/6/7 LEUCO Kernel Fabrication Readiness Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, Douglas W.; Bailey, Kirk W.

2015-02-01

In preparation for forming low-enriched uranium carbide/oxide (LEUCO) fuel kernels for the Advanced Gas Reactor (AGR) fuel development and qualification program, Idaho National Laboratory conducted an operational readiness review of the Babcock & Wilcox Nuclear Operations Group – Lynchburg (B&W NOG-L) procedures, processes, and equipment from January 14 – January 16, 2015. The readiness review focused on requirements taken from the American Society Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008, 1a-2009), a recent occurrence at the B&W NOG-L facility related to preparation of acid-deficient uranyl nitrate solution (ADUN), and a relook at concerns noted in a previous review. Topicmore » areas open for the review were communicated to B&W NOG-L in advance of the on-site visit to facilitate the collection of objective evidences attesting to the state of readiness.« less

Progress in understanding fission-product behaviour in coated uranium-dioxide fuel particles

NASA Astrophysics Data System (ADS)

Barrachin, M.; Dubourg, R.; Kissane, M. P.; Ozrin, V.

2009-03-01

Supported by results of calculations performed with two analytical tools (MFPR, which takes account of physical and chemical mechanisms in calculating the chemical forms and physical locations of fission products in UO2, and MEPHISTA, a thermodynamic database), this paper presents an investigation of some important aspects of the fuel microstructure and chemical evolutions of irradiated TRISO particles. The following main conclusions can be identified with respect to irradiated TRISO fuel: first, the relatively low oxygen potential within the fuel particles with respect to PWR fuel leads to chemical speciation that is not typical of PWR fuels, e.g., the relatively volatile behaviour of barium; secondly, the safety-critical fission-product caesium is released from the urania kernel but the buffer and pyrolytic-carbon coatings could form an important chemical barrier to further migration (i.e., formation of carbides). Finally, significant releases of fission gases from the urania kernel are expected even in nominal conditions.

Thermodynamic assessment of oxygen diffusion in non-stoichiometric UO2±x from experimental data and Frenkel pair modeling

NASA Astrophysics Data System (ADS)

Berthinier, C.; Rado, C.; Chatillon, C.; Hodaj, F.

2013-02-01

The self and chemical diffusion of oxygen in the non-stoichiometric domain of the UO2 compound is analyzed from the point of view of experimental determinations and modeling from Frenkel pair defects. The correlation between the self-diffusion and the chemical diffusion coefficients is analyzed using the Darken coefficient calculated from a thermodynamic description of the UO2±x phase. This description was obtained from an optimization of thermodynamic and phase diagram data and modeling with different point defects, including the Frenkel pair point defects. The proposed diffusion coefficients correspond to the 300-2300 K temperature range and to the full composition range of the non stoichiometric UO2 compound. These values will be used for the simulation of the oxidation and ignition of the uranium carbide in different oxygen atmospheres that starts at temperatures as low as 400 K.

Ab initio study on structural stability of uranium carbide

NASA Astrophysics Data System (ADS)

Sahoo, B. D.; Joshi, K. D.; Gupta, Satish C.

2013-06-01

First principles calculations have been performed using plane wave pseudopotential and full potential linearized augmented plane wave (FP-LAPW) methods to analyze structural, elastic and dynamic stability of UC under hydrostatic compression. Our calculations within pseudopotential method suggest that the rocksalt (B1) structure will transform to body centered orthorhombic (bco) structure at ˜21.5 GPa. The FP-LAPW calculations put this transition at 23 GPa. The transition pressures determined from our calculations though agree reasonably with the experimental value of 27 GPa, the high pressure bco structure suggested by theory differs slightly from the experimentally reported pseudo bco phase. The elastic stability analysis of B1 phase suggests that the B1 to bco transition is driven by the failure of C44 modulus. This finding is further substantiated by the lattice dynamic calculations which demonstrate that the B1 phase becomes dynamically unstable around the transition pressure and the instability is of long wavelength nature.

Uranium potential of precambrian rocks in the Raft River area of northwestern Utah and south-central Idaho. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Black, B.A.

1980-09-01

A total of 1214 geochemical samples were collected and analyzed. The sampling media included 334 waters, 616 stream sediments, and 264 rocks. In addition, some stratigraphic sections of Elba and Yost Quartzites and Archean metasedimentary rock were measured and sampled and numerous radiation determinations made of the various target units. Statistical evaluation of the geochemical data permitted recognition of 156 uranium anomalies, 52 in water, 79 in stream sediment, and 25 in rock. Geographically, 68 are located in the Grouse Creek Mountains, 43 in the Raft River Mountains, and 41 in the Albion Range. Interpretation of the various data leadsmore » to the conclusion that uranium anomalies relate to sparingly and moderately soluble uraniferous heavy minerals, which occur as sparse but widely distributed magmatic, detrital, and/or metamorphically segregated components in the target lithostratigraphic units. The uraniferous minerals known to occur and believed to account for the geochemical anomalies include allanite, monazite, zircon, and apatite. In some instances samarskite may be important. These heavy minerals contain uranium and geochemically related elements, such as Th, Ce, Y, and Zr, in sufficient quantities to account for both the conspicuous lithologic preference and the generally observed low amplitude of the anomalies. The various data generated in connection with this study, as well as those available in the published literature, collectively support the conclusion that the various Precambrian W and X lithostratigraphic units pre-selected for evaluation probably lack potential to host important Precambrian quartz-pebble conglomerate uranium deposits. Moreover it is also doubted that they possess any potential to host Proterozoic unconformity-type uranium deposits.« less

Nuclear Excitation by Electronic Transition of U-235

NASA Astrophysics Data System (ADS)

Chodash, Perry

2017-01-01

Nuclear excitation by electronic transition (NEET) is a rare nuclear excitation that is theorized to exist in numerous isotopes. NEET is the inverse of bound internal conversion and occurs when an electronic transition couples to a nuclear transition causing the nucleus to enter an excited state. This process can only occur for isotopes with low-lying nuclear levels due to the requirement that the electronic and nuclear transitions have similar energies. One of the candidate isotopes for NEET, 235U, has been studied several times over the past 40 years and NEET of 235U has never been conclusively observed. These past experiments generated conflicting results with some experiments claiming to observe NEET of 235U and others setting limits for the NEET rate. If NEET of 235U were to occur, the uranium would be excited to its first excited nuclear state. The first excited nuclear state in 235U is only 76 eV, the second lowest known nuclear state. Additionally, the 76 eV state is a nuclear isomer that decays by internal conversion with a half-life of 26 minutes. In order to measure whether NEET occurs in 235U and at what rate, a uranium plasma was required. The plasma was generated using a Q-switched Nd:YAG laser outputting 789 mJ pulses of 1064 nm light. The laser light was focused onto uranium targets generating an intensity on target of order 1012 W/cm2. The resulting plasma was captured on a catcher plate and electrons emitted from the catcher plate were accelerated and focused onto a microchannel plate detector. Measurements performed using a variety of uranium targets spanning depleted uranium up to 99.4% enriched uranium did not observe a 26 minute decay. An upper limit for the NEET rate of 235U was determined. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. The U.S. DHS, UC Berkeley, the NNIS fellowship and the NSSC further supported this work.

A Plutonium Ceramic Target for MASHA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilk, P A; Shaughnessy, D A; Moody, K J

2004-07-06

We are currently developing a plutonium ceramic target for the MASHA mass separator. The MASHA separator will use a thick plutonium ceramic target capable of tolerating temperatures up to 2000 C. Promising candidates for the target include oxides and carbides, although more research into their thermodynamic properties will be required. Reaction products will diffuse out of the target into an ion source, where they will then be transported through the separator to a position-sensitive focal-plane detector array. Experiments on MASHA will allow us to make measurements that will cement our identification of element 114 and provide for future experiments wheremore » the chemical properties of the heaviest elements are studied.« less

Bulk metallic glass matrix composites: Processing, microstructure, and application as a kinetic energy penetrator

NASA Astrophysics Data System (ADS)

Dandliker, Richard B.

The development of alloys with high glass forming ability allows fabrication of bulk samples of amorphous metal. This capability makes these materials available for applications which require significant material thickness in all three dimensions. Superior mechanical properties and advantages in processing make metallic glass a choice candidate as a matrix material for composites. This study reports techniques for making composites by melt-infiltration casting using the alloy Zrsb{41.2}Tisb{13.8}Cusb{12.5}Nisb{10.0}Besb{22.5} (VitreloyspTM 1) as a matrix material. Composite rods 5 cm in length and 7 mm in diameter were made and found to have a nearly fully amorphous matrix; there was less than 3 volume percent crystallized matrix material. The samples were reinforced by continuous metal wires, tungsten powder, or silicon carbide particulate preforms. The most easily processed samples were made with uniaxially aligned tungsten and carbon steel continuous wire reinforcement; the majority of the analysis presented is of these samples. The measured porosity was typically less than 3%. The results also indicate necessary guidelines for developing processing techniques for large scale production, new reinforcement materials, and other metallic glass compositions. Analysis of the microstructure of the tungsten wire and steel wire reinforced composites was performed by x-ray diffraction, scanning electron microscopy, scanning Auger microscopy, transmission electron microscopy, and energy dispersive x-ray spectroscopy. The most common phase in the crystallized matrix is most likely a Laves phase with the approximate formula Besb{12}Zrsb3TiNiCu. In tungsten-reinforced composites, a crystalline reaction layer 240 nm thick of tungsten nanocrystals in an amorphous matrix formed. In the steel reinforced composites, the reaction layer was primarily composed of a mixed metal carbide, mainly ZrC. One promising application of the metallic glass matrix composite is as a kinetic energy penetrator material. Ballistic tests show that a composite of 80 volume percent uniaxially aligned tungsten wires and a VitreloyspTM 1 matrix has self-sharpening behavior, which is a necessary characteristic of superior penetrator materials. Small-scale tests with both aluminum and steel targets show that this composite performs better than tungsten heavy alloys typically used for penetrator applications, and comparably with depleted uranium.

Surface Water-Groundwater Interactions as a Critical Component of Uranium Plume Persistence

NASA Astrophysics Data System (ADS)

Williams, K. H.; Christensen, J. N.; Hobson, C.

2015-12-01

Residual contamination of soils, sediments and groundwater by uranium milling operations presents a lingering problem at former mill sites throughout the upper Colorado River Basin in the western USA. Remedial strategies predicated upon natural flushing by low uranium recharge waters have frequently failed to achieve target concentrations set by national and state regulators. Flushing times of tens of years have often yielded negligible decreases in groundwater uranium concentrations, with extrapolated trends suggesting multiple decades or longer may be required to achieve regulatory goals. The U.S. Department of Energy's Rifle, Colorado field site serves as a natural laboratory for investigating the underlying causes for uranium plume persistence, with recent studies there highlighting the important role that surface water-groundwater interactions play in sustaining uranium delivery to the aquifer. Annual snowmelt-driven increases in Colorado River discharge induce 1-2 m excursions in groundwater elevation at the Rifle site, which enables residual tailings-contaminated materials (so-called Supplemental Standards) to become hydrologically connected to the aquifer for short periods of time during peak discharge. The episodic contact between shallow groundwater and residual contamination leads to abrupt 20-fold increases in groundwater uranium concentration, which serve to seasonally replenish the plume given the location of the Supplemental Standards along the upgradient edge of the aquifer. Uranium isotope composition changes abruptly as uranium concentrations increase reflecting the contribution of a temporally distinct contaminant reservoir. The release of uranium serves to potentially replenish organic matter rich sediments located within the alluvial aquifer at downstream locations, which have been postulated to serve as a parallel contributor to plume persistence following the uptake, immobilization, and slow re-oxidation of uranium.

Ablation Resistant Zirconium and Hafnium Ceramics

NASA Technical Reports Server (NTRS)

Bull, Jeffrey (Inventor); White, Michael J. (Inventor); Kaufman, Larry (Inventor)

1998-01-01

High temperature ablation resistant ceramic composites have been made. These ceramics are composites of zirconium diboride and zirconium carbide with silicon carbide, hafnium diboride and hafnium carbide with silicon carbide and ceramic composites which contain mixed diborides and/or carbides of zirconium and hafnium. along with silicon carbide.

An Overview of Process Monitoring Related to the Production of Uranium Ore Concentrate

DOE Office of Scientific and Technical Information (OSTI.GOV)

McGinnis, Brent

2014-04-01

Uranium ore concentrate (UOC) in various chemical forms, is a high-value commodity in the commercial nuclear market, is a potential target for illicit acquisition, by both State and non-State actors. With the global expansion of uranium production capacity, control of UOC is emerging as a potentially weak link in the nuclear supply chain. Its protection, control and management thus pose a key challenge for the international community, including States, regulatory authorities and industry. This report evaluates current process monitoring practice and makes recommendations for utilization of existing or new techniques for managing the inventory and tracking this material.

Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer Coefficient (HTC), bulk-fluid, sheath and fuel centerline temperature profiles were generated along the heated length of 5.772 m for a generic fuel channel. The fuel centerline and sheath temperature profiles have been determined at four Axial Heat Flux Profiles (AHFPs) using an average thermal power per channel of 8.5 MWth. The four examined AHFPs are the uniform, cosine, upstream-skewed and downstream-skewed profiles. Additionally, this study focuses on investigating a possibility of using low, enhanced and high thermal-conductivity fuels. The low thermal-conductivity fuels, which have been examined in this study, are uranium dioxide (UO 2), Mixed Oxide (MOX) and Thoria (ThO2) fuels. The examined enhanced thermal-conductivity fuels are uranium dioxide - silicon carbide (UO2 - SiC) and uranium dioxide - beryllium oxide (UO2 - BeO). Lastly, uranium carbide (UC), uranium dicarbide (UC2) and uranium nitride (UN) are the selected high thermal-conductivity fuels, which have been proposed for use in SCWRs. A comparison has been made between the low, enhanced and high thermal-conductivity fuels in order to identify the fuel centerline temperature behaviour when different nuclear fuels are used. Also, in the process of conducting the sensitivity analysis, the HTC was calculated using the Mokry et al. correlation, which is the most accurate supercritical water heat-transfer correlation so far. The sheath and the fuel centerline temperature profiles were determined for two cases. In Case 1, the HTC was calculated based on the Mokry et al. correlation, while in Case 2, the HTC values calculated for Case 1 were multiplied by a factor of 2. This factor was used in order to identify the amount of decrease in temperatures if the heat transfer is enhanced with appendages. Results of this analysis indicate that the use of the newly developed 54-element fuel bundle along with the proposed fuels is promising when compared with the Variant-20 (43-element) fuel bundle. Overall, the fuel centerline and sheath temperatures were below the industry and design limits when most of the proposed fuels were examined in the 54-element fuel bundle, however, the fuel centerline temperature limit was exceeded while MOX fuel was examined. Keywords: SCWRs, Fuel Centerline Temperature, Sheath Temperature, High Thermal Conductivity Fuels, Low Thermal Conductivity Fuels, HTC.

Method for fabricating .sup.99 Mo production targets using low enriched uranium, .sup.99 Mo production targets comprising low enriched uranium

DOEpatents

Wiencek, Thomas C.; Matos, James E.; Hofman, Gerard L.

1997-01-01

A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.

Method for fabricating .sup.99 Mo production targets using low enriched uranium, .sup.99 Mo production targets comprising low enriched uranium

DOEpatents

Wiencek, Thomas C [Orland Park, IL; Matos, James E [Oak Park, IL; Hofman, Gerard L [Downers Grove, IL

2000-12-12

A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.

Effect of Co-Contaminants Uranium and Nitrate on Iodine Remediation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szecsody, James E.; Lee, Brady D.; Lawter, Amanda R.

The objective of this study is to evaluate the significance of co-contaminants on the migration and transformation of iodine species in the Hanford subsurface environment. These impacts are relevant because remedies that target individual contaminants like iodine, may not only impact the fate and transport of other contaminants in the subsurface, but also inhibit the effectiveness of a targeted remedy. For example, iodine (as iodate) co-precipitates with calcite, and has been identified as a potential remedy because it immobilizes iodine. Since uranium also co-precipitates with calcite in field sediments, the presence of uranium may also inhibit iodine co-precipitation. Another potentiallymore » significant impact from co-existing contaminants is iodine and nitrate. The presence of nitrate has been shown to promote biogeochemical reduction of iodate to iodide, thereby increasing iodine species subsurface mobility (as iodide exhibits less sorption). Hence, this study reports on both laboratory batch and column experiments that investigated a) the change in iodate uptake mass and rate of uptake into precipitating calcite due to the presence of differing amounts of uranium, b) the amount of change of the iodate bio-reduction rate due to the presence of differing nitrate concentrations, and c) whether nitrite can reduce iodate in the presence of microbes and/or minerals acting as catalysts.« less

Improved toughness of silicon carbide

NASA Technical Reports Server (NTRS)

Palm, J. A.

1976-01-01

Impact energy absorbing layers (EALs) comprised of partially densified silicon carbide were formed in situ on fully sinterable silicon carbide substrates. After final sintering, duplex silicon carbide structures resulted which were comprised of a fully sintered, high density silicon carbide substrate or core, overlayed with an EAL of partially sintered silicon carbide integrally bonded to its core member. Thermal cycling tests proved such structures to be moderately resistant to oxidation and highly resistant to thermal shock stresses. The strength of the developed structures in some cases exceeded but essentially it remained the same as the fully sintered silicon carbide without the EAL. Ballistic impact tests indicated that substantial improvements in the toughness of sintered silicon carbide were achieved by the use of the partially densified silicon carbide EALs.

The effects of carbide column to swelling potential and Atterberg limit on expansive soil with column to soil drainage

NASA Astrophysics Data System (ADS)

Muamar Rifa'i, Alfian; Setiawan, Bambang; Djarwanti, Noegroho

2017-12-01

The expansive soil is soil that has a potential for swelling-shrinking due to changes in water content. Such behavior can exert enough force on building above to cause damage. The use of columns filled with additives such as Calcium Carbide is done to reduce the negative impact of expansive soil behavior. This study aims to determine the effect of carbide columns on expansive soil. Observations were made on swelling and spreading of carbides in the soil. 7 Carbide columns with 5 cm diameter and 20 cm height were installed into the soil with an inter-column spacing of 8.75 cm. Wetting is done through a pipe at the center of the carbide column for 20 days. Observations were conducted on expansive soil without carbide columns and expansive soil with carbide columns. The results showed that the addition of carbide column could reduce the percentage of swelling by 4.42%. Wetting through the center of the carbide column can help spread the carbide into the soil. The use of carbide columns can also decrease the rate of soil expansivity. After the addition of carbide column, the plasticity index value decreased from 71.76% to 4.3% and the shrinkage index decreased from 95.72% to 9.2%.

Friction and wear behavior of single-crystal silicon carbide in contact with titanium

NASA Technical Reports Server (NTRS)

Miyoshi, K.; Buckley, D. H.

1977-01-01

Sliding friction experiments were conducted with single crystal silicon carbide in sliding contact with titanium. Results indicate that the friction coefficient is greater in vacuum than in argon and that this is due to the greater adhesion or adhesive transfer in vacuum. Thin films of silicon carbide transferred to titanium also adhered to silicon carbide both in argon at atmospheric pressure and in high vacuum. Cohesive bonds fractured on both the silicon carbide and titanium surfaces. The wear debris of silicon carbide created by fracture plowed the silicon carbide surface in a plastic manner. The friction characteristics of titanium in contact with silicon carbide were sensitive to the surface roughness of silicon carbide, and the friction coefficients were higher for a rough surface of silicon carbide than for a smooth one. The difference in friction results was due to plastic deformation (plowing of titanium).

Silicon carbide semiconductor technology for high temperature and radiation environments

NASA Technical Reports Server (NTRS)

Matus, Lawrence G.

1993-01-01

Viewgraphs on silicon carbide semiconductor technology and its potential for enabling electronic devices to function in high temperature and high radiation environments are presented. Topics covered include silicon carbide; sublimation growth of 6H-SiC boules; SiC chemical vapor deposition reaction system; 6H silicon carbide p-n junction diode; silicon carbide MOSFET; and silicon carbide JFET radiation response.

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunt, Rodney Dale; Johnson, Jared A.; Collins, Jack Lee

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC 2), which is UC 1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UCmore » 2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90–92% of TD with full conversion of UC to UC 2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC 2. Lastly, the selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.« less

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Johnson, J. A.; Collins, J. L.; McMurray, J. W.; Reif, T. J.; Brown, D. R.

2018-01-01

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC2), which is UC1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UC2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90-92% of TD with full conversion of UC to UC2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC2. The selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

DOE PAGES

Hunt, Rodney Dale; Johnson, Jared A.; Collins, Jack Lee; ...

2017-10-12

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC 2), which is UC 1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UCmore » 2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90–92% of TD with full conversion of UC to UC 2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC 2. Lastly, the selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.« less

Carbides Evolution in a Ni-16Mo-7Cr Base Superalloy during Long-Term Thermal Exposure

PubMed Central

Han, Fenfen; Jiang, Li; Ye, Xiangxi; Lu, Yanling; Li, Zhijun; Zhou, Xingtai

2017-01-01

The effect of long-term thermal exposure on the carbide evolution in a Ni-16Mo-7Cr base superalloy was investigated. The results show that M12C carbides are mainly precipitated on the grain boundaries during thermal exposure, and the primary massive M6C carbides can be completely transformed to M12C carbides in situ at temperatures above 750 °C for long-term thermal exposure. The transformation from M6C carbides to M12C carbides is attributed to the release of C atoms from M6C, which results in the morphology changes of massive carbides, and stabilization of the sizes of M12C carbides precipitated on the grain boundaries. PMID:28772881

Morphological Characteristics of Particulate Material Formed from High Velocity Impact of Depleted Uranium Projectiles with Armor Targets

DTIC Science & Technology

1978-11-01

Magnification Showing Aggregation of Ultrafine Particles ; Gap Between Bars Represents 0.5 pm. .......... ... 15 iv LIST OF FIGURES (CONCLUDED) Figure Title...subsequent forma- tion of smaller particulates. An unexpected phenomenon was the formation of ultrafine particles less than 0.1 pm in diameter. These...and the highly reactive nature of pyrophoric depleted uranium. Ti ese ultrafine particles exhibited an extreme tendency to coalesce, probably due to

Biomarkers for Uranium Risk Assessment for the Development of the CURE (Concerted Uranium Research in Europe) Molecular Epidemiological Protocol.

PubMed

Guéguen, Yann; Roy, Laurence; Hornhardt, Sabine; Badie, Christophe; Hall, Janet; Baatout, Sarah; Pernot, Eileen; Tomasek, Ladislav; Laurent, Olivier; Ebrahimian, Teni; Ibanez, Chrystelle; Grison, Stephane; Kabacik, Sylwia; Laurier, Dominique; Gomolka, Maria

2017-01-01

Despite substantial experimental and epidemiological research, there is limited knowledge of the uranium-induce health effects after chronic low-dose exposures in humans. Biological markers can objectively characterize pathological processes or environmental responses to uranium and confounding agents. The integration of such biological markers into a molecular epidemiological study would be a useful approach to improve and refine estimations of uranium-induced health risks. To initiate such a study, Concerted Uranium Research in Europe (CURE) was established, and involves biologists, epidemiologists and dosimetrists. The aims of the biological work package of CURE were: 1. To identify biomarkers and biological specimens relevant to uranium exposure; 2. To define standard operating procedures (SOPs); and 3. To set up a common protocol (logistic, questionnaire, ethical aspects) to perform a large-scale molecular epidemiologic study in uranium-exposed cohorts. An intensive literature review was performed and led to the identification of biomarkers related to: 1. retention organs (lungs, kidneys and bone); 2. other systems/organs with suspected effects (cardiovascular system, central nervous system and lympho-hematopoietic system); 3. target molecules (DNA damage, genomic instability); and 4. high-throughput methods for the identification of new biomarkers. To obtain high-quality biological materials, SOPs were established for the sampling and storage of different biospecimens. A questionnaire was developed to assess potential confounding factors. The proposed strategy can be adapted to other internal exposures and should improve the characterization of the biological and health effects that are relevant for risk assessment.

Simulation of uranium and plutonium oxides compounds obtained in plasma

NASA Astrophysics Data System (ADS)

Novoselov, Ivan Yu.; Karengin, Alexander G.; Babaev, Renat G.

2018-03-01

The aim of this paper is to carry out thermodynamic simulation of mixed plutonium and uranium oxides compounds obtained after plasma treatment of plutonium and uranium nitrates and to determine optimal water-salt-organic mixture composition as well as conditions for their plasma treatment (temperature, air mass fraction). Authors conclude that it needs to complete the treatment of nitric solutions in form of water-salt-organic mixtures to guarantee energy saving obtainment of oxide compounds for mixed-oxide fuel and explain the choice of chemical composition of water-salt-organic mixture. It has been confirmed that temperature of 1200 °C is optimal to practice the process. Authors have demonstrated that condensed products after plasma treatment of water-salt-organic mixture contains targeted products (uranium and plutonium oxides) and gaseous products are environmental friendly. In conclusion basic operational modes for practicing the process are showed.

Determination of uranium and thorium in materials associated with real time electronic solar neutrino detectors

NASA Astrophysics Data System (ADS)

Fassett, J. D.; Kelly, W. R.

1992-07-01

The application of isotope dilution thermal ionization mass spectrometry to the determination of both uranium and thorium in four different target materials used or proposed for electronic neutrino detectors is described. Isotope dilution analysis is done using highly enriched 233U and 230Th separated isotopes. Sensitivity of the technique is such that sub-picogram amounts of material are readily measured. The overall limit to measurement is caused by contamination of these elements during the measurement process. Uranium is more easily measured than thorium because both the instrumental sensitivity is higher and contamination is better controlled. The materials analyzed were light and heavy water, pseudocumene, and mineral oil.

Preliminary report on the use of LANDSAT-1 (ERTS-1) reflectance data in locating alteration zones associated with uranium mineralization near Cameron, Arizona

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spirakis, C.S.; Condit, C.D.

1975-01-01

LANDSAT-1 (ERTS-1) multispectral reflectance data were used to enhance the detection of alteration around uranium deposits near Cameron, Ariz. The technique involved stretching and ratioing computer-enhanced data from which electronic noise and atmospheric haze had been removed. Using present techniques, the work proves that LANDSAT-1 data are useful in detecting alteration around uranium deposits, but the method may still be improved. Bluish-gray mudstone in the target area could not be differentiated from the altered zones on the ratioed images. Further experiments involving combinations of ratioed and nonratioed data will be required to uniquely define the altered zones.

Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

2014-09-01

One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g (× 106 cm-1) the limits were exceeded.

Porous silicon carbide (SIC) semiconductor device

NASA Technical Reports Server (NTRS)

Shor, Joseph S. (Inventor); Kurtz, Anthony D. (Inventor)

1996-01-01

Porous silicon carbide is fabricated according to techniques which result in a significant portion of nanocrystallites within the material in a sub 10 nanometer regime. There is described techniques for passivating porous silicon carbide which result in the fabrication of optoelectronic devices which exhibit brighter blue luminescence and exhibit improved qualities. Based on certain of the techniques described porous silicon carbide is used as a sacrificial layer for the patterning of silicon carbide. Porous silicon carbide is then removed from the bulk substrate by oxidation and other methods. The techniques described employ a two-step process which is used to pattern bulk silicon carbide where selected areas of the wafer are then made porous and then the porous layer is subsequently removed. The process to form porous silicon carbide exhibits dopant selectivity and a two-step etching procedure is implemented for silicon carbide multilayers.

Formation of dysprosium carbide on the graphite (0001) surface

DOE PAGES

Lii-Rosales, Ann; Zhou, Yinghui; Wallingford, Mark; ...

2017-07-12

When using scanning tunneling microscopy, we characterize a surface carbide that forms such that Dy is deposited on the basal plane of graphite. In order to form carbide islands on terraces, Dy is first deposited at 650–800 K, which forms large metallic islands. Upon annealing at 1000 K, these clusters convert to carbide. Deposition directly at 1000 K is ineffective because nucleation on terraces is inhibited. Reaction is signaled by the fact that each carbide cluster is partially or totally surrounded by an etch pit. The etch pit is one carbon layer deep for most carbide clusters. Carbide clusters aremore » also identifiable by striations on their surfaces. Based on mass balance, and assuming that only the surface layer of carbon is involved in the reaction, the carbide has stoichiometry D y 2 C . This is Dy-rich compared with the most common bulk carbide Dy C 2 , which may reflect limited surface carbon transport to the carbide.« less

Determining the release of radionuclides from tank waste residual solids. FY2015 report

DOE Office of Scientific and Technical Information (OSTI.GOV)

King, William D.; Hobbs, David T.

Methodology development for pore water leaching studies has been continued to support Savannah River Site High Level Waste tank closure efforts. For FY2015, the primary goal of this testing was the achievement of target pH and Eh values for pore water solutions representative of local groundwater in the presence of grout or grout-representative (CaCO 3 or FeS) solids as well as waste surrogate solids representative of residual solids expected to be present in a closed tank. For oxidizing conditions representative of a closed tank after aging, a focus was placed on using solid phases believed to be controlling pH andmore » E h at equilibrium conditions. For three pore water conditions (shown below), the target pH values were achieved to within 0.5 pH units. Tank 18 residual surrogate solids leaching studies were conducted over an E h range of approximately 630 mV. Significantly higher Eh values were achieved for the oxidizing conditions (ORII and ORIII) than were previously observed. For the ORII condition, the target Eh value was nearly achieved (within 50 mV). However, E h values observed for the ORIII condition were approximately 160 mV less positive than the target. E h values observed for the RRII condition were approximately 370 mV less negative than the target. Achievement of more positive and more negative E h values is believed to require the addition of non-representative oxidants and reductants, respectively. Plutonium and uranium concentrations measured during Tank 18 residual surrogate solids leaching studies under these conditions (shown below) followed the general trends predicted for plutonium and uranium oxide phases, assuming equilibrium with dissolved oxygen. The highest plutonium and uranium concentrations were observed for the ORIII condition and the lowest concentrations were observed for the RRII condition. Based on these results, it is recommended that these test methodologies be used to conduct leaching studies with actual Tank 18 residual solids material. Actual waste testing will include leaching evaluations of technetium and neptunium, as well as plutonium and uranium.« less

Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

NASA Astrophysics Data System (ADS)

Grande, Lisa Christine

A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

Removal of uranium from soil sample digests for ICP-OES analysis of trace metals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Foust, R.D. Jr.; Bidabad, M.

1996-10-01

An analytical procedure has been developed to quantitatively remove uranium from soil sample digests, permitting ICP-OES analysis of trace metals. The procedure involves digesting a soil sample with standard procedures (EPA SW-846, Method 3050), and passing the sample digestate through commercially available resin (U/TEVA{sm_bullet}Spec, Eichrom Industries, Inc.) containing diarryl amylphosphonate as the stationary phase. Quantitative removal of uranium was achieved with soil samples containing up to 60% uranium, and percent recoveries averaged better than 85% for 9 of the 10 metals evaluated (Ag, As, Cd. Cr, Cu, Ni, Pb, Se and Tl). The U/TEVA{sm_bullet}Spec column was regenerated by washing withmore » 200 mL of a 0.01 M oxalic acid/0.02 M nitric acid solution, permitting re-use of the column. GFAAS analysis of a sample spiked with 56.5% uranium, after treatment of the digestate with a U/TEVA{sm_bullet}Spec resin column, resulted in percent recoveries of 97% or better for all target metals.« less

Uranium hydrogeochemical and stream sediment reconnaissance data release for the Red Creek quartzite special study area, Vernal NTMS Quadrangle, Utah/Colorado, including concentrations of forty-six additional elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, S.; George, W.E.; Apel, C.T.

1981-04-01

Totals of 22 water and 140 sediment samples were collected from 148 locations in the study area. The study area, in the north-central portion of the Vernal NTMS quadrangle, is covered by four 7-1/2' topographic maps: Dutch John, Goslin Mountain, and Clav Basin, Utah; and Willow Creek Butte, Utah/Colorado. Additional HSSR data are available for the entire Vernal quadrangle (Purson, 1980). All field and analytical data are presented in Appendix I. Figure 1 is an index and sample location map that can be used in conjunction with the 1:250,000-scale topographic map of the Vernal quadrangle (USGS, 1954). Standarized field, analytical,more » and data base management procedures were followed in all phases of the study. These procedures are described briefly in Appendix II-A and in reports by Sharp (1977), Hues et al (1977), Sharp and Aamodt (1978), Cheadle (1977), and Kosiewicz (1979). The data presented in Appendix I are available on magnetic tape from GJOIS Project, Union Carbide Corporation (UCC-ND), Computer Applications Department, 4500 North Building, Oak Ridge National Laboratory, P.O. Box X, Oak Ridge, Tennessee 37830. Because this is simply a data release, intended to make the data available to the DOE and the public as quickly as possible, no discussion of the geology of the region, uranium occurrences, or data evaluation is included.« less

Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrani, Kurt A; Kiggans Jr, James O; McMurray, Jake W

2016-01-01

Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhancemore » heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.« less

Method of preparing silicon carbide particles dispersed in an electrolytic bath for composite electroplating of metals

DOEpatents

Peng, Yu-Min; Wang, Jih-Wen; Liue, Chun-Ying; Yeh, Shinn-Horng

1994-01-01

A method for preparing silicon carbide particles dispersed in an electrolytic bath for composite electroplating of metals includes the steps of washing the silicon carbide particles with an organic solvent; washing the silicon carbide particles with an inorganic acid; grinding the silicon carbide particles; and heating the silicon carbide particles in a nickel-containing solution at a boiling temperature for a predetermined period of time.

Method of fabricating porous silicon carbide (SiC)

NASA Technical Reports Server (NTRS)

Shor, Joseph S. (Inventor); Kurtz, Anthony D. (Inventor)

1995-01-01

Porous silicon carbide is fabricated according to techniques which result in a significant portion of nanocrystallites within the material in a sub 10 nanometer regime. There is described techniques for passivating porous silicon carbide which result in the fabrication of optoelectronic devices which exhibit brighter blue luminescence and exhibit improved qualities. Based on certain of the techniques described porous silicon carbide is used as a sacrificial layer for the patterning of silicon carbide. Porous silicon carbide is then removed from the bulk substrate by oxidation and other methods. The techniques described employ a two-step process which is used to pattern bulk silicon carbide where selected areas of the wafer are then made porous and then the porous layer is subsequently removed. The process to form porous silicon carbide exhibits dopant selectivity and a two-step etching procedure is implemented for silicon carbide multilayers.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lii-Rosales, Ann; Zhou, Yinghui; Wallingford, Mark

When using scanning tunneling microscopy, we characterize a surface carbide that forms such that Dy is deposited on the basal plane of graphite. In order to form carbide islands on terraces, Dy is first deposited at 650–800 K, which forms large metallic islands. Upon annealing at 1000 K, these clusters convert to carbide. Deposition directly at 1000 K is ineffective because nucleation on terraces is inhibited. Reaction is signaled by the fact that each carbide cluster is partially or totally surrounded by an etch pit. The etch pit is one carbon layer deep for most carbide clusters. Carbide clusters aremore » also identifiable by striations on their surfaces. Based on mass balance, and assuming that only the surface layer of carbon is involved in the reaction, the carbide has stoichiometry D y 2 C . This is Dy-rich compared with the most common bulk carbide Dy C 2 , which may reflect limited surface carbon transport to the carbide.« less

Structural evolution of molybdenum carbides in hot aqueous environments and impact on low-temperature hydroprocessing of acetic acid

DOE PAGES

Choi, Jae -Soon; Schwartz, Viviane; Santillan-Jimenez, Eduardo; ...

2015-03-13

In this paper, we investigated the structural evolution of molybdenum carbides subjected to hot aqueous environments and their catalytic performance in low-temperature hydroprocessing of acetic acid. While bulk structures of Mo carbides were maintained after aging in hot liquid water, a portion of carbidic Mo sites were converted to oxidic sites. Water aging also induced changes to the non-carbidic carbon deposited during carbide synthesis and increased surface roughness, which in turn affected carbide pore volume and surface area. The extent of these structural changes was sensitive to the initial carbide structure and was lower under actual hydroprocessing conditions indicating themore » possibility of further improving the hydrothermal stability of Mo carbides by optimizing catalyst structure and operating conditions. Mo carbides were active in acetic acid conversion in the presence of liquid water, their activity being comparable to that of Ru/C. Finally, the results suggest that effective and inexpensive bio-oil hydroprocessing catalysts could be designed based on Mo carbides, although a more detailed understanding of the structure-performance relationships is needed, especially in upgrading of more complex reaction mixtures or real bio-oils.« less

Sandstone type uranium deposits in the Ordos Basin, Northwest China: A case study and an overview

NASA Astrophysics Data System (ADS)

Akhtar, Shamim; Yang, Xiaoyong; Pirajno, Franco

2017-09-01

This paper provides a comprehensive review on studies of sandstone type uranium deposits in the Ordos Basin, Northwest China. As the second largest sedimentary basin, the Ordos Basin has great potential for targeting sandstone type U mineralization. The newly found and explored Dongsheng and Diantou sandstone type uranium deposits are hosted in the Middle Jurassic Zhilou Formation. A large number of investigations have been conducted to trace the source rock compositions and relationship between lithic subarkose sandstone host rock and uranium mineralization. An optical microscopy study reveals two types of alteration associated with the U mineralization: chloritization and sericitization. Some unusual mineral structures, with compositional similarity to coffinite, have been identified in a secondary pyrite by SEM These mineral phases are proposed to be of bacterial origin, following high resolution mapping of uranium minerals and trace element determinations in situ. Moreover, geochemical studies of REE and trace elements constrained the mechanism of uranium enrichment, displaying LREE enrichment relative to HREE. Trace elements such as Pb, Mo and Ba have a direct relationship with uranium enrichment and can be used as index for mineralization. The source of uranium ore forming fluids and related geological processes have been studied using H, O and C isotope systematics of fluid inclusions in quartz veins and the calcite cement of sandstone rocks hosting U mineralization. Both H and O isotopic compositions of fluid inclusions reveal that ore forming fluids are a mixture of meteoric water and magmatic water. The C and S isotopes of the cementing material of sandstone suggest organic origin and bacterial sulfate reduction (BSR), providing an important clue for U mineralization. Discussion of the ore genesis shows that the greenish gray sandstone plays a crucial role during processes leading to uranium mineralization. Consequently, an oxidation-reduction model for sandstone-type uranium deposit is proposed, which can elucidate the source of uranium in the deposits of the Ordos Basin, based on the role of organic materials and sulfate reducing bacteria. We discuss the mechanism of uranium deposition responsible for the genesis of these large sandstone type uranium deposits in this unique sedimentary basin.

Three-dimensional inversion of magnetotelluric data for mineral exploration: An example from the McArthur River uranium deposit, Saskatchewan, Canada

NASA Astrophysics Data System (ADS)

Farquharson, Colin G.; Craven, James A.

2009-08-01

Shallow exploration targets are becoming scarce, meaning interest is turning towards deeper targets. The magnetotelluric method has the necessary depth capability, unlike many of the controlled-source electromagnetic prospecting techniques traditionally used. The geological setting of ore deposits is usually complex, requiring three-dimensional Earth models for their representation. An example of the applicability of three-dimensional inversion of magnetotelluric data to mineral exploration is presented here. Inversions of an audio-magnetotelluric data-set from the McArthur River uranium mine in the Athabasca Basin were carried out. A sub-set comprising data from eleven frequencies distributed over almost three decades was inverted. The form of the data used in the inversion was impedance. All four elements of the tensor were included. No decompositions of the data were done, nor rotation to a preferred strike direction, nor correction for static shifts. The inversions were successful: the observations were adequately reproduced and the main features in the conductivity model corresponded to known geological features. These included the graphitic basement fault along which the McArthur River uranium deposit is located.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dale, Gregory E.

There is currently a serious shortage of 99Mo, from which to generate the medically significant isotope 99mTc. Most of the world's supply comes from the fission of highly enriched uranium targets--this is a proliferation concern. This document focuses on the technology involved in two alternative methods: electron accelerator production of 99Mo from the 100Mo(γ,n) 99Mo reaction and production of 99Mo as a fission product in a subcritical, DT accelerator-driven low enriched uranium salt solution.

JPRS Report Science and Technology, Japan: Atomic Energy Society 1989 Annual Meeting.

DTIC Science & Technology

1989-10-13

Control Rod Hole in VHTRC-1 Core [F, Akino, T, Yamane, et al.] ,,, 5 Measurement of MEU [Medium Enriched Uranium ] Fuel Element Characteristics in...K. Yoshida, K. Kobayashi, I. Kimura , C. Yamanaka, and S. Nakai, Laser Laboratory,, Osaka University. Nuclear Reactor Laboratory, Kyoto University...1 core loaded with 278 fuel rods (4 percent enriched uranium ). The PNS target was placed at the back center of the 1/2 assembly on the fixed side

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tufic Madi Filho; Adonis Marcelo Saliba Silva; Jose Patricio Nahuel Cardenas

2015-07-01

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to createmore » samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)« less

Design and synthesis of target-responsive hydrogel for portable visual quantitative detection of uranium with a microfluidic distance-based readout device.

PubMed

Huang, Yishun; Fang, Luting; Zhu, Zhi; Ma, Yanli; Zhou, Leiji; Chen, Xi; Xu, Dunming; Yang, Chaoyong

2016-11-15

Due to uranium's increasing exploitation in nuclear energy and its toxicity to human health, it is of great significance to detect uranium contamination. In particular, development of a rapid, sensitive and portable method is important for personal health care for those who frequently come into contact with uranium ore mining or who investigate leaks at nuclear power plants. The most stable form of uranium in water is uranyl ion (UO2(2+)). In this work, a UO2(2+) responsive smart hydrogel was designed and synthesized for rapid, portable, sensitive detection of UO2(2+). A UO2(2+) dependent DNAzyme complex composed of substrate strand and enzyme strand was utilized to crosslink DNA-grafted polyacrylamide chains to form a DNA hydrogel. Colorimetric analysis was achieved by encapsulating gold nanoparticles (AuNPs) in the DNAzyme-crosslinked hydrogel to indicate the concentration of UO2(2+). Without UO2(2+), the enzyme strand is not active. The presence of UO2(2+) in the sample activates the enzyme strand and triggers the cleavage of the substrate strand from the enzyme strand, thereby decreasing the density of crosslinkers and destabilizing the hydrogel, which then releases the encapsulated AuNPs. As low as 100nM UO2(2+) was visually detected by the naked eye. The target-responsive hydrogel was also demonstrated to be applicable in natural water spiked with UO2(2+). Furthermore, to avoid the visual errors caused by naked eye observation, a previously developed volumetric bar-chart chip (V-Chip) was used to quantitatively detect UO2(2+) concentrations in water by encapsulating Au-Pt nanoparticles in the hydrogel. The UO2(2+) concentrations were visually quantified from the travelling distance of ink-bar on the V-Chip. The method can be used for portable and quantitative detection of uranium in field applications without skilled operators and sophisticated instruments. Copyright © 2016 Elsevier B.V. All rights reserved.

Reference binding energies of transition metal carbides by core-level x-ray photoelectron spectroscopy free from Ar+ etching artefacts

NASA Astrophysics Data System (ADS)

Greczynski, G.; Primetzhofer, D.; Hultman, L.

2018-04-01

We report x-ray photoelectron spectroscopy (XPS) core level binding energies (BE's) for the widely-applicable groups IVb-VIb transition metal carbides (TMCs) TiC, VC, CrC, ZrC, NbC, MoC, HfC, TaC, and WC. Thin film samples are grown in the same deposition system, by dc magnetron co-sputtering from graphite and respective elemental metal targets in Ar atmosphere. To remove surface contaminations resulting from exposure to air during sample transfer from the growth chamber into the XPS system, layers are either (i) Ar+ ion-etched or (ii) UHV-annealed in situ prior to XPS analyses. High resolution XPS spectra reveal that even gentle etching affects the shape of core level signals, as well as BE values, which are systematically offset by 0.2-0.5 eV towards lower BE. These destructive effects of Ar+ ion etch become more pronounced with increasing the metal atom mass due to an increasing carbon-to-metal sputter yield ratio. Systematic analysis reveals that for each row in the periodic table (3d, 4d, and 5d) C 1s BE increases from left to right indicative of a decreased charge transfer from TM to C atoms, hence bond weakening. Moreover, C 1s BE decreases linearly with increasing carbide/metal melting point ratio. Spectra reported here, acquired from a consistent set of samples in the same instrument, should serve as a reference for true deconvolution of complex XPS cases, including multinary carbides, nitrides, and carbonitrides.

Method for fabricating {sup 99}Mo production targets using low enriched uranium, {sup 99}Mo production targets comprising low enriched uranium

DOEpatents

Wiencek, T.C.; Matos, J.E.; Hofman, G.L.

1997-03-25

A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate. 3 figs.

Degradation of Silicon Carbide Reflective Surfaces in the LEO Environment

NASA Astrophysics Data System (ADS)

Mileti, Sandro; Coluzzi, Plinio; Marchetti, Mario

2009-01-01

Space mirrors in Low Earth Orbit (LEO) encounter a degradation problem caused by the impact of atomic oxygen (ATOX) in the space environment. This paper presents an experiment of the atomic oxygen impact degradation and UV synergic effects on ground simulation. The experiment was carried out in a dedicated ATOX simulation vacuum chamber. As target materials, a polished CVD Beta-silicon carbide (SiC) coating was investigated. The selection of silicon carbide is due to its high potential candidate as a mirror layer substrate material for its good reflectance at UV wavelengths and excellent thermal diffusivity. It has highly desirable mechanical and thermal properties and can achieve an excellent surface finish. The deposition of the coatings were on carbon-based material substrate; i.e., silicon impregnated carbon fiber composite (C/SiC). Mechanical and thermal properties of the coatings such as hardness and Coefficient of Thermal Expansion (CTE) were achieved. Several atomic oxygen impact angles were studied tilting the target samples respect to the flux direction. The various impact angles permitted to analyze the different erosion rates and typologies which the mirrors would encounter in LEO environment. The degradation was analyzed in various aspects. Macroscopic mass loss per unit area, surface roughness and morphology change were basically analyzed. The exposed surfaces of the materials were observed through a Scanning Electron Microscope (SEM). Secondly, optical diagnostic of the surfaces were performed in order to investigate their variation in optical properties as the evaluation of reflectance degradation. The presence of micro-cracks caused by shrinkage, grinding, polishing or thermal cycling and the porosity in the coatings, could have led to the undercutting phenomenon. Observation of uprising of undercutting was also conducted. Remarks are given regarding capabilities in short-term mission exposures to the LEO environment of this coating.

Growth characteristics of primary M7C3 carbide in hypereutectic Fe-Cr-C alloy.

PubMed

Liu, Sha; Zhou, Yefei; Xing, Xiaolei; Wang, Jibo; Ren, Xuejun; Yang, Qingxiang

2016-09-06

The microstructure of the hypereutectic Fe-Cr-C alloy is observed by optical microscopy (OM). The initial growth morphology, the crystallographic structure, the semi-molten morphology and the stacking faults of the primary M7C3 carbide are observed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The in-suit growth process of the primary M7C3 carbide was observed by confocal laser microscope (CLM). It is found that the primary M7C3 carbide in hypereutectic Fe-Cr-C alloy is irregular polygonal shape with several hollows in the center and gaps on the edge. Some primary M7C3 carbides are formed by layers of shell or/and consist of multiple parts. In the initial growth period, the primary M7C3 carbide forms protrusion parallel to {} crystal planes. The extending and revolving protrusion forms the carbide shell. The electron backscattered diffraction (EBSD) maps show that the primary M7C3 carbide consists of multiple parts. The semi-molten M7C3 carbide contains unmelted shell and several small-scale carbides inside, which further proves that the primary M7C3 carbide is not an overall block. It is believed that the coalescence of the primary M7C3 carbides is ascribed to the growing condition of the protrusion and the gap filling process.

Growth characteristics of primary M7C3 carbide in hypereutectic Fe-Cr-C alloy

PubMed Central

Liu, Sha; Zhou, Yefei; Xing, Xiaolei; Wang, Jibo; Ren, Xuejun; Yang, Qingxiang

2016-01-01

The microstructure of the hypereutectic Fe-Cr-C alloy is observed by optical microscopy (OM). The initial growth morphology, the crystallographic structure, the semi-molten morphology and the stacking faults of the primary M7C3 carbide are observed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The in-suit growth process of the primary M7C3 carbide was observed by confocal laser microscope (CLM). It is found that the primary M7C3 carbide in hypereutectic Fe-Cr-C alloy is irregular polygonal shape with several hollows in the center and gaps on the edge. Some primary M7C3 carbides are formed by layers of shell or/and consist of multiple parts. In the initial growth period, the primary M7C3 carbide forms protrusion parallel to {} crystal planes. The extending and revolving protrusion forms the carbide shell. The electron backscattered diffraction (EBSD) maps show that the primary M7C3 carbide consists of multiple parts. The semi-molten M7C3 carbide contains unmelted shell and several small-scale carbides inside, which further proves that the primary M7C3 carbide is not an overall block. It is believed that the coalescence of the primary M7C3 carbides is ascribed to the growing condition of the protrusion and the gap filling process. PMID:27596718

Microwave sintering of boron carbide

DOEpatents

Blake, R.D.; Katz, J.D.; Petrovic, J.J.; Sheinberg, H.

1988-06-10

A method for forming boron carbide into a particular shape and densifying the green boron carbide shape. Boron carbide in powder form is pressed into a green shape and then sintered, using a microwave oven, to obtain a dense boron carbide body. Densities of greater than 95% of theoretical density have been obtained. 1 tab.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Jae -Soon; Schwartz, Viviane; Santillan-Jimenez, Eduardo

In this paper, we investigated the structural evolution of molybdenum carbides subjected to hot aqueous environments and their catalytic performance in low-temperature hydroprocessing of acetic acid. While bulk structures of Mo carbides were maintained after aging in hot liquid water, a portion of carbidic Mo sites were converted to oxidic sites. Water aging also induced changes to the non-carbidic carbon deposited during carbide synthesis and increased surface roughness, which in turn affected carbide pore volume and surface area. The extent of these structural changes was sensitive to the initial carbide structure and was lower under actual hydroprocessing conditions indicating themore » possibility of further improving the hydrothermal stability of Mo carbides by optimizing catalyst structure and operating conditions. Mo carbides were active in acetic acid conversion in the presence of liquid water, their activity being comparable to that of Ru/C. Finally, the results suggest that effective and inexpensive bio-oil hydroprocessing catalysts could be designed based on Mo carbides, although a more detailed understanding of the structure-performance relationships is needed, especially in upgrading of more complex reaction mixtures or real bio-oils.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curtis, Michael M.

As a result of NSG restrictions, India cannot import the natural uranium required to fuel its Pressurized Heavy Water Reactors (PHWRs); consequently, it is forced to rely on the expediency of domestic uranium production. However, domestic production from mines and byproduct sources has not kept pace with demand from commercial reactors. This shortage has been officially confirmed by the Indian Planning Commission’s Mid-Term Appraisal of the country’s current Five Year Plan. The report stresses that as a result of the uranium shortage, Indian PHWR load factors have been continually decreasing. The Uranium Corporation of India Ltd (UCIL) operates a numbermore » of underground mines in the Singhbhum Shear Zone of Jharkhand, and it is all processed at a single mill in Jaduguda. UCIL is attempting to aggrandize operations by establishing new mines and mills in other states, but the requisite permit-gathering and development time will defer production until at least 2009. A significant portion of India’s uranium comes from byproduct sources, but a number of these are derived from accumulated stores that are nearing exhaustion. A current maximum estimate of indigenous uranium production is 430t/yr (230t from mines and 200t from byproduct sources); whereas, the current uranium requirement for Indian PHWRs is 455t/yr (depending on plant capacity factor). This deficit is exacerbated by the additional requirements of the Indian weapons program. Present power generation capacity of Indian nuclear plants is 4350 MWe. The power generation target set by the Indian Department of Atomic Energy (DAE) is 20,000 MWe by the year 2020. It is expected that around half of this total will be provided by PHWRs using indigenously supplied uranium with the bulk of the remainder provided by breeder reactors or pressurized water reactors using imported low-enriched uranium.« less

Process for preparing metal-carbide-containing microspheres from metal-loaded resin beads

DOEpatents

Beatty, Ronald L.

1976-01-01

An improved method for treating metal-loaded resin microspheres is described which comprises heating a metal-loaded resin charge in an inert atmosphere at a pre-carbide-forming temperature under such conditions as to produce a microsphere composition having sufficient carbon as to create a substantially continuous carbon matrix and a metal-carbide or an oxide-carbide mixture as a dispersed phase(s) during carbide-forming conditions, and then heating the thus treated charge to a carbide-forming temperature.

Methods of producing continuous boron carbide fibers

DOEpatents

Garnier, John E.; Griffith, George W.

2015-12-01

Methods of producing continuous boron carbide fibers. The method comprises reacting a continuous carbon fiber material and a boron oxide gas within a temperature range of from approximately 1400.degree. C. to approximately 2200.degree. C. Continuous boron carbide fibers, continuous fibers comprising boron carbide, and articles including at least a boron carbide coating are also disclosed.

Depleted uranium investigation at missile impact sites in White Sands Missile Range

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Etten, D.M.; Purtymun, W.D.

1994-01-01

An investigation for residual depleted uranium was conducted at Pershing missile impact sites on the White Sands Missile Range. Subsurface core soil samples were taken at Chess, Salt Target, and Mine Impact Sites. A sampling pump was installed in a monitoring well at Site 65 where a Pershing earth penetrator was not recovered. Pumping tests and water samples were taken at this site. Chess Site, located in a gypsum flat, was the only location showing elevated levels of depleted uranium in the subsurface soil or perched groundwater. Small fragments can still be found on the surface of the impact sites.more » The seasonal flooding and near surface water has aided in the movement of surface fragments.« less

Natural precursor based hydrothermal synthesis of sodium carbide for reactor applications

NASA Astrophysics Data System (ADS)

Swapna, M. S.; Saritha Devi, H. V.; Sebastian, Riya; Ambadas, G.; Sankararaman, S.

2017-12-01

Carbides are a class of materials with high mechanical strength and refractory nature which finds a wide range of applications in industries and nuclear reactors. The existing synthesis methods of all types of carbides have problems in terms of use of toxic chemical precursors, high-cost, etc. Sodium carbide (Na2C2) which is an alkali metal carbide is the least explored one and also that there is no report of low-cost and low-temperature synthesis of sodium carbide using the eco-friendly, easily available natural precursors. In the present work, we report a simple low-cost, non-toxic hydrothermal synthesis of refractory sodium carbide using the natural precursor—Pandanus. The formation of sodium carbide along with boron carbide is evidenced by the structural and morphological characterizations. The sample thus synthesized is subjected to field emission scanning electron microscopy (FESEM), x-ray powder diffraction (XRD), ultraviolet (UV)—visible spectroscopy, Fourier transform infrared spectroscopy (FTIR), Raman, and photoluminescent (PL) spectroscopic techniques.

Surface and Bulk Carbide Transformations in High-Speed Steel

PubMed Central

Godec, M.; Večko Pirtovšek, T.; Šetina Batič, B.; McGuiness, P.; Burja, J.; Podgornik, B.

2015-01-01

We have studied the transformation of carbides in AISI M42 high-speed steels in the temperature window used for forging. The annealing was found to result in the partial transformation of the large, metastable M2C carbides into small, more stable grains of M6C, with an associated change in the crystal orientation. In addition, MC carbides form during the transformation of M2C to M6C. From the high-speed-steel production point of view, it is beneficial to have large, metastable carbides in the cast structure, which later during annealing, before the forging, transform into a structure of polycrystalline carbides. Such carbides can be easily decomposed into several small carbides, which are then randomly distributed in the microstructure. The results also show an interesting difference in the carbide-transformation reactions on the surface versus the bulk of the alloy, which has implications for in-situ studies of bulk phenomena that are based on surface observations. PMID:26537780

Improved toughness of silicon carbide

NASA Technical Reports Server (NTRS)

Palm, J. A.

1975-01-01

Several techniques were employed to apply or otherwise form porous layers of various materials on the surface of hot-pressed silicon carbide ceramic. From mechanical properties measurements and studies, it was concluded that although porous layers could be applied to the silicon carbide ceramic, sufficient damage was done to the silicon carbide surface by the processing required so as to drastically reduce its mechanical strength. It was further concluded that there was little promise of success in forming an effective energy absorbing layer on the surface of already densified silicon carbide ceramic that would have the mechanical strength of the untreated or unsurfaced material. Using a process for the pressureless sintering of silicon carbide powders it was discovered that porous layers of silicon carbide could be formed on a dense, strong silicon carbide substrate in a single consolidation process.

Summary Report of Depleted Uranium (DU) Survey Actions at Nevada Test and Training Range (NTTR), Airspace Region 63B, Active Target Complex 10 (63-10)

DTIC Science & Technology

2016-09-15

currently valid OMB control number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 15 Sep 2016 2. REPORT TYPE...Consultative Letter 3. DATES COVERED (From – To) Feb – Jun 2013 4. TITLE AND SUBTITLE Summary Report of Depleted Uranium (DU) Survey Actions at...USAF RADIOISOTOPE COMMITTEE SECRETARIAT ATTN: DR. RAMACHANDRA BHAT 7700 ARLINGTON BLVD, STE 5151 FALLS CHURCH, VA 22042-5151

Sensitivity of geological, geochemical and hydrologic parameters in complex reactive transport systems for in-situ uranium bioremediation

NASA Astrophysics Data System (ADS)

Yang, G.; Maher, K.; Caers, J.

2015-12-01

Groundwater contamination associated with remediated uranium mill tailings is a challenging environmental problem, particularly within the Colorado River Basin. To examine the effectiveness of in-situ bioremediation of U(VI), acetate injection has been proposed and tested at the Rifle pilot site. There have been several geologic modeling and simulated contaminant transport investigations, to evaluate the potential outcomes of the process and identify crucial factors for successful uranium reduction. Ultimately, findings from these studies would contribute to accurate predictions of the efficacy of uranium reduction. However, all these previous studies have considered limited model complexities, either because of the concern that data is too sparse to resolve such complex systems or because some parameters are assumed to be less important. Such simplified initial modeling, however, limits the predictive power of the model. Moreover, previous studies have not yet focused on spatial heterogeneity of various modeling components and its impact on the spatial distribution of the immobilized uranium (U(IV)). In this study, we study the impact of uncertainty on 21 parameters on model responses by means of recently developed distance-based global sensitivity analysis (DGSA), to study the main effects and interactions of parameters of various types. The 21 parameters include, for example, spatial variability of initial uranium concentration, mean hydraulic conductivity, and variogram structures of hydraulic conductivity. DGSA allows for studying multi-variate model responses based on spatial and non-spatial model parameters. When calculating the distances between model responses, in addition to the overall uranium reduction efficacy, we also considered the spatial profiles of the immobilized uranium concentration as target response. Results show that the mean hydraulic conductivity and the mineral reaction rate are the two most sensitive parameters with regard to the overall uranium reduction. But in terms of spatial distribution of immobilized uranium, initial conditions of uranium concentration and spatial uncertainty in hydraulic conductivity also become important. These analyses serve as the first step of further prediction practices of the complex uranium transport and reaction systems.

Production method for making rare earth compounds

DOEpatents

McCallum, R.W.; Ellis, T.W.; Dennis, K.W.; Hofer, R.J.; Branagan, D.J.

1997-11-25

A method of making a rare earth compound, such as a earth-transition metal permanent magnet compound, without the need for producing rare earth metal as a process step, comprises carbothermically reacting a rare earth oxide to form a rare earth carbide and heating the rare earth carbide, a compound-forming reactant (e.g., a transition metal and optional boron), and a carbide-forming element (e.g., a refractory metal) that forms a carbide that is more thermodynamically favorable than the rare earth carbide whereby the rare earth compound (e.g., Nd{sub 2}Fe{sub 14}B or LaNi{sub 5}) and a carbide of the carbide-forming element are formed.

Production method for making rare earth compounds

DOEpatents

McCallum, R. William; Ellis, Timothy W.; Dennis, Kevin W.; Hofer, Robert J.; Branagan, Daniel J.

1997-11-25

A method of making a rare earth compound, such as a earth-transition metal permanent magnet compound, without the need for producing rare earth metal as a process step, comprises carbothermically reacting a rare earth oxide to form a rare earth carbide and heating the rare earth carbide, a compound-forming reactant (e.g. a transition metal and optional boron), and a carbide-forming element (e.g. a refractory metal) that forms a carbide that is more thermodynamically favorable than the rare earth carbide whereby the rare earth compound (e.g. Nd.sub.2 Fe.sub.14 B or LaNi.sub.5) and a carbide of the carbide-forming element are formed.

The efficiency of ceramic-faced metal targets at high-velocity impact

NASA Astrophysics Data System (ADS)

Tolkachev, V. F.; Konyaev, A. A.; Pakhnutova, N. V.

2017-11-01

The paper represents experimental results and engineering evaluation concerning the efficiency of composite materials to be used as an additional protection during the high- velocity interaction of a tungsten rod with a target in the velocity range of 1...5 km/s. The main parameter that characterizes the high-velocity interaction of a projectile with a layered target is the penetration depth. Experimental data, numerical simulation and engineering evaluation by modified models are used to determine the penetration depth. Boron carbide, aluminum oxide, and aluminum nickelide are applied as a front surface of targets. Based on experimental data and numerical simulation, the main characteristics of ceramics are determined, which allows composite materials to be effectively used as additional elements of protection.

Reactive Fusion Welding for Ultra-High Temperature Ceramic Composite Joining

DTIC Science & Technology

2015-03-16

Titanium diboride TiC-Titanium carbide C-Carbon SiC - Silicon carbide B4C-Boron carbide 67 W-Tungsten WC-Tungsten carbide ZrB2-20ZrC-ZrB2...ceramics with a nominal carbide content of 20 vol% were prepared. Starting powders were mechanically mixed by ball milling ZrB2 (H.C. Starck; Grade B...0.50 wt%, or ~1.5 vol%. Milling was carried out in acetone for 2 hours using tungsten carbide media. After milling, the powder slurry was dried

A Systematic Theoretical Study of UC6: Structure, Bonding Nature, and Spectroscopy.

PubMed

Du, Jiguang; Jiang, Gang

2017-11-20

The study of uranium carbides has received renewed attention in recent years due to the potential use of these compounds as fuels in new generations of nuclear reactors. The isomers of the UC 6 cluster were determined by DFT and ab initio methods. The structures obtained using SC-RECP for U were generally consistent with those obtained using an all-electron basis set (ZORA-SARC). The CCSD(T) calculations indicated that two isomers had similar energies and may coexist in laser evaporation experiments. The nature of the U-C bonds in the different isomers was examined via a topological analysis of the electron density, and the results indicated that the U-C bonds are predominantly closed-shell (ionic) interactions with a certain degree of covalent character in all cases, particularly in the linear species. The IR and UV-vis spectra of the isomers were theoretically simulated to provide information that can be used to identify the isomers of UC 6 in future experiments.

High pressure phase transformation in uranium carbide: A first principle study

NASA Astrophysics Data System (ADS)

Sahoo, B. D.; Joshi, K. D.; Gupta, Satish C.

2013-02-01

First principles calculations have been carried out to analyze structural, elastic and dynamic stability, of UC under hydrostatic compression. The comparison of enthalpies of rocksalt type (B1) and body centered orthorhombic (bco) structures as a function of pressure suggests the B1 →bco transition at ˜ 23 GPa, in good agreement with experimental value of 27 GPa. From the lattice dynamic calculations we have determined the phonon dispersion relations for B1 phase at various compressions. It is found that TA phonon branch along Γ-X direction becomes imaginary around the transition pressure. Further, the phonon instability so caused is of long wavelength nature as it occurs near the Brillouin zone centre. This long wavelength phonon instability at the transition point indicates that the B1 →bco transition is driven by elastic failure (the vanishing of C44 modulus). Various physical quantities such as equilibrium volume, bulk modulus, pressure derivative of bulk modulus and elastic constants have been determined at zero pressure and compared with data available in literature.

Preparation and crystal structure of U3Fe2C5: An original uranium-iron carbide

NASA Astrophysics Data System (ADS)

Henriques, M. S.; Paixão, J. A.; Henriques, M. S. C.; Gonçalves, A. P.

2015-09-01

The U3Fe2C5 compound was prepared from the elements by arc-melting, followed by an heat-treatment in an induction furnace, at 1250 °C for 1 h and 1300 °C for 2 h. The crystal structure of this phase was determined by direct methods from single crystal X-ray diffraction data. U3Fe2C5 crystallizes in an original tetragonal crystal structure, with space group I4/mmm, a = 3.4980(3) Å and c = 19.8380(15) Å as lattice constants and two formula units per cell. This new type structure is characterized by the simultaneous presence of isolated and pairs of carbon atoms, the interatomic distances in the pairs being similar to a typical carbon-carbon double bond length found in a molecule. U3Fe2C5 is closely related to UC and UFeC2, and can be seen as build from two (distorted) UFeC2 unit cells and a UC layer.

Analyzing the impact of reactive transport on the repository performance of TRISO fuel

NASA Astrophysics Data System (ADS)

Schmidt, Gregory

One of the largest determiners of the amount of electricity generated by current nuclear reactors is the efficiency of the thermodynamic cycle used for power generation. Current light water reactors (LWR) have an efficiency of 35% or less for the conversion of heat energy generated by the reactor to electrical energy. If this efficiency could be improved, more power could be generated from equivalent volumes of nuclear fuel. One method of improving this efficiency is to use a coolant flow that operates at a much higher temperature for electricity production. A reactor design that is currently proposed to take advantage of this efficiency is a graphite-moderated, helium-cooled reactor known as a High Temperature Gas Reactor (HTGR). There are significant differences between current LWR's and the proposed HTGR's but most especially in the composition of the nuclear fuel. For LWR's, the fuel elements consist of pellets of uranium dioxide or plutonium dioxide that are placed in long tubes made of zirconium metal alloys. For HTGR's, the fuel, known as TRISO (TRIstructural-ISOtropic) fuel, consists of an inner sphere of fissile material, a layer of dense pyrolytic carbon (PyC), a ceramic layer of silicon carbide (SiC) and a final dense outer layer of PyC. These TRISO particles are then compacted with graphite into fuel rods that are then placed in channels in graphite blocks. The blocks are then arranged in an annular fashion to form a reactor core. However, this new fuel form has unanswered questions on the environmental post-burn-up behavior. The key question for current once-through fuel operations is how these large irradiated graphite blocks with spent fuel inside will behave in a repository environment. Data in the literature to answer this question is lacking, but nevertheless this is an important question that must be answered before wide-spread adoption of HTGR's could be considered. This research has focused on answering the question of how the large quantity of graphite surrounding the spent HTGR fuel will impact the release of aqueous uranium from the TRISO fuel. In order to answer this question, the sorption and partitioning behavior of uranium to graphite under a variety of conditions was investigated. Key systematic variables that were analyzed include solution pH, dissolved carbonate concentration, uranium metal concentration and ionic strength. The kinetics and desorption characteristics of uranium/graphite partitioning were studied as well. The graphite used in these experiments was also characterized by a variety of techniques and conclusions are drawn about the relevant surface chemistry of graphite. This data was then used to generate a model for the reactive transport of uranium in a graphite matrix. This model was implemented with the software code CXTFIT and validated through the use of column studies mirroring the predicted system.

Methods of repairing a substrate

NASA Technical Reports Server (NTRS)

Riedell, James A. (Inventor); Easler, Timothy E. (Inventor)

2011-01-01

A precursor of a ceramic adhesive suitable for use in a vacuum, thermal, and microgravity environment. The precursor of the ceramic adhesive includes a silicon-based, preceramic polymer and at least one ceramic powder selected from the group consisting of aluminum oxide, aluminum nitride, boron carbide, boron oxide, boron nitride, hafnium boride, hafnium carbide, hafnium oxide, lithium aluminate, molybdenum silicide, niobium carbide, niobium nitride, silicon boride, silicon carbide, silicon oxide, silicon nitride, tin oxide, tantalum boride, tantalum carbide, tantalum oxide, tantalum nitride, titanium boride, titanium carbide, titanium oxide, titanium nitride, yttrium oxide, zirconium boride, zirconium carbide, zirconium oxide, and zirconium silicate. Methods of forming the ceramic adhesive and of repairing a substrate in a vacuum and microgravity environment are also disclosed, as is a substrate repaired with the ceramic adhesive.

Temperature Stabilization of the NIFFTE Time Projection Chamber

NASA Astrophysics Data System (ADS)

Hicks, Caleb

2017-09-01

The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) is a collaboration measuring nuclear fission cross sections for use in advanced nuclear reactors. A neutron beam incident on targets of Uranium-235, Uranium-238, and Plutonium-239 is used to measure the neutron induced fission cross sections for these isotopes. A Time Projection Chamber (TPC) is used to record these reactions. Significant heat is generated by the readout cards mounted on the TPC, which are cooled by fans. One proposed measurement of the experiment is to compare the cross sections of the target to a proton target of gaseous hydrogen. A constant temperature inside the TPC's pressure vessel is desirable to maintain a constant number of hydrogen target atoms. In addition, a constant temperature minimizes the strain and wrinkles on an amplifying mesh inside the TPC. This poster describes the successful work to develop, build, and install a fan controller using a Raspberry Pi, an Arduino, and a custom circuit board to implement an algorithm called Proportional-Integral-Derivative control. This research was supported by US DOE MENP Grant DE-FG02-03ER41243.

40 CFR 415.30 - Applicability; description of the calcium carbide production subcategory.

Code of Federal Regulations, 2010 CFR

2010-07-01

... calcium carbide production subcategory. 415.30 Section 415.30 Protection of Environment ENVIRONMENTAL... SOURCE CATEGORY Calcium Carbide Production Subcategory § 415.30 Applicability; description of the calcium... the production of calcium carbide in uncovered furnaces. ...

40 CFR 415.30 - Applicability; description of the calcium carbide production subcategory.

Code of Federal Regulations, 2011 CFR

2011-07-01

... calcium carbide production subcategory. 415.30 Section 415.30 Protection of Environment ENVIRONMENTAL... SOURCE CATEGORY Calcium Carbide Production Subcategory § 415.30 Applicability; description of the calcium... the production of calcium carbide in uncovered furnaces. ...

40 CFR 415.30 - Applicability; description of the calcium carbide production subcategory.

Code of Federal Regulations, 2012 CFR

2012-07-01

... calcium carbide production subcategory. 415.30 Section 415.30 Protection of Environment ENVIRONMENTAL... SOURCE CATEGORY Calcium Carbide Production Subcategory § 415.30 Applicability; description of the calcium... the production of calcium carbide in uncovered furnaces. ...

40 CFR 415.30 - Applicability; description of the calcium carbide production subcategory.

Code of Federal Regulations, 2014 CFR

2014-07-01

... calcium carbide production subcategory. 415.30 Section 415.30 Protection of Environment ENVIRONMENTAL... SOURCE CATEGORY Calcium Carbide Production Subcategory § 415.30 Applicability; description of the calcium... the production of calcium carbide in uncovered furnaces. ...

40 CFR 415.30 - Applicability; description of the calcium carbide production subcategory.

Code of Federal Regulations, 2013 CFR

2013-07-01

... calcium carbide production subcategory. 415.30 Section 415.30 Protection of Environment ENVIRONMENTAL... SOURCE CATEGORY Calcium Carbide Production Subcategory § 415.30 Applicability; description of the calcium... the production of calcium carbide in uncovered furnaces. ...

Characterization of the carbides and the martensite phase in powder-metallurgy high-speed steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Godec, Matjaz, E-mail: matjaz.godec@imt.si; Batic, Barbara Setina; Mandrino, Djordje

2010-04-15

A microstructural characterization of the powder-metallurgy high-speed-steel S390 Microclean was performed based on an elemental distribution of the carbide phase as well as crystallographic analyses. The results showed that there were two types of carbides present: vanadium-rich carbides, which were not chemically homogeneous and exhibited a tungsten-enriched or tungsten-depleted central area; and chemically homogeneous tungsten-rich M{sub 6}C-type carbides. Despite the possibility of chemical inhomogenities, the crystallographic orientation of each of the carbides was shown to be uniform. Using electron backscatter diffraction the vanadium-rich carbides were determined to be either cubic VC or hexagonal V{sub 6}C{sub 5}, while the tungsten-rich carbidesmore » were M{sub 6}C. The electron backscatter diffraction results were also verified using X-ray diffraction. Several electron backscatter diffraction pattern maps were acquired in order to define the fraction of each carbide phase as well as the amount of martensite phase. The fraction of martensite was estimated using band-contrast images, while the fraction of carbides was calculated using the crystallographic data.« less

Three-dimensional studies of intergranular carbides in austenitic stainless steel.

PubMed

Ochi, Minoru; Kawano, Rika; Maeda, Takuya; Sato, Yukio; Teranishi, Ryo; Hara, Toru; Kikuchi, Masao; Kaneko, Kenji

2017-04-01

A large number of morphological studies of intergranular carbides in steels have always been carried out in two dimensions without considering their dispersion manners. In this article, focused ion beam serial-sectioning tomography was carried out to study the correlation among the grain boundary characteristics, the morphologies and the dispersions of intergranular carbides in 347 austenitic stainless steel. More than hundred intergranular carbides were characterized in three dimensions and finally classified into three different types, two types of carbides probably semi-coherent to one of the neighboring grains with plate-type morphology, and one type of carbides incoherent to both grains with rod-type morphology. In addition, the rod-type carbide was found as the largest number of carbides among three types. Since large numbers of defects, such as misfit dislocations, may be present at the grain boundaries, which can be ideal nucleation sites for intergranular rod-type carbide precipitation. © The Author 2016. Published by Oxford University Press on behalf of The Japanese Society of Microscopy. All rights reserved.For permissions, please e-mail: journals.permissions@oup.com.

HEU Holdup Measurements on 321-M A-Lathe

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dewberry, R.A.

The Analytical Development Section of SRTC was requested by the Facilities Disposition Division (FDD) of the Savannah River Site to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the solid waste Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. Three measurement systems were used to determine highly enrichedmore » uranium (HEU) holdup. This report covers holdup measurements on the A-Lathe that was used to machine uranium-aluminum-alloy (U-Al). Our results indicated that the lathe contained more than the limits stated in the Waste Acceptance Criteria (WAC) for the solid waste E-Area Vaults. Thus the lathe was decontaminated three times and assayed four times in order to bring the amounts of uranium to an acceptable content. This report will discuss the methodology, Non-Destructive Assay (NDA) measurements, and results of the U-235 holdup on the lathe.« less

Deposition method for producing silicon carbide high-temperature semiconductors

DOEpatents

Hsu, George C.; Rohatgi, Naresh K.

1987-01-01

An improved deposition method for producing silicon carbide high-temperature semiconductor material comprising placing a semiconductor substrate composed of silicon carbide in a fluidized bed silicon carbide deposition reactor, fluidizing the bed particles by hydrogen gas in a mildly bubbling mode through a gas distributor and heating the substrate at temperatures around 1200.degree.-1500.degree. C. thereby depositing a layer of silicon carbide on the semiconductor substrate.

MC Carbide Characterization in High Refractory Content Powder-Processed Ni-Based Superalloys

NASA Astrophysics Data System (ADS)

Antonov, Stoichko; Chen, Wei; Huo, Jiajie; Feng, Qiang; Isheim, Dieter; Seidman, David N.; Sun, Eugene; Tin, Sammy

2018-04-01

Carbide precipitates in Ni-based superalloys are considered to be desirable phases that can contribute to improving high-temperature properties as well as aid in microstructural refinement of the material; however, they can also serve as crack initiation sites during fatigue. To date, most of the knowledge pertaining to carbide formation has originated from assessments of cast and wrought Ni-based superalloys. As powder-processed Ni-based superalloys are becoming increasingly widespread, understanding the different mechanisms by which they form becomes increasingly important. Detailed characterization of MC carbides present in two experimental high Nb-content powder-processed Ni-based superalloys revealed that Hf additions affect the resultant carbide morphologies. This morphology difference was attributed to a higher magnitude of elastic strain energy along the interface associated with Hf being soluble in the MC carbide lattice. The composition of the MC carbides was studied through atom probe tomography and consisted of a complex carbonitride core, which was rich in Nb and with slight Hf segregation, surrounded by an Nb carbide shell. The characterization results of the segregation behavior of Hf in the MC carbides and the subsequent influence on their morphology were compared to density functional theory calculations and found to be in good agreement, suggesting that computational modeling can successfully be used to tailor carbide features.

Tribological properties of sintered polycrystalline and single crystal silicon carbide

NASA Technical Reports Server (NTRS)

Miyoshi, K.; Buckley, D. H.; Srinivasan, M.

1982-01-01

Tribological studies and X-ray photoelectron spectroscopy analyses were conducted with sintered polycrystalline and single crystal silicon carbide surfaces in sliding contact with iron at various temperatures to 1500 C in a vacuum of 30 nPa. The results indicate that there is a significant temperature influence on both the friction properties and the surface chemistry of silicon carbide. The main contaminants on the as received sintered polycrystalline silicon carbide surfaces are adsorbed carbon, oxygen, graphite, and silicon dioxide. The surface revealed a low coefficient of friction. This is due to the presence of the graphite on the surface. At temperatures of 400 to 600 C graphite and copious amount of silicon dioxide were observed on the polycrystalline silicon carbide surface in addition to silicon carbide. At 800 C, the amount of the silicon dioxide decreased rapidly and the silicon carbide type silicon and carbon peaks were at a maximum intensity in the XPS spectra. The coefficients of friction were high in the temperature range 400 to 800 C. Small amounts of carbon and oxygen contaminants were observed on the as received single crystal silicon carbide surface below 250 C. Silicon carbide type silicon and carbon peaks were seen on the silicon carbide in addition to very small amount of graphite and silicon dioxide at temperatures of 450 to 800 C.

MC Carbide Characterization in High Refractory Content Powder-Processed Ni-Based Superalloys

NASA Astrophysics Data System (ADS)

Antonov, Stoichko; Chen, Wei; Huo, Jiajie; Feng, Qiang; Isheim, Dieter; Seidman, David N.; Sun, Eugene; Tin, Sammy

2018-06-01

Carbide precipitates in Ni-based superalloys are considered to be desirable phases that can contribute to improving high-temperature properties as well as aid in microstructural refinement of the material; however, they can also serve as crack initiation sites during fatigue. To date, most of the knowledge pertaining to carbide formation has originated from assessments of cast and wrought Ni-based superalloys. As powder-processed Ni-based superalloys are becoming increasingly widespread, understanding the different mechanisms by which they form becomes increasingly important. Detailed characterization of MC carbides present in two experimental high Nb-content powder-processed Ni-based superalloys revealed that Hf additions affect the resultant carbide morphologies. This morphology difference was attributed to a higher magnitude of elastic strain energy along the interface associated with Hf being soluble in the MC carbide lattice. The composition of the MC carbides was studied through atom probe tomography and consisted of a complex carbonitride core, which was rich in Nb and with slight Hf segregation, surrounded by an Nb carbide shell. The characterization results of the segregation behavior of Hf in the MC carbides and the subsequent influence on their morphology were compared to density functional theory calculations and found to be in good agreement, suggesting that computational modeling can successfully be used to tailor carbide features.

Uranium isotope separation from 1941 to the present

NASA Astrophysics Data System (ADS)

Maier-Komor, Peter

2010-02-01

Uranium isotope separation was the key development for the preparation of highly enriched isotopes in general and thus became the seed for target development and preparation for nuclear and applied physics. In 1941 (year of birth of the author) large-scale development for uranium isotope separation was started after the US authorities were warned that NAZI Germany had started its program for enrichment of uranium and might have confiscated all uranium and uranium mines in their sphere of influence. Within the framework of the Manhattan Projects the first electromagnetic mass separators (Calutrons) were installed and further developed for high throughput. The military aim of the Navy Department was to develop nuclear propulsion for submarines with practically unlimited range. Parallel to this the army worked on the development of the atomic bomb. Also in 1941 plutonium was discovered and the production of 239Pu was included into the atomic bomb program. 235U enrichment starting with natural uranium was performed in two steps with different techniques of mass separation in Oak Ridge. The first step was gas diffusion which was limited to low enrichment. The second step for high enrichment was performed with electromagnetic mass spectrometers (Calutrons). The theory for the much more effective enrichment with centrifugal separation was developed also during the Second World War, but technical problems e.g. development of high speed ball and needle bearings could not be solved before the end of the war. Spying accelerated the development of uranium separation in the Soviet Union, but also later in China, India, Pakistan, Iran and Iraq. In this paper, the physical and chemical procedures are outlined which lead to the success of the project. Some security aspects and Non-Proliferation measures are discussed.

Integrated modeling/analyses of thermal-shock effects in SNS targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taleyarkhan, R.P.; Haines, J.

1996-06-01

In a spallation neutron source (SNS), extremely rapid energy pulses are introduced in target materials such as mercury, lead, tungsten, uranium, etc. Shock phenomena in such systems may possibly lead to structural material damage beyond the design basis. As expected, the progression of shock waves and interaction with surrounding materials for liquid targets can be quite different from that in solid targets. The purpose of this paper is to describe ORNL`s modeling framework for `integrated` assessment of thermal-shock issues in liquid and solid target designs. This modeling framework is being developed based upon expertise developed from past reactor safety studies,more » especially those related to the Advanced Neutron Source (ANS) Project. Unlike previous separate-effects modeling approaches employed (for evaluating target behavior when subjected to thermal shocks), the present approach treats the overall problem in a coupled manner using state-of-the-art equations of state for materials of interest (viz., mercury, tungsten and uranium). That is, the modeling framework simultaneously accounts for localized (and distributed) compression pressure pulse generation due to transient heat deposition, the transport of this shock wave outwards, interaction with surrounding boundaries, feedback to mercury from structures, multi-dimensional reflection patterns & stress induced (possible) breakup or fracture.« less

Carbide coated fibers in graphite-aluminum composites

NASA Technical Reports Server (NTRS)

Imprescia, R. J.; Levinson, L. S.; Reiswig, R. D.; Wallace, T. C.; Williams, J. M.

1975-01-01

The study of protective-coupling layers of refractory metal carbides on the graphite fibers prior to their incorporation into composites is presented. Such layers should be directly wettable by liquid aluminum and should act as diffusion barriers to prevent the formation of aluminum carbide. Chemical vapor deposition was used to uniformly deposit thin, smooth, continuous coats of ZrC on the carbon fibers of tows derived from both rayon and polyacrylonitrile. A wet chemical coating of the fibers, followed by high-temperature treatment, was used, and showed promise as an alternative coating method. Experiments were performed to demonstrate the ability of aluminum alloys to wet carbide surfaces. Titanium carbide, zirconium carbide and carbide-coated graphite surfaces were successfully wetted. Results indicate that initial attempts to wet surfaces of ZrC-coated carbon fibers appear successful.

An aerosol particle containing enriched uranium encountered during routine sampling

NASA Astrophysics Data System (ADS)

Murphy, Daniel; Froyd, Karl; Evangeliou, NIkolaos; Stohl, Andreas

2017-04-01

The composition of single aerosol particles has been measured using a laser ionization mass spectrometer during the global Atmospheric Tomography mission. The measurements were targeting the background atmosphere, not radiochemical emissions. One sub-micron particle sampled at about 7 km altitude near the Aleutian Islands contained uranium with approximately 3% 235U. It is the only particle with enriched uranium out of millions of particles sampled over several decades of measurements with this instrument. The particle also contained vanadium, alkali metals, and organic material similar to that present in emissions from combustion of heavy oil. No zirconium or other metals that might be characteristic of nuclear reactors were present, probably suggesting a source other than Fukushima or Chernobyl. Back trajectories suggest several areas in Asia that might be sources for the particle.

Preparation and electrocatalytic activity of tungsten carbide and titania nanocomposite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Sujuan; Shi, Binbin; Yao, Guoxing

2011-10-15

Graphical abstract: The electrocatalytic activity of tungsten carbide and titania nanocomposite is related to the structure, crystal phase and chemical components of the nanocomposite, and is also affected by the property of electrolyte. A synergistic effect exists between tungsten carbide and titania of the composite. Highlights: {yields} Electrocatalytic activity of tungsten carbide and titania nanocomposite with core-shell structure. {yields} Activity is related to the structure, crystal phase and chemical component of the nanocomposite. {yields} The property of electrolyte affects the electrocatalytic activity. {yields} A synergistic effect exists between tungsten carbide and titania of the composite. -- Abstract: Tungsten carbide andmore » titania nanocomposite was prepared by combining a reduced-carbonized approach with a mechanochemical approach. The samples were characterized by X-ray diffraction, transmission electron microscope under scanning mode and X-ray energy dispersion spectrum. The results show that the crystal phases of the samples are composed of anatase, rutile, nonstoichiometry titanium oxide, monotungsten carbide, bitungsten carbide and nonstoichiometry tungsten carbide, and they can be controlled by adjusting the parameters of the reduced-carbonized approach; tungsten carbide particles decorate on the surface of titania support, the diameter of tungsten carbide particle is smaller than 20 nm and that of titania is around 100 nm; the chemical components of the samples are Ti, O, W and C. The electrocatalytic activity of the samples was measured by a cyclic voltammetry with three electrodes. The results indicate that the electrocatalytic activities of the samples are related to their crystal phases and the property of electrolyte in aqueous solution. A synergistic effect between titania and tungsten carbide is reported for the first time.« less

VapC toxins drive cellular dormancy under uranium stress for the extreme thermoacidophile Metallosphaera prunae.

PubMed

Mukherjee, Arpan; Wheaton, Garrett H; Counts, James A; Ijeomah, Brenda; Desai, Jigar; Kelly, Robert M

2017-07-01

When abruptly exposed to toxic levels of hexavalent uranium, the extremely thermoacidophilic archaeon Metallosphaera prunae, originally isolated from an abandoned uranium mine, ceased to grow, and concomitantly exhibited heightened levels of cytosolic ribonuclease activity that corresponded to substantial degradation of cellular RNA. The M. prunae transcriptome during 'uranium-shock' implicated VapC toxins as possible causative agents of the observed RNA degradation. Identifiable VapC toxins and PIN-domain proteins encoded in the M. prunae genome were produced and characterized, three of which (VapC4, VapC7, VapC8) substantially degraded M. prunae rRNA in vitro. RNA cleavage specificity for these VapCs mapped to motifs within M. prunae rRNA. Furthermore, based on frequency of cleavage sequences, putative target mRNAs for these VapCs were identified; these were closely associated with translation, transcription, and replication. It is interesting to note that Metallosphaera sedula, a member of the same genus and which has a nearly identical genome sequence but not isolated from a uranium-rich biotope, showed no evidence of dormancy when exposed to this metal. M. prunae utilizes VapC toxins for post-transcriptional regulation under uranium stress to enter a cellular dormant state, thereby providing an adaptive response to what would otherwise be a deleterious environmental perturbation. © 2017 Society for Applied Microbiology and John Wiley & Sons Ltd.

Experimental studies of the medical radioisotopes production in neutron spectra generated by 660 MeV protons and 1-8 GeV deuterons in massive uranium target

NASA Astrophysics Data System (ADS)

Zhadan, A.; Sotnikov, V.; Adam, J.; Solnyshkin, A.; Tyutyunnikov, S.; Voronko, V.; Zhivkov, P.; Zavorka, L.

2017-06-01

The possibility of medical radionuclide 64,67Cu production in spallation neutron spectrum induced by proton and deuteron beams has been studied. Experiments were performed on a massive natural uranium target at the accelerators Phasotron and Nuclotron JINR, Dubna. The main disadvantage of this method is a high 64Cu/67Cu ratio in the final product at EOB. Significantly reduce 64Cu/67Cu ratio is only possible if you use zinc target enriched with 68Zn or 67Zn. The MCNPX simulation of 67,64Cu production and definition of the theoretical limit of the specific activity of 67,64Cu by irradiation of natural zinc and zinc enriched by the 68 isotope were performed. The neutron flux density shouldnot be less than 5.1013 n/cm2/s if we want to obtain high specific activity (>200 GBq/mg) of 67Cu.

Fertile-to-fissile and fission measurements for depleted uranium bombarded by 800-MeV protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, G.J.; Gilmore, J.S.; Robinson, H.

Axial distributions of fertile-to-fissile conversions (/sup 238/U to /sup 239/Pu) and fissions have been measured for a thick depleted uranium target bombarded by 800-MeV protons. The /sup 239/Pu production was determined by measuring the amount of /sup 239/Np produced. The axial distributions were integrated to get the total conversions and fissions occurring in the target. Preliminary experimental results give 3.81 +- 0.19 /sup 239/Np atoms produced per incident proton and 5.59 +- 0.56 fissions per incident proton. Corresponding calculated results are 3.46 +- 0.05 and 3.93 +- 0.06. The computations did not include the effects of high-energy fission competition withmore » evaporation. Measured axial disributions of /sup 237/U and eleven fission products produced in the target are reported. Preliminary experimental data give 0.95 +- 0.05 /sup 237/U atoms made per incident proton.« less

Monte carlo simulations of Yttrium reaction rates in Quinta uranium target

NASA Astrophysics Data System (ADS)

Suchopár, M.; Wagner, V.; Svoboda, O.; Vrzalová, J.; Chudoba, P.; Tichý, P.; Kugler, A.; Adam, J.; Závorka, L.; Baldin, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Solnyshkin, A.; Tsoupko-Sitnikov, V.; Tyutyunnikov, S.; Bielewicz, M.; Kilim, S.; Strugalska-Gola, E.; Szuta, M.

2017-03-01

The international collaboration Energy and Transmutation of Radioactive Waste (E&T RAW) performed intensive studies of several simple accelerator-driven system (ADS) setups consisting of lead, uranium and graphite which were irradiated by relativistic proton and deuteron beams in the past years at the Joint Institute for Nuclear Research (JINR) in Dubna, Russia. The most recent setup called Quinta, consisting of natural uranium target-blanket and lead shielding, was irradiated by deuteron beams in the energy range between 1 and 8 GeV in three accelerator runs at JINR Nuclotron in 2011 and 2012 with yttrium samples among others inserted inside the setup to measure the neutron flux in various places. Suitable activation detectors serve as one of possible tools for monitoring of proton and deuteron beams and for measurements of neutron field distribution in ADS studies. Yttrium is one of such suitable materials for monitoring of high energy neutrons. Various threshold reactions can be observed in yttrium samples. The yields of isotopes produced in the samples were determined using the activation method. Monte Carlo simulations of the reaction rates leading to production of different isotopes were performed in the MCNPX transport code and compared with the experimental results obtained from the yttrium samples.

Comparison of Homogeneous and Heterogeneous CFD Fuel Models for Phase I of the IAEA CRP on HTR Uncertainties Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom; Su-Jong Yoon

2014-04-01

Computational Fluid Dynamics (CFD) evaluation of homogeneous and heterogeneous fuel models was performed as part of the Phase I calculations of the International Atomic Energy Agency (IAEA) Coordinate Research Program (CRP) on High Temperature Reactor (HTR) Uncertainties in Modeling (UAM). This study was focused on the nominal localized stand-alone fuel thermal response, as defined in Ex. I-3 and I-4 of the HTR UAM. The aim of the stand-alone thermal unit-cell simulation is to isolate the effect of material and boundary input uncertainties on a very simplified problem, before propagation of these uncertainties are performed in subsequent coupled neutronics/thermal fluids phasesmore » on the benchmark. In many of the previous studies for high temperature gas cooled reactors, the volume-averaged homogeneous mixture model of a single fuel compact has been applied. In the homogeneous model, the Tristructural Isotropic (TRISO) fuel particles in the fuel compact were not modeled directly and an effective thermal conductivity was employed for the thermo-physical properties of the fuel compact. On the contrary, in the heterogeneous model, the uranium carbide (UCO), inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers of the TRISO fuel particles are explicitly modeled. The fuel compact is modeled as a heterogeneous mixture of TRISO fuel kernels embedded in H-451 matrix graphite. In this study, a steady-state and transient CFD simulations were performed with both homogeneous and heterogeneous models to compare the thermal characteristics. The nominal values of the input parameters are used for this CFD analysis. In a future study, the effects of input uncertainties in the material properties and boundary parameters will be investigated and reported.« less

40 CFR 424.50 - Applicability; description of the other calcium carbide furnaces subcategory.

Code of Federal Regulations, 2010 CFR

2010-07-01

... calcium carbide furnaces subcategory. 424.50 Section 424.50 Protection of Environment ENVIRONMENTAL... CATEGORY Other Calcium Carbide Furnaces Subcategory § 424.50 Applicability; description of the other calcium carbide furnaces subcategory. The provisions of this subpart are applicable to discharges...

40 CFR 424.50 - Applicability; description of the other calcium carbide furnaces subcategory.

Code of Federal Regulations, 2011 CFR

2011-07-01

... calcium carbide furnaces subcategory. 424.50 Section 424.50 Protection of Environment ENVIRONMENTAL... CATEGORY Other Calcium Carbide Furnaces Subcategory § 424.50 Applicability; description of the other calcium carbide furnaces subcategory. The provisions of this subpart are applicable to discharges...

40 CFR 424.50 - Applicability; description of the other calcium carbide furnaces subcategory.

Code of Federal Regulations, 2014 CFR

2014-07-01

... calcium carbide furnaces subcategory. 424.50 Section 424.50 Protection of Environment ENVIRONMENTAL... CATEGORY Other Calcium Carbide Furnaces Subcategory § 424.50 Applicability; description of the other calcium carbide furnaces subcategory. The provisions of this subpart are applicable to discharges...

40 CFR 424.50 - Applicability; description of the other calcium carbide furnaces subcategory.

Code of Federal Regulations, 2013 CFR

2013-07-01

... calcium carbide furnaces subcategory. 424.50 Section 424.50 Protection of Environment ENVIRONMENTAL... CATEGORY Other Calcium Carbide Furnaces Subcategory § 424.50 Applicability; description of the other calcium carbide furnaces subcategory. The provisions of this subpart are applicable to discharges...

Method for homogenizing alloys susceptible to the formation of carbide stringers and alloys prepared thereby

DOEpatents

Braski, David N.; Leitnaker, James M.

1980-01-01

A novel fabrication procedure prevents or eliminates the reprecipitation of segregated metal carbides such as stringers in Ti-modified Hastelloy N and stainless steels to provide a novel alloy having carbides uniformly dispersed throughout the matrix. The fabrication procedure is applicable to other alloys prone to the formation of carbide stringers. The process comprises first annealing the alloy at a temperature above the single phase temperature for sufficient time to completely dissolve carbides and then annealing the single phase alloy for an additional time to prevent the formation of carbide stringers upon subsequent aging or thermomechanical treatment.

Anisotropic Tribological Properties of Silicon Carbide

NASA Technical Reports Server (NTRS)

Miyoshi, K.; Buckley, D. H.

1980-01-01

The anisotropic friction, deformation and fracture behavior of single crystal silicon carbide surfaces were investigated in two categories. The categories were called adhesive and abrasive wear processes, respectively. In the adhesive wear process, the adhesion, friction and wear of silicon carbide were markedly dependent on crystallographic orientation. The force to reestablish the shearing fracture of adhesive bond at the interface between silicon carbide and metal was the lowest in the preferred orientation of silicon carbide slip system. The fracturing of silicon carbide occurred near the adhesive bond to metal and it was due to primary cleavages of both prismatic (10(-1)0) and basal (0001) planes.

40 CFR 424.50 - Applicability; description of the other calcium carbide furnaces subcategory.

Code of Federal Regulations, 2012 CFR

2012-07-01

... other calcium carbide furnaces subcategory. 424.50 Section 424.50 Protection of Environment... SOURCE CATEGORY Other Calcium Carbide Furnaces Subcategory § 424.50 Applicability; description of the other calcium carbide furnaces subcategory. The provisions of this subpart are applicable to discharges...

Cobalt Doping of Semiconducting Boron Carbide Using Cobaltocene

DTIC Science & Technology

2007-03-01

COBALT DOPING OF SEMICONDUCTING BORON CARBIDE USING COBALTOCENE THESIS Lonnie Carlson, Major...DOPING OF SEMICONDUCTING BORON CARBIDE USING COBALTOCENE THESIS Presented to the Faculty Department of Engineering Physics Graduate School...DISTRIBUTION UNLIMITED AFIT/GNE/ENP/07-01 COBALT DOPING OF SEMICONDUCTING BORON CARBIDE USING COBALTOCENE Lonnie

10 CFR 110.42 - Export licensing criteria.

Code of Federal Regulations, 2012 CFR

2012-01-01

... research on or development of any nuclear explosive device. (3) Adequate physical security measures will be... to exports of high-enriched uranium to be used as a fuel or target in a nuclear research or test... can be used in the reactor. (iii) A fuel or target “can be used” in a nuclear research or test reactor...

Role of microstructure in caustic stress corrosion cracking of Alloy 690

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mertz, D.A.; Duda, P.T.; Pica, P.N.

1995-12-31

Alloy 690 has been selected for nuclear heat transport system tubing application in recent commercial reactor plants due to its resistance to multiple types of corrosion attack. Typical corn final heat treatments for this material are a mill-anneal (MA, approximately 1,070 C) to completely dissolve the carbides and develop the final grain structure plus a thermal treatment (TT, approximately 700 C) to precipitate carbides at the grain boundaries. Tubing with grain boundary carbides and no or few intragranular carbides has been found resistant to intergranular stress corrosion cracking (IGSCC) in caustic environments. In this work, first, Alloy 690 plate wasmore » subjected to a variety of MA and MA-TT heat treatments to create microstructures of carbide-decorated grain boundaries and undecorated boundaries. Caustic IGSCC test results were consistent with tubing data. Second, experiments were conducted to understand the mechanism by which caustic-corrosion resistance is imparted to Alloy 690 by grain boundary carbides. Tubing with a fully-developed MA-TT carbide microstructure was strained and heat-treated to create a mixed microstructure of new grain boundaries with no carbide precipitate decoration, intermixed with intragranular carbide strings from prior grain boundaries. Caustic SCC performance of this material was identical to that of material with the MA-TT carbide-decorated grain boundaries. This work suggests that the fundamental cause of good IGSCC resistance of MA-TT Alloy 690 in caustic does not derive solely from grain boundary carbides. It is suggested that matrix strength, as measured by yield stress, could be a controlling factor.« less

Effect of Spheroidizing Annealing on Microstructure and Mechanical Properties of High-Carbon Martensitic Stainless Steel 8Cr13MoV

NASA Astrophysics Data System (ADS)

Yu, Wen-Tao; Li, Jing; Shi, Cheng-Bin; Zhu, Qin-Tian

2017-02-01

The effects of holding time during both austenitizing and spheroidizing on microstructure and mechanical properties of high-carbon martensitic stainless steel 8Cr13MoV were experimentally studied. The results showed that the amount of carbides and the proportion of fine carbides decrease first and then increase with the increase in austenitizing time ( t 1) in the case of short spheroidizing time ( t 2), whereas the amount of the lamellar carbides increases. In the case of long t 2, both the amount of carbides and the proportion of fine carbides decrease, and the amount of the lamellar carbides did not increase. The hardness of the steel decreases first and then increases with the increase of t 1. Under the conditions of different t 1, the change in the size of carbides and hardness of the steel show a same trend with the variation of t 2. The size of spheroidized carbides increases, whereas the hardness of the steel decreases with increasing t 2. The longer the holding time of austenitizing, the higher is the spheroidizing rate at the earlier stage. However, the spheroidizing rate shows an opposite trend with t 1 at the later stage of spheroidizing. The effect of cooling rate on microstructure is similar with t 2. With increasing cooling rate, the dimension of carbides became smaller, and the amount of lamellar carbides increased. The elongation of the sample fracture exhibits no corresponding relationship with holding time, whereas it is closely related to the precipitation of secondary carbides caused by the alloying elements segregation.

Methods for producing silicon carbide fibers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garnier, John E.; Griffith, George W.

Methods of producing silicon carbide fibers. The method comprises reacting a continuous carbon fiber material and a silicon-containing gas in a reaction chamber at a temperature ranging from approximately 1500.degree. C. to approximately 2000.degree. C. A partial pressure of oxygen in the reaction chamber is maintained at less than approximately 1.01.times.10.sup.2 Pascal to produce continuous alpha silicon carbide fibers. Continuous alpha silicon carbide fibers and articles formed from the continuous alpha silicon carbide fibers are also disclosed.

Tunable carbon nanotube-tungsten carbide nanoparticles heterostructures by vapor deposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xia, Min; Guo, Hongyan; Ge, Changchun

2014-05-14

A simple, versatile route for the synthesis of carbon nanotube (CNT)-tungsten carbide nanoparticles heterostructures was set up via vapor deposition process. For the first time, amorphous CNTs (α-CNTs) were used to immobilized tungsten carbide nanoparticles. By adjusting the synthesis and annealing temperature, α-CNTs/amorphous tungsten carbide, α-CNTs/W{sub 2}C, and CNTs/W{sub 2}C/WC heterostructures were prepared. This approach provides an efficient method to attach other metal carbides and other nanoparticles to carbon nanotubes with tunable properties.

Silicon carbide fibers and articles including same

DOEpatents

Garnier, John E; Griffith, George W

2015-01-27

Methods of producing silicon carbide fibers. The method comprises reacting a continuous carbon fiber material and a silicon-containing gas in a reaction chamber at a temperature ranging from approximately 1500.degree. C. to approximately 2000.degree. C. A partial pressure of oxygen in the reaction chamber is maintained at less than approximately 1.01.times.10.sup.2 Pascal to produce continuous alpha silicon carbide fibers. Continuous alpha silicon carbide fibers and articles formed from the continuous alpha silicon carbide fibers are also disclosed.

Method of joining ceramics

DOEpatents

Henager, Jr., Charles H.; Brimhall, John L.

2000-01-01

According to the method of the present invention, joining a first bi-element carbide to a second bi-element carbide, has the steps of: (a) forming a bond agent containing a metal carbide and silicon; (b) placing the bond agent between the first and second bi-element carbides to form a pre-assembly; and (c) pressing and heating the pre-assembly in a non-oxidizing atmosphere to a temperature effective to induce a displacement reaction creating a metal silicon phase bonding the first and second bi-element carbides.

Low-temperature electrical resistivity of transition-metal carbides

NASA Astrophysics Data System (ADS)

Allison, C. Y.; Finch, C. B.; Foegelle, M. D.; Modine, F. A.

1988-10-01

The electrical resistivities of single crystals of ZrC 0.93, VC 0.88, NbC 0.95, and TaC 0.99 were measured from liquid helium temperature to 350 K. The Bloch-Gruneisen theory of electrical resistivity gives a good fit to the zirconium carbide and the vanadium carbide measurements. In contrast, the resistivities of the two superconducting crystals, tantalum carbide and niobium carbide, show excellent agreement with the Wilson model. The appropriate model appears to depend upon the superconducting properties of the crystals.

Dilatometry Analysis of Dissolution of Cr-Rich Carbides in Martensitic Stainless Steels

NASA Astrophysics Data System (ADS)

Huang, Qiuliang; Volkova, Olena; Biermann, Horst; Mola, Javad

2017-12-01

The dissolution of Cr-rich carbides formed in the martensitic constituent of a 13 pct Cr stainless steel was studied by dilatometry and correlative electron channeling contrast examinations. The dissolution of carbides subsequent to the martensite reversion to austenite was associated with a net volume expansion which in turn increased the dilatometry-based apparent coefficient of thermal expansion (CTEa) during continuous heating. The effects of carbides fraction and size on the CTEa variations during carbides dissolution are discussed.

(Reaction mechanism studies of heavy ion induced nuclear reactions): Annual progress report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mignerey, A.C.

1988-10-01

A major experiment was performed at the Oak Ridge National Laboratory Holifield Heavy Ion Research Facility in January 1988. The primary goal of the experiment was to determine the excitation energy division in the initial stages of damped reactions. The reaction of /sup 35/Cl on /sup 209/Bi was chosen because the excited projectile-like fragments would preferentially emit light charged particles and the target-like fragments deexcite via neutron emission. This provides a means by which projectile excitations can be selected over target excitations through detection of light charged particles in coincidence with projectile-like fragments. Two experiments were performed during the pastmore » year at the Lawrence Berkeley Laboratory Bevalac in collaboration with the Wozniak-Moretto group. The first was in February 1988 and was a continuation of earlier work on La-induced reactions at intermediate energies. Beams of La with E/A = 80 and 100 MeV were used to bombard targets of C, Al, and Cu. At this time a test run was also performed using the uranium beam to see if the intensity was sufficient to use this very heavy beam for future experiments. The high intensities obtained for uranium showed that it was feasible to extend the studies of inverse reactions begun with the lanthanum beam to a heavier beam. Gold rather than uranium was chosen for our major run in August due to its low fission probability and higher beam intensity. No results are yet available for that experiment.« less

NUCLEAR FUEL MATERIAL

DOEpatents

Goeddel, W.V.

1962-06-26

An improved method is given for making the carbides of nuclear fuel material. The metal of the fuel material, which may be a fissile and/or fertile material, is transformed into a silicide, after which the silicide is comminuted to the desired particle size. This silicide is then carburized at an elevated temperature, either above or below the melting point of the silicide, to produce an intimate mixture of the carbide of the fuel material and the carbide of silicon. This mixture of the fuel material carbide and the silicon carbide is relatively stable in the presence of moisture and does not exhibit the highly reactive surface condition which is observed with fuel material carbides made by most other known methods. (AEC)

Boron containing multilayer coatings and method of fabrication

DOEpatents

Makowiecki, D.M.; Jankowski, A.F.

1997-09-23

Hard coatings are fabricated from multilayer boron/boron carbide, boron carbide/cubic boron nitride, and boron/boron nitride/boron carbide, and the fabrication thereof involves magnetron sputtering in a selected atmosphere. These hard coatings may be applied to tools and engine and other parts, as well to reduce wear on tribological surfaces and electronic devices. These boron coatings contain no morphological growth features. For example, the boron and boron carbide used in forming the multilayers are formed in an inert (e.g. argon) atmosphere, while the cubic boron nitride is formed in a reactive (e.g. nitrogen) atmosphere. The multilayer boron/boron carbide, and boron carbide/cubic boron nitride is produced by depositing alternate layers of boron, cubic boron nitride or boron carbide, with the alternate layers having a thickness of 1 nanometer to 1 micrometer, and at least the interfaces of the layers may be of a discrete or a blended or graded composition. 6 figs.

Boron containing multilayer coatings and method of fabrication

DOEpatents

Makowiecki, Daniel M.; Jankowski, Alan F.

1997-01-01

Hard coatings are fabricated from multilayer boron/boron carbide, boron carbide/cubic boron nitride, and boron/boron nitride/boron carbide, and the fabrication thereof involves magnetron sputtering in a selected atmosphere. These hard coatings may be applied to tools and engine and other parts, as well to reduce wear on tribological surfaces and electronic devices. These boron coatings contain no morphological growth features. For example, the boron and boron carbide used in forming the multilayers are formed in an inert (e.g. argon) atmosphere, while the cubic boron nitride is formed in a reactive (e.g. nitrogen) atmosphere. The multilayer boron/boron carbide, and boron carbide/cubic boron nitride is produced by depositing alternate layers of boron, cubic boron nitride or boron carbide, with the alternate layers having a thickness of 1 nanometer to 1 micrometer, and at least the interfaces of the layers may be of a discrete or a blended or graded composition.

Ceramic material suitable for repair of a space vehicle component in a microgravity and vacuum environment, method of making same, and method of repairing a space vehicle component

NASA Technical Reports Server (NTRS)

Riedell, James A. (Inventor); Easler, Timothy E. (Inventor)

2009-01-01

A precursor of a ceramic adhesive suitable for use in a vacuum, thermal, and microgravity environment. The precursor of the ceramic adhesive includes a silicon-based, preceramic polymer and at least one ceramic powder selected from the group consisting of aluminum oxide, aluminum nitride, boron carbide, boron oxide, boron nitride, hafnium boride, hafnium carbide, hafnium oxide, lithium aluminate, molybdenum silicide, niobium carbide, niobium nitride, silicon boride, silicon carbide, silicon oxide, silicon nitride, tin oxide, tantalum boride, tantalum carbide, tantalum oxide, tantalum nitride, titanium boride, titanium carbide, titanium oxide, titanium nitride, yttrium oxide, zirconium diboride, zirconium carbide, zirconium oxide, and zirconium silicate. Methods of forming the ceramic adhesive and of repairing a substrate in a vacuum and microgravity environment are also disclosed, as is a substrate repaired with the ceramic adhesive.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu Yong; Department of Materials Science and Engineering, University of Tennessee, Knoxville, TN 37996; Liu Fengxiao

Cemented carbides with a functionally graded structure have significantly improved mechanical properties and lifetimes in cutting, drilling and molding. In this work, WC-6 wt.% Co cemented carbides with three-layer graded structure (surface layer rich in WC, mid layer rich in Co and the inner part of the average composition) were prepared by carburizing pre-sintered {eta}-phase-containing cemented carbides. The three-point bending fatigue tests based on the total-life approach were conducted on both WC-6wt%Co functionally graded cemented carbides (FGCC) and conventional WC-6wt%Co cemented carbides. The functionally graded cemented carbide shows a slightly higher fatigue limit ({approx}100 MPa) than the conventional ones undermore » the present testing conditions. However, the fatigue crack nucleation behavior of FGCC is different from that of the conventional ones. The crack nucleates preferentially along the Co-gradient and perpendicular to the tension surface in FGCC, while parallel to the tension surface in conventional cemented carbides.« less

Novel hard compositions and methods of preparation

DOEpatents

Sheinberg, Haskell

1983-08-23

Novel very hard compositions of matter are prepared by using in all embodiments only a minor amount of a particular carbide (or materials which can form the carbide in situ when subjected to heat and pressure); and no strategic cobalt is needed. Under a particular range of conditions, densified compositions of matter of the invention are prepared having hardnesses on the Rockwell A test substantially equal to the hardness of pure tungsten carbide and to two of the hardest commercial cobalt-bonded tungsten carbides. Alternately, other compositions of the invention which have slightly lower hardnesses than those described above in one embodiment also possess the advantage of requiring no tungsten and in another embodiment possess the advantage of having a good fracture toughness value. Photomicrographs show that the shapes of the grains of the alloy mixture with which the minor amount of carbide (or carbide-formers) is mixed are radically altered from large, rounded to small, very angular by the addition of the carbide. Superiority of one of these hard compositions of matter over cobalt-bonded tungsten carbide for ultra-high pressure anvil applications was demonstrated.

Carbide Precipitation in 2.25 Cr-1 Mo Bainitic Steel: Effect of Heating and Isothermal Tempering Conditions

NASA Astrophysics Data System (ADS)

Dépinoy, Sylvain; Toffolon-Masclet, Caroline; Urvoy, Stéphane; Roubaud, Justine; Marini, Bernard; Roch, François; Kozeschnik, Ernst; Gourgues-Lorenzon, Anne-Françoise

2017-05-01

The effect of the tempering heat treatment, including heating prior to the isothermal step, on carbide precipitation has been determined in a 2.25 Cr-1 Mo bainitic steel for thick-walled applications. The carbides were identified using their amount of metallic elements, morphology, nucleation sites, and diffraction patterns. The evolution of carbide phase fraction, morphology, and composition was investigated using transmission electron microscopy, X-ray diffraction, as well as thermodynamic calculations. Upon heating, retained austenite into the as-quenched material decomposes into ferrite and cementite. M7C3 carbides then nucleate at the interface between the cementite and the matrix, triggering the dissolution of cementite. M2C carbides precipitate separately within the bainitic laths during slow heating. M23C6 carbides precipitate at the interfaces (lath boundaries or prior austenite grain boundaries) and grow by attracting nearby chromium atoms, which results in the dissolution of M7C3 and, depending on the temperature, coarsening, or dissolution of M2C carbides, respectively.

Novel hard compositions and methods of preparation

DOEpatents

Sheinberg, H.

1983-08-23

Novel very hard compositions of matter are prepared by using in all embodiments only a minor amount of a particular carbide (or materials which can form the carbide in situ when subjected to heat and pressure); and no strategic cobalt is needed. Under a particular range of conditions, densified compositions of matter of the invention are prepared having hardnesses on the Rockwell A test substantially equal to the hardness of pure tungsten carbide and to two of the hardest commercial cobalt-bonded tungsten carbides. Alternately, other compositions of the invention which have slightly lower hardnesses than those described above in one embodiment also possess the advantage of requiring no tungsten and in another embodiment possess the advantage of having a good fracture toughness value. Photomicrographs show that the shapes of the grains of the alloy mixture with which the minor amount of carbide (or carbide-formers) is mixed are radically altered from large, rounded to small, very angular by the addition of the carbide. Superiority of one of these hard compositions of matter over cobalt-bonded tungsten carbide for ultra-high pressure anvil applications was demonstrated. 3 figs.

Effect of electroslag remelting on carbides in 8Cr13MoV martensitic stainless steel

NASA Astrophysics Data System (ADS)

Zhu, Qin-tian; Li, Jing; Shi, Cheng-bin; Yu, Wen-tao

2015-11-01

The effect of electroslag remelting (ESR) on carbides in 8Cr13MoV martensitic stainless steel was experimentally studied. Phases precipitated from liquid steel during solidification were calculated using the Thermo-Calc software. The carbon segregation was analyzed by original position analysis (OPA), and the carbides were analyzed by optical microscopy (OM), scanning electron microscopy (SEM), energy- dispersive X-ray spectroscopy (EDS) and X-ray diffraction (XRD). The results indicated that more uniform carbon distribution and less segregation were obtained in the case of samples subjected to the ESR process. After ESR, the amount of netty carbides decreased significantly, and the chromium and vanadium contents in the grain-boundary carbides was reduced. The total area and average size of carbides were obviously smaller after the ESR process. In the sample subjected to ESR, the morphology of carbides changed from lamellar and angular to globular or lump, whereas the types of carbides did not change; both M23C6 and M7C3 were present before and after the ESR process.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Hongfen, E-mail: wanghongfen11@163.com; Wang, Zhiqi; Chen, Shougang

Molybdenum carbides with surfactants as carbon sources were prepared using the carbothermal reduction of the appropriate precursors (molybdenum oxides deposited on surfactant micelles) at 1023 K under hydrogen gas. The carburized products were characterized using scanning electron microscopy (SEM), X-ray diffraction and BET surface area measurements. From the SEM images, hollow microspherical and rod-like molybdenum carbides were observed. X-ray diffraction patterns showed that the annealing time of carburization had a large effect on the conversion of molybdenum oxides to molybdenum carbides. And BET surface area measurements indicated that the difference of carbon sources brought a big difference in specific surfacemore » areas of molybdenum carbides. - Graphical abstract: Molybdenum carbides having hollow microspherical and hollow rod-like morphologies that are different from the conventional monodipersed platelet-like morphologies. Highlights: Black-Right-Pointing-Pointer Molybdenum carbides were prepared using surfactants as carbon sources. Black-Right-Pointing-Pointer The kinds of surfactants affected the morphologies of molybdenum carbides. Black-Right-Pointing-Pointer The time of heat preservation at 1023 K affected the carburization process. Black-Right-Pointing-Pointer Molybdenum carbides with hollow structures had larger specific surface areas.« less

2005 22nd International Symposium on Ballistics Volume 2 Wednesday

DTIC Science & Technology

2005-11-18

Information 1 Experimental and Numerical Study of the Penetration of Tungsten Carbide Into Steel Targets During High Rates of Strain John F . Moxnes...QinetiQ; Vladimir Titarev, Eleuterio Toro , Umeritek Limited The Mechanism Analysis of Interior Ballistics of Serial Chamber Gun, Dr. Sanjiu Ying, Charge...Elements and Meshless Particles, Gordon R. Johnson and Robert A. Stryk, Network Computing Services, Inc. Experimental and Numerical Study of the

Oxide or carbide nanoparticles synthesized by laser ablation of a bulk Hf target in liquids and their structural, optical, and dielectric properties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Semaltianos, N. G., E-mail: nsemaltianos@yahoo.com; Friedt, J.-M.; Blondeau-Patissier, V.

2016-05-28

Laser ablation of a bulk Hf target in deionized (DI) water, ethanol, or toluene was carried out for the production of nanoparticles' colloidal solutions. Due to the interaction of the ablation plasma plume species with the species which are produced by the liquid decomposition at the plume-liquid interface, hafnia (HfO{sub 2}) nanoparticles are synthesized in DI water, hafnium carbide (HfC) nanoparticles in toluene, and a mixture of these in ethanol. The hafnia nanoparticles are in the monoclinic low temperature phase and in the tetragonal and fcc high temperature phases. Their size distribution follows log-normal function with a median diameter inmore » the range of 4.3–5.3 nm. Nanoparticles synthesized in DI water have band gaps of 5.6 and 5.4 eV, in ethanol 5.72 and 5.65 eV (using low and high pulse energy), and in toluene 3 eV. The values for the relative permittivity in the range of 7.74–8.90 were measured for hafnia nanoparticles' thin films deposited on substrates by drop-casting (self-assembled layers) in parallel plate capacitor structures.« less

Formation Energies and Electronic Properties of Vanadium Carbides Found in High Strength Steel Alloys

NASA Astrophysics Data System (ADS)

Limmer, Krista; Medvedeva, Julia

2013-03-01

Carbide formation and stabilization in steels is of great interest owing to its effect on the microstructure and properties of the Fe-based alloys. The appearance of carbides with different metal/C ratios strongly depends on the carbon concentration, alloy composition as well as the heat treatment. Strong carbide-forming elements such as Ti, V, and Nb have been used in microalloyed steels; with VC showing an increased solubility in the iron matrix as compared with TiC and NbC. This allows for dissolution of the VC into the steel during heating and fine precipitation during cooling. In addition to VC, the primary vanadium carbide with cubic structure, a wide range of non-stoichiometric compositions VCy with y varying from 0.72 to 0.88, has been observed. This range includes two ordered compounds, V8C7 and V6C5. In this study, first-principles density functional theory (DFT) is employed to examine the stability of the binary carbides by calculating their formation energies. We compare the local structures (atomic coordination, bond distances and angles) and the density of states in optimized geometries of the carbides. Further, the effect of alloying additions, such as niobium and titanium, on the carbide stabilization is investigated. We determine the energetically preferable substitutional atom location in each carbide and study the impurity distribution as well as its role in the carbide formation energy and electronic structure.

Crystallographic characterizations of eutectic and secondary carbides in a Fe-12Cr-2.5Mo-1.5W-3V-1.25C alloy

NASA Astrophysics Data System (ADS)

Guo, Jing; Liu, Ligang; Feng, Yunli; Liu, Sha; Ren, Xuejun; Yang, Qingxiang

2017-03-01

In this work, the morphology and structures of the eutectic and secondary carbides in a new high chromium Fe-12Cr-2.5Mo-1.5W-3V-1.25C designed for cold-rolling work roll were systematically studied. The precipitated carbides inside the grains and along the grain boundaries were investigated with optical microscope, scanning electron microscopy with energy dispersive spectroscopy, transmission electron microscopy and X-Ray diffraction. Selected area diffraction patterns have been successfully used to identify the crystal formation and lattice constants of the carbides with different alloying elements. The results show that the eutectic carbides precipitated contain MC and M2C distributed along the grain boundaries with dendrite feature. The composition and crystal structure analysis shows that the eutectic MC carbides contain VC and WC with a cubic and hexagonal crystal lattice structures respectively, while the eutectic M2C carbides predominantly contain V2C and Mo2C with orthorhombic and hexagonal crystal lattices respectively. The secondary carbides contain MC, M2C, M7C3 formed along the grain boundaries and their sizes are much larger than the eutectic carbides ones. The secondary M23C6 is much small (0.3-0.5μm) and is distributed dispersively inside the grain. Similar to the eutectic carbides, the secondary carbides also contain VC, WC, V2C, and Mo2C. M7C3 is hexagonal (Fe,Cr)7C3, while M23C6 is indexed to be in a cubic crystal form.

Aeromagnetic and Aeroradiometric Data for the Conterminous United States and Alaska from the National Uranium Resource Evaluation (NURE) Program of the U.S. Department of Energy

USGS Publications Warehouse

Hill, Patricia L.; Kucks, Robert P.; Ravat, Dhananjay

2009-01-01

The National Uranium Resource Evaluation (NURE) program was initiated in 1973 with a primary goal of identifying uranium resources in the United States. The airborne program's main purpose was to collect radiometric data of the conterminous United States and Alaska. Magnetic data were also collected. After the program ended, most of the data were given to the U.S. Geological Survey (USGS). All areas were flown at about 400 feet above ground, the optimum height for collecting radiometric data, and the line spacing varied from 3 to 6 mile intervals. A few selected quadrangles or parts of quadrangles were flown at 1- or 2-mile line spacing. About forty smaller areas were targeted and flown at 0.25-mile to 1 mile line spacing.

Studies on separation and purification of fission (99)Mo from neutron activated uranium aluminum alloy.

PubMed

Rao, Ankita; Kumar Sharma, Abhishek; Kumar, Pradeep; Charyulu, M M; Tomar, B S; Ramakumar, K L

2014-07-01

A new method has been developed for separation and purification of fission (99)Mo from neutron activated uranium-aluminum alloy. Alkali dissolution of the irradiated target (100mg) results in aluminum along with (99)Mo and a few fission products passing into solution, while most of the fission products, activation products and uranium remain undissolved. Subsequent purification steps involve precipitation of aluminum as Al(OH)3, iodine as AgI/AgIO3 and molybdenum as Mo-α-benzoin oxime. Ruthenium is separated by volatilization as RuO4 and final purification of (99)Mo was carried out using anion exchange method. The radiochemical yield of fission (99)Mo was found to be >80% and the purity of the product was in conformity with the international pharmacopoeia standards. Copyright © 2014 Elsevier Ltd. All rights reserved.

235U Holdup Measurements in the 321-M Exhaust Elbows

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salaymeh, S.R.

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Disposition Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The facility also includes the 324-M storage building and the passageway connecting it to 321-M. The results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control and Accountability, and to meetmore » criticality safety controls. This report covers holdup measurements of uranium residue in the exhaust piping elbows removed from the roof the 321-M facility.« less

Microstructural characterization of Ni-based self-fluxing alloy after selective surface-engineering using diode laser

NASA Astrophysics Data System (ADS)

Chun, Eun-Joon; Park, Changkyoo; Nishikawa, Hiroshi; Kim, Min-Su

2018-06-01

The microstructural characterization of thermal-sprayed Ni-based self-fluxing alloy (Metco-16C®) after laser-assisted homogenization treatment was performed. To this end, a high-power diode laser system was used. This supported the real-time control of the target homogenization temperature at the substrate surface. Non-homogeneities of the macrosegregation of certain elements (C and Cu) and the local concentration of Cr-based carbides and borides in certain regions in the as-sprayed state could be enhanced with the application of homogenization. After homogenization at 1423 K, the hardness of the thermal-sprayed layer was found to have increased by 1280 HV from the as-sprayed state (750 HV). At this homogenization temperature, the microstructure of the thermal-sprayed layer consisted of a lamellar structuring of the matrix phase (austenite and Ni3Si) with fine (<5 μm) carbides and borides (the rod-like phase of Cr5B3, the lumpy phase of M23C6, and the extra-fine phase of M7C3). Despite the formation of several kinds of carbides and borides during homogenization at 1473 K, the lowest hardness level was found to be less than that of the as-sprayed state, because of the liquid-state homogenization treatment without formation of lamellar structuring between austenite and Ni3Si.

Microbial Functional Gene Diversity Predicts Groundwater Contamination and Ecosystem Functioning

DOE Office of Scientific and Technical Information (OSTI.GOV)

He, Zhili; Zhang, Ping; Wu, Linwei

Contamination from anthropogenic activities has significantly impacted Earth’s biosphere. However, knowledge about how environmental contamination affects the biodiversity of groundwater microbiomes and ecosystem functioning remains very limited. Here, we used a comprehensive functional gene array to analyze groundwater microbiomes from 69 wells at the Oak Ridge Field Research Center (Oak Ridge, TN), representing a wide pH range and uranium, nitrate, and other contaminants. We hypothesized that the functional diversity of groundwater microbiomes would decrease as environmental contamination (e.g., uranium or nitrate) increased or at low or high pH, while some specific populations capable of utilizing or resistant to those contaminantsmore » would increase, and thus, such key microbial functional genes and/or populations could be used to predict groundwater contamination and ecosystem functioning. Our results indicated that functional richness/diversity decreased as uranium (but not nitrate) increased in groundwater. In addition, about 5.9% of specific key functional populations targeted by a comprehensive functional gene array (GeoChip 5) increased significantly (P < 0.05) as uranium or nitrate increased, and their changes could be used to successfully predict uranium and nitrate contamination and ecosystem functioning. Here, this study indicates great potential for using microbial functional genes to predict environmental contamination and ecosystem functioning.« less

Microbial Functional Gene Diversity Predicts Groundwater Contamination and Ecosystem Functioning

PubMed Central

Zhang, Ping; Wu, Linwei; Rocha, Andrea M.; Shi, Zhou; Wu, Bo; Qin, Yujia; Wang, Jianjun; Yan, Qingyun; Curtis, Daniel; Ning, Daliang; Van Nostrand, Joy D.; Wu, Liyou; Watson, David B.; Adams, Michael W. W.; Alm, Eric J.; Adams, Paul D.; Arkin, Adam P.

2018-01-01

ABSTRACT Contamination from anthropogenic activities has significantly impacted Earth’s biosphere. However, knowledge about how environmental contamination affects the biodiversity of groundwater microbiomes and ecosystem functioning remains very limited. Here, we used a comprehensive functional gene array to analyze groundwater microbiomes from 69 wells at the Oak Ridge Field Research Center (Oak Ridge, TN), representing a wide pH range and uranium, nitrate, and other contaminants. We hypothesized that the functional diversity of groundwater microbiomes would decrease as environmental contamination (e.g., uranium or nitrate) increased or at low or high pH, while some specific populations capable of utilizing or resistant to those contaminants would increase, and thus, such key microbial functional genes and/or populations could be used to predict groundwater contamination and ecosystem functioning. Our results indicated that functional richness/diversity decreased as uranium (but not nitrate) increased in groundwater. In addition, about 5.9% of specific key functional populations targeted by a comprehensive functional gene array (GeoChip 5) increased significantly (P < 0.05) as uranium or nitrate increased, and their changes could be used to successfully predict uranium and nitrate contamination and ecosystem functioning. This study indicates great potential for using microbial functional genes to predict environmental contamination and ecosystem functioning. PMID:29463661

Microbial Functional Gene Diversity Predicts Groundwater Contamination and Ecosystem Functioning

DOE PAGES

He, Zhili; Zhang, Ping; Wu, Linwei; ...

2018-02-20

Contamination from anthropogenic activities has significantly impacted Earth’s biosphere. However, knowledge about how environmental contamination affects the biodiversity of groundwater microbiomes and ecosystem functioning remains very limited. Here, we used a comprehensive functional gene array to analyze groundwater microbiomes from 69 wells at the Oak Ridge Field Research Center (Oak Ridge, TN), representing a wide pH range and uranium, nitrate, and other contaminants. We hypothesized that the functional diversity of groundwater microbiomes would decrease as environmental contamination (e.g., uranium or nitrate) increased or at low or high pH, while some specific populations capable of utilizing or resistant to those contaminantsmore » would increase, and thus, such key microbial functional genes and/or populations could be used to predict groundwater contamination and ecosystem functioning. Our results indicated that functional richness/diversity decreased as uranium (but not nitrate) increased in groundwater. In addition, about 5.9% of specific key functional populations targeted by a comprehensive functional gene array (GeoChip 5) increased significantly (P < 0.05) as uranium or nitrate increased, and their changes could be used to successfully predict uranium and nitrate contamination and ecosystem functioning. Here, this study indicates great potential for using microbial functional genes to predict environmental contamination and ecosystem functioning.« less

Effects of uranium concentration on microbial community structure and functional potential.

PubMed

Sutcliffe, Brodie; Chariton, Anthony A; Harford, Andrew J; Hose, Grant C; Greenfield, Paul; Elbourne, Liam D H; Oytam, Yalchin; Stephenson, Sarah; Midgley, David J; Paulsen, Ian T

2017-08-01

Located in the Northern Territory of Australia, Ranger uranium mine is directly adjacent to the UNESCO World Heritage listed Kakadu National Park, with rehabilitation targets needed to ensure the site can be incorporated into the park following the mine's closure in 2026. This study aimed to understand the impact of uranium concentration on microbial communities, in order to identify and describe potential breakpoints in microbial ecosystem services. This is the first study to report in situ deployment of uranium-spiked sediments along a concentration gradient (0-4000 mg U kg -1 ), with the study design maximising the advantages of both field surveys and laboratory manipulative studies. Changes to microbial communities were characterised through the use of amplicon and shotgun metagenomic next-generation sequencing. Significant changes to taxonomic and functional community assembly occurred at a concentration of 1500 mg U kg -1 sediment and above. At uranium concentrations of ≥ 1500 mg U kg -1 , genes associated with methanogenic consortia and processes increased in relative abundance, while numerous significant changes were also seen in the relative abundances of genes involved in nitrogen cycling. Such alterations in carbon and nitrogen cycling pathways suggest that taxonomic and functional changes to microbial communities may result in changes in ecosystem processes and resilience. © 2017 Society for Applied Microbiology and John Wiley & Sons Ltd.

Quantifying Uranium Isotope Ratios Using Resonance Ionization Mass Spectrometry: The Influence of Laser Parameters on Relative Ionization Probability

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isselhardt, Brett H.

2011-09-01

Resonance Ionization Mass Spectrometry (RIMS) has been developed as a method to measure relative uranium isotope abundances. In this approach, RIMS is used as an element-selective ionization process to provide a distinction between uranium atoms and potential isobars without the aid of chemical purification and separation. We explore the laser parameters critical to the ionization process and their effects on the measured isotope ratio. Specifically, the use of broad bandwidth lasers with automated feedback control of wavelength was applied to the measurement of 235U/ 238U ratios to decrease laser-induced isotopic fractionation. By broadening the bandwidth of the first laser inmore » a 3-color, 3-photon ionization process from a bandwidth of 1.8 GHz to about 10 GHz, the variation in sequential relative isotope abundance measurements decreased from >10% to less than 0.5%. This procedure was demonstrated for the direct interrogation of uranium oxide targets with essentially no sample preparation. A rate equation model for predicting the relative ionization probability has been developed to study the effect of variation in laser parameters on the measured isotope ratio. This work demonstrates that RIMS can be used for the robust measurement of uranium isotope ratios.« less

A new criterion for predicting rolling-element fatigue lives of through-hardened steels

NASA Technical Reports Server (NTRS)

Chevalier, J. L.; Zaretsky, E. V.; Parker, R. J.

1972-01-01

A carbide factor was derived based upon a statistical analysis which related rolling-element fatigue life to the total number of residual carbide particles per unit area, median residual carbide size, and percent residual carbide area. An equation was experimentally determined which predicts material hardness as a function of temperature. The limiting temperatures of all of the materials studied were dependent on initial room temperature hardness and tempering temperature. An equation was derived combining the effects of material hardness, carbide factor, and bearing temperature to predict rolling-element bearing life.

METHOD OF JOINING CARBIDES TO BASE METALS

DOEpatents

Krikorian, N.H.; Farr, J.D.; Witteman, W.G.

1962-02-13

A method is described for joining a refractory metal carbide such as UC or ZrC to a refractory metal base such as Ta or Nb. The method comprises carburizing the surface of the metal base and then sintering the base and carbide at temperatures of about 2000 deg C in a non-oxidizing atmosphere, the base and carbide being held in contact during the sintering step. To reduce the sintering temperature and time, a sintering aid such as iron, nickel, or cobait is added to the carbide, not to exceed 5 wt%. (AEC)

COATED CARBON ELEMENT FOR USE IN NUCLEAR REACTORS AND THE PROCESS OF MAKING THE ELEMENT

DOEpatents

Pyle, R.J.; Allen, G.L.

1963-01-15

S>This patent relates to a carbide-nitride-carbide coating for carbon bodies that are to be subjected to a high temperature nuclear reactor atmosphere, and a method of applying the same. This coating is a highly efficient diffusion barrier and protects the C body from corrosion and erosion by the reactor atmosphere. Preferably, the innermost coating is Zr carbide, the middle coatlng is Zr nitride, and the outermost coating is a mixture of Zr and Nb carbide. The nitride coating acts as a diffusion barrier, while the innermost carbide bonds the nitride to the C body and prevents deleterious reaction between the nitride and C body. The outermost carbide coating protects the nitride coating from the reactor atmosphere. (AEC)

Measuring Aerosols Generated Inside Armoured Vehicles Perforated by Depleted Uranium Ammunition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parkhurst, MaryAnn

2003-01-01

In response to questions raised after the Gulf War about the health significance of exposure to depleted uranium (DU), the U.S. Department of Defense initiated a study designed to provide an improved scientific basis for assessment of possible health effects of soldiers in vehicles struck by these munitions. As part of this study, a series of DU penetrators were fired at an Abrams tank and a Bradley fighting vehicle, and the aerosols generated by vehicle perforation were collected and characterized. A robust sampling system was designed to collect aerosols in this difficult environment and to monitor continuously the sampler flowmore » rates. Interior aerosols collected were analyzed for uranium concentration and particle size distribution as a function of time. They were also analyzed for uranium oxide phases, particle morphology, and dissolution in vitro. These data will provide input for future prospective and retrospective dose and health risk assessments of inhaled or ingested DU aerosols. This paper briefly discusses the target vehicles, firing trajectories, aerosol samplers and instrumentation control systems, and the types of analyses conducted on the samples.« less

X-ray Emission from Highly Charged Heavy Ions Studied at Storage Rings

NASA Astrophysics Data System (ADS)

Ma, X.; Stöhlker, Th.; Bosch, F.; Gumberidze, A.; Kozhuharov, C.; Muthig, A.; Mokler, P. H.; Warczak, A.

2003-01-01

Radiative electron capture at low projectile energies is studied via angular differential cross sections for collisions of bare uranium with low-Z target atoms. Our results show that for high-Z systems relativistic effects such as spin-flip transitions show up in an unambiguous fashion which still persist even in the low-energy domain. Moreover, following REC into the 2p3/2 state a strong alignment of this level was observed by measuring the angular distribution of the Lyα1 transition in H-like uranium. Here, an interference between the leading E1 decay channel and the much weaker M2 multipole transition gives rise to a remarkable modified angular distribution of the emitted photons. For the particular case of hydrogen-like uranium the former variance of the experimental data with theoretical findings is removed when this E1/M2 multipole mixing is taken into account. Finally, with respect to atomic structure studies, a very recent experiment will be discussed aiming on a precise determination of the electron-electron QED contribution to the groundstate ionization potential in He-like uranium.

GIS prospectivity mapping and 3D modeling validation for potential uranium deposit targets in Shangnan district, China

NASA Astrophysics Data System (ADS)

Xie, Jiayu; Wang, Gongwen; Sha, Yazhou; Liu, Jiajun; Wen, Botao; Nie, Ming; Zhang, Shuai

2017-04-01

Integrating multi-source geoscience information (such as geology, geophysics, geochemistry, and remote sensing) using GIS mapping is one of the key topics and frontiers in quantitative geosciences for mineral exploration. GIS prospective mapping and three-dimensional (3D) modeling can be used not only to extract exploration criteria and delineate metallogenetic targets but also to provide important information for the quantitative assessment of mineral resources. This paper uses the Shangnan district of Shaanxi province (China) as a case study area. GIS mapping and potential granite-hydrothermal uranium targeting were conducted in the study area combining weights of evidence (WofE) and concentration-area (C-A) fractal methods with multi-source geoscience information. 3D deposit-scale modeling using GOCAD software was performed to validate the shapes and features of the potential targets at the subsurface. The research results show that: (1) the known deposits have potential zones at depth, and the 3D geological models can delineate surface or subsurface ore-forming features, which can be used to analyze the uncertainty of the shape and feature of prospectivity mapping at the subsurface; (2) single geochemistry anomalies or remote sensing anomalies at the surface require combining the depth exploration criteria of geophysics to identify potential targets; and (3) the single or sparse exploration criteria zone with few mineralization spots at the surface has high uncertainty in terms of the exploration target.

Characterization of transition carbides in quench and partitioned steel microstructures by Mössbauer spectroscopy and complementary techniques

DOE PAGES

Pierce, D. T.; Coughlin, D. R.; Williamson, D. L.; ...

2015-05-01

Quenching and partitioning (Q&P) produces steel microstructures with martensite and austenite that exhibit promising property combinations for third generation advanced high strength steels. Understanding the kinetics of reactions that compete for available carbon, such as carbide formation, is critical for alloying and processing design and achieving austenite enrichment and retention during Q&P. Mössbauer effect spectroscopy (MES) was used to characterize Q&P microstructures in a 0.38C-1.54Mn-1.48Si wt.% steel after quenching to 225 °C and partitioning at 400 °C for 10 or 300 s, with an emphasis on transition carbides. The recoilless fraction for η-carbide was calculated and a correction for saturationmore » of the MES absorption spectrum was applied, making quantitative measurements of small amounts of η-carbide, including non-stoichiometric η-carbide, possible in Q&P microstructures. Complementary transmission electron microscopy confirmed the presence of η-carbides, and MES and X-ray diffraction were used to characterize the austenite. The amount of η-carbide formed during Q&P ranged from 1.4 to 2.4 at.%, accounting for a substantial portion (~24% to 41%) of the bulk carbon content of the steel. The amount (5.0 at.%) of η-carbide that formed after quenching and tempering (Q&T) at 400 °C for 300 s was significantly greater than after partitioning at 400 °C for 300 s (2.4 at.%), suggesting that carbon partitioning from martensite to austenite occurs in conjunction with η-carbide formation during Q&P in these specimens.« less

A silicon carbide array for electrocorticography and peripheral nerve recording.

PubMed

Diaz-Botia, C A; Luna, L E; Neely, R M; Chamanzar, M; Carraro, C; Carmena, J M; Sabes, P N; Maboudian, R; Maharbiz, M M

2017-10-01

Current neural probes have a limited device lifetime of a few years. Their common failure mode is the degradation of insulating films and/or the delamination of the conductor-insulator interfaces. We sought to develop a technology that does not suffer from such limitations and would be suitable for chronic applications with very long device lifetimes. We developed a fabrication method that integrates polycrystalline conductive silicon carbide with insulating silicon carbide. The technology employs amorphous silicon carbide as the insulator and conductive silicon carbide at the recording sites, resulting in a seamless transition between doped and amorphous regions of the same material, eliminating heterogeneous interfaces prone to delamination. Silicon carbide has outstanding chemical stability, is biocompatible, is an excellent molecular barrier and is compatible with standard microfabrication processes. We have fabricated silicon carbide electrode arrays using our novel fabrication method. We conducted in vivo experiments in which electrocorticography recordings from the primary visual cortex of a rat were obtained and were of similar quality to those of polymer based electrocorticography arrays. The silicon carbide electrode arrays were also used as a cuff electrode wrapped around the sciatic nerve of a rat to record the nerve response to electrical stimulation. Finally, we demonstrated the outstanding long term stability of our insulating silicon carbide films through accelerated aging tests. Clinical translation in neural engineering has been slowed in part due to the poor long term performance of current probes. Silicon carbide devices are a promising technology that may accelerate this transition by enabling truly chronic applications.

A silicon carbide array for electrocorticography and peripheral nerve recording

NASA Astrophysics Data System (ADS)

Diaz-Botia, C. A.; Luna, L. E.; Neely, R. M.; Chamanzar, M.; Carraro, C.; Carmena, J. M.; Sabes, P. N.; Maboudian, R.; Maharbiz, M. M.

2017-10-01

Objective. Current neural probes have a limited device lifetime of a few years. Their common failure mode is the degradation of insulating films and/or the delamination of the conductor-insulator interfaces. We sought to develop a technology that does not suffer from such limitations and would be suitable for chronic applications with very long device lifetimes. Approach. We developed a fabrication method that integrates polycrystalline conductive silicon carbide with insulating silicon carbide. The technology employs amorphous silicon carbide as the insulator and conductive silicon carbide at the recording sites, resulting in a seamless transition between doped and amorphous regions of the same material, eliminating heterogeneous interfaces prone to delamination. Silicon carbide has outstanding chemical stability, is biocompatible, is an excellent molecular barrier and is compatible with standard microfabrication processes. Main results. We have fabricated silicon carbide electrode arrays using our novel fabrication method. We conducted in vivo experiments in which electrocorticography recordings from the primary visual cortex of a rat were obtained and were of similar quality to those of polymer based electrocorticography arrays. The silicon carbide electrode arrays were also used as a cuff electrode wrapped around the sciatic nerve of a rat to record the nerve response to electrical stimulation. Finally, we demonstrated the outstanding long term stability of our insulating silicon carbide films through accelerated aging tests. Significance. Clinical translation in neural engineering has been slowed in part due to the poor long term performance of current probes. Silicon carbide devices are a promising technology that may accelerate this transition by enabling truly chronic applications.

Solid oxide membrane-assisted controllable electrolytic fabrication of metal carbides in molten salt.

PubMed

Zou, Xingli; Zheng, Kai; Lu, Xionggang; Xu, Qian; Zhou, Zhongfu

2016-08-15

Silicon carbide (SiC), titanium carbide (TiC), zirconium carbide (ZrC), and tantalum carbide (TaC) have been electrochemically produced directly from their corresponding stoichiometric metal oxides/carbon (MOx/C) precursors by electrodeoxidation in molten calcium chloride (CaCl2). An assembled yttria stabilized zirconia solid oxide membrane (SOM)-based anode was employed to control the electrodeoxidation process. The SOM-assisted controllable electrochemical process was carried out in molten CaCl2 at 1000 °C with a potential of 3.5 to 4.0 V. The reaction mechanism of the electrochemical production process and the characteristics of these produced metal carbides (MCs) were systematically investigated. X-ray diffraction, scanning electron microscopy, and transmission electron microscopy analyses clearly identify that SiC, TiC, ZrC, and TaC carbides can be facilely fabricated. SiC carbide can be controlled to form a homogeneous nanowire structure, while the morphologies of TiC, ZrC, and TaC carbides exhibit porous nodular structures with micro/nanoscale particles. The complex chemical/electrochemical reaction processes including the compounding, electrodeoxidation, dissolution-electrodeposition, and in situ carbonization processes in molten CaCl2 are also discussed. The present results preliminarily demonstrate that the molten salt-based SOM-assisted electrodeoxidation process has the potential to be used for the facile and controllable electrodeoxidation of MOx/C precursors to micro/nanostructured MCs, which can potentially be used for various applications.

A Novel Method to Calculate the Carbides Fraction from Dilatometric Measurements During Cooling in Hot-Work Tool Steel

NASA Astrophysics Data System (ADS)

Zhao, Xiaoli; Li, Chuanwei; Han, Lizhan; Gu, Jianfeng

2018-06-01

Dilatometry is a useful technique to obtain experimental data concerning transformation. In this paper, a dilation conversional model was established to calculate carbides fraction in AISI H13 hot-work tool steel based on the measured length changes. After carbides precipitation, the alloy contents in the matrix changed. In the usual models, the content of carbon atoms after precipitation is considered as the only element that affects the lattice constant and the content of the alloy elements such as Cr, Mo, Mn, V are often ignored. In the model introduced in this paper, the alloying elements (Cr, Mo, Mn, V) changes caused by carbides precipitation are incorporated. The carbides were identified using scanning electron microscope and transmission electron microscope. The relationship between lattice constant of carbides and temperature are measured by high-temperature X-ray diffraction. The results indicate that the carbides observed in all specimens cooled at different rates are V-rich MC and Cr-rich M23C6, and most of them are V-rich MC, only very few are Cr-rich M23C6. The model including the effects of substitutional alloying elements shows a good improvement on carbides fraction predictions. In addition, lower cooling rate advances the carbides precipitation for AISI H13 specimens. The results between experiments and mathematical model agree well.

New Structural Materials

DTIC Science & Technology

1989-03-25

3887) Tantalum Carbide (TaC) 4150 (3877) Niobium Carbide 4023 (3750) Carbon (Graphite) 3970 (3697) Zirconium Carbide 3805 (3532) Tungsten 3643 3643...process. Some fibers, especially those made of tungsten , silicon carbide, and zirconia, survived the reaction conditions. However, the ceramic bodies...displayed cracks and voids. Examination by SEM of cross’sections of the reacted parts made with tungsten fibers disclosed the presence of "whiskers

Co-precipitated and collocated carbides and Cu-rich precipitates in a Fe-Cu steel characterized by atom-probe tomography.

PubMed

Kolli, R Prakash; Seidman, David N

2014-12-01

The composition of co-precipitated and collocated NbC carbide precipitates, Fe3C iron carbide (cementite), and Cu-rich precipitates are studied experimentally by atom-probe tomography (APT). The Cu-rich precipitates located at a grain boundary (GB) are also studied. The APT results for the carbides are supplemented with computational thermodynamics predictions of composition at thermodynamic equilibrium. Two types of NbC carbide precipitates are distinguished based on their stoichiometric ratio and size. The Cu-rich precipitates at the periphery of the iron carbide and at the GB are larger than those distributed in the α-Fe (body-centered cubic) matrix, which is attributed to short-circuit diffusion of Cu along the GB. Manganese segregation is not observed at the heterophase interfaces of the Cu-rich precipitates that are located at the periphery of the iron carbide or at the GB, which is unlike those located at the edge of the NbC carbide precipitates or distributed in the α-Fe matrix. This suggests the presence of two populations of NiAl-type (B2 structure) phases at the heterophase interfaces in multicomponent Fe-Cu steels.

RIBE at an inter-organismic level: A study on genotoxic effects in Daphnia magna exposed to waterborne uranium and a uranium mine effluent.

PubMed

Reis, P; Lourenço, J; Carvalho, F P; Oliveira, J; Malta, M; Mendo, S; Pereira, R

2018-05-01

The induction of RIBE (Radiation Induced Bystander Effect) is a non-target effect of low radiation doses that has already been verified at an inter-organismic level in fish and small mammals. Although the theoretical impact in the field of environmental risk assessment (ERA) is possible, there is a gap of knowledge regarding this phenomenon in invertebrate groups and following environmentally relevant exposures. To understand if RIBE should be considered for ERA of radionuclide-rich wastewaters, we exposed Daphnia magna (<24 h and 5d old) to a 2% diluted uranium mine effluent for 48 h, and to a matching dose of waterborne uranium (55.3 μg L -1 ). Then the exposed organisms were placed (24 and 48 h) in a clean medium together with non-exposed neonates. The DNA damage observed for the non-exposed organisms was statistically significant after the 24 h cohabitation for both uranium (neonates p = 0.002; 5 d-old daphnids p = <0.001) and uranium mine effluent exposure (only for neonates p = 0.042). After 48 h cohabitation significant results were obtained only for uranium exposure (neonates p = 0.017; 5 d-old daphnids p = 0.013). Although there may be some variability associated to age and exposure duration, the significant DNA damage detected in non-exposed organisms clearly reveals the occurrence of RIBE in D. magna. The data obtained and here presented are a valuable contribution for the discussion about the relevance of RIBE for environmental risk assessment. Copyright © 2018 Elsevier B.V. All rights reserved.

Austenite decomposition to carbide-rich products in Fe-0.30C-6.3W

NASA Astrophysics Data System (ADS)

Hackenberg, R. E.; Granada, D. G.; Shiflet, G. J.

2002-12-01

The kinetics, morphology, and elemental distributions associated with the decomposition of austenite in Fe-0.30C-6.3W were surveyed, especially in the bay region of the time-temperature-transformation (TTT) diagram. Carbide precipitation characteristics were of particular interest. Similar to Fe-C-Mo and Fe-C-Cr alloys, grain- and twin-boundary bainite containing sheets of alloy carbides dominated the microstructure at and above the bay, while popcorn-like bainite was observed immediately below the bay. Nonequilibrium carbide-phase combinations were obtained both above and below the bay, although W partitioning to the alloy carbides was always observed. The carbon level in the remaining austenite increased with reaction time at a given temperature, which, at the later stages of reaction, helped trigger the growth of a constituent containing a high density of nonlamellar carbides. These nonequilibrium reaction-path characteristics are considered to originate from crystallographic and interfacial structure constraints affecting the nucleation of carbides at ferrite-austenite interfaces.

SiC formation for a solar cell passivation layer using an RF magnetron co-sputtering system

PubMed Central

2012-01-01

In this paper, we describe a method of amorphous silicon carbide film formation for a solar cell passivation layer. The film was deposited on p-type silicon (100) and glass substrates by an RF magnetron co-sputtering system using a Si target and a C target at a room-temperature condition. Several different SiC [Si1-xCx] film compositions were achieved by controlling the Si target power with a fixed C target power at 150 W. Then, structural, optical, and electrical properties of the Si1-xCx films were studied. The structural properties were investigated by transmission electron microscopy and secondary ion mass spectrometry. The optical properties were achieved by UV-visible spectroscopy and ellipsometry. The performance of Si1-xCx passivation was explored by carrier lifetime measurement. PMID:22221730

Searching for U-235m produced by Nuclear Excitation by Electronic Transition

NASA Astrophysics Data System (ADS)

Chodash, Perry; Norman, Eric; Burke, Jason; Wilks, Scott; Casperson, Robert

2014-09-01

Nuclear excitation by electronic transition (NEET) is a rare nuclear excitation that is predicted to occur in numerous isotopes, including U-235. When a nuclear transition matches the energy and the multipolarity of an electronic transition, there is a possibility that NEET will occur. If NEET were to occur in U-235, the nucleus would be excited to its 1/2 + isomeric state that subsequently decays by internal conversion with a decay energy of 77 eV and a half-life of 26 minutes. Theory predicts that NEET can occur in partially ionized uranium plasma with a charge state of 23 +. A pulsed Nd:YAG laser operating at 1064 nm with a pulse energy of 780 mJ and a pulse width of 9 ns was used to generate the uranium plasma. The laser was focused on small samples of both depleted uranium and highly enriched uranium. The plasma conditions created by the intense laser pulse were varied by changing the spot size of the laser on the target. The resulting plasma was collected on a plate and the internal conversion electrons were focused onto a microchannel plate detector by a series of electrostatic lenses. First results will be presented. Nuclear excitation by electronic transition (NEET) is a rare nuclear excitation that is predicted to occur in numerous isotopes, including U-235. When a nuclear transition matches the energy and the multipolarity of an electronic transition, there is a possibility that NEET will occur. If NEET were to occur in U-235, the nucleus would be excited to its 1/2 + isomeric state that subsequently decays by internal conversion with a decay energy of 77 eV and a half-life of 26 minutes. Theory predicts that NEET can occur in partially ionized uranium plasma with a charge state of 23 +. A pulsed Nd:YAG laser operating at 1064 nm with a pulse energy of 780 mJ and a pulse width of 9 ns was used to generate the uranium plasma. The laser was focused on small samples of both depleted uranium and highly enriched uranium. The plasma conditions created by the intense laser pulse were varied by changing the spot size of the laser on the target. The resulting plasma was collected on a plate and the internal conversion electrons were focused onto a microchannel plate detector by a series of electrostatic lenses. First results will be presented. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. The U.S. DHS, UC Berkeley, the NNIS fellowship and the NSSC further supported this work.

CHEMICAL DIFFERENCES BETWEEN SLUDGE SOLIDS AT THE F AND H AREA TANK FARMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reboul, S.

2012-08-29

The primary source of waste solids received into the F Area Tank Farm (FTF) was from PUREX processing performed to recover uranium and plutonium from irradiated depleted uranium targets. In contrast, two primary sources of waste solids were received into the H Area Tank Farm (HTF): a) waste from PUREX processing; and b) waste from H-modified (HM) processing performed to recover uranium and neptunium from burned enriched uranium fuel. Due to the differences between the irradiated depleted uranium targets and the burned enriched uranium fuel, the average compositions of the F and H Area wastes are markedly different from onemore » another. Both F and H Area wastes contain significant amounts of iron and aluminum compounds. However, because the iron content of PUREX waste is higher than that of HM waste, and the aluminum content of PUREX waste is lower than that of HM waste, the iron to aluminum ratios of typical FTF waste solids are appreciably higher than those of typical HTF waste solids. Other constituents present at significantly higher concentrations in the typical FTF waste solids include uranium, nickel, ruthenium, zinc, silver, cobalt and copper. In contrast, constituents present at significantly higher concentrations in the typical HTF waste solids include mercury, thorium, oxalate, and radionuclides U-233, U-234, U-235, U-236, Pu-238, Pu-242, Cm-244, and Cm-245. Because of the higher concentrations of Pu-238 in HTF, the long-term concentrations of Th-230 and Ra-226 (from Pu-238 decay) will also be higher in HTF. The uranium and plutonium distributions of the average FTF waste were found to be consistent with depleted uranium and weapons grade plutonium, respectively (U-235 comprised 0.3 wt% of the FTF uranium, and Pu-240 comprised 6 wt% of the FTF plutonium). In contrast, at HTF, U-235 comprised 5 wt% of the uranium, and Pu-240 comprised 17 wt% of the plutonium, consistent with enriched uranium and high burn-up plutonium. X-ray diffraction analyses of various FTF and HTF samples indicated that the primary crystalline compounds of iron in sludge solids are Fe{sub 2}O{sub 3}, Fe{sub 3}O{sub 4}, and FeO(OH), and the primary crystalline compounds of aluminum are Al(OH){sub 3} and AlO(OH). Also identified were carbonate compounds of calcium, magnesium, and sodium; a nitrated sodium aluminosilicate; and various uranium compounds. Consistent with expectations, oxalate compounds were identified in solids associated with oxalic acid cleaning operations. The most likely oxidation states and chemical forms of technetium are assessed in the context of solubility, since technetium-99 is a key risk driver from an environmental fate and transport perspective. The primary oxidation state of technetium in SRS sludge solids is expected to be Tc(IV). In salt waste, the primary oxidation state is expected to be Tc(VII). The primary form of technetium in sludge is expected to be a hydrated technetium dioxide, TcO{sub 2} {center_dot} xH{sub 2}O, which is relatively insoluble and likely co-precipitated with iron. In salt waste solutions, the primary form of technetium is expected to be the very soluble pertechnetate anion, TcO{sub 4}{sup -}. The relative differences between the F and H Tank Farm waste provide a basis for anticipating differences that will occur as constituents of FTF and HTF waste residue enter the environment over the long-term future. If a constituent is significantly more dominant in one of the Tank Farms, its long-term environmental contribution will likely be commensurately higher, assuming the environmental transport conditions of the two Tank Farms share some commonality. It is in this vein that the information cited in this document is provided - for use during the generation, assessment, and validation of Performance Assessment modeling results.« less

Versatile Boron Carbide-Based Visual Obscurant Compositions for Smoke Munitions

DTIC Science & Technology

2015-04-17

Versatile Boron Carbide-Based Visual Obscurant Compositions for Smoke Munitions Anthony P. Shaw,*,† Giancarlo Diviacchi,‡ Ernest L. Black,‡ Jared D...have been demonstrated to produce thick white smoke clouds upon combustion. These compositions use powdered boron carbide (B4C) as a pyrotechnic...ignition and are safe to handle. KEYWORDS: Smoke, Obscurants, Pyrotechnics, Boron carbide, Sustainable chemistry ■ INTRODUCTION Visible obscuration

Method for making hot-pressed fiber-reinforced carbide-graphite composite

DOEpatents

Riley, Robert E.; Wallace Sr., Terry C.

1979-01-01

A method for the chemical vapor deposition of a uniform coating of tantalum metal on fibers of a woven graphite cloth is described. Several layers of the coated cloth are hot pressed to produce a tantalum carbide-graphite composite having a uniformly dispersed, fine grained tantalum carbide in graphite with compositions in the range of 15 to 40 volume percent tantalum carbide.

Method of making silicon carbide-silicon composite having improved oxidation resistance

NASA Technical Reports Server (NTRS)

Wang, Hongyu (Inventor); Luthra, Krishan Lal (Inventor)

2002-01-01

A Silicon carbide-silicon matrix composite having improved oxidation resistance at high temperatures in dry or water-containing environments is provided. A method is given for sealing matrix cracks in situ in melt infiltrated silicon carbide-silicon matrix composites. The composite cracks are sealed by the addition of various additives, such as boron compounds, into the melt infiltrated silicon carbide-silicon matrix.

Silicon carbide-silicon composite having improved oxidation resistance and method of making

NASA Technical Reports Server (NTRS)

Wang, Hongyu (Inventor); Luthra, Krishan Lal (Inventor)

1999-01-01

A Silicon carbide-silicon matrix composite having improved oxidation resistance at high temperatures in dry or water-containing environments is provided. A method is given for sealing matrix cracks in situ in melt infiltrated silicon carbide-silicon matrix composites. The composite cracks are sealed by the addition of various additives, such as boron compounds, into the melt infiltrated silicon carbide-silicon matrix.

Silicon Carbide Capacitive High Temperature MEMS Strain Transducer

DTIC Science & Technology

2012-03-22

SILICON CARBIDE CAPACITIVE HIGH TEMPURATURE MEMS STRAIN TRANSDUCER THESIS Richard P. Weisenberger, DR01, USAF AFIT/GE/ENG...declared a work of the U.S. Government and is not subject to copyright protection in the United States AFIT/GE/ENG/12-43 SILICON CARBIDE CAPACITIVE...STATEMENT A. APPROVED FOR PUBLIC RELEASE; DISTRIBUTION UNLIMITED AFIT/GE/ENG/12-43 SILICON CARBIDE CAPACITIVE IDGH TEMPURATURE MEMS STRAIN TRANSDUCER

Study of the Effect of Trace Mg Additions on Carbides in Die Steel H13

NASA Astrophysics Data System (ADS)

Li, Ji; Li, Jing; Wang, Liang-liang; Zhu, Qin-tian

2016-09-01

Carbides in annealed steel H13 without magnesium and with a micro-addition of magnesium (0.0010%) are studied. Trace amounts of magnesium strengthen carbide segregation and reduce their size. Carbides phases M7C3, M6 C, and M(C, N) are detected in steel H13, and this agrees with results of thermodynamic calculations.

Characterization of boron carbide with an electron microprobe

NASA Technical Reports Server (NTRS)

Matteudi, G.; Ruste, J.

1983-01-01

Within the framework of a study of heterogeneous materials (Matteudi et al., 1971: Matteudi and Verchery, 1972) thin deposits of boron carbide were characterized. Experiments using an electronic probe microanalyzer to analyze solid boron carbide or boron carbide in the form of thick deposits are described. Quantitative results on boron and carbon are very close to those obtained when applying the Monte Carlo-type correction calculations.

Cell metal interactions: A comparison of natural uranium to other common metals in renal cells and bone osteoblasts

NASA Astrophysics Data System (ADS)

Milgram, S.; Carrière, M.; Thiebault, C.; Berger, P.; Khodja, H.; Gouget, B.

2007-07-01

Uranium acute intoxication has been documented to induce nephrotoxicity. Kidneys are the main target organs after short term exposures to high concentrations of the toxic, while chronic exposures lead to its accumulation in the skeleton. In this paper, chemical toxicity of uranium is investigated for rat osteoblastic bone cells and compared to results previously obtained on renal cells. We show that bone cells are less sensitive to uranium than renal cells. The influence of the chemical form on U cytotoxicity is demonstrated. For both cell types, a comparison of uranium toxicity with other metals or metalloids toxicities (Mn, Ni, Co, Cu, Zn, Se and Cd) permits classification of Cd, Zn, Se IV and Cu as the most toxic and Ni, Se VI, Mn and U as the least toxic. Chemical toxicity of natural uranium proves to be far less than that of cadmium. To try to explain the differences in sensitivities observed between metals and different cell types, cellular accumulations in cell monolayers are quantified by inductively coupled plasma-mass spectroscopy (ICP-MS), function of time or function of dose: lethal doses which simulate acute intoxications and sub-lethal doses which are more realistic with regard to environmentally metals concentrations. In addition to being more resistant, bone cells accumulated much more uranium than did renal cells. Moreover, for both cell models, Mn, U-citrate and U-bicarbonate are strongly accumulated whereas Cu, Zn and Ni are weakly accumulated. On the other hand, a strong difference in Cd behaviour between the two cell types is shown: whereas Cd is very weakly accumulated in bone cells, it is very strongly accumulated in renal cells. Finally, elemental distribution of the toxics is determined on a cellular scale using nuclear microprobe analysis. For both renal and osteoblastic cells, uranium was accumulated in as intracellular precipitates similar to those observed previously by SEM/EDS.

Electrocatalysis using transition metal carbide and oxide nanocrystals

NASA Astrophysics Data System (ADS)

Regmi, Yagya N.

Carbides are one of the several families of transition metal compounds that are considered economic alternatives to catalysts based on noble metals and their compounds. Phase pure transition metal carbides of group 4-6 metals, in the first three periods, were synthesized using a common eutectic salt flux synthesis method, and their electrocatalytic activities compared under uniform electrochemical conditions. Mo2C showed highest hydrogen evolution reaction (HER) and oxygen reduction reaction (ORR) activities among the nine metal carbides investigated, but all other metal carbides also showed substantial activities. All the metal carbides showed remarkable enhancement in catalytic activities as supports, when compared to traditional graphitic carbon as platinum support. Mo2C, the most active transition metal carbide electrocatalyst, was prepared using four different synthesis routes, and the synthesis route dependent activities compared. Bifunctional Mo 2C that is HER as well as oxygen evolution reaction (OER) active, was achieved when the carbide was templated on a multiwalled carbon nanotube using carbothermic reduction method. Bimetallic carbides of Fe, Co, and Ni with Mo or W were prepared using a common carbothermic reduction method. Two different stoichiometries of bimetallic carbides were obtained for each system within a 60 °C temperature window. While the bimetallic carbides showed relatively lower electrocatalytic activities towards HER and ORR in comparison to Mo2C and WC, they revealed remarkably higher OER activities than IrO2 and RuO2, the state-of-the-art OER catalysts. Bimetallic oxides of Fe, Co, and Ni with Mo and W were also prepared using a hydrothermal synthesis method and they also revealed OER activities that are much higher than RuO2 and IrO2. Additionally, the OER activities were dependent on the degree and nature of hydration in the bimetallic oxide crystal lattice, with the completely hydrated, as synthesized, cobalt molybdate and nickel molybdate showing the highest OER activities.

Development of refractory armored silicon carbide by infrared transient liquid phase processing

NASA Astrophysics Data System (ADS)

Hinoki, Tatsuya; Snead, Lance L.; Blue, Craig A.

2005-12-01

Tungsten (W) and molybdenum (Mo) were coated on silicon carbide (SiC) for use as a refractory armor using a high power plasma arc lamp at powers up to 23.5 MW/m 2 in an argon flow environment. Both tungsten powder and molybdenum powder melted and formed coating layers on silicon carbide within a few seconds. The effect of substrate pre-treatment (vapor deposition of titanium (Ti) and tungsten, and annealing) and sample heating conditions on microstructure of the coating and coating/substrate interface were investigated. The microstructure was observed by scanning electron microscopy (SEM) and optical microscopy (OM). The mechanical properties of the coated materials were evaluated by four-point flexural tests. A strong tungsten coating was successfully applied to the silicon carbide substrate. Tungsten vapor deposition and pre-heating at 5.2 MW/m 2 made for a refractory layer containing no cracks propagating into the silicon carbide substrate. The tungsten coating was formed without the thick reaction layer. For this study, small tungsten carbide grains were observed adjacent to the interface in all conditions. In addition, relatively large, widely scattered tungsten carbide grains and a eutectic structure of tungsten and silicon were observed through the thickness in the coatings formed at lower powers and longer heating times. The strength of the silicon carbide substrate was somewhat decreased as a result of the processing. Vapor deposition of tungsten prior to powder coating helped prevent this degradation. In contrast, molybdenum coating was more challenging than tungsten coating due to the larger coefficient of thermal expansion (CTE) mismatch as compared to tungsten and silicon carbide. From this work it is concluded that refractory armoring of silicon carbide by Infrared Transient Liquid Phase Processing is possible. The tungsten armored silicon carbide samples proved uniform, strong, and capable of withstanding thermal fatigue testing.

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coffey, Greg W.; Meinhardt, Kerry D.; Joshi, Vineet V.

2015-03-01

The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce amore » uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that plating can be performed under conditions that would greatly reduce the quantity of intermetallics that form at the interface between the zirconium and U-10Mo; unlike roll bonding, the molten salt plating approach would allow for complete coverage of the U-10Mo foil with zirconium. When utilizing the experimental parameters developed for zirconium plating onto molybdenum, a uranium fluoride reaction product was formed at the Zr/U-10Mo interface. By controlling the initial plating potential, the uranium fluoride could be prevented; however, the targeted zirconium thickness (25 ±12.5 μm) could not be achieved while maintaining 100% coverage.« less

On carbide dissolution in an as-cast ASTM F-75 alloy.

PubMed

Caudillo, M; Herrera-Trejo, M; Castro, M R; Ramírez, E; González, C R; Juárez, J I

2002-02-01

The solution treatment of an as-cast ASTM F-75 alloy was investigated. Microstructural evolution was followed during thermal processing, in particular with regard to the content and type of carbides formed. To evidence any probable carbide transformations occurring during the heating stage, as well as to clarify their effect on the carbide dissolution kinetics, three heating rates were studied. Image analysis and scanning electron microscopy techniques were used for microstructural characterization. For the identification of precipitates, these were electrolytically extracted from the matrix and then analyzed by X-ray diffraction. It was found that the precipitates in the as-cast alloy were constituted by both a M(23)C(6) carbide and a sigma intermetallic phase. The M(23)C(6) carbide was the only phase identified in solution-treated specimens, regardless of the heating rate employed, which indicated that this carbide dissolved directly into the matrix without being transformed first into an M(6)C carbide, as reported in the literature. It was found that the kinetics of dissolution for the M(23)C(6) carbide decreased progressively during the solution treatment, and that it was sensitive to the heating rate, decreasing whenever the latter was decreased. Because the M(23)C(6) carbide was not observed to suffer a phase transformation prior to its dissolution into the matrix, the effect of the heating rate was associated to the morphological change occurred as the specimens were heated. The occurrence of the observed phases was analyzed with the aid of phase diagrams computed for the system Co-Cr-Mo-C. Copyright 2001 John Wiley & Sons, Inc. J Biomed Mater Res 59: 378-385, 2002

Crystallographic texture of straight-rolled ?-uranium foils via neutron and X-ray diffraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Einhorn, J. R.; Steiner, M. A.; Vogel, S. C.

The texture of recrystallized straight-rolled ?-uranium foils, a component in prospective irradiation target designs for medical isotope production, has been measured by neutron diffraction, as well as X-ray diffraction using both Cu and Mo sources. Variations in the penetration depth of neutron and X-ray radiation allow for determination of both the bulk and surface textures. The bulk ?-uranium foil texture is similar to the warm straight-rolled plate texture, with the addition of a notable splitting of the (001) poles along the transverse direction. The surface texture of the foils is similar to the bulk, with an additional (001) texture componentmore » that is oriented between the rolling and normal directions. Differences between the surface and bulk textures are expected to arise from shear forces during the rolling process and the influence that distinct strain histories have on subsequent texture evolution during recrystallization.« less

Crystallographic texture of straight-rolled ?-uranium foils via neutron and X-ray diffraction

DOE PAGES

Einhorn, J. R.; Steiner, M. A.; Vogel, S. C.; ...

2017-05-25

The texture of recrystallized straight-rolled ?-uranium foils, a component in prospective irradiation target designs for medical isotope production, has been measured by neutron diffraction, as well as X-ray diffraction using both Cu and Mo sources. Variations in the penetration depth of neutron and X-ray radiation allow for determination of both the bulk and surface textures. The bulk ?-uranium foil texture is similar to the warm straight-rolled plate texture, with the addition of a notable splitting of the (001) poles along the transverse direction. The surface texture of the foils is similar to the bulk, with an additional (001) texture componentmore » that is oriented between the rolling and normal directions. Differences between the surface and bulk textures are expected to arise from shear forces during the rolling process and the influence that distinct strain histories have on subsequent texture evolution during recrystallization.« less

Bioremediation of uranium-contaminated groundwater: a systems approach to subsurface biogeochemistry.

PubMed

Williams, Kenneth H; Bargar, John R; Lloyd, Jonathan R; Lovley, Derek R

2013-06-01

Adding organic electron donors to stimulate microbial reduction of highly soluble U(VI) to less soluble U(IV) is a promising strategy for immobilizing uranium in contaminated subsurface environments. Studies suggest that diagnosing the in situ physiological status of the subsurface community during uranium bioremediation with environmental transcriptomic and proteomic techniques can identify factors potentially limiting U(VI) reduction activity. Models which couple genome-scale in silico representations of the metabolism of key microbial populations with geochemical and hydrological models may be able to predict the outcome of bioremediation strategies and aid in the development of new approaches. Concerns remain about the long-term stability of sequestered U(IV) minerals and the release of co-contaminants associated with Fe(III) oxides, which might be overcome through targeted delivery of electrons to select microorganisms using in situ electrodes. Copyright © 2012 Elsevier Ltd. All rights reserved.

Pressure-induced structural phase transition in transition metal carbides TMC (TM = Ru, Rh, Pd, Os, Ir, Pt): a DFT study

NASA Astrophysics Data System (ADS)

Manikandan, M.; Rajeswarapalanichamy, R.; Iyakutti, K.

2018-03-01

First-principles calculations based on density functional theory was performed to analyse the structural stability of transition metal carbides TMC (TM = Ru, Rh, Pd, Os, Ir, Pt). It is observed that zinc-blende phase is the most stable one for these carbides. Pressure-induced structural phase transition from zinc blende to NiAs phase is predicted at the pressures of 248.5 GPa, 127 GPa and 142 GPa for OsC, IrC and PtC, respectively. The electronic structure reveals that RuC exhibits a semiconducting behaviour with an energy gap of 0.7056 eV. The high bulk modulus values of these carbides indicate that these metal carbides are super hard materials. The high B/G value predicts that the carbides are ductile in their most stable phase.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pierce, D. T.; Coughlin, D. R.; Williamson, D. L.

Quenching and partitioning (Q&P) produces steel microstructures with martensite and austenite that exhibit promising property combinations for third generation advanced high strength steels. Understanding the kinetics of reactions that compete for available carbon, such as carbide formation, is critical for alloying and processing design and achieving austenite enrichment and retention during Q&P. Mössbauer effect spectroscopy (MES) was used to characterize Q&P microstructures in a 0.38C-1.54Mn-1.48Si wt.% steel after quenching to 225 °C and partitioning at 400 °C for 10 or 300 s, with an emphasis on transition carbides. The recoilless fraction for η-carbide was calculated and a correction for saturationmore » of the MES absorption spectrum was applied, making quantitative measurements of small amounts of η-carbide, including non-stoichiometric η-carbide, possible in Q&P microstructures. Complementary transmission electron microscopy confirmed the presence of η-carbides, and MES and X-ray diffraction were used to characterize the austenite. The amount of η-carbide formed during Q&P ranged from 1.4 to 2.4 at.%, accounting for a substantial portion (~24% to 41%) of the bulk carbon content of the steel. The amount (5.0 at.%) of η-carbide that formed after quenching and tempering (Q&T) at 400 °C for 300 s was significantly greater than after partitioning at 400 °C for 300 s (2.4 at.%), suggesting that carbon partitioning from martensite to austenite occurs in conjunction with η-carbide formation during Q&P in these specimens.« less

Parallel computation of multigroup reactivity coefficient using iterative method

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter

2013-09-01

One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

Extreme-Environment Silicon-Carbide (SiC) Wireless Sensor Suite

NASA Technical Reports Server (NTRS)

Yang, Jie

2015-01-01

Phase II objectives: Develop an integrated silicon-carbide wireless sensor suite capable of in situ measurements of critical characteristics of NTP engine; Compose silicon-carbide wireless sensor suite of: Extreme-environment sensors center, Dedicated high-temperature (450 deg C) silicon-carbide electronics that provide power and signal conditioning capabilities as well as radio frequency modulation and wireless data transmission capabilities center, An onboard energy harvesting system as a power source.

Low Temperature Processing of Boron Carbide Cement Composite for Tough, Wear Resistant Structures

DTIC Science & Technology

1997-12-15

TITLE AND SUBTITLE Low Temperature Processing of Boron Carbide Cement Composite for Tough, Wear Resistant Structures 6. AUTHOR(S) Kristen J. Law...project has developed a low temperature polymer ceramic composite consisting of boron carbide layers bonded by cement, laminated with polymer...composite have been shown to compare favorably to those of partially sintered boron carbide. Applications for this material have been identified in

Mechanical Testing of Silicon Carbide on MISSE-7

DTIC Science & Technology

2012-07-15

JS) ii Abstract Silicon carbide ( SiC ) mechanical test specimens were included on the second Optical and Reflector Materials Experiment (ORMatE II...2. Vendor 2 EFS Weibull Results (normalized to Extra Disks Weibull parameters) 12 1. Introduction Silicon carbide ( SiC ) mechanical test...AEROSPACE REPORT NO ATR-2012(8921)-5 Mechanical Testing of Silicon Carbide on MISSE-7 Jul> 15. 2012 David B. Witkin Space Materials Laboratory

Development and Performance of Boron Carbide-Based Smoke Compositions

DTIC Science & Technology

2013-03-06

DOI: 10.1002/prep.201200166 Development and Performance of Boron Carbide -Based Smoke Compositions Anthony P. Shaw,*[a] Jay C. Poret,[a] Robert A...volatilized and recondense to give smoke. Boron carbide was recognized as a pyrotechnic fuel many years ago, but it has since been overlooked. A 1961...Abstract : Pyrotechnic smoke compositions for visual ob- scuration containing boron carbide , potassium nitrate, po- tassium chloride, and various lubricants

Salt flux synthesis of single and bimetallic carbide nanowires

NASA Astrophysics Data System (ADS)

Leonard, Brian M.; Waetzig, Gregory R.; Clouser, Dale A.; Schmuecker, Samantha M.; Harris, Daniel P.; Stacy, John M.; Duffee, Kyle D.; Wan, Cheng

2016-07-01

Metal carbide compounds have a broad range of interesting properties and are some of the hardest and highest melting point compounds known. However, their high melting points force very high reaction temperatures and thus limit the formation of high surface area nanomaterials. To avoid the extreme synthesis temperatures commonly associated with these materials, a new salt flux technique has been employed to reduce reaction temperatures and form these materials in the nanometer regime. Additionally, the use of multiwall carbon nanotubes as a reactant further reduces the diffusion distance and provides a template for the final carbide materials. The metal carbide compounds produced through this low temperature salt flux technique maintain the nanowire morphology of the carbon nanotubes but increase in size to ˜15-20 nm diameter due to the incorporation of metal in the carbon lattice. These nano-carbides not only have nanowire like shape but also have much higher surface areas than traditionally prepared metal carbides. Finally, bimetallic carbides with composition control can be produced with this method by simply using two metal precursors in the reaction. This method provides the ability to produce nano sized metal carbide materials with size, morphology, and composition control and will allow for these compounds to be synthesized and studied in a whole new size and temperature regime.

The effect of carbide precipitation on the hydrogen-enhanced fracture behavior of alloy 690

DOE Office of Scientific and Technical Information (OSTI.GOV)

Symons, D.M.

1998-04-01

Alloy 690 is susceptible to hydrogen embrittlement where hydrogen reduces the ductility and causes the fracture morphology to change to predominantly intergranular. The role of carbide precipitation in the embrittlement behavior is not well defined. The objective of this work is to understand the effect of intergranular carbide precipitation on the hydrogen embrittlement of alloy 690. The work reported herein used tensile and compact-tension specimens in both the solution-annealed condition (minimal grain-boundary carbide precipitation) and in the solution-annealed condition followed by an aging treatment to precipitate grain-boundary carbides. By performing the mechanical tests on materials in both uncharged and hydrogen-chargedmore » conditions, it was possible to evaluate the degree of embrittlement as a function of the carbide precipitation. It is shown that the embrittlement due to hydrogen increased as the material was aged to allow grain-boundary carbide precipitation. It is proposed that the increase in embrittlement was caused by increased hydrogen at the carbide/matrix interface due to the trapping and increased stresses at the precipitate interface, which developed from strain incompatibility of the precipitate with the matrix. It is further shown that increasing the hydrostatic stress increased the tendency for intergranular fracture, as is consistent with other nickel-base alloys.« less

XPS, AES and friction studies of single-crystal silicon carbide

NASA Technical Reports Server (NTRS)

Miyoshi, K.; Buckley, D. H.

1982-01-01

The surface chemistry and friction behavior of a single crystal silicon carbide surface parallel to the 0001 plane in sliding contact with iron at various temperatures to 1500 C in a vacuum of 3 x 10 nPa are investigated using X-ray photoelectron and Auger electron spectroscopy. Results show that graphite and carbide-type carbon are seen primarily on the silicon carbide surface in addition to silicon at temperatures to 800 C by both types of spectroscopy. The coefficients of friction for iron sliding against a silicon carbide surface parallel to the 0001 plane surface are found to be high at temperatures up to 800 C, with the silicon and carbide-type carbon at maximum intensity in the X-ray photoelectron spectroscopy at 800 C. The concentration of the graphite increases rapidly on the surface as the temperature is increased above 800 C, while the concentrations of the carbide-type carbon and silicon decrease rapidly and this presence of graphite is accompanied by a significant decrease in friction. Preheating the surfaces to 1500 C also gives dramatically lower coefficients of friction when reheating in the sliding temperature range of from room temperature to 1200 C, with this reduction in friction due to the graphite layer on the silicon carbide surface.

Method for preparing boron-carbide articles

DOEpatents

Benton, S.T.; Masters, D.R.

1975-10-21

The invention is directed to the preparation of boron carbide articles of various configurations. A stoichiometric mixture of particulate boron and carbon is confined in a suitable mold, heated to a temperature in the range of about 1250 to 1500$sup 0$C for effecting a solid state diffusion reaction between the boron and carbon for forming the boron carbide (B$sub 4$C), and thereafter the resulting boron-carbide particles are hot-pressed at a temperature in the range of about 1800 to 2200$sup 0$C and a pressure in the range of about 1000 to 4000 psi for densifying and sintering the boron carbide into the desired article.

Low Cost Fabrication of Silicon Carbide Based Ceramics and Fiber Reinforced Composites

NASA Technical Reports Server (NTRS)

Singh, M.; Levine, S. R.

1995-01-01

A low cost processing technique called reaction forming for the fabrication of near-net and complex shaped components of silicon carbide based ceramics and composites is presented. This process consists of the production of a microporous carbon preform and subsequent infiltration with liquid silicon or silicon-refractory metal alloys. The microporous preforms are made by the pyrolysis of a polymerized resin mixture with very good control of pore volume and pore size thereby yielding materials with tailorable microstructure and composition. Mechanical properties (elastic modulus, flexural strength, and fracture toughness) of reaction-formed silicon carbide ceramics are presented. This processing approach is suitable for various kinds of reinforcements such as whiskers, particulates, fibers (tows, weaves, and filaments), and 3-D architectures. This approach has also been used to fabricate continuous silicon carbide fiber reinforced ceramic composites (CFCC's) with silicon carbide based matrices. Strong and tough composites with tailorable matrix microstructure and composition have been obtained. Microstructure and thermomechanical properties of a silicon carbide (SCS-6) fiber reinforced reaction-formed silicon carbide matrix composites are discussed.

Laser-Induced Ionization Efficiency Enhancement On A Filament For Thermal Ionization Mass Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siegfried, M.

2015-10-14

The evaluation of trace Uranium and Plutonium isotope ratios for nanogram to femtogram material quantities is a vital tool for nuclear counter-proliferation and safeguard activities. Thermal Ionization Mass Spectrometry (TIMS) is generally accepted as the state of the art technology for highly accurate and ultra-trace measurements of these actinide ratios. However, the very low TIMS ionization yield (typically less than 1%) leaves much room for improvement. Enhanced ionization of Nd and Sm from a TIMS filament was demonstrated using wavelength resonance with a nanosecond (pulse width) laser operating at 10 Hz when light was directed toward the filament.1 For thismore » study, femtosecond and picosecond laser capabilities were to be employed to study the dissociation and ionization mechanisms of actinides/lanthanides and measure the enhanced ionization of the metal of interest. Since the underlying chemistry of the actinide/lanthanide carbides produced and dissociated on a TIMS filament is not well understood, the experimental parameters affecting the photodissociation and photoionization with one and two laser beams were to be investigated.« less

Numerical simulation of systems of shear bands in ductile metal with inclusions

NASA Astrophysics Data System (ADS)

Plohr, Jeeyeon

2017-06-01

We develop a method for numerical simulations of high strain-rate loading of mesoscale samples of ductile metal with inclusions. Because of its small-scale inhomogeneity, the composite material is prone to localized shear deformation. This method employs the Generalized Method of Cells to ensure that the micro mechanical behavior of the metal and inclusions is reflected properly in the behavior of the composite at the mesoscale. To find the effective plastic strain rate when shear bands are present, we extend and apply the analytic and numerical analysis of shear bands of Glimm, Plohr, and Sharp. Our tests of the method focus on the stress/strain response in uniaxial-strain flow, both compressive and tensile, of depleted uranium metal containing silicon carbide inclusions. In results, we verify the elevated temperature and thermal softening at shear bands in our simulations of pure DU and DU/SiC composites. We also note that in composites, due the asymmetry caused by the inclusions, shear band form at different times in different subcells. In particular, in the subcells near inclusions, shear band form much earlier than they do in pure DU.

Recent advances in computational actinoid chemistry.

PubMed

Wang, Dongqi; van Gunsteren, Wilfred F; Chai, Zhifang

2012-09-07

We briefly review advances in computational actinoid (An) chemistry during the past ten years in regard to two issues: the geometrical and electronic structures, and reactions. The former addresses the An-O, An-C, and M-An (M is a metal atom including An) bonds in the actinoid molecular systems, including actinoid oxo and oxide species, actinoid-carbenoid, dinuclear and diatomic systems, and the latter the hydration and ligand exchange, the disproportionation, the oxidation, the reduction of uranyl, hydroamination, and the photolysis of uranium azide. Concerning their relevance to the electronic structures and reactions of actinoids and their importance in the development of an advanced nuclear fuel cycle, we also mentioned the work on actinoid carbides and nitrides, which have been proposed to be candidates of the next generation of nuclear fuel, and the oxidation of PuO(x), which is important to understand the speciation of actinoids in the environment, followed by a brief discussion on the urgent need for a heavier involvement of computational actinoid chemistry in developing advanced reprocessing protocols of spent nuclear fuel. The paper is concluded with an outlook.

The structure and function of supported molybdenum nitride and molybdenum carbide hydrotreating catalysts

NASA Astrophysics Data System (ADS)

Dolce, Gregory Martin

1997-11-01

A series of gamma-Alsb2Osb3 supported molybdenum nitrides and carbides were prepared by the temperature programmed reaction of supported molybdates with ammonia and methane/hydrogen mixtures, respectively. In the first part of this research, the effects of synthesis heating rates and molybdenum loading on the catalytic properties of the materials were examined. A significant amount of excess carbon was deposited on the surface of the carbides during synthesis. The materials consisted of small particles which were very highly dispersed. Oxygen chemisorption indicated that the nitride particles may have been two-dimensional. The dispersion of the carbides, however, appeared to decrease as the loading increased. The catalysts were evaluated for hydrodenitrogenation (HDN), hydrodesulfurization (HDS), and hydrodeoxygenation (HDO). The molybdenum loading had the largest effect on the activity of the materials. For the nitrides, the HDN and HDS activities were inverse functions of the loading. This suggested that the most active HDN and HDS sites were located at the perimeter of the two-dimensional particles. The HDN and HDS activities of the carbides followed the same trend as the oxygen uptake. This result suggested that oxygen titrated the active sites on the supported carbides. Selected catalysts were evaluated for methylcarbazole HDN, dibenzothiophene HDS, and dibenzofuran HDO. The activity and selectivity of the nitrides and carbides were competitive with a presulfided commercial catalyst. In the second part of this work, a series of supported nitrides and carbides were prepared using a wider range of loadings (5-30 wt% Mo). Thermogravimetric analysis was used to determine the temperature at which excess carbon was deposited on the carbides. By modifying the synthesis parameters, the deposition of excess carbon was effectively inhibited. The dispersions of the supported nitrides and carbides were constant and suggested that the materials consisted of two-dimensional raft-like particles. The HDN activity of the nitrides decreased as the loading increased, while that of the carbides was relatively constant. Carbon monoxide and methylamine adsorbed on the same types of sites on the nitrides and carbides. Infrared spectroscopy and temperature programmed desorption revealed that some methylamine underwent HDN on the nitrides and carbides. Carbon monoxide appeared to adsorb on two types of sites. One type of site adsorbed CO which desorbed upon heating while the other type of site adsorbed CO which dissociated when the material was heated. The relative amounts of desorbed CO and methylamine scaled with the activity of the nitrides suggesting that CO and methylamine titrated the active sites. It appeared that the active sites of the supported carbides were different from those on the supported nitrides. It was proposed that the active sites on the supported nitrides were at the perimeter of the two-dimensional particles while the active sites of the carbides were "on top" of the particles.

Russian Experience in the Regulatory Supervision of the Uranium Legacy Sites - 12441

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kiselev, M.F.; Romanov, V.V.; Shandala, N.K.

2012-07-01

Management of the uranium legacy is accompanied with environmental impact intensity of which depends on the amount of the waste generated, the extent of that waste localization and environmental spreading. The question is: how hazardous is such impact on the environment and human health? The criterion for safety assurance is adequate regulation of the uranium legacy. Since the establishment of the uranium industry, the well done regulatory system operates in the FMBA of Russia. Such system covers inter alia, the uranium legacy. This system includes the extent laboratory network of independent control and supervision, scientific researches, regulative practices. The currentmore » Russian normative and legal basis of the regulation and its application practice has a number of problems relating to the uranium legacy, connected firstly with the environmental remediation. To improve the regulatory system, the urgent tasks are: -To introduce the existing exposure situation into the national laws and standards in compliance with the ICRP system. - To develop criteria for site remediation and return, by stages, to uncontrolled uses. The similar criteria have been developed within the Russian-Norwegian cooperation for the purpose of remediation of the sites for temporary storage of SNF and RW. - To consider possibilities and methods of optimization for the remediation strategies under development. - To separate the special category - RW resulted from uranium ore mining and dressing. The current Russian RW classification is based on the waste subdivision in terms of the specific activities. Having in mind the new RW-specific law, we receive the opportunity to separate some special category - RW originated from the uranium mining and milling. Introduction of such category can simplify significantly the situation with management of waste of uranium mining and milling processes. Such approach is implemented in many countries and approved by IAEA. The category of 'RW originated from uranium mining and milling' is to be introduced as the legal acts and regulatory documents. The recent ICRP recommendations provide the flexible approaches for solving of such tasks. The FMBA of Russia recognizes the problems of radiation safety assurance related to the legacy of the former USSR in the uranium mining industry. Some part of the regulatory problems assumes to be solved within the EurAsEC inter-state target program 'Reclamation of the territories of the EurAsEC member states affected by the uranium mining and milling facilities'. Using the example of the uranium legacy sites in Kyrgyz and Tajikistan which could result in the tran-boundary disasters and require urgent reclamation, the experience will be gained to be used in other states as well. Harmonization of the national legislations and regulative documents on radiation safety assurance is envisaged. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mehta, Vrajesh; Maillot, Fabien; Wang, Zheming

Uranyl phosphate solids are often found with uranium ores, and their low solubility makes them promising target phases for in situ remediation of uranium-contaminated subsurface environments. The products and solubility of uranium(VI) precipitated with phosphate can be affected by the pH, dissolved inorganic carbon (DIC) concentration, and co-solute composition (e.g. Na+/Ca2+) of the groundwater. Batch experiments were performed to study the effect of these parameters on the products and extent of uranium precipitation induced by phosphate addition. In the absence of co-solute cations, chernikovite [H3O(UO2)(PO4)•3H2O] precipitated despite uranyl orthophosphate [(UO2)3(PO4)2•4H2O] being thermodynamically more favorable under certain conditions. As determined usingmore » X-ray diffraction, electron microscopy, and laser induced fluorescence spectroscopy, the presence of Na+ or Ca2+ as a co-solute led to the precipitation of sodium autunite ([Na2(UO2)2(PO4)2] and autunite [Ca(UO2)2(PO4)2]), which are structurally similar to chernikovite. In the presence of sodium, the dissolved U(VI) concentrations were generally in agreement with equilibrium predictions of sodium autunite solubility. However, in the calcium-containing systems, the observed concentrations were below the predicted solubility of autunite, suggesting the possibility of uranium adsorption to or incorporation in a calcium phosphate precipitate in addition to the precipitation of autunite.« less

Data Compilation for AGR-3/4 Designed-to-Fail (DTF) Fuel Particle Batch LEU04-02DTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D; Miller, James Henry

2008-10-01

This document is a compilation of coating and characterization data for the AGR-3/4 designed-to-fail (DTF) particles. The DTF coating is a high density, high anisotropy pyrocarbon coating of nominal 20 {micro}m thickness that is deposited directly on the kernel. The purpose of this coating is to fail early in the irradiation, resulting in a controlled release of fission products which can be analyzed to provide data on fission product transport. A small number of DTF particles will be included with standard TRISO driver fuel particles in the AGR-3 and AGR-4 compacts. The ORNL Coated Particle Fuel Development Laboratory 50-mm diametermore » fluidized bed coater was used to coat the DTF particles. The coatings were produced using procedures and process parameters that were developed in an earlier phase of the project as documented in 'Summary Report on the Development of Procedures for the Fabrication of AGR-3/4 Design-to-Fail Particles', ORNL/TM-2008/161. Two coating runs were conducted using the approved coating parameters. NUCO425-06DTF was a final process qualification batch using natural enrichment uranium carbide/uranium oxide (UCO) kernels. After the qualification run, LEU04-02DTF was produced using low enriched UCO kernels. Both runs were inspected and determined to meet the specifications for DTF particles in section 5 of the AGR-3 & 4 Fuel Product Specification (EDF-6638, Rev.1). Table 1 provides a summary of key properties of the DTF layer. For comparison purposes, an archive sample of DTF particles produced by General Atomics was characterized using identical methods. This data is also summarized in Table 1.« less

Geoantineutrino spectrum and slow nuclear burning on the boundary of the liquid and solid phases of the Earth's core

NASA Astrophysics Data System (ADS)

Rusov, V. D.; Pavlovich, V. N.; Vaschenko, V. N.; Tarasov, V. A.; Zelentsova, T. N.; Bolshakov, V. N.; Litvinov, D. A.; Kosenko, S. I.; Byegunova, O. A.

2007-09-01

We give an alternative description of the data produced in the KamLAND experiment. Assuming the existence of a natural nuclear reactor on the boundary of the liquid and solid phases of the Earth's core, a geoantineutrino spectrum is obtained. This assumption is based on the experimental results of V. Anisichkin and his collaborators on the interaction of uranium dioxide and uranium carbide with iron-nickel and silica-alumina melts at high pressure (5-10 GPa) and temperature (1600-2200°C), which led to the proposal of the existence of an actinide shell in the Earth's core. We describe the operating mechanism of this reactor as solitary waves of nuclear burning in 238U and/or 232Th medium, in particular, as neutron fission progressive waves of Feoktistov and/or Teller et al. type. Next, we propose a simplified model for the accumulation and burn-up kinetics in Feoktistov's U-Pu fuel cycle. We also apply this model for numerical simulations of neutron fission wave in a two-phase UO2/Fe medium on the surface of the Earth's solid core. The proposed georeactor model offers a mechanism for the generation of 3He. The 3He/4He distribution in the Earth's interior is calculated, which in turn can be used as a natural quantitative criterion of the georeactor thermal power. Finally, we give a tentative estimation of the geoantineutrino intensity and spectrum on the Earth's surface. For this purpose we use the O'Nions et al. geochemical model of mantle differentiation and crust growth complemented by a nuclear energy source (georeactor with power of 30 TW).

Abrasive slurry composition for machining boron carbide

DOEpatents

Duran, Edward L.

1985-01-01

An abrasive slurry particularly suited for use in drilling or machining boron carbide consists essentially of a suspension of boron carbide and/or silicon carbide grit in a carrier solution consisting essentially of a dilute solution of alkylaryl polyether alcohol in octyl alcohol. The alkylaryl polyether alcohol functions as a wetting agent which improves the capacity of the octyl alcohol for carrying the grit in suspension, yet without substantially increasing the viscosity of the carrier solution.

Abrasive slurry composition for machining boron carbide

DOEpatents

Duran, E.L.

1984-11-29

An abrasive slurry particularly suited for use in drilling or machining boron carbide consists essentially of a suspension of boron carbide and/or silicon carbide grit in a carrier solution consisting essentially of a dilute solution of alkylaryl polyether alcohol in octyl alcohol. The alkylaryl polyether alcohol functions as a wetting agent which improves the capacity of the octyl alcohol for carrying the grit in suspension, yet without substantially increasing the viscosity of the carrier solution.

Crystallography of in-situ transformations of the M 7C3 carbide in the cast Fe-Cr-Ni alloy

NASA Astrophysics Data System (ADS)

Kraposhin, V. S.; Kondrat'ev, S. Yu.; Talis, A. L.; Anastasiadi, G. P.

2017-03-01

In the process of holding of the cast heat-resistant Fe-Cr-Ni (0.45C-25Cr-35Ni) alloy at 1150°C, the eutectic chromium carbide present in its structure undergoes a gradual transition M 7C3 → M 23C6. The gradual formation of domains of the M 23C6 carbide inside the particles of the M 7C3 carbide makes it possible to assume that the observed phase transition is the well-known carbide transformation of the in situ type. The mechanism of the in situ transformation of the crystal structure of the carbide from M 7C3 into M 23C6 with a change in the number of nearest metal neighbors of carbon atoms is explained within the previously developed combinatory model of polymorphic transitions in the metals.

The dependence of carbide morphology on grain boundary character in the highly twinned Alloy 690

NASA Astrophysics Data System (ADS)

Li, Hui; Xia, Shuang; Zhou, Bangxin; Chen, Wenjue; Hu, Changliang

2010-04-01

The dependence of morphology of grain boundary carbides on grain boundary character in Alloy 690 (Ni-30Cr-10Fe, mass fraction, %) with high fraction of low Σ coincidence site lattice (CSL) grain boundaries was investigated by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Some of the surface grains were removed by means of deep etching. It was observed that carbides grow dendritically at grain boundaries. The carbide bars observed near incoherent twin boundaries and twin related Σ9 grain boundaries are actually secondary dendrites of the carbides on these boundaries. Higher order dendrites could be observed on random grain boundaries, however, no bar-like dendrites were observed near Σ27 grain boundaries and random grain boundaries. The morphology difference of carbides precipitated at grain boundaries with different characters is discussed based on the experimental results in this paper.

Eutectic equilibria in the quaternary system Fe-Cr-Mn-C

NASA Technical Reports Server (NTRS)

Nowotny, H.; Wayne, S.; Schuster, J. C.

1982-01-01

The constitution of the quaternary system, Fe-Cr-Mn-C and to a lesser extent of the quinary system, Fe-Cr-Mn-Al-C were examined for in situ composite alloy candidates. Multivariant eutectic compositions were determined from phase equilibria studies wherein M7C3 carbides (approximately 30% by volume) formed from the melt within gamma iron. An extended field of the hexagonal carbide, (Cr, Fe, Mn)7 C3, was found without undergoing transformation to the orthorhombic structure. Increasing stability for this carbide was found for higher ratios of Cr/Fe(+) Cr + Mn. Aluminum additions promoted a ferritic matrix while manganese favored the desired gamma austenitic matrix. In coexistence with the matrix phase, chromium enters preferentially the carbide phase while manganese distributes equally between the gamma matrix and the M7C3 carbide. The composition and lattice parameters of the carbide and matrix phases were determined to establish their respective stabilities.

Hugoniot equation of state and dynamic strength of boron carbide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grady, Dennis E.

Boron carbide ceramics have been particularly problematic in attempts to develop adequate constitutive model descriptions for purposes of analysis of dynamic response in the shock and impact environment. Dynamic strength properties of boron carbide ceramic differ uniquely from comparable ceramics. Furthermore, boron carbide is suspected, but not definitely shown, to undergoing polymorphic phase transformation under shock compression. In the present paper, shock-wave compression measurements conducted over the past 40 years are assessed for the purpose of achieving improved understanding of the dynamic equation of state and strength of boron carbide. In particular, attention is focused on the often ignored Losmore » Alamos National Laboratory (LANL) Hugoniot measurements performed on porous sintered boron carbide ceramic. The LANL data are shown to exhibit two compression anomalies on the shock Hugoniot within the range of 20–60 GPa that may relate to crystallographic structure transitions. More recent molecular dynamics simulations on the compressibility of the boron carbide crystal lattice reveal compression transitions that bear similarities to the LANL Hugoniot results. The same Hugoniot data are complemented with dynamic isentropic compression data for boron carbide extracted from Hugoniot measurements on boron carbide and copper granular mixtures. Other Hugoniot measurements, however, performed on near-full-density boron carbide ceramic differ markedly from the LANL Hugoniot data. These later data exhibit markedly less compressibility and tend not to show comparable anomalies in compressibility. Alternative Hugoniot anomalies, however, are exhibited by the near-full-density data. Experimental uncertainty, Hugoniot strength, and phase transformation physics are all possible explanations for the observed discrepancies. It is reasoned that experimental uncertainty and Hugoniot strength are not likely explanations for the observed differences. The notable mechanistic difference in the processes of shock compression between the LANL data and that of the other studies is the markedly larger inelastic deformation and dissipation experienced in the shock event brought about by compaction of the substantially larger porosity LANL test ceramics. High-pressure diamond anvil cell experiments reveal extensive amorphization, reasoned to be a reversion product of a higher-pressure crystallographic phase, which is a consequence of application of both high pressure and shear deformation to the boron carbide crystal structure. A dependence of shock-induced high-pressure phase transformation in boron carbide on the extent of shear deformation experienced in the shock process offers a plausible explanation for the differences observed in the LANL Hugoniot data on porous ceramic and that of other shock data on near-full-density boron carbide.« less

Molybdenum carbides, active and in situ regenerable catalysts in hydroprocessing of fast pyrolysis bio-oil

DOE PAGES

Choi, Jae -Soon; Zacher, Alan; Wang, Huamin; ...

2016-05-19

This paper describes properties of molybdenum carbides as a potential catalyst for fast pyrolysis bio-oil hydroprocessing. Currently, high catalyst cost, short catalyst lifetime, and lack of effective regeneration methods are hampering the development of this otherwise attractive renewable hydrocarbon technology. A series of metal-doped bulk Mo carbides were synthesized, characterized, and evaluated in sequential low-temperature stabilization and high-temperature deoxygenation of a pine-derived bio-oil. During a typical 60 h run, Mo carbides were capable of upgrading raw bio-oil to a level suitable for direct insertion into the current hydrocarbon infrastructure with residual oxygen content and total acid number of upgraded oilsmore » below 2 wt % and 0.01 mg KOH g –1, respectively. The performance was shown to be sensitive to the type of metal dopant, Ni-doped Mo carbides outperforming Co-, Cu-, or Ca-doped counterparts; a higher Ni loading led to a superior catalytic performance. No bulk oxidation or other significant structural changes were observed. Besides the structural robustness, another attractive property of Mo carbides was in situ regenerability. The effectiveness of regeneration was demonstrated by successfully carrying out four consecutive 60 h runs with a reductive decoking between two adjacent runs. These results strongly suggest that Mo carbides are a good catalyst candidate which could lead to a significant cost reduction in hydroprocessing bio-oils. Furthermore, we highlight areas for future research which will be needed to further understand carbide structure–function relationships and help design practical bio-oil upgrading catalysts based on Mo carbides.« less

Molybdenum Carbides, Active and In Situ Regenerable Catalysts in Hydroprocessing of Fast Pyrolysis Bio-Oil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Jae-Soon; Zacher, Alan H.; Wang, Huamin

We assessed molybdenum carbides as a potential catalyst for fast pyrolysis bio-oil hydroprocessing. Currently, high catalyst cost, short catalyst lifetime, and lack of effective regeneration methods are hampering the development of this otherwise attractive renewable hydrocarbon technology. A series of metal-doped bulk Mo carbides were synthesized, characterized and evaluated in sequential low-temperature stabilization and high-temperature deoxygenation of a pine-derived bio-oil. During a typical 60-h run, Mo carbides were capable of upgrading raw bio-oil to a level suitable for direct insertion into the current hydrocarbon infrastructure with residual oxygen content and total acid number of upgraded oils below 2 wt% andmore » 0.01 mg KOH g-1, respectively. The performance was shown to be sensitive to the type of metal dopant, Ni-doped Mo carbides outperforming Co-, Cu-, or Ca-doped counterparts; a higher Ni loading led to a superior catalytic performance. No bulk oxidation or other significant structural changes were observed. Besides the structural robustness, another attractive property of Mo carbides was in situ regenerability. The effectiveness of regeneration was demonstrated by successfully carrying out four consecutive 60-h runs with a reductive decoking between two adjacent runs. These results strongly suggest that Mo carbides are promising catalytic materials which could lead to a significant cost reduction in hydroprocessing bio-oils. This paper highlights areas for future research which will be needed to further understand carbide structure-function relationships and help design practical bio-oil upgrading catalysts based on Mo carbides.« less

Evaluation of titanium carbide metal matrix composites deposited via laser cladding

NASA Astrophysics Data System (ADS)

Cavanaugh, Daniel Thomas

Metal matrix composites have been widely studied in terms of abrasion resistance, but a particular material system may behave differently as particle size, morphology, composition, and distribution of the hardening phase varies. The purpose of this thesis was to understand the mechanical and microstructural effects of combining titanium carbide with 431 series stainless steel to create a unique composite via laser cladding, particularly regarding wear properties. The most predominant effect in increasing abrasion resistance, measured via ASTM G65, was confirmed to be volume fraction of titanium carbide addition. Macrohardness was directly proportional to the amount of carbide, though there was an overall reduction in individual particle microhardness after cladding. The reduction in particle hardness was obscured by the effect of volume fraction carbide and did not substantially contribute to the wear resistance changes. A model evaluating effective mean free path of the titanium carbide particles was created and correlated to the measured data. The model proved successful in linking theoretical mean free path to overall abrasion resistance. The effects of the titanium carbide particle distributions were limited, while differences in particle size were noticeable. The mean free path model did not correlate well with the particle size, but it was shown that the fine carbides were completely removed by the coarse abrasive particles in the ASTM G65 test. The particle morphology showed indications of influencing the wear mode, but no statistical reduction was observed in the volume loss figures. Future studies may more specifically focus on particle morphology or compositional effects of the carbide particles.

Molybdenum carbides, active and in situ regenerable catalysts in hydroprocessing of fast pyrolysis bio-oil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Jae -Soon; Zacher, Alan; Wang, Huamin

This paper describes properties of molybdenum carbides as a potential catalyst for fast pyrolysis bio-oil hydroprocessing. Currently, high catalyst cost, short catalyst lifetime, and lack of effective regeneration methods are hampering the development of this otherwise attractive renewable hydrocarbon technology. A series of metal-doped bulk Mo carbides were synthesized, characterized, and evaluated in sequential low-temperature stabilization and high-temperature deoxygenation of a pine-derived bio-oil. During a typical 60 h run, Mo carbides were capable of upgrading raw bio-oil to a level suitable for direct insertion into the current hydrocarbon infrastructure with residual oxygen content and total acid number of upgraded oilsmore » below 2 wt % and 0.01 mg KOH g –1, respectively. The performance was shown to be sensitive to the type of metal dopant, Ni-doped Mo carbides outperforming Co-, Cu-, or Ca-doped counterparts; a higher Ni loading led to a superior catalytic performance. No bulk oxidation or other significant structural changes were observed. Besides the structural robustness, another attractive property of Mo carbides was in situ regenerability. The effectiveness of regeneration was demonstrated by successfully carrying out four consecutive 60 h runs with a reductive decoking between two adjacent runs. These results strongly suggest that Mo carbides are a good catalyst candidate which could lead to a significant cost reduction in hydroprocessing bio-oils. Furthermore, we highlight areas for future research which will be needed to further understand carbide structure–function relationships and help design practical bio-oil upgrading catalysts based on Mo carbides.« less

64. INTERIOR VIEW OF THE CARBIDE COOLING SHED. VIEW IS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

64. INTERIOR VIEW OF THE CARBIDE COOLING SHED. VIEW IS SHOWING CALCIUM CARBIDE IN COOLING CARS ON THE FLOOR. DECEMBER 26, 1918. - United States Nitrate Plant No. 2, Reservation Road, Muscle Shoals, Muscle Shoals, Colbert County, AL

Titanium Carbide Bipolar Plate for Electrochemical Devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

LaConti, Anthony B.; Griffith, Arthur E.; Cropley, Cecelia C.

Titanium carbide comprises a corrosion resistant, electrically conductive, non-porous bipolar plate for use in an electrochemical device. The process involves blending titanium carbide powder with a suitable binder material, and molding the mixture, at an elevated temperature and pressure.