WorldWideScience

Sample records for uranium carbide dispersed

  1. Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

    International Nuclear Information System (INIS)

    Muenzer, W.A.

    1978-01-01

    A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

  2. A study on the formation of uranium carbide in an induction furnace

    International Nuclear Information System (INIS)

    Song, In Young; Lee, Yoon Sang; Kim, Eung Soo; Lee, Don Bae; Kim, Chang Kyu

    2005-01-01

    Uranium is a typical carbide-forming element. Three carbides, UC, U 2 C 3 and UC 2 , are formed in the uranium-carbon system. The most important of these as fuel is uranium monocarbide UC. It is well known that Uranium carbides can be obtained by three basic methods: 1) by reaction of uranium metal with carbon; 2) by reaction of uranium metal powder with gaseous hydrocarbons; 3) by reaction of uranium oxides with carbon. The use of uranium monocarbide, or materials based on it, has great prospects as fuel for nuclear reactors. It is quite possible that uranium dicarbide UC 2 may also acquire great importance as a fuel, particularly in dispersion fuel elements with graphite matrix. In the present study, uranium carbides are obtained by direct reaction of uranium metal with graphite in a high frequency induction furnace

  3. Fabrication of chamfered uranium-plutonium mixed carbide pellets

    International Nuclear Information System (INIS)

    Arai, Yasuo; Iwai, Takashi; Shiozawa, Kenichi; Handa, Muneo

    1985-10-01

    Chamfered uranium-plutonium mixed carbide pellets for high burnup irradiation test in JMTR were fabricated in glove boxes with purified argon gas. The size of die and punch in a press was decided from pellet densities and dimensions including the angle of chamfered parts. No chip or crack caused by adopting chamfered pellets was found in both pressing and sintering stages. In addition to mixed carbide pellets, uranium carbide pellets used as insulators were also successfully fabricated. (author)

  4. Study on niobium carbide dispersed superconducting tapes

    Energy Technology Data Exchange (ETDEWEB)

    Wada, H; Tachikawa, K [National Research Inst. for Metals, Tokyo (Japan); Oh' asa, M [Science Univ. of Tokyo (Japan)

    1977-11-01

    Niobium carbide (NbC) dispersed superconducting tapes have been fabricated by two metallurgical processes. In the first process, Ni-Nb-C alloys are directly arc melted and hot worked in air and the NbC phase is distributed in the form of fine discrete particles. In the second process, Ni-Nb and Ni-Nb-Cu alloys are arc melted, hot worked and subjected to solid-state carburization. NbC then precipitates along the grain boundaries, forming a network. The highest superconducting transition temperature attained is about 11 K. Taken together with the lattice parameter measurement, this indicates that NbC with a nearly perfect NaCl structure is formed in both processes. Measured values of the upper critical field, the critical current density and the volume fraction of the NbC phase are also discussed.

  5. Ternary carbide uranium fuels for advanced reactor design applications

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    1999-01-01

    Solid-solution mixed uranium/refractory metal carbides such as the pseudo-ternary carbide, (U, Zr, Nb)C, hold significant promise for advanced reactor design applications because of their high thermal conductivity and high melting point (typically greater than 3200 K). Additionally, because of their thermochemical stability in a hot-hydrogen environment, pseudo-ternary carbides have been investigated for potential space nuclear power and propulsion applications. However, their stability with regard to sodium and improved resistance to attack by water over uranium carbide portends their usefulness as a fuel for advanced terrestrial reactors. An investigation into processing techniques was conducted in order to produce a series of (U, Zr, Nb)C samples for characterization and testing. Samples with densities ranging from 91% to 95% of theoretical density were produced by cold pressing and sintering the mixed constituent carbides at temperatures as high as 2650 K. (author)

  6. Pilot production of 325 kg of uranium carbide

    International Nuclear Information System (INIS)

    Clozet, C.; Dessus, J.; Devillard, J.; Guibert, M.; Morlot, G.

    1969-01-01

    This report describes the pilot fabrication of uranium carbide rods to be mounted in bundles and assayed in two channels of the EL 4 reactor. The fabrication process includes: - elaboration of uranium carbide granules by carbothermic reduction of uranium dioxide; - electron bombardment melting and continuous casting of the granules; - machining of the raw ingots into rods of the required dimensions; finally, the rods will be piled-up to make the fuel elements. Both qualitative and quantitative results of this pilot production chain are presented and discussed. (authors) [fr

  7. URANIUM BISMUTHIDE DISPERSION IN MOLTEN METAL

    Science.gov (United States)

    Teitel, R.J.

    1959-10-27

    The formation of intermetallic bismuth compounds of thorium or uranium dispersed in a liquid media containing bismuth and lead is described. A bismuthide of uranium dispersed in a liquid metal medium is formed by dissolving uranium in composition of lead and bismuth containing less than 80% lead and lowering the temperature of the composition to a temperature below the point at which the solubility of uranium is exceeded and above the melting point of the composition.

  8. Thermal conductivity and emissivity measurements of uranium carbides

    International Nuclear Information System (INIS)

    Corradetti, S.; Manzolaro, M.; Andrighetto, A.; Zanonato, P.; Tusseau-Nenez, S.

    2015-01-01

    Highlights: • Thermal conductivity and emissivity measurements of uranium carbides were performed. • The tested materials are candidates as targets for radioactive ion beam production. • The results are correlated with the materials composition and microstructure. - Abstract: Thermal conductivity and emissivity measurements on different types of uranium carbide are presented, in the context of the ActiLab Work Package in ENSAR, a project within the 7th Framework Program of the European Commission. Two specific techniques were used to carry out the measurements, both taking place in a laboratory dedicated to the research and development of materials for the SPES (Selective Production of Exotic Species) target. In the case of thermal conductivity, estimation of the dependence of this property on temperature was obtained using the inverse parameter estimation method, taking as a reference temperature and emissivity measurements. Emissivity at different temperatures was obtained for several types of uranium carbide using a dual frequency infrared pyrometer. Differences between the analyzed materials are discussed according to their compositional and microstructural properties. The obtainment of this type of information can help to carefully design materials to be capable of working under extreme conditions in next-generation ISOL (Isotope Separation On-Line) facilities for the generation of radioactive ion beams.

  9. Reaction of uranium and plutonium carbides with austenitic steels

    International Nuclear Information System (INIS)

    Mouchnino, M.

    1967-01-01

    The reaction of uranium and plutonium carbides with austenitic steels has been studied between 650 and 1050 deg. C using UC, steel and (UPu)C, steel diffusion couples. The steels are of the type CN 18.10 with or without addition of molybdenum. The carbides used are hyper-stoichiometric. Tests were also carried out with UCTi, UCMo, UPuCTi and UPuCMo. Up to 800 deg. C no marked diffusion of carbon into stainless steel is observed. Between 800 and 900 deg. C the carbon produced by the decomposition of the higher carbides diffuses into the steel. Above 900 deg. C, decomposition of the monocarbide occurs according to a reaction which can be written schematically as: (U,PuC) + (Fe,Ni,Cr) → (U,Pu) Fe 2 + Cr 23 C 6 . Above 950 deg. C the behaviour of UPuCMo and that of the titanium (CN 18.12) and nickel (NC 38. 18) steels is observed to be very satisfactory. (author) [fr

  10. Gravimetric determination of carbon in uranium-plutonium carbide materials

    International Nuclear Information System (INIS)

    Kavanaugh, H.J.; Dahlby, J.W.; Lovell, A.P.

    1979-12-01

    A gravimetric method for determining carbon in uranium-plutonium carbide materials was developed to analyze six samples simultaneously. The samples are burned slowly in an oxygen atmosphere at approximately 900 0 C, and the gases generated are passed through Schuetze's oxidizing reagent (iodine pentoxide on silica gel) to assure quantitative oxidation of the CO to CO 2 . The CO 2 is collected on Ascarite and weighed. This method was tested using a tungsten carbide reference material (NBS-SRM-276) and a (U,Pu)C sample. For 42 analyses of the tungsten carbide, which has a certified carbon content of 6.09%, an average value of 6.09% was obtained with a standard deviation of 0.01 7 % or a relative standard deviation of 0.28%. For 17 analyses of the (U,Pu)C sample, an average carbon content of 4.97% was found with a standard deviation of 0.01 2 % or a relative standard deviation of 0.24%

  11. Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    2002-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

  12. Steady State Sputtering Yields and Surface Compositions of Depleted Uranium and Uranium Carbide bombarded by 30 keV Gallium or 16 keV Cesium Ions.

    Energy Technology Data Exchange (ETDEWEB)

    Siekhaus, W. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Teslich, N. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Weber, P. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-10-23

    Depleted uranium that included carbide inclusions was sputtered with 30-keV gallium ions or 16-kev cesium ions to depths much greater than the ions’ range, i.e. using steady-state sputtering. The recession of both the uranium’s and uranium carbide’s surfaces and the ion corresponding fluences were used to determine the steady-state target sputtering yields of both uranium and uranium carbide, i.e. 6.3 atoms of uranium and 2.4 units of uranium carbide eroded per gallium ion, and 9.9 uranium atoms and 3.65 units of uranium carbide eroded by cesium ions. The steady state surface composition resulting from the simultaneous gallium or cesium implantation and sputter-erosion of uranium and uranium carbide were calculated to be U₈₆Ga₁₄, (UC)₇₀Ga₃₀ and U₈₁Cs₉, (UC)₇₉Cs₂₁, respectively.

  13. Characterisation of nuclear dispersion fuels. The non-destructive examination of silicon carbide by selenium immersion

    Energy Technology Data Exchange (ETDEWEB)

    Ambler, J.F.R.; Ferguson, I.F.

    1974-07-15

    The non-destructive microscopic examination of silicon-carbide-coated spheres containing uranium carbide, which involves immersing the coated spheres in selenium, is particularly suited for the examination of flaws in the coats but it is not possible to measure coating thicknesses by this method. Some coats are found to be opaque and this is related to their porosity. (auth)

  14. Metallographic preparation of sintered oxides, carbides and nitrides of uranium and plutonium

    International Nuclear Information System (INIS)

    Martin, A.; Arles, L.

    1967-12-01

    We describe the methods of polishing, attack and coloring used at the section of plutonium base ceramics studies. These methods have stood the test of experience on the uranium and plutonium carbides, nitrides and carbonitrides as well on the mixed uranium and plutonium oxides. These methods have been particularly adapted to fit to the low dense and sintered samples [fr

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

    International Nuclear Information System (INIS)

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

    2013-01-01

    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 HNO 3 concentration

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

    Science.gov (United States)

    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.

  17. Uranium Dispersion and Dosimetry (UDAD) Code

    International Nuclear Information System (INIS)

    Momeni, M.H.; Yuan, Y.; Zielen, A.J.

    1979-05-01

    The Uranium Dispersion and Dosimetry (UDAD) Code provides estimates of potential radiation exposure to individuals and to the general population in the vicinity of a uranium processing facility. The UDAD Code incorporates the radiation dose from the airborne release of radioactive materials, and includes dosimetry of inhalation, ingestion, and external exposures. The removal of raioactive particles from a contaminated area by wind action is estimated, atmospheric concentrations of radioactivity from specific sources are calculated, and source depletion as a result of deposition, fallout, and ingrowth of radon daughters are included in a sector-averaged Gaussian plume dispersion model. The average air concentration at any given receptor location is assumed to be constant during each annual release period, but to increase from year to year because of resuspension. Surface contamination and deposition velocity are estimated. Calculation of the inhalation dose and dose rate to an individual is based on the ICRP Task Group Lung Model. Estimates of the dose to the bronchial epithelium of the lung from inhalation of radon and its short-lived daughters are calculated based on a dose conversion factor from the BEIR report. External radiation exposure includes radiation from airborne radionuclides and exposure to radiation from contaminated ground. Terrestrial food pathways include vegetation, meat, milk, poultry, and eggs. Internal dosimetry is based on ICRP recommendations. In addition, individual dose commitments, population dose commitments, and environmental dose commitments are computed. This code also may be applied to dispersion of any other pollutant

  18. Diamond dispersed cemented carbide produced without using ultra high pressure equipment

    International Nuclear Information System (INIS)

    Moriguchi, H.; Tsuzuki, K.; Ikegaya, A.

    2001-01-01

    We have developed a composite material of dispersed diamond particles in cemented carbide without using ultra high pressure equipment. The developed diamond dispersed cemented carbide combines the excellent properties of cemented carbide with diamond and also provides 1.5 times improved fracture toughness over that of cemented carbide. They also show 10 times higher wear resistance over that of cemented carbide in a wear resistance test against bearing steel, and 5 times greater grindability than diamond compacts. Because ultra high pressure equipment is not used to produce the developed material, large compacts over 100 mm in diameter can be manufactured. The developed material showed 10-25 times higher wear resistance in real use as wear-resistant tools such as centerless blades and work-rests. (author)

  19. Plasma spraying process of disperse carbides for spraying and facing

    International Nuclear Information System (INIS)

    Blinkov, I.V.; Vishnevetskaya, I.A.; Kostyukovich, T.G.; Ostapovich, A.O.

    1989-01-01

    A possibility to metallize carbides in plasma of impulsing capacitor discharge is considered. Powders granulation occurs during plasma spraying process, ceramic core being completely capped. X-ray phase and chemical analyses of coatings did not show considerable changes of carbon content in carbides before and after plasma processing. This distinguishes the process of carbides metallization in impulsing plasma from the similar processing in arc and high-frequency plasma generator. Use of powder composites produced in the impulsing capacitor discharge, for plasma spraying and laser facing permits 2-3 times increasing wear resistance of the surface layer as against the coatings produced from mechanical powders mixtures

  20. Present status of uranium-plutonium mixed carbide fuel development for LMFBRs

    International Nuclear Information System (INIS)

    Handa, Muneo; Suzuki, Yasufumi

    1984-01-01

    The feature of carbide fuel is that it has the doubling time as short as about 13 years, that is, close to one half as compared with oxide fuel. The development of the carbide fuel in the past 10 years has been started in amazement. Especially in the program of new fuel development in USA started in 1974, He and Na bond fuel attained the burnup of 16 a/o without causing the breaking of cladding tubes. In 1984, the irradiation of the assembly composed of 91 fuel pins in the FFTF is expected. On the other hand in Japan, the fuel research laboratory was constructed in 1974 in the Oarai Laboratory, Japan Atomic Energy Research Institute, to carry out the studies on carbide fuel. In the autumn of 1982, two carbide fuel pins with different chemical composition have been successfully made. Accordingly, the recent status of the development is explained. The uranium-plutonium mixed carbide fuel is suitable to liquid metal-cooled fast breeder reactors because of large heat conductivity and the high density of nuclear fission substances. The thermal and nuclear characteristics of carbide fuel, the features of the reactor core using carbide fuel, the chemical and mechanical interaction of fuel and cladding tubes, the selection of bond materials, the manufacturing techniques for the fuel, the development of the analysis code for fuel behavior, and the research and development of carbide fuel in Japan are described. (Kako, I.)

  1. Review of thermal expansion and density of uranium and plutonium carbides

    International Nuclear Information System (INIS)

    Andrew, J.F.; Latimer, T.W.

    1975-07-01

    The published literature on linear thermal expansion and density of uranium and plutonium carbide nuclear fuels, including UC, PuC, (U,Pu)C, U 2 C 3 , Pu 2 C 3 , and (U,Pu) 2 C 3 , is critically reviewed. Recommended values are given in tabular form and additional experimental studies needed for completeness are outlined. 16 tables, 52 references

  2. Quantitative phase analysis of uranium carbide from x-ray diffraction data using the Rietveld method

    International Nuclear Information System (INIS)

    Singh Mudher, K.D.; Krishnan, K.

    2003-01-01

    Quantitative phase analysis of a uranium carbide sample was carried out from the x-ray diffraction data by Rietveld profile fitting method. The method does not require the addition of any reference material. The percentage of UC, UC 2 and UO 2 phases in the sample were determined. (author)

  3. Equation of state and transport properties of uranium and plutonium carbides in the liquid region

    International Nuclear Information System (INIS)

    Sheth, A.; Leibowitz, L.

    1975-09-01

    By the use of available low-temperature data for various thermophysical and transport properties for uranium and plutonium carbides, values above the melting point were estimated. Sets of recommended values have been prepared for the compounds UC, PuC, and (U,Pu)C. The properties that have been evaluated are density, heat capacity, enthalpy, vapor pressure, thermal conductivity, viscosity, and emissivity

  4. Analysis of refabricated fuel: determination of carbon in uranium plutonium mixed carbide

    International Nuclear Information System (INIS)

    Huwyler, S.

    1977-09-01

    In developing uranium plutonium mixed carbide which represents an advanced fuel for breeder reactors carbon analysis is an important means of determining the stoichiometry. Methods of carbon determination are briefly reviewed. The carbon determination using a LECO WR-12 Carbon Determinator is treated in detail and experience of three years operation communicated. Problems arising from operating the LECO-apparatus in a glove box are discussed. It is pointed out that carbon determination with the LECO-apparatus is a very fast method with good precision and well suited for the routine analysis of mixed carbide fuel. The accuracy of the method is checked by means of a standard. (Auth.)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    Highlights: → High resolution imagery (FIB, SEM and SIMS) of carbide inclusions in uranium metal. → Real time images following the reaction of the carbide inclusions with water vapour. → Shown preferential consumption of carbide over that of the bulk metal. → Quantity of impurities in the metal therefore seriously influence reaction rate. → Metal purity must be considered when storing uranium in air or moist conditions. - Abstract: 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.

  6. Preconcentration of uranium in water samples using dispersive ...

    African Journals Online (AJOL)

    Preconcentration of uranium in water samples using dispersive liquid-liquid micro- extraction coupled with solid-phase extraction and determination with inductively coupled plasma-optical emission spectrometry.

  7. Study of the machining of uranium carbide rods obtained by continuous casting under electronic bombardment

    International Nuclear Information System (INIS)

    Rousset, P.; Accary, A.

    1965-01-01

    The authors consider the various methods of machining uranium mono-carbide and compare them critically in the case of their application to uranium carbide obtained by fusion under an electronic bombardment and continuous casting. This study leads them to propose two mechanical machining methods: cylindrical rectification and center-less rectification, preceded by a preliminary roughing out of a cylinder, the latter appearing more suitable. A study of the machining yields as a function of the diameter of the rough bars and of the diameter of the finished rods has shown that an optimum value of the rough bar diameter exists for each value of the finished rod diameter. It is found that the yield increases as the diameter itself increases, this yield rising from 45 per cent to around 70 per cent as the diameter of the rough bars increases from 25-26 mm to 37-38 mm. (authors) [fr

  8. Synthesis and magnetic properties of highly dispersed tantalum carbide nanoparticles decorated on carbon spheres

    CSIR Research Space (South Africa)

    Bhattacharjee, K

    2016-01-01

    Full Text Available The decoration of carbon spheres (CS) by highly dispersed tantalum carbide nanoparticles (TaC NPs) was achieved, for the first time by a unique carbothermal reduction method at 1350 °C for 30 min under reduced oxygen partial pressure. TaC NPs...

  9. Electrical and thermal transport properties of uranium and plutonium carbides

    International Nuclear Information System (INIS)

    Lewis, H.D.; Kerrisk, J.F.

    1976-09-01

    Contributions of many authors are outlined with respect to the experimental measurement methods used and characteristics of the sample materials. Discussions treat the qualitative effects of sample material composition; oxygen, nitrogen, and nickel concentrations; porosity; microstructural variations; and the variability in transport property values obtained by the various investigators. Temperature-dependent values are suggested for the electrical resistivities and thermal conductivities of selected carbide compositions based on a comparative evaluation of the available data and the effects of variation in the characteristics of sample materials

  10. Irradiation performance of helium-bonded uranium--plutonium carbide fuel elements

    International Nuclear Information System (INIS)

    Latimer, T.W.; Petty, R.L.; Kerrisk, J.F.; DeMuth, N.S.; Levine, P.J.; Boltax, A.

    1979-01-01

    The current irradiation program of helium-bonded uranium--plutonium carbide elements is achieving its original goals. By August 1978, 15 of the original 171 helium-bonded elements had reached their goal burnups including one that had reached the highest burnup of any uranium--plutonium carbide element in the U.S.--12.4 at.%. A total of 66 elements had attained burnups over 8 at.%. Only one cladding breach had been identified at that time. In addition, the systematic and coordinated approach to the current steady-state irradiation tests is yielding much needed information on the behavior of helium-bonded carbide fuel elements that was not available from the screening tests (1965 to 1974). The use of hyperstoichiometric (U,Pu)C containing approx. 10 vol% (U,Pu) 2 C 3 appears to combine lower swelling with only a slightly greater tendency to carburize the cladding than single-phase (U,Pu)C. The selected designs are providing data on the relationship between the experimental parameters of fuel density, fuel-cladding gap size, and cladding type and various fuel-cladding mechanical interaction mechanisms

  11. Preparation and study of the nitrides and mixed carbide-nitrides of uranium and of plutonium

    International Nuclear Information System (INIS)

    Anselin, F.

    1966-06-01

    A detailed description is given of a simple method for preparing uranium and plutonium nitrides by the direct action of nitrogen under pressure at moderate temperatures (about 400 C) on the partially hydrogenated bulk metal. It is shown that there is complete miscibility between the UN and PuN phases. The variations in the reticular parameters of the samples as a function of temperature and in the presence of oxide have been used to detect and evaluate the solubility of oxygen in the different phases. A study has been made of the sintering of these nitrides as a function of the preparation conditions with or without sintering additives. A favorable but non-reproducible, effect has been found for traces of oxide. The best results were obtained for pure UN at 1600 C (96 per cent theoretical density) on condition that a well defined powder, was used. The criterion used is the integral width of the X-ray diffraction lines. The compounds UN and PuN are completely miscible with the corresponding carbides. This makes it possible to prepare carbide-nitrides of the general formula (U,Pu) (C,N) by solid-phase diffusion, at around 1400 C. The sintering of these carbide-nitrides is similar to that of the carbides if the nitrogen content is low; in particular, nickel is an efficient sintering agent. For high contents, the sintering is similar to that of pure nitrides. (author) [fr

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

    International Nuclear Information System (INIS)

    McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; Reif, Tyler J.; Morris, Robert N.; Hunn, John D.

    2017-01-01

    Highlights: • The minimum required uranium carbide content for HTGR UCO fuel kernels is calculated. • More nuclear and chemical factors have been included for more useful predictions. • The effect of transmutation products, like Pu and Np, on the oxygen distribution is included for the first time. - Abstract: 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 O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

  13. Radionuclide Inventories for DOE SNF Waste Stream and Uranium/Thorium Carbide Fuels

    International Nuclear Information System (INIS)

    K.L. Goluoglu

    2000-01-01

    The objective of this calculation is to generate radionuclide inventories for the Department of Energy (DOE) spent nuclear fuel (SNF) waste stream destined for disposal at the potential repository at Yucca Mountain. The scope of this calculation is limited to the calculation of two radionuclide inventories; one for all uranium/thorium carbide fuels in the waste stream and one for the entire waste stream. These inventories will provide input in future screening calculations to be performed by Performance Assessment to determine important radionuclides

  14. preconcentration of uranium in water samples using dispersive

    African Journals Online (AJOL)

    B. S. Chandravanshi

    Atomic Energy Organization of Iran, P.O. Box 14395-836, Tehran, Iran. 2Department of ... A new liquid phase microextraction method based on the dispersion of an extraction solvent into aqueous phase ... optical emission spectrometry, Uranium, Water samples ..... The validation of the presented procedure was performed ...

  15. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, J.A.B.; Durazzo, M.

    2010-01-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm 3 by using the U 3 Si 2 -Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm 3 for the U 3 Si 2 -Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  16. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Jose Antonio Batista de; Durazzo, Michelangelo, E-mail: jasouza@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 g U/c m3 by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 g U/c m3 for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian- Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  17. The compatibility of stainless steels with particles and powders of uranium carbide and low-sulphur UCS fuels

    International Nuclear Information System (INIS)

    Venter, S.

    1978-05-01

    Slightly hyperstoichiometric (U,Pu)C is a potential nuclear fuel for fast breeder reactors. The excess carbon above the stoichiometric amount results in a higher carbon activity in the fuel, and carbon is transferred to the stainless steel cladding, resulting in embrittlement of the cladding. It is with this problem of carbon transfer from the fuel to the cladding that this thesis is concerned. For practical reasons, UC and not (U,Pu)C was used as the fuel. The theory of decarburisation of carbide fuel and the carburisation of stainless steel, the facilities constructed for the project at the Atomic Energy Board, and the experimental techniques used, including preparation of the fuels, are discussed. The effect of a number of variables of uranium carbide fuel on its compatibility behaviour with stainless steels was investigated, as well as the effect om microstructure and type of stainless steel (304, 304 L and 316) on the rate of carburisation. These studies can be briefly summarised under the following headings: powder-particle size; surface oxidation of uranium carbide; preparation temperature of uranium carbide; low sulfur UCS fuels; uranium sulfide and the microstructure and type of steel. The author concludes that: the effect of surface oxidation and particle size must be taken into account when evaluating out-of-pile tests; the possible effects of surface oxidation must be taken into account when considering vibro-compacted carbide fuels; there is no advantage in replacing a fraction of the carbon atoms by sulphur atoms in slightly hyperstoichiometric carbide fuels, and the type and thermo-mechanical treatment of the stainless steel used as cladding material in a fuel pin is not important as far as the rate of carburisation by the fuel is concerned

  18. Geochemical dispersion of uranium near prospects in Pennsylvania

    International Nuclear Information System (INIS)

    Rose, A.W.; Schmiermund, R.L.; Mahar, D.L.

    1977-06-01

    The geochemical dispersion of U was investigated near sedimentary uranium prospects in eastern and north-central Pennsylvania. Near Jim Thorpe, known uranium occurrences in the Catskill Fm. are limited to the base of the Duncannon member. At Penn Haven Junction, roll-type U deposits with appreciable Pb and Se are localized adjacent to an oxidized tongue of channel-filling conglomeratic sandstone. The channel and encircling U occurrences furnish a large target for geochemical exploration. Selective extractions show that the organic, Fe-oxide, sand and silt fractions of stream sediments are the major hosts for U in stream sediments. Fe-oxides have a greater affinity for U than organic matter but are less abundant. The U content of organic matter is about 10 5 times the U content of stream water. Stream sediments furnish a representative sample of the average content of U, Zn, Cu, and major elements in soils of a drainage basin in north-central Pennsylvania, so a semiquantitative appraisal of weathering uranium occurrences can be made from stream sediments in climates and topography like Pennsylvania. The flux of uranium leaving the basin in solution is about equal to that leaving as sediment. Uranium is considerably less mobile than Ca and Na. A new method of extracting uranium from water samples, using a liquid ion exchanger (Amberlite LA-1), shows promise for simple field application

  19. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, Jose Antonio Batista de

    2011-01-01

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm 3 for U 3 Si 2 -Al dispersion-based and 2.3 gU/cm 3 for U 3 O 8 -Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm 3 in U 3 Si 2 -Al dispersion and 3.2 gU/cm 3 U 3 O 8 -Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U 3 Si 2 -Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U 3 O 8 -Al dispersion fuel plates with 3.2 gU/cm 3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U 3 Si 2 production at 4.8 gU/cm 3 , with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  20. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Hofmann, G.L.; Ryu, Woo-Seog

    1991-01-01

    Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and micro structural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide dispersion fuel. (orig.)

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

    Science.gov (United States)

    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.

  2. Study on the preparation and stability of uranium carbide samples for the determination of oxygen, hydrogen and nitrogen by fusion under high vacuum

    International Nuclear Information System (INIS)

    Perez Garcia, M.

    1966-01-01

    In view of the high reactivity of uranium carbide, the method employed for the preparation of the sample for the analysis of its gas content: oxygen, hydrogen and nitrogen, has a decisive influence on the analytical results. The variation in the O 2 , H 2 and N 2 content of the uranium carbide has been studied in this paper with the methods utilized for the sample preparation (grinding and cutting). (Author) 9 refs

  3. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  4. Report on the R&D of Uranium Carbide targets by the PLOG collaboration at PNPI-Gatchina

    CERN Document Server

    A.E. Barzakh, D.V. Fedorov, A.M. Ionan, V.S. Ivanov, M.P. Levchenko, K.A. Mezilev, F.V. Moroz, S.Yu. Orlov, V.N. Panteleev, Yu.M. Volkov,O. Alyakrinskiy, A. Andrighetto, A. Lanchais, G. Lhersonneau*, V. Rizzi, L. Stroe#, L.B. Tecchio,O. Bajeat, M. Cheikh Mhamed, S. Essabaa, C. Lau, B. Roussière,M. Dubois, C. Eléon, G. Gaubert, P. Jardin, N. Lecesne, R. Leroy, J.Y. Pacquet, M. -G. Saint Laurent, A.C.C. Villari.

    The aim of this report is to summarize the experimental results of the R&D program on Uranium Carbide targets for Radioactive Ion Beam (RIB) production performed at the Petersburg Nuclear Physics Institute (PNPI) of Gatchina (Russia). The targets have been irradiated with 1 GeV protons delivered by the Synchrocyclotron and the measurements were carried out at the IRIS isotope separator on-line. Different compositions of Uranium Carbide targets as well as different kinds of ion sources have been tested in order to evaluate efficiency and release times of the reaction products. The report includes the results of experiments performed in the period of time going from November 2001 up to March 2006. This R&D program was performed in the framework of the collaboration with the EURISOL, SPES and SPIRAL-2 projects and ISTC program.

  5. Rheological Characteristics of 2D Titanium Carbide (MXene) Dispersions: A Guide for Processing MXenes.

    Science.gov (United States)

    Akuzum, Bilen; Maleski, Kathleen; Anasori, Babak; Lelyukh, Pavel; Alvarez, Nicolas Javier; Kumbur, E Caglan; Gogotsi, Yury

    2018-03-27

    Understanding the rheological properties of two-dimensional (2D) materials in suspension is critical for the development of various solution processing and manufacturing techniques. 2D carbides and nitrides (MXenes) constitute one of the largest families of 2D materials with >20 synthesized compositions and applications already ranging from energy storage to medicine to optoelectronics. However, in spite of a report on clay-like behavior, not much is known about their rheological response. In this study, rheological behavior of single- and multilayer Ti 3 C 2 T x in aqueous dispersions was investigated. Viscous and viscoelastic properties of MXene dispersions were studied over a variety of concentrations from colloidal dispersions to high loading slurries, showing that a multilayer MXene suspension with up to 70 wt % can exhibit flowability. Processing guidelines for the fabrication of MXene films, coatings, and fibers have been established based on the rheological properties. Surprisingly, high viscosity was observed at very low concentrations for solutions of single-layer MXene flakes. Single-layer colloidal solutions were found to exhibit partial elasticity even at the lowest tested concentrations (<0.20 mg/mL) due to the presence of strong surface charge and excellent hydrophilicity of MXene, making them amenable to fabrication at dilute concentrations. Overall, the findings of this study provide fundamental insights into the rheological response of this quickly growing 2D family of materials in aqueous environments as well as offer guidelines for processing of MXenes.

  6. Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

    International Nuclear Information System (INIS)

    Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

    1988-11-01

    Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

  7. Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.

    1985-01-01

    Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235 U. Fuel plates containing 33 v/o U 3 Si and U 3 Si 2 behaved very well up to this burnup. Plates containing 33 v/o U 3 Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U 3 Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U 3 Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs

  8. Stabilization of mixed carbides of uranium-plutonium by zirconium. Part 1.: uranium carbide with small additions of zirconium; Etude de la stabilisation des carbures mixtes d'uranium et de plutonium par addition de zirconium. 1. partie: etude des carbures d'uranium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Bocker, S [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-07-01

    Cast carbide samples, being of a high density and purity, are preferable for research purposes, to samples produced by powder metallurgy methods. Samples of uranium carbide with small additions of zirconium (1 to 5 per cent) were cast, as rods, in an arc furnace. A single phase carbide with interesting qualities was produced. As cast, a dendrite structure is observed, which does not disappear, after a treatment at 1900 deg. C during 110 hours. In comparison with uranium monocarbide the compatibility with stainless steel is much improved. The specific heat (between room temperature and 2500 deg. C) is similar to the specific heat of uranium monocarbide. A study of these mixed carbides, but having a part of the uranium replaced by plutonium is under way. (author) [French] Les echantillons de monocarbures obtenus par coulee sont tres interessants pour les recherches experimentales a cause de leur grande purete, de leur densite tres elevee et de la facilite d'obtention des lingots de forme et dimensions variees. On a prepare et coule dans un four a arc des echantillons de carbures d'uranium avec de faibles additions de zirconium (1 a 5 at. pour cent). On obtient ainsi des carbures monophases presentant de meilleures proprietes que le monocarbure d'uranium. A l'etat brut de coulee on observe une structure dendritique qui n'est pas detruite par un traitement thermique de 110 heures a 1900 deg. C. La compatibilite avec l'acier inoxydable 316 (a 925 deg. C pendant 500 heures) est nettement amelioree par rapport a UC. La chaleur specifique (entre la temperature ordinaire et 2500 deg. C) et la densite sont tres peu differentes de celles du monocarbure d'uranium. Une etude concernant les composes U-Pu-Zr-C est actuellement en cours. (auteur)

  9. Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.

    2002-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 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 and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . 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–UO 2 compositions.

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

    Science.gov (United States)

    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.

  12. High temperature monitoring of silicon carbide ceramics by confocal energy dispersive X-ray fluorescence spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, Fangzuo; Liu, Zhiguo; Sun, Tianxi, E-mail: stx@bnu.edu.cn

    2016-04-15

    Highlights: • X-ray scattering was used for monitoring oxidation situation of SiC ceramics. • A calibration curve was obtained. • The confocal X-ray scattering technology was based on polycapillary X-ray optics. • The variations of contents of components of SiC ceramics were obtained. - Abstract: In the present work, we presented an alternative method for monitoring of the oxidation situation of silicon carbide (SiC) ceramics at various high temperatures in air by measuring the Compton-to-Rayleigh intensity ratios (I{sub Co}/I{sub Ra}) and effective atomic numbers (Z{sub eff}) of SiC ceramics with the confocal energy dispersive X-ray fluorescence (EDXRF) spectrometer. A calibration curve of the relationship between I{sub Co}/I{sub Ra} and Z{sub eff} was established by using a set of 8 SiC calibration samples. The sensitivity of this approach is so high that it can be easily distinguished samples of Z{sub eff} differing from each other by only 0.01. The linear relationship between the variation of Z{sub eff} and the variations of contents of C, Si and O of SiC ceramics were found, and the corresponding calculation model of the relationship between the ΔZ and the ΔC{sub C}, ΔC{sub Si}, and ΔC{sub O} were established. The variation of contents of components of the tested SiC ceramics after oxidation at high temperature was quantitatively calculated based on the model. It was shown that the results of contents of carbon, silicon and oxygen obtained by this method were in good agreement with the results obtained by XPS, giving values of relative deviation less than 1%. It was concluded that the practicality of this proposed method for monitoring of the oxidation situation of SiC ceramics at high temperatures was acceptable.

  13. Recent developments and on-line tests of uranium carbide targets for production of nuclides far from

    CERN Document Server

    V.N. Panteleev et al.

    The capacity of uranium carbide target materials of different structure and density for production of neutron-rich and heavy neutron-deficient isotopes have been investigated at the IRIS facility (PNPI) in the collaboration with Legnaro – GANIL – Orsay laboratories. The yields and release times of the species produced in the targets by the reactions induced by a 1 GeV proton beam of the PNPI synchrocyclotron have been measured. For the purpose to elaborate the most efficient and fast uranium carbide target prototype three kinds of the target materials were studied: a) a high density UC target material having ceramic-like structure with the density of 11 g/cm3 and the grain dimensions of about 200 microns; b) a high density UC target material with the density of 12 g/cm3 and the grain dimensions of about 20 microns prepared by the method of the powder metallurgy; c) a low density UCx target material with the density 3g/cm3 and the grain dimensions of about 20 microns prepared by the ISOLDE method. The comp...

  14. Present status of uranium-plutonium mixed carbide fuel development for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo; Suzuki, Yasufumi.

    One Oarai characteristic of a carbide fuel is that its doubling time is about 13 years which is only about half as long as that of an oxide fuel. The development of carbide fuels in the past ten years has been truly remarkable. Especially, through the new fuel development program initiated in 1974 in the United States, success has been achieved with respect to He- and Na-bond fuels in obtaining a 16 a/o burning rate without damage to cladding tubes. In 1984 at FFTF, a radiation of a fuel assembly consisting 91 fuel pins is contemplated. On the other hand, in Japan, in 1974, a Fuel Research Wing specializing in the study of carbide fuels was constructed in the Oarai Laboratory of the Atomic Energy Research Institute and in the fall of 1982, was successful in fabricating two carbide fuel pins having different chemical compositions

  15. Postirradiation examination of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Strain, R.V.

    1998-01-01

    Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U 2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U 2 Mo and U 3 Si 2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U 235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

  16. DEVELOPMENT OF HIGH-DENSITY U/AL DISPERSION PLATES FOR MO-99 PRODUCTION USING ATOMIZED URANIUM POWDER

    OpenAIRE

    RYU, HO JIN; KIM, CHANG KYU; SIM, MOONSOO; PARK, JONG MAN; LEE, JONG HYUN

    2013-01-01

    Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99) production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 g-U/cm3 were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compou...

  17. The solubility of solid fission products in carbides and nitrides of uranium and plutonium. Part I: literature review on experimental results

    International Nuclear Information System (INIS)

    Benedict, U.

    1977-01-01

    This review compiles the available data on the solubility of the most important non-volatile fission products in the carbides, nitrides, and carbonitrides of uranium and plutonium. It includes some elements which are not fission products, but belong to a group of the Periodic Table which contains one or more fission products elements

  18. A novel method for the preparation of uranium metal, oxide and carbide via electrolytic amalgamation

    International Nuclear Information System (INIS)

    Wang, L.C.; Lee, H.C.; Lee, T.S.; Lai, W.C.; Chang, C.T.

    1978-01-01

    A solid uranium amalgam was prepared electrolytically using a two-compartment cell separated with an ion exchange membrane for the purpose of regulating pH value within a narrowly restricted region of 2 to 3. The mercury cathode was kept at -1.8V vs SCE during electrolysis. The thereby obtained amalgam containing as high as 1.9gm U/ml Hg is easily converted into uranium metal by heating in vacuo above 1300 0 C. Uranium dioxide and uranium monocarbide could be easily obtained at relatively low temperature by reacting the amalgam with water vapor and methane. (author)

  19. Reaction of uranium and plutonium carbides with nitrogen; Reaction avec l'azote des carbures d'uranium et de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzelli, R; Martin, A; Schickel, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1966-03-01

    Uranium and plutonium carbides react with nitrogen during the grinding process preceding the final sintering. The reaction occurs even in argon atmospheres containing a few percent of residual nitrogen. The resulting contamination is responsible for the appearance of an equivalent quantity of higher carbide in the sintered products; nitrogen remains quantitatively in the monocarbide phase. UC can be transformed completely into nitride under a nitrogen pressure, at a temperature as low as 400 C. The reaction is more sluggish with PuC. The following reactions take places: UC + 0,8 N{sub 2} {yields}> UN{sub 1.60} + C and PuC + 0,5 N{sub 2} {yields} PuN + C. (authors) [French] Les carbures d'uranium et de plutonium reagissent avec l'azote au cours du broyage qui precede le frittage final. Cette reaction est sensible meme sous des atmospheres d'argon ne contenant que quelques pour cent d'azote. Cette contamination se traduit sur les produits frittes par l'apparition d'une quantite equivalente de carbure superieur, l'azote restant fixe quantitativement dans la phase monocarbure. On peut transformer entierement UC en nitrure par action de l'azote sous pression des 400 C. La reaction est plus difficile avec PuC. Les reactions sont les suivantes: UC + 0,8 N{sub 2} {yields} UN{sub 1.60} + C et PuC + 0,5 N{sub 2} {yields} PuN + C.

  20. Reaction of uranium and plutonium carbides with austenitic steels; Reaction des carbures d'uranium et de plutonium avec des aciers austenitiques

    Energy Technology Data Exchange (ETDEWEB)

    Mouchnino, M [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-07-01

    The reaction of uranium and plutonium carbides with austenitic steels has been studied between 650 and 1050 deg. C using UC, steel and (UPu)C, steel diffusion couples. The steels are of the type CN 18.10 with or without addition of molybdenum. The carbides used are hyper-stoichiometric. Tests were also carried out with UCTi, UCMo, UPuCTi and UPuCMo. Up to 800 deg. C no marked diffusion of carbon into stainless steel is observed. Between 800 and 900 deg. C the carbon produced by the decomposition of the higher carbides diffuses into the steel. Above 900 deg. C, decomposition of the monocarbide occurs according to a reaction which can be written schematically as: (U,PuC) + (Fe,Ni,Cr) {yields} (U,Pu) Fe{sub 2} + Cr{sub 23}C{sub 6}. Above 950 deg. C the behaviour of UPuCMo and that of the titanium (CN 18.12) and nickel (NC 38. 18) steels is observed to be very satisfactory. (author) [French] La reaction des carbures d'uranium et de plutonium avec des aciers austenitiques a ete etudiee entre 650 deg. C et 1050 deg. C a partir de couples de diffusion UC, acier et (UPu)C, acier. Les aciers sont du type CN 18.10 avec ou sans addition de molybdene. Les carbures utilises sont hyper-stoechiometriques. En outre on a fait des essais avec UCTi, UCMo, UPuCTi, UPuCMo. Jusqu'a 800 deg. C on ne detecte pas de diffusion sensible du carbone dans l'acier inoxydable. Entre 800 et 900 deg. C il y a diffusion dans l'acier du carbone provenant de la decomposition des carbures superieurs. A partir de 900 deg. C il y a decomposition du monocarbure selon une reaction que l'on ecrit schematiquement: (U,PuC) + (Fe, Ni, Cr) {yields} (U,Pu)Fe{sub 2} + Cr{sub 23}C{sub 6}. Nous notons a 950 deg. C le bon comportement de UPuCMo ainsi que celui des aciers au titane (CN 18. 12) et au nickel (NC 38.18). (auteur)

  1. Electron bombardment fusion and continuous casting of uranium carbide. Fundamental study of the metallurgical and thermal processes

    International Nuclear Information System (INIS)

    Trouve, J.

    1968-02-01

    During a pilot production run, about 1.200 kg of uranium carbide cylindrical rods were prepared by electron bombardment fusion and continuous casting in an apparatus making it possible to operate in a constant vacuum automatically. In order to make the most of the fusion technique used, it was necessary to resolve a certain number of problems involved in this production. It was found that the energy yield for the electron bombardment heating using accelerating voltages of about 10 kV was 100 per cent; about 40 per cent of the electrons are re-emitted by back-scattering. These electrons leave the surface with practically zero energy. The fusion technique leads to the elimination of the majority of the metallic impurities. In order to explain the variations in the non-metallic impurity contents the different reactions occurring in the molten uranium monocarbide have been determined. A micrographic study of the rods obtained has shown various types of crystallization depending on the rate of casting and, despite the uniaxial symmetry of the cooling, no texture has been observed, whatever the rate of fusion employed. The aspects of the fracture surfaces observed on certain rods can be explained by theory in the domain where the material is elastic. Furthermore it has been shown that a decrease in the brittleness occurs as a result of the formation of fine precipitates of the Wiedmanstatten structure type. (authors) [fr

  2. Fracture toughness and fatigue crack propagation in cast irons with spheroidal vanadium carbides dispersed within martensitic matrix microstructure

    International Nuclear Information System (INIS)

    Uematsu, Y.; Tokaji, K.; Horie, T.; Nishigaki, K.

    2007-01-01

    Fracture toughness and fatigue crack propagation (FCP) have been studied using compact tension (CT) specimens of as-cast and subzero-treated materials in a cast iron with spheroidal vanadium carbides (VCs) dispersed in the martensitic matrix microstructure. X-ray diffraction (XRD) analysis revealed that retained austenite was transformed to martensite by subzero treatment. Vickers hardness was increased from 738 for the as-cast material to 782 for the subzero-treated material, which could be attributed to retained austenite to martensite transformation. The subzero-treated material exhibited lower fracture toughness than the as-cast material because soft and ductile retained austenite which possesses high fracture toughness was transformed to martensite in the subzero-treated material. Intrinsic FCP resistance after taking account of crack closure was decreased by the subzero treatment, which was attributed to the predominant crack propagation through the interface between VCs and the matrix and the straight crack path in the matrix microstructure

  3. Optimization of uranium carbide fabrication by carbothermic reduction with limited oxygen content

    International Nuclear Information System (INIS)

    Raveu, Gaelle

    2014-01-01

    Mixed carbides (U, Pu)C, are good fuel candidate for generation IV reactors because of their high fissile atoms density and excellent thermal properties for economical (more compact and efficient cores) and safety reasons (high melting margin). UC can be imagine as a surrogate material ror R and D studies on (U,Pu)C fuel behavior, because of their similar structures. The carbothermic reaction was used because it is the most studied and now consider for industrial process. However, it involves powders manipulation: in air, carbide can strongly react at room temperature and under controlled atmosphere it can absorb impurities. An inerted installation under Ar, BaGCARA, was therefore used. Process improvements were carried out, including the sintering atmosphere in order to evaluate the impact on the sample purity (about oxygen content). The original method by ion beam analysis was used to determine the surface composition (oxygen in-depth profiles in the first microns and stoichiometry). This oxygen analysis was set for the first time in carbonaceous materials. XRD analysis showed the formation of an intermediate compound during the carbothermic reaction and a better crystallization of the samples fabricated in BaGCARA. They also have a better microstructure, density, and visual appearance if compared to former samples. Vacuum sintering leads to a denser UC with fewer second phases if compared to Ar, Ar/H 2 or controlled PC atmospheres. However, it was not possible to analyze carbides without air contact which may impact their lattice parameter and lead to their deterioration. When the carbide is initially free of oxygen, it oxidizes faster, more intensely and heterogeneously. The mechanical stress induced between the grains lead to fracturing the material, to corrosion cracking and then a de-bonding of the material. A study of oxidation mechanisms would be interesting to validate and understand the evolution of the material in contact with oxygen. A study of the

  4. DEVELOPMENT OF HIGH-DENSITY U/AL DISPERSION PLATES FOR MO-99 PRODUCTION USING ATOMIZED URANIUM POWDER

    Directory of Open Access Journals (Sweden)

    HO JIN RYU

    2013-12-01

    Full Text Available Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99 production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 g-U/cm3 were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compounds by a chemical reaction of the uranium particles and aluminum matrix. Thus, these target plates can be treated with the same alkaline dissolution process that is used for conventional UAlx dispersion targets, while increasing the uranium density in the target plates

  5. Irradiation behaviour of mixed uranium-plutonium carbides, nitrides and carbonitrides; Comportement a l'irradiation de carbures, nitrures et carbonitrures mixtes d'uranium et de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Mikailoff, H; Mustelier, J P; Bloch, J; Leclere, J; Hayet, L [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-07-01

    In the framework of the research program of fast reactor fuels two irradiation experiments have been carried out on mixed uranium-plutonium carbides, nitrides and carbo-nitrides. In the first experiment carried out with thermal neutrons, the fuel consisted of sintered pellets sheathed in a stainless steel can with a small gap filled with helium. There were three mixed mono-carbide samples and the maximum linear power was 715 W/cm. After a burn-up slightly lower than 20000 MW day/tonne, a swelling of the fuel which had ruptured the cans was observed. In the second experiment carried out in the BR2 reactor with epithermal neutrons, the samples consisted of sintered pellets sodium bonded in a stainless steel tube. There were three samples containing different fuels and the linear power varies between 1130 and 1820 W/cm. Post-irradiation examination after a maximal burn-up of 1550 MW day/tonne showed that the behaviour of the three fuel elements was satisfactory. (authors) [French] Dans le cadre du programme d'etude des conibustiles pour reacteurs rapides, on a realise deux experiences d'irradiation de carbures, nitrures et carbonitrures mixtes d'uranium et de plutonium. Dans la premiere experience, faite en neutrons thermiques, le combustible etait constitue de,pastilles frittees gainees dans un tube d'acier inoxydable avec un faible jeu rempli d'helium. Il y avait trois echantillons de monocarbures mixtes, et la puissance lineaire maximale etait de 715 W/cm. Apres un taux de combustion legerement inferieur a 20 000 MWj/t, on a observe un gonflement des combustible qui a provoque, la rupture des gaines. Pans la seconde experience, realisee dans le reacteur BR2 en neutrons epithermiques, les echantillons etaient constitues de pastilles frittees gainees dans un tube d'acier avec un joint sodium. Il y avait trois echantillons contenant des combustibles differents, et la puissance lineaire variait de 1130 a 1820 W/cm. Les examens apres irradiation a un taux maximal de

  6. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Hofman, G.L.; Ryu, Woo-Seog.

    1989-01-01

    Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and microstructural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide disperson fuel. 5 refs., 10 figs

  7. Milling uranium silicide powder for dispersion nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, E.; Silva, D.G.; Souza, J.A.B.; Durazzo, M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Riella, H.G. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)

    2009-07-01

    Full text: Uranium silicide (U3Si2) is presently considered the best fuel qualified so far in terms of uranium loading and performance. Stability of the U3Si2 fuel with uranium density of 4.8 g/cm3 was confirmed by burnup stability tests performed during the Reduced Enrichment for Research and Test Reactors (RERTR) program. This fuel was chosen to compose the first core of the new Brazilian Multipurpose Research Reactor (RMB), planned to be constructed in the next years. This new reactor will consume bigger quantities of U3Si2 powder, when compared with the small consumption of the IEA-R1 research reactor of IPEN-CNEN/SP, the unique MTR type research reactor operating in the country. At the present time, the milling operation of U3Si2 ingots is made manually. In order to increase the powder production capacity, the manual milling must be replaced by an automated procedure. This paper describes a new milling machine and procedure developed to produce U3Si2 powder with higher efficiency. (author)

  8. Draft environmental statement related to the Union Carbide Corporation, Gas Hills Uranium Project (Natrona County, Wyoming)

    International Nuclear Information System (INIS)

    1979-01-01

    The proposed action is the renewal of Source Material License SUA-648 issued for the operation of the Gas Hills Uranium Project in Wyoming, near Moneta. The project is an acid leach, ion-exchange, and solvent-extraction uranium ore processing mill at an increased capacity of 500,000 tons per year and the construction of two heap leach facilities in Natrona and Fremont Counties for initial processing of low-grade ore. After analysis of environmental impacts and adverse effects, it is the proposed position of NRC that the license be renewed subject to conditions relating to stabilization of the tailings, reclamation, environmental monitoring, evaluation of any future activity not evaluated by NRC, archeological survey, analysis of unexpected harmful effects, and decommissioning

  9. The improvement of technology for high-uranium-density Al-base dispersion fuel plates

    International Nuclear Information System (INIS)

    Shouhui, Dai; Rongxian, Sun; Hejian, Mao; Baosheng, Zhao; Changgen, Yin

    1987-01-01

    An improved rolling process was developed for manufacturing Al-base dispersion fuel plates. When the fuel content in the meat increased up to 50 vol%, the non-uniformity of uranium is not more than ± 7.2%, and the minimum cladding thickness is not less than 0.32 mm. (Author)

  10. Studies relating to construction materials to be used in different options for head end treatment in reprocessing of mixed carbide fuel of plutonium and uranium

    International Nuclear Information System (INIS)

    Rajan, S.K.; Palamalai, A.; Ravi, T.N.; Sampath, M.; Raman, V.R.; Balasubramanian, G.R.

    1993-01-01

    Mixed carbide of uranium and plutonium has been chosen as the fuel for the first core of Fast Breeder Test Reactor, installed in the Indira Gandhi Centre for Atomic Research. Reprocessing of this fuel is one of the vital steps to prove the viability of the fuel cycle. The head end treatment process introduces constraints in the reprocessing of carbide fuel when compared to the commonly used mixed oxide fuel. Three head end processes, namely direct oxidation, pyrohydrolysis and direct dissolution in nitric acid with oxidation of organic acids were considered for study for exercising the choice. The paper briefly describes the three processes. In each process probable material of construction and related problems are discussed. (author). 3 refs, 5 figs, 7 tabs

  11. On the mechanism of dispersion hardening in molybdenum-carbide alloy systems

    International Nuclear Information System (INIS)

    Shulepov, V.I.; Yudkovskij, S.I.; Batenina, O.I. et al.

    1975-01-01

    The effect of heat treatment of the forming alloys of the Mo-Ti-C and Mo-Ti-Zr-C systems (at the temperatures below the recrystallization temperature) on the structure, distribution of carbon and mechanical properties of the alloys is studied. It is shown that the dispersion-strengthened state of the molybdenum alloys may be obtained on the account of the deformation ageing effect, rather than through the use of the standard heat-treatment procedure (hardening plus ageing). On the basis of the experimental results a theoretical explanation of strengthening of the high-alloy molybdenum-titanum-carbon system is given

  12. Dispersion of silicon carbide nanoparticles in a AA2024 aluminum alloy by a high-energy ball mill

    International Nuclear Information System (INIS)

    Carreño-Gallardo, C.; Estrada-Guel, I.; López-Meléndez, C.; Martínez-Sánchez, R.

    2014-01-01

    Highlights: • Synthesis of 2024-SiC NP nanocomposite by mechanical milling process. • SiC nanoparticles improved mechanical properties of aluminum alloy 2024 matrix. • A homogeneous distribution of SiC nanoparticles were observed in the matrix • Compressive and hardness properties of the composite are improved significantly. -- Abstract: Al 2024 alloy was reinforced with silicon carbide nanoparticles (SiC NP ), whose concentration was varied in the range from 0 to 5 wt.%; some composites were synthesized with the mechanical milling (MM) process. Structure and microstructure of the consolidated samples were studied by X-ray diffraction and transmission electron microscopy, while mechanical properties were investigated by compressive tests and hardness measurements. The microstructural evidence shows that SiC NP were homogeneously dispersed into the Al 2024 alloy using high-energy MM after 2 h of processing. On the other hand, an increase of the mechanical properties (yield stress, maximum strength and hardness) was observed in the synthesized composites as a direct function of the SiC NP content. In this research several strengthening mechanisms were observed, but the main was the obstruction of dislocations movement by the addition of SiC NP

  13. Pre-concentration of uranium from water samples by dispersive liquid-liquid micro-extraction

    Energy Technology Data Exchange (ETDEWEB)

    Khajeh, Mostafa; Nemch, Tabandeh Karimi [Zabol Univ. (Iran, Islamic Republic of). Dept. of Chemistry

    2014-07-01

    In this study, a simple and rapid dispersive liquid-liquid microextraction (DLLME) was developed for the determination of uranium in water samples prior to high performance liquid chromatography with diode array detection. 1-(2-pyridylazo)-2-naphthol (PAN) was used as complexing agent. The effect of various parameters on the extraction step including type and volume of extraction and dispersive solvents, pH of solution, concentration of PAN, extraction time, sample volume and ionic strength were studied and optimized. Under the optimum conditions, the limit of detection (LOD) and preconcentration factor were 0.3 μg L{sup -1} and 194, respectively. Furthermore, the relative standard deviation of the ten replicate was <2.6%. The developed procedure was then applied to the extraction and determination of uranium in the water samples.

  14. Pre-concentration of uranium from water samples by dispersive liquid-liquid micro-extraction

    International Nuclear Information System (INIS)

    Khajeh, Mostafa; Nemch, Tabandeh Karimi

    2014-01-01

    In this study, a simple and rapid dispersive liquid-liquid microextraction (DLLME) was developed for the determination of uranium in water samples prior to high performance liquid chromatography with diode array detection. 1-(2-pyridylazo)-2-naphthol (PAN) was used as complexing agent. The effect of various parameters on the extraction step including type and volume of extraction and dispersive solvents, pH of solution, concentration of PAN, extraction time, sample volume and ionic strength were studied and optimized. Under the optimum conditions, the limit of detection (LOD) and preconcentration factor were 0.3 μg L -1 and 194, respectively. Furthermore, the relative standard deviation of the ten replicate was <2.6%. The developed procedure was then applied to the extraction and determination of uranium in the water samples.

  15. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

  16. Irradiation behavior of uranium oxide-aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

  17. Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Ray, Allison E.

    1998-01-01

    Uranium alloys are candidates for the fuel phase in aluminium matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminium interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic γ-phase during fabrication and irradiation, at temperatures at which αU is the equilibrium phase. transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analysed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data. (author)

  18. Irradiation testing of high density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U 2 Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions

  19. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  20. Irradiation behavior of uranium-silicide dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.

    1984-01-01

    This paper describes and analyzes the irradiation behavior of experimental fuel plates containing U 3 Si, U 3 Si-1.5 w/o Al, and U 3 Si 2 particulate fuel dispersed and clad in aluminum. The fuel is nominally 19.9%-enriched 235 U and the fuel volume fraction in the central ''meat'' section of the plates is approximately 33%. Sets of fuel plates were removed from the Oak Ridge Research reactor at burnup levels of 35, 83, and 94% 235 U depletion and examined at the Alpha-Gamma Hot-Cell Facility at Argonne National Laboratory. The results of the examination may be summarized as follows. The dimensional stability of the U 3 Si 2 and pure U 3 Si fuel was excellent throughout the entire burnup range, with uniform plate thickness increases up to a maximum of 4 mils at the highest burnup level (94% 235 U depletion). This corresponds to a meat volume increase of 11%. The swelling was partially due to solid fission products but to a larger extent to fission gas bubbles. The fission gas bubbles in U 3 Si 2 were small (submicrometer size) and very uniformly distributed, indicating great stability. To a large extent this was also the case for U 3 Si; however, larger bubbles ( 3 Si-1.5 w/o Al fuel became unstable at the higher burnup levels. Fission gas bubbles were larger than in the other two fuels and were present throughout the fuel particles. At 94% 235 U depletion, the formation of fission gas bubbles with diameters up to 20 mils caused the plates to pillow. It is proposed that aluminum in U 3 Si destabilizes fission gas bubble formation to the point of severe breakaway swelling in the prealloyed silicide fuel. (author)

  1. Composites comprising silicon carbide fibers dispersed in magnesia-aluminate matrix and fabrication thereof and of other composites by sinter forging

    Science.gov (United States)

    Panda, Prakash C.; Seydel, Edgar R.; Raj, Rishi

    1989-10-03

    A novel ceramic-ceramic composite of a uniform dispersion of silicon carbide fibers in a matrix of MgO.multidot.nAl.sub.2 O.sub.3 wherein n ranges from about 1 to about 4.5, said composite comprising by volume from 1 to 50% silicon carbide fibers and from 99 to 50% MgO.multidot.nAl.sub.2 O.sub.3. The composite is readily fabricated by forming a powder comprising a uniform dispersion of silicon carbide fibers in poorly crystalline phase comprising MgO and Al.sub.2 O.sub.3 in a mole ratio of n and either (a) hot pressing or preferably (b) cold pressing to form a preform and then forging utilizing a temperature in the range of 1100.degree. C. to 1900.degree. C. and a strain rate ranging from about 10.sup.-5 seconds .sup.-1 to about 1 seconds .sup.-1 so that surfaces cracks do not appear to obtain a shear deformation greater than 30%.

  2. The obtainment of highly concentrated uranium pellets for plate type (MTR) fuel by dispersion of uranium aluminides in aluminium

    International Nuclear Information System (INIS)

    Morando, R.A.; Raffaeli, H.A.; Balzaretti, D.E.

    1980-01-01

    The use of the intermetallic UAl 3 for manufacturing plate type MTR fuel with 20% U 235 enriched uranium and a density of about 20 kg/m 3 is analyzed. The technique used is the dispersion of UAl 3 particles in aluminium powder. The obtainment of the UAl 3 intermetallic was performed by fusion in an induction furnace in an atmosphere of argon at a pressure of 0.7 BAR (400 mm) using an alumina melting pot. To make the aluminide powder and attain the wished granulometry a cutting and a rotating crusher were used. Aluminide powders of different granulometries and different pressures of compactation were analyzed. In each case the densities were measured. The compacts were colaminated with the 'Picture Frame' technique at temperatures of 490 and 0 deg C with excellent results from the manufacturing view point. (M.E.L.) [es

  3. Uranium and base metal dispersion studies in the Maquire Lake area, Saskatchewan

    International Nuclear Information System (INIS)

    Sopuck, V.J.; Lehto, D.A.W.; Alley, D.W.

    1980-03-01

    The objective of this study was to study uranium and base metal dispersion in various sample media occurring in the Maguire Lake area of Saskatchewan: bedrock, overburden, lake water, and lake sediments. Factors controlling partitioning of metals among various sample media were investigated, and lake sediment data were interpreted in terms of the factors to determine the significance of lake sediment data in indicating local mineralization. The association between organic matter contents and metal contents was found to vary between lake-center and nearshore sediments. Nickel, cobalt and zinc in lake sediments are strongly controlled by hydroxide precipitation and are less dependent on bedrock type. The concentration of Fe in center-lake sediments appears to reflect only the physicochemical parameters in the lake. Uranium and copper are strongly controlled by and preferentially concentrated in the organic matter; however, in center-lake sediments with >12 percent organic matter, U and Cu strongly reflect rock type

  4. A comparison of the metallurgical behaviour of dispersion fuels with uranium silicides and U6Fe as dispersants

    International Nuclear Information System (INIS)

    Nazare, S.

    1984-01-01

    In the past few years metallurgical studies have been carried out to develop fuel dispersions with U-densities up to 7.0 Mg U m -3 . Uranium silicides have been considered to be the prime candidates as dispersants; U 6 Fe being a potential alternative on account of its higher U-density. The objective of this paper is to compare the metallurgical behaviour of these two material combinations with regard to the following aspects: (1) preparation of the compounds U 3 Si, U 3 Si 2 and U 6 Fe; (2) powder metallurgical processing to miniature fuel element plates; (3) reaction behaviour under equilibrium conditions in the relevant portions of the ternary U-Si-Al and U-Fe-Al systems; (4) dimensional stability of the fuel plates after prolonged thermal treatment; (5) thermochemical behaviour of fuel plates at temperatures near the melting point of the cladding. Based on this data, the possible advantages of each fuel combination are discussed. (author)

  5. Study of the machining of uranium carbide rods obtained by continuous casting under electronic bombardment; Etude de l'usinage de barreaux de carbure d'uranium obtenus par coulee continue sous bombardement electronique

    Energy Technology Data Exchange (ETDEWEB)

    Rousset, P; Accary, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The authors consider the various methods of machining uranium mono-carbide and compare them critically in the case of their application to uranium carbide obtained by fusion under an electronic bombardment and continuous casting. This study leads them to propose two mechanical machining methods: cylindrical rectification and center-less rectification, preceded by a preliminary roughing out of a cylinder, the latter appearing more suitable. A study of the machining yields as a function of the diameter of the rough bars and of the diameter of the finished rods has shown that an optimum value of the rough bar diameter exists for each value of the finished rod diameter. It is found that the yield increases as the diameter itself increases, this yield rising from 45 per cent to around 70 per cent as the diameter of the rough bars increases from 25-26 mm to 37-38 mm. (authors) [French] Les auteurs envisagent les differentes methodes d'usinage du monocarbure d'uranium et se livrent a une etude critique de celles-ci, dans le cas de leur application a l'usinage de barreaux de carbure d'uranium obtenus par fusion sous bombardement electronique et coulee continue. Cette etude les conduit a proposer deux methodes d'usinage mecanique: la rectification cylindrique et la rectification 'centerless', precedee d'un ebauchage par carottage, la seconde paraissant la plus appropriee. L'etude des rendements d'usinage en fonction du diametre des barreaux bruts et du diametre des barreaux finis, a mis en evidence une valeur optimale du diametre des barreaux bruts pour chaque valeur du diametre des barreaux usines. Elle a montre que le rendement croit lorsque le diametre croit lui-meme, ce rendement passant d'environ 45 pour cent a environ 70 pour cent, lorsque le diametre des barreaux bruts passe de 25-26 mm a 37-38 mm.

  6. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  7. Contribution to the study of second phases particles dispersion in polycrystalline uranium dioxide

    International Nuclear Information System (INIS)

    Peres, V.

    1994-06-01

    To reduce fission gas release of irradiated polycrystalline uranium dioxide, the dispersion of intragranular nanometric particles of second phase necessary to pin gas bubbles may complete the advantage of a large-grained fuel microstructure. Moreover, intergranular glass films may improve high temperatures mechanical properties of UO 2 . In this study, mixtures of additives composed of ''corindon'' structure oxides that enhance the fuel grain growth and composed of different oxides with variable solid solubilities in the UO 2 matrix were achieved. Additives with a negligible solubility inhibit grain boundaries motion except those, such as silica, that involve a liquid phase at the sintering temperature. Rare earth oxides that form stable solid solutions with UO 2 cannot lead to precipitation, but have no effect on the fuel grain growth doped with ''corindon'' type oxides. A chromium oxide excess allows the creation of a fuel microstructure described by large grains and intragranular spherical Cr 2 O 3 inclusions observed by scanning electron microscopy. Values for the bulk lattice diffusion coefficient of Cr 3+ cations in UO 2 can be deduced from the experimental growth of those dispersed particles by an Ostwald ripening mechanism. The formation of small precipitated metal particles inside the uranium dioxide matrix induced by the internal reduction of a solid solution has not been performed. However, direct reduction of insoluble chromium oxide particles is easy and produces metallic intragranular precipitates. (author). 119 refs., 112 figs., 33 tabs., 5 annexes

  8. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    International Nuclear Information System (INIS)

    Vasudevamurthy, G.; Radecka, A.; Massey, C.

    2015-01-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  9. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

    2015-07-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  10. The solubility of solid fission products in carbides and nitrides of uranium and plutonium: Pt.2. Solubility rules based on lattice parameter differences

    International Nuclear Information System (INIS)

    Benedict, U.

    1977-01-01

    The Relative Lattice Parameter Difference (RLPD) is defined for a solute element with respect to cubic carbides and nitrides of uranium and plutonium as solvents. Rules are given for the relationship between the solubility and the RLPD. NaCl type monocarbides with RLPD's from -10.2% to +7.8% are completely miscible with UC and PuC. NaCl type mononitrides with RLPD's from -7.5% to +8.5% are completely miscible with UN and PuN. The solubility in the sesquicarbides increases with decreasing RPLD and becomes complete in Pu 2 C 3 at RLPD = +4%, and in U 2 C 3 at RLPD approximately +1.5%. Solubilities are predicted on the basis of these rules for the cases where no experimental results are available

  11. Dose evaluation by radon dispersion in the uranium mining tailings of Malargue

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.; Ferrer, F.; Munoz, G.

    1996-01-01

    The objective of this work is the environmental impact evaluation of the uranium mining tailings actually sited in Malargue, due to atmospheric dispersion of radon gas and its inhalation by inhabitants of the vicinity Malargue factory complex is located at the south of Mendoza-Argentina. the complex was an industrial installation to treat uranium mineral, which started operation in 1954 and was operable until December 1986. As a by-product of this industrial process approximately 700 thousand tons of tailings were generated. They actually cover about 9 Ha in four piles with a ∼6 m height above the ground. their final disposal is at present under discussion, and in this frame the present work is performed. In the work a Gaussian-plume dispersion model was developed, using local meteorological data, with specific correlations to consider the different stability classes, according to the Pasquill-Gifford concept. The model is fed with the meteorological information provided by two towers mounted by the Atomic Energy Commission, that bring information each half hour. This information for 211 full days is available. The model considers the wind induced[dispersion, the corrections for emission period (discretized in half hour intervals), the stability class (automatically chosen as a function of day time and wind speed) and radioactive decay. The model calculates the inhalation dose with a non-equilibrium factor which depends on the time since release. Finally the model integrates the dose in time, and provides yearly doses in any coordinates point. Special attention was devoted to the fact that the source is extended, and several source discretization were performed until reliable results were obtained. the model allows for comparative calculations. As the work result the yearly doses were obtained, for four 'critical groups' of interest. They correspond to three nearby houses, and downtown Malargue. (authors). 6 refs., 3 tabs

  12. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  13. Energy dispersive X-ray fluorescence determination of cadmium in uranium matrix using Cd Kα line excited by continuum

    International Nuclear Information System (INIS)

    Dhara, Sangita; Misra, N.L.; Aggarwal, S.K.; Venugopal, V.

    2010-01-01

    An energy dispersive X-ray fluorescence method for determination of cadmium (Cd) in uranium (U) matrix using continuum source of excitation was developed. Calibration and sample solutions of cadmium, with and without uranium were prepared by mixing different volumes of standard solutions of cadmium and uranyl nitrate, both prepared in suprapure nitric acid. The concentration of Cd in calibration solutions and samples was in the range of 6 to 90 μg/mL whereas the concentration of Cd with respect to U ranged from 90 to 700 μg/g of U. From the calibration solutions and samples containing uranium, the major matrix uranium was selectively extracted using 30% tri-n-butyl phosphate in dodecane. Fixed volumes (1.5 mL) of aqueous phases thus obtained were taken directly in specially designed in-house fabricated leak proof Perspex sample cells for the energy dispersive X-ray fluorescence measurements and calibration plots were made by plotting Cd Kα intensity against respective Cd concentration. For the calibration solutions not having uranium, the energy dispersive X-ray fluorescence spectra were measured without any extraction and Cd calibration plots were made accordingly. The results obtained showed a precision of 2% (1σ) and the results deviated from the expected values by < 4% on average.

  14. Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

    International Nuclear Information System (INIS)

    Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

    1976-02-01

    An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

  15. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

  16. Improvement of the homogeneity of atomized particles dispersed in high uranium density research reactor fuels

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Yoon-Sang; Lee, Don-Bae; Sohn, Woong-Hee; Hong, Soon-Hyung

    1998-01-01

    A study on improving the homogeneous dispersion of atomized spherical particles in fuel meats has been performed in connection with the development of high uranium density fuel. In comparing various mixing methods, the better homogeneity of the mixture could be obtained as in order of Spex mill, V-shape tumbler mixer, and off-axis rotating drum mixer. The Spex mill mixer required some laborious work because of its small capacity per batch. Trough optimizing the rotating speed parameter for the V-shape tumbler mixer, almost the same homogeneity as with the Spex mill could be obtained. The homogeneity of the extruded fuel meats appeared to improve through extrusion. All extruded fuel meats with U 3 Si powder of 50-volume % had fairly smooth surfaces. The homogeneity of fuel meats by V-shaped tumbler mixer revealed to be fairly good on micrographs. (author)

  17. Study on the preparation and stability of uranium carbide samples for the determination of oxygen, hydrogen and nitrogen by fusion under high vacuum; Estudio sobre la preparacion y estabilidad de las muestras de carburo de uranio para la determinacion de oxigeno, hidrogeno y nitrogeno por fusion en alto vacio

    Energy Technology Data Exchange (ETDEWEB)

    Perez Garcia, M

    1966-07-01

    In view of the high reactivity of uranium carbide, the method employed for the preparation of the sample for the analysis of its gas content: oxygen, hydrogen and nitrogen, has a decisive influence on the analytical results. The variation in the O{sub 2}, H{sub 2} and N{sub 2} content of the uranium carbide has been studied in this paper with the methods utilized for the sample preparation (grinding and cutting). (Author) 9 refs.

  18. HGSYSTEMUF6, Simulating Dispersion Due to Atmospheric Release of Uranium Hexafluoride (UF6)

    International Nuclear Information System (INIS)

    Hanna, G; Chang, J.C.; Zhang, J.X.; Bloom, S.G.; Goode, W.D. Jr; Lombardi, D.A.; Yambert, M.W.

    2001-01-01

    1 - Description of program or function: HGSYSTEMUF6 is a suite of models designed for use in estimating consequences associated with accidental, atmospheric release of Uranium Hexafluoride (UF 6 ) and its reaction products, namely Hydrogen Fluoride (HF), and other non-reactive contaminants which are either negatively, neutrally, or positively buoyant. It is based on HGSYSTEM Version 3.0 of Shell Research LTD., and contains specific algorithms for the treatment of UF 6 chemistry and thermodynamics. HGSYSTEMUF6 contains algorithms for the treatment of dense gases, dry and wet deposition, effects due to the presence of buildings (canyon and wake), plume lift-off, and the effects of complex terrain. The models components of the suite include (1) AEROPLUME/RK, used to model near-field dispersion from pressurized two-phase jet releases of UF6 and its reaction products, (2) HEGADAS/UF6 for simulating dense, ground based release of UF 6 , (3) PGPLUME for simulation of passive, neutrally buoyant plumes (4) UF6Mixer for modeling warm, potentially reactive, ground-level releases of UF 6 from buildings, and (5) WAKE, used to model elevated and ground-level releases into building wake cavities of non-reactive plumes that are either neutrally or positively buoyant. 2 - Methods: The atmospheric release and transport of UF 6 is a complicated process involving the interaction between dispersion, chemical and thermodynamic processes. This process is characterized by four separate stages (flash, sublimation, chemical reaction entrainment and passive dispersion) in which one or more of these processes dominate. The various models contained in the suite are applicable to one or more of these stages. For example, for modeling reactive, multiphase releases of UF 6 , the AEROPLUME/RK component employs a process-splitting scheme which numerically integrates the differential equations governing dispersion, UF 6 chemistry, and thermodynamics. This algorithm is based on the assumption that

  19. The effect of dispersed materials on baro-membrane treatment of uranium-containing waters

    International Nuclear Information System (INIS)

    Kryvoruchko, Antonina P.; Atamanenkoa, Irina D.

    2007-01-01

    The paper investigated a treatment process of uranium-containing waters in a membrane reactor while using natural mineral kizelgur and synthetic sorbent SKN-1K with subsequent ultra- and nano-filtration separation of the mixture. The retention coefficient of U(VI) by membrane UPM-20 under conditions of quasi-stationary equilibrium reached the levels of 0.87-0.89 and 0.89-0.91, respectively, while using natural mineral kizelgur and synthetic sorbent SKN-1K. In the case of membrane OPMN-P and natural mineral kizelgur the retention coefficient of U(VI) was 0.990-0.991 and 0.993-0.996, respectively, while using natural mineral kizelgur and synthetic sorbent SKN-1K. Data regarding the state of water in membranes formed from natural mineral or synthetic sorbent on the surface of substrate membranes UPM-20 and OPMN-P made it possible to conclude that dispersed materials of different chemical nature affect the process of baro-membrane treatment of uranium-containing waters. (authors)

  20. FIREPLUME model for plume dispersion from fires: Application to uranium hexafluoride cylinder fires

    International Nuclear Information System (INIS)

    Brown, D.F.; Dunn, W.E.

    1997-06-01

    This report provides basic documentation of the FIREPLUME model and discusses its application to the prediction of health impacts resulting from releases of uranium hexafluoride (UF 6 ) in fires. The model application outlined in this report was conducted for the Draft Programmatic Environmental Impact Statement for Alternative Strategies for the Long-Term Management and Use of Depleted UF 6 . The FIREPLUME model is an advanced stochastic model for atmospheric plume dispersion that predicts the downwind consequences of a release of toxic materials from an explosion or a fire. The model is based on the nonbuoyant atmospheric dispersion model MCLDM (Monte Carlo Lagrangian Dispersion Model), which has been shown to be consistent with available laboratory and field data. The inclusion of buoyancy and the addition of a postprocessor to evaluate time-varying concentrations lead to the current model. The FIREPLUME model, as applied to fire-related UF 6 cylinder releases, accounts for three phases of release and dispersion. The first phase of release involves the hydraulic rupture of the cylinder due to heating of the UF 6 in the fire. The second phase involves the emission of material into the burning fire, and the third phase involves the emission of material after the fire has died during the cool-down period. The model predicts the downwind concentration of the material as a function of time at any point downwind at or above the ground. All together, five fire-related release scenarios are examined in this report. For each scenario, downwind concentrations of the UF 6 reaction products, uranyl fluoride and hydrogen fluoride, are provided for two meteorological conditions: (1) D stability with a 4-m/s wind speed, and (2) F stability with a 1-m/s wind speed

  1. Study of the machining of uranium carbide rods obtained by continuous casting under electronic bombardment; Etude de l'usinage de barreaux de carbure d'uranium obtenus par coulee continue sous bombardement electronique

    Energy Technology Data Exchange (ETDEWEB)

    Rousset, P.; Accary, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The authors consider the various methods of machining uranium mono-carbide and compare them critically in the case of their application to uranium carbide obtained by fusion under an electronic bombardment and continuous casting. This study leads them to propose two mechanical machining methods: cylindrical rectification and center-less rectification, preceded by a preliminary roughing out of a cylinder, the latter appearing more suitable. A study of the machining yields as a function of the diameter of the rough bars and of the diameter of the finished rods has shown that an optimum value of the rough bar diameter exists for each value of the finished rod diameter. It is found that the yield increases as the diameter itself increases, this yield rising from 45 per cent to around 70 per cent as the diameter of the rough bars increases from 25-26 mm to 37-38 mm. (authors) [French] Les auteurs envisagent les differentes methodes d'usinage du monocarbure d'uranium et se livrent a une etude critique de celles-ci, dans le cas de leur application a l'usinage de barreaux de carbure d'uranium obtenus par fusion sous bombardement electronique et coulee continue. Cette etude les conduit a proposer deux methodes d'usinage mecanique: la rectification cylindrique et la rectification 'centerless', precedee d'un ebauchage par carottage, la seconde paraissant la plus appropriee. L'etude des rendements d'usinage en fonction du diametre des barreaux bruts et du diametre des barreaux finis, a mis en evidence une valeur optimale du diametre des barreaux bruts pour chaque valeur du diametre des barreaux usines. Elle a montre que le rendement croit lorsque le diametre croit lui-meme, ce rendement passant d'environ 45 pour cent a environ 70 pour cent, lorsque le diametre des barreaux bruts passe de 25-26 mm a 37-38 mm.

  2. Contribution to the study of the hydrolysis of uranium carbides (1963); Contribution a l'etude de l'hydrolyse des carbures d'uranium (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Spitz, J [Commisariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1963-06-15

    The hydrolysis of uranium monocarbide in neutral or acid medium leads to the formation of a complex mixture of hydrogen and hydrocarbons mostly saturated. When UC-U alloys are dissolved in hydrochloric-phosphoric medium, the free uranium contents can be determined with good accuracy from the composition of the gaseous phase. The hydrolysis of mixtures of uranium mono - and dicarbide in neutral or acid medium, leads to the formation of a complex mixture of hydrogen and gaseous and condensed hydrocarbons, the composition of which is principally dependent upon the UC{sub 2} content. The reaction mechanisms which are presented in this paper for the hydrolysis of UC and UC{sub 2} provide account for all experimental observations. (author) [French] L'hydrolyse en milieu neutre ou acide du monocarbure d'uranium conduit a la formation d'un melange complexe d'hydrogene et d'hydrocarbures, satures en grande majorite. L'attaque en milieu chlorhydrique-phosphorique des alliages UC-U permet la determination avec une bonne precision, des teneurs en uranium libre a partir de la composition des gaz degages. L'hydrolyse en milieu neutre ou acide des melanges de mono - et dicarbure d'uranium conduit a la formation d'un melange complexe d'hydrogene et d'hydrocarbures gazeux et condenses, dont la composition est essentiellement fonction de la teneur en UC{sub 2}. Les mecanismes reactionnels proposes pour l'hydrolyse de UC et UC{sub 2} rendent compte de tous les faits experimentaux observes. (auteur)

  3. Uranium

    International Nuclear Information System (INIS)

    Cuney, M.; Pagel, M.; Leroy, J.

    1992-01-01

    First, this book presents the physico-chemical properties of Uranium and the consequences which can be deduced from the study of numerous geological process. The authors describe natural distribution of Uranium at different scales and on different supports, and main Uranium minerals. A great place in the book is assigned to description and classification of uranium deposits. The book gives also notions on prospection and exploitation of uranium deposits. Historical aspects of Uranium economical development (Uranium resources, production, supply and demand, operating costs) are given in the last chapter. 7 refs., 17 figs

  4. Uranium facilitated transport by water-dispersible colloids in field and soil columns

    International Nuclear Information System (INIS)

    Crancon, P.; Pili, E.; Charlet, L.

    2010-01-01

    The transport of uranium through a sandy podzolic soil has been investigated in the field and in column experiments. Field monitoring, numerous years after surface contamination by depleted uranium deposits, revealed a 20 cm deep uranium migration in soil. Uranium retention in soil is controlled by the 238 U initially present in the soil column and 233 U brought by input solution are desorbed. The mobilization process observed experimentally after a drop of ionic strength may account for a rapid uranium migration in the field after a rainfall event, and for the significant uranium concentrations found in deep soil horizons and in groundwater, 1 km downstream from the pollution source.

  5. Transition metal carbide and boride abrasive particles

    International Nuclear Information System (INIS)

    Valdsaar, H.

    1978-01-01

    Abrasive particles and their preparation are discussed. The particles consist essentially of a matrix of titanium carbide and zirconium carbide, at least partially in solid solution form, and grains of crystalline titanium diboride dispersed throughout the carbide matrix. These abrasive particles are particularly useful as components of grinding wheels for abrading steel. 1 figure, 6 tables

  6. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The article briefly discusses the Australian government policy and the attitude of political party factions towards the mining and exporting of the uranium resources in Australia. Australia has a third of the Western World's low-cost uranium resources

  7. Uranium

    International Nuclear Information System (INIS)

    Poty, B.; Cuney, M.; Bruneton, P.; Virlogeux, D.; Capus, G.

    2010-01-01

    With the worldwide revival of nuclear energy comes the question of uranium reserves. For more than 20 years, nuclear energy has been neglected and uranium prospecting has been practically abandoned. Therefore, present day production covers only 70% of needs and stocks are decreasing. Production is to double by 2030 which represents a huge industrial challenge. The FBR-type reactors technology, which allows to consume the whole uranium content of the fuel, is developing in several countries and will ensure the long-term development of nuclear fission. However, the implementation of these reactors (the generation 4) will be progressive during the second half of the 21. century. For this reason an active search for uranium ores will be necessary during the whole 21. century to ensure the fueling of light water reactors which are huge uranium consumers. This dossier covers all the aspects of natural uranium production: mineralogy, geochemistry, types of deposits, world distribution of deposits with a particular attention given to French deposits, the exploitation of which is abandoned today. Finally, exploitation, ore processing and the economical aspects are presented. Contents: 1 - the uranium element and its minerals: from uranium discovery to its industrial utilization, the main uranium minerals (minerals with tetravalent uranium, minerals with hexavalent uranium); 2 - uranium in the Earth's crust and its geochemical properties: distribution (in sedimentary rocks, in magmatic rocks, in metamorphic rocks, in soils and vegetation), geochemistry (uranium solubility and valence in magmas, uranium speciation in aqueous solution, solubility of the main uranium minerals in aqueous solution, uranium mobilization and precipitation); 3 - geology of the main types of uranium deposits: economical criteria for a deposit, structural diversity of deposits, classification, world distribution of deposits, distribution of deposits with time, superficial deposits, uranium

  8. DISPERSION AND SORPTION CHARACTERISTICS OF URANIUM IN THE ZEOLITE-QUARTZ MIXTURE AS BACKFILL MATERIAL IN THE RADIOACTIVE WASTE REPOSITORY

    Directory of Open Access Journals (Sweden)

    Herry Poernomo

    2010-06-01

    Full Text Available The experiment of sorption and dispersion characteristics of uranium in the zeolite-quartz mixture as candidate of raw material of backfill material in the radioactive waste repository has been performed. The objective is to know the effect of zeolite and quartz grain size on the zeolite-to-quartz weight ratio that gives porosity (ε, permeability (K, and dispersivity (α of uranium in the zeolite-quartz mixture as backfill material. The experiment was carried out by fixed bed method in the column filled by the zeolite-quartz mixture with zeolite-to-quartz weight percent ratio of 100/0, 80/20, 60/40, 40/60, 20/80, 0/100 wt. % in the water saturated condition flowed by uranyl nitrate solution of 500 ppm concentration (Co as uranium simulation which was leached from immobilized radioactive waste in the repository. The concentration of uranium in the effluents represented as Ct were analyzed by spectrophotometer Corning Colorimeter 253 every 15 minutes, then using Co and Ct uranium dispersivity (α in the backfill material was determined. The experiment data shown that 0.196 mm particle size of zeolite and 0.116 mm particle size of quartz on the zeolite-to-quartz weight ratio of 60/40 wt. % with ε = 0.678, K = 3.345x10-4 cm/second, and α = 0.759 cm can be proposed as candidate of raw material of backfill material in the radioactive waste repository.   Keywords: backfill material, quartz, radioactive waste, zeolite

  9. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  10. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  11. Preparation and study of the nitrides and mixed carbide-nitrides of uranium and of plutonium; Preparation et etude des nitrures et carbonitrures d'uranium et de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Anselin, F [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1966-06-01

    A detailed description is given of a simple method for preparing uranium and plutonium nitrides by the direct action of nitrogen under pressure at moderate temperatures (about 400 C) on the partially hydrogenated bulk metal. It is shown that there is complete miscibility between the UN and PuN phases. The variations in the reticular parameters of the samples as a function of temperature and in the presence of oxide have been used to detect and evaluate the solubility of oxygen in the different phases. A study has been made of the sintering of these nitrides as a function of the preparation conditions with or without sintering additives. A favorable but non-reproducible, effect has been found for traces of oxide. The best results were obtained for pure UN at 1600 C (96 per cent theoretical density) on condition that a well defined powder, was used. The criterion used is the integral width of the X-ray diffraction lines. The compounds UN and PuN are completely miscible with the corresponding carbides. This makes it possible to prepare carbide-nitrides of the general formula (U,Pu) (C,N) by solid-phase diffusion, at around 1400 C. The sintering of these carbide-nitrides is similar to that of the carbides if the nitrogen content is low; in particular, nickel is an efficient sintering agent. For high contents, the sintering is similar to that of pure nitrides. (author) [French] On decrit en detail une methode simple de preparation des nitrures d'uranium et de plutonium par action directe de l'azote sous pression, a temperature moyenne (vers 400 C), sur les metaux massifs partiellement hydrures. On montre que la miscibilite est complete entre les phases UN et PuN. L'evolution des parametres reticulaires des echantillons en fonction de la temperature et en presence d'oxyde a ete utilisee pour detecter et estimer la solubilite de l'oxygene dans les diverses phases. On a etudie le frittage de ces nitrures en fonction des conditions de preparation, avec ou sans additif de

  12. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

    A discussion is given of uranium as an energy source in The Australian economy. Figures and predictions are presented on the world supply-demand position and also figures are given on the added value that can be achieved by the processing of uranium. Conclusions are drawn about Australia's future policy with regard to uranium (R.L.)

  13. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  14. Tungsten--carbide critical assembly

    International Nuclear Information System (INIS)

    Hansen, G.E.; Paxton, H.C.

    1975-06-01

    The tungsten--carbide critical assembly mainly consists of three close-fitting spherical shells: a highly enriched uranium shell on the inside, a tungsten--carbide shell surrounding it, and a steel shell on the outside. Ideal critical specifications indicate a rather low computed value of k/sub eff/. Observed and calculated fission-rate distributions for 235 U, 238 U, and 237 Np are compared, and calculated leakage neutrons per fission in various energy groups are given. (U.S.)

  15. Assessment of the meteorological data and atmospheric dispersion estimates in the Ranger 1 Uranium Mining Environmental Impact Statement

    International Nuclear Information System (INIS)

    Clark, G.H.

    1977-03-01

    Wind records from Jabiru, Northern Territory, Australia have been re-analysed to give atmospheric dispersion estimates of sulphur dioxide and radioactive contaminants associated with a proposed uranium mining and milling operation. Revisions in the plume rise equations have led to lower annual average sulphur dioxide air concentrations than those presented in the Ranger 1 Uranium Mining Environmental Impact Statement. Likewise, the short term peak air concentrations of sulphur dioxide were all within the United States Environment Protection Agency air quality standards. Even though the radon gas inventory was revised upwards, predicted concentrations were only slightly higher than those in the RUMEIS. An attempt was made at a first estimate of the uranium dust source term caused by wind suspension from stockpiled ore and waste rock. In a preliminary analysis using a 'surface depletion' model, it was estimated that uranium dust air concentrations would be decreased by about an order of magnitude when dry deposition was included in the atmospheric dispersion model. Integrating over all sources, radionuclides and meteorological conditions, the annual radiation dose to members of the public in the Regional Centre is estimated to be a maximum of 5 per cent of the recommended annual limits. (author)

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

    Science.gov (United States)

    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.

  17. Phonon dispersion relation of uranium nitrate above and below the Neel temperature

    International Nuclear Information System (INIS)

    Dolling, G.; Holden, T.M.; Evensson, E.C.; Buyers, W.J.L.; Lander, G.H.

    1977-01-01

    Neutron coherent inelastic scattering measurements have been made of the phonon dispersion relation of uranium nitride both above and below the Neel temperature T/sub N/ = 50 K. Within the precision of the measurements, about 1% in frequency and 10% in line width and in scattered neutron intensity, no significant changes in these phonon properties were observed as a function of temperature other than those arising from population factor changes and a small stiffening of the lattice as the temperature decreases. At 4.2 K, two acoustic and two optic branches have been determined for each of the [001], [110] and [111] directions. The optic mode measurements revealed (a) a 20% variation in frequency across the Brillouin zone and (b) an interesting disposition of the LO and TO modes, such that nu/sub LO/ > nu/sub TO/ along [001] and [110], while the reverse is true along the [111] directions. Within the experimental resolution, the LO and TO modes are degenerate near q = 0. We have been unable to obtain any satisfactory description of these results on the basis of conventional theoretical treatments (e.g. rigid-ion or shell models). Other possible interpretations of the results are discussed

  18. Phonon dispersion relation of uranium nitride above and below the Neel temperature

    International Nuclear Information System (INIS)

    Dolling, G.; Holden, T.M.; Svensson, E.C.; Buyers, W.J.L.; Lander, G.H.

    1977-01-01

    Neutron coherent inelastic scattering measurements have been made of the phonon dispersion relation of uranium nitride both above and below the Neel temperature T N = 50 K. Within the precision of the measurements, about 1% in frequency and 10% in line width and in scattered neutron intensity, no significant changes in these phonon properties were observed as a function of temperature other than those arising from population factor changes and a small stiffening of the lattice as the temperature decreases. At 4.2 K, two acoustic and two optic branches have been determined for each of the [001], [110] and [111] directions. The optic mode measurements revealed (a) a 20% variation in frequency across the Brillouin zone and (b) and interesting disposition of the LO and TO modes, such that ν LO > ν TO along [001] and [11-], while the reverse is true along the [111] directions. Within the experimental resolution, the LO and TO modes are degenerate near q = 0. We have been unable to obtain any satisfactory description of these results on the basis of conventional theoretical treatments (e.g. rigid-ion or shell models). Other possible interpretations of the results are discussed. (author)

  19. BASIC program to compute uranium density and void volume fraction in laboratory-scale uranium silicide aluminum dispersion plate-type fuel

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro

    1991-05-01

    BASIC program simple and easy to operate has been developed to compute uranium density and void volume fraction for laboratory-scale uranium silicide aluminum dispersion plate-type fuel, so called miniplate. An example of the result of calculation is given in order to demonstrate how the calculated void fraction correlates with the microstructural distribution of the void in a miniplate prepared in our laboratory. The program is also able to constitute data base on important parameters for miniplates from experimentally-determined values of density, weight of each constituent and dimensions of miniplates. Utility programs pertinent to the development of the BASIC program are also given which run in the popular MS-DOS environment. All the source lists are attached and brief description for each program is made. (author)

  20. Processing map and hot working mechanisms in a P/M TiAl alloy composite with in situ carbide and silicide dispersions

    International Nuclear Information System (INIS)

    Rao, K.P.; Prasad, Y.V.R.K.

    2010-01-01

    Research highlights: Mechanical alloying of Ti and Al with small additions of Si and C was used to synthesize metastable phases, which were incorporated in Ti-Al matrices using powder metallurgy techniques. These metastable phases (or also called as precursors), at higher temperatures, transformed in situ into very fine hard reinforcements that develop coherent interface with the surrounding matrix. Typically, Ti5Si3 and TiC are the end products after the synthesis of composite. In this study, hot working behavior of such composites has been studied using the concepts of processing maps to identify the safe and best processing conditions that should be adopted while forming this composite. Also, kinetic analysis of hot deformation has been performed to identify the dominant deformation mechanism. The results are compared with that of base TiAl matrix. The powder metallurgy route offers the advantage of working the material at much lower temperatures compared to the traditional cast and forge route. - Abstract: A titanium aluminide alloy composite with in situ carbide and silicide dispersions has been synthesized by mixing 90% of matrix with elemental composition of 46Ti-46Al-4Nb-2Cr-2Mn and 10% precursor with composition 55Ti-27Al-12Si-6C prepared by mechanical alloying. The powder mixture was blended for 2 h followed by hot isostatic pressing (HIP) at 1150 deg. C for 4 h under a pressure of 150 MPa. In addition to TiAl alloy matrix, the microstructure of the HIP'ed billet showed a small volume fraction of Nb-rich intermetallic phase along with carbide and silicide dispersions formed in situ during HIP'ing. Cylindrical specimens from the HIP'ed billets were compressed at temperatures and strain rates in the ranges of 800-1050 deg. C and 0.0001-1 s -1 . The flow curves exhibited flow softening leading to a steady-state flow at strain rates lower than 0.01 s -1 while fracture occurred at higher strain rates. The processing map developed on the basis of flow stress at

  1. Simple and Rapid Dual-Dispersive Liquid-Liquid Microextraction as an Innovative Extraction Method for Uranium in Real Water Samples Prior to the Determination of Uranium by a Spectrophotometric Technique.

    Science.gov (United States)

    Khan, Naeemullah; Tuzen, Mustafa; Kazi, Tasneem Gul

    2017-11-01

    An innovative, rapid, and simple dual-dispersive liquid-liquid microextraction (DDLL-ME) approach was used to extract uranium from real samples for the first time. The main objective of this study was to disperse extraction solvent by using an air-agitated syringe system to overcome matrix effects and avoid dispersion of hazardous dispersive organic solvents by using heat. The DDLL-ME method consisted of two dispersive liquid-liquid extraction steps with chloroform as the extracting solvent. Uranium formed complexes with 4-(2-thiazolylazo) resorcinol in the aqueous phase and was extracted in extracting solvent (chloroform) after the first dispersive liquid-liquid process. Uranium was then back-extracted in the acidic aqueous phase in a second dispersive liquid-liquid process. Finally, uranium was determined by a spectrophotometric detection technique. The variables that played a key role in the proposed method were studied and optimized. The LOD and sensitivity enhancement factor for uranium were found to be 0.60 µg/L and 45, respectively, under optimized conditions. Calibration graphs were found to be linear in the range of 5.0-600 µg/L. The RSD was 2.5%. Reliability of the proposed method was verified by analyzing certified reference material TM-28.3.

  2. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium

    International Nuclear Information System (INIS)

    Bocker, S.

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [fr

  3. Loading Actinides in Multilayered Structures for Nuclear Waste Treatment: The First Case Study of Uranium Capture with Vanadium Carbide MXene.

    Science.gov (United States)

    Wang, Lin; Yuan, Liyong; Chen, Ke; Zhang, Yujuan; Deng, Qihuang; Du, Shiyu; Huang, Qing; Zheng, Lirong; Zhang, Jing; Chai, Zhifang; Barsoum, Michel W; Wang, Xiangke; Shi, Weiqun

    2016-06-29

    Efficient nuclear waste treatment and environmental management are important hurdles that need to be overcome if nuclear energy is to become more widely used. Herein, we demonstrate the first case of using two-dimensional (2D) multilayered V2CTx nanosheets prepared by HF etching of V2AlC to remove actinides from aqueous solutions. The V2CTx material is found to be a highly efficient uranium (U(VI)) sorbent, evidenced by a high uptake capacity of 174 mg g(-1), fast sorption kinetics, and desirable selectivity. Fitting of the sorption isotherm indicated that the sorption followed a heterogeneous adsorption model, most probably due to the presence of heterogeneous adsorption sites. Density functional theory calculations, in combination with X-ray absorption fine structure characterizations, suggest that the uranyl ions prefer to coordinate with hydroxyl groups bonded to the V-sites of the nanosheets via forming bidentate inner-sphere complexes.

  4. Uranium facilitated transport by water-dispersible colloids in field and soil columns

    Energy Technology Data Exchange (ETDEWEB)

    Crancon, P.; Pili, E. [CEA Bruyeres-le-Chatel, DIF, 91 (France); Charlet, L. [Univ Grenoble 1, Lab Geophys Interne and Tectonophys LGIT OSUG, CNRS, UJF, UMR5559, F-38041 Grenoble 9 (France)

    2010-07-01

    The transport of uranium through a sandy podsolic soil has been investigated in the field and in column experiments. Field monitoring, numerous years after surface contamination by depleted uranium deposits, revealed a 20 cm deep uranium migration in soil. Uranium retention in soil is controlled by the {<=} 50 {mu}m mixed humic and clayey coatings in the first 40 cm i.e. in the E horizon. Column experiments of uranium transport under various conditions were run using isotopic spiking. After 100 pore volumes elution, 60% of the total input uranium is retained in the first 2 cm of the column. Retardation factor of uranium on E horizon material ranges from 1300 (column) to 3000 (batch). In parallel to this slow uranium migration, we experimentally observed a fast elution related to humic colloids of about 1-5% of the total-uranium input, transferred at the mean pore-water velocity through the soil column. In order to understand the effect of rain events, ionic strength of the input solution was sharply changed. Humic colloids are retarded when ionic strength increases, while a major mobilization of humic colloids and colloid-borne uranium occurs as ionic strength decreases. Isotopic spiking shows that both {sup 238}U initially present in the soil column and {sup 233}U brought by input solution are desorbed. The mobilization process observed experimentally after a drop of ionic strength may account for a rapid uranium migration in the field after a rainfall event, and for the significant uranium concentrations found in deep soil horizons and in groundwater, 1 km downstream from the pollution source. (authors)

  5. Uranium facilitated transport by water-dispersible colloids in field and soil columns

    Energy Technology Data Exchange (ETDEWEB)

    Crancon, P., E-mail: pierre.crancon@cea.fr [CEA, DAM, DIF, F-91297 Arpajon (France); Pili, E. [CEA, DAM, DIF, F-91297 Arpajon (France); Charlet, L. [Laboratoire de Geophysique Interne et Tectonophysique (LGIT-OSUG), University of Grenoble-I, UMR5559-CNRS-UJF, BP53, 38041 Grenoble cedex 9 (France)

    2010-04-01

    The transport of uranium through a sandy podzolic soil has been investigated in the field and in column experiments. Field monitoring, numerous years after surface contamination by depleted uranium deposits, revealed a 20 cm deep uranium migration in soil. Uranium retention in soil is controlled by the < 50 {mu}m mixed humic and clayey coatings in the first 40 cm i.e. in the E horizon. Column experiments of uranium transport under various conditions were run using isotopic spiking. After 100 pore volumes elution, 60% of the total input uranium is retained in the first 2 cm of the column. Retardation factor of uranium on E horizon material ranges from 1300 (column) to 3000 (batch). In parallel to this slow uranium migration, we experimentally observed a fast elution related to humic colloids of about 1-5% of the total-uranium input, transferred at the mean porewater velocity through the soil column. In order to understand the effect of rain events, ionic strength of the input solution was sharply changed. Humic colloids are retarded when ionic strength increases, while a major mobilization of humic colloids and colloid-borne uranium occurs as ionic strength decreases. Isotopic spiking shows that both {sup 238}U initially present in the soil column and {sup 233}U brought by input solution are desorbed. The mobilization process observed experimentally after a drop of ionic strength may account for a rapid uranium migration in the field after a rainfall event, and for the significant uranium concentrations found in deep soil horizons and in groundwater, 1 km downstream from the pollution source.

  6. Uranium determination in different compositions

    International Nuclear Information System (INIS)

    Bulyanitsa, L.S.; Ivanova, K.S.; Ryzhinskij, M.V.; Alekseeva, N.A.; Solntseva, L.F.; Shereshevskaya, I.I.

    1978-01-01

    For clarifying the suitability of two different methods of analysis for determining uranium without its previous purification, the analysis of uranium carbides (UC, UC 2 , UC - ZrC) and alloys (U - Al, U - Zr - Nb, U- Ti) has been carried out. Dissolution of the compositions examined was carried out either after previous calcining (UC, UC 2 ) or fusion with KHSO 4 (UC - ZrC), or in phosphoric acid (alloys). The first method, a variant of potentiometric titration, has been modified for small amounts of uranium. Titration was carried out on a semiautomatic titrating unit. The uranium amount per titration is about 4 to 5 mg. The second method of analysis is the coulombmetric titration at a constant current intensity. The quantity of uranium per titration was equal to 1 - 3 mg. The statistical processing of the results obtained was carried out by a dispersion analysis that allowed to reveal the influence of separate factors, such as method of analysis, type of composition, the non-uniformity of a sample, the enumerated factors influencing the dispersion of the analysis results. It has been shown that the both methods are equally suitable for analysis of the uranium compounds examined

  7. Effects of HIP and forging on fracture behaviour in cast iron with spheroidal vanadium carbides dispersed within martensitic-matrix microstructure

    International Nuclear Information System (INIS)

    Uematsu, Y.; Tokaji, K.; Nishigaki, K.; Okajima, D.; Ogasawara, M.

    2010-01-01

    The cast iron with spheroidal vanadium carbides dispersed within martensitic-matrix microstructure was developed as a die material due to its high hardness. In order to achieve high performances of dies, not only the hardness but also the mechanical properties such as fracture toughness and fatigue crack propagation (FCP) resistance should be improved. In this paper, hot isostatic pressing (HIP) or forging was applied to the cast iron to improve mechanical properties, and the fracture behaviour, such as flexural strength, fracture toughness and FCP, was studied. The average flexural strength was reduced by forging because of the enhanced notch sensitivity due to the increase in the hardness. The fracture toughness was not affected by HIP nor forging while its scatter was significantly reduced by both post-treatments. The intrinsic FCP resistance taking account of crack closure was the same regardless of the application of HIP or forging, indicating that a slight change in the microstructure resulting from both treatments and the presence of casting defects exerted little influence on FCP behaviour. It could be concluded that both HIP and forging could improve the hardness of the material, while fracture toughness and FCP resistance were maintained.

  8. An investigation on fuel meats extruded with atomized U-10wt% Mo powder for uranium high-density dispersion fuel

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Don-Bae; Sohn, Dong-Seong

    1997-01-01

    The RERTR program has been making an effort to develop dispersion fuels with uranium densities of 8 to 9 g U/cm3 for research and test reactors. Using atomized U-10wt%Mo powder, fuel meats have been fabricated successfully up to 55 volume % of fuel powder. The uranium density of an extruded meat with a 55 volume % of fuel powder was obtained to be 7.7 g/cm3. A relatively high porosity of 7.3% was formed due to cracking of particles, presumably induced by the impingement among agglomerated particles. Tensile test results indicated that the strength of fuel meats with 55% volume fraction decreased some and a little of ductility was maintained. Examination on the fracture surface revealed that some U-10%Mo particles appeared to be broken by the tensile force in brittle rupture mode. The increase of broken particles in high fuel fraction is considered to be induced mainly by the impingement among agglomerated particles. Uranium loading density is assumed to be improved through the development of the better homogeneous dispersion technology. (author)

  9. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

    Events in the Canadian uranium industry during 1980 are reviewed. Mine and mill expansions and exploration activity are described, as well as changes in governmental policy. Although demand for uranium is weak at the moment, the industry feels optimistic about the future. (LL)

  10. Geochemical dispersion associated with uranium deposits in sandstone roll front type and its relationship to the Orinoco Oil Belt, Venezuela

    International Nuclear Information System (INIS)

    Manrique, J.

    2014-01-01

    In Venezuela, there is a potential for the formation of uranium deposits in areas such as the Guiana Shield, the south of the Eastern Basin, the Andes and the massif of Baúl, among other areas. Especially great interest is the exploration of uranium redox interface type (roll front), in areas such as the southern part of the Orinoco Oil Belt, north and northwest of the Guiana Shield, where groundwater uranium collecting the weathering shield flowing northward in the sandstones and mudstones of the Cretaceous to Quaternary formations, which constitute the southern boundary of the Eastern basin Venezuela. The presence of gas, extra-heavy crude oil, bitumen and lignite of the Orinoco Oil Belt can be an effective barrier for uranium in solution, which may have precipitated at the redox interface of this groundwater. This process certainly was more effective before the Orinoco river take its course to the east and the waters of small rivers and large draining shield contributed to uranium aquifers became more deep north. This work was based on a qualitative model describing geochemical dispersion associated with uranium deposits in sandstone, roll front type, which indicates that the daughter isotopes "2"3"8U, which can migrate extensively are: "2"2"2Rn, "4He, and in a smaller proportion: "2"2"6Ra and "2"2"2Rn daughters ("2"1"4Bi, "2"1"0Pb). The main exploration methods were established, which can be applied in areas of the Orinoco Oil Belt, north of the Guiana Shield, and areas west of this, among the most important are: soil measurements of radon and helium near faults, sampling soils with gamma spectrometry analysis, log interpretation of oil wells in the area of interest to establish gamma – lithological anomalies, ground water analysis of uranium, radon, radium, helium, vanadium, selenium, molybdenum, analysis of samples oil drilling cores to locate anomalous stratigraphic levels. This research will provide the basis to establish methodologies for uraniferous

  11. Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  12. Major constituent quantitative determination in uranium alloys by coupled plasma atomic emission spectrometry and X ray fluorescence wavelength dispersive spectrometry

    International Nuclear Information System (INIS)

    Oliveira, Luis Claudio de; Silva, Adriana Mascarenhas Martins da; Gomide, Ricardo Goncalves; Silva, Ieda de Souza

    2013-01-01

    A wavelength-dispersive X-ray fluorescence (WD-XRF) spectrometric method for determination of major constituents elements (Zr, Nb, Mo) in Uranium/Zirconium/Niobium and Uranium/Molybdenum alloy samples were developed. The methods use samples taken in the form of chips that were dissolved in hot nitric acid and precipitate particles melted with lithium tetraborate and dissolved in hot nitric acid and finally analyzed as a solution. Studies on the determination by inductively coupled plasma optic emission spectrometry (ICP OES) using matched matrix in calibration curve were developed. The same samples solution were analyzed in both methods. The limits of detection (LOD), linearity of the calibrations curves, recovery study, accuracy and precision of the both techniques were carried out. The results were compared. (author)

  13. Highly dispersive ion exchangers in the analytical chemistry of uranium, particularly regarding separation methods

    International Nuclear Information System (INIS)

    Schoening, R.

    1975-01-01

    The reaction of water-insoluble polyvinyl pyrrolidon with uranium VI was investigated and a determination method for uranium was worked out in which the polyvinyl pyrrolidon was used as specific exchanger. Good separations of uranium from numerous transition metal ions were achieved here. The application of this exchanger for a fast and simple elution and determination method was of particular importance. A possible sorption mechanism was suggested based on the capacity curve of uranium with polyvinyl pyrrolidon and nitrogen and chloride content at maximum load. The sorption occurs by coordination of the carbonyl oxygen of single pyrrolidon rings with the protons of the complex acides and uranium. This assumption is supported by IR investigations. The sorbability of other inorganic acids was also investigated and possible structures were formulated for the sorption mechanism. In addition to this, ion exchangers were prepared based on cellulose by converting cellulose powder with aziridine and tris-1-aziridinyl-phosphine oxide. A polyethylene imine cellulose of high capacity was obtained in the conversion of cellulose powder with aziridine. This exchanger absorbs cobalt III very strongly. The exchanger loaded with cobalt III was used to separate the uranium as cyanato complex. The exchanger obtained in converting chlorated cellulose with tris-1-aziridinyl phosphine oxide also absorbs uranium VI very strongly. Thus a separation method of high specifity and selectivity was developed. (orig.) [de

  14. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    International Nuclear Information System (INIS)

    Sinha, V.P.; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P.

    2009-01-01

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes

  15. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, V.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)], E-mail: vedsinha@barc.gov.in; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2009-04-03

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and {gamma}-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes.

  16. Microstructure and Mechanical Properties of Nano-Size Zirconium Carbide Dispersion Strengthened Tungsten Alloys Fabricated by Spark Plasma Sintering Method

    International Nuclear Information System (INIS)

    Xie Zhuoming; Liu Rui; Fang Qianfeng; Zhang Tao; Jiang Yan; Wang Xianping; Liu Changsong

    2015-01-01

    W-(0.2, 0.5, 1.0)wt% ZrC alloys with a relative density above 97.5% were fabricated through the spark plasma sintering (SPS) method. The grain size of W-1.0wt% ZrC is about 2.7 μm, smaller than that of pure W and W-(0.2, 0.5)wt% ZrC. The results indicated that the W-ZrC alloys exhibit higher hardness at room temperature, higher tensile strength at high temperature, and a lower ductile to brittle transition temperature (DBTT) than pure W. The tensile strength and total elongation of W-0.5wt% ZrC alloy at 700 °C is 535 MPa and 24.8%, which are respectively 59% and 114% higher than those of pure W (337 MPa, 11.6%). The DBTT of W-(0.2, 0.5, 1.0)wt% ZrC materials is in the range of 500°C–600°C, which is about 100 °C lower than that of pure W. Based on microstructure analysis, the improved mechanical properties of the W-ZrC alloys were suggested to originate from the enhanced grain boundary cohesion by ZrC capturing the impurity oxygen in tungsten and nano-size ZrC dispersion strengthening. (paper)

  17. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

    Evidence of expanding markets, improved prices and the short supply of uranium became abundantly clear in 1975, providing the much needed impetus for widespread activity in all phases of uranium operations. Exploration activity that had been at low levels in recent years in Canada was evident in most provinces as well as the Northwest Territories. All producers were in the process of expanding their uranium-producing facilities. Canada's Atomic Energy Control Board (AECB) by year-end had authorized the export of over 73,000 tons of U/sub 3/0/sub 8/ all since September 1974, when the federal government announced its new uranium export guidelines. World production, which had been in the order of 25,000 tons of U/sub 3/0/sub 8/ annually, was expected to reach about 28,000 tons in 1975, principally from increased output in the United States.

  18. Electron bombardment fusion and continuous casting of uranium carbide. Fundamental study of the metallurgical and thermal processes; Fusion sous bombardement d'electrons et coulee continue de carbure d'uranium. Etude fondamentale des processus metallurgiques et thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Trouve, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-02-01

    During a pilot production run, about 1.200 kg of uranium carbide cylindrical rods were prepared by electron bombardment fusion and continuous casting in an apparatus making it possible to operate in a constant vacuum automatically. In order to make the most of the fusion technique used, it was necessary to resolve a certain number of problems involved in this production. It was found that the energy yield for the electron bombardment heating using accelerating voltages of about 10 kV was 100 per cent; about 40 per cent of the electrons are re-emitted by back-scattering. These electrons leave the surface with practically zero energy. The fusion technique leads to the elimination of the majority of the metallic impurities. In order to explain the variations in the non-metallic impurity contents the different reactions occurring in the molten uranium monocarbide have been determined. A micrographic study of the rods obtained has shown various types of crystallization depending on the rate of casting and, despite the uniaxial symmetry of the cooling, no texture has been observed, whatever the rate of fusion employed. The aspects of the fracture surfaces observed on certain rods can be explained by theory in the domain where the material is elastic. Furthermore it has been shown that a decrease in the brittleness occurs as a result of the formation of fine precipitates of the Wiedmanstatten structure type. (authors) [French] Au cours d'une fabrication pilote, environ 1 200 kg de barreaux cylindriques de carbure d'uranium ont ete prepares par fusion sous bombardement d'electrons et coulee continue dans un appareillage permettant d'operer d'une maniere automatique sous vide constant. Afin de tirer le meilleur parti possible de la technique de fusion utilisee, il importait de repondre a un certain nombre de questions soulevees par cette fabrication. Le rendement energetique du chauffage par bombardement d'electrons pour des tensions acceleratrices de l'ordre de 10 kV a ete

  19. PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

    Science.gov (United States)

    Holden, R.B.; Powers, R.M.; Blaber, O.J.

    1959-09-22

    The preparation of uranium metal by the carbon reduction of an oxide of uranium is described. In a preferred embodiment of the invention a charge composed of carbon and uranium oxide is heated to a solid mass after which it is further heated under vacuum to a temperature of about 2000 deg C to produce a fused uranium metal. Slowly ccoling the fused mass produces a dendritic structure of uranium carbide in uranium metal. Reacting the solidified charge with deionized water hydrolyzes the uranium carbide to finely divide uranium dioxide which can be separated from the coarser uranium metal by ordinary filtration methods.

  20. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

    Developments in the Australian uranium industry during 1980 are reviewed. Mine production increased markedly to 1841 t U 3 O 8 because of output from the new concentrator at Nabarlek and 1131 t of U 3 O 8 were exported at a nominal value of $37.19/lb. Several new contracts were signed for the sale of yellowcake from Ranger and Nabarlek Mines. Other developments include the decision by the joint venturers in the Olympic Dam Project to sink an exploration shaft and the release of an environmental impact statement for the Honeymoon deposit. Uranium exploration expenditure increased in 1980 and additions were made to Australia's demonstrated economic uranium resources. A world review is included

  1. Uranium

    International Nuclear Information System (INIS)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-01-01

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U 3 O 8 ; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables

  2. Method to manufacture a nuclear fuel from uranium-plutonium monocarbide or uranium-plutonium mononitride

    International Nuclear Information System (INIS)

    Krauth, A.; Mueller, N.

    1977-01-01

    Pure uranium carbide or nitride is converted with plutonium oxide and carbon (all in powder form) to uranium-plutonium monocarbide or mononitride by cold pressing and sintering at about 1600 0 C. Pure uranium carbide or uranium nitride powder is firstly prepared without extensive safety measures. The pure uranium carbide or nitride powder can also be inactivated by using chemical substances (e.g. stearic acid) and be handled in air. The sinterable uranium carbide or nitride powder (or also granulate) is then introduced into the plutonium line and mixed with a nonstoichiometrically adjusted, prereacted mixture of plutonium oxide and carbon, pressed to pellets and reaction sintered. The surface of the uranium-plutonium carbide (higher metal content) can be nitrated towards the end of the sinter process in a stream of nitrogen. The protective layer stabilizes the carbide against the water and oxygen content in air. (IHOE) [de

  3. An analysis of uranium dispersal and health effects using a Gulf War case study

    International Nuclear Information System (INIS)

    Marshall, Albert Christian

    2005-01-01

    The study described in this report used mathematical modeling to estimate health risks from exposure to depleted uranium (DU) during the 1991 Gulf War for both U.S. troops and nearby Iraqi civilians. The analysis found that the risks of DU-induced leukemia or birth defects are far too small to result in an observable increase in these health effects among exposed veterans or Iraqi civilians. Only a few veterans in vehicles accidentally struck by U.S. DU munitions are predicted to have inhaled sufficient quantities of DU particulate to incur any significant health risk (i.e., the possibility of temporary kidney damage from the chemical toxicity of uranium and about a 1% chance of fatal lung cancer). The health risk to all downwind civilians is predicted to be extremely small. Recommendations for monitoring are made for certain exposed groups. Although the study found fairly large calculational uncertainties, the models developed and used are generally valid. The analysis was also used to assess potential uranium health hazards for workers in the weapons complex. No illnesses are projected for uranium workers following standard guidelines; nonetheless, some research suggests that more conservative guidelines should be considered

  4. An analysis of uranium dispersal and health effects using a Gulf War case study.

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Albert Christian

    2005-07-01

    The study described in this report used mathematical modeling to estimate health risks from exposure to depleted uranium (DU) during the 1991 Gulf War for both U.S. troops and nearby Iraqi civilians. The analysis found that the risks of DU-induced leukemia or birth defects are far too small to result in an observable increase in these health effects among exposed veterans or Iraqi civilians. Only a few veterans in vehicles accidentally struck by U.S. DU munitions are predicted to have inhaled sufficient quantities of DU particulate to incur any significant health risk (i.e., the possibility of temporary kidney damage from the chemical toxicity of uranium and about a 1% chance of fatal lung cancer). The health risk to all downwind civilians is predicted to be extremely small. Recommendations for monitoring are made for certain exposed groups. Although the study found fairly large calculational uncertainties, the models developed and used are generally valid. The analysis was also used to assess potential uranium health hazards for workers in the weapons complex. No illnesses are projected for uranium workers following standard guidelines; nonetheless, some research suggests that more conservative guidelines should be considered.

  5. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Recent decisions by the Australian Government will ensure a significant expansion of the uranium industry. Development at Roxby Downs may proceed and Ranger may fulfil two new contracts but the decision specifies that apart from Roxby Downs, no new mines should be approved. The ACTU maintains an anti-uranium policy but reaction to the decision from the trade union movement has been muted. The Australian Science and Technology Council (ASTEC) has been asked by the Government to conduct an inquiry into a number of issues relating to Australia's role in the nuclear fuel cycle. The inquiry will examine in particular Australia's nuclear safeguards arrangements and the adequacy of existing waste management technology. In two additional decisions the Government has dissociated itself from a study into the feasibility of establishing an enrichment operation and has abolished the Uranium Advisory Council. Although Australian reserves account for 20% of the total in the Western World, Australia accounts for a relatively minor proportion of the world's uranium production

  6. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The French Government has decided to freeze a substantial part of its nuclear power programme. Work has been halted on 18 reactors. This power programme is discussed, as well as the effect it has on the supply of uranium by South Africa

  7. Manhattan Project Technical Series The Chemistry of Uranium (I) Chapters 1-10

    International Nuclear Information System (INIS)

    Rabinowitch, E. I.; Katz, J. J.

    1946-01-01

    This constitutes Chapters 1 through 10. inclusive, of The Survey Volume on Uranium Chemistry prepared for the Manhattan Project Technical Series. Chapters are titled: Nuclear Properties of Uranium; Properties of the Uranium Atom; Uranium in Nature; Extraction of Uranium from Ores and Preparation of Uranium Metal; Physical Properties of Uranium Metal; Chemical Properties of Uranium Metal; Intermetallic Compounds and Alloy systems of Uranium; the Uranium-Hydrogen System; Uranium Borides, Carbides, and Silicides; Uranium Nitrides, Phosphides, Arsenides, and Antimonides.

  8. Investigations of uraniumsilicide-based dispersion fuels for the use of low enrichment uranium (LEU) in research and test reactors

    International Nuclear Information System (INIS)

    Nazare, S.

    1982-07-01

    The work presents at the outset, a review of the preparation and properties of uranium silicides (U 3 Si and U 3 Si 2 ) in so far as these are relevant for their use as dispersants in research reactor fuels. The experimental work deals with the preparation and powder metallurgical processing of Al-clad miniature fuel element plates with U 3 Si- und U 3 Si-Al up to U-densities of 6.0 g U/cm 3 . The compatibility of these silicides with the Al-matrix under equilibrium conditions (873 K) and the influence of the reaction on the dimensional stability of the miniplates is described and discussed. (orig.) [de

  9. Status of fuel element technology for plate type dispersion fuels with high uranium density

    International Nuclear Information System (INIS)

    Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.

    1983-01-01

    A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration

  10. Standard test method for analysis of uranium and thorium in soils by energy dispersive X-Ray fluorescence spectroscopy

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 This test method covers the energy dispersive X-ray fluorescence (EDXRF) spectrochemical analysis of trace levels of uranium and thorium in soils. Any sample matrix that differs from the general ground soil composition used for calibration (that is, fertilizer or a sample of mostly rock) would have to be calibrated separately to determine the effect of the different matrix composition. 1.2 The analysis is performed after an initial drying and grinding of the sample, and the results are reported on a dry basis. The sample preparation technique used incorporates into the sample any rocks and organic material present in the soil. This test method of sample preparation differs from other techniques that involve tumbling and sieving the sample. 1.3 Linear calibration is performed over a concentration range from 20 to 1000 μg per gram for uranium and thorium. 1.4 The values stated in SI units are to be regarded as the standard. The inch-pound units in parentheses are for information only. 1.5 This standard...

  11. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements; Procedimentos de fabricacao de elementos combustiveis a base de dispersoes com alta concentracao de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Souza, J.A.B.; Durazzo, M., E-mail: jasouza@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2010-07-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm{sup 3} by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm{sup 3} for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  12. Dispersion of uranium in accessory apatite in crystalline rocks and its possible petrogenetic meaning

    International Nuclear Information System (INIS)

    Kral, J.; Burchart, J.

    1983-01-01

    The coefficient of variation for grain-by-grain fission track uranium analysis of apatites from igneous rocks seems to reflect the temperature of crystallization and the cooling rate. For metamorphic rocks the coefficient represents a complex record of the homogeneity of the source and of metamorphic neocrystallization. As a test case 41 West Carpathian rocks have been examined and the coefficients of variation for U in apatites found to be: granitic rocks 0.30-0.79, paragneisses 0.35-0.95, migmatites 0.55-0.87, and volcanic rocks 0.30-0.40. Most of the frequency distributions are lognormal, though for some cases a normal distribution gives a better fit, and some are incompatible with either of the two distributions. (orig.)

  13. Uranium dispersion in the coating of weak-acid-resin-deprived HTGR fuel microspheres

    International Nuclear Information System (INIS)

    Weber, G.W.; Beatty, R.L.; Tennery, V.J.; Lackey, W.J. Jr.

    1976-02-01

    The current reference HTGR recycle fuel particle is a UO 2 /UC 2 kernel with a Triso coating comprising a low-density pyrocarbon (PyC) buffer, a high-density PyC inner LTI coating, SiC, and a high-density PyC outer LTI. The kernel is fabricated from a weak-acid ion exchange resin (WAR). Microradiographic examination of coated WAR particles has demonstrated that considerable U can be transferred from the kernel to the buffer coating during fabrication. Investigation of causes of fuel dispersion has indicated several different factors that contribute to fuel redistribution if not properly controlled. The presence of a nonequilibrium UC/sub 1-x/O/sub x/ (0 less than or equal to x less than or equal to 0.3) phase had no significant effect on initiating fuel dispersion. Gross exposure of the completed fuel kernel to ambient atmosphere or to water vapor at room temperature produced very minimal levels of dispersion. Exposure of the fuel to perchloroethylene during buffer and inner LTI deposition produced massive redistribution. Fuel redistribution observed in Triso-coated particles results from permeation of the inner LTI by HCl during SiC deposition. As the decomposition of CH 3 Cl 3 Si is used to deposit SiC, chlorine is readily available during this process. The permeability of the inner LTI coating has a marked effect on the extent of this mode of fuel dispersion. LTI permeability was determined by chlorine leaching studies to be a strong function of density, coating gas dilution, and coating temperature but relatively unaffected by application of a seal coat, variations in coating thickness, and annealing at 1800 0 C. Mechanical attrition of the kernels during processing was identified as a potential source of U-bearing fines that may be incorporated into the coating in some circumstances

  14. Dispersion of long-lived radionuclides from uranium mining, milling and fuel fabrication facilities

    International Nuclear Information System (INIS)

    Pettersson, H.B.L.

    1990-11-01

    The principal aim of the study was to gain further insight into the environmental dispersion of long-lived U series radionuclides from selected part of the nuclear fuel cycle and to assess the resulting exposure of members of the public. The specific objectives of this study were: 1. To determine the levels of natural radioactivity in the vicinity of two U deposits in Sweden and to establish whether U prospecting had generated significant radiological impact on man. 2. To investigate the spatial distributions of long-lived U series radionuclides caused by the dispersion of dust from the Ranger open-pit U mine in Australia. 3. To study the uptakes of long-lived U and T series radionuclides by the waterlily in order to facilitate assessment of natural exposures to the public and predictions of exposures arising from consumption of the plant due to any subsequent discharges of water from the Ranger U mine. 4. To investigate the spatial distributions of U isotopes in environmental air as a result of the release of radionuclides from the ABB-ATOM nuclear fuel factory at Vaesteraas in Sweden. In these investigations special emphasis was given to - activity ratio techniques suitable for distinguishing between natural and operation-related concentrations and for facilitating determination of the source of radionuclide uptake in the waterlily, and - the use of passive air samplers such as 'sticky vinyl' and bioindicators in investigating the aerial dispersion of radionuclides. (author)

  15. Uranium

    International Nuclear Information System (INIS)

    Battey, G.C.; McKay, A.D.

    1988-01-01

    Production for 1986 was 4899 t U 3 O 8 (4154 t U), 30% greater than in 1985, mainly because of a 39% increase in production at Ranger. Exports for 1986 were 4166 t U 3 O 8 at an average f.o.b. unit value of $40.57/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1985-86 fiscal year was $50.2 million. Plans were announced to increase the nominal capacity of the processing plant at Ranger from 3000 t/year U 3 O 8 to 4500 t and later to 6000 t/year. Construction and initial mine development at Olympic Dam began in March. Production is planned for mid 1988 at an annual rate of 2000 t U 3 O 8 , 30 000 t Cu, and 90 000 oz (2800 kg) Au. The first long-term sales agreement was concluded in September 1986. At the Manyingee deposit, testing of the alkaline solution mining method was completed, and the treatment plant was dismantled. Spot market prices (in US$/lb U 3 O 8 ) quoted by Nuexco were generally stable. From January-October the exchange value fluctuated from US$17.00-US$17.25; for November and December it was US$16.75. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U at December 1986 were estimated as 462 000 t U, 3000 t U less than in 1985. This represents 30% of the total low-cost RAR in the WOCA (World Outside the Centrally Planned Economy Areas) countries. Australia also has 257 000 t U in the low-cost Estimated Additional Resources Category I, 29% of the WOCA countries' total resources in this category

  16. Comparison of irradiation behavior of different uranium silicide dispersion fuel element designs

    International Nuclear Information System (INIS)

    Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    1995-01-01

    Calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 were performed for various dispersion fuel element designs. Breakaway swelling criteria in the form of critical fuel volume fractions were derived with data obtained from U 3 SiAl-Al plate irradiations. The results of the analysis show that rod-type elements remain well below the pillowing threshold. However, tubular fuel elements, which behave essentially like plates, will likely develop pillows or blisters at around 90% 235 U burnup. The U 3 Si 2 -Al compounds demonstrate stable swelling behavior throughout the entire burnup range for all fuel element designs

  17. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements; Procedimentos de fabricacao de elementos combustiveis a base de dispersoes com alta concentracao de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Jose Antonio Batista de

    2011-07-01

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm{sup 3} for U{sub 3}Si{sub 2}-Al dispersion-based and 2.3 gU/cm{sup 3} for U{sub 3}O{sub 8}-Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm{sup 3} in U{sub 3}Si{sub 2}-Al dispersion and 3.2 gU/cm{sup 3} U{sub 3}O{sub 8}-Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U{sub 3}Si{sub 2}-Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U{sub 3}O{sub 8}-Al dispersion fuel plates with 3.2 gU/cm{sup 3} showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U{sub 3}Si{sub 2} production at 4.8 gU/cm{sup 3}, with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  18. Microwave-assisted of dispersive liquid-liquid microextraction and spectrophotometric determination of uranium after optimization based on Box-Behnken design and chemometrics methods

    Science.gov (United States)

    Niazi, Ali; Khorshidi, Neda; Ghaemmaghami, Pegah

    2015-01-01

    In this study an analytical procedure based on microwave-assisted dispersive liquid-liquid microextraction (MA-DLLME) and spectrophotometric coupled with chemometrics methods is proposed to determine uranium. In the proposed method, 4-(2-pyridylazo) resorcinol (PAR) is used as a chelating agent, and chloroform and ethanol are selected as extraction and dispersive solvent. The optimization strategy is carried out by using two level full factorial designs. Results of the two level full factorial design (24) based on an analysis of variance demonstrated that the pH, concentration of PAR, amount of dispersive and extraction solvents are statistically significant. Optimal condition for three variables: pH, concentration of PAR, amount of dispersive and extraction solvents are obtained by using Box-Behnken design. Under the optimum conditions, the calibration graphs are linear in the range of 20.0-350.0 ng mL-1 with detection limit of 6.7 ng mL-1 (3δB/slope) and the enrichment factor of this method for uranium reached at 135. The relative standard deviation (R.S.D.) is 1.64% (n = 7, c = 50 ng mL-1). The partial least squares (PLS) modeling was used for multivariate calibration of the spectrophotometric data. The orthogonal signal correction (OSC) was used for preprocessing of data matrices and the prediction results of model, with and without using OSC, were statistically compared. MA-DLLME-OSC-PLS method was presented for the first time in this study. The root mean squares error of prediction (RMSEP) for uranium determination using PLS and OSC-PLS models were 4.63 and 0.98, respectively. This procedure allows the determination of uranium synthesis and real samples such as waste water with good reliability of the determination.

  19. Microwave-assisted of dispersive liquid-liquid microextraction and spectrophotometric determination of uranium after optimization based on Box-Behnken design and chemometrics methods.

    Science.gov (United States)

    Niazi, Ali; Khorshidi, Neda; Ghaemmaghami, Pegah

    2015-01-25

    In this study an analytical procedure based on microwave-assisted dispersive liquid-liquid microextraction (MA-DLLME) and spectrophotometric coupled with chemometrics methods is proposed to determine uranium. In the proposed method, 4-(2-pyridylazo) resorcinol (PAR) is used as a chelating agent, and chloroform and ethanol are selected as extraction and dispersive solvent. The optimization strategy is carried out by using two level full factorial designs. Results of the two level full factorial design (2(4)) based on an analysis of variance demonstrated that the pH, concentration of PAR, amount of dispersive and extraction solvents are statistically significant. Optimal condition for three variables: pH, concentration of PAR, amount of dispersive and extraction solvents are obtained by using Box-Behnken design. Under the optimum conditions, the calibration graphs are linear in the range of 20.0-350.0 ng mL(-1) with detection limit of 6.7 ng mL(-1) (3δB/slope) and the enrichment factor of this method for uranium reached at 135. The relative standard deviation (R.S.D.) is 1.64% (n=7, c=50 ng mL(-1)). The partial least squares (PLS) modeling was used for multivariate calibration of the spectrophotometric data. The orthogonal signal correction (OSC) was used for preprocessing of data matrices and the prediction results of model, with and without using OSC, were statistically compared. MA-DLLME-OSC-PLS method was presented for the first time in this study. The root mean squares error of prediction (RMSEP) for uranium determination using PLS and OSC-PLS models were 4.63 and 0.98, respectively. This procedure allows the determination of uranium synthesis and real samples such as waste water with good reliability of the determination. Copyright © 2014. Published by Elsevier B.V.

  20. Phase equilibrium study on system uranium-plutonium-tungsten-carbon

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro

    1976-11-01

    Metallurgical properties of the U-Pu-W-C system have been studied with emphasis on phases and reactions. Free energy of compound formation, carbon activity and U/Pu segregation in the W-doped carbide fuel are estimated using phase diagram data. The results indicate that tungsten metal is useful as a thermochemical stabilizer of the carbide fuel. Tungsten has high temperature stability in contact with uranium carbide and mixed uranium-plutonium carbide. (auth.)

  1. The dispersion of radon in the Straz-Hamr area of the Czech Republic as an effect of uranium mining and related activities

    International Nuclear Information System (INIS)

    Smetana, R.; Novak, J.

    1997-01-01

    Uranium is exploited in the deposit Straz pod Ralskem-Hamr since 1968. During all the time two mining methods have existed side by side - the deep mining and the ''in situ leaching'' technology using the sulphuric acid. The uranium mining culminated in the second half of 1980s in the deposit. Higher concentrations of radon is expected in the uranium mining area. It is caused for one thing by higher content of the mother elements in the crust of the earth, for another by the various mining and reprocessing processes. To evaluate a radon exposure in the Straz-Hamr area an analysis of radon distribution was worked out. The analysis was prepared in 1986 in the mining company Uranove doly Hamr (now DIAMO s.p.) and it described dispersion of radon emitted to the air in connection with the mining activities. The sources of radon could be divided into two groups - area sources (leaching fields, ore depots, water basins) and point sources (stacks, ventilation boreholes, ventilators). This paper describes the methodology of the radon dispersion calculation, based on the stationary Gaussian model of dispersion of the gaseous contaminants from the point source. Modification of the methodology for the area sources and extension for the radioactive decay are also presented. Results of the calculations are represented graphically in the contour maps of the ground-level concentrations of radon and an assessment of the doses for the critical group is presented. (author)

  2. Fission product phases in irradiated carbide fuels

    International Nuclear Information System (INIS)

    Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

    1975-09-01

    Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

  3. Plasma metallization of refractory carbide powders

    International Nuclear Information System (INIS)

    Koroleva, E.B.; Klinskaya, N.A.; Rybalko, O.F.; Ugol'nikova, T.A.

    1986-01-01

    The effect of treatment conditions in plasma on properties of produced metallized powders of titanium, tungsten and chromium carbides with the main particle size of 40-80 μm is considered. It is shown that plasma treatment permits to produce metallized powders of carbide materials with the 40-80 μm particle size. The degree of metallization, spheroidization, chemical and phase composition of metallized carbide powders are controlled by dispersivity of the treated material, concentration of a metal component in the treated mixtures, rate of plasma flow and preliminary spheroidization procedure

  4. Hydrolysis of uranium monocarbide

    International Nuclear Information System (INIS)

    Hajek, B.; Karen, P.; Brozek, V.

    1984-01-01

    The substoichiometric uranium monocarbide UCsub(0.95) was hydrolyzed in acid medium at 80 degC. The composition of the products of hydrolysis corresponds to published data but it correlates better with the stoichiometric composition of the hydrolyzable carbide. The mechanisms of the hydrolytic reaction are discussed and a modified radical mechanism is suggested based on the concept of initiation of the radical process by Hsup(.) radicals formed owing to the nonstoichiometry of the substance. A relation is proposed for calculating the content of free hydrogen in the hydrolysis products of carbides of metallic nature for which a radical mechanism of their reaction with water can be assumed. Some effects occurring during the hydrolysis of uranium carbide, as described in literature, are explained in terms of the concept suggested. The results obtained by the authors for carbides of manganese (Mn 7 C 3 ) and for rare earth elements are discussed. (author)

  5. Radiation damage of metal uranium

    International Nuclear Information System (INIS)

    Mihajlovic, A.

    1965-01-01

    This report is concerned with the role of dispersion second phase in uranium and burnup rate. The role of dispersion phases in radiation stability of metal uranium was studies by three methods: variation of electric conductivity dependent on the neutron flux and temperature of pure uranium for different states of dispersion second phase; influence of dispersion phase on the radiation creep; transmission electron microscopy of fresh and irradiated uranium

  6. The Recovery of Uranium From The Rejected Fuel Plate Dispersion Type of U3O8-Al and U3Si2Al by NaOH

    International Nuclear Information System (INIS)

    Widodo, G; Aji, D

    1998-01-01

    The recovery of uranium from the rejected fuel plate dispersion type of U 3 O 8 -AI And U 3 Si 2 -AI with a dissolution has been performed.Each of 5 fragment of fuel plate dispersion of U 3 O 8 -AI or U 3 Si 2 Al of 1x4 cm size was put in the distilled glass content of 250 ml NaOH solution whit The concentration variation 10,15,20,25,and 30%,and than was heated at temperature of 102 o C and was stirred constantly by magnetic stirred.Uranium in the form of U 3 O 8 or U 3 Si 2 was separated by filtration and Either residu and filtrate was analyzed by potentiometry using modified Devies Gray method. From the experiment data it was found in the residu that presentation of uranium was 83.99-84.05% and 84.67-86.556% while in filtrate it was found 53.90 ppm and 69.3 ppm

  7. Geochemical study for primary dispersion of trace elements in uranium bearing black slates of the Ogcheon Group, Korea

    International Nuclear Information System (INIS)

    Kim, O.B.

    1980-01-01

    Total 145 boring core samples of Deogpyongri, Geosan and Mogsori, Geumsan in Ogcheon Group have been collected and analyzed for uranium and trace elements such as lead, zinc, copper, chromium, cadmium, vanadium and mloybdenium. All the data of the elments analyzed have been processed statistically by computer in order to estimate the correlation co-efficient between elements. The vertical distribution pattern of trace elements has been discussed. The results obtained are summarized as follows: Uranium has high correlation co-efficients with vanadium and molybdenium. And the last two can be used as indicator elements for the geochemical prospecting of uranium. The occurrence of uranium is closely related with the carbonaceous material in boring core of Ogcheon Group. Considering the vertical distribution pattern of uranium, it can't be said that the epigenetic uranium absorption to the carbonaceous material is in progress. The uranium minerals in the carbonanceous material must be correctly defined to resolve the genetic problems of uranium deposit in Ogcheon Group. (Author)

  8. Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

    Science.gov (United States)

    Grisaffe, Salvatore J.; Caves, Robert M.

    1964-01-01

    An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

  9. Viability utilization of one Se sup(75) source in the analysis of uranium, thorium and rare earths for use on energy dispersive x-ray fluorescence

    International Nuclear Information System (INIS)

    Nova Mussel, W. da.

    1989-01-01

    This work is a study about the viable utilization of one Se sup(75) source as an excitation source for the use of Energy Dispersive X-Ray Fluorescence (EDXRF), in the analysis of Uranium, Thorium and the Rare Earths. The following arrangement was build up: a HPGE detector, two Se sup(75) sources in 30 sup(0) positions of castle, deadtime of 5%. Using this arrangement the calibration curve for U and Th was measured and the angular correlation coeficient was r+ 0,999, and for the Rare Earths was superior r+ 0,960. The answer given for this system was considered very fine. (author)

  10. Influence of oxygen, nitrogen and carbon on the lattice parameter of uranium mono-carbide; Influence de l'oxygene, de l'azote et du carbone sur le parametre reticulaire du monocarbure d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Magnier, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1966-04-15

    The author studies the influence of oxygen and nitrogen contents on the lattice parameter of U(C,O,N) solid solutions around UC composition. The whole data conducts to a determination of the solubility of oxygen in UC: a U(C(1-x)O(x)) solid solution exist if x if smaller than 0.37. The author studies also the influence of carbon content on the lattice parameter of U-UC solid solutions around UC. This study conducts to the determination of the solubility of U in UC at the different temperatures. Consequences upon uranium-carbon diagram are envisaged. (author) [French] L'auteur etudie quantitativement l'influence de l'oxygene et de l'azote sur le parametre reticulaire des solutions solides U(C,O,N) proches de UC. Cette etude permet la determination de la solubilite de l'oxygene dans UC: on montre l'existence d'une solution solide U(C(1-x)O(x)) lorsque x est compris entre 0 et 0,37. Par ailleurs l'auteur etudie l'influence de la teneur en carbone sur le parametre des solutions solides U-UC proches de UC. Cette etude permet la determination de la solubilite de l'uranium dans UC aux differentes temperatures. On envisage enfin les modifications apportees par cette etude au diagramme uranium-carbone. (auteur)

  11. Study on the identification of organic and common anions in the pyrohydrolysis distillate of mixed uranium-plutonium carbide for the interference free determination of chlorine and fluorine by ion chromatography

    Energy Technology Data Exchange (ETDEWEB)

    Jeyakumar, Subbiah; Mishra, Vivekchandra Guruprasad; Das, Mrinal Kanti; Raut, Vaibhavi Vishwajeet; Sawant, Ramesh Mahadeo [Bhabha Atomic Research Centre, Mumbai (India). Radioanalytical Chemistry Div.; Ramakumar, Karanam Lakshminarayana [Bhabha Atomic Research Centre, Mumbai (India). Radiochemistry and Isotope Group

    2014-07-01

    Identification of various soluble organic acids formed during the pyrohydrolysis of uranium-plutonium mixed carbide [(U,Pu)C] was carried out using ion chromatography. This has significant importance as the soluble organic acids can cause severe interferences during the ion chromatography separation and determination of Cl{sup -} and F{sup -} in the pyrohydrolysis distillate of (U,Pu)C. Determination of Cl and F is important in the chemical quality control of nuclear materials as these two elements can cause corrosion and hence, their concentrations in all nuclear materials are restricted to certain specified values. Since the pyrohydrolysis distillates contain both inorganic and organic acid anions, for the sake of separating and identifying organic acid anions from the common inorganic anions, three independent isocratic elutions using varying concentrations of NaOH eluent were employed for the separation of weakly, moderately and strongly retained anions. It was observed that pyrohydrolysis of (U,Pu)C also produced soluble organic acids as in the case of nitric acid dissolution of UC. The present investigation revealed the presence of formic, acetic, propionic, butyric, oxalic acid anions in the pyrohydrolysis distillate of (U,Pu)C in trace or ultra-trace concentrations. The presence of each organic acid identified in the chromatogram was confirmed with spike addition as well as by separating them by capillary electrophoresis method. The presence of lower aliphatic acids viz. formic and acetic acids was reconfirmed by carrying out an independent separation with tetraborate eluent. It is suggested that nitric acid being formed during pyrohydrolysis could be responsible for the formation of organic acids. Based on the findings, an ion chromatography separation method has been proposed for the interference-free determination of chloride and fluoride in pyrohydrolysis distillate of (U,Pu)C. (orig.)

  12. Uranium hexafluoride reconversion used for dispersion fuel elements fabrication for IEAR-1/SP reactor; Reconversao de hexafluoreto de uranio para a fabricacao de combustiveis na forma de dispersoes para o reator IEA-R1/SP

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, E.F. Urano de; Lainetti, P.E.; Gomes, R.P. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    1996-07-01

    In this paper are described the main chemical process employed in the Chemical Processes Division of the Fuel Technology Department - IPEN for conversion of enriched UF{sub 6} in ammonium diuranate - DUA and uranium tetrafluoride - UF{sub 4}. These activities have assured the continuity of fuel elements production at IPEN since 1984. The uranium recovery from scraps of the fuel elements production and the purification processes are also described. Those compounds are important intermediate products in the fabrication routine and in development dispersed fuel elements with higher uranium loading for IEA{sub R}1 research reactor power increase program. (author)

  13. Calculation of vapour pressures over mixed carbide fuels

    International Nuclear Information System (INIS)

    Joseph, M.; Mathews, C.K.

    1988-01-01

    Vapour pressure over the uranium-plutonium mixed carbide (Usub(l-p) Pusub(p C) was calculated in the temperature range of 1300-9000 for various compositions (p=0.1 to 0.7). Effects of variation of the sesquicarbide content were also studied. The principle of corresponding states was applied to UC and mixed carbides to obtain the equation of state. (author)

  14. Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

    International Nuclear Information System (INIS)

    Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

    1999-01-01

    This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

  15. Nuclear-fuel-cycle education: Module 2. Exploration, reserve estimation, mining, milling, conversion, and properties of uranium

    International Nuclear Information System (INIS)

    Brookins, D.G.

    1981-12-01

    In this module geological and geochemical data pertinent to locating, mining, and milling of uranium are examined. Chapters are devoted to: uranium source characteristics; uranium ore exploration methods; uranium reserve estimation for sandstone deposits; mining; milling; conversion processes for uranium; and properties of uranium, thorium, plutonium and their oxides and carbides

  16. Three-dimensional charge dispersion curves from interactions of 11--29 GeV protons with uranium

    International Nuclear Information System (INIS)

    Yu, Y.

    1980-01-01

    Experimental nuclear charge dispersion curves from interactions of 11--29 Gev protons with 238 U have been used in the construction of three-dimensional charge dispersion curves. They show the yield variation with mass number A. Neutron-deficient products are distributed over the entire mass range with a peak at A near 87, while the yield of neutron-excessive products is distributed only in the relatively narrow mass region between A=70 and A=150 and has a maximum around A=115. An isobaric yield curve has been obtained by summing up each of the charge dispersion curves and shows a peak, rather than the flat top, in the mass region A=80 to 140 reported previously. The mass yield curves of neutron-excessive and neutron-deficient products are obtained by a decomposition of the charge dispersion curve with two Gaussians, and the mechanism of formation is suggested

  17. THE INTERPLAY BETWEEN GEOCHEMICAL REACTIONS AND ADVECTION-DISPERSION IN CONTAMINANT TRANSPORT AT A URANIUM MILL TAILINGS SITE

    Science.gov (United States)

    It is well known that the fate and transport of contaminants in the subsurface are controlled by complex processes including advection, dispersion-diffusion, and chemical reactions. However, the interplay between the physical transport processes and chemical reactions, and their...

  18. Application of the HGSYSTEM/UF6 model to simulate atmospheric dispersion of UF6 releases from uranium enrichment plants

    International Nuclear Information System (INIS)

    Goode, W.D. Jr.; Bloom, S.G.; Keith, K.D. Jr.

    1995-01-01

    Uranium hexafluoride is a dense, reactive gas used in Gaseous Diffusion Plants (GDPs) to make uranium enriched in the 235 U isotope. Large quantities of UF 6 exist at the GDPs in the form of in-process gas and as a solid in storage cylinders; smaller amounts exist as hot liquid during transfer operations. If liquid UF 6 is released to the environment, it immediately flashes to a solid and a dense gas that reacts rapidly with water vapor in the air to form solid particles of uranyl fluoride and hydrogen fluoride gas. Preliminary analyses were done on various accidental release scenarios to determine which scenarios must be considered in the safety analyses for the GDPS. These scenarios included gas releases due to failure of process equipment and liquid/gas releases resulting from a breach of transfer piping from a cylinder. A major goal of the calculations was to estimate the response time for mitigating actions in order to limit potential off-site consequences of these postulated releases. The HGSYSTEM/UF 6 code was used to assess the consequences of these release scenarios. Inputs were developed from release calculations which included two-phase, choked flow followed by expansion to atmospheric pressure. Adjustments were made to account for variable release rates and multiple release points. Superpositioning of outputs and adjustments for exposure time were required to evaluate consequences based on health effects due to exposures to uranium and HF at a specific location

  19. Morphology study of refractory carbide powders

    International Nuclear Information System (INIS)

    Vavrda, J.; Blazhikova, Ya.

    1982-01-01

    Refractory carbides were investigated using JSM-U3 electron microscope of Joelco company at 27 KV accelerating voltage. Some photographs of each powder were taken with different enlargements to characterise the sample upon the whole. It was shown that morphological and especially topographic study of powders enables to learn their past history (way of fabrication and treatment). The presence of steps of compact particle fractures and cracks is accompanied by occurence of fine dispersion of carbides subjected to machining after facrication. On the contrary, the character of crystallographic surfaces and features of surface growth testify to the way of crystallization

  20. Tool steel for cold worck niobium carbides

    International Nuclear Information System (INIS)

    Goldenstein, H.

    1984-01-01

    A tool steel was designed so as to have a microstructure with the matrix similar a cold work tool steel of D series, containing a dispersion of Niobium carbides, with no intention of putting Niobium in solution on the matrix. The alloy was cast, forged and heat treated. The alloy was easily forged; the primary carbide morfology, after forging, was faceted, tending to equiaxed. The hardness obtained was equivalent to the maximum hardness of a D-3 sttel when quenched from any temperature between 950 0 C, and 1200 0 , showing a very small sensitivy to the quenching temperature. (Author) [pt

  1. Corrosion resistant cemented carbide

    International Nuclear Information System (INIS)

    Hong, J.

    1990-01-01

    This paper describes a corrosion resistant cemented carbide composite. It comprises: a granular tungsten carbide phase, a semi-continuous solid solution carbide phase extending closely adjacent at least a portion of the grains of tungsten carbide for enhancing corrosion resistance, and a substantially continuous metal binder phase. The cemented carbide composite consisting essentially of an effective amount of an anti-corrosion additive, from about 4 to about 16 percent by weight metal binder phase, and with the remaining portion being from about 84 to about 96 percent by weight metal carbide wherein the metal carbide consists essentially of from about 4 to about 30 percent by weight of a transition metal carbide or mixtures thereof selected from Group IVB and of the Periodic Table of Elements and from about 70 to about 96 percent tungsten carbide. The metal binder phase consists essentially of nickel and from about 10 to about 25 percent by weight chromium, the effective amount of an anti-corrosion additive being selected from the group consisting essentially of copper, silver, tine and combinations thereof

  2. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Bocker, S. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

  3. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Bocker, S [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

  4. Process for producing uranium carbide spheroids

    International Nuclear Information System (INIS)

    Shennan, J.V.; Ford, L.H.

    1977-01-01

    The invention deals with a method to fabricate UC spheroids which are filled into moulds made of refractory material for fuel elements. The UC fuel particles are double-coated: a first thin layer of pyrolytic carbon is coated at low temperature 1200-1400 0 C, a record layer of pyrolytic material (e.g. Si c) is coated at a higher temperature (above 1500 0 C) which holds back the fission products. The method is described more closely by means of an example. (GSC) [de

  5. Process for producing uranium carbide spheroids

    International Nuclear Information System (INIS)

    Shennan, J.V.; Ford, L.H.

    1976-01-01

    The invention deals with a method to produce UC spheroids which are filled into molded bodies of fire-proof material for fuel elements. The UC fuel particles are doubly coated: a first thin layer of pyrolytic carbon is coated at low temperature (1,200-1,400 0 C), a second layer of fire-proof material (e.g. SiC) is coated at a higher temperature (above 1,500 0 C) which holds back the fission products. The process is explained in more detail using an example. (GSCH) [de

  6. Colloidal characterization of ultrafine silicon carbide and silicon nitride powders

    Science.gov (United States)

    Whitman, Pamela K.; Feke, Donald L.

    1986-01-01

    The effects of various powder treatment strategies on the colloid chemistry of aqueous dispersions of silicon carbide and silicon nitride are examined using a surface titration methodology. Pretreatments are used to differentiate between the true surface chemistry of the powders and artifacts resulting from exposure history. Silicon nitride powders require more extensive pretreatment to reveal consistent surface chemistry than do silicon carbide powders. As measured by titration, the degree of proton adsorption from the suspending fluid by pretreated silicon nitride and silicon carbide powders can both be made similar to that of silica.

  7. High loading uranium plate

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.

    1990-01-01

    Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pari of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat hiving a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process

  8. Loading ion exchange resins with uranium for HTGR fuel kernels

    International Nuclear Information System (INIS)

    Notz, K.J.; Greene, C.W.

    1976-12-01

    Uranium-loaded ion exchange beads provide an excellent starting material in the production of uranium carbide microspheres for nuclear fuel applications. Both strong-acid (sulfonate) and weak-acid (carboxylate) resins can be fully loaded with uranium from a uranyl nitrate solution utilizing either a batch method or a continuous column technique

  9. Investigation of aerial dispersion of radioactive dust from an open-pit uranium mine by passive vinyl collectors

    International Nuclear Information System (INIS)

    Pettersson, H.B.; Koperski, J.

    1991-01-01

    Detailed investigations of the aerial dispersion of radioactive dust from the biggest open-pit U mining and milling operation in Australia were carried out. Spatial distributions of the long-lived radionuclides of 238 U series and their origin, i.e., mining and milling operations vs. natural background radiation, have been studied. Horizontal flux, dry deposition, and ground resuspension of the radionuclides were investigated along a 50-km transect in the direction of the prevailing monsoonal winds in the region. The study was performed by means of unconventional 'sticky vinyl' passive dust collectors, occasionally supported by high-volume air filter samplers. The data from the flux measurements show an inverse square to inverse cubic dependence, and the dry deposition exhibits an inverse square dependence, of radionuclide load vs. distance. The pit has been the predominant contributor of long-lived U series radionuclides to the environment within the radius of several kilometers from the operations. An aerial dispersion computer code (LUCIFER), based on a Gaussian plume model, was developed for the project. Experimental data were used as the code input data. Good agreement between the measured data and the normalized computed results was obtained

  10. Oxalate complexation in dissolved carbide systems

    International Nuclear Information System (INIS)

    Choppin, G.R.; Bokelund, H.; Valkiers, S.

    1983-01-01

    It has been shown that the oxalic acid produced in the dissolution of mixed uranium, plutonium carbides in nitric acid can account for the problems of incomplete uranium and plutonium extraction on the Purex process. Moreover, it was demonstrated that other identified products such as benzene polycarboxylic acids are either too insoluble or insufficiently complexing to be of concern. The stability constants for oxalate complexing of UO 2 +2 and Pu +4 ions (as UO 2 (C 2 O 4 ), Pu(C 2 O 4 ) 2+ and Pu(C 2 O 4 ) 2 , respectively) were measured in nitrate solutions of 4.0 molar ionic strength (0-4 M HNO 3 ) by extraction of these species with TBP. (orig.)

  11. Combination of solid phase extraction and dispersive liquid–liquid microextraction for separation/preconcentration of ultra trace amounts of uranium prior to its fiber optic-linear array spectrophotometry determination

    International Nuclear Information System (INIS)

    Dadfarnia, Shayessteh; Shabani, Ali Mohammad Haji; Shakerian, Farid; Shiralian Esfahani, Golnaz

    2013-01-01

    Graphical abstract: Pass the sample through the basic alumina column ⇒ elute retained uranium along with the cations ⇒ convert the uranium to its anionic benzoate complex ⇒ extract its ion pair with malachite green into small volume of chloroform by DLLME ⇒ measure its absorption at 621 nm using fiber optic-linear array detection spectrophotometry. -- Highlights: • By combination of SPE and DDLME a high preconcentration factor of 2500 was obtained. • Development of SPE-DDLME-Spectrophotometric method for det. of trace amounts of uranium. • Ultra trace amount of uranium in water samples was det. by the proposed method. • The detection limit of the proposed method is comparable to the most sensitive method. • The proposed method is a free interference spectrophotometric method for uranium det. -- Abstract: A simple and sensitive method for the separation and preconcentration of the ultra trace amounts of uranium and its determination by spectrophotometry was developed. The method is based on the combination of solid phase extraction and dispersive liquid–liquid microextraction. Thus, by passing the sample through the basic alumina column, the uranyl ion and some cations are separated from the sample matrix. The retained uranyl ion along with the cations are eluted with 5 mL of nitric acid (2 mol L −1 ) and after neutralization of the eluent, the extracted uranyl ion is converted to its anionic benzoate complex and is separated from other cations by extraction of its ion pair with malachite green into small volume of chloroform using dispersive liquid–liquid microextraction. The amount of uranium is then determined by the absorption measurement of the extracted ion pair at 621 nm using flow injection spectrophotometry. Under the optimum conditions, with 500 mL of the sample, a preconcentration factor of 1980, a detection limit of 40 ng L −1 , and a relative standard deviation of 4.1% (n = 6) at 400 ng L −1 were obtained. The method was

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

    International Nuclear Information System (INIS)

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

    2001-01-01

    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

  13. Shock Response of Boron Carbide

    National Research Council Canada - National Science Library

    Dandekar, D. P. (Dattatraya Purushottam)

    2001-01-01

    .... The present work was undertaken to determine tensile/spall strength of boron carbide under plane shock wave loading and to analyze all available shock compression data on boron carbide materials...

  14. A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Drera, Saleem S., E-mail: saleem.drera@gmail.com [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); Hofman, Gerard L. [Argonne National Laboratory, Chicago, IL 60439 (United States); Kee, Robert J. [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); King, Jeffrey C. [Metallurgical and Materials Engineering, Colorado School of Mines, Golden, CO 80401 (United States)

    2014-10-15

    Highlights: • This article presents a cellular automata (CA) algorithm to synthesize the growth of intermetallic interaction layers in U–Mo/Al dispersion fuel. • The method utilizes a 3D representation of the fuel, which is discretized into separate voxels that can change identy based on derived CA rules. • The CA model is compared to ILT measurements for RERTR experimental data. • The primary objective of the model is to synthesize three-dimensional microstructures that can be used in subsequent thermal and mechanical modeling. • The CA model can be used for predictive analysis. For example, it can be used to study the dependence of temperature on interaction layer growth. - Abstract: Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium–molybdenum (U–Mo) particles within an aluminum matrix. Fresh U–Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction–diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

  15. Radiation damage of metal uranium; Radijaciono ostecenje metalnog urana

    Energy Technology Data Exchange (ETDEWEB)

    Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This report is concerned with the role of dispersion second phase in uranium and burnup rate. The role of dispersion phases in radiation stability of metal uranium was studies by three methods: variation of electric conductivity dependent on the neutron flux and temperature of pure uranium for different states of dispersion second phase; influence of dispersion phase on the radiation creep; transmission electron microscopy of fresh and irradiated uranium.

  16. Hadfield steels with Nb and Ti carbides

    International Nuclear Information System (INIS)

    Vatavuk, J.; Goldenstein, H.

    1987-01-01

    The Hadfield Steels and the mechanisms responsible for its high strain hardening rate were reviewed. Addition of carbide forming alloying elements to the base compostion was discussed, using the matrix sttel concept. Three experimental crusher jaws were cast, with Nb and Nb + Ti added to the usual Hadfiedl compostion, with enough excess carbon to allow the formation of MC carbides. Samples for metallographic analysis were prepared from both as cast and worn out castings. The carbic morphology was described. Partition of alloying elements was qualitatively studied, using Energy Dispersive Espectroscopy in SEM. The structure of the deformed layer near the worn surface was studied by optical metalography and microhardness measurements. The results showed that fatigue cracking is one of the wear mechanisms is operation in association with the ciclic work hardening of the surface of worn crusher jaws. (Author) [pt

  17. Combination of solid phase extraction and dispersive liquid-liquid microextraction for separation/preconcentration of ultra trace amounts of uranium prior to its fiber optic-linear array spectrophotometry determination.

    Science.gov (United States)

    Dadfarnia, Shayessteh; Shabani, Ali Mohammad Haji; Shakerian, Farid; Shiralian Esfahani, Golnaz

    2013-12-15

    A simple and sensitive method for the separation and preconcentration of the ultra trace amounts of uranium and its determination by spectrophotometry was developed. The method is based on the combination of solid phase extraction and dispersive liquid-liquid microextraction. Thus, by passing the sample through the basic alumina column, the uranyl ion and some cations are separated from the sample matrix. The retained uranyl ion along with the cations are eluted with 5 mL of nitric acid (2 mol L(-1)) and after neutralization of the eluent, the extracted uranyl ion is converted to its anionic benzoate complex and is separated from other cations by extraction of its ion pair with malachite green into small volume of chloroform using dispersive liquid-liquid microextraction. The amount of uranium is then determined by the absorption measurement of the extracted ion pair at 621 nm using flow injection spectrophotometry. Under the optimum conditions, with 500 mL of the sample, a preconcentration factor of 1980, a detection limit of 40 ng L(-1), and a relative standard deviation of 4.1% (n=6) at 400 ng L(-1) were obtained. The method was successfully applied to the determination of uranium in mineral water, river water, well water, spring water and sea water samples. Copyright © 2013 Elsevier B.V. All rights reserved.

  18. Electrocatalysis on tungsten carbide

    International Nuclear Information System (INIS)

    Fleischmann, R.

    1975-01-01

    General concepts of electrocatalysis, the importance of the equilibrium rest potential and its standardization on polished WC-electrodes, the influence of oxygen in the catalysts upon the oxidation of hydrogen, and the attained results of the hydrogen oxidation on tungsten carbide are treated. (HK) [de

  19. Health hazards of uranium dust from radioactive battlefields of the Balkan conflicts, Eastern Afghanistan and Iraq after the Gulf wars. Lessons for civil protection in the terrorist scenario of radiological dispersion devices

    International Nuclear Information System (INIS)

    Durakovic, A.; Klimaschewski, F.

    2007-01-01

    Complete text of publication follows. Purpose: The purpose of this study is to identify key health hazards of uranium dust from the radioactive battlefields (Balkan, Middle East and Eastern Afghanistan conflicts) to draw lessons for civil protection in the terrorist scenario of radiological dispersion devices (RDD). Gulf War I (GW I) in 1991 resulted in 350 metric tons of depleted uranium (DU) deposited in the environment and 3 to 6 million grams of DU aerosol dust particles released into the atmosphere, by the most conservative estimates. Its possible legacy (Gulf War disease) continues after the military conflicts (Operation Enduring Freedom, OEF, in Afghanistan and Gulf War II in Iraq). The symptoms of the multiorgan incapacitating progressive disease have been as numerous as their names, including incapacitating fatigue, musculoskeletal and joint pains, headaches, neuropsychiatric disorders, affects changes, confusion, visual problems, changes of gait, loss of memory, lympadenopathies, respiratory impairment, impotence, and urinary tract morphological and functional alterations. The disease is still a matter of controversy regarding etiology and pathogenesis of the syndrome commonly named Gulf War disease. It was underestimated and subsequently evolved in its clinical description through recognition of progressive symptomatology. Methods: UMRC's studies of the human contamination with uranium isotopes were conducted with the exposed subjects of Jalalabad, Spin Gar, Tora Bora, and Kabul areas in Afghanistan after OEF as well as Samawah, Baghdad and Basrah in Iraq after GW II. The urine samples of the subjects were analysed by the plasma mass spectrometry. The analytical methodology involved pre-concentration of the uranium using co-precipitation and/or evaporation, oxidation of organic matter, purification of uranium with ion exchange chromatography, and mass spectrometry with a double focusing Thermo-Elemental Plasma54 multi-collector ICP-MS equipped with a

  20. Hollow microspheres with a tungsten carbide kernel for PEMFC application.

    Science.gov (United States)

    d'Arbigny, Julien Bernard; Taillades, Gilles; Marrony, Mathieu; Jones, Deborah J; Rozière, Jacques

    2011-07-28

    Tungsten carbide microspheres comprising an outer shell and a compact kernel prepared by a simple hydrothermal method exhibit very high surface area promoting a high dispersion of platinum nanoparticles, and an exceptionally high electrochemically active surface area (EAS) stability compared to the usual Pt/C electrocatalysts used for PEMFC application.

  1. Analysis of carbides and inclusions in high speed tool steels

    DEFF Research Database (Denmark)

    Therkildsen, K.T.; Dahl, K.V.

    2002-01-01

    The fracture surfaces of fatigued specimens were investigated using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS). The aim was to quantify the distribution of cracked carbides and non-metallic inclusions on the fracturesurfaces as well as on polished cross...

  2. Characterization of Nanometric-Sized Carbides Formed During Tempering of Carbide-Steel Cermets

    Directory of Open Access Journals (Sweden)

    Matus K.

    2016-06-01

    Full Text Available The aim of this article of this paper is to present issues related to characterization of nanometric-sized carbides, nitrides and/or carbonitrides formed during tempering of carbide-steel cermets. Closer examination of those materials is important because of hardness growth of carbide-steel cermet after tempering. The results obtained during research show that the upswing of hardness is significantly higher than for high-speed steels. Another interesting fact is the displacement of secondary hardness effect observed for this material to a higher tempering temperature range. Determined influence of the atmosphere in the sintering process on precipitations formed during tempering of carbide-steel cermets. So far examination of carbidesteel cermet produced by powder injection moulding was carried out mainly in the scanning electron microscope. A proper description of nanosized particles is both important and difficult as achievements of nanoscience and nanotechnology confirm the significant influence of nanocrystalline particles on material properties even if its mass fraction is undetectable by standard methods. The following research studies have been carried out using transmission electron microscopy, mainly selected area electron diffraction and energy dispersive spectroscopy. The obtained results and computer simulations comparison were made.

  3. Determination of carbon in uranium and its compounds

    International Nuclear Information System (INIS)

    Perez-Garcia, M. M.

    1972-01-01

    This paper collects the analytical methods used our laboratories for the determination of carbon in uranium metal, uranate salts and the oxides, fluorides and carbides of uranium. The carbon is usually burned off in a induction or resistance oven under oxygen flow. The CO 2 is collected in barite solution. Where it is backtitrated with potassium biphthalate. (Author)

  4. Joining elements of silicon carbide

    International Nuclear Information System (INIS)

    Olson, B.A.

    1979-01-01

    A method of joining together at least two silicon carbide elements (e.g.in forming a heat exchanger) is described, comprising subjecting to sufficiently non-oxidizing atmosphere and sufficiently high temperature, material placed in space between the elements. The material consists of silicon carbide particles, carbon and/or a precursor of carbon, and silicon, such that it forms a joint joining together at least two silicon carbide elements. At least one of the elements may contain silicon. (author)

  5. Three-dimensional studies of intergranular carbides in austenitic stainless steel.

    Science.gov (United States)

    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.

  6. High density dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.

    1996-01-01

    A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm 3 of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm -3 with U 3 Si 2 as fuel. High-density uranium compounds offer no real density advantage over U 3 Si 2 and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U 3 Si has approximately a 30% higher uranium density but the density of the U 6 X compounds would yield the factor 1.5 needed to achieve 9 g cm -3 uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure α-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic γ phase at low temperatures where normally α phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing

  7. Recovery of uranium values

    International Nuclear Information System (INIS)

    Rowden, G.A.

    1982-01-01

    A process is provided for the recovery of uranium from an organic extractant phase containing an amine. The extractant phase is contacted in a number of mixing stages with an acidic aqueous stripping phase containing sulphate ions, and the phases are passed together through a series of mixing stages while maintaining a dispersion of droplets of one phase in the other. Uranium is precipitated from the final stage by raising the pH. An apparatus having several mixing chambers is described

  8. Carbon potential measurement on some actinide carbides

    International Nuclear Information System (INIS)

    Anthonysamy, S.; Ananthasivan, K.; Kaliappan, I.; Chandramouli, V.; Vasudeva Rao, P.R.; Mathews, C.K.; Jacob, K.T.

    1994-01-01

    Uranium-Plutonium mixed carbides with a Pu/(U+Pu) ratio of 0.55 are to be used as the fuel in the Fast Breeder Test Reactor (FBTR) at Kalpakkam, India. Carburization of the stainless steel clad by this fuel is determined by its carbon potential. Because the carbon potential of this fuel composition is not available in the literature, it was measured by the methane-hydrogen gas equilibration technique. The sample was equilibrated with purified hydrogen and the equilibrium methane-to-hydrogen ratio in the gas phase was measured with a flame ionization detector. The carbon potential of the ThC-ThC 2 as well as Mo-Mo 2 C system, which is an important binary in the actinide-fission product-carbon systems, were also measured by this technique in the temperature range 973 to 1,173 K. The data for the Mo-Mo 2 C system are in agreement with values reported in the literature. The results for the ThC-ThC 2 system are different from estimated values with large uncertainty limits given in the literature. The data on (U, Pu) mixed carbides indicates the possibility of stainless steel clad attack under isothermal equilibrium conditions

  9. Metal Carbides for Biomass Valorization

    Directory of Open Access Journals (Sweden)

    Carine E. Chan-Thaw

    2018-02-01

    Full Text Available Transition metal carbides have been utilized as an alternative catalyst to expensive noble metals for the conversion of biomass. Tungsten and molybdenum carbides have been shown to be effective catalysts for hydrogenation, hydrodeoxygenation and isomerization reactions. The satisfactory activities of these metal carbides and their low costs, compared with noble metals, make them appealing alternatives and worthy of further investigation. In this review, we succinctly describe common synthesis techniques, including temperature-programmed reaction and carbothermal hydrogen reduction, utilized to prepare metal carbides used for biomass transformation. Attention will be focused, successively, on the application of transition metal carbide catalysts in the transformation of first-generation (oils and second-generation (lignocellulose biomass to biofuels and fine chemicals.

  10. ENTIRELY AQUEOUS SOLUTION-GEL ROUTE FOR THE PREPARATION OF ZIRCONIUM CARBIDE, HAFNIUM CARBIDE AND THEIR TERNARY CARBIDE POWDERS

    Directory of Open Access Journals (Sweden)

    Zhang Changrui

    2016-07-01

    Full Text Available An entirely aqueous solution-gel route has been developed for the synthesis of zirconium carbide, hafnium carbide and their ternary carbide powders. Zirconium oxychloride (ZrOCl₂.8H₂O, malic acid (MA and ethylene glycol (EG were dissolved in water to form the aqueous zirconium carbide precursor. Afterwards, this aqueous precursor was gelled and transformed into zirconium carbide at a relatively low temperature (1200 °C for achieving an intimate mixing of the intermediate products. Hafnium and the ternary carbide powders were also synthesized via the same aqueous route. All the zirconium, hafnium and ternary carbide powders exhibited a particle size of ∼100 nm.

  11. Depleted uranium hexafluoride: Waste or resource?

    International Nuclear Information System (INIS)

    Schwertz, N.; Zoller, J.; Rosen, R.; Patton, S.; Bradley, C.; Murray, A.

    1995-07-01

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

  12. Depleted uranium

    International Nuclear Information System (INIS)

    Huffer, E.; Nifenecker, H.

    2001-02-01

    This document deals with the physical, chemical and radiological properties of the depleted uranium. What is the depleted uranium? Why do the military use depleted uranium and what are the risk for the health? (A.L.B.)

  13. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)], E-mail: mfarhan_73@yahoo.co.uk; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)

    2008-09-15

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.

  14. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2008-01-01

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease

  15. Microstructural Study of Titanium Carbide Coating on Cemented Carbide

    DEFF Research Database (Denmark)

    Vuorinen, S.; Horsewell, Andy

    1982-01-01

    Titanium carbide coating layers on cemented carbide substrates have been investigated by transmission electron microscopy. Microstructural variations within the typically 5µm thick chemical vapour deposited TiC coatings were found to vary with deposit thickness such that a layer structure could...... be delineated. Close to the interface further microstructural inhomogeneities were obsered, there being a clear dependence of TiC deposition mechanism on the chemical and crystallographic nature of the upper layers of the multiphase substrate....

  16. Dispersion toughened ceramic composites and method for making same

    Science.gov (United States)

    Stinton, D.P.; Lackey, W.J.; Lauf, R.J.

    1984-09-28

    Ceramic composites exhibiting increased fracture toughness are produced by the simultaneous codeposition of silicon carbide and titanium disilicide by chemical vapor deposition. A mixture of hydrogen, methyltrichlorosilane and titanium tetrachloride is introduced into a furnace containing a substrate such as graphite or silicon carbide. The thermal decomposition of the methyltrichlorosilane provides a silicon carbide matrix phase and the decomposition of the titanium tetrachloride provides a uniformly dispersed second phase of the intermetallic titanium disilicide within the matrix phase. The fracture toughness of the ceramic composite is in the range of about 6.5 to 7.0 MPa..sqrt..m which represents a significant increase over that of silicon carbide.

  17. Preconcentration of uranium in water samples using dispersive liquid-liquid micro- extraction coupled with solid-phase extraction and determination with inductively coupled plasma-optical emission spectrometry

    Directory of Open Access Journals (Sweden)

    M. Rezaee,

    2015-10-01

    Full Text Available A new liquid phase microextraction method based on the dispersion of an extraction solvent into aqueous phase coupled with solid-phase extraction was investigated for the extraction, preconcentration and determination of uranium in water samples. 1-(2-Pyridylazo-2-naphthol reagent (PAN at pH 6.0 was used as a chelating agent prior to extraction. After concentration and purification of the samples in SPE C18 sorbent, 1.5 mL elution sample containing 40.0 µL chlorobenzene was injected into the 5.0 mL pure water. After extraction and centrifuging, the sedimented phase was evaporated and the residue was dissolved in nitric acid (0.5 M and was injected by injection valve into the ICP-OES. Some important extraction parameters, such as sample solution flow rate, sample pH, type and volume of extraction and disperser solvents as well as the salt addition were studied and optimized. Under the optimum conditions, the calibration graph was linear in the range of 0.5-500 µg L-1. The detection limit was 0.1 µg L-1. The relative standard deviation (RSD at 5.0 µg L-1 concentration level was 6.6%. Finally, the developed method was successfully applied to the extraction and determination of uranium in the well, river, mineral, waste and tap water samples and satisfactory results were obtained.DOI: http://dx.doi.org/10.4314/bcse.v29i3.4

  18. Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

    2009-08-15

    The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

  19. Uranium conversion

    International Nuclear Information System (INIS)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

    2006-03-01

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF 6 and UF 4 are present require equipment that is made of corrosion resistant material

  20. Uranium exploration

    International Nuclear Information System (INIS)

    De Voto, R.H.

    1984-01-01

    This paper is a review of the methodology and technology that are currently being used in varying degrees in uranium exploration activities worldwide. Since uranium is ubiquitous and occurs in trace amounts (0.2 to 5 ppm) in virtually all rocks of the crust of the earth, exploration for uranium is essentially the search of geologic environments in which geologic processes have produced unusual concentrations of uranium. Since the level of concentration of uranium of economic interest is dependent on the present and future price of uranium, it is appropriate here to review briefly the economic realities of uranium-fueled power generation. (author)

  1. Effect of Ti additive on (Cr, Fe)7C3 carbide in arc surfacing layer and its refined mechanism

    International Nuclear Information System (INIS)

    Zhou Yefei; Yang Yulin; Yang Jian; Hao Feifei; Li Da; Ren Xuejun; Yang Qingxiang

    2012-01-01

    Arc surfacing layer of hypoeutectic high chromium cast iron (HCCI) expects refiner carbides in the microstructure to improve its mechanical properties. In this paper, Ti additive as a strong carbide forming element was added in the hypoeutectic HCCI arc surfacing layer. Microstructure of titaniferous hypoeutectic HCCI was studied by optical microscopy, X-ray diffraction and field emission scanning electronic microscopy with energy dispersive spectrometer. Furthermore, the M(M = Cr, Fe) 7 C 3 carbide refinement mechanism was explained by the phase diagram calculation and lattice misfit theory. The results show that, the M 7 C 3 carbide in arc surfacing microstructure of hypoeutectic HCCI has been refined with 2 wt.% Ti additive, and TiC carbide can be observed in/around the M 7 C 3 carbide. With Ti addictive increasing, the micro-hardness along the depth in profile section of layer becomes more uniform, and the wear resistance has been improved. According to the phase diagram calculation, MC carbide precipitates prior to M 7 C 3 carbide in Fe-Cr-C-Ti alloy. In addition, the lattice misfit between (1 1 0) TiC and (010) Cr 7 C 3 is 9.257%, which indicates that the TiC as heterogeneous nuclei of the M 7 C 3 is medium effective. Therefore, the M 7 C 3 carbide can be refined.

  2. Effect of electroslag remelting on carbides in 8Cr13MoV martensitic stainless steel

    Science.gov (United States)

    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.

  3. TRANSFORMATIONS IN NANO-DIAMONDS WITH FORMATION OF NANO-POROUS SILICON CARBIDE AT HIGH PRESSURE

    Directory of Open Access Journals (Sweden)

    V. N. Kovalevsky

    2010-01-01

    Full Text Available The paper contains investigations on regularities of diamond - silicon carbide composite structure formation at impact-wave excitation. It has been determined that while squeezing a porous blank containing Si (SiC nano-diamond by explosive detonation products some processes are taking place such as diamond nano-particles consolidation, reverse diamond transition into graphite, fragments formation from silicon carbide. A method for obtaining high-porous composites with the presence of ultra-disperse diamond particles has been developed. Material with three-dimensional high-porous silicon-carbide structure has been received due to nano-diamond graphitation at impact wave transmission and plastic deformation. The paper reveals nano-diamonds inverse transformation into graphite and its subsequent interaction with the silicon accompanied by formation of silicon-carbide fragments with dimensions of up to 100 nm.

  4. Porous silicon carbide (SIC) semiconductor device

    Science.gov (United States)

    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.

  5. Uranium and the use of depleted uranium in weaponry

    International Nuclear Information System (INIS)

    Roussel, R.

    2000-01-01

    In this brief report the author shows that the use of shells involving a load of depleted uranium might lead to lasting hazards to civil population and environment. These hazards come from the part of the shell that has been dispersed as contaminating radioactive dusts. The author describes some features of radioactivity and highlights the role of Uranium-238 as a provider of energy to the planet. (A.C.)

  6. New Icosahedral Boron Carbide Semiconductors

    Science.gov (United States)

    Echeverria Mora, Elena Maria

    Novel semiconductor boron carbide films and boron carbide films doped with aromatic compounds have been investigated and characterized. Most of these semiconductors were formed by plasma enhanced chemical vapor deposition. The aromatic compound additives used, in this thesis, were pyridine (Py), aniline, and diaminobenzene (DAB). As one of the key parameters for semiconducting device functionality is the metal contact and, therefore, the chemical interactions or band bending that may occur at the metal/semiconductor interface, X-ray photoemission spectroscopy has been used to investigate the interaction of gold (Au) with these novel boron carbide-based semiconductors. Both n- and p-type films have been tested and pure boron carbide devices are compared to those containing aromatic compounds. The results show that boron carbide seems to behave differently from other semiconductors, opening a way for new analysis and approaches in device's functionality. By studying the electrical and optical properties of these films, it has been found that samples containing the aromatic compound exhibit an improvement in the electron-hole separation and charge extraction, as well as a decrease in the band gap. The hole carrier lifetimes for each sample were extracted from the capacitance-voltage, C(V), and current-voltage, I(V), curves. Additionally, devices, with boron carbide with the addition of pyridine, exhibited better collection of neutron capture generated pulses at ZERO applied bias, compared to the pure boron carbide samples. This is consistent with the longer carrier lifetimes estimated for these films. The I-V curves, as a function of external magnetic field, of the pure boron carbide films and films containing DAB demonstrate that significant room temperature negative magneto-resistance (> 100% for pure samples, and > 50% for samples containing DAB) is possible in the resulting dielectric thin films. Inclusion of DAB is not essential for significant negative magneto

  7. Production of silicon carbide bodies

    International Nuclear Information System (INIS)

    Parkinson, K.

    1981-01-01

    A body consisting essentially of a coherent mixture of silicon carbide and carbon for subsequent siliconising is produced by casting a slip comprising silicon carbide and carbon powders in a porous mould. Part of the surface of the body, particularly internal features, is formed by providing within the mould a core of a material which retains its shape while casting is in progress but is compressed by shrinkage of the cast body as it dries and is thereafter removable from the cast body. Materials which are suitable for the core are expanded polystyrene and gelatinous products of selected low elastic modulus. (author)

  8. High yield silicon carbide prepolymers

    International Nuclear Information System (INIS)

    Baney, R.H.

    1982-01-01

    Prepolymers which exhibit good handling properties, and are useful for preparing ceramics, silicon carbide ceramic materials and articles containing silicon carbide, are polysilanes consisting of 0 to 60 mole% (CH 3 ) 2 Si units and 40 to 100 mole% CH 3 Si units, all Si valences being satisfied by CH 3 groups, other Si atoms, or by H atoms, the latter amounting to 0.3 to 2.1 weight% of the polysilane. They are prepared by reducing the corresponding chloro- or bromo-polysilanes with at least the stoichiometric amount of a reducing agent, e.g. LiAlH 4 . (author)

  9. LIQUID METAL COMPOSITIONS CONTAINING URANIUM

    Science.gov (United States)

    Teitel, R.J.

    1959-04-21

    Liquid metal compositions containing a solid uranium compound dispersed therein is described. Uranium combines with tin to form the intermetallic compound USn/sub 3/. It has been found that this compound may be incorporated into a liquid bath containing bismuth and lead-bismuth components, if a relatively small percentage of tin is also included in the bath. The composition has a low thermal neutron cross section which makes it suitable for use in a liquid metal fueled nuclear reactor.

  10. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Clark, C.R.; Meyer, M.K.

    1998-01-01

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  11. Precipitation Behavior of Carbides in H13 Hot Work Die Steel and Its Strengthening during Tempering

    Directory of Open Access Journals (Sweden)

    Angang Ning

    2017-02-01

    Full Text Available The properties of carbides, such as morphology, size, and type, in H13 hot work die steel were studied with optical microscopy, transmission electron microscopy, electron diffraction, and energy dispersive X-ray analysis; their size distribution and quantity after tempering, at different positions within the ingot, were analyzed using Image-Pro Plus software. Thermodynamic calculations were also performed for these carbides. The microstructures near the ingot surface were homogeneous and had slender martensite laths. Two kinds of carbide precipitates have been detected in H13: (1 MC and M6C, generally smaller than 200 nm; and (2 M23C6, usually larger than 200 nm. MC and M6C play the key role in precipitation hardening. These are the most frequent carbides precipitating at the halfway point from the center of the ingot, and the least frequent at the surface. From the center of the ingot to its surface, the size and volume fraction of the carbides decrease, and the toughness improves, while the contribution of the carbides to the yield strength increases.

  12. Effect of magnetic field on the carbide precipitation during tempering of a molybdenum-containing steel

    International Nuclear Information System (INIS)

    Hou, T.P.; Li, Y.; Zhang, J.J.; Wu, K.M.

    2012-01-01

    The influence of a high magnetic field on the carbide precipitation during the tempering of an Fe–2.8C–3.0Mo(wt%) steel was investigated. As-quenched steels were tempered at 200 °C for various times with and without the presence of 12-T magnetic field. The applied field effectively promoted the precipitation of the relatively high-temperature monoclinic χ-Fe 5 C 2 carbide, compared to the usual ε-Fe 2 C and η-Fe 2 C carbides precipitated without magnetic field. It is believed that the effect of applying a magnetic field is due to the reduction in the Gibbs free energy of the relatively higher magnetization phase. The denser distributions of the metastable carbides are attributed to the increased nucleation rate due to additional transformation force. The dispersed precipitation strengthening compensated for the decrease of hardness due to the loss of supersaturation of carbon atoms in the matrix. - Highlights: ► Applied field promoted the precipitation of χ-Fe 5 C 2 carbide. ► Promotion of the transition carbide was attributed to its higher magnetization. ► Increase in hardness was counterbalanced by the reduction in carbon content.

  13. Ab initio study on structural stability of uranium carbide

    International Nuclear Information System (INIS)

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

    2013-01-01

    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 C 44 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

  14. Czechoslovak uranium

    International Nuclear Information System (INIS)

    Pluskal, O.

    1992-01-01

    Data and knowledge related to the prospecting, mining, processing and export of uranium ores in Czechoslovakia are presented. In the years between 1945 and January 1, 1991, 98,461.1 t of uranium were extracted. In the period 1965-1990 the uranium industry was subsidized from the state budget to a total of 38.5 billion CSK. The subsidies were put into extraction, investments and geologic prospecting; the latter was at first, ie. till 1960 financed by the former USSR, later on the two parties shared costs on a 1:1 basis. Since 1981 the prospecting has been entirely financed from the Czechoslovak state budget. On Czechoslovak territory uranium has been extracted from deposits which may be classified as vein-type deposits, deposits in uranium-bearing sandstones and deposits connected with weathering processes. The future of mining, however, is almost exclusively being connected with deposits in uranium-bearing sandstones. A brief description and characteristic is given of all uranium deposits on Czechoslovak territory, and the organization of uranium mining in Czechoslovakia is described as is the approach used in the world to evaluate uranium deposits; uranium prices and actual resources are also given. (Z.S.) 3 figs

  15. Superconductivity in borides and carbides

    International Nuclear Information System (INIS)

    Muranaka, Takahiro

    2007-01-01

    It was thought that intermetallic superconductors do not exhibit superconductivity at temperatures over 30 K because of the Bardeen-Cooper-Schrieffer (BCS) limit; therefore, researchers have been interested in high-T c cuprates. Our group discovered high-T c superconductivity in MgB 2 at 39 K in 2001. This discovery has initiated a substantial interest in the potential of high-T c superconductivity in intermetallic compounds that include 'light' elements (borides, carbides, etc.). (author)

  16. Helium diffusion in irradiated boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1981-03-01

    Boron carbide has been internationally adopted as the neutron absorber material in the control and safety rods of large fast breeder reactors. Its relatively large neutron capture cross section at high neutron energies provides sufficient reactivity worth with a minimum of core space. In addition, the commercial availability of boron carbide makes it attractive from a fabrication standpoint. Instrumented irradiation experiments in EBR-II have provided continuous helium release data on boron carbide at a variety of operating temperatures. Although some microstructural and compositional variations were examined in these experiments most of the boron carbide was prototypic of that used in the Fast Flux Test Facility. The density of the boron carbide pellets was approximately 92% of theoretical. The boron carbide pellets were approximately 1.0 cm in diameter and possessed average grain sizes that varied from 8 to 30 μm. Pellet centerline temperatures were continually measured during the irradiation experiments

  17. The structure and function of supported molybdenum nitride and molybdenum carbide hydrotreating catalysts

    Science.gov (United States)

    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

  18. Determination of carbon in uranium and its compounds; Determinacion de carbono en uranio metal y sus compuestos

    Energy Technology Data Exchange (ETDEWEB)

    Perez-Garcia, M M

    1972-07-01

    This paper collects the analytical methods used our laboratories for the determination of carbon in uranium metal, uranate salts and the oxides, fluorides and carbides of uranium. The carbon is usually burned off in a induction or resistance oven under oxygen flow. The CO{sub 2} is collected in barite solution. Where it is backtitrated with potassium biphthalate. (Author)

  19. Crystallization of nodular cast iron with carbides

    Directory of Open Access Journals (Sweden)

    S. Pietrowski

    2008-12-01

    Full Text Available In this paper a crystallization process of nodular cast iron with carbides having a different chemical composition have been presented. It have been found, that an increase of molybdenum above 0,30% causes the ledeburutic carbides crystallization after (γ+ graphite eutectic phase crystallization. When Mo content is lower, these carbides crystallize as a pre-eutectic phase. In this article causes of this effect have been given.

  20. Uranium ores

    International Nuclear Information System (INIS)

    Poty, B.; Roux, J.

    1998-01-01

    The processing of uranium ores for uranium extraction and concentration is not much different than the processing of other metallic ores. However, thanks to its radioactive property, the prospecting of uranium ores can be performed using geophysical methods. Surface and sub-surface detection methods are a combination of radioactive measurement methods (radium, radon etc..) and classical mining and petroleum prospecting methods. Worldwide uranium prospecting has been more or less active during the last 50 years, but the rise of raw material and energy prices between 1970 and 1980 has incited several countries to develop their nuclear industry in order to diversify their resources and improve their energy independence. The result is a considerable increase of nuclear fuels demand between 1980 and 1990. This paper describes successively: the uranium prospecting methods (direct, indirect and methodology), the uranium deposits (economical definition, uranium ores, and deposits), the exploitation of uranium ores (use of radioactivity, radioprotection, effluents), the worldwide uranium resources (definition of the different categories and present day state of worldwide resources). (J.S.)

  1. Uranium market

    International Nuclear Information System (INIS)

    Rubini, L.A.; Asem, M.A.D.

    1990-01-01

    The historical development of the uranium market is present in two periods: The initial period 1947-1970 and from 1970 onwards, with the establishment of a commercial market. The world uranium requirements are derived from the corresponding forecast of nuclear generating capacity, with, particular emphasis to the brazilian requirements. The forecast of uranium production until the year 2000 is presented considering existing inventories and the already committed demand. The balance between production and requirements is analysed. Finally the types of contracts currently being used and the development of uranium prices in the world market are considered. (author)

  2. Uranium enrichment

    International Nuclear Information System (INIS)

    1990-01-01

    This report looks at the following issues: How much Soviet uranium ore and enriched uranium are imported into the United States and what is the extent to which utilities flag swap to disguise these purchases? What are the U.S.S.R.'s enriched uranium trading practices? To what extent are utilities required to return used fuel to the Soviet Union as part of the enriched uranium sales agreement? Why have U.S. utilities ended their contracts to buy enrichment services from DOE?

  3. Production and mechanical properties of sintered carbides (hard steels WC-Co)

    International Nuclear Information System (INIS)

    Batalha, G.F.

    1987-09-01

    Densification and mechanical characteristics or WC-Co Cemented Carbides, were investigated by dilatometry, Hardness and bending tests, as a function of the two principal micro-structural parameters: the cobalt content and the particle size of carbide crystals. Vickers hardness of the studied compositions showed a linear variation with the increase of the cobalt content. By three point bending, the transverse rupture strenght increases with cobalt content, however, for larger grain size reaches a maximum, eventually reduced by brittle phases and incomplete dispersion. The results of brittle facture tests were statistically analised and fitted better to the 'Weakest Link Model' (Weibull distribution) than the 'Chain Model' (Gaussian distribution). (author) [pt

  4. Hafnium carbide nanocrystal chains for field emitters

    International Nuclear Information System (INIS)

    Tian, Song; Li, Hejun; Zhang, Yulei; Ren, Jincui; Qiang, Xinfa; Zhang, Shouyang

    2014-01-01

    A hafnium carbide (HfC) nanostructure, i.e., HfC nanocrystal chain, was synthesized by a chemical vapor deposition (CVD) method. X-ray diffractometer, field-emission scanning electron microscope, transmission electron microscope, and energy-dispersive X-ray spectrometer were employed to characterize the product. The synthesized one-dimensional (1D) nanostructures with many faceted octahedral nanocrystals possess diameters of tens of nanometers to 500 nm and lengths of a few microns. The chain-like structures possess a single crystalline structure and preferential growth direction along the [1 0 0] crystal orientation. The growth of the chains occurred through the vapor–liquid–solid process along with a negative-feedback mechanism. The field emission (FE) properties of the HfC nanocrystal chains as the cold cathode emitters were examined. The HfC nanocrystal chains display good FE properties with a low turn-on field of about 3.9 V μm −1 and a high field enhancement factor of 2157, implying potential applications in vacuum microelectronics.

  5. Tribology of carbide derived carbon films synthesized on tungsten carbide

    Science.gov (United States)

    Tlustochowicz, Marcin

    Tribologically advantageous films of carbide derived carbon (CDC) have been successfully synthesized on binderless tungsten carbide manufactured using the plasma pressure compaction (P2CRTM) technology. In order to produce the CDC films, tungsten carbide samples were reacted with chlorine containing gas mixtures at temperatures ranging from 800°C to 1000°C in a sealed tube furnace. Some of the treated samples were later dechlorinated by an 800°C hydrogenation treatment. Detailed mechanical and structural characterizations of the CDC films and sliding contact surfaces were done using a series of analytical techniques and their results were correlated with the friction and wear behavior of the CDC films in various tribosystems, including CDC-steel, CDC-WC, CDC-Si3N4 and CDC-CDC. Optimum synthesis and treatment conditions were determined for use in two specific environments: moderately humid air and dry nitrogen. It was found that CDC films first synthesized at 1000°C and then hydrogen post-treated at 800°C performed best in air with friction coefficient values as low as 0.11. However, for dry nitrogen applications, no dechlorination was necessary and both hydrogenated and as-synthesized CDC films exhibited friction coefficients of approximately 0.03. A model of tribological behavior of CDC has been proposed that takes into consideration the tribo-oxidation of counterface material, the capillary forces from adsorbed water vapor, the carbon-based tribofilm formation, and the lubrication effect of both chlorine and hydrogen.

  6. Some evidence of uranium in volcanic feldspar rocks in the state of Sonora

    Energy Technology Data Exchange (ETDEWEB)

    Marquina M, O. E. [Uranio Mexicano, Mexico City

    1983-05-15

    Description is given of four projects of exploration and survey for uranium associated with tertiary volcanic feldspar rocks importantly dispersed in the State of Sonora and being carried out by Uranium Mexicano.

  7. Uranium mining

    International Nuclear Information System (INIS)

    Lange, G.

    1975-01-01

    The winning of uranium ore is the first stage of the fuel cycle. The whole complex of questions to be considered when evaluating the profitability of an ore mine is shortly outlined, and the possible mining techniques are described. Some data on uranium mining in the western world are also given. (RB) [de

  8. Synthesis of carbide fuels from nano-structured precursors: impact on carbo-reduction and physico-chemical properties

    International Nuclear Information System (INIS)

    Saravia, Alvaro

    2015-01-01

    The classical way classically used for manufacturing carbide fuels consists of carbo-reducing at high temperature (1600 C) and under primary vacuum a mixture of AnO 2 and graphite powders. These conditions are disadvantageous for the synthesis of mixed (U,Pu)C carbides on account of plutonium volatilization. Therefore, one of the main aims of these studies is to decrease the carbo-reduction temperature. The experiments focused mainly on the lowering of the uranium oxide temperature. This result has been obtained with the use of uranium oxide and carbon nano-structured precursors. To achieve this goal colloidal suspensions of uranium oxide have been prepared and stabilized by cellulosic ethers. Cellulosic ethers are both stabiliser for uranium oxide nanoparticles and carbon source for carbo-reduction. It has been shown that these precursors are more efficient for carbo-reduction than the standard precursors: a reduction of 300 C of carbo-reduction temperature has been obtained. The impact of these precursors on carbo-reduction and on physico-chemical properties as well as the structural and microstructural characterizations of the obtained carbides have been carried out. (author) [fr

  9. Uranium enrichment

    International Nuclear Information System (INIS)

    1989-01-01

    GAO was asked to address several questions concerning a number of proposed uranium enrichment bills introduced during the 100th Congress. The bill would have restructured the Department of Energy's uranium enrichment program as a government corporation to allow it to compete more effectively in the domestic and international markets. Some of GAO's findings discussed are: uranium market experts believe and existing market models show that the proposed DOE purchase of a $750 million of uranium from domestic producers may not significantly increase production because of large producer-held inventories; excess uranium enrichment production capacity exists throughout the world; therefore, foreign producers are expected to compete heavily in the United States throughout the 1990s as utilities' contracts with DOE expire; and according to a 1988 agreement between DOE's Offices of Nuclear Energy and Defense Programs, enrichment decommissioning costs, estimated to total $3.6 billion for planning purposes, will be shared by the commercial enrichment program and the government

  10. Uranium resources

    International Nuclear Information System (INIS)

    1976-01-01

    This is a press release issued by the OECD on 9th March 1976. It is stated that the steep increases in demand for uranium foreseen in and beyond the 1980's, with doubling times of the order of six to seven years, will inevitably create formidable problems for the industry. Further substantial efforts will be needed in prospecting for new uranium reserves. Information is given in tabular or graphical form on the following: reasonably assured resources, country by country; uranium production capacities, country by country; world nuclear power growth; world annual uranium requirements; world annual separative requirements; world annual light water reactor fuel reprocessing requirements; distribution of reactor types (LWR, SGHWR, AGR, HWR, HJR, GG, FBR); and world fuel cycle capital requirements. The information is based on the latest report on Uranium Resources Production and Demand, jointly issued by the OECD's Nuclear Energy Agency (NEA) and the International Atomic Energy Agency. (U.K.)

  11. Effect of nano size 3% wt TaC particles dispersion in two different metallic matrix composites

    International Nuclear Information System (INIS)

    Gomes, U.U.; Oliveira, L.A.; Souza, C.P.; Menezes, R.C.; Furukava, M.; Torres, Y.

    2009-01-01

    This work studies the characteristics of two different metallic matrixes composites, ferritic and austenitic steels, reinforced with 3% wt nano size tantalum carbide by powder metallurgy. The starting powders were characterized by X-ray diffraction (XRD) and scanning electron microscopy (SEM). The effects of the nano sized carbide dispersion on the matrix microstructures and its consequences on the mechanical properties were identified. The preliminary results showed that the sintering were influenced by morphology and the distribution of carbide and the alloys. (author)

  12. The carbide M7C3 in low-temperature-carburized austenitic stainless steel

    International Nuclear Information System (INIS)

    Ernst, Frank; Li, Dingqiang; Kahn, Harold; Michal, Gary M.; Heuer, Arthur H.

    2011-01-01

    Prolonged low-temperature gas-phase carburization of AISI 316L-type austenitic stainless steel can cause intragranular precipitation of the carbide M 7 C 3 (M: randomly dispersed Fe, Cr, Ni). Transmission electron microscopy revealed that the carbide particles have the shape of needles. They grow by a ledge-migration mechanism and in a crystallographic orientation relationship to the austenite matrix that enables highly coherent interphase interfaces. A small solubility limit of Ni in the carbide and restricted Ni diffusivity at the processing temperature leads to Ni pileup around the particles and may explain the extreme aspect ratio of the particle shape. These characteristics closely resemble what has been observed earlier for precipitates of M 5 C 2 under slightly different processing conditions and can be rationalized by considering the particular constraints imposed by carburization at low temperature.

  13. Depleted uranium hexafluoride: Waste or resource?

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  14. Thermal and electrochemical stability of tungsten carbide catalyst supports

    Energy Technology Data Exchange (ETDEWEB)

    Chhina, H. [Ballard Power Systems, 9000 Glenlyon Parkway, Burnaby, BC (Canada); Department of Materials Engineering, University of British Columbia, Vancouver, BC (Canada); Campbell, S. [Ballard Power Systems, 9000 Glenlyon Parkway, Burnaby, BC (Canada); Kesler, O. [Department of Mechanical Engineering, University of British Columbia, Vancouver, BC (Canada)

    2007-02-10

    The thermal and electrochemical stability of tungsten carbide (WC), with and without a catalyst dispersed on it, have been investigated to evaluate the potential suitability of the material as an oxidation-resistant catalyst support. Standard techniques currently used to disperse Pt on carbon could not be used to disperse Pt on WC, so an alternative method was developed and used to disperse Pt on both commercially available WC and on carbon for comparison of stability. Electrochemical testing was performed by applying oxidation cycles between +0.6 V and +1.8 V to the support-catalyst material combinations and monitoring the activity of the supported catalyst over 100 oxidation cycles. Comparisons of activity change with cumulative oxidation cycles were made between C and WC supports with comparable loadings of catalyst by weight, solid volume, and powder volume. WC was found to be more thermally and electrochemically stable than currently used carbon support material Vulcan XC-72R. However, further optimization of the particle sizes and dispersion of Pt/WC catalyst/support materials and of comparison standards between new candidate materials and existing carbon-based supports are required. (author)

  15. Muonium states in silicon carbide

    International Nuclear Information System (INIS)

    Patterson, B.D.; Baumeler, H.; Keller, H.; Kiefl, R.F.; Kuendig, W.; Odermatt, W.; Schneider, J.W.; Estle, T.L.; Spencer, D.P.; Savic, I.M.

    1986-01-01

    Implanted muons in samples of silicon carbide have been observed to form paramagnetic muonium centers (μ + e - ). Muonium precession signals in low applied magnetic fields have been observed at 22 K in a granular sample of cubic β-SiC, however it was not possible to determine the hyperfine frequency. In a signal crystal sample of hexagonal 6H-SiC, three apparently isotropic muonium states were observed at 20 K and two at 300 K, all with hyperfine frequencies intermediate between those of the isotropic muonium centers in diamond and silicon. No evidence was seen of an anisotropic muonium state analogous to the Mu * state in diamond and silicon. (orig.)

  16. Uranium supply and demand

    Energy Technology Data Exchange (ETDEWEB)

    Spriggs, M J

    1976-01-01

    Papers were presented on the pattern of uranium production in South Africa; Australian uranium--will it ever become available; North American uranium resources, policies, prospects, and pricing; economic and political environment of the uranium mining industry; alternative sources of uranium supply; whither North American demand for uranium; and uranium demand and security of supply--a consumer's point of view. (LK)

  17. Low temperature study of nonstoichiometric titanium carbide

    International Nuclear Information System (INIS)

    Tashmetov, M.Yu.

    2005-05-01

    By low temperature neutron diffraction method was studied structure in nonstoichiometric titanium carbide from room temperature up to 12K. It is found of low temperature phase in titanium carbide- TiC 0.71 . It is established region and borders of this phase. It is determined change of unit cell parameter. (author)

  18. Elastic modulus and fracture of boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Walther, G.

    1978-12-01

    The elastic modulus of hot-pressed boron carbide with 1 to 15% porosity was measured at room temperature. K/sub IC/ values were determined for the same porosity range at 500 0 C by the double torsion technique. The critical stress intensity factor of boron carbide with 8% porosity was evaluated from 25 to 1200 0 C

  19. Ligand sphere conversions in terminal carbide complexes

    DEFF Research Database (Denmark)

    Morsing, Thorbjørn Juul; Reinholdt, Anders; Sauer, Stephan P. A.

    2016-01-01

    Metathesis is introduced as a preparative route to terminal carbide complexes. The chloride ligands of the terminal carbide complex [RuC(Cl)2(PCy3)2] (RuC) can be exchanged, paving the way for a systematic variation of the ligand sphere. A series of substituted complexes, including the first...... example of a cationic terminal carbide complex, [RuC(Cl)(CH3CN)(PCy3)2]+, is described and characterized by NMR, MS, X-ray crystallography, and computational studies. The experimentally observed irregular variation of the carbide 13C chemical shift is shown to be accurately reproduced by DFT, which also...... demonstrates that details of the coordination geometry affect the carbide chemical shift equally as much as variations in the nature of the auxiliary ligands. Furthermore, the kinetics of formation of the sqaure pyramidal dicyano complex, trans-[RuC(CN)2(PCy3)2], from RuC has been examined and the reaction...

  20. Microsegregation in Nodular Cast Iron with Carbides

    Directory of Open Access Journals (Sweden)

    S. Pietrowski

    2012-12-01

    Full Text Available In this paper results of microsegregation in the newly developed nodular cast iron with carbides are presented. To investigate the pearlitic and bainitic cast iron with carbides obtained by Inmold method were chosen. The distribution of linear elements on the eutectic cell radius was examined. To investigate the microsegregation pearlitic and bainitic cast iron with carbides obtained by Inmold method were chosen.The linear distribution of elements on the eutectic cell radius was examined. Testing of the chemical composition of cast iron metal matrix components, including carbides were carried out. The change of graphitizing and anti-graphitizing element concentrations within eutectic cell was determined. It was found, that in cast iron containing Mo carbides crystallizing after austenite + graphite eutectic are Si enriched.

  1. Microsegregation in Nodular Cast Iron with Carbides

    Directory of Open Access Journals (Sweden)

    Pietrowski S.

    2012-12-01

    Full Text Available In this paper results of microsegregation in the newly developed nodular cast iron with carbides are presented. To investigate the pearlitic and bainitic cast iron with carbides obtained by Inmold method were chosen. The distribution of linear elements on the eutectic cell radius was examined. To investigate the microsegregation pearlitic and bainitic cast iron with carbides obtained by Inmold method were chosen. The linear distribution of elements on the eutectic cell radius was examined. Testing of the chemical composition of cast iron metal matrix components, including carbides were carried out. The change of graphitizing and anti-graphitizing element concentrations within eutectic cell was determined. It was found, that in cast iron containing Mo carbides crystallizing after austenite + graphite eutectic are Si enriched.

  2. Mössbauer study of iron carbide nanoparticles produced by laser ablation in alcohols

    Energy Technology Data Exchange (ETDEWEB)

    Amagasa, S., E-mail: B115608@ed.tus.ac.jp; Nishida, N. [Tokyo University of Science, Department of Chemistry (Japan); Kobayashi, Y. [The University of Electro-Communications, Graduate School of Informatics and Engineering (Japan); Yamada, Y. [Tokyo University of Science, Department of Chemistry (Japan)

    2016-12-15

    Iron carbide nanoparticles were synthesized by laser ablation of iron in alcohols (methanol and ethanol). A new cell, designed to allow the ablation to be conducted in a flowing solvent, enabled separation and collection of the nanoparticles immediately after production, thus preventing further photochemical reactions of the colloids. The nanoparticles were investigated using Mössbauer spectroscopy, X-ray diffraction, and transmission electron microscopy. In methanol, they consisted of α-iron, γ-iron, iron carbide, and amorphous paramagnetic iron carbides, whereas in ethanol they consisted of iron carbides and amorphous paramagnetic iron carbides. The difference in products depending on the alcohol was attributed to the different carbon supplies for methanol and ethanol. For both solvents, the average particle size was found to be 16 nm, and the nanoparticles were dispersed in amorphous carbon. We also examined the effect of further laser irradiation of the colloids using stagnant solvent, and the particle size was found to increase and a very small amount of carbonization was observed.

  3. Effect of composition and heat treatment on carbide phases in Ni-Mo alloys

    International Nuclear Information System (INIS)

    Svistunova, T.V.; Tsvigunov, A.N.; Stegnukhina, L.V.; Sakuta, N.D.

    1984-01-01

    The investigation results of vanadium, iron, carbon and silicon effect and heat treatment regime on the type and composition of carbides in Ni-(26...31)%Mo alloys are presented. It is shown that type, composition and quantity of carbide phases forming in alloys are determined not only by molybdenum and carbon content, but presence of other elements (V, Fe), admixtures (C, Si) and reducers as well as by regime of thermal treatment. In the alloy, containing 26...31% Mo, 0.01...0.03% C ( 12 C type with a=1.083...1.089 nm lattice parameter, in which V and Ti, Fe and Si are presented besides Mo and Ni. In the temperature range of 600-800 deg C high dispersed carbides segregate on grain boundaries. Silicon initiates segregation of the carbide phases among them by grain boundaries at the temperatures of 800 deg C as well as regulates carbide of M 12 C type with a=1.094...1.098 nm lattice parameter

  4. Uranium briquettes for irradiation target

    Energy Technology Data Exchange (ETDEWEB)

    Saliba-Silva, Adonis Marcelo; Garcia, Rafael Henrique Lazzari; Martins, Ilson Carlos; Carvalho, Elita Fontenele Urano de; Durazzo, Michelangelo, E-mail: saliba@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Direct irradiation on targets inside nuclear research or multiple purpose reactors is a common route to produce {sup 99}Mo-{sup 99m}Tc radioisotopes. Nevertheless, since the imposed limits to use LEU uranium to prevent nuclear armament production, the amount of uranium loaded in target meats has physically increased and new processes have been proposed for production. Routes using metallic uranium thin film and UAl{sub x} dispersion have been used for this purpose. Both routes have their own issues, either by bringing difficulties to disassemble the aluminum case inside hot cells or by generating great amount of alkaline radioactive liquid rejects. A potential route might be the dispersion of powders of LEU metallic uranium and nickel, which are pressed as a blend inside a die and followed by pulse electroplating of nickel. The electroplating provides more strength to the briquettes and creates a barrier for gas evolution during neutronic disintegration of {sup 235}U. A target briquette platted with nickel encapsulated in an aluminum case to be irradiated may be an alternative possibility to replace other proposed targets. This work uses pulse Ni-electroplating over iron powder briquette to simulate the covering of uranium by nickel. The following parameters were applied 10 times for each sample: 900Hz, -0.84A/square centimeters with duty cycle of 0.1 in Watts Bath. It also presented the optical microscopy analysis of plated microstructure section. (author)

  5. Uranium briquettes for irradiation target

    International Nuclear Information System (INIS)

    Saliba-Silva, Adonis Marcelo; Garcia, Rafael Henrique Lazzari; Martins, Ilson Carlos; Carvalho, Elita Fontenele Urano de; Durazzo, Michelangelo

    2011-01-01

    Direct irradiation on targets inside nuclear research or multiple purpose reactors is a common route to produce 99 Mo- 99m Tc radioisotopes. Nevertheless, since the imposed limits to use LEU uranium to prevent nuclear armament production, the amount of uranium loaded in target meats has physically increased and new processes have been proposed for production. Routes using metallic uranium thin film and UAl x dispersion have been used for this purpose. Both routes have their own issues, either by bringing difficulties to disassemble the aluminum case inside hot cells or by generating great amount of alkaline radioactive liquid rejects. A potential route might be the dispersion of powders of LEU metallic uranium and nickel, which are pressed as a blend inside a die and followed by pulse electroplating of nickel. The electroplating provides more strength to the briquettes and creates a barrier for gas evolution during neutronic disintegration of 235 U. A target briquette platted with nickel encapsulated in an aluminum case to be irradiated may be an alternative possibility to replace other proposed targets. This work uses pulse Ni-electroplating over iron powder briquette to simulate the covering of uranium by nickel. The following parameters were applied 10 times for each sample: 900Hz, -0.84A/square centimeters with duty cycle of 0.1 in Watts Bath. It also presented the optical microscopy analysis of plated microstructure section. (author)

  6. Uranium, depleted uranium, biological effects; Uranium, uranium appauvri, effets biologiques

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  7. Uranium toxicology

    International Nuclear Information System (INIS)

    Ferreyra, Mariana D.; Suarez Mendez, Sebastian

    1997-01-01

    In this paper are presented the methods and procedures optimized by the Nuclear Regulatory Authority (ARN) for the determination of: natural uranium mass, activity of enriched uranium in samples of: urine, mucus, filters, filter heads, rinsing waters and Pu in urine, adopted and in some cases adapted, by the Environmental Monitoring and Internal Dosimetry Laboratory. The analyzed material corresponded to biological and environmental samples belonging to the staff professionally exposed that work in plants of the nuclear fuel cycle. For a better comprehension of the activities of this laboratory, it is included a brief description of the uranium radiochemical toxicity and the limits internationally fixed to preserve the workers health

  8. Chemical and phase composition of powders obtained by electroerosion dispersion from alloys WC-Co

    International Nuclear Information System (INIS)

    Putintseva, M.N.

    2004-01-01

    A consideration is given to the dependence of chemical and phase compositions of dispersed powders on the conditions, the medium of electroerosion dispersing and the content of cobalt in an initial alloy. It is shown that dissociation of carbon from tungsten carbide proceeds even on dispersing in liquid hydrocarbon-containing media (kerosene and machine oil). The phase composition is determined to a large extent by a medium of dispersing and a cobalt content in the initial alloy. In all powders complex tungsten-cobalt carbides and even Co 7 W 6 intermetallic compounds are found [ru

  9. Chemical and Phase Composition of Powders Obtained by Electroerosion Dispersion from WC - Co Alloys

    Science.gov (United States)

    Putintseva, M. N.

    2004-03-01

    The dependence of the chemical and phase composition of dispersed powders on the mode and medium of electroerosion dispersion and the content of cobalt in the initial alloy is considered. It is shown that the dissociation of carbon from tungsten carbide occurs even in dispersion in liquid hydrocarbon-bearing media (kerosene and industrial oils). The phase composition is primarily determined by the dispersion medium and the content of cobalt in the initial alloy. Compound tungsten-cobalt carbides and even a Co7W6 intermetallic are determined in all the powders.

  10. Rossing uranium

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    In this article the geology of the deposits of the Rossing uranium mine in Namibia is discussed. The planning of the open-pit mining, the blasting, drilling, handling and the equipment used for these processes are described

  11. Uranium recovering from slags generated in the metallic uranium by magnesiothermic reduction

    International Nuclear Information System (INIS)

    Fornarolo, F.; Carvalho, E.F. Urano de; Durazzo, M.; Riella, H.G.

    2008-01-01

    The Nuclear Fuel Center of IPEN/CNEN-SP has recent/y concluded a program for developing the fabrication technology of the nuclear fuel based on the U 3 Si 2 -Al dispersion, which is being used in the IEA-R1 research reactor. The uranium silicide (U 3 Si 2 ) fuel production starts with the uranium hexafluoride (UF 6 ) processing and uranium tetrafluoride (UF 4 ) precipitation. Then, the UF 4 is converted to metallic uranium by magnesiothermic reduction. The UF 4 reduction by magnesium generates MgF 2 slag containing considerable concentrations of uranium, which could reach 20 wt%. The uranium contained in that slag should be recovered and this work presents the results obtained in recovering the uranium from that slag. The uranium recovery is accomplished by acidic leaching of the calcined slag. The calcination transforms the metallic uranium in U 3 O 8 , promoting the pulverization of the pieces of metallic uranium and facilitating the leaching operation. As process variables, have been considered the nitric molar concentration, the acid excess regarding the stoichiometry and the leaching temperature. As result, the uranium recovery reached a 96% yield. (author)

  12. Uranium, depleted uranium, biological effects

    International Nuclear Information System (INIS)

    2001-01-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  13. Uranium loans

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    When NUEXCO was organized in 1968, its founders conceived of a business based on uranium loans. The concept was relatively straightforward; those who found themselves with excess supplies of uranium would deposit those excesses in NUEXCO's open-quotes bank,close quotes and those who found themselves temporarily short of uranium could borrow from the bank. The borrower would pay interest based on the quantity of uranium borrowed and the duration of the loan, and the bank would collect the interest, deduct its service fee for arranging the loan, and pay the balance to those whose deposits were borrowed. In fact, the original plan was to call the firm Nuclear Bank Corporation, until it was discovered that using the word open-quotes Bankclose quotes in the name would subject the firm to various US banking regulations. Thus, Nuclear Bank Corporation became Nuclear Exchange Corporation, which was later shortened to NUEXCO. Neither the nuclear fuel market nor NUEXCO's business developed quite as its founders had anticipated. From almost the very beginning, the brokerage of uranium purchases and sales became a more significant activity for NUEXCO than arranging uranium loans. Nevertheless, loan transactions have played an important role in the international nuclear fuel market, requiring the development of special knowledge and commercial techniques

  14. Steady- and transient-state analysis of fully ceramic microencapsulated fuel with randomly dispersed tristructural isotropic particles via two-temperature homogenized model-II: Applications by coupling with COREDAX

    International Nuclear Information System (INIS)

    Lee, Yoon Hee; Cho, Bum Hee; Cho, Nam Zin

    2016-01-01

    In Part I of this paper, the two-temperature homogenized model for the fully ceramic microencapsulated fuel, in which tristructural isotropic particles are randomly dispersed in a fine lattice stochastic structure, was discussed. In this model, the fuel-kernel and silicon carbide matrix temperatures are distinguished. Moreover, the obtained temperature profiles are more realistic than those obtained using other models. Using the temperature-dependent thermal conductivities of uranium nitride and the silicon carbide matrix, temperature-dependent homogenized parameters were obtained. In Part II of the paper, coupled with the COREDAX code, a reactor core loaded by fully ceramic microencapsulated fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure is analyzed via a two-temperature homogenized model at steady and transient states. The results are compared with those from harmonic- and volumetric-average thermal conductivity models; i.e., we compare keff eigenvalues, power distributions, and temperature profiles in the hottest single channel at a steady state. At transient states, we compare total power, average energy deposition, and maximum temperatures in the hottest single channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized model for Doppler temperature feedback lead to significant differences

  15. Uranium extraction from gold-uranium ores

    Energy Technology Data Exchange (ETDEWEB)

    Laskorin, B.N.; Golynko, Z.Sh.

    1981-01-01

    The process of uranium extraction from gold-uranium ores in the South Africa is considered. Flowsheets of reprocessing gold-uranium conglomerates, pile processing and uranium extraction from the ores are presented. Continuous counter flow ion-exchange process of uranium extraction using strong-active or weak-active resins is noted to be the most perspective and economical one. The ion-exchange uranium separation with the succeeding extraction is also the perspective one.

  16. X-ray photoelectron spectroscopy study of CO2 reaction with polycrystalline uranium surface

    International Nuclear Information System (INIS)

    Liu Kezhao; Yu Yong; Zhou Juesheng; Wu Sheng; Wang Xiaolin; Fu Yibei

    1999-10-01

    The adsorption of CO 2 on 'clean' depleted polycrystalline uranium metal surface has been studied by X-ray photoelectron spectroscopy (XPS) at 300 K. The 'clean' surface were prepared by Ar + ion sputtering under ultra-high vacuum (UHV) condition with a base pressure 6.7 x 10 -8 Pa. The result s shows that adsorption of CO 2 on 'clean' uranium metal took place in total dissociation, and leads to the formation of uranium dioxide, uranium carbides and free carbon. The total dissociation of CO 2 produced carbon, oxygen species, CO 2 2- and CO 3 2- species. The diffusion tendency of carbon was much stronger than that of oxygen, and led to form a carbide in oxide-metal interface while the oxygen remained on their surface as an oxide

  17. Uranium mining

    International Nuclear Information System (INIS)

    2008-01-01

    Full text: The economic and environmental sustainability of uranium mining has been analysed by Monash University researcher Dr Gavin Mudd in a paper that challenges the perception that uranium mining is an 'infinite quality source' that provides solutions to the world's demand for energy. Dr Mudd says information on the uranium industry touted by politicians and mining companies is not necessarily inaccurate, but it does not tell the whole story, being often just an average snapshot of the costs of uranium mining today without reflecting the escalating costs associated with the process in years to come. 'From a sustainability perspective, it is critical to evaluate accurately the true lifecycle costs of all forms of electricity production, especially with respect to greenhouse emissions, ' he says. 'For nuclear power, a significant proportion of greenhouse emissions are derived from the fuel supply, including uranium mining, milling, enrichment and fuel manufacture.' Dr Mudd found that financial and environmental costs escalate dramatically as the uranium ore is used. The deeper the mining process required to extract the ore, the higher the cost for mining companies, the greater the impact on the environment and the more resources needed to obtain the product. I t is clear that there is a strong sensitivity of energy and water consumption and greenhouse emissions to ore grade, and that ore grades are likely to continue to decline gradually in the medium to long term. These issues are critical to the current debate over nuclear power and greenhouse emissions, especially with respect to ascribing sustainability to such activities as uranium mining and milling. For example, mining at Roxby Downs is responsible for the emission of over one million tonnes of greenhouse gases per year and this could increase to four million tonnes if the mine is expanded.'

  18. Uranium and the use of depleted uranium in weaponry; L'uranium et les armes a l'uranium appauvri

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, R

    2000-07-01

    In this brief report the author shows that the use of shells involving a load of depleted uranium might lead to lasting hazards to civil population and environment. These hazards come from the part of the shell that has been dispersed as contaminating radioactive dusts. The author describes some features of radioactivity and highlights the role of Uranium-238 as a provider of energy to the planet. (A.C.)

  19. Vanadium carbide coatings: deposition process and properties

    International Nuclear Information System (INIS)

    Borisova, A.; Borisov, Y.; Shavlovsky, E.; Mits, I.; Castermans, L.; Jongbloed, R.

    2001-01-01

    Vanadium carbide coatings on carbon and alloyed steels were produced by the method of diffusion saturation from the borax melt. Thickness of the vanadium carbide layer was 5-15 μm, depending upon the steel grade and diffusion saturation parameters. Microhardness was 20000-28000 MPa and wear resistance of the coatings under conditions of end face friction without lubrication against a mating body of WC-2Co was 15-20 times as high as that of boride coatings. Vanadium carbide coatings can operate in air at a temperature of up to 400 o C. They improve fatigue strength of carbon steels and decrease the rate of corrosion in sea and fresh water and in acid solutions. The use of vanadium carbide coatings for hardening of various types of tools, including cutting tools, allows their service life to be extended by a factor of 3 to 30. (author)

  20. Uranium enrichment

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.

    1988-06-01

    Canada is the world's largest producer and exporter of uranium, most of which is enriched elsewhere for use as fuel in LWRs. The feasibility of a Canadian uranium-enrichment enterprise is therefore a perennial question. Recent developments in uranium-enrichment technology, and their likely impacts on separative work supply and demand, suggest an opportunity window for Canadian entry into this international market. The Canadian opportunity results from three particular impacts of the new technologies: 1) the bulk of the world's uranium-enrichment capacity is in gaseous diffusion plants which, because of their large requirements for electricity (more than 2000 kW·h per SWU), are vulnerable to competition from the new processes; 2) the decline in enrichment costs increases the economic incentive for the use of slightly-enriched uranium (SEU) fuel in CANDU reactors, thus creating a potential Canadian market; and 3) the new processes allow economic operation on a much smaller scale, which drastically reduces the investment required for market entry and is comparable with the potential Canadian SEU requirement. The opportunity is not open-ended. By the end of the century the enrichment supply industry will have adapted to the new processes and long-term customer/supplier relationships will have been established. In order to seize the opportunity, Canada must become a credible supplier during this century

  1. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    International Nuclear Information System (INIS)

    Radulescu, H.

    2001-01-01

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

  2. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    Energy Technology Data Exchange (ETDEWEB)

    H. radulescu

    2001-09-28

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

  3. Geochemical prospecting for thorium and uranium deposits

    International Nuclear Information System (INIS)

    Boyle, R.W.

    1982-01-01

    The basic purpose of this book is to present an analysis of the various geochemical methods applicable in the search for all types of thorium and uranium deposits. The general chemistry and geochemistry of thorium and uranium are briefly described in the opening chapter, and this is followed by a chapter on the deposits of the two elements with emphasis on their indicator (pathfinder) elements and on the primary and secondary dispersion characteristics of thorium and uranium in the vicinity of their deposits. The next seven chapters form the main part of the book and describe geochemical prospecting for thorium and uranium, stressing selection of areas in which to prospect, radiometric surveys, analytical geochemical surveys based on rocks (lithochemical surveys), unconsolidated materials (pedochemical surveys), natural waters and sediments (hydrochemical surveys), biological materials (biogeochemical surveys), gases (atmochemical surveys), and miscellaneous methods. A final brief chapter reviews radiometric and analytical methods for the detection and estimation of thorium and uranium. (Auth.)

  4. Stable carbides in transition metal alloys

    International Nuclear Information System (INIS)

    Piotrkowski, R.

    1991-01-01

    In the present work different techniques were employed for the identification of stable carbides in two sets of transition metal alloys of wide technological application: a set of three high alloy M2 type steels in which W and/or Mo were total or partially replaced by Nb, and a Zr-2.5 Nb alloy. The M2 steel is a high speed steel worldwide used and the Zr-2.5 Nb alloy is the base material for the pressure tubes in the CANDU type nuclear reactors. The stability of carbide was studied in the frame of Goldschmidt's theory of interstitial alloys. The identification of stable carbides in steels was performed by determining their metallic composition with an energy analyzer attached to the scanning electron microscope (SEM). By these means typical carbides of the M2 steel, MC and M 6 C, were found. Moreover, the spatial and size distribution of carbide particles were determined after different heat treatments, and both microstructure and microhardness were correlated with the appearance of the secondary hardening phenomenon. In the Zr-Nb alloy a study of the α and β phases present after different heat treatments was performed with optical and SEM metallographic techniques, with the guide of Abriata and Bolcich phase diagram. The α-β interphase boundaries were characterized as short circuits for diffusion with radiotracer techniques and applying Fisher-Bondy-Martin model. The precipitation of carbides was promoted by heat treatments that produced first the C diffusion into the samples at high temperatures (β phase), and then the precipitation of carbide particles at lower temperature (α phase or (α+β)) two phase field. The precipitated carbides were identified as (Zr, Nb)C 1-x with SEM, electron microprobe and X-ray diffraction techniques. (Author) [es

  5. Point defects and transport properties in carbides

    International Nuclear Information System (INIS)

    Matzke, Hj.

    1984-01-01

    Carbides of transition metals and of actinides are interesting and technologically important. The transition-metal carbides (or carbonitrides) are extensively being used as hard materials and some of them are of great interest because of the high transition temperature for superconductivity, e.g. 17 K for Nb(C,N). Actinide carbides and carbonitrides, (U,Pu)C and (U,Pu)(C,N) are being considered as promising advanced fuels for liquid metal cooled fast breeder nuclear reactors. Basic interest exists in all these materials because of their high melting points (e.g. 4250 K for TaC) and the unusually broad range of homogeneity of nonstoichiometric compositions (e.g. from UCsub(0.9) to UCsub(1.9) at 2500 K). Interaction of point defects to clusters and short-range ordering have recently been studied with elastic neutron diffraction and diffuse scattering techniques, and calculations of energies of formation and interaction of point defects became available for selected carbides. Diffusion measurements also exist for a number of carbides, in particular for the actinide carbides. The existing knowledge is discussed and summarized with emphasis on informative examples of particular technological relevance. (Auth.)

  6. Controlled formation of iron carbides and their performance in Fischer-Tropsch synthesis

    KAUST Repository

    Wezendonk, Tim A.

    2018-04-19

    Iron carbides are unmistakably associated with the active phase for Fischer-Tropsch synthesis (FTS). The formation of these carbides is highly dependent on the catalyst formulation, the activation method and the operational conditions. Because of this highly dynamic behavior, studies on active phase performance often lack the direct correlation between catalyst performance and iron carbide phase. For the above reasons, an extensive in situ Mössbauer spectroscopy study on highly dispersed Fe on carbon catalysts (Fe@C) produced through pyrolysis of a Metal Organic Framework was coupled to their FTS performance testing. The preparation of Fe@C catalysts via this MOF mediated synthesis allows control over the active phase formation and therefore provides an ideal model system to study the performance of different iron carbides. Reduction of fresh Fe@C followed by low-temperature Fischer-Tropsch (LTFT) conditions resulted in the formation of the ε′-Fe2.2C, whereas carburization of the fresh catalysts under high-temperature Fischer-Tropsch (HTFT) resulted in the formation of χ-Fe5C2. Furthermore, the different activation methods did not alter other important catalyst properties, as pre- and post-reaction transmission electron microscopy (TEM) characterization confirmed that the iron nanoparticle dispersion was preserved. The weight normalized activities (FTY) of χ-Fe5C2 and ε′-Fe2.2C are virtually identical, whilst it is found that ε′-Fe2.2C is a better hydrogenation catalyst than χ-Fe5C2. The absence of differences under subsequent HTFT experiments, where χ-Fe5C2 is the dominating phase, is a strong indication that the iron carbide phase is responsible for the differences in selectivity.

  7. Uranium update

    International Nuclear Information System (INIS)

    Steane, R.

    1997-01-01

    This paper is about the current uranium mining situation, especially that in Saskatchewan. Canada has a unique advantage with the Saskatchewan uranium deposits. Making the most of this opportunity is important to Canada. The following is reviewed: project development and the time and capital it takes to bring a new project into production; the supply and demand situation to show where the future production fits into the world market; and our foreign competition and how we have to be careful not to lose our opportunity. (author)

  8. From nitrides to carbides: topotactic synthesis of the eta-carbides Fe3Mo3C and Co3Mo3C.

    Science.gov (United States)

    Alconchel, Silvia; Sapiña, Fernando; Martínez, Eduardo

    2004-08-21

    The molybdenum bimetallic interstitial carbides Fe(3)Mo(3)C and Co(3)Mo(3)C have been synthesized by temperature-programmed reaction (TPR) between the molybdenum bimetallic interstitial nitrides Fe(3)Mo(3)N and Co(3)Mo(3)N and a flowing mixture of CH(4) and H(2) diluted in Ar. These compounds have been characterized by X-ray diffraction, laser Raman spectroscopy, elemental analysis, energy dispersive analysis of X rays, thermal analysis (in air) and scanning electron microscopy (field emission). Their structures have been refined from X-ray powder diffraction data. These carbides crystallize in the cubic system, space group Fd3m[a= 11.11376(6) and 11.0697(3)[Angstrom] for Fe and Co compounds, respectively].

  9. The composition and character of oxycarbide phase in uranium metal

    International Nuclear Information System (INIS)

    Liu Kezhao; Lai Xinchun; Yu Yong; Ni Ranfu

    1999-08-01

    The oxide layer of uranium metal formed by vacuum heating were examined with X-ray photoelectron spectroscopy (XPS) and Auger Electron Spectroscopy (AES). XPS results indicated that the air-exposed surface of the oxide layer were mainly consisted of UO 2 and free carbon. After the air-exposed surface were removed by low energy argon ion sputtering, C1s spectra shifted from 284.8 eV to 281.8 eV, indicating the existence of carbide phase. AES results of C(KVV) Auger transitions confirmed this result. Resolved and fitted using a combination of Gaussian and Lorentzian peak shape, U4f 7/2 spectra showed that three uranium chemical states existed in the layer, there were uranium dioxide, uranium carbide (or oxycarbide, UC x O 1-x ) and uranium metal phase. Calculated the AES data by relatively sensitive factor, the composition of oxycarbide was given as UC 0.41+-0.04 O 0.62+-0.01

  10. Structure and properties of hot-pressed boron carbide ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Koval' chenko, M S; Tkachenko, IU G; Koval' chuk, V V; Iurchenko, D Z; Satanin, S V [Institut Problem Materialovedeniia, Kiev (Ukrainian SSR)

    1990-07-01

    The microstructure and strength of B4C-TiB2-TiO{sub 2} ceramics samples, hot-compacted from a mixture of two types of B4C-TiO2-C powder, are examined. The two types are obtained by combining boric acid with either sucrose or carbon black. The grain-sizes of the two powders are found to be distinctly different from one another both before and after the grinding procedure and the degree of dispersion is not high. The strength tests show 600 MPa, the Vicker's hardness is 34.5 GPa, and the crack resistance coefficient of ceramics containing 15 percent TiB2 by mass is 5 MPa m exp 1/2. The use of soluble boron carbide powder helps achieve higher levels of strength and crack resistance. 5 refs.

  11. Machining of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Morris, T.O.

    1981-01-01

    Uranium and uranium alloys can be readily machined by conventional methods in the standard machine shop when proper safety and operating techniques are used. Material properties that affect machining processes and recommended machining parameters are discussed. Safety procedures and precautions necessary in machining uranium and uranium alloys are also covered. 30 figures

  12. Debye temperatures of uranium chalcogenides from their lattice ...

    Indian Academy of Sciences (India)

    Phonon dispersion relations in uranium chalcogenides have been investigated using a modified three-body force shell model. From the phonon frequencies, their Debye temperatures are evaluated. Further, on the basis of the spin fluctuation in the heavy fermion uranium compounds, UPt3 and UBe13, the possible ...

  13. Plasma spraying of zirconium carbide – hafnium carbide – tungsten cermets

    Czech Academy of Sciences Publication Activity Database

    Brožek, Vlastimil; Ctibor, Pavel; Cheong, D.-I.; Yang, S.-H.

    2009-01-01

    Roč. 9, č. 1 (2009), s. 49-64 ISSN 1335-8987 Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma spraying * cermet coatings * microhardness * zirconium carbide * hafnium carbide * tungsten * water stabilized plasma Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass

  14. Hydrodynamic disperser

    Energy Technology Data Exchange (ETDEWEB)

    Bulatov, A.I.; Chernov, V.S.; Prokopov, L.I.; Proselkov, Yu.M.; Tikhonov, Yu.P.

    1980-01-15

    A hydrodynamic disperser is suggested which contains a housing, slit nozzles installed on a circular base arranged opposite from each other, resonators secured opposite the nozzle and outlet sleeve. In order to improve the effectiveness of dispersion by throttling the flow, each resonator is made in the form of a crimped plate with crimpings that decrease in height in a direction towards the nozzle.

  15. Tungsten carbide and tungsten-molybdenum carbides as automobile exhaust catalysts

    International Nuclear Information System (INIS)

    Leclercq, L.; Daubrege, F.; Gengembre, L.; Leclercq, G.; Prigent, M.

    1987-01-01

    Several catalyst samples of tungsten carbide and W, Mo mixed carbides with different Mo/W atom ratios, have been prepared to test their ability to remove carbon monoxide, nitric oxide and propane from a synthetic exhaust gas simulating automobile emissions. Surface characterization of the catalysts has been performed by X-ray photoelectron spectroscopy (XPS) and selective chemisorption of hydrogen and carbon monoxide. Tungsten carbide exhibits good activity for CO and NO conversion, compared to a standard three-way catalyst based on Pt and Rh. However, this W carbide is ineffective in the oxidation of propane. The Mo,W mixed carbides are markedly different having only a very low activity. 9 refs.; 10 figs.; 5 tabs

  16. High temperature evaporation of titanium, zirconium and hafnium carbides

    International Nuclear Information System (INIS)

    Gusev, A.I.; Rempel', A.A.

    1991-01-01

    Evaporation of cubic nonstoichiometric carbides of titanium, zirconium and hafnium in a comparatively low-temperature interval (1800-2700) with detailed crystallochemical sample certification is studied. Titanium carbide is characterized by the maximum evaporation rate: at T>2300 K it loses 3% of sample mass during an hour and at T>2400 K titanium carbide evaporation becomes extremely rapid. Zirconium and hafnium carbide evaporation rates are several times lower than titanium carbide evaporation rates at similar temperatures. Partial pressures of metals and carbon over the carbides studied are calculated on the base of evaporation rates

  17. Enhanced thermal conductivity of nano-SiC dispersed water based ...

    Indian Academy of Sciences (India)

    Silicon carbide (SiC) nanoparticle dispersed water based nanofluids were prepared using up to 0.1 vol% of nanoparticles. Use of suitable stirring routine ensured uniformity and stability of dispersion. Thermal conductivity ratio of nanofluid measured using transient hot wire device shows a significant increase of up to 12% ...

  18. Liquid membranes and process for uranium recovery therewith

    International Nuclear Information System (INIS)

    Frankenfeld, J.W.; Li, N.N.T.; Bruncati, R.L.

    1981-01-01

    A liquid membrane system consisting of water-in-oil type emulsions dispersed in water, which is capable of extracting uranium-containing ions from an aqueous feed solution containing uranium ions at a temperature in the range of 25 0 C to 80 0 C, is described. The emulsion comprises an aqueous interior phase surrounded by a surfactant-containing exterior phase. The exterior phase is immiscible with the interior phase and comprises a transfer agent capable of transporting selectively the desired uranium-containing ions and a solvent for the transfer agent. The interior phase comprises a reactant capable of removing uranium-containing ions from the transfer agent and capable of changing the valency of the uranium in uranium-containing ions to a second valency state and converting the uranium-containing ions into a nonpermeable form. (U.K.)

  19. Uranium mining

    International Nuclear Information System (INIS)

    Cheeseman, E.W.

    1980-01-01

    The international uranium market appears to be currently over-supplied with a resultant softening in prices. Buyers on the international market are unhappy about some of the restrictions placed on sales by the government, and Canadian sales may suffer as a result. About 64 percent of Canada's shipments come from five operating Ontario mines, with the balance from Saskatchewan. Several other properties will be producing within the next few years. In spite of the adverse effects of the Three Mile Island incident and the default by the T.V.A. of their contract, some 3 600 tonnes of new uranium sales were completed during the year. The price for uranium had stabilized at US $42 - $44 by mid 1979, but by early 1980 had softened somewhat. The year 1979 saw the completion of major environmental hearings in Ontario and Newfoundland and the start of the B.C. inquiry. Two more hearings are scheduled for Saskatchewan in 1980. The Elliot Lake uranium mining expansion hearings are reviewed, as are other recent hearings. In the production of uranium for nuclear fuel cycle, environmental matters are of major concern to the industry, the public and to governments. Research is being conducted to determine the most effective method for removing radium from tailings area effluents. Very stringent criteria are being drawn up by the regulatory agencies that must be met by the industry in order to obtain an operating licence from the AECB. These criteria cover seepages from the tailings basin and through the tailings retention dam, seismic stability, and both short and long term management of the tailings waste management area. (auth)

  20. Uranium industry annual 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

  1. Uranium industry annual 1996

    International Nuclear Information System (INIS)

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs

  2. Remedial action and site design for stabilization of the inactive uranium mill tailings sites at Slick Rock, Colorado

    International Nuclear Information System (INIS)

    1993-07-01

    The US Environmental Protection Agency (EPA) has established health and environmental protection regulations to correct and prevent groundwater contamination resulting from processing activities at inactive uranium milling sites (EPA, 1987). According to the Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978 Public Law (PL) 95-604 (PL 95-604), the US Department of Energy (DOE) is responsible for assessing the inactive uranium processing sites. The DOE has determined that for Slick Rock, this assessment shall include hydrogeologic site characterization for two separate uranium processing sites, the Union Carbide (UC) site and the North Continent (NC) site, and for the proposed Burro Canyon disposal site

  3. Uranium industry annual, 1991

    International Nuclear Information System (INIS)

    1992-10-01

    In the Uranium Industry Annual 1991, data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2. A feature article entitled ''The Uranium Industry of the Commonwealth of Independent States'' is included in this report

  4. Joining of boron carbide using nickel interlayer

    International Nuclear Information System (INIS)

    Vosughi, A.; Hadian, A. M.

    2008-01-01

    Carbide ceramics such as boron carbide due to their unique properties such as low density, high refractoriness, and high strength to weight ratio have many applications in different industries. This study focuses on direct bonding of boron carbide for high temperature applications using nickel interlayer. The process variables such as bonding time, temperature, and pressure have been investigated. The microstructure of the joint area was studied using electron scanning microscope technique. At all the bonding temperatures ranging from 1150 to 1300 d eg C a reaction layer formed across the ceramic/metal interface. The thickness of the reaction layer increased by increasing temperature. The strength of the bonded samples was measured using shear testing method. The highest strength value obtained was about 100 MPa and belonged to the samples bonded at 1250 for 75 min bonding time. The strength of the joints decreased by increasing the bonding temperature above 1250 d eg C . The results of this study showed that direct bonding technique along with nickel interlayer can be successfully utilized for bonding boron carbide ceramic to itself. This method may be used for bonding boron carbide to metals as well.

  5. Effect of carbide precipitates on high temperature creep of a 20Cr-25Ni austenitic stainless steel

    International Nuclear Information System (INIS)

    Yamane, T.; Takahashi, Y.; Nakagawa, K.

    1984-01-01

    The high temperature creep of an austenitic stainless steel having carbide precipitates, is different from that of the carbide precipitate-free one. Strain rates of the steady state creep d(epsilonsub(s))/dt, or minimum strain rates of the creep in precipitate hardened and dispersion strengthened alloys at the creep temperature T, can be expressed by Sherby-Dorn's equation d(epsilonsub(s))/dt = Aσsup(n) exp (-Qsub(c)/RT). The stress exponent n, and the activation energy for creep Qsub(c), in a power law creep region, are more than those of unstrengthened alloys, where σ is the creep stress, R the gas constant and A the constant. In this research, the influence of carbide precipitates on steady creep rates, is investigated. Experimental details are given. Results are given and discussed. (author)

  6. The carbide M{sub 7}C{sub 3} in low-temperature-carburized austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Ernst, Frank, E-mail: frank.ernst@cwru.edu [Department of Materials Science and Engineering, Case Western Reserve University, Cleveland, OH 44106-7204 (United States); Li, Dingqiang; Kahn, Harold; Michal, Gary M.; Heuer, Arthur H. [Department of Materials Science and Engineering, Case Western Reserve University, Cleveland, OH 44106-7204 (United States)

    2011-04-15

    Prolonged low-temperature gas-phase carburization of AISI 316L-type austenitic stainless steel can cause intragranular precipitation of the carbide M{sub 7}C{sub 3} (M: randomly dispersed Fe, Cr, Ni). Transmission electron microscopy revealed that the carbide particles have the shape of needles. They grow by a ledge-migration mechanism and in a crystallographic orientation relationship to the austenite matrix that enables highly coherent interphase interfaces. A small solubility limit of Ni in the carbide and restricted Ni diffusivity at the processing temperature leads to Ni pileup around the particles and may explain the extreme aspect ratio of the particle shape. These characteristics closely resemble what has been observed earlier for precipitates of M{sub 5}C{sub 2} under slightly different processing conditions and can be rationalized by considering the particular constraints imposed by carburization at low temperature.

  7. stabilization of ikpayongo laterite with cement and calcium carbide

    African Journals Online (AJOL)

    PROF EKWUEME

    Laterite obtained from Ikpayongo was stabilized with 2-10 % cement and 2-10 % Calcium Carbide waste, for use .... or open dumping which have effect on surface and ... Table 1: Chemical Composition of Calcium Carbide Waste and Cement.

  8. Method of fabricating porous silicon carbide (SiC)

    Science.gov (United States)

    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.

  9. Dispersion Forces

    CERN Document Server

    Buhmann, Stefan Yoshi

    2012-01-01

    In this book, a modern unified theory of dispersion forces on atoms and bodies is presented which covers a broad range of advanced aspects and scenarios. Macroscopic quantum electrodynamics is shown to provide a powerful framework for dispersion forces which allows for discussing general properties like their non-additivity and the relation between microscopic and macroscopic interactions. It is demonstrated how the general results can be used to obtain dispersion forces on atoms in the presence of bodies of various shapes and materials. Starting with a brief recapitulation of volume I, this volume II deals especially with bodies of irregular shapes, universal scaling laws, dynamical forces on excited atoms, enhanced forces in cavity quantum electrodynamics, non-equilibrium forces in thermal environments and quantum friction. The book gives both the specialist and those new to the field a thorough overview over recent results in the field. It provides a toolbox for studying dispersion forces in various contex...

  10. The diffusion bonding of silicon carbide and boron carbide using refractory metals

    International Nuclear Information System (INIS)

    Cockeram, B.V.

    1999-01-01

    Joining is an enabling technology for the application of structural ceramics at high temperatures. Metal foil diffusion bonding is a simple process for joining silicon carbide or boron carbide by solid-state, diffusive conversion of the metal foil into carbide and silicide compounds that produce bonding. Metal diffusion bonding trials were performed using thin foils (5 microm to 100 microm) of refractory metals (niobium, titanium, tungsten, and molybdenum) with plates of silicon carbide (both α-SiC and β-SiC) or boron carbide that were lapped flat prior to bonding. The influence of bonding temperature, bonding pressure, and foil thickness on bond quality was determined from metallographic inspection of the bonds. The microstructure and phases in the joint region of the diffusion bonds were evaluated using SEM, microprobe, and AES analysis. The use of molybdenum foil appeared to result in the highest quality bond of the metal foils evaluated for the diffusion bonding of silicon carbide and boron carbide. Bonding pressure appeared to have little influence on bond quality. The use of a thinner metal foil improved the bond quality. The microstructure of the bond region produced with either the α-SiC and β-SiC polytypes were similar

  11. Uranium - what role

    International Nuclear Information System (INIS)

    Grey, T.; Gaul, J.; Crooks, P.; Robotham, R.

    1980-01-01

    Opposing viewpoints on the future role of uranium are presented. Topics covered include the Australian Government's uranium policy, the status of nuclear power around the world, Australia's role as a uranium exporter and problems facing the nuclear industry

  12. Brazilian uranium exploration program

    International Nuclear Information System (INIS)

    Marques, J.P.M.

    1981-01-01

    General information on Brazilian Uranium Exploration Program, are presented. The mineralization processes of uranium depoits are described and the economic power of Brazil uranium reserves is evaluated. (M.C.K.) [pt

  13. Joining of porous silicon carbide bodies

    Science.gov (United States)

    Bates, Carl H.; Couhig, John T.; Pelletier, Paul J.

    1990-05-01

    A method of joining two porous bodies of silicon carbide is disclosed. It entails utilizing an aqueous slip of a similar silicon carbide as was used to form the porous bodies, including the sintering aids, and a binder to initially join the porous bodies together. Then the composite structure is subjected to cold isostatic pressing to form a joint having good handling strength. Then the composite structure is subjected to pressureless sintering to form the final strong bond. Optionally, after the sintering the structure is subjected to hot isostatic pressing to further improve the joint and densify the structure. The result is a composite structure in which the joint is almost indistinguishable from the silicon carbide pieces which it joins.

  14. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-11-01

    This paper analyzes under four different scenarios the adequacy of a $500 million annual deposit into a fund to pay for the cost of cleaning up the Department of Energy's (DOE) three aging uranium enrichment plants. These plants are located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. In summary the following was found: A fixed annual $500 million deposit made into a cleanup fund would not be adequate to cover total expected cleanup costs, nor would it be adequate to cover expected decontamination and decommissioning (D and D) costs. A $500 million annual deposit indexed to an inflation rate would likely be adequate to pay for all expected cleanup costs, including D and D costs, remedial action, and depleted uranium costs

  15. Uranium production

    International Nuclear Information System (INIS)

    Spriggs, M.

    1980-01-01

    The balance between uranium supply and demand is examined. Should new resources become necessary, some unconventional sources which could be considered include low-grade extensions to conventional deposits, certain types of intrusive rock, tuffs, and lake and sea-bed sediments. In addition there are large but very low grade deposits in carbonaceous shales, granites, and seawater. The possibility of recovery is discussed. Programmes of research into the feasibility of extraction of uranium from seawater, as a by-product from phosphoric acid production, and from copper leach solutions, are briefly discussed. Other possible sources are coal, old mine dumps and tailings, the latter being successfully exploited commercially in South Africa. The greatest constraints on increased development of U from lower grade sources are economics and environmental impact. It is concluded that apart from U as a by-product from phosphate, other sources are unlikely to contribute much to world requirements in the foreseeable future. (U.K.)

  16. Determination of free carbon content in boron carbide ceramic powders

    International Nuclear Information System (INIS)

    Castro, A.R.M. de; Lima, N.B. de; Paschoal, J.O.A.

    1990-01-01

    Boron carbide is a ceramic material of technological importance due to its hardness and high chemical and thermal stabilities. Free carbon is always found as a process dependent impurity in boron carbide. The development of procedures for its detection is required because its presence leads to a degradation of the boron carbide properties. In this work, several procedures for determining free carbon content in boron carbide specimens are reported and discussed for comparison purposes. (author) [pt

  17. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-08-01

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector

  18. Derived enriched uranium market

    International Nuclear Information System (INIS)

    Rutkowski, E.

    1996-01-01

    The potential impact on the uranium market of highly enriched uranium from nuclear weapons dismantling in the Russian Federation and the USA is analyzed. Uranium supply, conversion, and enrichment factors are outlined for each country; inventories are also listed. The enrichment component and conversion components are expected to cause little disruption to uranium markets. The uranium component of Russian derived enriched uranium hexafluoride is unresolved; US legislation places constraints on its introduction into the US market

  19. Effect of phosphorus addition on the hydrotreating activity of NiMo/Al{sub 2}O{sub 3} carbide catalyst

    Energy Technology Data Exchange (ETDEWEB)

    Sundaramurthy, V.; Dalai, A.K. [Catalysis and Chemical Reaction Engineering Laboratories, Department of Chemical Engineering, University of Saskatchewan, Saskatoon, SK S7N 5A9 (Canada); Adjaye, J. [Syncrude Edmonton Research Centre, Edmonton, AB T6N 1H4 (Canada)

    2007-07-30

    A series of phosphorus promoted {gamma}-Al{sub 2}O{sub 3} supported NiMo carbide catalysts with 0-4.5 wt.% P, 13 wt.% Mo and 2.5 wt.% Ni were synthesized and characterized by elemental analysis, pulsed CO chemisorption, BET surface area measurement, X-ray diffraction, near-edge X-ray absorption fine structure, DRIFT spectroscopy of CO adsorption and H{sub 2} temperature programmed reduction. X-ray diffraction patterns and CO uptake showed the P addition to NiMo/{gamma}-Al{sub 2}O{sub 3} carbide, increased the dispersion of {beta}-Mo{sub 2}C particles. DRIFT spectra of adsorbed CO revealed that P addition to NiMo/{gamma}-Al{sub 2}O{sub 3} carbide catalyst not only increases the dispersion of Ni-Mo carbide phase, but also changes the nature of surface active sites. The hydrodenitrogenation (HDN) and hydrodesulfurization (HDS) activities of these P promoted NiMo/{gamma}-Al{sub 2}O{sub 3} carbide catalysts were performed in trickle bed reactor using light gas oil (LGO) derived from Athabasca bitumen and model feed containing quinoline and dibenzothiophene at industrial conditions. The P added NiMo/{gamma}-Al{sub 2}O{sub 3} carbide catalysts showed enhanced HDN activity compared to the NiMo/{gamma}-Al{sub 2}O{sub 3} catalysts with both the feed stocks. The P had almost no influence on the HDS activity of NiMo/{gamma}-Al{sub 2}O{sub 3} carbide with LGO and dibenzothiophene. P addition to NiMo/{gamma}-Al{sub 2}O{sub 3} carbide accelerated C-N bond breaking and thus increased the HDN activity. (author)

  20. Silicon carbide microsystems for harsh environments

    CERN Document Server

    Wijesundara, Muthu B J

    2011-01-01

    Silicon Carbide Microsystems for Harsh Environments reviews state-of-the-art Silicon Carbide (SiC) technologies that, when combined, create microsystems capable of surviving in harsh environments, technological readiness of the system components, key issues when integrating these components into systems, and other hurdles in harsh environment operation. The authors use the SiC technology platform suite the model platform for developing harsh environment microsystems and then detail the current status of the specific individual technologies (electronics, MEMS, packaging). Additionally, methods

  1. Inhalation of uranium ores

    International Nuclear Information System (INIS)

    Stuart, B.O.; Jackson, P.O.

    1975-01-01

    In previous studies the biological dispositions of individual long-lived alpha members of the uranium chain ( 238 U, 234 U and 230 Th) were determined during and following repeated inhalation exposures of rats to pitchblende (26 percent U 3 O 8 ) ore. Although finely dispersed ore in secular equilibrium was inhaled, 230 Th/ 234 U radioactivity ratios in the lungs rose from 1.0 to 2.5 during 8 weeks of exposures and increased to 9.2 by four months after cessation of exposures. Marked non-equilibrium levels were also found in the tracheobronchial lymph nodes, kidneys, liver, and femur. Daily exposures of beagle dogs to high levels of this ore for 8 days resulted in lung 230 Th/ 234 U ratios of >2.0. Daily exposures of dogs to lower levels (0.1 mg/1) for 6 months, with sacrifice 15 months later, resulted in lung and thoracic lymph node 230 Th/ 234 U ratios ranging from 3.6 to 9 and nearly 7, respectively. The lungs of hamsters exposed to carnotite (4 percent U 3 O 8 ) ore in current lifespan studies show 230 Th/ 234 U ratios as high as 2.0 during daily inhalation of this ore in secular equilibrium. Beagle dogs sacrificed after several years of daily inhalations of the same carnotite ore plus radon daughters also showed marked non-equilibrium ratios of 230 Th/ 234 U, ranging from 5.6 to 7.4 in lungs and 6.2 to 9.1 in thoracic lymph nodes. This pattern of higher retention of 230 Th than 234 U in lungs, thoracic lymph nodes, and other tissues is thus consistent for two types of uranium ore among several species and suggests a reevaluation of maximum permissible air concentrations of ore, currently based only on uranium content

  2. Uranium industry annual, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Uranium industry data collected in the EIA-858 survey provide a comprehensive statistical characterization of annual activities of the industry and include some information about industry plans over the next several years. This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities utility market requirements and related topics

  3. Uranium Industry. Annual 1984

    International Nuclear Information System (INIS)

    Lawrence, M.S.S.

    1985-01-01

    This report provides a statistical description of activities of the US uranium industry during 1984 and includes a statistical profile of the status of the industry at the end of 1984. It is based on the results of an Energy Information Administration (EIA) survey entitled ''Uranium Industry Annual Survey'' (Form EIA-858). The principal findings of the survey are summarized under two headings - Uranium Raw Materials Activities and Uranium Marketing Activities. The first heading covers exploration and development, uranium resources, mine and mill production, and employment. The second heading covers uranium deliveries and delivery commitments, uranium prices, foreign trade in uranium, inventories, and other marketing activities. 32 figs., 48 tabs

  4. Microstructural Characterization of Reaction-Formed Silicon Carbide Ceramics. Materials Characterization

    Science.gov (United States)

    Singh, M.; Leonhardt, T. A.

    1995-01-01

    Microstructural characterization of two reaction-formed silicon carbide ceramics has been carried out by interference layering, plasma etching, and microscopy. These specimens contained free silicon and niobium disilicide as minor phases with silicon carbide as the major phase. In conventionally prepared samples, the niobium disilicide cannot be distinguished from silicon in optical micrographs. After interference layering, all phases are clearly distinguishable. Back scattered electron (BSE) imaging and energy dispersive spectrometry (EDS) confirmed the results obtained by interference layering. Plasma etching with CF4 plus 4% O2 selectively attacks silicon in these specimens. It is demonstrated that interference layering and plasma etching are very useful techniques in the phase identification and microstructural characterization of multiphase ceramic materials.

  5. Synthesis of niobium carbide by a high energy milling technique of powder metallurgy

    International Nuclear Information System (INIS)

    Antonello, Rodrigo Tecchio; Gonzalez, Cezar Henrique; Urtiga Filho, Severino Leopoldino; Araujo Filho, Oscar Olimpio de; Ambrozio Filho, Francisco

    2010-01-01

    The aim of this work is to obtain and characterize the Niobium Carbide (NbC) by a suitable high energy milling technique using a SPEX Mill vibratory type and niobium and carbon (graphite) powders. Since this carbide is scarced in the national market and it's necessary to apply this NbC as a reinforcement in two molybdenum high speed steels (AISI M2 and AISI M3:2) object of another work motivated this research. The powders were submitted to a high energy milling procedure for suitable times and conditions and then were characterized by means of Scanning Electronic Microscopy (SEM), Energy Dispersive Spectroscopy (EDS) and X-ray diffraction (DRX) techniques. The ball-to-powder weight ratio was 10:1. The analysed samples showed that the high-energy milling is an alternative route of the NbC synthesis. (author)

  6. Laboratory simulation studies of uranium mobility in natural waters

    International Nuclear Information System (INIS)

    Giblin, A.M.; Swaine, D.J.; Batts, B.D.

    1981-01-01

    The effects of imposed variations of pH and Eh on aqueous uranium mobility at 25 0 C have been studied in three simulations of natural water systems. Constituents tested for their effect on uranium mobility were: (a) hydrous ferric oxide, to represent adsorptive solids which precipitate or dissolve in response to variations in pH and Eh; (b) kaolinite, representing minerals which, although modified by pH and Eh changes, are present as solids over the pH-Eh range of natural waters; and (c) carbonate, to represent a strong uranium-complexing species. Uranium mobility measurements from each simulation were regressed against pH and Eh within a range appropriate to natural waters. Hydrous ferric oxide and kaolinite each affected uranium mobility, but in separate pH-Eh domains. Aqueous carbonate increased mobility of uranium, and adsorption of UO 2 (CO 3 ) 3 4- caused colloidal dispersion of hydrous ferric oxide, possibly explaining the presence of 'hydrothermal hematite' in some uranium deposits. Enhanced uranium mobility observed in the pH-Eh domains of thermodynamically insoluble uranium oxides could be explained if the oxides were present as colloids. Uranium persisting as a mobile species, even after reduction, has implications for the near surface genesis of uranium ores. (author)

  7. Uranium isotopes in ground water as a prospecting technique

    International Nuclear Information System (INIS)

    Cowart, J.B.; Osmond, J.K.

    1980-02-01

    The isotopic concentrations of dissolved uranium were determined for 300 ground water samples near eight known uranium accumulations to see if new approaches to prospecting could be developed. It is concluded that a plot of 234 U/ 238 U activity ratio (A.R.) versus uranium concentration (C) can be used to identify redox fronts, to locate uranium accumulations, and to determine whether such accumulations are being augmented or depleted by contemporary aquifer/ground water conditions. In aquifers exhibiting flow-through hydrologic systems, up-dip ground water samples are characterized by high uranium concentration values (> 1 to 4 ppB) and down-dip samples by low uranium concentration values (less than 1 ppB). The boundary between these two regimes can usually be identified as a redox front on the basis of regional water chemistry and known uranium accumulations. Close proximity to uranium accumulations is usually indicated either by very high uranium concentrations in the ground water or by a combination of high concentration and high activity ratio values. Ground waters down-dip from such accumulations often exhibit low uranium concentration values but retain their high A.R. values. This serves as a regional indicator of possible uranium accumulations where conditions favor the continued augmentation of the deposit by precipitation from ground water. Where the accumulation is being dispersed and depleted by the ground water system, low A.R. values are observed. Results from the Gulf Coast District of Texas and the Wyoming districts are presented

  8. Stability with temperature of mixed uranium plutonium monocarbides

    International Nuclear Information System (INIS)

    Riglet-Martial, Ch.; Dumas, J.C.; Piron, J.P.; Gueneau, Ch.

    2008-01-01

    Full text: Among the different advanced fuel materials of concern for Generation IV systems, the mixed carbide of uranium and plutonium fuel is considered as one of the key materials for Gas Fast Reactors (GFR) systems. For purposes of optimising its fabrication process as well as its performances in various operating conditions, the losses of gaseous plutonium specially at elevated temperatures have to be controlled and minimized. The paper is therefore concerned with a parametric analysis of the stability with temperature of mixed carbides of uranium and plutonium. Previous published experimental studies have shown that mixed (U ,Pu) carbides undergo a highly incongruent sublimation at high temperatures: the vapour phase in equilibrium with the solid is mainly composed of gaseous plutonium (P Pu /P total > 99 % ) while the contribution of gaseous U and C remains very low. The composition of the system U 1-z Pu z C 1+x ' (z =Pu/(U+Pu) and x C/(U+Pu)), the temperature (T) and the expansion volume (V) of the gas are the main parameters in the loss of gaseous Pu. The calculations are carried out using the SAGE (Solgasmix Advanced Gibbs Energy) software, by assuming ideal solid solutions between UC and PuC, as well as between U 2 C 3 and Pu 2 C 3 . The validity of the model is previously tested using published equilibrium vapour pressure data. This work gives rise to a large description of the variations of Pu losses from mixed uranium plutonium carbides and leads to some basic recommendations in connection with the use of this advanced fuel materials

  9. Uranium price reporting systems

    International Nuclear Information System (INIS)

    1987-09-01

    This report describes the systems for uranium price reporting currently available to the uranium industry. The report restricts itself to prices for U 3 O 8 natural uranium concentrates. Most purchases of natural uranium by utilities, and sales by producers, are conducted in this form. The bulk of uranium in electricity generation is enriched before use, and is converted to uranium hexafluoride, UF 6 , prior to enrichment. Some uranium is traded as UF 6 or as enriched uranium, particularly in the 'secondary' market. Prices for UF 6 and enriched uranium are not considered directly in this report. However, where transactions in UF 6 influence the reported price of U 3 O 8 this influence is taken into account. Unless otherwise indicated, the terms uranium and natural uranium used here refer exclusively to U 3 O 8 . (author)

  10. Uranium Industry Annual, 1992

    International Nuclear Information System (INIS)

    1993-01-01

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ''Decommissioning of US Conventional Uranium Production Centers,'' is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2

  11. Uranium Industry Annual, 1992

    Energy Technology Data Exchange (ETDEWEB)

    1993-10-28

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

  12. Preparation of aluminum nitride-silicon carbide nanocomposite powder by the nitridation of aluminum silicon carbide

    NARCIS (Netherlands)

    Itatani, K.; Tsukamoto, R.; Delsing, A.C.A.; Hintzen, H.T.J.M.; Okada, I.

    2002-01-01

    Aluminum nitride (AlN)-silicon carbide (SiC) nanocomposite powders were prepared by the nitridation of aluminum-silicon carbide (Al4SiC4) with the specific surface area of 15.5 m2·g-1. The powders nitrided at and above 1400°C for 3 h contained the 2H-phases which consisted of AlN-rich and SiC-rich

  13. Growth and structure of carbide nanorods

    International Nuclear Information System (INIS)

    Lieber, C.M.; Wong, E.W.; Dai, H.; Maynor, B.W.; Burns, L.D.

    1996-01-01

    Recent research on the growth and structure of carbide nanorods is reviewed. Carbide nanorods have been prepared by reacting carbon nanotubes with volatile transition metal and main group oxides and halides. Using this approach it has been possible to obtain solid carbide nanorods of TiC, SiC, NbC, Fe 3 C, and BC x having diameters between 2 and 30 nm and lengths up to 20 microm. Structural studies of single crystal TiC nanorods obtained through reactions of TiO with carbon nanotubes show that the nanorods grow along both [110] and [111] directions, and that the rods can exhibit either smooth or saw-tooth morphologies. Crystalline SiC nanorods have been produced from reactions of carbon nanotubes with SiO and Si-iodine reactants. The preferred growth direction of these nanorods is [111], although at low reaction temperatures rods with [100] growth axes are also observed. The growth mechanisms leading to these novel nanomaterials have also been addressed. Temperature dependent growth studies of TiC nanorods produced using a Ti-iodine reactant have provided definitive proof for a template or topotactic growth mechanism, and furthermore, have yielded new TiC nanotube materials. Investigations of the growth of SiC nanorods show that in some cases a catalytic mechanism may also be operable. Future research directions and applications of these new carbide nanorod materials are discussed

  14. Surface metallurgy of cemented carbide tools

    International Nuclear Information System (INIS)

    Chopra, K.L.; Kashyap, S.C.; Rao, T.V.; Rajagopalan, S.; Srivastava, P.K.

    1983-01-01

    Transition metal carbides, owing to their high melting point, hardness and wear resistance, are potential candidates for specific application in rockets, nuclear engineering equipment and cutting tools. Tungsten carbide sintered with a binder (either cobalt metal or a mixture of Co + TiC and/or TaC(NbC)) is used for cutting tools. The surface metallurgy of several commercially available cemented carbide tools was studied by Auger electron spectroscopy and X-ray photoelectron spectroscopy techniques. The tool surfaces were contaminated by adsorbed oxygen up to a depth of nearly 0.3 μm causing deterioration of the mechanical properties of the tools. Studies of fractured samples indicated that the tool surfaces were prone to oxygen adsorption. The fracture path passes through the cobalt-rich regions. The ineffectiveness of a worn cutting tool is attributed to the presence of excessive iron from the steel workpiece and carbon and oxygen in the surface layers of the tool. The use of appropriate hard coatings on cemented carbide tools is suggested. (Auth.)

  15. Silicon Carbide Power Devices and Integrated Circuits

    Science.gov (United States)

    Lauenstein, Jean-Marie; Casey, Megan; Samsel, Isaak; LaBel, Ken; Chen, Yuan; Ikpe, Stanley; Wilcox, Ted; Phan, Anthony; Kim, Hak; Topper, Alyson

    2017-01-01

    An overview of the NASA NEPP Program Silicon Carbide Power Device subtask is given, including the current task roadmap, partnerships, and future plans. Included are the Agency-wide efforts to promote development of single-event effect hardened SiC power devices for space applications.

  16. Low temperature CVD deposition of silicon carbide

    International Nuclear Information System (INIS)

    Dariel, M.; Yeheskel, J.; Agam, S.; Edelstein, D.; Lebovits, O.; Ron, Y.

    1991-04-01

    The coating of graphite on silicon carbide from the gaseous phase in a hot-well, open flow reactor at 1150degC is described. This study constitutes the first part of an investigation of the process for the coating of nuclear fuel by chemical vapor deposition (CVD)

  17. Anomalous Seebeck coefficient in boron carbides

    International Nuclear Information System (INIS)

    Aselage, T.L.; Emin, D.; Wood, C.; Mackinnon, I.D.R.; Howard, I.A.

    1987-01-01

    Boron carbides exhibit an anomalously large Seebeck coefficient with a temperature coefficient that is characteristic of polaronic hopping between inequivalent sites. The inequivalence in the sites is associated with disorder in the solid. The temperature dependence of the Seebeck coefficient for materials prepared by different techniques provides insight into the nature of the disorder

  18. Reaction of boron carbide with molybdenum disilicide

    International Nuclear Information System (INIS)

    Novikov, A.V.; Melekhin, V.F.; Pegov, V.S.

    1989-01-01

    The investigation results of interaction in the B 4 C-MoSi 2 system during sintering in vacuum are presented. Sintering of boron carbide with molybdenum disilicide is shown to lead to the formation of MoB 2 , SiC, Mo 5 Si 3 compounds, the presence of carbon-containing covering plays an important role in sintering

  19. Mechanical characteristics of microwave sintered silicon carbide

    Indian Academy of Sciences (India)

    In firing of products by conventionally sintered process, SiC grain gets oxidized producing SiO2 (∼ 32 wt%) and deteriorates the quality of the product substantially. Partially sintered silicon carbide by such a method is a useful material for a varieties of applications ranging from kiln furniture to membrane material.

  20. Visible light emission from porous silicon carbide

    DEFF Research Database (Denmark)

    Ou, Haiyan; Lu, Weifang

    2017-01-01

    Light-emitting silicon carbide is emerging as an environment-friendly wavelength converter in the application of light-emitting diode based white light source for two main reasons. Firstly, SiC has very good thermal conductivity and therefore a good substrate for GaN growth in addition to the small...

  1. Provision by the uranium and uranium products

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2005-01-01

    International uranium market is converted from the buyer market into the seller market. The prices of uranium are high and the market attempts to adapt to changing circumstances. The industry of uranium enrichment satisfies the increasing demands but should to increase ots capacities. On the whole the situation is not stable and every year may change the existing position [ru

  2. Uranium recovery from slags of metallic uranium

    International Nuclear Information System (INIS)

    Fornarolo, F.; Frajndlich, E.U.C.; Durazzo, M.

    2006-01-01

    The Center of the Nuclear Fuel of the Institute of Nuclear Energy Research - IPEN finished the program of attainment of fuel development for research reactors the base of Uranium Scilicet (U 3 Si 2 ) from Hexafluoride of Uranium (UF 6 ) with enrichment 20% in weight of 235 U. In the process of attainment of the league of U 3 Si 2 we have as Uranium intermediate product the metallic one whose attainment generates a slag contend Uranium. The present work shows the results gotten in the process of recovery of Uranium in slags of calcined slags of Uranium metallic. Uranium the metallic one is unstable, pyrophoricity and extremely reactive, whereas the U 3 O 8 is a steady oxide of low chemical reactivity, what it justifies the process of calcination of slags of Uranium metallic. The calcination of the Uranium slag of the metallic one in oxygen presence reduces Uranium metallic the U 3 O 8 . Experiments had been developed varying it of acid for Uranium control and excess, nitric molar concentration gram with regard to the stoichiometric leaching reaction of temperature of the leaching process. The 96,0% income proves the viability of the recovery process of slags of Uranium metallic, adopting it previous calcination of these slags in nitric way with low acid concentration and low temperature of leaching. (author)

  3. Uranium enrichment

    International Nuclear Information System (INIS)

    Mohrhauer, H.

    1982-01-01

    The separation of uranium isotopes in order to enrich the fuel for light water reactors with the light isotope U-235 is an important part of the nuclear fuel cycle. After the basic principals of isotope separation the gaseous diffusion and the centrifuge process are explained. Both these techniques are employed on an industrial scale. In addition a short review is given on other enrichment techniques which have been demonstrated at least on a laboratory scale. After some remarks on the present situation on the enrichment market the progress in the development and the industrial exploitation of the gas centrifuge process by the trinational Urenco-Centec organisation is presented. (orig.)

  4. Uranium conversion; Urankonvertering

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina [Swedish Defence Research Agency (FOI), Stockholm (Sweden)

    2006-03-15

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF{sub 6} and UF{sub 4} are present require equipment that is made of corrosion resistant material.

  5. Chemical dispersants

    NARCIS (Netherlands)

    Rahsepar, Shokouhalsadat; Smit, Martijn P.J.; Murk, Albertinka J.; Rijnaarts, Huub H.M.; Langenhoff, Alette A.M.

    2016-01-01

    Chemical dispersants were used in response to the Deepwater Horizon oil spill in the Gulf of Mexico, both at the sea surface and the wellhead. Their effect on oil biodegradation is unclear, as studies showed both inhibition and enhancement. This study addresses the effect of Corexit on oil

  6. Uranium under its depleted state

    International Nuclear Information System (INIS)

    2001-01-01

    This day organised by the SFRP, with the help of the Army Health service, the service of radiation protection of Army and IPSN is an information day to inform the public about the real toxicity of uranium, and its becoming in man and environment, about the risks during the use of depleted uranium and eventual consequences of its dispersion after a conflict, to give information on how is managed the protection of workers (civil or military ones) and what is really the situation of French military personnel in these conflicts. The news have brought to the shore cases of leukemia it is necessary to bring some information to the origin of this disease. (N.C.)

  7. Issues in uranium availability

    International Nuclear Information System (INIS)

    Schanz, J.J. Jr.; Adams, S.S.; Gordon, R.L.

    1982-01-01

    The purpose of this publication is to show the process by which information about uranium reserves and resources is developed, evaluated and used. The following three papers in this volume have been abstracted and indexed for the Energy Data Base: (1) uranium reserve and resource assessment; (2) exploration for uranium in the United States; (3) nuclear power, the uranium industry, and resource development

  8. Australian uranium industry

    Energy Technology Data Exchange (ETDEWEB)

    Warner, R K

    1976-04-01

    Various aspects of the Australian uranium industry are discussed including the prospecting, exploration and mining of uranium ores, world supply and demand, the price of uranium and the nuclear fuel cycle. The market for uranium and the future development of the industry are described.

  9. Irradiated uranium reprocessing

    International Nuclear Information System (INIS)

    Gal, I.

    1961-12-01

    Task concerned with reprocessing of irradiated uranium covered the following activities: implementing the method and constructing the cell for uranium dissolving; implementing the procedure for extraction of uranium, plutonium and fission products from radioactive uranium solutions; studying the possibilities for using inorganic ion exchangers and adsorbers for separation of U, Pu and fission products

  10. Uranium processing and properties

    CERN Document Server

    2013-01-01

    Covers a broad spectrum of topics and applications that deal with uranium processing and the properties of uranium Offers extensive coverage of both new and established practices for dealing with uranium supplies in nuclear engineering Promotes the documentation of the state-of-the-art processing techniques utilized for uranium and other specialty metals

  11. Recovering uranium from phosphates

    Energy Technology Data Exchange (ETDEWEB)

    Bergeret, M [Compagnie de Produits Chimiques et Electrometallurgiques Pechiney-Ugine Kuhlmann, 75 - Paris (France)

    1981-06-01

    Processes for the recovery of the uranium contained in phosphates have today become competitive with traditional methods of working uranium sources. These new possibilities will make it possible to meet more rapidly any increases in the demand for uranium: it takes ten years to start working a new uranium deposit, but only two years to build a recovery plant.

  12. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Gagne, R.W.; Thomas, D.C.

    1977-01-01

    The status of existing uranium enrichment contracts in the US is reviewed and expected natural uranium requirements for existing domestic uranium enrichment contracts are evaluated. Uncertainty in natural uranium requirements associated with requirements-type and fixed-commitment type contracts is discussed along with implementation of variable tails assay

  13. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Thomas, D.C.; Gagne, R.W.

    1978-01-01

    The following topics are covered: the status of the Government's existing uranium enrichment services contracts, natural uranium requirements based on the latest contract information, uncertainty in predicting natural uranium requirements based on uranium enrichment contracts, and domestic and foreign demand assumed in enrichment planning

  14. The uranium in Kvanefjeld

    International Nuclear Information System (INIS)

    Nielsen, J.S.

    1983-08-01

    The report is a final thesis at the study of biology at the University of Copenhagen. It examines on a theoretical basis a number of possible environmental effects from a uranium mining and milling project under consideration at the Kvanefjeld site near Narssaq in South Greenland. An introductory description and discussion of the advantages and limitations of ecological baseline studies and dose committment assessments as a tool for planning and decision making is given. The leaching and atmospheric dispersion of particles, heavy metals, radionuclides and other elements from future waste rock and ore piles as well as from mill tailings at the Kvanefjeld site are analysed and discussed. Also, the mobility, transport and accumulation of potentially toxic elements in local terrestrial and aquatic ecosystems and food chains are examined. The resulting human burden are discussed with special attention given to the impact on the local population from eating lamb and seafood. A special quantitative analysis of the dispersion and biological uptake of fluoride, which is found in high concentrations in the ore, is given, focusing on the possible human intake of fluoride-polluted arctic char (Salvelinus alpinus) caught in Narrssaq River. The report at the end gives consideration to the long term problems of controlling mill tailings, discussing among other things the long term human exposure to radon and thoron daughters. (author)

  15. Uranium industry annual 1985

    International Nuclear Information System (INIS)

    1986-11-01

    This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities, utility market requirements, and related topics. A glossary and appendices are included to assist the reader in interpreting the substantial array of statistical data in this report and to provide background information about the survey

  16. Uranium industry framework

    International Nuclear Information System (INIS)

    Riley, K.

    2008-01-01

    The global uranium market is undergoing a major expansion due to an increase in global demand for uranium, the highest uranium prices in the last 20 years and recognition of the potential greenhouse benefits of nuclear power. Australia holds approximately 27% of the world's uranium resources (recoverable at under US$80/kg U), so is well placed to benefit from the expansion in the global uranium market. Increasing exploration activity due to these factors is resulting in the discovery and delineation of further high grade uranium deposits and extending Australia's strategic position as a reliable and safe supplier of low cost uranium.

  17. Elemental profiling of laser cladded multilayer coatings by laser induced breakdown spectroscopy and energy dispersive X-ray spectroscopy

    Science.gov (United States)

    Lednev, V. N.; Sdvizhenskii, P. A.; Filippov, M. N.; Grishin, M. Ya.; Filichkina, V. A.; Stavertiy, A. Ya.; Tretyakov, R. S.; Bunkin, A. F.; Pershin, S. M.

    2017-09-01

    Multilayer tungsten carbide wear resistant coatings were analyzed by laser induced breakdown spectroscopy (LIBS) and energy dispersive X-ray (EDX) spectroscopy. Coaxial laser cladding technique was utilized to produce tungsten carbide coating deposited on low alloy steel substrate with additional inconel 625 interlayer. EDX and LIBS techniques were used for elemental profiling of major components (Ni, W, C, Fe, etc.) in the coating. A good correlation between EDX and LIBS data was observed while LIBS provided additional information on light element distribution (carbon). A non-uniform distribution of tungsten carbide grains along coating depth was detected by both LIBS and EDX. In contrast, horizontal elemental profiling showed a uniform tungsten carbide particles distribution. Depth elemental profiling by layer-by-layer LIBS analysis was demonstrated to be an effective method for studying tungsten carbide grains distribution in wear resistant coating without any sample preparation.

  18. Reduction of uranium hexafluoride to uranium tetrafluoride

    International Nuclear Information System (INIS)

    Chang, I.S.; Do, J.B.; Choi, Y.D.; Park, M.H.; Yun, H.H.; Kim, E.H.; Kim, Y.W.

    1982-01-01

    The single step continuous reduction of uranium hexafluoride (UF 6 ) to uranium tetrafluoride (UF 4 ) has been investigated. Heat required to initiate and maintain the reaction in the reactor is supplied by the highly exothermic reaction of hydrogen with a small amount of elemental fluorine which is added to the uranium hexafluoride stream. When gases uranium hexafluoride and hydrogen react in a vertical monel pipe reactor, the green product, UF 4 has 2.5g/cc in bulk density and is partly contaminated by incomplete reduction products (UF 5 ,U 2 F 9 ) and the corrosion product, presumably, of monel pipe of the reactor itself, but its assay (93% of UF 4 ) is acceptable for the preparation of uranium metal with magnesium metal. Remaining problems are the handling of uranium hexafluoride, which is easily clogging the flowmeter and gas feeding lines because of extreme sensitivity toward moisture, and a development of gas nozzel for free flow of uranium hexafluoride gas. (Author)

  19. Uranium - the world picture

    International Nuclear Information System (INIS)

    Silver, J.M.; Wright, W.J.

    1976-01-01

    The world resources of uranium and the future demand for uranium are discussed. The amount of uranium available depends on the price which users are prepared to pay for its recovery. As the price is increased, there is an incentive to recover uranium from lower grade or more difficult deposits. In view of this, attention is drawn to the development of the uranium industry in Australias

  20. Chemical vapor deposition of tantalum on graphite cloth for making hot pressed fiber reinforced carbide-graphite composite

    International Nuclear Information System (INIS)

    Hollabaugh, C.M.; Davidson, K.V.; Radosevich, C.L.; Riley, R.E.; Wallace, T.C.

    1977-01-01

    Conditions for the CVD of a uniform coating of Ta on fibers of a woven graphite cloth were established. The effect of gas composition, pressure, and temperature were investigated, and the conditions that gave the desired results are presented. Several layers of the coated cloth were hot pressed to produce a TaC--C composite having uniformly dispersed, fine-grained TaC in graphite. Three compositions were hot pressed: 15, 25, and 40 volume percent carbide. 8 figures, 2 tables

  1. Assessment Of Depleted Uranium Contamination In Selective IRAQI Soils

    International Nuclear Information System (INIS)

    Mohammed, A.A.; Hussien, A.Sh.M.; Tawfiq, N.F.

    2008-01-01

    The aim of this research was to measure the radiation exposure rates in three selected Locations in southren part of Iraq (two in Nassireya, and one in Amara) resulted from the existence of depleted uranium in soil and metal pieces have been taken from destroyed tank and study mathmatically the concentration of Depleted Uranium by its dispersion from soil surface by winds and rains from 2003 to 2007. The exposure rates were measured using inspector device, while depleted uranium concentration in soil samples and tank's matal pieces were detected with Solid State Nuclear Track Detectors(SSNTDs). The wind and rain effects were considered in the calculation of dispersion effect on depleted uranium concentration in soil, where the wind effect were calculated with respect to the sites nature and soil conditions, and rain effect with respect to dispersive-convective equation for radionuclide in soil. The results obtained for the exposure rates were high near the penetrated surfac, moderate and low in soil and metal pices. The Depleted Uranium concentration in soil and metal pieces have the highest value in Nassireya. The results from dispersion calculation (wind & rain) showed that the depleted uranium concentration in 2008 will be less than the danger level and in allowable contamination range

  2. Verification of the visual control efficacity on quenching effect produced by Cr, Mn, Co and Fe in the uranium fluorimetric determination, by quantifying these metals with energy dispersive X-ray fluorescence

    International Nuclear Information System (INIS)

    Morales Sanchez, E.A.

    1986-08-01

    It's described the APDC preconcentration procedure, applied to the quantification of chromium, manganese, iron and cobalt in river sediments which have been digested in order to evaluate their uranium content by fluorimetry. The purpose was to show the lack of correlation between the visual control of the quenching effect and its actual presence. A Cd-109 source and a Si-Li detector were used. Thin film method for avoiding matrix effects and a deconvolution soft-ware, AXIL, for correcting peaks overlapping were applied. Precision and accuracy were 2% and 10% respectively in the 2 to 5 micro-grams/milliliter range. It was shown that for the samples analyzed there was not a good correlation between the visual observations and the quenching effect. It was found that the principal - quencher was the iron, present in so high concentration that is caused up to 80% of attenuation. (author)

  3. Natural uranium

    International Nuclear Information System (INIS)

    Ammerich, Marc; Frot, Patricia; Gambini, Denis-Jean; Gauron, Christine; Moureaux, Patrick; Herbelet, Gilbert; Lahaye, Thierry; Pihet, Pascal; Rannou, Alain

    2014-08-01

    This sheet belongs to a collection which relates to the use of radionuclides essentially in unsealed sources. Its goal is to gather on a single document the most relevant information as well as the best prevention practices to be implemented. These sheets are made for the persons in charge of radiation protection: users, radioprotection-skill persons, labor physicians. Each sheet treats of: 1 - the radio-physical and biological properties; 2 - the main uses; 3 - the dosimetric parameters; 4 - the measurement; 5 - the protection means; 6 - the areas delimitation and monitoring; 7 - the personnel classification, training and monitoring; 8 - the effluents and wastes; 9 - the authorization and declaration administrative procedures; 10 - the transport; and 11 - the right conduct to adopt in case of incident or accident. This sheet deals specifically with natural uranium

  4. Rf-plasma synthesis of nanosize silicon carbide and nitride. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Buss, R.J.

    1997-02-01

    A pulsed rf plasma technique is capable of generating ceramic particles of 10 manometer dimension. Experiments using silane/ammonia and trimethylchlorosilane/hydrogen gas mixtures show that both silicon nitride and silicon carbide powders can be synthesized with control of the average particle diameter from 7 to 200 nm. Large size dispersion and much agglomeration appear characteristic of the method, in contrast to results reported by another research group. The as produced powders have a high hydrogen content and are air and moisture sensitive. Post-plasma treatment in a controlled atmosphere at elevated temperature (800{degrees}C) eliminates the hydrogen and stabilizes the powder with respect to oxidation or hydrolysis.

  5. Tribological Characteristics of Tungsten Carbide Reinforced Arc Sprayed Coatings using Different Carbide Grain Size Fractions

    Directory of Open Access Journals (Sweden)

    W. Tillmann

    2017-06-01

    Full Text Available Tungsten carbide reinforced coatings play an important role in the field of surface engineering to protect stressed surfaces against wear. For thermally sprayed coatings, it is already shown that the tribological properties get mainly determined by the carbide grain size fraction. Within the scope of this study, the tribological characteristics of iron based WC-W2C reinforced arc sprayed coatings deposited using cored wires consisting of different carbide grain size fractions were examined. Microstructural characteristics of the produced coatings were scrutinized using electron microscopy and x-ray diffraction analyses. Ball-on-disk test as well as Taber Abraser and dry sand rubber wheel test were employed to analyze both the dry sliding and the abrasive wear behavior. It was shown that a reduced carbide grain size fraction as filling leads to an enhanced wear resistance against sliding. In terms of the Taber Abraser test, it is also demonstrated that a fine carbide grain size fraction results in an improved wear resistant against abrasion. As opposed to that, a poorer wear resistance was found within the dry sand rubber wheel tests. The findings show that the operating mechanisms for both abrasion tests affect the stressed surface in a different way, leading either to microcutting or microploughing.

  6. Neutron irradiation damage in transition metal carbides

    International Nuclear Information System (INIS)

    Matsui, Hisayuki; Nesaki, Kouji; Kiritani, Michio

    1991-01-01

    Effects of neutron irradiation on the physical properties of light transition metal carbides, TiC x , VC x and NbC x , were examined, emphasizing the characterization of irradiation induced defects in the nonstoichiometric composition. TiC x irradiated with 14 MeV (fusion) neutrons showed higher damage rates with increasing C/Ti (x) ratio. A brief discussion is made on 'cascade damage' in TiC x irradiated with fusion neutrons. Two other carbides (VC x and NbC x ) were irradiated with fission reactor neutrons. The irradiation effects on VC x were not so simple, because of the complex irradiation behavior of 'ordered' phases. For instance, complete disordering was revealed in an ordered phase, 'V 8 C 7 ', after an irradiation dose of 10 25 n/m 2 . (orig.)

  7. Seebeck effect of some thin film carbides

    International Nuclear Information System (INIS)

    Beensh-Marchwicka, G.; Prociow, E.

    2002-01-01

    Several materials have been investigated for high-temperature thin film thermocouple applications. These include silicon carbide with boron (Si-C-B), ternary composition based on Si-C-Mn, fourfold composition based on Si-C-Zr-B and tantalum carbide (TaC). All materials were deposited on quartz or glass substrates using the pulse sputter deposition technique. Electrical conduction and thermoelectric power were measured for various compositions at 300-550 K. It has been found, that the efficiency of thermoelectric power of films containing Si-C base composition was varied from 0.0015-0.034 μW/cmK 2 . However for TaC the value about 0.093 μW/cmK 2 was obtained. (author)

  8. METHOD FOR PRODUCING CEMENTED CARBIDE ARTICLES

    Science.gov (United States)

    Onstott, E.I.; Cremer, G.D.

    1959-07-14

    A method is described for making molded materials of intricate shape where the materials consist of mixtures of one or more hard metal carbides or oxides and matrix metals or binder metals thereof. In one embodiment of the invention 90% of finely comminuted tungsten carbide powder together with finely comminuted cobalt bonding agent is incorporated at 60 deg C into a slurry with methyl alcohol containing 1.5% paraffin, 3% camphor, 3.5% naphthalene, and 1.8% toluene. The compact is formed by the steps of placing the slurry in a mold at least one surface of which is porous to the fluid organic system, compacting the slurry, removing a portion of the mold from contact with the formed object and heating the formed object to remove the remaining organic matter and to sinter the compact.

  9. Radiation stability of proton irradiated zirconium carbide

    International Nuclear Information System (INIS)

    Yang, Yong; Dickerson, Clayton A.; Allen, Todd R.

    2009-01-01

    The use of zirconium carbide (ZrC) is being considered for the deep burn (DB)-TRISO fuel as a replacement for the silicon carbide coating. The radiation stability of ZrC was studied using 2.6 MeV protons, across the irradiation temperature range from 600 to 900degC and to doses up to 1.75 dpa. The microstructural characterization shows that the irradiated microstructure is comprised of a high density of nanometer-sized dislocation loops, while no irradiation induced amorphization or voids are observed. The lattice expansion induced by point defects is found to increase as the dose increases for the samples irradiated at 600 and 800degC, while for the 900degC irradiation, a slight lattice contraction is observed. The radiation hardening is also quantified using a micro indentation technique for the temperature and doses studies. (author)

  10. Oxidation of boron carbide at high temperatures

    International Nuclear Information System (INIS)

    Steinbrueck, Martin

    2005-01-01

    The oxidation kinetics of various types of boron carbides (pellets, powder) were investigated in the temperature range between 1073 and 1873 K. Oxidation rates were measured in transient and isothermal tests by means of mass spectrometric gas analysis. Oxidation of boron carbide is controlled by the formation of superficial liquid boron oxide and its loss due to the reaction with surplus steam to volatile boric acids and/or direct evaporation at temperatures above 1770 K. The overall reaction kinetics is paralinear. Linear oxidation kinetics established soon after the initiation of oxidation under the test conditions described in this report. Oxidation is strongly influenced by the thermohydraulic boundary conditions and in particular by the steam partial pressure and flow rate. On the other hand, the microstructure of the B 4 C samples has a limited influence on oxidation. Very low amounts of methane were produced in these tests

  11. An improved method of preparing silicon carbide

    International Nuclear Information System (INIS)

    Baney, R.H.

    1979-01-01

    A method of preparing silicon carbide is described which comprises forming a desired shape from a polysilane of the average formula:[(CH 3 ) 2 Si][CH 3 Si]. The polysilane contains from 0 to 60 mole percent (CH 3 ) 2 Si units and from 40 to 100 mole percent CH 3 Si units. The remaining bonds on the silicon are attached to another silicon atom or to a halogen atom in such manner that the average ratio of halogen to silicon in the polysilane is from 0.3:1 to 1:1. The polysilane has a melt viscosity at 150 0 C of from 0.005 to 500 Pa.s and an intrinsic viscosity in toluene of from 0.0001 to 0.1. The shaped polysilane is heated in an inert atmosphere or in a vacuum to an elevated temperature until the polysilane is converted to silicon carbide. (author)

  12. Uranium management activities

    International Nuclear Information System (INIS)

    Jackson, D.; Marshall, E.; Sideris, T.; Vasa-Sideris, S.

    2001-01-01

    One of the missions of the Department of Energy's (DOE) Oak Ridge Office (ORO) has been the management of the Department's uranium materials. This mission has been accomplished through successful integration of ORO's uranium activities with the rest of the DOE complex. Beginning in the 1980's, several of the facilities in that complex have been shut down and are in the decommissioning process. With the end of the Cold War, the shutdown of many other facilities is planned. As a result, inventories of uranium need to be removed from the Department facilities. These inventories include highly enriched uranium (HEU), low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU). The uranium materials exist in different chemical forms, including metals, oxides, solutions, and gases. Much of the uranium in these inventories is not needed to support national priorities and programs. (author)

  13. Effect of boron on the microstructure and mechanical properties of carbidic austempered ductile iron

    International Nuclear Information System (INIS)

    Peng Yuncheng; Jin Huijin; Liu Jinhai; Li Guolu

    2011-01-01

    Highlights: → Boron are applied to carbidic austempered ductile iron (CADI). → Boron microalloying CADI is a new high hardenability of wear-resistant cast iron. → Addition of boron to CADI significantly improves hardenability. → Effect of boron on the CADI grinding ball were investigated. → Optimum property is obtained when boron content at 0.03 wt%. - Abstract: Carbidic austempered ductile iron (CADI) castings provide a unique combination of high hardness and toughness coupled with superior wear resistance properties, but their hardenability restricts their range of applications. The purpose of this study was to investigate the influence of boron on the microstructure and mechanical properties of CADI. The experimental results indicate that the CADI comprises graphite nodules, which are dispersive boron-carbides that are distributed in the form of strips, and the matrix is a typical ausferritic matrix. Microscopic amounts of boron can improve the hardenability of CADI, but higher boron content reduces the hardenability and toughness of CADI. The results are discussed in the context of the influence of boron content on the microstructure and mechanical properties of grinding balls.

  14. Study of nano-metric silicon carbide powder sintering. Application to fibers processing

    International Nuclear Information System (INIS)

    Malinge, A.

    2011-01-01

    Silicon carbide ceramic matrix composites (SiCf/SiCm) are of interest for high temperature applications in aerospace or nuclear components for their relatively high thermal conductivity and low activation under neutron irradiation. While most of silicon carbide fibers are obtained through the pyrolysis of a poly-carbo-silane precursor, sintering of silicon carbide nano-powders seems to be a promising route to explore. For this reason, pressureless sintering of SiC has been studied. Following the identification of appropriate sintering aids for the densification, optimization of the microstructure has been achieved through (i) the analysis of the influence of operating parameters and (ii) the control of the SiC β a SiC α phase transition. Green fibers have been obtained by two different processes involving the extrusion of SiC powder dispersion in polymer solution or the coagulation of a water-soluble polymer containing ceramic particles. Sintering of these green fibers led to fibers of around fifty microns in diameter. (author) [fr

  15. Uranium industry annual 1998

    International Nuclear Information System (INIS)

    1999-01-01

    The Uranium Industry Annual 1998 (UIA 1998) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. It contains data for the period 1989 through 2008 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data provides a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Data on uranium raw materials activities for 1989 through 1998, including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment, are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2008, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, and uranium inventories, are shown in Chapter 2. The methodology used in the 1998 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ''Uranium Industry Annual Survey'' is provided in Appendix C. The Form EIA-858 ''Uranium Industry Annual Survey'' is shown in Appendix D. For the readers convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix E along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 24 figs., 56 tabs

  16. Uranium industry annual 1994

    International Nuclear Information System (INIS)

    1995-01-01

    The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data collected on the ''Uranium Industry Annual Survey'' (UIAS) provide a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ''Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,'' is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2

  17. High resolution imaging of boron carbide microstructures

    International Nuclear Information System (INIS)

    MacKinnon, I.D.R.; Aselage, T.; Van Deusen, S.B.

    1986-01-01

    Two samples of boron carbide have been examined using high resolution transmission electron microscopy (HRTEM). A hot-pressed B 13 C 2 sample shows a high density of variable width twins normal to (10*1). Subtle shifts or offsets of lattice fringes along the twin plane and normal to approx.(10*5) were also observed. A B 4 C powder showed little evidence of stacking disorder in crystalline regions

  18. Spark plasma sintering of tantalum carbide

    International Nuclear Information System (INIS)

    Khaleghi, Evan; Lin, Yen-Shan; Meyers, Marc A.; Olevsky, Eugene A.

    2010-01-01

    A tantalum carbide powder was consolidated by spark plasma sintering. The specimens were processed under various temperature and pressure conditions and characterized in terms of relative density, grain size, rupture strength and hardness. The results are compared to hot pressing conducted under similar settings. It is shown that high densification is accompanied by substantial grain growth. Carbon nanotubes were added to mitigate grain growth; however, while increasing specimens' rupture strength and final density, they had little effect on grain growth.

  19. Recycling of uranium by a perennial vegetation

    International Nuclear Information System (INIS)

    Thiry, Y.

    2005-01-01

    At sites of large scale mining and processing of uranium ore, tailings and waste rock piles are today the most visible relics of the uranium extractive industry. These mining relics are constantly subjected to weathering and leaching processes causing the dissemination of radioactive and toxic elements and sometimes requiring remedial operations. The in situ remediation of waste rock piles usually includes their revegetation for minimizing the water infiltration and for increasing surface soil stability. Thanks to its biomass density and longevity, the perennial vegetation plays an important role in stabilisation of the water cycling. The buffer role of forest vegetation can reduce water export from watersheds as well as erosion and hydrological losses of chemicals including radionuclides from contaminated sites. If long term reduction of contaminant dispersion at revegetated uranium mining sites is to be fully appreciated, then the extent of radioactive contaminant availability to forest vegetation and ecosystem cycling as well as the possible economic valorisation of the woody products must be considered. Concerned study focused on a Scots pine plantation established 35 years ago on a uranium waste rock pile (Wismuth GmbH) situated near Schlema (Germany). This investigation aimed at quantifying the mobility of uranium in the mining debris and its transport to the different tree compartments with emphasis on the processes involved. The influence of pine vegetation on uranium cycling dynamics was further assessed in terms of annual fluxes)

  20. HCl removal using cycled carbide slag from calcium looping cycles

    International Nuclear Information System (INIS)

    Xie, Xin; Li, Yingjie; Wang, Wenjing; Shi, Lei

    2014-01-01

    Highlights: • Cycled carbide slag from calcium looping cycles is used to remove HCl. • The optimum temperature for HCl removal of cycled carbide slag is 700 °C. • The presence of CO 2 restrains HCl removal of cycled carbide slag. • CO 2 capture conditions have important effects on HCl removal of cycled carbide slag. • HCl removal capacity of carbide slag drops with cycle number rising from 1 to 50. - Abstract: The carbide slag is an industrial waste from chlor-alkali plants, which can be used to capture CO 2 in the calcium looping cycles, i.e. carbonation/calcination cycles. In this work, the cycled carbide slag from the calcium looping cycles for CO 2 capture was proposed to remove HCl in the flue gas from the biomass-fired and RDFs-fired boilers. The effects of chlorination temperature, HCl concentration, particle size, presence of CO 2 , presence of O 2 , cycle number and CO 2 capture conditions in calcium looping cycles on the HCl removal behavior of the carbide slag experienced carbonation/calcination cycles were investigated in a triple fixed-bed reactor. The chlorination product of the cycled carbide slag from the calcium looping after absorbing HCl is not CaCl 2 but CaClOH. The optimum temperature for HCl removal of the cycled carbide slag from the carbonation/calcination cycles is 700 °C. The chlorination conversion of the cycled carbide slag increases with increasing the HCl concentration. The cycled carbide slag with larger particle size exhibits a lower chlorination conversion. The presence of CO 2 decreases the chlorination conversions of the cycled carbide slag and the presence of O 2 has a trifling impact. The chlorination conversion of the carbide slag experienced 1 carbonation/calcination cycle is higher than that of the uncycled calcined sorbent. As the number of carbonation/calcination cycles increases from 1 to 50, the chlorination conversion of carbide slag drops gradually. The high calcination temperature and high CO 2

  1. Electronic specific heat of transition metal carbides

    International Nuclear Information System (INIS)

    Conte, R.

    1964-07-01

    The experimental results that make it possible to define the band structure of transition metal carbides having an NaCI structure are still very few. We have measured the electronic specific heat of some of these carbides of varying electronic concentration (TiC, either stoichiometric or non-stoichiometric, TaC and mixed (Ti, Ta) - C). We give the main characteristics (metallography, resistivity, X-rays) of our samples and we describe the low temperature specific heat apparatus which has been built. In one of these we use helium as the exchange gas. The other is set up with a mechanical contact. The two use a germanium probe for thermometer. The measurement of the temperature using this probe is described, as well as the various measurement devices. The results are presented in the form of a rigid band model and show that the density of the states at the Fermi level has a minimum in the neighbourhood of the group IV carbides. (author) [fr

  2. Laser deposition of carbide-reinforced coatings

    International Nuclear Information System (INIS)

    Cerri, W.; Martinella, R.; Mor, G.P.; Bianchi, P.; D'Angelo, D.

    1991-01-01

    CO 2 laser cladding with blown powder presents many advantages: fusion bonding with the substrate with low dilution, metallurgical continuity in the metallic matrix, high solidification rates, ease of automation, and reduced environmental contamination. In the present paper, laser cladding experimental results using families of carbides (tungsten and titanium) mixed with metallic alloys are reported. As substrates, low alloy construction steel (AISI 4140) (austenitic stainless steel) samples have been utilized, depending on the particular carbide reinforcement application. The coating layers obtained have been characterized by metallurgical examination. They show low dilution, absence of cracks, and high abrasion resistance. The WC samples, obtained with different carbide sizes and percentages, have been characterized with dry and rubber wheel abrasion tests and the specimen behaviour has been compared with the behaviour of materials used for similar applications. The abrasion resistance proved to be better than that of other widely used hardfacing materials and the powder morphology have a non-negligible influence on the tribological properties. (orig.)

  3. Doping of silicon carbide by ion implantation

    International Nuclear Information System (INIS)

    Gimbert, J.

    1999-01-01

    It appeared that in some fields, as the hostile environments (high temperature or irradiation), the silicon compounds showed limitations resulting from the electrical and mechanical properties. Doping of 4H and 6H silicon carbide by ion implantation is studied from a physicochemical and electrical point of view. It is necessary to obtain n-type and p-type material to realize high power and/or high frequency devices, such as MESFETs and Schottky diodes. First, physical and electrical properties of silicon carbide are presented and the interest of developing a process technology on this material is emphasised. Then, physical characteristics of ion implantation and particularly classical dopant implantation, such as nitrogen, for n-type doping, and aluminium and boron, for p-type doping are described. Results with these dopants are presented and analysed. Optimal conditions are extracted from these experiences so as to obtain a good crystal quality and a surface state allowing device fabrication. Electrical conduction is then described in the 4H and 6H-SiC polytypes. Freezing of free carriers and scattering processes are described. Electrical measurements are carried out using Hall effect on Van der Panw test patterns, and 4 point probe method are used to draw the type of the material, free carrier concentrations, resistivity and mobility of the implanted doped layers. These results are commented and compared to the theoretical analysis. The influence of the technological process on electrical conduction is studied in view of fabricating implanted silicon carbide devices. (author)

  4. Geochemical exploration for uranium

    International Nuclear Information System (INIS)

    Rose, A.W.

    1977-01-01

    The processes and types of dispersion that produce anomalies in stream water, stream sediment, and ground water, and the factors that must be considered in planning and interpreting geochemical surveys are reviewed. Examples of surveys near known deposits show the types of results to be expected. Background values depend mainly on the content of U in rocks of the drainage area. In igneous rocks, U tends to increase with potassium from ultramafic rocks (0.01 ppM) to granitic rocks (1 to 5 ppM). Some alkalic rocks have unusually high contents of U (15 to 100 ppM). Uranium-rich provinces marked by igneous rocks unusually rich in U are recognized in several areas and appear to have a deep crustal or mantle origin. In western U.S., many tertiary tuffaceous rocks have a high U content. Sandstones, limestones, and many shales approximate the crustal abundance at 0.5 to 4 ppM, but black shales, phosphates, and some organic materials are notably enriched in U. Uranium is very soluble in most oxidizing waters at the earth's surface, but is precipitated by reducing agents (organic matter, H 2 S) and adsorbed by organic material and some Fe oxides. In most surface and ground waters, U correlates approximately with the total dissolved solids, conductivity, and bicarbonate concentration of the water, and with the U content of rocks it comes into contact with. Most surveys of stream water near known districts show distinct anomalies extending a few km to tens of km downstream. A complication with water is the large variability with time, up to x 50, as a result of changes in the ratio of ground water to direct runoff, and changes in rate of oxidation and leaching. Collection and analysis of water samples also pose some difficulties

  5. Uranium: a basic evaluation

    International Nuclear Information System (INIS)

    Crull, A.W.

    1978-01-01

    All energy sources and technologies, including uranium and the nuclear industry, are needed to provide power. Public misunderstanding of the nature of uranium and how it works as a fuel may jeopardize nuclear energy as a major option. Basic chemical facts about uranium ore and uranium fuel technology are presented. Some of the major policy decisions that must be made include the enrichment, stockpiling, and pricing of uranium. Investigations and lawsuits pertaining to uranium markets are reviewed, and the point is made that oil companies will probably have to divest their non-oil energy activities. Recommendations for nuclear policies that have been made by the General Accounting Office are discussed briefly

  6. Uranium health physics

    International Nuclear Information System (INIS)

    1980-01-01

    This report contains the papers delivered at the Summer School on Uranium Health Physics held in Pretoria on the 14 and 15 April 1980. The following topics were discussed: uranium producton in South Africa; radiation physics; internal dosimetry and radiotoxicity of long-lived uranium isotopes; uranium monitoring; operational experience on uranium monitoring; dosimetry and radiotoxicity of inhaled radon daughters; occupational limits for inhalation of radon-222, radon-220 and their short-lived daughters; radon monitoring techniques; radon daughter dosimeters; operational experience on radon monitoring; and uranium mill tailings management

  7. Uranium: one utility's outlook

    International Nuclear Information System (INIS)

    Gass, C.B.

    1983-01-01

    The perspective of the Arizona Public Service Company (APS) on the uncertainty of uranium as a fuel supply is discussed. After summarizing the history of nuclear power and the uranium industries, a projection is made for the future uranium market. An uncrtain uranium market is attributed to various determining factors that include international politics, production costs, non-commercial government regulation, production-company stability, and questionable levels of uranium sales. APS offers its solutions regarding type of contract, choice of uranium producers, pricing mechanisms, and aids to the industry as a whole. 5 references, 10 figures, 1 table

  8. Microhardness and grain size of disordered nonstoichiometric titanium carbide

    International Nuclear Information System (INIS)

    Lipatnikov, V.N.; Zueva, L.V.; Gusev, A.I.

    1999-01-01

    Effect of the disordered nonstoichiometric titanium carbide on its microhardness and grain size is studied. It is established that decrease in defectiveness of carbon sublattice of disordered carbide is accompanied by microhardness growth and decrease in grain size. Possible causes of the TiC y microhardness anomalous behaviour in the area 0.8 ≤ y ≤ 0.9 connected with plastic deformation mechanism conditioned by peculiarities of the electron-energetic spectrum of nonstoichiometric carbide are discussed [ru

  9. Recovery of uranium from crude uranium tetrafluoride

    International Nuclear Information System (INIS)

    Ghosh, S.K.; Bellary, M.P.; Keni, V.S.

    1994-01-01

    An innovative process has been developed for recovery of uranium from crude uranium tetrafluoride cake. The process is based on direct dissolution of uranium tetrafluoride in nitric acid in presence of aluminium hydroxide and use of solvent extraction for removal of fluorides and other bulk impurities to make uranium amenable for refining. It is a simple process requiring minimum process step and has advantage of lesser plant corrosion. This process can be applied for processing of uranium tetrafluoride generated from various sources like uranium by-product during thorium recovery from thorium concentrate, first stage product of uranium recovery from phosphoric acid by OPPA process and off grade uranium tetrafluoride material. The paper describes the details of the process developed and demonstrated on bench and pilot scale and its subsequent modification arising out of bulky solid waste generation. The modified process uses a lower quantity of aluminium hydroxide by allowing a lower dissolution of uranium per cycle and recycles the undissolved material to the next cycle, maintaining the overall recovery at high level. This innovation has reduced the solid waste generated by a factor of four at the cost of a slightly larger dissolution vessel and its increased corrosion rate. (author)

  10. Recovery of uranium from crude uranium tetrafluoride

    Energy Technology Data Exchange (ETDEWEB)

    Ghosh, S K; Bellary, M P; Keni, V S [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    An innovative process has been developed for recovery of uranium from crude uranium tetrafluoride cake. The process is based on direct dissolution of uranium tetrafluoride in nitric acid in presence of aluminium hydroxide and use of solvent extraction for removal of fluorides and other bulk impurities to make uranium amenable for refining. It is a simple process requiring minimum process step and has advantage of lesser plant corrosion. This process can be applied for processing of uranium tetrafluoride generated from various sources like uranium by-product during thorium recovery from thorium concentrate, first stage product of uranium recovery from phosphoric acid by OPPA process and off grade uranium tetrafluoride material. The paper describes the details of the process developed and demonstrated on bench and pilot scale and its subsequent modification arising out of bulky solid waste generation. The modified process uses a lower quantity of aluminium hydroxide by allowing a lower dissolution of uranium per cycle and recycles the undissolved material to the next cycle, maintaining the overall recovery at high level. This innovation has reduced the solid waste generated by a factor of four at the cost of a slightly larger dissolution vessel and its increased corrosion rate. (author). 4 refs., 1 fig., 3 tabs.

  11. Carbides in Nodular Cast Iron with Cr and Mo

    Directory of Open Access Journals (Sweden)

    S. Pietrowski

    2007-07-01

    Full Text Available In these paper results of elements microsegregation in carbidic nodular cast iron have been presented. A cooling rate in the centre of the cross-section and on the surface of casting and change of moulding sand temperature during casting crystallization and its self-cooling have been investigated. TDA curves have been registered. The linear distribution of elements concentration in an eutectic grain, primary and secondary carbides have been made. It was found, that there are two kinds of carbides: Cr and Mo enriched. A probable composition of primary and secondary carbides have been presented.

  12. Silicon Carbide Corrugated Mirrors for Space Telescopes, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Trex Enterprises Corporation (Trex) proposes technology development to manufacture monolithic, lightweight silicon carbide corrugated mirrors (SCCM) suitable for...

  13. Colorado's prospectus on uranium milling

    International Nuclear Information System (INIS)

    Hazle, A.J.; Franz, G.A.; Gamewell, R.

    1982-01-01

    The first part of this paper will discuss Colorado's control of uranium mill tailings under Titles I and II of the Uranium Mill Tailings Radiation Control Act of 1978. Colorado has a legacy of nine inactive mill sites requiring reclamation under Title I, and two presently active plus a number of new mill proposals which must be regulated in accordance with Title II. Past failures in siting and control on the part of federal jurisdictions have left the state with a heavy legacy requiring extensive effort to address impacts to the state's environment and population. The second part of this paper will discuss the remedial action programme authorized under Public Law 92-314 for Mesa Country, where lack of federal control led to the dispersal of several hundred thousand tons of uranium mill tailings on thousands of properties, including hundreds of homes, schools and other structures. Successful completion of the State efforts under both programmes will depend on a high level of funding and on the maintenance of adequate regulatory standards. (author)

  14. Preparation of silicon carbide nanowires via a rapid heating process

    International Nuclear Information System (INIS)

    Li Xintong; Chen Xiaohong; Song Huaihe

    2011-01-01

    Silicon carbide (SiC) nanowires were fabricated in a large quantity by a rapid heating carbothermal reduction of a novel resorcinol-formaldehyde (RF)/SiO 2 hybrid aerogel in this study. SiC nanowires were grown at 1500 deg. C for 2 h in an argon atmosphere without any catalyst via vapor-solid (V-S) process. The β-SiC nanowires were characterized by field-emission scanning electron microscope (FE-SEM), X-ray diffraction (XRD), transmission electron microscope (TEM), high-resolution transmission electron microscope (HRTEM) equipped with energy dispersive X-ray (EDX) facility, Fourier transformed infrared spectroscopy (FTIR), and thermogravimetric analysis (TGA). The analysis results show that the aspect ratio of the SiC nanowires via the rapid heating process is much larger than that of the sample produced via gradual heating process. The SiC nanowires are single crystalline β-SiC phase with diameters of about 20-80 nm and lengths of about several tens of micrometers, growing along the [1 1 1] direction with a fringe spacing of 0.25 nm. The role of the interpenetrating network of RF/SiO 2 hybrid aerogel in the carbothermal reduction was discussed and the possible growth mechanism of the nanowires is analyzed.

  15. Porosity and wear resistance of flame sprayed tungsten carbide coatings

    Science.gov (United States)

    Winarto, Winarto; Sofyan, Nofrijon; Rooscote, Didi

    2017-06-01

    Thermal-sprayed coatings offer practical and economical solutions for corrosion and wear protection of components or tools. To improve the coating properties, heat treatment such as preheat is applied. The selection of coating and substrate materials is a key factor in improving the quality of the coating morphology after the heat treatment. This paper presents the experimental results regarding the effect of preheat temperatures, i.e. 200°C, 300°C and 400°C, on porosity and wear resistance of tungsten carbide (WC) coating sprayed by flame thermal coating. The powders and coatings morphology were analyzed by a Field Emission Scanning Electron Microscope equipped with Energy Dispersive Spectrometry (FE-SEM/EDS), whereas the phase identification was performed by X-Ray diffraction technique (XRD). In order to evaluate the quality of the flame spray obtained coatings, the porosity, micro-hardness and wear rate of the specimens was determined. The results showed that WC coating gives a higher surface hardness from 1391 HVN up to 1541 HVN compared to that of the non-coating. Moreover, the wear rate increased from 0.072 mm3/min. to 0.082 mm3/min. when preheat temperature was increased. Preheat on H13 steel substrate can reduce the percentage of porosity level from 10.24 % to 3.94% on the thermal spray coatings.

  16. Solid phase epitaxy of amorphous silicon carbide: Ion fluence dependence

    International Nuclear Information System (INIS)

    Bae, I.-T.; Ishimaru, Manabu; Hirotsu, Yoshihiko; Sickafus, Kurt E.

    2004-01-01

    We have investigated the effect of radiation damage and impurity concentration on solid phase epitaxial growth of amorphous silicon carbide (SiC) as well as microstructures of recrystallized layer using transmission electron microscopy. Single crystals of 6H-SiC with (0001) orientation were irradiated with 150 keV Xe ions to fluences of 10 15 and 10 16 /cm 2 , followed by annealing at 890 deg. C. Full epitaxial recrystallization took place in a specimen implanted with 10 15 Xe ions, while retardation of recrystallization was observed in a specimen implanted with 10 16 /cm 2 Xe ions. Atomic pair-distribution function analyses and energy dispersive x-ray spectroscopy results suggested that the retardation of recrystallization of the 10 16 Xe/cm 2 implanted sample is attributed to the difference in amorphous structures between the 10 15 and 10 16 Xe/cm 2 implanted samples, i.e., more chemically disordered atomistic structure and higher Xe impurity concentration in the 10 16 Xe/cm 2 implanted sample

  17. Synthesis of silicon carbide by carbothermal reduction of silica

    International Nuclear Information System (INIS)

    Abel, Joao Luis

    2009-01-01

    The production of silicon carbide (SiC) in an industrial scale still by carbothermal reduction of silica. This study aims to identify, in a comparative way, among the common reducers like petroleum coke, carbon black, charcoal and graphite the carbothermal reduction of silica from the peat. It is shown, that the peat, also occurs in nature together with high purity silica sand deposits, where the proximity of raw materials and their quality are key elements that determine the type, purity and cost of production of SiC. Tests were running from samples produced in the electric resistance furnace with controlled atmosphere at temperatures of 1550 degree C, 1600 degree C and 1650 degree C, both the precursors and products of reaction of carbothermal reduction were characterized by applying techniques of X-ray diffraction, scanning electron microscopy (SEM) and Energy-Dispersive X-ray analysis Spectroscopy (EDS). The results showed the formation of SiC for all common reducers, as well as for peat, but it was not possible to realize clearly the difference between them, being necessary, specific tests. (author)

  18. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  19. Uranium production

    International Nuclear Information System (INIS)

    Jones, J.Q.

    1981-01-01

    The domestic uranium industry is in a state of stagflation. Costs continue to rise while the market for the product remains stagnant. During the last 12 months, curtailments and closures of mines and mills have eliminated over 5000 jobs in the industry, plus many more in those industries that furnish supplies and services. By January 1982, operations at four mills and the mines that furnish them ore will have been terminated. Other closures may follow, depending on cost trends, duration of current contracts, the degree to which mills have been amortized, the feasibility of placing mines on standby, the grade of the ore, and many other factors. Open-pit mines can be placed on standby without much difficulty, other than the possible cost of restoration before all the ore has been removed. There are a few small, dry, underground mines that could be mothballed; however, the major underground producers are wet sandstone mines that in most cases could not be reopened after a prolonged shutdown; mills can be mothballed for several years. Figure 8 shows the location of all the production centers in operation, as well as those that have operated or are on standby. Table 1 lists the same production centers plus those that have been deferred, showing nominal capacity of conventional mills in tons of ore per calendar day, and the industry production rate for those mills as of October 1, 1981

  20. Studies on the influence of surface pre-treatments on electroless copper coating of boron carbide particles

    International Nuclear Information System (INIS)

    Deepa, J.P.; Resmi, V.G.; Rajan, T.P.D.; Pavithran, C.; Pai, B.C.

    2011-01-01

    Boron carbide is one of the hard ceramic particles which find application as structural materials and neutron shielding material due to its high neutron capture cross section. Copper coating on boron carbide particle is essential for the synthesis of metal-ceramic composites with enhanced sinterability and dispersibility. Surface characteristics of the substrate and the coating parameters play a foremost role in the formation of effective electroless coating. The effect of surface pre-treatment conditions and pH on electroless copper coating of boron carbide particles has been studied. Surface pre-treatement of B 4 C when compared to acid treated and alkali treated particles were carried out. Uniform copper coating was observed at pH 12 in alkali treated particles when compared to others due to the effective removal of inevitable impurities during the production and processing of commercially available B 4 C. A threshold pH 11 was required for initiation of copper coating on boron carbide particles. The growth pattern of the copper coating also varies depending on the surface conditions from acicular to spherical morphology.

  1. Construction of reduced graphene oxide supported molybdenum carbides composite electrode as high-performance anode materials for lithium ion batteries

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Minghua; Zhang, Jiawei [Key Laboratory of Engineering Dielectric and Applications (Ministry of Education), and School of Applied Science, Harbin University of Science and Technology, Harbin 150080 (China); Chen, Qingguo, E-mail: qgchen@263.net [Key Laboratory of Engineering Dielectric and Applications (Ministry of Education), and School of Applied Science, Harbin University of Science and Technology, Harbin 150080 (China); Qi, Meili [Key Laboratory of Engineering Dielectric and Applications (Ministry of Education), and School of Applied Science, Harbin University of Science and Technology, Harbin 150080 (China); Xia, Xinhui, E-mail: helloxxh@zju.edu.cn [State Key Laboratory of Silicon Materials, Key Laboratory of Advanced Materials and Applications for Batteries of Zhejiang Province, and School of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China)

    2016-01-15

    Highlights: • Reduced graphene oxide supported molybdenum carbides are prepared by two-step strategy. • A unique sheet-on-sheet integrated nanostructure is favorable for fast ion/electron transfer. • The integrated electrode shows excellent Li ion storage performance. - Abstract: Metal carbides are emerging as promising anodes for advanced lithium ion batteries (LIBs). Herein we report reduced graphene oxide (RGO) supported molybdenum carbides (Mo{sub 2}C) integrated electrode by the combination of solution and carbothermal methods. In the designed integrated electrode, Mo{sub 2}C nanoparticles are uniformly dispersed among graphene nanosheets, forming a unique sheet-on-sheet integrated nanostructure. As anode of LIBs, the as-prepared Mo{sub 2}C-RGO integrated electrode exhibits noticeable electrochemical performances with a high reversible capacity of 850 mAh g{sup −1} at 100 mA g{sup −1}, and 456 mAh g{sup −1} at 1000 mA g{sup −1}, respectively. Moreover, the Mo{sub 2}C-RGO integrated electrode shows excellent cycling life with a capacity of ∼98.6 % at 1000 mA g{sup −1} after 400 cycles. Our research may pave the way for construction of high-performance metal carbides anodes of LIBs.

  2. Stress corrosion cracking of stainless steel under deaerated high-temperature water. Influence of grain boundary carbide precipitation

    International Nuclear Information System (INIS)

    Yamada, Takuyo; Terachi, Takumi; Arioka, Koji

    2006-01-01

    In order to evaluate the influence of grain boundary carbide on IGSCC susceptibility, crack growth rate tests were performed under deaerated and 0.3 ppm hydrogenated pure water environments at 320degC using half-inch compact tension specimens. To investigate various grain boundary carbide conditions, three kinds of SUS316 - non-sensitized, sensitized at 650degC for 1 hour or 48 hours - were prepared. To examine the influence of grain boundary carbide, the grain boundary conditions of those materials were investigated by transmission electron microscopy and energy dispersive x-ray spectroscopy. As a result, (1) IGSCC crack growth was observed on non sensitized and cold worked SUS316 under deaerated and 0.3 ppm hydrogenated water environments at 320degC; (2) Any trace of IGSCC crack growth was not observed on sensitized at 650degC for 48 hours and cold worked SUS316 under the same water environments; (3) The SUS316 sensitized at 650degC for 48 hours showed extensive M 23 C 6 precipitation as well as Cr depletion at grain boundaries. These differences in IGSCC crack growth rate indicate that grain boundary carbide has the beneficial effect of improving IGSCC susceptibility, at least under deaerated and 0.3 ppm hydrogenated water environments, despite chromium depletion at the grain boundary. (author)

  3. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Known uranium deposits and the companies involved in uranium mining and exploration in Australia are listed. The status of the development of the deposits is outlined and reasons for delays to mining are given

  4. Uranium Processing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — An integral part of Y‑12's transformation efforts and a key component of the National Nuclear Security Administration's Uranium Center of Excellence, the Uranium...

  5. Uranium in Niger

    International Nuclear Information System (INIS)

    Gabelmann, E.

    1978-03-01

    This document presents government policy in the enhancement of uranium resources, existing mining companies and their productions, exploitation projects and economical outcome related to the uranium mining and auxiliary activities [fr

  6. Price of military uranium

    International Nuclear Information System (INIS)

    Klimenko, A.V.

    1998-01-01

    The theoretical results about optimum strategy of use of military uranium confirmed by systems approach accounts are received. The numerical value of the system approach price of the highly enriched military uranium also is given

  7. Uranium market and resources

    International Nuclear Information System (INIS)

    Capus, G.; Arnold, Th.

    2004-01-01

    The controversy about the extend of the uranium resources worldwide is still important, this article sheds some light on this topic. Every 2 years IAEA and NEA (nuclear energy agency) edit an inventory of uranium resources as reported by contributing countries. It appears that about 4.6 millions tons of uranium are available at a recovery cost less than 130 dollars per kg of uranium and a total of 14 millions tons of uranium can be assessed when including all existing or supposed resources. In fact there is enough uranium to sustain a moderate growth of the park of nuclear reactors during next decades and it is highly likely that the volume of uranium resources can allow a more aggressive development of nuclear energy. It is recalled that a broad use of the validated breeder technology can stretch the durability of uranium resources by a factor 50. (A.C.)

  8. Uranium from phosphate ores

    International Nuclear Information System (INIS)

    Hurst, F.J.

    1983-01-01

    The following topics are described briefly: the way phosphate fertilizers are made; how uranium is recovered in the phosphate industry; and how to detect covert uranium recovery operations in a phsophate plant

  9. Industrial realities: Uranium

    International Nuclear Information System (INIS)

    Thiron, H.

    1990-01-01

    In this special issue are examined ores and metals in France and in the world for 1988. The chapter on uranium gives statistical data on the uranium market: Demand, production, prices and reserves [fr

  10. Brazilian uranium deposits

    International Nuclear Information System (INIS)

    Santos, L.C.S. dos.

    1985-01-01

    Estimatives of uranium reserves carried out in Figueira, Itataia, Lagoa Real and Espinharas, in Brazil are presented. The samples testing allowed to know geological structures, and the characteristics of uranium mineralization. (M.C.F.) [pt

  11. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The mining of uranium in Australia is criticised in relation to it's environmental impact, economics and effects on mine workers and Aborigines. A brief report is given on each of the operating and proposed uranium mines in Australia

  12. Depleted uranium and the Gulf War syndrome

    International Nuclear Information System (INIS)

    1999-01-01

    Some military personnel involved in the 1991Gulf War have complained of continuing stress-like symptoms for which no obvious cause has been found. These symptoms have at times been attributed to the use of depleted uranium (DU) in shell casings which are believed to have caused toxic effects. Depleted uranium is natural uranium which is depleted in the rarer U-235 isotope. It is a heavy metal and in common with other heavy metals is chemically toxic. It is also slightly radioactive and could give rise to a radiological hazard if dispersed in finely divided form so that it was inhaled. In response to concerns, the possible effects of DU have been extensively studied along with other possible contributors to G ulf War sickness . This article looks at the results of some of the research that has been done on DU. (author)

  13. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    Western world requirements for uranium based on increasing energy consumption and a changing energy mix, will warrant the development of Australia's resources. By 1985 Australian mines could be producing 9500 tonnes of uranium oxide yearly and by 1995 the export value from uranium could reach that from wool. In terms of benefit to the community the economic rewards are considerable but, in terms of providing energy to the world, Australias uranium is vital

  14. Radiation damage of uranium

    International Nuclear Information System (INIS)

    Lazarevic, Dj.

    1966-11-01

    Study of radiation damage covered the following: Kinetics of electric resistance of uranium and uranium alloy with 1% of molybdenum dependent on the second phase and burnup rate; Study of gas precipitation and diffusion of bubbles by transmission electron microscopy; Numerical analysis of the influence of defects distribution and concentration on the rare gas precipitation in uranium; study of thermal sedimentation of uranium alloy with molybdenum; diffusion of rare gas in metal by gas chromatography method

  15. Bicarbonate leaching of uranium

    International Nuclear Information System (INIS)

    Mason, C.

    1998-01-01

    The alkaline leach process for extracting uranium from uranium ores is reviewed. This process is dependent on the chemistry of uranium and so is independent on the type of mining system (conventional, heap or in-situ) used. Particular reference is made to the geochemical conditions at Crownpoint. Some supporting data from studies using alkaline leach for remediation of uranium-contaminated sites is presented

  16. Bicarbonate leaching of uranium

    Energy Technology Data Exchange (ETDEWEB)

    Mason, C.

    1998-12-31

    The alkaline leach process for extracting uranium from uranium ores is reviewed. This process is dependent on the chemistry of uranium and so is independent on the type of mining system (conventional, heap or in-situ) used. Particular reference is made to the geochemical conditions at Crownpoint. Some supporting data from studies using alkaline leach for remediation of uranium-contaminated sites is presented.

  17. Uranium in fossil bones

    International Nuclear Information System (INIS)

    Koul, S.L.

    1978-01-01

    An attempt has been made to determine the uranium content and thus the age of certain fossil bones Haritalyangarh (Himachal Pradesh), India. The results indicate that bones rich in apatite are also rich in uranium, and that the radioactivity is due to radionuclides in the uranium series. The larger animals apparently have a higher concentration of uranium than the small. The dating of a fossil jaw (elephant) places it in the Pleistocene. (Auth.)

  18. Method for converting uranium oxides to uranium metal

    International Nuclear Information System (INIS)

    Duerksen, W.K.

    1988-01-01

    A method for converting uranium oxide to uranium metal is described comprising the steps of heating uranium oxide in the presence of a reducing agent to a temperature sufficient to reduce the uranium oxide to uranium metal and form a heterogeneous mixture of a uranium metal product and oxide by-products, heating the mixture in a hydrogen atmosphere at a temperature sufficient to convert uranium metal in the mixture to uranium hydride, cooling the resulting uranium hydride-containing mixture to a temperature sufficient to produce a ferromagnetic transition in the uranium hydride, magnetically separating the cooled uranium hydride from the mixture, and thereafter heating the separated uranium hydride in an inert atmosphere to a temperature sufficient to convert the uranium hydride to uranium metal

  19. Dispersion strengthening

    International Nuclear Information System (INIS)

    Scattergood, R.O.; Das, E.S.P.

    1976-01-01

    Using digital computer-based methods, models for dispersion strengthening can now be developed which take into account many of the important effects that have been neglected in the past. In particular, the self interaction of a dislocation can be treated, and a computer simulation method was developed to determine the flow stress of a random distribution of circular, impenetrable obstacles, taking into account all such interactions. The flow stress values depended on the obstacle sizes and spacings, over and above the usual 1/L dependence where L is the average obstacle spacing. From an analysis of the results, it was found that the main effects of the self interactions can be captured in a line tension analogue in which the obstacles appear to be penetrable

  20. RADIONUCLIDE DISPERSION RATES BY AEOLIAN, FLUVIAL, AND POROUS MEDIA TRANSPORT

    International Nuclear Information System (INIS)

    Walton, J.; Goodell, P.; Brashears, C.; French, D.; Kelts, A.

    2005-01-01

    Radionuclide transport was measured from high grade uranium ore boulders near the Nopal I Site, Chihuahua, Mexico. High grade uranium ore boulders were left behind after removal of a uranium ore stockpile at the Prior High Grade Stockpile (PHGS). During the 25 years when the boulder was present, radionuclides were released and transported by sheetflow during precipitation events, wind blown resuspension, and infiltration into the unsaturated zone. In this study, one of the boulders was removed, followed by grid sampling of the surrounding area. Measured gamma radiation levels in three dimensions were used to derive separate dispersion rates by the three transport mechanisms

  1. RADIONUCLIDE DISPERSION RATES BY AEOLIAN, FLUVIAL, AND POROUS MEDIA TRANSPORT

    Energy Technology Data Exchange (ETDEWEB)

    J. Walton; P. Goodell; C. Brashears; D. French; A. Kelts

    2005-07-11

    Radionuclide transport was measured from high grade uranium ore boulders near the Nopal I Site, Chihuahua, Mexico. High grade uranium ore boulders were left behind after removal of a uranium ore stockpile at the Prior High Grade Stockpile (PHGS). During the 25 years when the boulder was present, radionuclides were released and transported by sheetflow during precipitation events, wind blown resuspension, and infiltration into the unsaturated zone. In this study, one of the boulders was removed, followed by grid sampling of the surrounding area. Measured gamma radiation levels in three dimensions were used to derive separate dispersion rates by the three transport mechanisms.

  2. Yalina booster subcritical assembly performance with low enriched uranium fuel

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gohar, Yousry

    2011-01-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  3. Yalina booster subcritical assembly performance with low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

    2011-07-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  4. Thermodynamic studies of thorium carbide fuel preparation and fuel-clad comptability

    International Nuclear Information System (INIS)

    Besmann, T.M.; Beahm, E.C.

    1979-01-01

    The carbothermic reduction of thorium and uranium-thorium dioxide to monocarbide has been assessed. Equilibrium calculations have yielded Th-C-O and U-Th-C-O phase equilibria and (CO) pressures generated during reduction. The (CO) pressures were found to be at least five orders of magnitude greater than any of the other 15 gaseous species considered. This confirms that the monocarbide can successfully be prepared by carbothermic reduction. The chemical compatibility of thorium carbides with the Cr-Fe-Ni content of clad alloys has been thermodynamically avaluated. Solid solutions of 5 > and 5 > and of 7 C 3 > and 7 C 3 > were the principal reaction products. The Cr-Fe-Ni content of 316 stainless steel showed much less reaction product than that for any of the other six alloys considered. (orig.) [de

  5. Preliminary design study for a carbide LEU-nuclear thermal rocket

    International Nuclear Information System (INIS)

    Venneri, P.F.; Kim, Y.

    2014-01-01

    Nuclear space propulsion is a requirement for the successful exploration of the solar system. It offers the possibility of having both a high specific impulse and a relatively high thrust, allowing rapid transit times with a minimum usage of fuel. This paper proposes a nuclear thermal rocket design based on heritage NERVA rockets that makes use of Low Enriched Uranium (LEU) fuel. The Carbide LEU Nuclear Thermal Rocket (C-LEU-NTR) is designed to fulfill the rocket requirements as set forth in the NASA 2009 Mars Mission Design Reference Architecture 5.0, that is provide 25,000 lbf of thrust, operate at full power condition for at least two hours, and have a specific impulse close to 900 s. The neutronics analysis was done using MCNP5 with the ENDF/B-VII.1 neutron library. The thermal hydraulic calculations and size optimization were completed with a finite difference code being developed at the Center for Space Nuclear Research. (authors)

  6. Interaction of noble-metal fission products with pyrolytic silicon carbide

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  7. Thermal behavior analysis of U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

  8. Thermal behavior analysis of U-Mo/Al dispersion fuel

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu

    2004-01-01

    According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

  9. Microbial accumulation of uranium

    International Nuclear Information System (INIS)

    Zhang Wei; Dong Faqin; Dai Qunwei

    2005-01-01

    The mechanism of microbial accumulation of uranium and the effects of some factors (including pH, initial uranium concentration, pretreatment of bacteria, and so on) on microbial accumulation of uranium are discussed briefly. The research direction and application prospect are presented. (authors)

  10. Uranium energy dependence

    International Nuclear Information System (INIS)

    Erkes, P.

    1981-06-01

    Uranium supply and demand as projected by the Uranium Institute is discussed. It is concluded that for the industrialized countries, maximum energy independence is a necessity. Hence it is necessary to achieve assurance of supply for uranium used in thermal power reactors in current programs and eventually to move towards breeders

  11. Australian uranium today

    International Nuclear Information System (INIS)

    Fisk, B.

    1978-01-01

    The subject is covered in sections, entitled: Australia's resources; Northern Territory uranium in perspective; the government's decision [on August 25, 1977, that there should be further development of uranium under strictly controlled conditions]; Government legislation; outlook [for the Australian uranium mining industry]. (U.K.)

  12. Uranium resources, 1983

    International Nuclear Information System (INIS)

    1983-01-01

    The specific character of uranium as energy resources, the history of development of uranium resources, the production and reserve of uranium in the world, the prospect regarding the demand and supply of uranium, Japanese activity of exploring uranium resources in foreign countries and the state of development of uranium resources in various countries are reported. The formation of uranium deposits, the classification of uranium deposits and the reserve quantity of each type are described. As the geological environment of uranium deposits, there are six types, that is, quartz medium gravel conglomerate deposit, the deposit related to the unconformity in Proterozoic era, the dissemination type magma deposit, pegmatite deposit and contact deposit in igneaus rocks and metamorphic rocks, vein deposit, sandstone type deposit and the other types of deposit. The main features of respective types are explained. The most important uranium resources in Japan are those in the Tertiary formations, and most of the found reserve belongs to this type. The geological features, the state of yield and the scale of the deposits in Ningyotoge, Tono and Kanmon Mesozoic formation are reported. Uranium minerals, the promising districts in the world, and the matters related to the exploration and mining of uranium are described. (Kako, I.)

  13. Recycling of reprocessed uranium

    International Nuclear Information System (INIS)

    Randl, R.P.

    1987-01-01

    Since nuclear power was first exploited in the Federal Republic of Germany, the philosophy underlying the strategy of the nuclear fuel cycle has been to make optimum use of the resource potential of recovered uranium and plutonium within a closed fuel cycle. Apart from the weighty argument of reprocessing being an important step in the treatment and disposal of radioactive wastes, permitting their optimum ecological conditioning after the reprocessing step and subsequent storage underground, another argument that, no doubt, carried weight was the possibility of reducing the demand of power plants for natural uranium. In recent years, strategies of recycling have emerged for reprocessed uranium. If that energy potential, too, is to be exploited by thermal recycling, it is appropriate to choose a slightly different method of recycling from the one for plutonium. While the first generation of reprocessed uranium fuel recycled in the reactor cuts down natural uranium requirement by some 15%, the recycling of a second generation of reprocessed, once more enriched uranium fuel helps only to save a further three per cent of natural uranium. Uranium of the second generation already carries uranium-232 isotope, causing production disturbances, and uranium-236 isotope, causing disturbances of the neutron balance in the reactor, in such amounts as to make further fabrication of uranium fuel elements inexpedient, even after mixing with natural uranium feed. (orig./UA) [de

  14. Atmospheric corrosion of uranium-carbon alloys

    International Nuclear Information System (INIS)

    Rousset, P.; Accary, A.

    1965-01-01

    The authors study the corrosion of uranium-carbon alloys having compositions close to that of the mono-carbide; they show that the extent of the observed corrosion effects increases with the water vapour content of the surrounding gas and they conclude that the atmospheric corrosion of these alloys is due essentially to the humidity of the air, the effect of the oxygen being very slight at room temperature. They show that the optimum conditions for preserving U-C alloys are either a vacuum or a perfectly dry argon atmosphere. The authors have also established that the type of corrosion involved is a corrosion which 'cracks under stress' and is transgranular (it can also be intergranular in the case of sub-stoichiometric alloys). They propose, finally, two hypotheses for explaining this mechanism, one of which is illustrated by the existence, at the fissure interface, of corrosion products which can play the role of 'corners' in the mono-carbide grains. (authors) [fr

  15. Uranium series disequilibrium studies at the Broubster analogue site

    International Nuclear Information System (INIS)

    Longworth, G.; Ivanovich, M.; Wilkins, M.A.

    1990-11-01

    Uranium series measurements at a natural analogue site at Broubster, Caithness have been used to investigate radionuclide migration over periods ranging from several hundred to 10 6 years. The measured values for the uranium concentration and activity values 234 U/ 238 U and 230 Th/ 234 U indicate that the geochemical system is more complicated than that originally proposed of uranium dispersion and water transport into a peat bog. There appears to be little thorium mobility although there is evidence for an appreciable fraction of thorium on the colloidal phase. (author)

  16. Uranium series disequilibrium studies at the Broubster analogue site

    International Nuclear Information System (INIS)

    Longworth, G.; Ivanovich, M.; Wilkins, M.A.

    1989-09-01

    Uranium series measurements at a natural analogue site at Broubster, Caithness have been used to investigate radionuclide migration over a period of several hundred to 10 6 years. The measured values for the uranium concentration and activity ratios 234 U/ 238 U and 230 Th/ 234 U indicate that the geochemical system is more complicated than that originally proposed of uranium dispersion and water transport into a peat bog. There appears to be little thorium mobility although there is evidence for an appreciable fraction of thorium on the colloidal phases. (author)

  17. Fracture and Residual Characterization of Tungsten Carbide Cobalt Coatings on High Strength Steel

    National Research Council Canada - National Science Library

    Parker, Donald S

    2003-01-01

    Tungsten carbide cobalt coatings applied via high velocity oxygen fuel thermal spray deposition are essentially anisotropic composite structures with aggregates of tungsten carbide particles bonded...

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

  20. Reactor irradiation effect on the physical-mechanical properties of zirconium carbides and niobium carbides

    International Nuclear Information System (INIS)

    Andrievskij, R.A.; Vlasov, K.P.; Shevchenko, A.S.; Lanin, A.G.; Pritchin, S.A.; Klyushin, V.V.; Kurushin, S.P.; Maskaev, A.S.

    1978-01-01

    A study has been made of the effect of the reactor radiation by a flux of neutrons 1.5x10 20 n/cm 2 (E>=1 meV) at radiation temperatures of 150 and 1100 deg C on the physico-mechanical properties of carbides of zirconium and niobium and their equimolar hard solution. A difference has been discovered in the behaviour of the indicated carbides under the effect of radiation. Under the investigated conditions of radiation the density of zirconium carbide is being decreased, while in the niobium carbide no actual volumetric changes occur. The increase of the lattice period in ZrC is more significant than in NbC. The electric resistance of ZrC is also changed more significantly than in the case of NbC, while for the microhardness a reverse relationship is observed. Strength and elasticity modulus change insignificantly in both cases. Resistance to crack formation shows a higher reduction for ZrC than for NbC, while the thermal strength shows an approximately similar increase. The equimolar hard solution of ZrC and NbC behaves to great extent similar to ZrC, although the change in electric resistance reminds of NbC while thermal strength changes differently. The study of the microstructure of the specimens has shown that radiation causes a large number of etching patterns-dislocations in NbC which are almost absent in ZrC

  1. Reuse of ammonium fluoride generated in the uranium hexafluoride conversion

    International Nuclear Information System (INIS)

    Silva Neto, J.B.; Carvalho, E.F. Urano de; Durazzo, M.; Riella, H.G

    2010-01-01

    The Nuclear Fuel Centre of IPEN / CNEN - SP develops and manufactures dispersion fuel with high uranium concentration to meet the demand of the IEA-R1 reactor and future research reactors planned to be constructed in Brazil. The fuel uses uranium silicide (U 3 Si 2 ) dispersed in aluminum. For producing the fuel, the processes for uranium hexafluoride (UF 6 ) conversion consist in obtaining U 3 Si 2 and / or U 3 O 8 through the preparation of intermediate compounds, among them ammonium uranyl carbonate - AUC, ammonium diuranate - DUA and uranium tetrafluoride - UF 4 . This work describes a procedure for preparing uranium tetrafluoride by a dry route using as raw material the filtrate generated when producing routinely ammonium uranyl carbonate. The filtrate consists primarily of a solution containing high concentrations of ammonium (NH 4 + ), fluoride (F - ), carbonate (CO 3 -- ) and low concentrations of uranium. The procedure is basically the recovery of NH 4 F and uranium, as UF 4 , through the crystallization of ammonium bifluoride (NH 4 HF 2 ) and, in a later step, the addition of UO 2 , occurring fluoridation and decomposition. The UF 4 obtained is further diluted in the UF 4 produced routinely at IPEN / CNEN-SP by a wet route process. (author)

  2. Manufacturing Experience for Oxide Dispersion Strengthened Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, Wendy D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Henager, Charles H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Montgomery, Robert O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Mark T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, Ryan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-22

    This report documents the results of the development and the manufacturing experience gained at the Pacific Northwest National Laboratories (PNNL) while working with the oxide dispersion strengthened (ODS) materials MA 956, 14YWT, and 9YWT. The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. ODS materials have the potential to provide improved performance for the U-Mo concept.

  3. Investigation of the fire at the Uranium Enrichment Laboratory. Analysis of samples and pressurization experiment/analysis of container

    International Nuclear Information System (INIS)

    Akabori, Mitsuo; Minato, Kazuo; Watanabe, Kazuo

    1998-05-01

    To investigate the cause of the fire at the Uranium Enrichment Laboratory of the Tokai Research Establishment on November 20, 1997, samples of uranium metal waste and scattered residues were analyzed. At the same time the container lid that had been blown off was closely inspected, and the pressurization effects of the container were tested and analyzed. It was found that 1) the uranium metal waste mainly consisted of uranium metal, carbides and oxides, whose relative amounts were dependent on the particle size, 2) the uranium metal waste hydrolyzed to produce combustible gases such as methane and hydrogen, and 3) the lid of the outer container could be blown off by an explosive rise of the inner pressure caused by combustion of inflammable gas mixture. (author)

  4. Nondestructive neutron activation analysis of silicon carbide

    Energy Technology Data Exchange (ETDEWEB)

    Vandergraaf, T. T.; Wikjord, A. G.

    1973-10-15

    Instrumentel neutron activation analysis was used to determine trace constituents in silicon carbide. Four commercial powders of different origin, an NBS reference material, and a single crystal were characterized. A total of 36 activation species were identified nondestructively by high resolution gamma spectrometry; quantitative results are given for 12 of the more predominant elements. The limitations of the method for certain elements are discussed. Consideration is given to the depression of the neutron flux by impurities with large neutron absorption cross sections. Radiation fields from the various specimens were estimated assuming all radionuclides have reached their saturation activities. (auth)

  5. Crack propagation and fracture in silicon carbide

    International Nuclear Information System (INIS)

    Evans, A.G.; Lange, F.F.

    1975-01-01

    Fracture mechanics and strength studies performed on two silicon carbides - a hot-pressed material (with alumina) and a sintered material (with boron) - have shown that both materials exhibit slow crack growth at room temperature in water, but only the hot-pressed material exhibits significant high temperature slow crack growth (1000 to 1400 0 C). A good correlation of the observed fracture behaviour with the crack growth predicted from the fracture mechanics parameters shows that effective failure predictions for this material can be achieved using macro-fracture mechanics data. (author)

  6. An improved method for preparing silicon carbide

    International Nuclear Information System (INIS)

    Baney, R.H.

    1980-01-01

    A desired shape is formed from a polysilane and the shape is heated in an inert atmosphere or under vacuum to 1150 to 1600 0 C until the polysilane is converted to silicon carbide. The polysilane contains from 0 to 60 mole percent of (CH 3 ) 2 Si units and from 40 to 100 mole percent of CH 3 Si units. The remaining bonds on silicon are attached to another silicon atom or to a chlorine or bromine atom, such that the polysilane contains from 10 to 43 weight percent of hydrolyzable chlorine or from 21 to 63 weight percent of hydrolyzable bromine. (author)

  7. Hardness of carbides, nitrides, and borides

    International Nuclear Information System (INIS)

    Schroeter, W.

    1981-01-01

    Intermetallic compounds of metals with non-metals such as C, N, and B show different hardness. Wagner's interaction parameter characterizes manner and extent of the interaction between the atoms of the substance dissolved and the additional elements in metallic mixed phases. An attempt has been made to correlate the hardness of carbides, nitrides, and borides (data taken from literature) with certain interaction parameters and associated thermodynamic quantities (ΔH, ΔG). For some metals of periods 4, 5, and 6 corresponding relations were found between microhardness, interaction parameters, heat of formation, and atomic number

  8. The chemical vapor deposition of zirconium carbide onto ceramic substrates

    International Nuclear Information System (INIS)

    Glass A, John Jr.; Palmisiano, Nick Jr.; Welsh R, Edward

    1999-01-01

    Zirconium carbide is an attractive ceramic material due to its unique properties such as high melting point, good thermal conductivity, and chemical resistance. The controlled preparation of zirconium carbide films of superstoichiometric, stoichiometric, and substoichiometric compositions has been achieved utilizing zirconium tetrachloride and methane precursor gases in an atmospheric pressure high temperature chemical vapor deposition system

  9. Influence of nanometric silicon carbide on phenolic resin composites ...

    Indian Academy of Sciences (India)

    Abstract. This paper presents a preliminary study on obtaining and characterization of phenolic resin-based com- posites modified with nanometric silicon carbide. The nanocomposites were prepared by incorporating nanometric silicon carbide (nSiC) into phenolic resin at 0.5, 1 and 2 wt% contents using ultrasonication to ...

  10. Determination of free and combined carbon in boron carbide

    International Nuclear Information System (INIS)

    Shankaran, P.S.; Kulkarni, Amit S.; Pandey, K.L.; Ramanjaneyulu, P.S.; Yadav, C.S.; Sayi, Y.S.; Ramakumar, K.L.

    2009-01-01

    A simple, sensitive and fast method for the determination of free and combined carbon in boron carbide samples, based on combustion in presence of oxygen at different temperatures, has been developed. Method has been standardized by analyzing mixture of two different boron carbide samples. Error associated with the method in the determination of free carbon is less than 5%. (author)

  11. Stress in tungsten carbide-diamond like carbon multilayer coatings

    NARCIS (Netherlands)

    Pujada, B.R.; Tichelaar, F.D.; Janssen, G.C.A.M.

    2007-01-01

    Tungsten carbide-diamond like carbon (WC-DLC) multilayer coatings have been prepared by sputter deposition from a tungsten-carbide target and periodic switching on and off of the reactive acetylene gas flow. The stress in the resulting WC-DLC multilayers has been studied by substrate curvature.

  12. Process for the preparation of fine grain metal carbide powders

    International Nuclear Information System (INIS)

    Gortsema, F.P.

    1976-01-01

    Fine grain metal carbide powders are conveniently prepared from the corresponding metal oxide by heating in an atmosphere of methane in hydrogen. Sintered articles having a density approaching the theoretical density of the metal carbide itself can be fabricated from the powders by cold pressing, hot pressing or other techniques. 8 claims, no drawings

  13. stabilization of ikpayongo laterite with cement and calcium carbide

    African Journals Online (AJOL)

    PROF EKWUEME

    the stabilization of soil will ensure economy in road construction, while providing an effective way of disposing calcium carbide waste. KEYWORDS: Cement, Calcium carbide waste, Stabilization, Ikpayongo laterite, Pavement material. INTRODUCTION. Road building in the developing nations has been a major challenge to ...

  14. Primary dispersal patterns of uraninite in the Proterozoic Vaal Reef placer deposit, Witwatersrand, South Africa

    International Nuclear Information System (INIS)

    Minter, W.E.L.

    1981-01-01

    The Vaal Reef is a strata-bound uranium orebody. Uraninite, together with nodular pyrite and gold, is a detrital heavy-mineral component of the sediment. Consequently, the areal distribution patterns of uranium content match the braided, southeasterly dispersal pattern evident in the Vaal Reef. Deeper channelways contain more uraninite

  15. PROCESS OF RECOVERING URANIUM

    Science.gov (United States)

    Carter, J.M.; Larson, C.E.

    1958-10-01

    A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

  16. Study on microstructure change of Uranium nitride coated U-7wt%Mo powder by heat treatment

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Woo Hyoung; Park, Jae Soon; Lee, Hae In; Kim, Woo Jeong; Yang, Jae Ho; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Uranium-molybdenum alloy particle dispersion fuel in an aluminum matrix with a high uranium density has been developed for a high performance research reactor in the RERTR program. In order to retard the fuel-matrix interaction in U-Mo/Al dispersion fuel in which the U-Mo fuel particles were dispersed in Al matrix, nitride layer coated U-Mo fuel particle has been designed and techniques to fabricate nitride-layer coated U-7wt%Mo particles have been developed in our lab. In this study, uranium nitride coated U-Mo particle has heat treatment for several times and degree. And we suggested for interaction layer remedy in U-Mo dispersion fuel. We investigate effect of heat treatment interaction layer evolution on uranium nitride coated U-Mo powder. The EDS and XRD analysis to investigate the phase evolution in uranium nitride coated layer is also a part of the present work

  17. INTRAVAL phase 2, test case 8. Alligator Rivers Natural Analogue - Modelling of uranium transport in the weathered zone at Koongarra (Australia). Progress report

    NARCIS (Netherlands)

    van der Weerd H; Hassanizadeh SM; Richardson-van der Poel MA; LBG

    1993-01-01

    A study of uranium transport in the Koongarra site of Alligator Rivers Uranium deposit (Australia) is carried out. The analysis of the solid phase uranium concentration measured at various depths provides a useful picture of the dispersion process. Results of this analysis seem to support the

  18. Method for converting uranium oxides to uranium metal

    Science.gov (United States)

    Duerksen, Walter K.

    1988-01-01

    A process is described for converting scrap and waste uranium oxide to uranium metal. The uranium oxide is sequentially reduced with a suitable reducing agent to a mixture of uranium metal and oxide products. The uranium metal is then converted to uranium hydride and the uranium hydride-containing mixture is then cooled to a temperature less than -100.degree. C. in an inert liquid which renders the uranium hydride ferromagnetic. The uranium hydride is then magnetically separated from the cooled mixture. The separated uranium hydride is readily converted to uranium metal by heating in an inert atmosphere. This process is environmentally acceptable and eliminates the use of hydrogen fluoride as well as the explosive conditions encountered in the previously employed bomb-reduction processes utilized for converting uranium oxides to uranium metal.

  19. Uranium speciation in plants

    International Nuclear Information System (INIS)

    Guenther, A.; Bernhard, G.; Geipel, G.; Reich, T.; Rossberg, A.; Nitsche, H.

    2003-01-01

    Detailed knowledge of the nature of uranium complexes formed after the uptake by plants is an essential prerequisite to describe the migration behavior of uranium in the environment. This study focuses on the determination of uranium speciation after uptake of uranium by lupine plants. For the first time, time-resolved laser-induced fluorescence spectroscopy and X-ray absorption spectroscopy were used to determine the chemical speciation of uranium in plants. Differences were detected between the uranium speciation in the initial solution (hydroponic solution and pore water of soil) and inside the lupine plants. The oxidation state of uranium did not change and remained hexavalent after it was taken up by the lupine plants. The chemical speciation of uranium was identical in the roots, shoot axis, and leaves and was independent of the uranium speciation in the uptake solution. The results indicate that the uranium is predominantly bound as uranyl(VI) phosphate to the phosphoryl groups. Dandelions and lamb's lettuce showed uranium speciation identical to lupine plants. (orig.)

  20. Design and Thermal Analysis for Irradiation of Pyrolytic Carbon/Silicon Carbide Diffusion Couples in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    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 Department 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.

  1. Hydrodynamic dispersion

    International Nuclear Information System (INIS)

    Pryce, M.H.L.

    1985-01-01

    A dominant mechanism contributing to hydrodynamic dispersion in fluid flow through rocks is variation of travel speeds within the channels carrying the fluid, whether these be interstices between grains, in granular rocks, or cracks in fractured crystalline rocks. The complex interconnections of the channels ensure a mixing of those parts of the fluid which travel more slowly and those which travel faster. On a macroscopic scale this can be treated statistically in terms of the distribution of times taken by a particle of fluid to move from one surface of constant hydraulic potential to another, lower, potential. The distributions in the individual channels are such that very long travel times make a very important contribution. Indeed, while the mean travel time is related to distance by a well-defined transport speed, the mean square is effectively infinite. This results in an asymmetrical plume which differs markedly from a gaussian shape. The distribution of microscopic travel times is related to the distribution of apertures in the interstices, or in the microcracks, which in turn are affected in a complex way by the stresses acting on the rock matrix

  2. 31 CFR 540.317 - Uranium feed; natural uranium feed.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

  3. Are stirring and sonication pre-dispersion methods equivalent for in vitro toxicology evaluation of SiC and TiC?

    International Nuclear Information System (INIS)

    Mejia, Jorge; Valembois, Vanessa; Piret, Jean-Pascal; Tichelaar, Frans; Huis, Marijn van; Masereel, Bernard; Toussaint, Olivier; Delhalle, Joseph; Mekhalif, Zineb; Lucas, Stéphane

    2012-01-01

    The evolution of the particle size distribution and the surface composition of silicon carbide and titanium carbide nanoparticle (NP) dispersions were studied. The pre-dispersions were prepared using two commonly used protocols for dispersion: stirring and sonication. Two dispersants were investigated (water and Pluronic F108 1 %) at two stages: pre-dispersion and during in vitro assays. Our data show that for each tested condition, different time-dependent results for the surface chemical composition as well as size and percentage of the agglomerates and the primary particles are observed. De-agglomeration and successive or simultaneous cleaning-wrapping cycles of the nanomaterial are observed and are related to the dispersion method and the medium as well as to the chemical stability of the NP surface. Biological response during in vitro assessment was also performed for one given pre-dispersion time condition and demonstrates that the preparation method significantly alters the results.

  4. Are stirring and sonication pre-dispersion methods equivalent for in vitro toxicology evaluation of SiC and TiC?

    Energy Technology Data Exchange (ETDEWEB)

    Mejia, Jorge, E-mail: jorge.mejiamendoza@fundp.ac.be; Valembois, Vanessa [University of Namur-FUNDP, Research Centre for the Physics of Matter and Radiation (LARN-PMR), NARILIS (Belgium); Piret, Jean-Pascal [University of Namur-FUNDP, Research Unit in Cellular Biology (URBC), NARILIS (Belgium); Tichelaar, Frans; Huis, Marijn van [Delft University of Technology, National Centre for HRTEM, Kavli Institute of Nanoscience (Netherlands); Masereel, Bernard [University of Namur-FUNDP, Department of Pharmacy NAMEDIC, Namur Thrombosis and Hemostasis Center (NTHC) (Belgium); Toussaint, Olivier [University of Namur-FUNDP, Research Unit in Cellular Biology (URBC), NARILIS (Belgium); Delhalle, Joseph; Mekhalif, Zineb [University of Namur-FUNDP, Laboratory of Chemistry and Electrochemistry of Surfaces-CES (Belgium); Lucas, Stephane [University of Namur-FUNDP, Research Centre for the Physics of Matter and Radiation (LARN-PMR), NARILIS (Belgium)

    2012-03-15

    The evolution of the particle size distribution and the surface composition of silicon carbide and titanium carbide nanoparticle (NP) dispersions were studied. The pre-dispersions were prepared using two commonly used protocols for dispersion: stirring and sonication. Two dispersants were investigated (water and Pluronic F108 1 %) at two stages: pre-dispersion and during in vitro assays. Our data show that for each tested condition, different time-dependent results for the surface chemical composition as well as size and percentage of the agglomerates and the primary particles are observed. De-agglomeration and successive or simultaneous cleaning-wrapping cycles of the nanomaterial are observed and are related to the dispersion method and the medium as well as to the chemical stability of the NP surface. Biological response during in vitro assessment was also performed for one given pre-dispersion time condition and demonstrates that the preparation method significantly alters the results.

  5. Method for fabricating boron carbide articles

    International Nuclear Information System (INIS)

    Ardary, Z.; Reynolds, C.

    1980-01-01

    Described is a method for fabricating an essentially uniformly dense boron carbide article of a length-to-diameter or width ratio greater than 2 to 1 comprising the steps of providing a plurality of article segments to be joined together to form the article with each of said article segments having a length-to-diameter or width ratio less than 1.5 to 1. Each segment is fabricated by hot pressing a composition consisting of boron carbide powder at a pressure and temperature effective to provide the article segment with a density greater than about 85% of theoretical density, providing each article segment with parallel planar end surfaces, placing a plurality of said article segments in a hot-pressing die in a line with the planar surfaces of adjacent article segments being disposed in intimate contact, and hot pressing the aligned article segments at a temperature and pressure effective to provide said article with a density over the length thereof in the range of about 94 to 98 percent theoretical density and greater than the density provided in the discrete hot pressing of each of the article segments and to provide a bond between adjacent article segments with said bond being at least equivalent in hardness, strength and density to a remainder of said article

  6. Development of silicon carbide composites for fusion

    International Nuclear Information System (INIS)

    Snead, L.L.

    1993-01-01

    The use of silicon carbide composites for structural materials is of growing interest in the fusion community. However, radiation effects in these materials are virtually unexplored, and the general state of ceramic matrix composites for nonnuclear applications is still in its infancy. Research into the radiation response of the most popular silicon carbide composite, namely, the chemically vapor-deposited (CVD) SiC-carbon-Nicalon fiber system is discussed. Three areas of interest are the stability of the fiber and matrix materials, the stability of the fiber-matrix interface, and the true activation of these open-quotes reduced activityclose quotes materials. Two methods are presented that quantitatively measure the effect of radiation on fiber and matrix elastic modulus as well as the fiber-matrix interfacial strength. The results of these studies show that the factor limiting the radiation performance of the CVD SiC-carbon-Nicalon system is degradation of the Nicalon fiber, which leads to a weakened carbon interface. The activity of these composites is significantly higher than expected and is dominated by impurity isotopes. 52 refs., 12 figs., 3 tabs

  7. Metal-boride phase formation on tungsten carbide (WC-Co) during microwave plasma chemical vapor deposition

    Energy Technology Data Exchange (ETDEWEB)

    Johnston, Jamin M.; Catledge, Shane A., E-mail: catledge@uab.edu

    2016-02-28

    Graphical abstract: - Highlights: • A detailed phase analysis after PECVD boriding shows WCoB, CoB and/or W{sub 2}CoB{sub 2}. • EDS of PECVD borides shows boron diffusion into the carbide grain structure. • Nanoindentation hardness and modulus of borides is 23–27 GPa and 600–780 GPa. • Scratch testing shows hard coating with cracking at 40N and spallation at 70N. - Abstract: Strengthening of cemented tungsten carbide by boriding is used to improve the wear resistance and lifetime of carbide tools; however, many conventional boriding techniques render the bulk carbide too brittle for extreme conditions, such as hard rock drilling. This research explored the variation in metal-boride phase formation during the microwave plasma enhanced chemical vapor deposition process at surface temperatures from 700 to 1100 °C. We showed several well-adhered metal-boride surface layers consisting of WCoB, CoB and/or W{sub 2}CoB{sub 2} with average hardness from 23 to 27 GPa and average elastic modulus of 600–730 GPa. The metal-boride interlayer was shown to be an effective diffusion barrier against elemental cobalt; migration of elemental cobalt to the surface of the interlayer was significantly reduced. A combination of glancing angle X-ray diffraction, electron dispersive spectroscopy, nanoindentation and scratch testing was used to evaluate the surface composition and material properties. An evaluation of the material properties shows that plasma enhanced chemical vapor deposited borides formed at substrate temperatures of 800 °C, 850 °C, 900 °C and 1000 °C strengthen the material by increasing the hardness and elastic modulus of cemented tungsten carbide. Additionally, these boride surface layers may offer potential for adhesion of ultra-hard carbon coatings.

  8. Metal-boride phase formation on tungsten carbide (WC-Co) during microwave plasma chemical vapor deposition

    International Nuclear Information System (INIS)

    Johnston, Jamin M.; Catledge, Shane A.

    2016-01-01

    Graphical abstract: - Highlights: • A detailed phase analysis after PECVD boriding shows WCoB, CoB and/or W_2CoB_2. • EDS of PECVD borides shows boron diffusion into the carbide grain structure. • Nanoindentation hardness and modulus of borides is 23–27 GPa and 600–780 GPa. • Scratch testing shows hard coating with cracking at 40N and spallation at 70N. - Abstract: Strengthening of cemented tungsten carbide by boriding is used to improve the wear resistance and lifetime of carbide tools; however, many conventional boriding techniques render the bulk carbide too brittle for extreme conditions, such as hard rock drilling. This research explored the variation in metal-boride phase formation during the microwave plasma enhanced chemical vapor deposition process at surface temperatures from 700 to 1100 °C. We showed several well-adhered metal-boride surface layers consisting of WCoB, CoB and/or W_2CoB_2 with average hardness from 23 to 27 GPa and average elastic modulus of 600–730 GPa. The metal-boride interlayer was shown to be an effective diffusion barrier against elemental cobalt; migration of elemental cobalt to the surface of the interlayer was significantly reduced. A combination of glancing angle X-ray diffraction, electron dispersive spectroscopy, nanoindentation and scratch testing was used to evaluate the surface composition and material properties. An evaluation of the material properties shows that plasma enhanced chemical vapor deposited borides formed at substrate temperatures of 800 °C, 850 °C, 900 °C and 1000 °C strengthen the material by increasing the hardness and elastic modulus of cemented tungsten carbide. Additionally, these boride surface layers may offer potential for adhesion of ultra-hard carbon coatings.

  9. Standard test methods for analysis of sintered gadolinium oxide-uranium dioxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 These test methods cover procedures for the analysis of sintered gadolinium oxide-uranium dioxide pellets to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Section Carbon (Total) by Direct CombustionThermal Conductivity Method C1408 Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method Chlorine and Fluorine by Pyrohydrolysis Ion-Selective Electrode Method C1502 Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide Gadolinia Content by Energy-Dispersive X-Ray Spectrometry C1456 Test Method for Determination of Uranium or Gadolinium, or Both, in Gadolinium Oxide-Uranium Oxide Pellets or by X-Ray Fluorescence (XRF) Hydrogen by Inert Gas Fusion C1457 Test Method for Determination of Total Hydrogen Content of Uranium Oxide Powders and Pellets by Carrier Gas Extraction Isotopic Uranium Composition by Multiple-Filament Surface-Ioni...

  10. Biogeochemical prospecting for uranium with conifers: results from the Midnite mine area, Washington

    International Nuclear Information System (INIS)

    Nash, J.T.; Ward, F.N.

    1977-01-01

    The ash of needles, cones, and duff from Ponderosa pine (Pinus ponderosa Laws) growing near uranium deposits of the Midnite mine, Stevens County, Wash., contain as much as 200 ppM uranium. Needle samples containing more than 10 ppM uranium define zones that correlate well with known uranium deposits or dumps. Dispersion is as much as 300 m but generally is less. Background is about 1 ppM. Tree roots are judged to be sampling ore, low-grade uranium halo, or ground water to a depth of about 15 m. Uptake of uranium by Douglas fir (Pseudotsuga menziesii (Mirb.) Franco) needles appears to be about the same as by Ponderosa pine needles. Cones and duff are generally enriched in uranium relative to needles. Needles, cones, and duff are recommended as easily collected, uncomplicated sample media for geochemical surveys. Samples can be analyzed by standard methods and total cost per sample kept to about $6

  11. Formation of hexagonal silicon carbide by high energy ion beam irradiation on Si (1 0 0) substrate

    International Nuclear Information System (INIS)

    Bhuyan, H; Favre, M; Valderrama, E; Avaria, G; Chuaqui, H; Mitchell, I; Wyndham, E; Saavedra, R; Paulraj, M

    2007-01-01

    We report the investigation of high energy ion beam irradiation on Si (1 0 0) substrates at room temperature using a low energy plasma focus (PF) device operating in methane gas. The unexposed and ion exposed substrates were characterized by x-ray diffraction, scanning electron microscopy (SEM), photothermal beam deflection, energy-dispersive x-ray analysis and atomic force microscopy (AFM) and the results are reported. The interaction of the pulsed PF ion beams, with characteristic energy in the 60-450 keV range, with the Si surface, results in the formation of a surface layer of hexagonal silicon carbide. The SEM and AFM analyses indicate clear step bunching on the silicon carbide surface with an average step height of 50 nm and a terrace width of 800 nm

  12. Uranium in Malwa region of Punjab, India

    International Nuclear Information System (INIS)

    Kochhar, Naresh; Dadwal, Veena; Balaram, V.

    2012-01-01

    It is well known in Punjab that the Malwa region shows a very high incidence of cancer, stunted growth and other neurological disorders. The high values of uranium have been attributed to Kota nuclear power plant; Khushab heavy water plant in Pakistan; and uranium - carrying winds from Afghanistan, without any scientific basis. Though Malwa is a part of Punjab, geologically it is more akin to Haryana and Rajasthan. Uranium is a naturally occurring radioactive element which is present in trace in rocks, minerals plants and natural waters. It occurs along with thorium and potassium in granitic rocks. It has the property to get dissolved in water in hexavalent form at a normal pH of 5 to 7. It gets precipitated in the reducing environment in tetravalent form and form complexes such as hydroxides, phosphate, sulfate, carbonate etc. Uranium compounds are soluble in water, very mobile and travel kilometers. When the bed rocks containing uranium and thorium and other elements are exposed to sun, rain, wind, they get weathered and breakdown to form soil. Uranium gets dispersed in matrix, soil and finally gets re-deposited in areas/pockets where reducing conditions are present. Hence we get higher concentration of uranium in pockets. There are no rocks exposed on the surface in the SW Punjab. However the rocks of Aravalli-Delhi ridge and Malani granites and rhyohtes are exposed at Tusham district Bhiwani just south of the region. These rocks take a northwest turn from Tusham and become submerged under the Punjab Plains only to get resurfaced at Kirana Hills Pakistan. The gravity data have delineated 6 km wide and 240 km long per shaped body under the Punjab plains covering the SW Punjab. The Tusham granites are high heat producing granites that is they are enriched in uranium, thorium and Potasium. The uranium concentration in Tusham granites is 8 to 11 .5 parts per million (ppm) as compared to the normal value of 4.5 in granites in general. The average crustal values is 2

  13. Recovery of uranium from phosphatic rock and its derivatives

    International Nuclear Information System (INIS)

    Romero Guzman, E.T.

    1992-01-01

    The recovery of uranium present in the manufacture process of phosphoric acid and fertilizers has been one interesting field of study in chemistry. It is true that the recovery of uranium it is not very attractive from the commercial point of view, however the phosphatic fertilizers have an important amount of uranium which comes from the starting materials (phosphatic rock), therefore there must be many tons of uranium that are dispersed in the environmental together with the fertilizers used in agriculture every year. They are utilized for the enrichment of the nutrients which are exhausted in the soil. In this work, uranium was identified and quantified in the phosphatic rocks and in inorganic fertilizers using Gamma Spectroscopy, Neutron Activation Analysis, UV/Visible Spectrophotometry, Alpha Spectroscopy. On the other hand, it was done a correlation of the behaviour of uranium with inorganic elements present in the samples such as phosphorus, calcium and iron; which were determined by UV/Visible Spectrophotometry for phosphorus and Atomic Absorption Spectrometry for calcium and iron. The quantity of uranium found in the phosphatic rock, phosphoric acid and fertilizers was considerable (70-200 ppm). The adequate conditions for the recovery of 40% of total of uranium from the phosphatic rock with the addition of leaching solutions were stablished. (Author)

  14. Development of MTR fuel plate with U-Al dispersion core constituents

    International Nuclear Information System (INIS)

    Bressiani, Jose Carlos

    1979-01-01

    This work is a contribution to the development of fuel plates for Research Nuclear Reaction Materials Test Reactors. The plates have the core constituted by dispersions of metallic uranium in aluminum. The main topics of this work are: 1) The preparation of uranium powder with particle sizes in the 53-105μm diameter range; 2) The mixture and cold-pressing of uranium and aluminum powders for different uranium concentrations; 3) The behavior of the dispersions in the roll milling conditions; 4) Blister, radiographic, metallographic and irradiation tests for quality control of the plates. The irradiation test was performed in the IEA-R1 swimming-pool reactor using a prototype with a dispersion of aluminum and natural uranium (45 w/o ), reaching an integrated neutron flux of 8.663 X 10 18 n/cm 2 , no visual changes being noticed after the completion of the experiment. The behavior of the uranium-aluminum reaction for dispersions with 45% w/o uranium also studied. X-ray diffraction experiments showed the formation of UAl 2 UAl 3 and UAl 4 , while energy dispersive analysis of X-rays(EDAX) demonstrated that the diffusion of aluminum in uranium is the mechanism responsible for that reaction. The activation energy for the U-Al reaction was determined by dilatometric experiments yielding 20.2 kcal/mol.The aluminum-uranium reaction reaches an end when extended to 96 h at 600 deg C, namely, when all the uranium is found in the UAl 4 composition. (author)

  15. Uranium Elemental and Isotopic Constraints on Groundwater Flow Beneath the Nopal I Uranium Deposit, Pena Blanca, Mexico

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Goldstein; M.T. Murrell; A.M. Simmons

    2005-07-11

    The Nopal I uranium deposit in Chihuahua, Mexico, is an excellent analogue for evaluating the fate of spent fuel, associated actinides, and fission products over long time scales for the proposed Yucca Mountain high-level nuclear waste repository. In 2003, three groundwater wells were drilled directly adjacent to (PB-1) and 50 m on either side of the uranium deposit (PB-2 and PB-3) in order to evaluate uranium-series transport in three dimensions. After drilling, uranium concentrations were elevated in all of the three wells (0.1-18 ppm) due to drilling activities and subsequently decreased to {approx}5-20% of initial values over the next several months. The {sup 234}U/{sup 238}U activity ratios were similar for PB-1 and PB-2 (1.005 to 1.079) but distinct for PB-3 (1.36 to 1.83) over this time period, suggesting limited mixing between groundwater from these wells over these short time and length scales. Regional groundwater wells located up to several km from the deposit also have distinct uranium isotopic characteristics and constrain mixing over larger length and time scales. We model the decreasing uranium concentrations in the newly drilled wells with a simple one-dimensional advection-dispersion model, assuming uranium is introduced as a slug to each of the wells and transported as a conservative tracer. Using this model for our data, the relative uranium concentrations are dependent on both the longitudinal dispersion as well as the mean groundwater flow velocity. These parameters have been found to be correlated in both laboratory and field studies of groundwater velocity and dispersion (Klotz et al., 1980). Using typical relationships between velocity and dispersion for field and laboratory studies along with the relationship observed from our uranium data, both velocity (1-10 n/yr) and dispersion coefficient (1E-5 to 1E-2 cm{sup 2}/s) can be derived from the modeling. As discussed above, these relatively small flow velocities and dispersivities agree with

  16. Uranium Elemental and Isotopic Constraints on Groundwater Flow Beneath the Nopal I Uranium Deposit, Pena Blanca, Mexico

    International Nuclear Information System (INIS)

    S.J. Goldstein; M.T. Murrell; A.M. Simmons

    2005-01-01

    The Nopal I uranium deposit in Chihuahua, Mexico, is an excellent analogue for evaluating the fate of spent fuel, associated actinides, and fission products over long time scales for the proposed Yucca Mountain high-level nuclear waste repository. In 2003, three groundwater wells were drilled directly adjacent to (PB-1) and 50 m on either side of the uranium deposit (PB-2 and PB-3) in order to evaluate uranium-series transport in three dimensions. After drilling, uranium concentrations were elevated in all of the three wells (0.1-18 ppm) due to drilling activities and subsequently decreased to ∼5-20% of initial values over the next several months. The 234 U/ 238 U activity ratios were similar for PB-1 and PB-2 (1.005 to 1.079) but distinct for PB-3 (1.36 to 1.83) over this time period, suggesting limited mixing between groundwater from these wells over these short time and length scales. Regional groundwater wells located up to several km from the deposit also have distinct uranium isotopic characteristics and constrain mixing over larger length and time scales. We model the decreasing uranium concentrations in the newly drilled wells with a simple one-dimensional advection-dispersion model, assuming uranium is introduced as a slug to each of the wells and transported as a conservative tracer. Using this model for our data, the relative uranium concentrations are dependent on both the longitudinal dispersion as well as the mean groundwater flow velocity. These parameters have been found to be correlated in both laboratory and field studies of groundwater velocity and dispersion (Klotz et al., 1980). Using typical relationships between velocity and dispersion for field and laboratory studies along with the relationship observed from our uranium data, both velocity (1-10 n/yr) and dispersion coefficient (1E-5 to 1E-2 cm 2 /s) can be derived from the modeling. As discussed above, these relatively small flow velocities and dispersivities agree with mixing

  17. Uranium of Kazakhstan

    International Nuclear Information System (INIS)

    Tsalyuk, Yu.; Gurevich, D.

    2000-01-01

    Over 25 % of the world's uranium reserves are concentrated in Kazakhstan. So, the world's largest Shu-Sarysu uranium province is situated on southern Kazakhstan, with resources exceeding 1 billion tonnes of uranium. No less, than 3 unique deposits with resources exceeding 100,000 tonnes are situated here. From the economic point of view the most important thing is that these deposits are suitable for in-situ leaching, which is the cheapest, environmentally friendly and most efficient method available for uranium extracting. In 1997 the Kazatomprom National Joint-Stock Company united all Kazakhstan's uranium enterprises (3 mine and concentrating plants, Volkovgeologiya Joint-Stock Company and the Ulbinskij Metallurgical plant). In 1998 uranium production came to 1,500 tonnes (860 kg in 1997). In 1999 investment to the industry were about $ 30 million. Plans for development of Kazakhstan's uranium industry provide a significant role for foreign partners. At present, 2 large companies (Comeco (Canada), Cogema (France) working in Kazakhstan. Kazakatomprom continues to attract foreign investors. The company's administration announced that in that in next year they have plan to make a radical step: to sell 67 % of stocks to strategic investors (at present 100 % of stocks belongs to state). Authors of the article regard, that the Kazakhstan's uranium industry still has significant reserves to develop. Even if the scenario for the uranium industry could be unfavorable, uranium production in Kazakhstan may triple within the next three to four years. The processing of uranium by the Ulbinskij Metallurgical Plant and the production of some by-products, such as rhenium, vanadium and rare-earth elements, may provide more profits. Obviously, the sale of uranium (as well as of any other reserves) cannot make Kazakhstan a prosperous country. However, country's uranium industry has a god chance to become one of the most important and advanced sectors of national economy

  18. Chemical, mechanical, and tribological properties of pulsed-laser-deposited titanium carbide and vanadium carbide

    International Nuclear Information System (INIS)

    Krzanowski, J.E.; Leuchtner, R.E.

    1997-01-01

    The chemical, mechanical, and tribological properties of pulsed-laser-deposited TiC and VC films are reported in this paper. Films were deposited by ablating carbide targets using a KrF (λ = 248 nm) laser. Chemical analysis of the films by XPS revealed oxygen was the major impurity; the lowest oxygen concentration obtained in a film was 5 atom%. Oxygen was located primarily on the carbon sublattice of the TiC structure. The films were always substoichiometric, as expected, and the carbon in the films was identified primarily as carbidic carbon. Nanoindentation hardness tests gave values of 39 GPa for TiC and 26 GPa for VC. The friction coefficient for the TiC films was 0.22, while the VC film exhibited rapid material transfer from the steel ball to the substrate resulting in steel-on-steel tribological behavior

  19. Study on the performance of fuel elements with carbide and carbide-nitride fuel

    International Nuclear Information System (INIS)

    Golovchenko, Yu.M.; Davydov, E.F.; Maershin, A.A.

    1985-01-01

    Characteristics, test conditions and basic results of material testing of fuel elements with carbide and carbonitride fuel irradiated in the BOR-60 reactor up to 3-10% burn-up at specific power rate of 55-70 kW/m and temperatures of the cladding up to 720 deg C are described. Increase of cladding diameter is stated mainly to result from pressure of swelling fuel. The influence of initial efficient porosity of the fuel on cladding deformation and fuel stoichiometry on steel carbonization is considered. Utilization of carbide and carbonitride fuel at efficient porosity of 20% at the given test modes is shown to ensure their operability up to 10% burn-up

  20. Preparation and Fatigue Properties of Functionally Graded Cemented Carbides

    International Nuclear Information System (INIS)

    Liu Yong; Liu Fengxiao; Liaw, Peter K.; He Yuehui

    2008-01-01

    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 η-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 (∼100 MPa) than the conventional ones under 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

  1. Graphite and boron carbide composites made by hot-pressing

    International Nuclear Information System (INIS)

    Miyazaki, K.; Hagio, T.; Kobayashi, K.

    1981-01-01

    Composites consisting of graphite and boron carbide were made by hot-pressing mixed powders of coke carbon and boron carbide. The change of relative density, mechanical strength and electrical resistivity of the composites and the X-ray parameters of coke carbon were investigated with increase of boron carbide content and hot-pressing temperature. From these experiments, it was found that boron carbide powder has a remarkable effect on sintering and graphitization of coke carbon powder above the hot-pressing temperature of 2000 0 C. At 2200 0 C, electrical resistivity of the composite and d(002) spacing of coke carbon once showed minimum values at about 5 to 10 wt% boron carbide and then increased. The strength of the composite increased with increase of boron carbide content. It was considered that some boron from boron carbide began to diffuse substitutionally into the graphite structure above 2000 0 C and densification and graphitization were promoted with the diffusion of boron. Improvements could be made to the mechanical strength, density, oxidation resistance and manufacturing methods by comparing with the properties and processes of conventional graphites. (author)

  2. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  3. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  4. Seed dispersal in fens

    NARCIS (Netherlands)

    Middleton, Beth; van Diggelen, Rudy; Jensen, Kai

    Question: How does seed dispersal reduce fen isolation and contribute to biodiversity? Location: European and North American fens. Methods: This paper reviews the literature on seed dispersal to fens. Results: Landscape fragmentation may reduce dispersal opportunities thereby isolating fens and

  5. Titrimetric determination of uranium

    International Nuclear Information System (INIS)

    Florence, T.M.

    1989-01-01

    Titrimetric methods are almost invariably used for the high precision assay of uranium compounds, because gravimetric methods are nonselective, and not as reliable. Although precipitation titrations have been used, for example with cupferron and ferrocyanide, and chelate titrations with EDTA and oxine give reasonable results, in practice only redox titrations find routine use. With all redox titration methods for uranium a precision of 01 to 02 percent can be achieved, and precisions as high as 0.003 percent have been claimed for the more refined techniques. There are two types of redox titrations for uranium in common use. The first involves the direct titration of uranium (VI) to uranium (IV) with a standard solution of a strong reductant, such as chromous chloride or titanous chloride, and the second requires a preliminary reduction of uranium to the (IV) or (III) state, followed by titration back to the (VI) state with a standard oxidant. Both types of redox titrations are discussed. 4 figs

  6. Politics of Uranium

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    Uranium is the most political of all the elements, the material for the production of both the large amounts of electricity and the most destructive weapons in the world. The problems that its dual potential creates are only now beginning to become evident. Author Norman Moss looks at this situation and sheds light on many of the questions that emerge. The nuclear issue always comes back to how much uranium there is, what can be done with it, and which countries have it. Starting with a concise history of uranium and explaining its technology in terms the nonspecialist can understand, The Politics of Uranium considers the political issues that technical arguments obscure. It tells the little-known story of the international uranium cartel, explains the entanglements of governments with the uranium trade, and describes the consequences of wrong decisions and blunders-especially the problems of nuclear waste. It also examines the intellectual and emotional roots of the anti-nuclear movement

  7. Uranium resources and supply

    International Nuclear Information System (INIS)

    Cameron, J.

    1973-01-01

    The future supply of uranium has to be considered against a background of forecasts of uranium demand over the next decades which show increases of a spectacular nature. It is not necessary to detail these forecasts, they are well known. A world survey by the Joint NEA/IAEA Working Party on 'Uranium Resources, Production and Demand', completed this summer, indicates that from a present production level of just over 19,000 tonnes uranium per year, the demand will rise to the equivalent of an annual production requirement of 50,000 tonnes uranium by 1980, 100,000 by 1985 and 180,000 by 1990. Few, if any, mineral production industries have been called upon to plan for a near tenfold increase in production in a space of about 15 years as these forecasts imply. This might possibly mean that, perhaps, ten times the present number of uranium mines will have to be planned and engineered by 1990

  8. How much uranium

    International Nuclear Information System (INIS)

    Kenward, M.

    1976-01-01

    Comment is made on the latest of a series of reports on world uranium resources from the OECD's Nuclear Energy Agency and the UN's International Atomic Energy Agency (Uranium resources, production and demand (including other nuclear fuel cycle data), published by the Organisation for Economic Cooperation and Development, Paris). The report categories uranium reserves by their recovery cost and looks at power demand and the whole of the nuclear fuel cycle, including uranium enrichment and spent fuel reprocessing. The effect that fluctuations in uranium prices have had on exploration for new uranium resources is considered. It is stated that increased exploration is essential considering the long lead times involved but that thanks to today's higher prices there are distinct signs that prospecting activities are increasing again. (U.K.)

  9. Uranium Mill Tailings Management

    International Nuclear Information System (INIS)

    Nelson, J.D.

    1982-01-01

    This book presents the papers given at the Fifth Symposium on Uranium Mill Tailings Management. Advances made with regard to uranium mill tailings management, environmental effects, regulations, and reclamation are reviewed. Topics considered include tailings management and design (e.g., the Uranium Mill Tailings Remedial Action Project, environmental standards for uranium mill tailings disposal), surface stabilization (e.g., the long-term stability of tailings, long-term rock durability), radiological aspects (e.g. the radioactive composition of airborne particulates), contaminant migration (e.g., chemical transport beneath a uranium mill tailings pile, the interaction of acidic leachate with soils), radon control and covers (e.g., radon emanation characteristics, designing surface covers for inactive uranium mill tailings), and seepage and liners (e.g., hydrologic observations, liner requirements)

  10. Geochemical exploration for uranium

    International Nuclear Information System (INIS)

    1988-01-01

    This Technical Report is designed mainly to introduce the methods and techniques of uranium geochemical exploration to exploration geologists who may not have had experience with geochemical exploration methods in their uranium programmes. The methods presented have been widely used in the uranium exploration industry for more than two decades. The intention has not been to produce an exhaustive, detailed manual, although detailed instructions are given for a field and laboratory data recording scheme and a satisfactory analytical method for the geochemical determination of uranium. Rather, the intention has been to introduce the concepts and methods of uranium exploration geochemistry in sufficient detail to guide the user in their effective use. Readers are advised to consult general references on geochemical exploration to increase their understanding of geochemical techniques for uranium

  11. Classification of Uranium deposits

    International Nuclear Information System (INIS)

    Dahlkamp, F.J.

    1978-01-01

    A listing of the recognized types of uranium mineralization shows nineteen determinable types out of which only six can be classified as of economic significance at present: Oligomiitic quartz pebble conglomerates, sandstone types, calcretes, intra-intrusive types, hydrothermal veins, veinlike types. The different types can be genetically related to prevalent geological environments, i.e. 1. the primary uranium occurrences formed by endogenic processes, 2. the secondary derived from the primary by subsequent exogenic processes, 3. the tertiary occurrences are assumed to be formed by endogenic metamorphic processes, although little is known about the behaviour of the uranium during the metamorphosis and therefore the metallogenesis of this tertiary uranium generation is still vague. A metallotectonic-geochronologic correlation of the uranium deposits shows a distinct affinity of the uranium to certain geological epochs: The Upper Archean, Lower Proterozoic, the Hercynian and, in a less established stage, the Upper Proterozoic. (orig.) 891 HP/orig. 892 MKO [de

  12. Uranium Newsletter. No. 1

    International Nuclear Information System (INIS)

    1987-03-01

    The new Uranium Newsletter is presented as an IAEA annual newsletter. The organization of the IAEA and its involvement with uranium since its founding in 1957 is described. The ''Red Book'' (Uranium Resources, Production and Demand) is mentioned. The Technical Assistance Programme of the IAEA in this field is also briefly mentioned. The contents also include information on the following meetings: The Technical Committee Meeting on Uranium Deposits in Magmatic and Metamorphic Rocks, Advisory Group Meeting on the Use of Airborne Radiometric Data, and the Technical Committee Meeting on Metallogenesis. Recent publications are listed. Current research contracts in uranium exploration are mentioned. IAEA publications on uranium (in press) are listed also. Country reports from the following countries are included: Australia, Brazil, Canada, China (People's Republic of), Denmark, Finland, Germany (Federal Republic of), Malaysia, Philippines, Portugal, South Africa (Republic of), Spain, Syrian Arab Republic, United Kingdom, United States of America, Zambia, and Greece. There is also a report from the Commission of European Communities

  13. Uranium purchases report 1992

    International Nuclear Information System (INIS)

    1993-01-01

    Data reported by domestic nuclear utility companies in their responses to the 1991 and 1992 ''Uranium Industry Annual Survey,'' Form EIA-858, Schedule B ''Uranium Marketing Activities,are provided in response to the requirements in the Energy Policy Act 1992. Data on utility uranium purchases and imports are shown on Table 1. Utility enrichment feed deliveries and secondary market acquisitions of uranium equivalent of US DOE separative work units are shown on Table 2. Appendix A contains a listing of firms that sold uranium to US utilities during 1992 under new domestic purchase contracts. Appendix B contains a similar listing of firms that sold uranium to US utilities during 1992 under new import purchase contracts. Appendix C contains an explanation of Form EIA-858 survey methodologies with emphasis on the processing of Schedule B data

  14. The development of the production process for the thorium/uranium dicarbide fuel kernels for the first charge of the Dragon Reactor

    International Nuclear Information System (INIS)

    Burnett, R.C.; Hankart, L.J.; Horsley, G.W.

    1965-05-01

    The development of methods of producing spheroidal sintered porous kernels of hyperstoichiometric thorium/uranium dicarbide solid solution from thorium/uranium monocarbide/carbon and thoria/urania/carbon powder mixes is described. The work has involved study of (i) Methods of preparing green kernels from UC/Th/C powder mixes using the rotary sieve technique. (ii) Methods of producing green kernels from UO2/Th02/C powder mixes using the planetary mill technique. (iii) The conversion by appropriate heat treatment of green kernels produced by both routes to sintered porous kernels of thorium/uranium carbide. (iv) The efficiency of the processes. (author)

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

    Science.gov (United States)

    Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

    1995-01-01

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

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

    Science.gov (United States)

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

    1995-06-06

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

  17. New french uranium mineral species

    International Nuclear Information System (INIS)

    Branche, G.; Chervet, J.; Guillemin, C.

    1952-01-01

    In this work, the authors study the french new uranium minerals: parsonsite and renardite, hydrated phosphates of lead and uranium; kasolite: silicate hydrated of uranium and lead uranopilite: sulphate of uranium hydrated; bayleyite: carbonate of uranium and of hydrated magnesium; β uranolite: silicate of uranium and of calcium hydrated. For all these minerals, the authors give the crystallographic, optic characters, and the quantitative chemical analyses. On the other hand, the following species, very rare in the french lodgings, didn't permit to do quantitative analyses. These are: the lanthinite: hydrated uranate oxide; the α uranotile: silicate of uranium and of calcium hydrated; the bassetite: uranium phosphate and of hydrated iron; the hosphuranylite: hydrated uranium phosphate; the becquerelite: hydrated uranium oxide; the curite: oxide of uranium and lead hydrated. Finally, the authors present at the end of this survey a primary mineral: the brannerite, complex of uranium titanate. (author) [fr

  18. Uranium demand. An exploration challenge

    Energy Technology Data Exchange (ETDEWEB)

    Roux, A J.A.

    1976-10-01

    The estimated world resources of uranium as well as the estimated consumption of uranium over the next 25 years are briefly discussed. Attention is also given to the prospecting for uranium in South Africa and elsewhere in the world.

  19. Precipitation behavior of carbides in high-carbon martensitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Qin-tian; Li, Jing; Shi, Cheng-bin; Yu, Wen-tao; Shi, Chang-min [University of Science and Technology, Beijing (China). State Key Laboratory of Advanced Metallurgy; Li, Ji-hui [Yang Jiang Shi Ba Zi Group Co., Ltd, Guangdong (China)

    2017-01-15

    A fundamental study on the precipitation behavior of carbides was carried out. Thermo-calc software, scanning electron microscopy, electron probe microanalysis, transmission electron microscopy, X-ray diffractometry and high-temperature confocal laser scanning microscopy were used to study the precipitation and transformation behaviors of carbides. Carbide precipitation was of a specific order. Primary carbides (M7C3) tended to be generated from liquid steel when the solid fraction reached 84 mol.%. Secondary carbides (M7C3) precipitated from austenite and can hardly transformed into M23C6 carbides with decreasing temperature in air. Primary carbides hardly changed once they were generated, whereas secondary carbides were sensitive to heat treatment and thermal deformation. Carbide precipitation had a certain effect on steel-matrix phase transitions. The segregation ability of carbon in liquid steel was 4.6 times greater that of chromium. A new method for controlling primary carbides is proposed.

  20. Performance of a sphere-pac mixed carbide fuel pin irradiated in the Dounreay Fast Reactor (DFR 527/1 experiment)

    International Nuclear Information System (INIS)

    Bischoff, K.; Smith, L.; Stratton, R.W.

    1980-10-01

    The DFR 527/1 experiment was the first irradiation of EIR sphere-pac uranium-plutonium mixed carbide fuel in a fast flux. The experiment has been successfully irradiated to a burn-up of 7.3% FIMA at ratings between 45 and 62 kW m - 1 and clad temperatures between 300 and 600 0 C. Restructuring and elemental redistribution has been found to be similar to the pattern established for pellet type fuel and follows effects seen in earlier sphere-pac carbide tests. Gas release of 12-14% has been measured. A preliminary comparison of radial temperature distribution calculations using a first version of the fuel behaviour modelling code SPECKLE with the actual metallography has been attempted. (Auth.)