WorldWideScience

Sample records for temperature uranium carbide

  1. 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)

  2. Universal high-temperature heat treatment furnace for FBR mixed uranium and plutonium carbide fuel

    International Nuclear Information System (INIS)

    Handa, Muneo; Takahashi, Ichiro; Watanabe, Hitoshi

    1978-10-01

    A universal high-temperature heat treatment furnace for LMFBR advanced fuels was installed in Plutonium Fuel Laboratory, Oarai Research Establishment. Design, construction and performance of the apparatus are described. With the apparatus, heat treatment of the fuel under a controlled gas atmosphere and quenching of the fuel with blowing helium gas are possible. Equipment to measure impurity gas release of the fuel is also provided. Various plutonium enclosure techniques, e.g., a gas line filter with new exchange mechanics, have been developed. In performance test, results of the enclosure techniques are described. (author)

  3. 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.

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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%

  16. 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

  17. 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)

  18. 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.

  19. 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.

  20. 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

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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)

  6. 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

  7. 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.

  8. 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.)

  9. 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)

  10. 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

  11. 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)

  12. 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

  13. 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.)

  14. 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.

  15. 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)

  16. 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

  17. 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

  18. Temperature Induced Voltage Offset Drifts in Silicon Carbide Pressure Sensors

    Science.gov (United States)

    Okojie, Robert S.; Lukco, Dorothy; Nguyen, Vu; Savrun, Ender

    2012-01-01

    We report the reduction of transient drifts in the zero pressure offset voltage in silicon carbide (SiC) pressure sensors when operating at 600 C. The previously observed maximum drift of +/- 10 mV of the reference offset voltage at 600 C was reduced to within +/- 5 mV. The offset voltage drifts and bridge resistance changes over time at test temperature are explained in terms of the microstructure and phase changes occurring within the contact metallization, as analyzed by Auger electron spectroscopy and field emission scanning electron microscopy. The results have helped to identify the upper temperature reliable operational limit of this particular metallization scheme to be 605 C.

  19. Stress envelope of silicon carbide composites at elevated temperatures

    International Nuclear Information System (INIS)

    Nozawa, Takashi; Kim, Sunghun; Ozawa, Kazumi; Tanigawa, Hiroyasu

    2014-01-01

    To identify a comprehensive stress envelope, i.e., strength anisotropy map, of silicon carbide fiber-reinforced silicon carbide matrix composite (SiC/SiC composite) for practical component design, tensile and compressive tests were conducted using the small specimen test technique specifically tailored for high-temperature use. In-plane shear properties were, however, estimated using the off-axial tensile method and assuming that the mixed mode failure criterion, i.e., Tsai–Wu criterion, is valid for the composites. The preliminary test results indicate no significant degradation to either proportional limit stress (PLS) or fracture strength by tensile loading at temperatures below 1000 °C. A similarly good tolerance of compressive properties was identified at elevated temperatures, except for a slight degradation in PLS. With the high-temperature test data of tensile, compressive and in-plane shear properties, the stress envelopes at elevated temperatures were finally obtained. A slight reduction in the design limit was obvious at elevated temperatures when the compressive mode is dominant, whereas a negligibly small impact on the design is expected by considering the tensile loading case

  20. Stress envelope of silicon carbide composites at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Nozawa, Takashi, E-mail: nozawa.takashi67@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Omotedate, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Kim, Sunghun [Graduate School of Energy Science, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Ozawa, Kazumi; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate, Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2014-10-15

    To identify a comprehensive stress envelope, i.e., strength anisotropy map, of silicon carbide fiber-reinforced silicon carbide matrix composite (SiC/SiC composite) for practical component design, tensile and compressive tests were conducted using the small specimen test technique specifically tailored for high-temperature use. In-plane shear properties were, however, estimated using the off-axial tensile method and assuming that the mixed mode failure criterion, i.e., Tsai–Wu criterion, is valid for the composites. The preliminary test results indicate no significant degradation to either proportional limit stress (PLS) or fracture strength by tensile loading at temperatures below 1000 °C. A similarly good tolerance of compressive properties was identified at elevated temperatures, except for a slight degradation in PLS. With the high-temperature test data of tensile, compressive and in-plane shear properties, the stress envelopes at elevated temperatures were finally obtained. A slight reduction in the design limit was obvious at elevated temperatures when the compressive mode is dominant, whereas a negligibly small impact on the design is expected by considering the tensile loading case.

  1. 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

  2. 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)

  3. Low-temperature synthesis of silicon carbide powder using shungite

    International Nuclear Information System (INIS)

    Gubernat, A.; Pichor, W.; Lach, R.; Zientara, D.; Sitarz, M.; Springwald, M.

    2017-01-01

    The paper presents the results of investigation the novel and simple method of synthesis of silicon carbide. As raw material for synthesis was used shungite, natural mineral rich in carbon and silica. The synthesis of SiC is possible in relatively low temperature in range 1500–1600°C. It is worth emphasising that compared to the most popular method of SiC synthesis (Acheson method where the temperature of synthesis is about 2500°C) the proposed method is much more effective. The basic properties of products obtained from different form of shungite and in wide range of synthesis temperature were investigated. The process of silicon carbide formation was proposed and discussed. In the case of synthesis SiC from powder of raw materials the product is also in powder form and not requires any additional process (crushing, milling, etc.). Obtained products are pure and after grain classification may be used as abrasive and polishing powders. (Author)

  4. Low-temperature synthesis of silicon carbide powder using shungite

    Energy Technology Data Exchange (ETDEWEB)

    Gubernat, A.; Pichor, W.; Lach, R.; Zientara, D.; Sitarz, M.; Springwald, M.

    2017-07-01

    The paper presents the results of investigation the novel and simple method of synthesis of silicon carbide. As raw material for synthesis was used shungite, natural mineral rich in carbon and silica. The synthesis of SiC is possible in relatively low temperature in range 1500–1600°C. It is worth emphasising that compared to the most popular method of SiC synthesis (Acheson method where the temperature of synthesis is about 2500°C) the proposed method is much more effective. The basic properties of products obtained from different form of shungite and in wide range of synthesis temperature were investigated. The process of silicon carbide formation was proposed and discussed. In the case of synthesis SiC from powder of raw materials the product is also in powder form and not requires any additional process (crushing, milling, etc.). Obtained products are pure and after grain classification may be used as abrasive and polishing powders. (Author)

  5. 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.

  6. Decomposition of silicon carbide at high pressures and temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Daviau, Kierstin; Lee, Kanani K. M.

    2017-11-01

    We measure the onset of decomposition of silicon carbide, SiC, to silicon and carbon (e.g., diamond) at high pressures and high temperatures in a laser-heated diamond-anvil cell. We identify decomposition through x-ray diffraction and multiwavelength imaging radiometry coupled with electron microscopy analyses on quenched samples. We find that B3 SiC (also known as 3C or zinc blende SiC) decomposes at high pressures and high temperatures, following a phase boundary with a negative slope. The high-pressure decomposition temperatures measured are considerably lower than those at ambient, with our measurements indicating that SiC begins to decompose at ~ 2000 K at 60 GPa as compared to ~ 2800 K at ambient pressure. Once B3 SiC transitions to the high-pressure B1 (rocksalt) structure, we no longer observe decomposition, despite heating to temperatures in excess of ~ 3200 K. The temperature of decomposition and the nature of the decomposition phase boundary appear to be strongly influenced by the pressure-induced phase transitions to higher-density structures in SiC, silicon, and carbon. The decomposition of SiC at high pressure and temperature has implications for the stability of naturally forming moissanite on Earth and in carbon-rich exoplanets.

  7. 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.

  8. 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.

  9. 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

  10. High Temperature Dynamic Pressure Measurements Using Silicon Carbide Pressure Sensors

    Science.gov (United States)

    Okojie, Robert S.; Meredith, Roger D.; Chang, Clarence T.; Savrun, Ender

    2014-01-01

    Un-cooled, MEMS-based silicon carbide (SiC) static pressure sensors were used for the first time to measure pressure perturbations at temperatures as high as 600 C during laboratory characterization, and subsequently evaluated in a combustor rig operated under various engine conditions to extract the frequencies that are associated with thermoacoustic instabilities. One SiC sensor was placed directly in the flow stream of the combustor rig while a benchmark commercial water-cooled piezoceramic dynamic pressure transducer was co-located axially but kept some distance away from the hot flow stream. In the combustor rig test, the SiC sensor detected thermoacoustic instabilities across a range of engine operating conditions, amplitude magnitude as low as 0.5 psi at 585 C, in good agreement with the benchmark piezoceramic sensor. The SiC sensor experienced low signal to noise ratio at higher temperature, primarily due to the fact that it was a static sensor with low sensitivity.

  11. 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.

  12. Uranium peroxide precipitate drying temperature relationships

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, C.; Dyck, B., E-mail: chick_rodgers@cameco.com [Cameco Corp., Saskatoon, SK (Canada)

    2010-07-01

    Cameco Corporation is in the process of revitalizing the mill at its Key Lake operation in northern Saskatchewan. The current Key Lake process employs ammonia stripping and ammonia precipitation. As part of the revitalization, the company is considering installing strong acid stripping in solvent extraction as used at its Rabbit Lake operation. This change would lead to using hydrogen peroxide for uranium precipitation. As part of the process evaluation, tests were carried out to study how changes in the temperature of an indirect fired dryer affected the properties of uranium peroxide [yellowcake] precipitate. This paper discusses the results of the test work, including the relationships between drying temperature and the following: (author)

  13. Medium temperature reaction between lanthanide and actinide carbides and hydrogen

    International Nuclear Information System (INIS)

    Dean, G.; Lorenzelli, R.; Pascard, R.

    1964-01-01

    Hydrogen is fixed reversibly by the lanthanide and actinide mono carbides in the range 25 - 400 C, as for pure corresponding metals. Hydrogen goes into the carbides lattice through carbon vacancies and the total fixed amount is approximately equal to two hydrogen atoms per initial vacancy. Final products c.n thus be considered as carbo-hydrides of general formula M(C 1-x , H 2x ). The primitive CFC, NaCl type, structure remains unchanged but expands strongly in the case of actinide carbides. With lanthanide carbides, hydrogenation induces a phase transformation with reappearance of the metal structure (HCP). Hydrogen decomposition pressures of all the studied carbo-hydrides are greater than those of the corresponding di-hydrides. (authors) [fr

  14. Spheroidization of transition metal carbides in low temperature plasma

    International Nuclear Information System (INIS)

    Klinskaya, N.A.; Koroleva, E.B.; Petrunichev, V.A.; Rybalko, O.F.; Solov'ev, P.V.; Ugol'nikova, T.A.

    1986-01-01

    Plasma process of preparation of titanium, tungsten and chromium carbide spherical powders with the main particle size 40-80 μm is considered. Spheroidization degree, granulometric and phase composition of the product are investigated

  15. 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.)

  16. Carbide Transformation in Haynes 230 during Long-term Exposure at High Temperature

    International Nuclear Information System (INIS)

    Lee, Ho Jung; Kim, Hyunmyung; Hong, Sunghoon; Jang, Changheui

    2014-01-01

    Long-term aging behaviors of a solid solution hardened Ni-base superalloy, Haynes 230 at high temperature have not been fully investigated yet. In this study, long-term aging tests of Haynes 230 was carried out to evaluate microstructure changes especially in carbide evolution. In addition, its consequential effects on tensile property such as tensile strength and elongation were discussed. In Haynes 230, a nucleation of the secondary carbides was dominant at 800 .deg. C ageing while growth at 900 .deg. C ageing. In addition, after aging at 800 .deg. C, transition of primary W-rich M 6 C carbides (break down) were observed and it showed high W content (up to 70 at.% W) compared to un-aged W-rich M 6 C carbides (around 30 at.% W). Coarsened Cr- and Ni-rich phase surrounded by carbide depleted region and high W-rich M 6 C carbide along the grain boundary were formed only at 900 .deg. C after long-term exposure above 10000 h. Tensile strength of aged Haynes 230 increased at 800 .deg. C while decreased at 900 .deg. C due to the formation of secondary carbide within the grains at 800 .deg. C. Decrease in elongation would be resulted from the coarsened and continuous carbides at the grain boundary as well as Cr- and Ni-rich phase along the grain boundary

  17. 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

  18. Reaction of Oxygen with Chromium and Chromium Carbide at Low O2 Pressures and High Temperatures

    International Nuclear Information System (INIS)

    Hur, Dong O.; Kang, Sung G.; Paik, Young N.

    1984-01-01

    The oxidation rate of chromium carbide has been measured continuously using thermogravimetric analysis at different oxygen pressures ranging from 1.33x10 -2 to 2.67x10 -1 Pa O 2 at 1000-1300 .deg. C. The oxidation of pure chromium has also been studied between 1000-1300 .deg. C under 6.67x10 -2 Pa O 2 and compared with that of chromium carbide. The oxidation of chromium carbide showed a linear behavior which was different from that of chromium. The oxidation rate of chromium carbide increased with increasing temperature and oxygen pressure was lower than of pure chromium. Above 1200 .deg. C, the volatile oxide was formed and evaporated causing a weight loss. The compositions and morphology of the oxide were studied with X-ray diffractometer and scanning electron microscope, respectively. The morphology of oxide changed with varying temperature and pressure. The oxide scale was consisted of mainly two different layers of Cr 2 O 3 and CrO, and the properties of oxide scale were correlated with oxidation behavior. The oxide film formed in the above test condition has been detached from the carbide surface. The crack and pore were thought to be from CO gas evolving at the interface of chromium carbide and its oxide and the major factor of the linear behavior of chromium carbide

  19. Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

    Energy Technology Data Exchange (ETDEWEB)

    Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

    2008-07-01

    Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

  20. Uranium casting furnace automatic temperature control development

    International Nuclear Information System (INIS)

    Lind, R.F.

    1992-01-01

    Development of an automatic molten uranium temperature control system for use on batch-type induction casting furnaces is described. Implementation of a two-color optical pyrometer, development of an optical scanner for the pyrometer, determination of furnace thermal dynamics, and design of control systems are addressed. The optical scanning system is shown to greatly improve pyrometer measurement repeatability, particularly where heavy floating slag accumulations cause surface temperature gradients. Thermal dynamics of the furnaces were determined by applying least-squares system identification techniques to actual production data. A unity feedback control system utilizing a proportional-integral-derivative compensator is designed by using frequency-domain techniques. 14 refs

  1. High pressure low temperature hot pressing method for producing a zirconium carbide ceramic

    Science.gov (United States)

    Cockeram, Brian V.

    2017-01-10

    A method for producing monolithic Zirconium Carbide (ZrC) is described. The method includes raising a pressure applied to a ZrC powder until a final pressure of greater than 40 MPa is reached; and raising a temperature of the ZrC powder until a final temperature of less than 2200.degree. C. is reached.

  2. 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.

  3. Influence of rolling direction and carbide precipitation on IGSCC susceptibility in hydrogenated high temperature water

    International Nuclear Information System (INIS)

    Arioka, Koji; Yamada, Takuyo; Terachi, Takumi; Chiba, Goro

    2005-01-01

    IGSCC growth behaviors of austenitic stainless steels in hydrogenated high temperature water were studied using compact type specimens (0.5T for cold worked materials). The effect of cold rolling direction, alloy composition and carbide precipitation on crack growth behaviors was studied in hydrogenated high temperature water. Then, to examine the effect of cold work and carbide precipitation on IGSCC behaviors, the role of grain boundary sliding studied in high temperature air using CT specimens. The similar dependences of carbide precipitation and cold work on IGSCC and creep behaviors suggest that grain boundary sliding might play an important role by itself or in conjunction with other reactions such as crack tip dissolution etc. (author)

  4. Mode of carbide TiC-ZrC alloy fracture at various temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Paderno, V.N.; Lesnaya, M.I.; Martynenko, A.N.; Chugunova, S.I. (AN Ukrainskoj SSR, Kiev. Inst. Problem Materialovedeniya; Khersonskij Gosudarstvennyj Pedagogicheskij Inst. (Ukrainian SSR))

    1983-05-01

    Bending strength is studied for mutual TiC-ZrC alloys within a temperature range of 20-600 deg C. Structure of the material failure surface is studied from replicas by scanning and transmission electron microscopy. The data obtained are compared with concentration dependences of some physical properties of these alloys. Bending strength is shown to be minimum for the alloy with 40 mol % of zirconium carbide. It is stated that within the temperature range under study the carbide alloys undergo macroscopic brittle failure. The materials are characterized by a mixed type of failure with transcrystalline failure somewhat prevailing.

  5. Mode of carbide TiC-ZrC alloy fracture at various temperatures

    International Nuclear Information System (INIS)

    Paderno, V.N.; Lesnaya, M.I.; Martynenko, A.N.; Chugunova, S.I.

    1983-01-01

    Bending strength is studied for mutual TiC-ZrC alloys within a temperature range of 20-600 deg C. Structure of the material failure surface is studied from replicas by scanning and transmission electron microscopy. The data obtained are compared with concentration dependences of some physical properties of these alloys. Bending strength is shown to be minimum for the alloy with 40 mol % of zirconium carbide. It is stated that within the temperature range under study the carbide alloys undergo macroscopic brittle failure. The materials are characterized by a mixed type of failure with transcrystalline failure somewhat prevailing

  6. 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...

  7. 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

  8. A Silicon Carbide Wireless Temperature Sensing System for High Temperature Applications

    Science.gov (United States)

    Yang, Jie

    2013-01-01

    In this article, an extreme environment-capable temperature sensing system based on state-of-art silicon carbide (SiC) wireless electronics is presented. In conjunction with a Pt-Pb thermocouple, the SiC wireless sensor suite is operable at 450 °C while under centrifugal load greater than 1,000 g. This SiC wireless temperature sensing system is designed to be non-intrusively embedded inside the gas turbine generators, acquiring the temperature information of critical components such as turbine blades, and wirelessly transmitting the information to the receiver located outside the turbine engine. A prototype system was developed and verified up to 450 °C through high temperature lab testing. The combination of the extreme temperature SiC wireless telemetry technology and integrated harsh environment sensors will allow for condition-based in-situ maintenance of power generators and aircraft turbines in field operation, and can be applied in many other industries requiring extreme environment monitoring and maintenance. PMID:23377189

  9. Single-Crystal Tungsten Carbide in High-Temperature In-Situ Additive Manufacturing Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Kolopus, James A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Boatner, Lynn A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-18

    Nanoindenters are commonly used for measuring the mechanical properties of a wide variety of materials with both industrial and scientific applications. Typically, these instruments employ an indenter made of a material of suitable hardness bonded to an appropriate shaft or holder to create an indentation on the material being tested. While a variety of materials may be employed for the indenter, diamond and boron carbide are by far the most common materials used due to their hardness and other desirable properties. However, as the increasing complexity of new materials demands a broader range of testing capabilities, conventional indenter materials exhibit significant performance limitations. Among these are the inability of diamond indenters to perform in-situ measurements at temperatures above 600oC in air due to oxidation of the diamond material and subsequent degradation of the indenters mechanical properties. Similarly, boron carbide also fails at high temperature due to fracture. [1] Transition metal carbides possess a combination of hardness and mechanical properties at high temperatures that offer an attractive alternative to conventional indenter materials. Here we describe the technical aspects for the growth of single-crystal tungsten carbide (WC) for use as a high-temperature indenter material, and we examine a possible approach to brazing these crystals to a suitable mount for grinding and attachment to the indenter instrument. The use of a by-product of the recovery process is also suggested as possibly having commercial value.

  10. 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

  11. The effects of annealing temperature and cooling rate on carbide precipitation behavior in H13 hot-work tool steel

    International Nuclear Information System (INIS)

    Kang, Minwoo; Park, Gyujin; Jung, Jae-Gil; Kim, Byung-Hoon; Lee, Young-Kook

    2015-01-01

    Highlights: • Unexpected Mo carbides formed during slow cooling from low annealing temperatures. • Mo carbides formed during the migration of Mo for a transition of Cr-rich carbide. • Mo carbides were precipitated at the boundaries of M 7 C 3 carbides and ferrite grains. • Annealing conditions for the precipitation of Mo carbides were discussed. - Abstract: The precipitation behavior of H13 hot-work tool steel was investigated as a function of both annealing temperature and cooling rate through thermodynamic calculations and microstructural analyses using transmission and scanning electron microscope and a dilatometer. The V-rich MC carbide and Cr-rich M 7 C 3 and M 23 C 6 carbides were observed in all annealed specimens regardless of annealing and cooling conditions, as expected from an equilibrium phase diagram of the steel used. However, Mo-rich M 2 C and M 6 C carbides were unexpectedly precipitated at a temperature between 675 °C and 700 °C during slow cooling at a rate of below 0.01 °C/s from the annealing temperatures of 830 °C and below. The solubility of Mo in both M 7 C 3 and ferrite reduces with decreasing temperature during cooling. Mo atoms diffuse out of both M 7 C 3 and ferrite, and accumulate locally at the interface between M 7 C 3 and ferrite. Mo carbides were form at the interface of M 7 C 3 carbides during the transition of Cr-rich M 7 C 3 to stable M 23 C 6

  12. Boron Carbide: Stabilization of Highly-Loaded Aqueous Suspensions, Pressureless Sintering, and Room Temperature Injection Molding

    Science.gov (United States)

    Diaz-Cano, Andres

    Boron carbide (B4C) is the third hardest material after diamond and cubic boron nitride. It's unique combination of properties makes B4C a highly valuable material. With hardness values around 35 MPa, a high melting point, 2450°C, density of 2.52 g/cm3, and high chemical inertness, boron carbide is used in severe wear components, like cutting tools and sandblasting nozzles, nuclear reactors' control rots, and finally and most common application, armor. Production of complex-shaped ceramic component is complex and represents many challenges. Present research presents a new and novel approach to produce complex-shaped B4C components. Proposed approach allows forming to be done at room temperatures and under very low forming pressures. Additives and binder concentrations are kept as low as possible, around 5Vol%, while ceramics loadings are maximized above 50Vol%. Given that proposed approach uses water as the main solvent, pieces drying is simple and environmentally safe. Optimized formulation allows rheological properties to be tailored and adjust to multiple processing approaches, including, injection molding, casting, and additive manufacturing. Boron carbide samples then were pressureless sintered. Due to the high covalent character of boron carbide, multiples sintering aids and techniques have been proposed in order to achieve high levels of densification. However, is not possible to define a clear sintering methodology based on literature. Thus, present research developed a comprehensive study on the effect of multiple sintering aids on the densification of boron carbide when pressureless sintered. Relative densities above 90% were achieved with values above 30MPa in hardness. Current research allows extending the uses and application of boron carbide, and other ceramic systems, by providing a new approach to produce complex-shaped components with competitive properties.

  13. Catastrophic degradation of the interface of epitaxial silicon carbide on silicon at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Pradeepkumar, Aiswarya; Mishra, Neeraj; Kermany, Atieh Ranjbar; Iacopi, Francesca [Queensland Micro and Nanotechnology Centre and Environmental Futures Research Institute, Griffith University, Nathan QLD 4111 (Australia); Boeckl, John J. [Materials and Manufacturing Directorate, Air Force Research Laboratories, Wright-Patterson Air Force Base, Ohio 45433 (United States); Hellerstedt, Jack; Fuhrer, Michael S. [Monash Centre for Atomically Thin Materials, Monash University, Monash, VIC 3800 (Australia)

    2016-07-04

    Epitaxial cubic silicon carbide on silicon is of high potential technological relevance for the integration of a wide range of applications and materials with silicon technologies, such as micro electro mechanical systems, wide-bandgap electronics, and graphene. The hetero-epitaxial system engenders mechanical stresses at least up to a GPa, pressures making it extremely challenging to maintain the integrity of the silicon carbide/silicon interface. In this work, we investigate the stability of said interface and we find that high temperature annealing leads to a loss of integrity. High–resolution transmission electron microscopy analysis shows a morphologically degraded SiC/Si interface, while mechanical stress measurements indicate considerable relaxation of the interfacial stress. From an electrical point of view, the diode behaviour of the initial p-Si/n-SiC junction is catastrophically lost due to considerable inter-diffusion of atoms and charges across the interface upon annealing. Temperature dependent transport measurements confirm a severe electrical shorting of the epitaxial silicon carbide to the underlying substrate, indicating vast predominance of the silicon carriers in lateral transport above 25 K. This finding has crucial consequences on the integration of epitaxial silicon carbide on silicon and its potential applications.

  14. High temperature oxidation of carbide-carbon materials of NbC-C, NbC-TiC-C systems

    International Nuclear Information System (INIS)

    Afonin, Yu.D.; Shalaginov, V.N.; Beketov, A.R.

    1981-01-01

    The effect of titanium carbide additions on the oxidation of carbide - carbon composition NbC-TiC-C in oxygen under the pressure of 10 mm Hg and in the air at atmospheric pressure in the temperature range 800-1300 deg is studied. It is shown that the region of negative temperature coefficient during oxidation in the system NbC+C is determined by the processes of sintering and polymorphous transformation. The specific character of the oxide film, formed during oxidation of Nbsub(x)Tisub(y)C+C composites is connected with non-equilibrium nature of carbide grain in its composition. Carbon gasification takes place with the formation of carbon dioxide. Composite materials, containing titanium carbide in complex carbide up to 50-83 mol. %, are the most corrosion resisting ones [ru

  15. High-temperature thermal conductivity of uranium chromite and uranium niobate

    International Nuclear Information System (INIS)

    Fedoseev, D.V.; Varshavskaya, I.G.; Lavrent'ev, A.V.; Oziraner, S.N.; Kuznetsova, D.G.

    1979-01-01

    The technique of determining thermal conductivity coefficient of uranium niobate and uranium chromite on heating with laser radiation is described. Determined is the coefficient of free-convective heat transfer (with provision for a conduction component) by means of a standard specimen. The thermal conductivity coefficients of uranium chromite and niobate were measured in the 1300-1700 K temperature range. The results are presented in a diagram form. It has been calculated, that the thermal conductivity coefficient for uranium niobate specimens is greater in comparison with uranium chromite specimens. The thermal conductivity coefficients of the materials mentioned depend on temperature very slightly. Thermal conductivity of the materials considerably depends on their porosity. The specimens under investigation were fabricated by the pressing method and had the following porosity: uranium chromite - 30 %, uranium niobate - 10 %. Calculation results show, that thermal conductivity of dense uranium chromite is higher than thermal conductivity of dense uranium niobate. The experimental error equals approximately 20 %, that is mainly due to the error of measuring the temperature equal to +-25 deg, with a micropyrometer

  16. 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)

  17. Low-temperature synthesis of homogeneous nanocrystalline cubic silicon carbide films

    International Nuclear Information System (INIS)

    Cheng Qijin; Xu, S.

    2007-01-01

    Silicon carbide films are fabricated by inductively coupled plasma chemical vapor deposition from feedstock gases silane and methane heavily diluted with hydrogen at a low substrate temperature of 300 deg. C. Fourier transform infrared absorption spectroscopy, Raman spectroscopy, x-ray photoelectron spectroscopy, and high-resolution transmission electron microscopy analyses show that homogeneous nanocrystalline cubic silicon carbide (3C-SiC) films can be synthesized at an appropriate silane fraction X[100%xsilane flow(SCCM)/silane+methane flow(SCCM)] in the gas mixture. The achievement of homogeneous nanocrystalline 3C-SiC films at a low substrate temperature of 300 deg. C is a synergy of a low deposition pressure (22 mTorr), high inductive rf power (2000 W), heavy dilution of feedstock gases silane and methane with hydrogen, and appropriate silane fractions X (X≤33%) in the gas mixture employed in our experiments

  18. 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

  19. 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 ...

  20. High Temperature All Silicon-Carbide (SiC) DC Motor Drives for Venus Exploration Vehicles, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This Small Business Innovation Research Phase I project seeks to prove the feasibility of creating high-temperature silicon-carbide (SiC) based motor drives for...

  1. Room-temperature Electrochemical Synthesis of Carbide-derived Carbons and Related Materials

    Energy Technology Data Exchange (ETDEWEB)

    Gogotsi, Yury [Drexel Univ., Philadelphia, PA (United States). Nanomaterials Group. Materials Science and Engineering Dept.

    2015-02-28

    This project addresses room-temperature electrochemical etching as an energy-efficient route to synthesis of 3D nanoporous carbon networks and layered 2D carbons and related structures, as well as provides fundamental understanding of structure and properties of materials produced by this method. Carbide-derived-carbons (CDCs) are a growing class of nanostructured carbon materials with properties that are desirable for many applications, such as electrical energy and gas storage. The structure of these functional materials is tunable by the choice of the starting carbide precursor, synthesis method, and process parameters. Moving from high-temperature synthesis of CDCs through vacuum decomposition above 1400°C and chlorination above 400°C, our studies under the previous DOE BES support led to identification of precursor materials and processing conditions for CDC synthesis at temperatures as low as 200°C, resulting in amorphous and highly reactive porous carbons. We also investigated synthesis of monolithic CDC films from carbide films at 250-1200°C. The results of our early studies provided new insights into CDC formation, led to development of materials for capacitive energy storage, and enabled fundamental understanding of the electrolyte ions confinement in nanoporous carbons.

  2. 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

  3. An X-ray photoelectron spectroscopic study of a nitric acid/argon ion cleaned uranium metal surface at elevated temperature

    International Nuclear Information System (INIS)

    Paul, A.J.; Sherwood, P.M.A.

    1987-01-01

    X-ray photoelectron spectroscopy has been used to study the surface of uranium metal cleaned by nitric acid treatment and argon ion etching, followed by heating in a high vacuum. The surface is shown to contain UOsub(2-x) species over the entire temperature range studied. Heating to temperatures in the range 400-600 0 C generates a mixture of this oxide, the metal and a carbide and/or oxycarbide species. (author)

  4. High temperature Hexoloy{trademark} SX silicon carbide. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, G.V.; Lau, S.K.; Storm, R.S. [Carborundum Co., Niagara Falls, NY (United States)

    1994-09-01

    HEXOLOY{reg_sign} SX-SiC, fabricated with Y and Al containing compounds as sintering aids, has been shown to possess significantly improved strength and toughness over HEXOLOY{reg_sign}SA-SiC. This study was undertaken to establish and benchmark the complete mechanical property database of a first generation material, followed by a process optimization task to further improve the properties. Mechanical characterization on the first generation material indicated that silicon-rich pools, presumably formed as a reaction product during sintering, controlled the strength from room temperature to 1,232 C. At 1,370 C in air, the material was failing due to a glass-phase formation at the surface. This glass-phase formation was attributed to the reaction of yttrium aluminates, which exist as a second phase in the material, with the ambient. This process was determined to be a time-dependent one that leads to slow crack growth. Fatigue experiments clearly indicated that the slow crack growth driven by the reaction occurred only at temperatures >1,300 C, above the melting point of the glass phase. Process optimization tasks conducted included the selection of the best SiC powder source, studies on mixing/milling conditions for SiC powder with the sintering aids, and a designed experiment involving a range of sintering and post-treatment conditions. The optimization study conducted on the densification variables indicated that lower sintering temperatures and higher post-treatment pressures reduce the Si-rich pool formation, thereby improving the room-temperature strength. In addition, it was also determined that furnacing configuration and atmosphere were critical in controlling the Si-rich formation.

  5. High temperature heat capacities and electrical conductivities of boron carbides

    International Nuclear Information System (INIS)

    Matsui, Tsuneo; Arita, Yuri; Naito, Keiji; Imai, Hisashi

    1991-01-01

    The heat capacities and the electrical conductivities of B x C(x=3, 4, 5) were measured by means of direct heating pulse calorimetry in the temperature range from 300 to 1500 K. The heat capacities of B x C increased with increasing x value. This increase in the heat capacity is probably related to the change of the lattice vibration mode originated from the reduction of the stiffness of the intericosahedral chain accompanied with a change from C-B-C to C-B-B chains. A linear relationship between the logarithm of σT (σ is the electrical conductivity and T is the absolute temperature) of B x C and the reciprocal temperature was observed, indicating the presence of small polaron hopping as the predominant conduction mechanism. The electrical conductivity of B x C also increased with increasing x value (from 4 to 5) due to an increase of the polaron hopping of holes between carbon atoms at geometrically nonequivalent sites, since these nonequivalent sites of carbon atoms were considered to increase in either B 11 C icosahedra or in icosahedral chains with increasing x. The electrical conductivity of B 3 C was higher than that of B 4 C, which is probably due to the precipitation of high-conducting carbon. The thermal conductivity and the thermodynamic quantities of B 4 C were also determined precisely from the heat capacity value. (orig.)

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

    Indian Academy of Sciences (India)

    Unknown

    From the phonon frequencies, their Debye temperatures are evaluated. Further, ... Keywords. Uranium chalcogenides; p-wave electronic superconductor; phonon frequency; Debye tempera- ture; spin ... to the ionic crystals of similar structure.

  7. High temperature behavior of metallic inclusions in uranium dioxide

    International Nuclear Information System (INIS)

    Yang, R.L.

    1980-08-01

    The object of this thesis was to construct a temperature gradient furnace to simulate the thermal conditions in the reactor fuel and to study the migration of metallic inclusions in uranium oxide under the influence of temperature gradient. No thermal migration of molybdenum and tungsten inclusions was observed under the experimental conditions. Ruthenium inclusions, however, dissolved and diffused atomically through grain boundaries in slightly reduced uranium oxide. An intermetallic compound (probably URu 3 ) was formed by reaction of Ru and UO/sub 2-x/. The diffusivity and solubility of ruthenium in uranium oxide were measured

  8. Silicon carbide production by Self-Propagating High Temperature (SHS) technique

    International Nuclear Information System (INIS)

    Lima, Eduardo de Souza; Schneider, Pedro Luiz; Mattoso, Irani Guedes; Costa, Carlos Roberto Correia da; Louro, Luis Henrique Leme

    1997-01-01

    Samples of silicon carbide (SiC) were synthesized from a mixture of silicon and carbon powders, using the Self-Propagating High Temperature Synthesis (SHS) technique. Three mixtures were tried, using silicon particles of the same average size but carbon particles of different average sizes. The method tried is characterized by an ignition temperature of 1450 deg C and the short duration of the synthesis ( 2-3 min). The samples were characterized by X-ray diffraction and scattering electron microscopy. (author)

  9. Silicon Carbide-Based Hydrogen Gas Sensors for High-Temperature Applications

    Directory of Open Access Journals (Sweden)

    Sangchoel Kim

    2013-10-01

    Full Text Available We investigated SiC-based hydrogen gas sensors with metal-insulator-semiconductor (MIS structure for high temperature process monitoring and leak detection applications in fields such as the automotive, chemical and petroleum industries. In this work, a thin tantalum oxide (Ta2O5 layer was exploited with the purpose of sensitivity improvement, because tantalum oxide has good stability at high temperature with high permeability for hydrogen gas. Silicon carbide (SiC was used as a substrate for high-temperature applications. We fabricated Pd/Ta2O5/SiC-based hydrogen gas sensors, and the dependence of their I-V characteristics and capacitance response properties on hydrogen concentrations were analyzed in the temperature range from room temperature to 500 °C. According to the results, our sensor shows promising performance for hydrogen gas detection at high temperatures.

  10. A Computational-Experimental Study of Plasma Processing of Carbides at High Temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Bronson, Arturo [Univ. of Texas, El Paso, TX (United States); Kumar, Vinod [Univ. of Texas, El Paso, TX (United States)

    2016-02-01

    appears to grow with Ti ions migrating outward from the Ti3AlC/Ti2AlC/TiC core and oxygen ions diffusing inwardly toward the core. The transient temperature distribution of a cylindrical, carbide packed bed (i.e., B4C) was simulated with COMSOL software to determine the response of the bed to a sudden temperature spike exposed to the outer wall of the bed. The temperature distribution of B4C was similarly heated and compared with Hf and Zr metal. The thermal conductivity of Hf and Zr is higher than the B4C packed bed and hence they respond quicker than B4C. The packed bed still takes approximately 1200 s to plateau the temperature distribution between the cylinder surfaces to the centerline of the carbide packed bed of 5 cm diameter. Though the modeling of the distributions in the carbide packed bed gives an understanding of the transient heat response behavior driven by radiation, the effect of the plasma on the surface temperature of individual carbide particles needs further investigation to understand the plasma contribution to densification of a carbide packed bed.

  11. 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

  12. Structural modifications induced by ion irradiation and temperature in boron carbide B{sub 4}C

    Energy Technology Data Exchange (ETDEWEB)

    Victor, G., E-mail: g.victor@ipnl.in2p3.fr [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); Pipon, Y.; Bérerd, N. [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); Institut Universitaire de Technologie (IUT) Lyon-1, Université Claude Bernard Lyon 1, 69622 Villeurbanne Cedex (France); Toulhoat, N. [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); CEA-DEN, Saclay, 91191 Gif-sur-Yvette (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); Djourelov, N. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, 72 Tzarigradsko chaussee blvd, BG-1784 Sofia (Bulgaria); ELI-NP, IFIN-HH, 30 Reactorului Str, MG-6 Bucharest-Magurele (Romania); Miro, S. [CEA-DEN, Service de Recherches de Métallurgie Physique, Laboratoire JANNUS, F-91191 Gif-sur-Yvette (France); Baillet, J. [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); Pradeilles, N.; Rapaud, O.; Maître, A. [SPCTS, UMR CNRS 7315, Centre Européen de la céramique, University of Limoges (France); Gosset, D. [CEA, Saclay, DMN-SRMA-LA2M, 91191 Gif-sur-Yvette (France)

    2015-12-15

    Already used as neutron absorber in the current French nuclear reactors, boron carbide (B{sub 4}C) is also considered in the future Sodium Fast Reactors of the next generation (Gen IV). Due to severe irradiation conditions occurring in these reactors, it is of primary importance that this material presents a high structural resistance under irradiation, both in the ballistic and electronic damage regimes. Previous works have shown an important structural resistance of boron carbide even at high neutron fluences. Nevertheless, the structural modification mechanisms due to irradiation are not well understood. Therefore the aim of this paper is to study structural modifications induced in B{sub 4}C samples in different damage regimes. The boron carbide pellets were shaped and sintered by using spark plasma sintering method. They were then irradiated in several conditions at room temperature or 800 °C, either by favoring the creation of ballistic damage (between 1 and 3 dpa), or by favoring the electronic excitations using 100 MeV swift iodine ions (S{sub e} ≈ 15 keV/nm). Ex situ micro-Raman spectroscopy and Doppler broadening of annihilation radiation technique with variable energy slow positrons were coupled to follow the evolution of the B{sub 4}C structure under irradiation.

  13. Structural modifications induced by ion irradiation and temperature in boron carbide B4C

    Science.gov (United States)

    Victor, G.; Pipon, Y.; Bérerd, N.; Toulhoat, N.; Moncoffre, N.; Djourelov, N.; Miro, S.; Baillet, J.; Pradeilles, N.; Rapaud, O.; Maître, A.; Gosset, D.

    2015-12-01

    Already used as neutron absorber in the current French nuclear reactors, boron carbide (B4C) is also considered in the future Sodium Fast Reactors of the next generation (Gen IV). Due to severe irradiation conditions occurring in these reactors, it is of primary importance that this material presents a high structural resistance under irradiation, both in the ballistic and electronic damage regimes. Previous works have shown an important structural resistance of boron carbide even at high neutron fluences. Nevertheless, the structural modification mechanisms due to irradiation are not well understood. Therefore the aim of this paper is to study structural modifications induced in B4C samples in different damage regimes. The boron carbide pellets were shaped and sintered by using spark plasma sintering method. They were then irradiated in several conditions at room temperature or 800 °C, either by favoring the creation of ballistic damage (between 1 and 3 dpa), or by favoring the electronic excitations using 100 MeV swift iodine ions (Se ≈ 15 keV/nm). Ex situ micro-Raman spectroscopy and Doppler broadening of annihilation radiation technique with variable energy slow positrons were coupled to follow the evolution of the B4C structure under irradiation.

  14. Low temperature sintering of hyperstoichiometric uranium dioxide

    International Nuclear Information System (INIS)

    Chevrel, H.

    1991-12-01

    In the lattice of uranium dioxide with hyperstoichiometric oxygen content (UO 2+x ), each additional oxygen atoms is introduced by shifting two anions from normal sites to interstitial ones, thereby creating two oxygen vacancies. The point defects then combine to form complex defects comprising several interstitials and vacancies. The group of anions (3x) in the interstitial position participate in equilibria promoting the creation of uranium vacancies thereby considerably increasing uranium self-diffusion. However, uranium grain boundaries diffusion governs densification during the first two stages of sintering of uranium dioxide with hyperstoichiometric oxygen content, i.e., up to 93% of the theoretical density. Surface diffusion and evaporation-condensation, which are considerably accentuated by the hyperstoichiometric deviation, play an active role during sintering by promoting crystalline growth during the second and third stages of sintering. U 8 O 8 can be added to adjust the stoichiometry and to form a finely porous structure and thus increase the pore area subjected to surface phenomena. The composition with an O/U ratio equal to 2.25 is found to densify the best, despite a linear growth in sintering activation energy with hyperstoichiometric oxygen content, increasing from 300 kj.mol -1 for UO 2.10 to 440 kJ.mol -1 for UO 2.25 . Seeds can be introduced to obtain original microstructures, for example the presence of large grains in small-grain matrix

  15. 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

  16. 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)

  17. High-temperature effect of hydrogen on sintered alpha-silicon carbide

    Science.gov (United States)

    Hallum, G. W.; Herbell, T. P.

    1986-01-01

    Sintered alpha-silicon carbide was exposed to pure, dry hydrogen at high temperatures for times up to 500 hr. Weight loss and corrosion were seen after 50 hr at temperatures as low as 1000 C. Corrosion of SiC by hydrogen produced grain boundary deterioration at 1100 C and a mixture of grain and grain boundary deterioration at 1300 C. Statistically significant strength reductions were seen in samples exposed to hydrogen for times greater than 50 hr and temperatures above 1100 C. Critical fracture origins were identified by fractography as either general grain boundary corrision at 1100 C or as corrosion pits at 1300 C. A maximum strength decrease of approximately 33 percent was seen at 1100 and 1300 C after 500 hr exposure to hydrogen. A computer assisted thermodynamic program was also used to predict possible reaction species of SiC and hydrogen.

  18. 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.

  19. 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

  20. High-temperature mechanical properties of a uniaxially reinforced zircon-silicon carbide composite

    International Nuclear Information System (INIS)

    Singh, R.N.

    1990-01-01

    This paper reports that mechanical properties of a monolithic zircon ceramic and zircon-matrix composites uniaxially reinforced with either uncoated or BN-coated silicon carbide monofilaments were measured in flexure between 25 degrees and 1477 degrees C. Monolithic zircon ceramics were weak and exhibited a brittle failure up to abut 1300 degrees C. An increasing amount of the plastic deformation was observed before failure above about 1300 degrees C. In contrast, composites reinforced with either uncoated or BN-coated Sic filaments were stronger and tougher than the monolithic zircon at all test temperatures between 25 degrees and 1477 degrees. The ultimate strength and work-of-fracture of composite samples decreased with increasing temperature. A transgranular matrix fracture was shown by the monolithic and composite samples tested up to about 1200 degrees C, whereas an increasing amount of the intergranular matrix fracture was displayed above 1200 degrees C

  1. Method for analyzing passive silicon carbide thermometry with a continuous dilatometer to determine irradiation temperature

    Science.gov (United States)

    Campbell, Anne A.; Porter, Wallace D.; Katoh, Yutai; Snead, Lance L.

    2016-03-01

    Silicon carbide is used as a passive post-irradiation temperature monitor because the irradiation defects will anneal out above the irradiation temperature. The irradiation temperature is determined by measuring a property change after isochronal annealing, i.e., lattice spacing, dimensions, electrical resistivity, thermal diffusivity, or bulk density. However, such methods are time-consuming since the steps involved must be performed in a serial manner. This work presents the use of thermal expansion from continuous dilatometry to calculate the SiC irradiation temperature, which is an automated process requiring minimal setup time. Analysis software was written that performs the calculations to obtain the irradiation temperature and removes possible user-introduced error while standardizing the analysis. This method has been compared to an electrical resistivity and isochronal annealing investigation, and the results revealed agreement of the calculated temperatures. These results show that dilatometry is a reliable and less time-intensive process for determining irradiation temperature from passive SiC thermometry.

  2. Microstructures of beta-silicon carbide after irradiation creep deformation at elevated temperatures

    International Nuclear Information System (INIS)

    Katoh, Yutai; Kondo, Sosuke; Snead, Lance L.

    2008-01-01

    Microstructures of silicon carbide were examined by transmission electron microscopy (TEM) after creep deformation under neutron irradiation. Thin strip specimens of polycrystalline and monocrystalline, chemically vapor-deposited, beta-phase silicon carbide were irradiated in the high flux isotope reactor to 0.7-4.2 dpa at nominal temperatures of 640-1080 deg. C in an elastically pre-strained bend stress relaxation configuration with the initial stress of ∼100 MPa. Irradiation creep caused permanent strains of 0.6 to 2.3 x 10 -4 . Tensile-loaded near-surface portions of the crept specimens were examined by TEM. The main microstructural features observed were dislocation loops in all samples, and appeared similar to those observed in samples irradiated in non-stressed conditions. Slight but statistically significant anisotropy in dislocation loop microstructure was observed in one irradiation condition, and accounted for at least a fraction of the creep strain derived from the stress relaxation. The estimated total volume of loops accounted for 10-45% of the estimated total swelling. The results imply that the early irradiation creep deformation of SiC observed in this work was driven by anisotropic evolutions of extrinsic dislocation loops and matrix defects with undetectable sizes

  3. Preplastic strain effect on chromium carbides precipitation of type 316 stainless steel during high-temperature ageing

    International Nuclear Information System (INIS)

    Mao, X.; Zhao, W.

    1992-01-01

    Long exposure of Type 316 stainless steel to elevated temperature (400-900 o C) is known to cause high-temperature embrittlement due to chromium carbides and σ-phase precipitating in grain boundaries. Numerous investigations have been published on the mechanical properties and microstructure changes occurring during such exposure. However, no investigations exist on the preplastic deformation effect on chromium carbide precipitation in the grain matrix and grain boundary during high-temperature ageing of Type 316 stainless steel and then its effects on the room-temperature tensile properties. Since the stainless steel sometimes is deformed before use at high temperatures, it is necessary to study the preplastic strain effect of the stainless steel on the microstructure change and mechanical property change during high-temperature exposure. The purpose of the present investigation was to carry out such a study. The conclusions reached are as follows. First, chromium carbides are precipitated in deformation lines (slip lines) and then the amount of chromium carbides precipitation in the grain boundary is relatively reduced in predeformed stainless steel after ageing. Secondly, plastic strain pretreatments of and subsequent ageing treatments of Type 316 stainless steel can improve its tensile ductility. Finally, secondary cracking of aged stainless steel occurs in a normal tensile test. The secondary cracking can be reduced by adding preplastic strain into the material. (Author)

  4. Release of gases from uranium metal at high temperatures

    International Nuclear Information System (INIS)

    Sayi, Y.S.; Ramanjaneyulu, P.S.; Yadav, C.S.; Shankaran, P.S.; Chhapru, G.C.; Ramakumar, K.L.; Venugopal, V.

    2008-01-01

    Depending on the ambient environmental conditions, different gaseous species could get entrapped in uranium metal ingots or pellets. On heating, melting or vapourising uranium metal, these get released and depending on the composition, may cause detrimental effects either within the metal matrix itself or on the surrounding materials/environment. For instance, these gases may affect the performance of the uranium metal, which is used as fuel in the heavy water moderated research reactors, CIRUS and DHRUVA. Hence, detailed investigations have been carried out on the release of gases over a temperature range 875-1500 K employing hot vacuum extraction technique, in specimen uranium pellets made from uranium rods/ingots. Employing an on-line quadrupole mass spectrometer, the analysis of released gases was carried out. The isobaric interference between carbon monoxide and nitrogen at m/e = 28 in the mass spectrometric analysis has been resolved by considering their fragmentation patterns. Since no standards are available to evaluate the results, only the reproducibility is tested. The precision (relative standard deviation at 3σ level) of the method is ±5%. The minimum detectable gas content employing the method is 5.00 x 10 -09 m 3 . About 4 x 10 -04 m 3 /kg of gas is released from uranium pellets, with hydrogen as the main constituent. The gas content increases with storage in air

  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. High temperature diffusion of hafnium in tungsten and a tungsten-hafnium carbide alloy

    International Nuclear Information System (INIS)

    Ozaki, Y.; Zee, R.H.

    1994-01-01

    Refractory metals and ceramics are used extensively in energy systems due to their high temperature properties. This is particularly important in direct conversion systems where thermal to electric conversion efficiency is a direct function of temperature. Tungsten, which has the highest melting temperature among elemental metals, does not possess sufficient creep resistance at temperature above 1,600 K. Different dispersion strengthened tungsten alloys have been developed to extend the usefulness of tungsten to higher temperatures. One of these alloys, tungsten with 0.4 mole percent of finely dispersed HfC particles (W-HfC), has the optimum properties for high temperature applications. Hafnium carbide is used as the strengthening agent due to its high chemical stability and its compatibility with tungsten. The presence of HfC particles retards the rate of grain growth as well as restricting dislocation motion. Both of which are beneficial for creep resistance. The long term behavior of this alloy depends largely on the evolution of its microstructure which is governed by the diffusion of its constituents. Data on the diffusion of carbon in tungsten and tungsten self-diffusion are available, but no direct measurements have been made on the diffusion of hafnium in tungsten. The only diffusion data available are estimated from a coarsening study and these data are highly unreliable. In this study, the diffusion behavior of hafnium in pure tungsten and in a W-HfC alloy was directly measured by means of Secondary Ion Mass Spectroscopy (SIMS). The selection of the W-HfC alloy is due to its importance in high temperature engineering applications, and its higher recrystallization temperature. The presence of HfC particles in tungsten restricts grain growth resulting in better high temperature creep resistance. The higher recrystallization temperature allows measurements to be made over a wider range of temperatures at a relatively constant grain size

  7. Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Rohbeck, Nadia, E-mail: nadia.rohbeck@manchester.ac.uk; Xiao, Ping, E-mail: p.xiao@manchester.ac.uk

    2016-09-15

    The HTR design envisions fuel operating temperatures of up to 1000 °C and in case of an accident even 1600 °C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600 °C up to 2200 °C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100 °C onwards, but initial signs of porosity formation were visible already at 1800 °C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600 °C to 2000 °C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500 °C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500 °C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

  8. Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

    International Nuclear Information System (INIS)

    Rohbeck, Nadia; Xiao, Ping

    2016-01-01

    The HTR design envisions fuel operating temperatures of up to 1000 °C and in case of an accident even 1600 °C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600 °C up to 2200 °C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100 °C onwards, but initial signs of porosity formation were visible already at 1800 °C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600 °C to 2000 °C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500 °C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500 °C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

  9. Synthesis of titanium carbide from wood by self-propagating high temperature synthesis

    Directory of Open Access Journals (Sweden)

    Sutham Niyomwas

    2010-05-01

    Full Text Available Titanium carbide (TiC particles were obtained in situ by a self-propagating high temperature synthesis (SHS of wooddust with TiO2 and Mg. The reaction was carried out in a SHS reactor under static argon gas at the pressure of 0.5 MPa. Thestandard Gibbs energy minimization method was used to calculate the equilibrium composition of the reacting species. Theeffects of increasing Mg mole ratio to the precursor mixture of TiO2 and wood dusts were investigated. XRD and SEManalyses indicate a complete reaction of the precursors to yield TiC-MgO as a product composite. The synthesized compositeswere leached with 0.1M HCl acid solution to obtain TiC particles as final products.

  10. Evaluation of the Mechanical Performance of Silicon Carbide in TRISO Fuel at High Temperatures

    International Nuclear Information System (INIS)

    Rohbeck, N.; Xiao, P.

    2014-01-01

    The HTR design envisions fuel operating temperatures of up to 1000°C and in case of an accident even 1600°C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600°C up to 2200°C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100°C onwards, but initial signs of porosity formation were visible already at 1800°C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600°C to 2000°C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in-situ up to 500°C using a high temperature nanoindentation facility. This approach allows conducting numerous tests on small sample volumes and thus promises to improve our knowledge of irradiation effects on the mechanical properties. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500°C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value. (author)

  11. Study on the Chemical Compatibility of Silicon Carbide regarding Uranium Dioxide

    International Nuclear Information System (INIS)

    Braun, J.; Sauder, C.; Guéneau, C.; Alpettaz, T.; Balbaud, F.; Allegri, P.; Brackx, E.

    2013-01-01

    Conclusions/perspectives: • SiC/SiC composites are considered in SFR reactors as structure materials. • Knudsen Cell Mass Spectrometry combined with thermodynamic calculations to investigate the compatibility of UO 2+x regarding SiC: - Open system; - Signals overlapping issues →New spectrometer to work at different energies; - Temperature limited by the molecular beam regime →Improvement of pumping capacity. • UO 2.15 /SiC at 1500°C: - Strong interaction; - Reduction of UO 2+x by SiC; - CO formation. To a lesser extend SiO, CO 2 ; - Formation of USi 1.88 ; U 3 Si 5 ; U 3 Si 2 C 2 ; UC; UC 2 . • UO 2.02 /SiC: - 1200°C – Limited reaction; • 1350/1500°C – Three areas (UC/USi x on the periphery - UO 2 at the center - USi x in-between)/ Characteristic of a gaseous diffusion; • 1650°C – Crucible strongly-attacked – liquid-phase detected. Perspectives: - Diffusion couples: closed system; - Alternative crucible (Ta -W)

  12. Method of enhanced lithiation of doped silicon carbide via high temperature annealing in an inert atmosphere

    Science.gov (United States)

    Hersam, Mark C.; Lipson, Albert L.; Bandyopadhyay, Sudeshna; Karmel, Hunter J; Bedzyk, Michael J

    2014-05-27

    A method for enhancing the lithium-ion capacity of a doped silicon carbide is disclosed. The method utilizes heat treating the silicon carbide in an inert atmosphere. Also disclosed are anodes for lithium-ion batteries prepared by the method.

  13. Emission characteristics of uranium hexafluoride at high temperatures

    International Nuclear Information System (INIS)

    Krascella, N.L.

    1976-01-01

    An experimental study was conducted to ascertain the spectral characteristics of uranium hexafluoride (UF 6 ) and possible UF 6 thermal decomposition products as a function of temperature and pressure. Relative emission measurements were made for UF 6 /Argon mixtures heated in a plasma torch over a range of temperatures from 800 to about 3600 0 K over a wavelength range from 80 to 600 nm. Total pressures were varied from 1 to approximately 1.7 atm. Similarly absorption measurements were carried out in the visible region from 420 to 580 nm over a temperature range from about 1000 to 1800 0 K. Total pressure for these measurements was 1.0 atm

  14. High Temperature Corrosion of Silicon Carbide and Silicon Nitride in Water Vapor

    Science.gov (United States)

    Opila, E. J.; Robinson, Raymond C.; Cuy, Michael D.; Gray, Hugh R. (Technical Monitor)

    2002-01-01

    Silicon carbide (SiC) and silicon nitride (Si3N4) are proposed for applications in high temperature combustion environments containing water vapor. Both SiC and Si3N4 react with water vapor to form a silica (SiO2) scale. It is therefore important to understand the durability of SiC, Si3N4 and SiO2 in water vapor. Thermogravimetric analyses, furnace exposures and burner rig results were obtained for these materials in water vapor at temperatures between 1100 and 1450 C and water vapor partial pressures ranging from 0.1 to 3.1 atm. First, the oxidation of SiC and Si3N4 in water vapor is considered. The parabolic kinetic rate law, rate dependence on water vapor partial pressure, and oxidation mechanism are discussed. Second, the volatilization of silica to form Si(OH)4(g) is examined. Mass spectrometric results, the linear kinetic rate law and a volatilization model based on diffusion through a gas boundary layer are discussed. Finally, the combined oxidation and volatilization reactions, which occur when SiC or Si3N4 are exposed in a water vapor-containing environment, are presented. Both experimental evidence and a model for the paralinear kinetic rate law are shown for these simultaneous oxidation and volatilization reactions.

  15. 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.

  16. 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

  17. High-temperature stability of laser-joined silicon carbide components

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, Marion, E-mail: marion.herrmann@tu-dresden.de; Lippmann, Wolfgang; Hurtado, Antonio

    2013-11-15

    Silicon carbide is recommended for applications in energy technology due to its good high-temperature corrosion resistance, mechanical durability, and abrasion resistance. The prerequisite for use is often the availability of suitable technologies for joining or sealing the components. A laser-induced process using fillers and local heating of the components represents a possible low-cost option. Investigations in which yttrium aluminosilicate glass was used for laser-induced brazing of SiC components of varying geometry are presented. A four-point bending strength of 112 MPa was found for these joints. In burst tests, laser-joined components were found to withstand internal pressures of up to 54 MPa. Helium leak tests yielded leak rates of less than 10{sup –8} mbar l s{sup −1}, even after 300 h at 900 °C. In contrast, the assemblies showed an increased leak rate after annealing at 1050 °C. The short process time of the laser technique – in the range of a few seconds to a few minutes – results in high temperature gradients and transients. SEM analysis showed that the filler in the seam predominantly solidifies in a glassy state. Crystallization occurred during later thermal loading of the joined components, with chemical equilibrium being established. Differences in seam structures yielded from different cooling rates in the laser process could not be equalized by annealing. The results demonstrated the long-term stability of laser-brazed SiC assemblies to temperatures in the range of glass transformation (900 °C) of the yttrium aluminosilicate filler. In technological investigations, the suitability of the laser joining technique for sealing of SiC components with a geometry approximating that of a fuel element sleeve pin (pin) in a gas-cooled fast reactor was proven.

  18. 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.

  19. Characterization, Modeling and Design Parameters Identification of Silicon Carbide Junction Field Effect Transistor for Temperature Sensor Applications

    Directory of Open Access Journals (Sweden)

    Sofiane Khachroumi

    2010-01-01

    Full Text Available Sensor technology is moving towards wide-band-gap semiconductors providing high temperature capable devices. Indeed, the higher thermal conductivity of silicon carbide, (three times more than silicon, permits better heat dissipation and allows better cooling and temperature management. Though many temperature sensors have already been published, little endeavours have been invested in the study of silicon carbide junction field effect devices (SiC-JFET as a temperature sensor. SiC-JFETs devices are now mature enough and it is close to be commercialized. The use of its specific properties versus temperatures is the major focus of this paper. The SiC-JFETs output current-voltage characteristics are characterized at different temperatures. The saturation current and its on-resistance versus temperature are successfully extracted. It is demonstrated that these parameters are proportional to the absolute temperature. A physics-based model is also presented. Relationships between on-resistance and saturation current versus temperature are introduced. A comparative study between experimental data and simulation results is conducted. Important to note, the proposed model and the experimental results reflect a successful agreement as far as a temperature sensor is concerned.

  20. Advanced Packaging Technology Used in Fabricating a High-Temperature Silicon Carbide Pressure Sensor

    Science.gov (United States)

    Beheim, Glenn M.

    2003-01-01

    The development of new aircraft engines requires the measurement of pressures in hot areas such as the combustor and the final stages of the compressor. The needs of the aircraft engine industry are not fully met by commercially available high-temperature pressure sensors, which are fabricated using silicon. Kulite Semiconductor Products and the NASA Glenn Research Center have been working together to develop silicon carbide (SiC) pressure sensors for use at high temperatures. At temperatures above 850 F, silicon begins to lose its nearly ideal elastic properties, so the output of a silicon pressure sensor will drift. SiC, however, maintains its nearly ideal mechanical properties to extremely high temperatures. Given a suitable sensor material, a key to the development of a practical high-temperature pressure sensor is the package. A SiC pressure sensor capable of operating at 930 F was fabricated using a newly developed package. The durability of this sensor was demonstrated in an on-engine test. The SiC pressure sensor uses a SiC diaphragm, which is fabricated using deep reactive ion etching. SiC strain gauges on the surface of the diaphragm sense the pressure difference across the diaphragm. Conventionally, the SiC chip is mounted to the package with the strain gauges outward, which exposes the sensitive metal contacts on the chip to the hostile measurement environment. In the new Kulite leadless package, the SiC chip is flipped over so that the metal contacts are protected from oxidation by a hermetic seal around the perimeter of the chip. In the leadless package, a conductive glass provides the electrical connection between the pins of the package and the chip, which eliminates the fragile gold wires used previously. The durability of the leadless SiC pressure sensor was demonstrated when two 930 F sensors were tested in the combustor of a Pratt & Whitney PW4000 series engine. Since the gas temperatures in these locations reach 1200 to 1300 F, the sensors were

  1. Formation of Porous Silicon Carbide and its Suitability as a Chemical and Temperature Detector

    National Research Council Canada - National Science Library

    Rittenhouse, Tilghman

    2004-01-01

    .... A novel electroless method of producing porous silicon carbide (PSiC) is presented. Unlike anodic methods of producing PSiC the electroless process does not require electrical contact during etching...

  2. 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

  3. AC measurements on uranium doped high temperature superconductors

    International Nuclear Information System (INIS)

    Eisterer, M.

    1999-11-01

    The subject of this thesis is the influence of fission tracks on the superconducting properties of melt textured Y-123. The critical current densities, the irreversibility lines and the transition temperature were determined by means of ac measurements. The corresponding ac techniques are explored in detail. Deviations of the ac signal from the expectations according to the Bean model were explained by the dependence of the shielding currents on the electric field. This explanation is supported by the influence of the ac amplitude and frequency on the critical current density but also by a comparison of the obtained data with other experimental techniques. Y-123 has to be doped with uranium in order to induce fission tracks. Uranium forms normal conducting clusters, which are nearly spherical, with a diameter of about 300 nm. Fission of uranium-235 by thermal neutrons creates two high energy ions with a total energy of about 160 MeV. Each of these fission products induces a linear defect with a diameter of about 10 nm. The length of one fission track is 2-4 μm. At 77 K the critical current density is enhanced by the pinning action of the uranium clusters, compared to undoped samples. With decreasing temperature this influence becomes negligible. The critical current densities are strongly enhanced due to the irradiation. At low magnetic fields we find extremely high values for melt textured materials, e.g. 2.5x10 9 Am -2 at 77 K and 0.25 T or 6x10 10 Am -2 at 5 K. Since the critical current was found to be inverse proportional to the square root of the applied magnetic field it decreases rapidly as the field increases. This behavior is predicted by simple theoretical considerations, but is only valid at low temperatures as well as in low magnetic fields at high temperatures. At high fields the critical current drops more rapidly. The irreversibility lines are only slightly changed by this irradiation technique. Only a small shift to higher fields and temperatures

  4. Coarsening-densification transition temperature in sintering of uranium dioxide

    International Nuclear Information System (INIS)

    Balakrishna, Palanki; Narasimha Murty, B.; Chakraborthy, K.P.; Jayaraj, R.N.; Ganguly, C.

    2001-01-01

    The concept of coarsening-densification transition temperature (CDTT) has been proposed to explain the experimental observations of the study of sintering undoped uranium dioxide and niobia-doped uranium dioxide powder compacts in argon atmosphere in a laboratory tubular furnace. The general method for deducing CDTT for a given material under the prevailing conditions of sintering and the likely variables that influence the CDTT are described. Though the present work is specific in nature for uranium dioxide sintering in argon atmosphere, the concept of CDTT is fairly general and must be applicable to sintering of any material and has immense potential to offer advantages in designing and/or optimizing the profile of a sintering furnace, in the diagnosis of the fault in the process conditions of sintering, and so on. The problems of viewing the effect of heating rate only in terms of densification are brought out in the light of observing the undesirable phenomena of coring and bloating and causes were identified and remedial measures suggested

  5. Comparison between Silicon-Carbide and diamond for fast neutron detection at room temperature

    Directory of Open Access Journals (Sweden)

    Obraztsova O.

    2018-01-01

    Full Text Available Neutron radiation detector for nuclear reactor applications plays an important role in getting information about the actual neutron yield and reactor environment. Such detector must be able to operate at high temperature (up to 600° C and high neutron flux levels. It is worth nothing that a detector for industrial environment applications must have fast and stable response over considerable long period of use as well as high energy resolution. Silicon Carbide is one of the most attractive materials for neutron detection. Thanks to its outstanding properties, such as high displacement threshold energy (20-35 eV, wide band gap energy (3.27 eV and high thermal conductivity (4.9 W/cm·K, SiC can operate in harsh environment (high temperature, high pressure and high radiation level without additional cooling system. Our previous analyses reveal that SiC detectors, under irradiation and at elevated temperature, respond to neutrons showing consistent counting rates as function of external reverse bias voltages and radiation intensity. The counting-rate of the thermal neutron-induced peak increases with the area of the detector, and appears to be linear with respect to the reactor power. Diamond is another semi-conductor considered as one of most promising materials for radiation detection. Diamond possesses several advantages in comparison to other semiconductors such as a wider band gap (5.5 eV, higher threshold displacement energy (40-50 eV and thermal conductivity (22 W/cm·K, which leads to low leakage current values and make it more radiation resistant that its competitors. A comparison is proposed between these two semiconductors for the ability and efficiency to detect fast neutrons. For this purpose the deuterium-tritium neutron generator of Technical University of Dresden with 14 MeV neutron output of 1010 n·s-1 is used. In the present work, we interpret the first measurements and results with both 4H-SiC and chemical vapor deposition (CVD

  6. Comparison between Silicon-Carbide and diamond for fast neutron detection at room temperature

    Science.gov (United States)

    Obraztsova, O.; Ottaviani, L.; Klix, A.; Döring, T.; Palais, O.; Lyoussi, A.

    2018-01-01

    Neutron radiation detector for nuclear reactor applications plays an important role in getting information about the actual neutron yield and reactor environment. Such detector must be able to operate at high temperature (up to 600° C) and high neutron flux levels. It is worth nothing that a detector for industrial environment applications must have fast and stable response over considerable long period of use as well as high energy resolution. Silicon Carbide is one of the most attractive materials for neutron detection. Thanks to its outstanding properties, such as high displacement threshold energy (20-35 eV), wide band gap energy (3.27 eV) and high thermal conductivity (4.9 W/cm·K), SiC can operate in harsh environment (high temperature, high pressure and high radiation level) without additional cooling system. Our previous analyses reveal that SiC detectors, under irradiation and at elevated temperature, respond to neutrons showing consistent counting rates as function of external reverse bias voltages and radiation intensity. The counting-rate of the thermal neutron-induced peak increases with the area of the detector, and appears to be linear with respect to the reactor power. Diamond is another semi-conductor considered as one of most promising materials for radiation detection. Diamond possesses several advantages in comparison to other semiconductors such as a wider band gap (5.5 eV), higher threshold displacement energy (40-50 eV) and thermal conductivity (22 W/cm·K), which leads to low leakage current values and make it more radiation resistant that its competitors. A comparison is proposed between these two semiconductors for the ability and efficiency to detect fast neutrons. For this purpose the deuterium-tritium neutron generator of Technical University of Dresden with 14 MeV neutron output of 1010 n·s-1 is used. In the present work, we interpret the first measurements and results with both 4H-SiC and chemical vapor deposition (CVD) diamond

  7. 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.

  8. Melting temperature of uranium - plutonium mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

  9. Melting temperature of uranium - plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Hirosawa, Takashi

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

  10. Anomalous thermal property behaviour of uranium at low temperatures

    International Nuclear Information System (INIS)

    Sandenaw, T.A.

    1975-01-01

    Low temperature heat capacity curves are presented for polycrystalline 235 U and 238 U metals in different microstructural states and of different purities. Thermal conductivity versus temperature curves are shown for low-purity, polycrystalline 238 U in the temperature range between approximately 80 and 373 0 K for metal having undergone varied fabrication procedures. Published information suggests that there will be no structural modification in very pure uranium below room temperature. The influence of impurities on low temperature transitions may be through their effects on dislocation formation. Thermal conductivity and heat capacity runs started at approximately 80 0 K, after holding specimens at the temperature of boiling liquid nitrogen, do not give results which match up with runs started below 36 to 43 0 K. Result of measurements started at approximately 80 0 K indicate that an ordering mechanism is predominating, with microstructure rather than purity being the important factor. This can be explained if ordering at approximately 80 0 K is through lattice imperfections remaining from prior specimen processing. The drop off in heat capacity appearing above 36 0 K in the C/sub p/ versus T curves of 235 U and 238 U suggest the possibility of: (1) heat evolution from a developing antiphase structure or (2) heat evolution similar to that noted with a quenched martensite. Physical property changes in 238 U at 250 to 270 0 K and at 325 to 350 0 K seem to be related to the heat evolution which starts at 36 0 K during adiabatic heat capacity measurements. The data from heat capacity and thermal conductivity measurements are analyzed to help explain the significance of the sometimes very slight physical property changes observed at 36 to 43, approximately 80, 250 to 270 and 325 to 350 0 K in uranium metal. (U.S.)

  11. Implementation Challenges for Sintered Silicon Carbide Fiber Bonded Ceramic Materials for High Temperature Applications

    Science.gov (United States)

    Singh, M.

    2011-01-01

    During the last decades, a number of fiber reinforced ceramic composites have been developed and tested for various aerospace and ground based applications. However, a number of challenges still remain slowing the wide scale implementation of these materials. In addition to continuous fiber reinforced composites, other innovative materials have been developed including the fibrous monoliths and sintered fiber bonded ceramics. The sintered silicon carbide fiber bonded ceramics have been fabricated by the hot pressing and sintering of silicon carbide fibers. However, in this system reliable property database as well as various issues related to thermomechanical performance, integration, and fabrication of large and complex shape components has yet to be addressed. In this presentation, thermomechanical properties of sintered silicon carbide fiber bonded ceramics (as fabricated and joined) will be presented. In addition, critical need for manufacturing and integration technologies in successful implementation of these materials will be discussed.

  12. Nanoporous, Metal Carbide, Surface Diffusion Membranes for High Temperature Hydrogen Separations

    Energy Technology Data Exchange (ETDEWEB)

    Way, J. Douglas [Colorado School of Mines, Golden, CO (United States). Dept. of Chemical and Biological Engineering; Wolden, Colin A. [Colorado School of Mines, Golden, CO (United States)

    2013-09-30

    Colorado School of Mines (CSM) developed high temperature, hydrogen permeable membranes that contain no platinum group metals with the goal of separating hydrogen from gas mixtures representative of gasification of carbon feedstocks such as coal or biomass in order to meet DOE NETL 2015 hydrogen membrane performance targets. We employed a dual synthesis strategy centered on transition metal carbides. In the first approach, novel, high temperature, surface diffusion membranes based on nanoporous Mo2C were fabricated on ceramic supports. These were produced in a two step process that consisted of molybdenum oxide deposition followed by thermal carburization. Our best Mo2C surface diffusion membrane achieved a pure hydrogen flux of 367 SCFH/ft2 at a feed pressure of only 20 psig. The highest H2/N2 selectivity obtained with this approach was 4.9. A transport model using “dusty gas” theory was derived to describe the hydrogen transport in the Mo2C coated, surface diffusion membranes. The second class of membranes developed were dense metal foils of BCC metals such as vanadium coated with thin (< 60 nm) Mo2C catalyst layers. We have fabricated a Mo2C/V composite membrane that in pure gas testing delivered a H2 flux of 238 SCFH/ft2 at 600 °C and 100 psig, with no detectable He permeance. This exceeds the 2010 DOE Target flux. This flux is 2.8 times that of pure Pd at the same membrane thickness and test conditions and over 79% of the 2015 flux target. In mixed gas testing we achieved a permeate purity of ≥99.99%, satisfying the permeate purity milestone, but the hydrogen permeance was low, ~0.2 SCFH/ft2.psi. However, during testing of a Mo2C coated Pd alloy membrane with DOE 1 feed gas mixture a hydrogen permeance of >2 SCFH/ft2.psi was obtained which was stable during the entire test, meeting the permeance associated with

  13. 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

  14. 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.

  15. Time-Dependent Stress Rupture Strength Degradation of Hi-Nicalon Fiber-Reinforced Silicon Carbide Composites at Intermediate Temperatures

    Science.gov (United States)

    Sullivan, Roy M.

    2016-01-01

    The stress rupture strength of silicon carbide fiber-reinforced silicon carbide composites with a boron nitride fiber coating decreases with time within the intermediate temperature range of 700 to 950 degree Celsius. Various theories have been proposed to explain the cause of the time-dependent stress rupture strength. The objective of this paper is to investigate the relative significance of the various theories for the time-dependent strength of silicon carbide fiber-reinforced silicon carbide composites. This is achieved through the development of a numerically based progressive failure analysis routine and through the application of the routine to simulate the composite stress rupture tests. The progressive failure routine is a time-marching routine with an iterative loop between a probability of fiber survival equation and a force equilibrium equation within each time step. Failure of the composite is assumed to initiate near a matrix crack and the progression of fiber failures occurs by global load sharing. The probability of survival equation is derived from consideration of the strength of ceramic fibers with randomly occurring and slow growing flaws as well as the mechanical interaction between the fibers and matrix near a matrix crack. The force equilibrium equation follows from the global load sharing presumption. The results of progressive failure analyses of the composite tests suggest that the relationship between time and stress-rupture strength is attributed almost entirely to the slow flaw growth within the fibers. Although other mechanisms may be present, they appear to have only a minor influence on the observed time-dependent behavior.

  16. A comparison between thorium-uranium and low enrichment uranium cycles in the high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cerles, J M

    1973-03-15

    In a previous report, it was shown that the Uranium cycle could be used as well with multi-hole block (GGA type) as with tubular elements. Now, in a F.S.V. geometry, a comparison is made between Thorium cycle and Uranium cycle. This comparison will be concerned with the physical properties of the materials, the needs of natural Uranium, the fissile material inventory and, at last, an attempt of economical considerations. In this report the cycle will be characterizd by the fertile material. So, we write ''Thorium cycle'' for Highly Enriched Uranium - Thorium cycle and ''Uranium cycle'' for low Enrichment Uranium cycle.

  17. 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.

  18. Detection of uranium extraction zone by axial temperature profiles in a pulsed column for Purex process

    International Nuclear Information System (INIS)

    Tsukada, T.; Takahashi, K.

    1991-01-01

    A new method was presented for detecting uranium extraction zone in a pulsed column by means of measuring axial temperature profile originated from reaction heat during uranium extraction. Key parameters of the temperature profiles were estimated with a code developed for calculating temperature profiles in a direct-contact heat exchanger such as a pulsed column, and were verified using data from a small pulsed column simulating reaction heat with injecting hot water. Finally, the results were compared with those from an actual uranium extraction tests, indicating that the method presented was promising for detecting uranium extraction zone in a pulsed column. (author)

  19. Temperature Dependence of Uranium and Vanadium Adsorption on Amidoxime-Based Adsorbents in Natural Seawater

    Energy Technology Data Exchange (ETDEWEB)

    Kuo, Li-Jung [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; Gill, Gary A. [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; Tsouris, Costas [Oak Ridge National Laboratory, Oak Ridge TN 37831 USA; Rao, Linfeng [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley CA 94720 USA; Pan, Horng-Bin [Department of Chemistry, University of Idaho, Moscow ID 83844 USA; Wai, Chien M. [Department of Chemistry, University of Idaho, Moscow ID 83844 USA; Janke, Christopher J. [Oak Ridge National Laboratory, Oak Ridge TN 37831 USA; Strivens, Jonathan E. [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; Wood, Jordana R. [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; Schlafer, Nicholas [Marine Sciences Laboratory, Pacific Northwest National Laboratory, Sequim WA 98382 USA; D' Alessandro, Evan K. [Rosensteil School of Marine and Atmospheric Chemistry, University of Miami, Miami FL 33149 USA

    2018-01-16

    The apparent enthalpy and entropy of the complexation of uranium (VI) and vanadium (V) with amidoxime ligands grafted onto polyethylene fiber was determined using time series measurements of adsorption capacities in natural seawater at three different temperatures. The complexation of uranium was highly endothermic, while the complexation of vanadium showed minimal temperature sensitivity. Amidoxime-based polymeric adsorbents exhibit significantly increased uranium adsorption capacities and selectivity in warmer waters.

  20. 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)

  1. 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

  2. High temperature corrosion of silicon carbide and silicon nitride in the presence of chloride compound

    International Nuclear Information System (INIS)

    McNallan, M.

    1993-01-01

    Silicon carbide and silicon nitride are resistant to oxidation because a protective silicon dioxide films on their surfaces in most oxidizing environments. Chloride compounds can attack the surface in two ways: 1) chlorine can attack the silicon directly to form a volatile silicon chloride compound or 2) alkali compounds combined with the chlorine can be transported to the surface where they flux the silica layer by forming stable alkali silicates. Alkali halides have enough vapor pressure that a sufficient quantity of alkali species to cause accelerated corrosion can be transported to the ceramic surface without the formation of a chloride deposit. When silicon carbide is attacked simultaneously by chlorine and oxygen, the corrosion products include both volatile and condensed spices. Silicon nitride is much more resistance to this type of attack than silicon carbide. Silicon based ceramics are exposed to oxidizing gases in the presence of alkali chloride vapors, the rate of corrosion is controlled primarily by the driving force for the formation of alkali silicate, which can be quantified as the activity of the alkali oxide in equilibrium with the corrosive gas mixture. In a gas mixture containing a fixed partial pressure of KCl, the rate of corrosion is accelerated by increasing the concentration of water vapor and inhibited by increasing the concentration of HCl. Similar results have been obtained for mixtures containing other alkalis and halogens. (Orig./A.B.)

  3. In situ spectroscopy and spectroelectrochemistry of uranium in high-temperature alkali chloride molten salts.

    Science.gov (United States)

    Polovov, Ilya B; Volkovich, Vladimir A; Charnock, John M; Kralj, Brett; Lewin, Robert G; Kinoshita, Hajime; May, Iain; Sharrad, Clint A

    2008-09-01

    Soluble uranium chloride species, in the oxidation states of III+, IV+, V+, and VI+, have been chemically generated in high-temperature alkali chloride melts. These reactions were monitored by in situ electronic absorption spectroscopy. In situ X-ray absorption spectroscopy of uranium(VI) in a molten LiCl-KCl eutectic was used to determine the immediate coordination environment about the uranium. The dominant species in the melt was [UO 2Cl 4] (2-). Further analysis of the extended X-ray absorption fine structure data and Raman spectroscopy of the melts quenched back to room temperature indicated the possibility of ordering beyond the first coordination sphere of [UO 2Cl 4] (2-). The electrolytic generation of uranium(III) in a molten LiCl-KCl eutectic was also investigated. Anodic dissolution of uranium metal was found to be more efficient at producing uranium(III) in high-temperature melts than the cathodic reduction of uranium(IV). These high-temperature electrolytic processes were studied by in situ electronic absorption spectroelectrochemistry, and we have also developed in situ X-ray absorption spectroelectrochemistry techniques to probe both the uranium oxidation state and the uranium coordination environment in these melts.

  4. Effect of sintering temperature and boron carbide content on the wear behavior of hot pressed diamond cutting segments

    Directory of Open Access Journals (Sweden)

    Islak S.

    2015-01-01

    Full Text Available The aim of this study was to investigate the effect of sintering temperature and boron carbide content on wear behavior of diamond cutting segments. For this purpose, the segments contained 2, 5 and 10 wt.% B4C were prepared by hot pressing process carried out under a pressure of 35 MPa, at 600, 650 and 700 °C for 3 minutes. The transverse rupture strength (TRS of the segments was assessed using a three-point bending test. Ankara andesite stone was cut to examine the wear behavior of segments with boron carbide. Microstructure, surfaces of wear and fracture of segments were determined by scanning electron microscopy (SEM-EDS, and X-ray diffraction (XRD analysis. As a result, the wear rate decreased significantly in the 0-5 wt.% B4C contents, while it increased in the 5-10 wt.% B4C contents. With increase in sintering temperature, the wear rate decreased due to the hard matrix.

  5. 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

  6. 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

  7. Medium temperature reaction between lanthanide and actinide carbides and hydrogen; Reaction a temperature moyenne entre les monocarbures de lanthanides et d'actinides et l'hydrogene

    Energy Technology Data Exchange (ETDEWEB)

    Dean, G; Lorenzelli, R; Pascard, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    Hydrogen is fixed reversibly by the lanthanide and actinide mono carbides in the range 25 - 400 C, as for pure corresponding metals. Hydrogen goes into the carbides lattice through carbon vacancies and the total fixed amount is approximately equal to two hydrogen atoms per initial vacancy. Final products c.n thus be considered as carbo-hydrides of general formula M(C{sub 1-x}, H{sub 2x}). The primitive CFC, NaCl type, structure remains unchanged but expands strongly in the case of actinide carbides. With lanthanide carbides, hydrogenation induces a phase transformation with reappearance of the metal structure (HCP). Hydrogen decomposition pressures of all the studied carbo-hydrides are greater than those of the corresponding di-hydrides. (authors) [French] Les monocarbures d'actinides et de lanthanides fixent reversiblement de l'hydrogene a temperature peu elevee, a peu pres dans les memes conditions que les metaux purs correspondants. L'hydrogene penetre dans le reseau des carbures par l'intermediaire des lacunes de carbone, et la quantite totale fixee est approximativement egale a deux atomes d'hydrogene par lacune initiale. Les produits obtenus peuvent donc etre consideres comme des carbohydrures de formule generale M(C{sub 1-x}, H{sub 2x}). La structure d'origine CFC, type NaCl est conservee, mais avec une forte expansion, dans le cas des carbures d'actinides. En revanche, l'hydrogenation entraine un changement de phase cristalline avec retour a la structure du metal (HC) pour les carbures de lanthanides. Tous les carbohydrures etudies ont des tensions de decomposition en hydrogene superieures a celles des dihydrures correspondants. (auteurs)

  8. 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

  9. 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

  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. Effect of Carbide Dissolution on Chlorine Induced High Temperature Corrosion of HVOF and HVAF Sprayed Cr3C2-NiCrMoNb Coatings

    Science.gov (United States)

    Fantozzi, D.; Matikainen, V.; Uusitalo, M.; Koivuluoto, H.; Vuoristo, P.

    2018-01-01

    Highly corrosion- and wear-resistant thermally sprayed chromium carbide (Cr3C2)-based cermet coatings are nowadays a potential highly durable solution to allow traditional fluidized bed combustors (FBC) to be operated with ecological waste and biomass fuels. However, the heat input of thermal spray causes carbide dissolution in the metal binder. This results in the formation of carbon saturated metastable phases, which can affect the behavior of the materials during exposure. This study analyses the effect of carbide dissolution in the metal matrix of Cr3C2-50NiCrMoNb coatings and its effect on chlorine-induced high-temperature corrosion. Four coatings were thermally sprayed with HVAF and HVOF techniques in order to obtain microstructures with increasing amount of carbide dissolution in the metal matrix. The coatings were heat-treated in an inert argon atmosphere to induce secondary carbide precipitation. As-sprayed and heat-treated self-standing coatings were covered with KCl, and their corrosion resistance was investigated with thermogravimetric analysis (TGA) and ordinary high-temperature corrosion test at 550 °C for 4 and 72 h, respectively. High carbon dissolution in the metal matrix appeared to be detrimental against chlorine-induced high-temperature corrosion. The microstructural changes induced by the heat treatment hindered the corrosion onset in the coatings.

  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. High temperature synthesis of ceramic composition by directed reaction of molten titanium or zirconium with boron carbide

    International Nuclear Information System (INIS)

    Johnson, W.B.

    1990-01-01

    Alternative methods of producing ceramics and ceramic composites include sintering, hot pressing and more recently hot isostatic pressing (HIP) and self-propagating high temperature synthesis (SHS). Though each of these techniques has its advantages, each suffers from several restrictions as well. Sintering may require long times at high temperatures and for most materials requires sintering aids to get full density. These additives can, and generally do, change (often degrade) the properties of the ceramic. Hot pressing and hot isostatic pressing are convenient methods to quickly prepare samples of some materials to full density, but generally are expensive and may damage some types of reinforcements during densification. This paper focuses on the preparation and processing of composites prepared by the directed reaction of molten titanium or zirconium with boron carbide. Advantages and disadvantages of this approach when compared to traditional methods are discussed, with reference to specific examples. Examples of microstructure are properties of these materials are reported

  16. Performance and metallography of a uranium tritide bed operated at elevated temperatures and tritium pressures

    International Nuclear Information System (INIS)

    Mote, M.W. Jr.; Mintz, J.M.

    1986-12-01

    A uranium gettering bed was cycled between room temperature/zero pressure and 600C/275 psi (D 2 ) for 210 cycles over a period of 8 months. Metallographic examination of the hardware revealed an acceptable amount of reaction between the uranium and the stainless steel container. This exposure is estimated to represent about ten years of normal use

  17. Review of experimental studies of zirconium carbide coated fuel particles for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Minato, Kazuo; Ogawa, Toru; Fukuda, Kousaku

    1995-03-01

    Experimental studies of zirconium carbide(ZrC) coated fuel particles were reviewed from the viewpoints of fuel particle designs, fabrication, characterization, fuel performance, and fission product retentiveness. ZrC is known as a refractory and chemically stable compound, so ZrC is a candidate to replace the silicon carbide(SiC) coating layer of the Triso-coated fuel particles. The irradiation experiments, the postirradiation heating tests, and the out-of-reactor experiments showed that the ZrC layer was less susceptible than the SiC layer to chemical attack by fission products and fuel kernels, and that the ZrC-coated fuel particles performed better than the standard Triso-coated fuel particles at high temperatures, especially above 1600degC. The ZrC-coated fuel particles demonstrated better cesium retention than the standard Triso-coated fuel particles though the ZrC layer showed a less effective barrier to ruthenium than the SiC layer. (author) 51 refs

  18. The Effect of High Temperature Annealing on the Grain Characteristics of a Thin Chemical Vapor Deposition Silicon Carbide Layer.

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen; Philippus M van Rooyen; Mary Lou Dunzik-Gougar

    2013-08-01

    The unique combination of thermo-mechanical and physiochemical properties of silicon carbide (SiC) provides interest and opportunity for its use in nuclear applications. One of the applications of SiC is as a very thin layer in the TRi-ISOtropic (TRISO) coated fuel particles for high temperature gas reactors (HTGRs). This SiC layer, produced by chemical vapor deposition (CVD), is designed to withstand the pressures of fission and transmutation product gases in a high temperature, radiation environment. Various researchers have demonstrated that macroscopic properties can be affected by changes in the distribution of grain boundary plane orientations and misorientations [1 - 3]. Additionally, various researchers have attributed the release behavior of Ag through the SiC layer as a grain boundary diffusion phenomenon [4 - 6]; further highlighting the importance of understanding the actual grain characteristics of the SiC layer. Both historic HTGR fission product release studies and recent experiments at Idaho National Laboratory (INL) [7] have shown that the release of Ag-110m is strongly temperature dependent. Although the maximum normal operating fuel temperature of a HTGR design is in the range of 1000-1250°C, the temperature may reach 1600°C under postulated accident conditions. The aim of this specific study is therefore to determine the magnitude of temperature dependence on SiC grain characteristics, expanding upon initial studies by Van Rooyen et al, [8; 9].

  19. 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

  20. 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.

  1. Structural and optical properties of silicon-carbide nanowires produced by the high-temperature carbonization of silicon nanostructures

    Energy Technology Data Exchange (ETDEWEB)

    Pavlikov, A. V., E-mail: pavlikov@physics.msu.ru [Moscow State University, Faculty of Physics (Russian Federation); Latukhina, N. V.; Chepurnov, V. I. [Samara National Researh University (Russian Federation); Timoshenko, V. Yu. [Moscow State University, Faculty of Physics (Russian Federation)

    2017-03-15

    Silicon-carbide (SiC) nanowire structures 40–50 nm in diameter are produced by the high-temperature carbonization of porous silicon and silicon nanowires. The SiC nanowires are studied by scanning electron microscopy, X-ray diffraction analysis, Raman spectroscopy, and infrared reflectance spectroscopy. The X-ray structural and Raman data suggest that the cubic 3C-SiC polytype is dominant in the samples under study. The shape of the infrared reflectance spectrum in the region of the reststrahlen band 800–900 cm{sup –1} is indicative of the presence of free charge carriers. The possibility of using SiC nanowires in microelectronic, photonic, and gas-sensing devices is discussed.

  2. Non-adiabatic ab initio molecular dynamics of supersonic beam epitaxy of silicon carbide at room temperature

    Energy Technology Data Exchange (ETDEWEB)

    Taioli, Simone [Interdisciplinary Laboratory for Computational Science, FBK-Center for Materials and Microsystems and University of Trento, Trento (Italy); Department of Physics, University of Trento, Trento (Italy); Istituto Nazionale di Fisica Nucleare, Sezione di Perugia (Italy); Department of Chemistry, University of Bologna, Bologna (Italy); Garberoglio, Giovanni [Interdisciplinary Laboratory for Computational Science, FBK-Center for Materials and Microsystems and University of Trento, Trento (Italy); Simonucci, Stefano [Interdisciplinary Laboratory for Computational Science, FBK-Center for Materials and Microsystems and University of Trento, Trento (Italy); Istituto Nazionale di Fisica Nucleare, Sezione di Perugia (Italy); Department of Physics, University of Camerino, Camerino (Italy); Beccara, Silvio a [Interdisciplinary Laboratory for Computational Science, FBK-Center for Materials and Microsystems and University of Trento, Trento (Italy); Department of Physics, University of Trento, Trento (Italy); Aversa, Lucrezia [Institute of Materials for Electronics and Magnetism, IMEM-CNR, Trento (Italy); Nardi, Marco [Institute of Materials for Electronics and Magnetism, IMEM-CNR, Trento (Italy); Institut fuer Physik, Humboldt-Universitaet zu Berlin, Berlin (Germany); Verucchi, Roberto [Institute of Materials for Electronics and Magnetism, FBK-CNR, Trento (Italy); Iannotta, Salvatore [Institute of Materials for Electronics and Magnetism, IMEM-CNR, Parma (Italy); Dapor, Maurizio [Interdisciplinary Laboratory for Computational Science, FBK-Center for Materials and Microsystems and University of Trento, Trento (Italy); Department of Materials Engineering and Industrial Technologies, University of Trento, Trento (Italy); Istituto Nazionale di Fisica Nucleare, Sezione di Padova (Italy); and others

    2013-01-28

    In this work, we investigate the processes leading to the room-temperature growth of silicon carbide thin films by supersonic molecular beam epitaxy technique. We present experimental data showing that the collision of fullerene on a silicon surface induces strong chemical-physical perturbations and, for sufficient velocity, disruption of molecular bonds, and cage breaking with formation of nanostructures with different stoichiometric character. We show that in these out-of-equilibrium conditions, it is necessary to go beyond the standard implementations of density functional theory, as ab initio methods based on the Born-Oppenheimer approximation fail to capture the excited-state dynamics. In particular, we analyse the Si-C{sub 60} collision within the non-adiabatic nuclear dynamics framework, where stochastic hops occur between adiabatic surfaces calculated with time-dependent density functional theory. This theoretical description of the C{sub 60} impact on the Si surface is in good agreement with our experimental findings.

  3. Spectroscopic measurements of plasma temperatures and electron number density in a uranium hollow cathode discharge lamp

    International Nuclear Information System (INIS)

    Shah, M.L.; Suri, B.M.; Gupta, G.P.

    2015-01-01

    The HCD (Hollow Cathode Discharge) lamps have been used as a source of free atoms of any metal, controllable by direct current in the lamp. The plasma parameters including neutral species temperature, atomic excitation temperature and electron number density in a see-through type, homemade uranium hollow cathode discharge lamp with neon as a buffer gas have been investigated using optical emission spectroscopic techniques. The neutral species temperature has been measured using the Doppler broadening of a neon atomic spectral line. The atomic excitation temperature has been measured using the Boltzmann plot method utilizing uranium atomic spectral lines. The electron number density has been determined from the Saha-Boltzmann equation utilizing uranium atomic and ionic spectral lines. To the best of our knowledge, all these three plasma parameters are simultaneously measured for the first time in a uranium hollow cathode discharge lamp

  4. Use of thermogravimetry and thermodynamic calculations for specifying chromium diffusion occurring in alloys containing chromium carbides during high temperature oxidation

    International Nuclear Information System (INIS)

    Berthod, Patrice; Conrath, Elodie

    2015-01-01

    The chromium diffusion is of great importance for the high temperature oxidation behaviour of the chromium-rich carbides-strengthened superalloys. These ones contain high chromium quantities for allowing them well resisting hot corrosion by constituting and maintaining a continuous external scale of chromia. Knowing how chromium can diffuse in such alloys is thus very useful for predicting the sustainability of their chromia-forming behaviour. Since Cr diffusion occurs through the external part of the alloy already affected by the previous steps of oxidation (decarburized subsurface) it is more judicious to specify this diffusion during the oxidation process itself. This was successfully carried out in this work in the case of a model chromia-forming nickel-based alloy containing chromium carbides, Ni(bal.)–25Cr–0.5C (in wt.%). This was done by specifying, using real-time thermogravimetry, the mass gain kinetic due to oxidation, and by combining it with the post-mortem determination of the Cr concentration profiles in subsurface. The values of D Cr thus obtained for 1000, 1050 and 1100 °C in the alloy subsurface are consistent with the values obtained in earlier works for similar alloy's chemical compositions. - Highlights: • A Ni25Cr0.50C alloy was oxidized at high temperature in a thermo-balance. • The mass gain files were analysed to specify the Cr 2 O 3 volatilization constant K v . • Concentration profiles were acquired to specify the chromium gradient. • The diffusion coefficient of chromium through the subsurface was deduced. • The obtained diffusion coefficient is consistent with values previously obtained.

  5. 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

  6. An unusual temperature dependence in the oxidation of oxycarbide layers on uranium

    Science.gov (United States)

    Ellis, Walton P.

    1981-09-01

    An anomalous temperature dependence has been observed for the oxidation kinetics of outermost oxycarbide layers on polycrystalline uranium metal. Normally, oxidation or corrosion reactions are expected to proceed more rapidly as the temperature is elevated. Thus, it came as a surprise when we observed that the removal of the outermost atomic layers of carbon from uranium oxycarbide by O 2 reproducibly proceeds at a much faster rate at 25°C than at 280°C.

  7. 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)

  8. 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)

  9. Recent advance in high manufacturing readiness level and high temperature CMOS mixed-signal integrated circuits on silicon carbide

    Science.gov (United States)

    Weng, M. H.; Clark, D. T.; Wright, S. N.; Gordon, D. L.; Duncan, M. A.; Kirkham, S. J.; Idris, M. I.; Chan, H. K.; Young, R. A. R.; Ramsay, E. P.; Wright, N. G.; Horsfall, A. B.

    2017-05-01

    A high manufacturing readiness level silicon carbide (SiC) CMOS technology is presented. The unique process flow enables the monolithic integration of pMOS and nMOS transistors with passive circuit elements capable of operation at temperatures of 300 °C and beyond. Critical to this functionality is the behaviour of the gate dielectric and data for high temperature capacitance-voltage measurements are reported for SiO2/4H-SiC (n and p type) MOS structures. In addition, a summary of the long term reliability for a range of structures including contact chains to both n-type and p-type SiC, as well as simple logic circuits is presented, showing function after 2000 h at 300 °C. Circuit data is also presented for the performance of digital logic devices, a 4 to 1 analogue multiplexer and a configurable timer operating over a wide temperature range. A high temperature micro-oven system has been utilised to enable the high temperature testing and stressing of units assembled in ceramic dual in line packages, including a high temperature small form-factor SiC based bridge leg power module prototype, operated for over 1000 h at 300 °C. The data presented show that SiC CMOS is a key enabling technology in high temperature integrated circuit design. In particular it provides the ability to realise sensor interface circuits capable of operating above 300 °C, accommodate shifts in key parameters enabling deployment in applications including automotive, aerospace and deep well drilling.

  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 of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-09-15

    By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the {gamma} {yields} {beta} transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the {beta} {yields} {alpha} transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

  12. Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-09-15

    By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the {gamma} {yields} {beta} transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the {beta} {yields} {alpha} transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

  13. 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.

  14. The uranium dioxide-uranium system at high temperature; Le systeme uranium-dioxyde d'uranium a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Guinet, Ph.; Vaugoyeau, H.; Blum, P. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1966-07-01

    The liquidus curve has been determined by a saturation method in which the thermal gradient was cancelled upon cooling, and the solidus curve by analyzing the deposits in equilibrium with the liquid at each temperature. The diagram, of a displaced eutectic type, presents a liquid immiscibility domain between 47 and 59 mol per cent of dioxide and a substoichiometry range UO{sub 2x}, the minimum O/U ratio being 1,6 at 3470 {+-} 30 C. The monotectic composition was found by chemical analysis to be 59 mol per cent of dioxide and the reaction temperature 2470 {+-} 30 C. (author) [French] La courbe liquidus a ete determinee par une methode de saturation en annulant le gradient thermique au cours du refroidissement, la courbe solidus par analyse des depots en equilibre avec le liquide a chaque temperature. Le diagramme du type a eutectique deporte comporte un domaine d'immiscibilite liquide entre 47 et 59 moles pour cent de dioxyde, ainsi qu'un domaine d'existence du compose sous stoechiometrique UO{sub 2x}, le rapport O/U minimum etant egal a 1,6 a 2470 {+-} 30 C. La composition du monotectique, obtenue par analyse chimique, est de 59 moles pour cent de dioxyde et la temperature de la reaction a ete trouvee egale a 2470 {+-} 30 C. (auteur)

  15. Influence of the temperature in the uranium (Vi) sorption in zirconium diphosphate

    International Nuclear Information System (INIS)

    Garcia G, N.; Solis, D.; Ordonez R, E.

    2012-10-01

    In the present work was evaluated the uranium (Vi) sorption at 10, 20, 30, 40 and 60 C on the zirconium diphosphate (ZrP 2 O 7 ). They were carried out kinetic and isotherms using the method by lots, these will allow to fix the sorption time (kinetic) and to explain the behavior of this sorption in different ph conditions and temperature (isotherm). The quantity of retained uranium in the surface was quantified by means of the fluorescence technique. (Author)

  16. Modeling thermal spike driven reactions at low temperature and application to zirconium carbide radiation damage

    Science.gov (United States)

    Ulmer, Christopher J.; Motta, Arthur T.

    2017-11-01

    The development of TEM-visible damage in materials under irradiation at cryogenic temperatures cannot be explained using classical rate theory modeling with thermally activated reactions since at low temperatures thermal reaction rates are too low. Although point defect mobility approaches zero at low temperature, the thermal spikes induced by displacement cascades enable some atom mobility as it cools. In this work a model is developed to calculate "athermal" reaction rates from the atomic mobility within the irradiation-induced thermal spikes, including both displacement cascades and electronic stopping. The athermal reaction rates are added to a simple rate theory cluster dynamics model to allow for the simulation of microstructure evolution during irradiation at cryogenic temperatures. The rate theory model is applied to in-situ irradiation of ZrC and compares well at cryogenic temperatures. The results show that the addition of the thermal spike model makes it possible to rationalize microstructure evolution in the low temperature regime.

  17. 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

  18. Measurement of surface temperature profiles on liquid uranium metal during electron beam evaporation

    Energy Technology Data Exchange (ETDEWEB)

    Ohba, Hironori; Shibata, Takemasa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-11-01

    Surface temperature distributions of liquid uranium in a water-cooled copper crucible during electron beam evaporation were measured. Evaporation surface was imaged by a lens through a band-path filter (650{+-}5 nm) and a double mirror system on a charge coupled device (CCD) camera. The video signals of the recorded image were connected to an image processor and converted to two-dimensional spectral radiance profiles. The surface temperatures were obtained from the spectral radiation intensity ratio of the evaporation surface and a freezing point of uranium and/or a reference light source using Planck`s law of radiation. The maximum temperature exceeded 3000 K and had saturation tendency with increasing electron beam input. The measured surface temperatures agreed with those estimated from deposition rates and data of saturated vapor pressure of uranium. (author)

  19. Thermodynamic approach to the synthesis of silicon carbide using tetramethylsilane as the precursor at high temperature

    Science.gov (United States)

    Jeong, Seong-Min; Kim, Kyung-Hun; Yoon, Young Joon; Lee, Myung-Hyun; Seo, Won-Seon

    2012-10-01

    Tetramethylsilane (TMS) is commonly used as a precursor in the production of SiC(β) films at relatively low temperatures. However, because TMS contains much more C than Si, it is difficult to produce solid phase SiC at high temperatures. In an attempt to develop a more efficient TMS-based SiC(α) process, computational thermodynamic simulations were performed under various temperatures, working pressures and TMS/H2 ratios. The findings indicate that each solid phase has a different dependency on the H2 concentration. Consequently, a high H2 concentration results in the formation of a single, solid phase SiC region at high temperatures. Finally, TMS appears to be useful as a precursor for the high temperature production of SiC(α).

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

    Energy Technology Data Exchange (ETDEWEB)

    Pai, Rajesh V., E-mail: pairajesh007@gmail.com [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai (India); Mollick, P.K. [Powder Metallurgy Division, Bhabha Atomic Research Centre, Mumbai (India); Kumar, Ashok [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai (India); Banerjee, J. [Radiometullurgy Division, Bhabha Atomic Research Centre, Mumbai (India); Radhakrishna, J. [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai (India); Chakravartty, J.K. [Powder Metallurgy Division, Bhabha Atomic Research Centre, Mumbai (India)

    2016-05-15

    UO{sub 2} microspheres prepared by internal gelation technique were coated with pyrolytic carbon and silicon carbide using CVD technique. The particles which were not meeting the specifications were rejected. The rejected/failed UO{sub 2} based coated particles prepared by CVD technique was used for oxidation and recovery and recycling. The oxidation behaviour of sintered UO{sub 2} microspheres coated with different layers of carbon and SiC was studied by thermal techniques to develop a method for recycling and recovery of uranium from the failed/rejected coated particles. It was observed that the complete removal of outer carbon from the spheres is difficult. The crushing of microspheres enabled easier accessibility of oxygen and oxidation of carbon and uranium at 800–1000 °C. With the optimized process of multiple crushing using die & plunger and sieving the broken coated layers, we could recycle around fifty percent of the UO{sub 2} microspheres which could be directly recoated. The rest of the particles were recycled using a wet recycling method. - Highlights: • The oxidation behaviour of coated particles was studied in air, O{sub 2} and moist O{sub 2}. • It was observed that coated layers cannot be completely removed by mere oxidation. • Complete recovery of uranium from the rejected coated particles has been carried out using a combination of dry and wet recovery scheme. • A crushing step prior to oxidation is needed for full recovery of uranium from the coated particles.

  1. Low threading dislocation density aluminum nitride on silicon carbide through the use of reduced temperature interlayers

    KAUST Repository

    Foronda, Humberto M.; Wu, Feng; Zollner, Christian; Alif, Muhammad Esmed; Saifaddin, Burhan; Almogbel, Abdullah; Iza, Michael; Nakamura, Shuji; DenBaars, Steven P.; Speck, James S.

    2017-01-01

    temperature on the AlN crystal quality, defect density, and surface morphology. The crystal quality was characterized using omega rocking curve scans and the threading dislocation density was determined by plan view transmission electron microscopy. The growth

  2. Synthesis of full-density nanocrystalline tungsten carbide by reduction of tungstic oxide at room temperature

    International Nuclear Information System (INIS)

    El-Eskandarany, M.S.; Omori, M.; Ishikuro, M.; Konno, T.J.; Takada, K.; Sumiyama, K.; Hirai, T.; Suzuki, K.

    1996-01-01

    Among the hard alloys, WC alloys find wide industrial applications as tips for cutting tools and wear-resistant parts. Their intrinsic resistance to oxidation and corrosion at high temperatures also makes them desirable as a protective coating for devices at elevated temperatures. In the industrial scale of production, WC is prepared by a direct union of the elements at a temperature of 3,273 to 3,473 K. Accordingly, the high cost of preparation is a disadvantage of this process. Here, the authors report a novel technique for preparing a large amount of WC powder using a simple method. This process is based on mechanical solid-state reduction (MSSR) followed y solid-state reaction (SSR) during room-temperature ball milling (a high energy ball mill, Fritsch P6, was used at a rotation speed of 4.2 s -1 ) of a mixture of WO 3 , Mg, and C powders

  3. High temperature strengthening mechanism of hafnium carbide in a tungsten-rhenium matrix

    International Nuclear Information System (INIS)

    Luo, A.; Shin, K.S.; Jacobson, D.L.

    1991-01-01

    The interrelationship between the testing temperature and HfC strength increment of an arc-melted W-3.6Re-0.4HfC was determined from 1950 K to 2980 K in a vacuum of better than 1.3x10 -5 Pa (10 -7 torr). The present research was focused on the characteristic temperature at which the rapid coarsening of HfC particles occurred and the effect of the second-phase particle size on the high temperature strength properties of this material. It was found that the HfC particle strengthening was effective in a W-Re matrix up to a characteristic temperature of 2450 K in the short-term tensile test. Carbon was found to be the rate-limiting solute in the HfC particle growth. The strength of HfC strengthened alloy at temperature above 0.5 T m is proportional to the square root of particle volume fraction. The yield strengths of W-3.6Re-0.26HfC calculated based on the particle statistical distribution had good agreement with the experimental values from 1950 K to 2980 K. Besides, an addition of 0.26 percent HfC in tungsten resulted in about 28 percent increase in the activation energy of plastic deformation at high temperatures

  4. Effect of deposition conditions on the properties of pyrolytic silicon carbide coatings for high-temperature gas-cooled reactor fuel particles

    International Nuclear Information System (INIS)

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

    1977-10-01

    Silicon carbide coatings on HTGR microsphere fuel act as the barrier to contain metallic fission products. Silicon carbide coatings were applied by the decomposition of CH 3 SiCl 3 in a 13-cm-diam (5-in.) fluidized-bed coating furnace. The effects of temperature, CH 3 SiCl 3 supply rate and the H 2 :CH 3 SiCl 3 ratio on coating properties were studied. Deposition temperature was found to control coating density, whole particle crushing strength, coating efficiency, and microstructure. Coating density and microstructure were also partially determined by the H 2 :CH 3 SiCl 3 ratio. From this work, it appears that the rate at which high quality SiC can be deposited can be increased from 0.2 to 0.5 μm/min

  5. 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

  6. 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

  7. Ultra-high temperature oxidation behavior of chemical vapor deposited silicon carbide layers

    International Nuclear Information System (INIS)

    Goto, Takashi

    2003-01-01

    The active oxidation, passive oxidation and bubble formation of CVD SiC were studied in O 2 and CO 2 at temperatures from 1650 to 2000 K. The active oxidation rates in O 2 increased with increasing oxygen partial pressure (P o2 ); however, those in CO 2 showed the maxima at specific P o2 . The passive oxidation kinetics in O 2 were either linear-parabolic or parabolic depending on temperature and P o2 , whereas that in CO 2 was always parabolic. The activation energies for the parabolic oxidation in O 2 and CO 2 were 210 and 150 kJ/mol, respectively, suggesting different rate-determining process between these atmospheres. The bubble formation was controlled by temperature and P o2 being independent of oxidant gas species. (author)

  8. 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

  9. Low threading dislocation density aluminum nitride on silicon carbide through the use of reduced temperature interlayers

    KAUST Repository

    Foronda, Humberto M.

    2017-11-23

    In this work, reduced threading dislocation density AlN on (0 0 0 1) 6H-SiC was realized through the use of reduced temperature AlN interlayers in the metalorganic chemical vapor deposition growth. We explored the dependence of the interlayer growth temperature on the AlN crystal quality, defect density, and surface morphology. The crystal quality was characterized using omega rocking curve scans and the threading dislocation density was determined by plan view transmission electron microscopy. The growth resulted in a threading dislocation density of 7 × 108 cm−2 indicating a significant reduction in the defect density of AlN in comparison to direct growth of AlN on SiC (∼1010 cm−2). Atomic force microscopy images demonstrated a clear step-terrace morphology that is consistent with step flow growth at high temperature. Reducing the interlayer growth temperature increased the TD inclination and thus enhanced TD-TD interactions. The TDD was decreased via fusion and annihilation reactions.

  10. Distribution of temperature and deformations during resistance butt welding of uranium rods with titanium

    International Nuclear Information System (INIS)

    Tatarinov, V.R.; Krasnorutskij, V.S.

    1977-01-01

    Results are described on studying time-temperature and deformation parameters for resistance welding of uranium rods with titanium. It is shown that in the first period of welding (approximately 2/3 tsub(wel.)) the maxima of weld temperature and weld deformation deviate to titanium, and in the final period uranium deformation reaches the level of maximum lateral deformation of titanium. For faying surfaces with minimum weld deformation the joint cleaning of contaminants and oxides is insufficient, which results in lower weld quality

  11. 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

  12. 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

  13. Preparation of silicon carbide/carbon fiber composites through high-temperature spark plasma sintering

    Directory of Open Access Journals (Sweden)

    Ehsan Ghasali

    2017-12-01

    Full Text Available This study discusses the potentials of spark plasma sintering (SPS integrated with high temperature process that can enable sintering of SiC/Cf composites without any sintering aids. The random distribution of carbon fibers was obtained through mixing composite components in ethanol by using a shaker mill for 10 min. The corresponding sintering process was carried out at 1900 and 2200 °C with 50 MPa pressure applied at maximum temperature. The results showed that 89 ± 0.9 and 97 ± 0.8% of the theoretical density can be obtained for sintering temperatures of 1900 and 2200 °C, respectively. The densification curves were plotted to monitor sintering behavior with punch displacement changes. The appropriate bonding between SiC particles and carbon fibers was detected using FE-SEM for sample which was sintered at 2200 °C. The clear maximum in hardness (2992 ± 33 Vickers, bending strength (427 ± 26 MPa and fracture toughness (4.2 ± 0.3 MPa m1/2 were identified for sample sintered at 2200 °C. XRD investigations supposed that SiC and carbon were the only crystalline phases in both sintered samples.

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

    Science.gov (United States)

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

    2013-09-03

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

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

    Science.gov (United States)

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

    2011-03-01

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

  16. The Effect of Grain Size on the Radiation Response of Silicon Carbide and its Dependence on Irradiation Species and Temperature

    Science.gov (United States)

    Jamison, Laura

    In recent years the push for green energy sources has intensified, and as part of that effort accident tolerant and more efficient nuclear reactors have been designed. These reactors demand exceptional material performance, as they call for higher temperatures and doses. Silicon carbide (SiC) is a strong candidate material for many of these designs due to its low neutron cross-section, chemical stability, and high temperature resistance. The possibility of improving the radiation resistance of SiC by reducing the grain size (thus increasing the sink density) is explored in this work. In-situ electron irradiation and Kr ion irradiation was utilized to explore the radiation resistance of nanocrystalline SiC (nc-SiC), SiC nanopowders, and microcrystalline SiC. Electron irradiation simplifies the experimental results, as only isolated Frenkel pairs are produced so any observed differences are simply due to point defect interactions with the original microstructure. Kr ion irradiation simulates neutron damage, as large radiation cascades with a high concentration of point defects are produced. Kr irradiation studies found that radiation resistance decreased with particle size reduction and grain refinement (comparing nc-SiC and microcrystalline SiC). This suggests that an interface-dependent amorphization mechanism is active in SiC, suggested to be interstitial starvation. However, under electron irradiation it was found that nc-SiC had improved radiation resistance compared to single crystal SiC. This was found to be due to several factors including increased sink density and strength and the presence of stacking faults. The stacking faults were found to improve radiation response by lowering critical energy barriers. The change in radiation response between the electron and Kr ion irradiations is hypothesized to be due to either the change in ion type (potential change in amorphization mechanism) or a change in temperature (at the higher temperatures of the Kr ion

  17. Self propagating high temperature synthesis of mixed carbide and boride powder systems for cutting tools manufacturing

    International Nuclear Information System (INIS)

    Vallauri, D.; Cola, P.L. de; Piscone, F.; Amato, I.

    2001-01-01

    TiC-TiB 2 composites have been produced via SHS technique starting from low cost raw materials like TiO 2 , B 4 C, Mg. The influence of the diluent phase (Mg, TiC) content on combustion temperature has been investigated. The use of magnesium as the reductant phase allowed acid leaching of the undesired oxide product (MgO), leaving pure hard materials with fine particle size suitable to be employed in cutting tools manufacturing through cold pressing and sintering route. The densification has shown to be strongly dependent on the wetting additions. The influence of the metal binder and wetting additions on the sintering process has been investigated. A characterization of the obtained materials was performed by the point of view of cutting tools life (hardness, toughness, strength). (author)

  18. 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.

  19. 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

  20. 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.

  1. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

    International Nuclear Information System (INIS)

    Forsberg, C. W.; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of 7 LiF and BeF 2 (FLiBe) possessing a boiling point above 1300 C and the figure of merit ρC p (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

  2. KINETICS OF THE REACTION OF ELEMENTAL FLUORINE WITH ZIRCONIUM CARBIDE AND ZIRCONIUM DIBORIDE AT HIGH TEMPERATURES

    Energy Technology Data Exchange (ETDEWEB)

    Kuriakose, A. K.; Margrave, J. L.

    1963-09-15

    The reaction between ZrC and F/sub 2/ was investigated at 278 to 410 deg C, using 31 mm HgF/syb 2/. The reaction was found to be linear with time, and linear rate constants were computed. The activation energy was determined to be 22.1 plus or minus 1.6 kcal/mole. ZrB/sub 2/ is not attacked by 31 mm HgF/sub 2/ below 500 deg C. The weight losses from reaction of ZrB/sub 2/ with F/sub 2/ at 600 to 900 deg C and of ZrC with F/sub 2/ at 700 to 950 deg C, were rneasured for a F/sub 2/ pressure of 2.7 mm Hg. Zero-time linear rate constants were calculated and found not to be strongly temperature-dependent above 600 deg C, and the activation energies are essentially zero for both ZrB/sub 2/ and ZrC. For ZrC at 350 deg C and for ZrB/sub 2/ at 700 deg C, the rate is approximately proportional to the square root of F/sub 2/ partial pressure, while for ZrC at 700 deg C, it is proportional to the 1.5 power of F/sub 2/ partial pressure. (D.L.C.)

  3. Densification of boron carbide at relatively low temperatures by hot pressing and hot isostatic pressing

    International Nuclear Information System (INIS)

    Telle, R.

    1988-01-01

    The poor sinterability of B 4 C limits its widespread application because both high temperatures and high pressures are required for a complete densification. Moreover, B 4 C suffers from a low strength and fracture toughness, possesses, however, a high potential because of its extreme hardness. Reaction hot pressing of B 4 C-WC-TiC-Si-Co mixtures resulting in B 4 C-TiB 2 -W 2 B 5 composites of high density exhibit remarkable mechanical properties. The influence of hot isostatic pressing (HIP) on the microstructure and the mechanical properties is investigated in cooperation with participants of the COST 503 activities and related to the strengthening and toughening mechanisms. Difficulties during densification by HIP arise from the evaporation of adsorbed volatiles as well as from the strong swelling of the powder compact due to the sintering reaction. Several HIP cycle designs were tested in order to prevent the bloating of the capsule and to control internal stresses due to the misfit of the thermal expansion of the entire phases. In comparison to single phase B 4 C ceramics, bending strength was improved to 1030 MPa, K Ic to 5.2 MPa/m, while hardness was comparable with HV1=38 GPa. Wear test were performed and related to the toughening mechanisms. (orig.) With 56 refs., 9 tabs., 64 figs

  4. Study about uranium oxides at high temperature by X-ray diffraction

    International Nuclear Information System (INIS)

    Costa, M.I.

    1978-01-01

    In this work a technique to study the lattice parameters in the crystalline substances at hight temperature by X-rays diffraction is developed. The results obtained agree very well with the experimental data found in the literature. The crystalline structure of uranium oxide at different temperature is studied in detail by this technique. At the range of the temperature investigated, i.e., 20 0 C to 640 0 C, the following forms for uranium oxide: U 3 O 8 in its hexagonal modification, cubic UO 2 , cubic U 4 O 9 and tetragonal U 3 O 7 is observed. The appearance of two hexagonal units observed in this work is identified by Milne. A good reproducibillity is observed for measurements at the same temperature [pt

  5. 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)

  6. Uranium luminescence in La2 Zr2 O7 : effect of concentration and annealing temperature.

    Science.gov (United States)

    Mohapatra, M; Rajeswari, B; Hon, N S; Kadam, R M

    2016-12-01

    The speciation of a particular element in any given matrix is a prerequisite to understanding its solubility and leaching properties. In this context, speciation of uranium in lanthanum zirconate pyrochlore (La 2 Zr 2 O 7  = LZO), prepared by a low-temperature combustion route, was carried out using a simple photoluminescence lifetime technique. The LZO matrix is considered to be a potential ceramic host for fixing nuclear and actinide waste products generated during the nuclear fuel cycle. Special emphasis has been given to understanding the dynamics of the uranium species in the host as a function of annealing temperature and concentration. It was found that, in the LZO host, uranium is stabilized as the commonly encountered uranyl species (UO 2 2+ ) up to a heat treatment of 500 °C at the surface. Above 500 °C, the uranyl ion is diffused into the matrix as the more symmetric octahedral uranate species (UO 6 6- ). The uranate ions thus formed replace the six-coordinated 'Zr' atoms at regular lattice positions. Further, it was observed that concentration quenching takes place beyond 5 mol% of uranium doping. The mechanism of the quenching was found to be a multipolar interaction. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

  7. Comparative study of the oxidation of various qualities of uranium in carbon dioxide at high temperatures

    International Nuclear Information System (INIS)

    Desrues, R.; Paidassi, J.

    1965-01-01

    Uranium samples of six different qualities were subjected, in the temperature range 400 - 1000 C, to the action of carbon dioxide carefully purified to eliminate oxygen and water vapour; the resulting oxidation was followed micro-graphically and also (but only in the range 400 - 700 C) gravimetrically using an Ugine-Eyraud microbalance. A comparison of the results leads to the following 3 observations. First, the oxidation of the six uraniums studied obeys a linear law, (followed at 700 C by an accelerating law). The rates of reaction differ by a maximum of 100 per cent, the higher purity grades being oxidized more slowly except at 700 C when the reverse is true. Secondly, simultaneously with the growth, of an approximately uniform film of uranium dioxide on the metal, there occurs a localized attack in the form of blisters in the immediate neighbourhood of the monocarbide inclusions in the uranium. The relative importance of this attack is greater for lower oxidation temperatures and for a larger size, number and inequality of distribution of the inclusions, that is to say for higher carbon concentrations in the uranium (which have values from 7 to 1000 ppm in our tests). Thirdly, for oxidation temperatures above 600 C blistering is much less pronounced, but at 700 C the beginning of a general deformation of the sample occurs, which, above 750 C, becomes much greater; this leads to an acceleration of the reaction rate with respect to the linear law. In view of the over-heating, the sample must already be in the γ-phase which is particularly easily deformed; furthermore this expansion phenomenon is more pronounced when the sample is more plastic and therefore purer. (authors) [fr

  8. 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

  9. 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

  10. Viscoplastic behavior of uranium dioxide at high temperature; Comportement viscoplastique du dioxyde d'uranium a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Sauter, F

    2001-02-01

    This work is a part of a project led by EDF the purpose of which is the development of more predictive models to describe the thermomechanical behavior of fuel assembly. First, we recall the baselines of the Power Water Reactors then we deal with the viscoplastic behavior of uranium dioxide (UO{sub 2}). This knowledge enables an accurate description of the stress relaxation during Pellet Cladding Interactions. The pellets we have used in the last part are similar to the industrial ones. They exhibit a yield point during strain hardening tests and a sigma creep curve. In order to describe these characteristics, we have adapted different kind of approaches: thermodynamical - the Distribution of Non Linear Relaxations, approaches based on dislocation glide inspired by Alexander and Haasen and introduced in the Pilvin polycrystalline model. We recall the purpose of internal variables in the thermodynamics of system far from equilibrium then in case of a viscoplastic flow controlled by dislocation glide, we establish a link between densities of dislocations and internal variables in the D.N.L.R. approach. As vacancy diffusion in the grain boundary has a contribution to the viscoplastic strain, a similar is presented in appendix. These models are able to reproduce the behavior of UO{sub 2} pellets in strain hardening, stress relaxation and creep tests. Much possible progress has been revealed by the analysis of the tests. Further more, we propose a model for yield point and sigma creep curve. We also have extended these results to the behavior of irradiated pellets and stressed the influence of damage. (author)

  11. Temperature and direction dependence of internal strain and texture evolution during deformation of uranium

    Energy Technology Data Exchange (ETDEWEB)

    Brown, D.W., E-mail: dbrown@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Bourke, M.A.M.; Clausen, B.; Korzekwa, D.R.; Korzekwa, R.C.; McCabe, R.J.; Sisneros, T.A.; Teter, D.F. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2009-06-25

    Depleted uranium is of current programmatic interest at Los Alamos National Lab due to its high density and nuclear applications. At room temperature, depleted uranium displays an orthorhombic crystal structure with highly anisotropic mechanical and thermal properties. For instance, the coefficient of thermal expansion is roughly 20 x 10{sup -6} deg. C{sup -1} in the a and c directions, but near zero or slightly negative in the b direction. The innate anisotropy combined with thermo-mechanical processing during manufacture results in spatially varying residual stresses and crystallographic texture, which can cause distortion, and failure in completed parts, effectively wasting resources. This paper focuses on the development of residual stresses and textures during deformation at room and elevated temperatures with an eye on the future development of computational polycrystalline plasticity models based on the known micro-mechanical deformation mechanisms of the material.

  12. Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element

    International Nuclear Information System (INIS)

    Delaplace, J.

    1960-09-01

    By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the γ → β transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the β → α transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form β at ordinary temperatures after quenching from the β and γ regions. The β phase is particularly unstable and changes into needles of the α form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The β phase obtained by quenching from the β phase region is more stable than that obtained by quenching from the γ region. Chromium is a more effective stabiliser of the β phase than is iron. Unfortunately it causes serious surface cracking. The β → α transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct γ → α transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C/s. He has however observed the formation of several martensitic structures. (author) [fr

  13. High temperature chlorination of uranium and some radionuclides from rich sulphide ores

    International Nuclear Information System (INIS)

    Mahdy, M.A.

    1992-01-01

    This work is concerned with the application of the high temperature chlorination technique upon a sulphide-rich uranium ore from elliot lake, ontario, canada. The purpose is to find a substitute to conventional sulphuric acid leaching which involves both acid drainage and radionuclide dissolution problems. Test work has therefore been directed towards studying some relevant factors of chlorination beside the effect of a number of additives

  14. 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

  15. Uranium-thorium fuel cycle in a very high temperature hybrid system

    International Nuclear Information System (INIS)

    Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.

    2011-01-01

    Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)

  16. Temperature effect on uranium retention onto Zr2O(PO4)2 surface

    International Nuclear Information System (INIS)

    Almazan Torres, M.G.

    2007-03-01

    Uranium sorption onto Zr 2 O(PO 4 ) 2 has been studied between 298 K and 363 K, in 0.1 M NaClO 4 medium. Potentiometric titrations were realized to determine temperature dependency of the acid-base properties (pH(pcn), acidity constants). Classical batch experiments were performed at different temperatures. The sorption experiments revealed that the uranium sorption onto Zr 2 O(PO 4 ) 2 is favoured with the temperature. Structural characterization of the surface complexes was performed by both Time-Resolved Laser-Induced Fluorescence (TRLIF) and EXAFS spectroscopy. The TRLIF measurements vs. temperature revealed two uranyl surface complexes. No influence of the temperature onto the nature surface complex was observed. The EXAFS analysis showed a splitting of the equatorial oxygen atoms in two shells, corresponding to uranyl bidentate, inner-sphere complexes. The obtained structural uranyl surface complex information was used to simulate (using a constant capacitance model) the sorption edges. The proposed complexes equilibrium model consists of the following surface complexes: (ZrOH) 2 UO 2 2+ and (PO) 2 UO 2 . Besides the stability constants for the surface complexes, the thermodynamic parameters ΔH 0 and ΔS 0 were determined using the van't Hoff equation. The enthalpy values associated to the U(VI) retention onto Zr 2 O(PO 4 ) 2 , determined by the temperature dependence of the stability constants, testify that the formation of the complex (PO) 2 UO 2 (55 kJ/mol) is endothermic, while no influence of the temperature was observed for the formation of the complex (ZrOH) 2 UO 2 2+ . The adsorption reaction of the last complex is then driven by entropy. In addition, calorimetric measurements of uranium sorption onto Zr 2 O(PO 4 ) 2 were carried out to directly quantify the enthalpy associated to the retention processes. (author)

  17. Existence and structure of rare-earth mono-carbides: study of their low-temperature magnetic properties; Existence et structure des monocarbures de terres rares. Etude de leurs proprietes magnetiques a basses temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Lallement, R [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

    1966-07-01

    There are two types of rare earth carbides, the first one is face-centered cubic, stable at high temperature, and very hypo-stoichiometric (formula MC{sub x} with 0.35 < x < 0.65); the other is rhombohedric, stable at lower temperature, with a formula M{sub 2}C. These two carbides are magnetically ordered at low temperatures (ferro or ferri-magnetism). They are highly anisotropic. A great part of the electric and magnetic properties can be explained from the following ideas: the M{sup 3+} ions are coupled via the conduction electrons, there are more conduction electrons in the carbides than in the metals, and there is some local order around the transition temperatures. (author) [French] Nous avons mis en evidence l'existence de deux carbures de terres rares, l'un de formule MC{sub x} (0,35 < x < 0,65) de structure cubique a faces centrees, stable a haute temperature, l'autre de formule M{sub 2}C, de structure rhomhoedrique, stable a temperature moyenne. Ces deux types de carbures presentent des phenomenes d'ordre magnetique a basses temperatures (ferro ou ferrimagnetisme). Ils sont caracterises par une forte anisotropie magnetique. Une grande partie des proprietes electriques et magnetiques s'explique a partir des hypotheses suivantes: a) Les ions M{sup 3+} sont couples entre eux par l'intermediaire des electrons de conduction; b) Le nombre d'electrons de conduction dans les carbures est plus grand que dans les metaux; c) Autour des temperatures de transition se manifestent des phenomenes d'ordre local. (auteur)

  18. Experimental studies of neutron irradiated uranium dioxide at high temperatures

    International Nuclear Information System (INIS)

    Tanke, R.H.J.

    1990-01-01

    In case of an accident situation, in which the heat of the nuclear fuel can no longer be transferred to coolin water, the temperature of the nuclear fuel ay rise very strongly, so that radioactive fission products may be released, which can ultimately lead to the release of radioactive substances to the environment. In this respect it is important to know more about the release rate of the various fission products and their fuel samples, used in the investigation, were UO-2 spheres of approximately 1 mm. The chemical forms of the particles which are being released from the sphees during evaporation have been determined using a mass spectrometer. At the same time, the activity of the fission products has been measured using a gamma spectrometer. A gamma tomographer has been developed for determining the three-dimensional distribution of the concentration of radioactive fission products in the sphere. With this tomographer the change of this distribution as a function of temperature could be measured. For interpretation of the results two models have been developed: a model of the evaporation of the non-stoichiometric UO-2, and a model of the diffusion of fission products in UO-2. The first model was used to determine the stoichiometry of the sphere while the second has been used to determine the activation energy for the diffusion of the fission products. The main conclusion is that the microstructure of the nuclear fuel has a great effect on both the amount of free oxygen atoms, the release rate and the chemical form of fission products. This microstructure has not been investigated in greater detail so that all other conclusions are of qualitative nature. (author). 111 refs.; 114 figs.; 13 tabs

  19. Room temperature photoluminescence spectrum modeling of hydrogenated amorphous silicon carbide thin films by a joint density of tail states approach and its application to plasma deposited hydrogenated amorphous silicon carbide thin films

    International Nuclear Information System (INIS)

    Sel, Kıvanç; Güneş, İbrahim

    2012-01-01

    Room temperature photoluminescence (PL) spectrum of hydrogenated amorphous silicon carbide (a-SiC x :H) thin films was modeled by a joint density of tail states approach. In the frame of these analyses, the density of tail states was defined in terms of empirical Gaussian functions for conduction and valance bands. The PL spectrum was represented in terms of an integral of joint density of states functions and Fermi distribution function. The analyses were performed for various values of energy band gap, Fermi energy and disorder parameter, which is a parameter that represents the width of the energy band tails. Finally, the model was applied to the measured room temperature PL spectra of a-SiC x :H thin films deposited by plasma enhanced chemical vapor deposition system, with various carbon contents, which were determined by X-ray photoelectron spectroscopy measurements. The energy band gap and disorder parameters of the conduction and valance band tails were determined and compared with the optical energies and Urbach energies, obtained by UV–Visible transmittance measurements. As a result of the analyses, it was observed that the proposed model sufficiently represents the room temperature PL spectra of a-SiC x :H thin films. - Highlights: ► Photoluminescence spectra (PL) of the films were modeled. ► In the model, joint density of tail states and Fermi distribution function are used. ► Various values of energy band gap, Fermi energy and disorder parameter are applied. ► The model was applied to the measured PL of the films. ► The proposed model represented the room temperature PL spectrum of the films.

  20. Uranium recovery by leaching with sodium carbonate at high temperature and pressure

    International Nuclear Information System (INIS)

    Soerensen, E.; Koefoed, S.; Lundgaard, T.

    1983-11-01

    The principal uranium bearing mineral in Greenland steenstrupine is a complex sodium REE phosphosilicate in which Fe, Mn, Th, U are minor constituents. The Na 2 CO 3 extractant is used for specially acidconsuming ores. However, steenstrupine is decomposed by Na 2 CO 3 only at temperatures above 220degC, so the leaching must be carried out under pressure. Laboratory tests have shown the optimal temperature to be 260degC and the leach liquor composition120 g/l of NaHCO 3 and 20 g/l of Na 2 CO 3 . Addition of oxygen is necessary as uranium will not dissolve in carbonate unless it is brought in its highest state of oxidation. According to the laboratory tests it may be estimated that 1 kg of ore suspended in 1 l of leach liquor and ground to 80% minus 200 mesh can be extracted in 20-40 minutes. On the basis of data obtained a process was suggested in which the ore is ground with carbonate leach liquor to a suitable suspension which is fed to an autoclave with a retentiontime of 20 minutes at 260degC. The residue is filtered off and the liquor reused for grinding and ex- traction. The demand for a reaction temperature near 300degC, a pressure up to 120 atm. and a continuos operation favours a tubular flow autoclave with so narrow a bore that the turbulence provides the mechanical agitation of the suspension. From the mined material it appears that more than 80% of the uranium can be extracted in the pipe autoclave. Some samples give off the obtainable uranium in 20 minutes. The precipitated yellow cake is contaminated with more Na and Si than admitted by international standards. (EG)

  1. The temperature coefficient of the resonance integral for uranium metal and oxide

    Energy Technology Data Exchange (ETDEWEB)

    Blomberg, P; Hellstrand, E; Homer, S

    1960-06-15

    The temperature coefficient of the resonance integral in uranium metal and oxide has been measured over a wide temperature range for rods with three different diameters. The results for metal agree with most earlier results from activation measurements but differ as much as a factor of two from results obtained with reactivity methods. For oxide only one measurement has been reported recently. Our value is considerably lower than the result of that measurement. The experiments will continue in order to find the reason for the large discrepancy mentioned above.

  2. The temperature coefficient of the resonance integral for uranium metal and oxide

    International Nuclear Information System (INIS)

    Blomberg, P.; Hellstrand, E.; Homer, S.

    1960-06-01

    The temperature coefficient of the resonance integral in uranium metal and oxide has been measured over a wide temperature range for rods with three different diameters. The results for metal agree with most earlier results from activation measurements but differ as much as a factor of two from results obtained with reactivity methods. For oxide only one measurement has been reported recently. Our value is considerably lower than the result of that measurement. The experiments will continue in order to find the reason for the large discrepancy mentioned above

  3. Plastic Flow Characteristics of Uranium-Niobium as a Function of Strain Rate and Temperature

    International Nuclear Information System (INIS)

    Cady, C.M.; Gray, G.T. III; Hecker, S.S; Thoma, D.J.; Korzekwa, D.R.; Patterson, R.A.; Dunn, P.S.; Bingert, J.F.

    1999-01-01

    The stress-strain response of uranium-niobium alloys as a function of temperature, strain-rate and stress-state was investigated. The yield and flow stresses of the U-Nb alloys were found to exhibit a pronounced strain rate sensitivity, while the hardening rates were found to be insensitive to strain rate and temperature. The overall stress-strain response of the U-6Nb exhibits a sinusoidal hardening response, which is consistent with multiple deformation modes and is thought to be related to shape-memory behavior

  4. Uranium recovery by leaching with sodium carbonate at high temperature and pressure

    International Nuclear Information System (INIS)

    Soerensen, E.; Koefoed, S.; Lundgaard, T.

    1990-09-01

    An alkaline rock from the Ilimaussaq instrusion, SW Greenland, was proposed as a source of uranium. Its principal uranium bearing mineral, Steenstrupine, is a complex sodium REE phosphosilicate in which Fe, Mn, Th and U are minor constituents. A special feature of this ore body is the content of water soluble minerals: NaF (Villiaumite), Na 2 Si 2 O 5 (Natrosilite) and an organic substance which displays the characteristics of humus. Sulfides are sparse, the most important one being ZnS (Sphalerite) of which the content is generally less than 0.5%. In the mineral under consideration (Lujavrite) the Steenstrupine is mainly finelay disseminated throughout the rock, yielding a uranium content of 300-400 ppm and thorium content of 800-1000 ppm. Laboratory tests indicated that high temperature carbonate leaching was necessary to decompose Steenstrupine. The optium temperature was shown to be 260 deg. C and the leach liquor composition 120 g/l of NaHCO 3 and 20 g/l of Na 2 C0 3 . Addition of oxygen is necessary. The process was developed to industrial scale in a continuous pipe autoclave with a retention time of 20 min. After filtering on a belt filter, the liquor was recycled several times to obtain a higher U-concentration. By reductive precipitation with iron powder a raw UO 2 was obtained. It was purified after dissolution in HNO 3 . An overall yield of 80% could be obtained. (author) 32 tabs., 13 ills., 24 refs

  5. Strength and rupture-life transitions caused by secondary carbide precipitation in HT-9 during high-temperature low-rate mechanical testing

    International Nuclear Information System (INIS)

    DiMelfi, R.J.; Gruber, E.E.; Kramer, J.M.; Hughes, T.H.

    1992-01-01

    The martensitic-ferritic alloy HT-9 is slated for long-term use as a fuel-cladding material in the Integral Fast Reactor. Analysis of published high-temperature mechanical property data suggests that secondary carbide precipitation would occur during service life causing substantial strengthening of the as-heat-treated material. Aspects of the kinetics of this precipitation process are extracted from calculations of the back stress necessary to produce the observed strengthening effect under various creep loading conditions. The resulting Arrhenius factor is shown to agree quantitatively with shifts to higher strength of crept material in reference to the intrinsic strength of HT-9. The results of very low constant strain-rate high-temperature tensile tests on as-heat-treated HT-9 that focus on the transition in strength with precipitation will be presented and related to rupture-life

  6. 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)

  7. The fracture mechanism of uranium-niobium alloys near hypoeutectoid composition aged at low temperature

    International Nuclear Information System (INIS)

    Wang Xiaoying; Ren Dapeng; Yang Jianxiong; Jiang Guifen

    2006-01-01

    The microstructures and the crack propagation of uranium-niobium alloys near hypoeutectoid composition aged at temperature 200 degree C for 2 hours during a tension was investigated by means of in situ tension tests using TEM. The results show that the twinning planes inside and between the martensite laths move and merge, and then disintegrate in uranium-niobium alloys with monoclinic α structure during the tension. The crack propagation can be described as follows. Under the tension, the thinning zone which is locally plastically deformed emerges in the front of the crack tip. After the process of nucleation, growth and conjunction, the microvoids connect with the main crack, which results in the fracture. Neither of emission, propagation and movement of dislocation was observed during the tension. (authors)

  8. Stress corrosion cracking of stainless steels under deaerated high-temperature water. Influence of grain boundary carbide precipitation, and effect of Mo and Cr in alloys

    International Nuclear Information System (INIS)

    Yamada, Takuyo; Terachi, Takumi; Miyamoto, Tomoki; Arioka, Koji

    2007-01-01

    In order to evaluate the influence of grain boundary carbide on IGSCC susceptibility of stainless steel, crack growth rate tests were performed under deaerated or 0.3 ppm hydrogenated pure water environments at 320degC using half-inch compact tension (CT) specimens. In our previous report, CT testing showed that the susceptibility of CW316 to IGSCC was inhibited by the precipitation of grain boundary carbide under these environments. The result suggested quite different behavior from that in an oxygenated high-temperature water environment. In this study, the influence of (1) Mo and (2) Cr content in alloys, and (3) Cr depletion at the grain boundary on the IGSCC growth behavior in stainless steel was studied at 320degC under a 0.3-ppm hydrogenated pure-water environment. As a result, (1) IGSCC growth was observed on non-sensitized CW20%316, CW20%304, CW20%20Cr316, and CW20%20Cr304 under a 0.3-ppm hydrogenated pure-water environment at 320degC. (2) IGSCC growth was not observed for sensitized CW20%316 and CW20%304 (at 650degC x 48 or 24 h) and healing heat-treated CW20%316 (at 650degC x 48 h + 900degC x 0.5 h) under the same water environment. (3) The susceptibility of high Cr content materials (CW20%20Cr316 and CW20% 20Cr304) to IGSCC resistance was improved that of conventional CW316 and CW304 under the same water environment. The higher Cr content is effective in inhibiting susceptibility to IGSCC, but the inhibiting effect of Cr content is smaller than the effect of the grain boundary carbide. (4) These differences in IGSCC suggest that grain boundary carbide has a beneficial effect in improving IGSCC resistance, at least in a 0.3-ppm hydrogenated pure-water environment, despite the Mo content and Cr depletion at grain boundary. (author)

  9. Solid-state interfacial reaction in molybdenum-carbide systems at high temperature-pressure, and its application to bonding technique

    International Nuclear Information System (INIS)

    Horiguchi, Akihiro; Suganuma, Katsuaki; Miyamoto, Yoshinari; Koizumi, Mitsue; Shimada, Masahiko.

    1986-01-01

    Diffusion couples of molybdenum with several carbides, i.e. SiC, B 4 C, TiC, ZrC, HfC and TaC, were heated at various temperatures ranging from 1500 to 1840 deg C under high pressures of 3 GPa and 100 MPa for up to 4 hr. The couples were then examined for the composition of reaction products, the growth rate of reaction layers, interfacial structures, and tensile strength. In case of Mo-transition metal carbides, Mo 2 C layer was mainly formed, so that the carbides, which had supplied carbon, resulted in having the nonstoichiometric composition near the interface. The activation energy for the growth of Mo 2 C layer in Mo-TiC system was 332 kJ/mol, and that in Mo-TaC system was 366 kJ/mol. In Mo-SiC system, Mo 2 C layer, the mixed phase of Mo 2 C and Mo 5 Si 3 , and Mo 5 Si 3 C layer were formed in order from the Mo side. In Mo-B 4 C system, the mixed phase of Mo 2 B and MoB, and Mo 2 BC layer appeared. The decomposed graphite from B 4 C was also observed between B 4 C and Mo 2 BC phase. The activation energy for the growth of total reaction layer in Mo-SiC system was 531 kJ/mol, and that in Mo-B 4 C system was 183 kJ/mol. It can be said that the growth of reaction layers is controlled by diffusion. The orientation of crystals was observed in all reaction products except for Mo 2 BC phase in Mo-B 4 C system and (Mo, Ta) 2 C phase in Mo-TaC system. In HIPed couples, the magnitude of tensile strength was dependent on the difference in thermal expansion coefficient between Mo and carbides. HIPed Mo-TaC couple had the best weldability among the systems examined in the present investigation. (author)

  10. 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.

  11. Thermodynamic functions and vapor pressures of uranium and plutonium oxides at high temperatures

    International Nuclear Information System (INIS)

    Green, D.W.; Reedy, G.T.; Leibowitz, L.

    1977-01-01

    The total energy release in a hypothetical reactor accident is sensitive to the total vapor pressure of the fuel. Thermodynamic functions which are accurate at high temperature can be calculated with the methods of statistical mechanics provided that needed spectroscopic data are available. This method of obtaining high-temperature vapor pressures should be greatly superior to the extrapolation of experimental vapor pressure measurements beyond the temperature range studied. Spectroscopic data needed for these calculations are obtained from infrared spectroscopy of matrix-isolated uranium and plutonium oxides. These data allow the assignments of the observed spectra to specific molecular species as well as the calculation of anharmonicities for monoxides, bond angles for dioxides, and molecular geometries for trioxides. These data are then employed, in combination with data on rotational and electronic molecular energy levels, to determine thermodynamic functions that are suitable for the calculation of high-temperature vapor pressures

  12. First-principles study on oxidation effects in uranium oxides and high-pressure high-temperature behavior of point defects in uranium dioxide

    Science.gov (United States)

    Geng, Hua Y.; Song, Hong X.; Jin, K.; Xiang, S. K.; Wu, Q.

    2011-11-01

    Formation Gibbs free energy of point defects and oxygen clusters in uranium dioxide at high-pressure high-temperature conditions are calculated from first principles, using the LSDA+U approach for the electronic structure and the Debye model for the lattice vibrations. The phonon contribution on Frenkel pairs is found to be notable, whereas it is negligible for the Schottky defect. Hydrostatic compression changes the formation energies drastically, making defect concentrations depend more sensitively on pressure. Calculations show that, if no oxygen clusters are considered, uranium vacancy becomes predominant in overstoichiometric UO2 with the aid of the contribution from lattice vibrations, while compression favors oxygen defects and suppresses uranium vacancy greatly. At ambient pressure, however, the experimental observation of predominant oxygen defects in this regime can be reproduced only in a form of cuboctahedral clusters, underlining the importance of defect clustering in UO2+x. Making use of the point defect model, an equation of state for nonstoichiometric oxides is established, which is then applied to describe the shock Hugoniot of UO2+x. Furthermore, the oxidization and compression behavior of uranium monoxide, triuranium octoxide, uranium trioxide, and a series of defective UO2 at 0 K are investigated. The evolution of mechanical properties and electronic structures with an increase of the oxidation degree are analyzed, revealing the transition of the ground state of uranium oxides from metallic to Mott insulator and then to charge-transfer insulator due to the interplay of strongly correlated effects of 5f orbitals and the shift of electrons from uranium to oxygen atoms.

  13. 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

  14. 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)

  15. 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.

  16. Microscopic mapping of specific contact resistances and long-term reliability tests on 4H-silicon carbide using sputtered titanium tungsten contacts for high temperature device applications

    Science.gov (United States)

    Lee, S.-K.; Zetterling, C.-M.; Ostling, M.

    2002-07-01

    We report on the microscopic mapping of specific contact resistances (rhoc) and long-term reliability tests using sputtered titanium tungsten (TiW) ohmic contacts to highly doped n-type epilayers of 4H-silicon carbide. The TiW ohmic contacts showed good uniformity with low contact resistivity of 3.3 x10-5 Omega cm2. Microscopic mapping of the rhoc showed that the rhoc had a distribution that decreased from the center to the edge of the wafer. This distribution of the rhoc is caused by variation of the doping concentration of the wafer. Sacrificial oxidation at high temperature partially recovered inductively coupled plasma etch damage. TiW contacts with platinum and gold capping layers have stable specific contact resistance at 500 and 600 degC in a vacuum chamber for 308 h.

  17. In-situ neutron diffraction characterization of temperature dependence deformation in α-uranium

    Science.gov (United States)

    Calhoun, C. A.; Garlea, E.; Sisneros, T. A.; Agnew, S. R.

    2018-04-01

    In-situ strain neutron diffraction measurements were conducted at temperature on specimens coming from a clock-rolled α-uranium plate, and Elasto-Plastic Self-Consistent (EPSC) modeling was employed to interpret the findings. The modeling revealed that the active slip systems exhibit a thermally activated response, while deformation twinning remains athermal over the temperature ranges explored (25-150 °C). The modeling also allowed assessment of the effects of thermal residual stresses on the mechanical response during compression. These results are consistent with those from a prior study of room-temperature deformation, indicating that the thermal residual stresses strongly influence the internal strain evolution of grain families, as monitored with neutron diffraction, even though accounting for these residual stresses has little effect on the macroscopic flow curve, except in the elasto-plastic transition.

  18. Vapor pressures and vapor compositions in equilibrium with hypostoichiometric uranium-plutonium dioxide at high temperatures

    International Nuclear Information System (INIS)

    Green, D.W.; Fink, J.K.; Leibowitz, L.

    1982-01-01

    Vapor pressures and vapor compositions in equilibrium with a hypostoichiometric uranium-plutonium dioxide condensed phase (U/sub 1-y/Pu/sub y/)O/sub 2-x/, as functions of T, x, and y, have been calculated for 0.0 less than or equal to x less than or equal to 0.1, 0.0 less than or equal to y less than or equal to 0.3, and for the temperature range 2500 less than or equal to T less than or equal to 6000 K. The range of compositions and temperatures was limited to the region of interest to reactor safety analysis. Thermodynamic functions for the condensed phase and for each of the gaseous species were combined with an oxygen potential model to obtain partial pressures of O, O 2 , Pu, PuO, PuO 2 , U, UO, UO 2 , and UO 3 as functions of T, x, and y

  19. 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.)

  20. 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

  1. 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.

  2. Waste arisings from a high-temperature reactor with a uranium-thorium fuel cycle

    International Nuclear Information System (INIS)

    1979-09-01

    This paper presents an equilibrium-recycle condition flow sheet for a high-temperature gas-cooled reactor (HTR) fuel cycle which uses thorium and high-enriched uranium (93% U-235) as makeup fuel. INFCE Working Group 7 defined percentage losses to various waste streams are used to adjust the heavy-element mass flows per gigawatt-year of electricity generated. Thorium and bred U-233 are recycled following Thorex reprocessing. Fissile U-235 is recycled one time following Purex reprocessing and then is discarded to waste. Plutonium and other transuranics are discarded to waste. Included are estimates of volume, radioactivity, and heavy-element content of wastes arising from HTR fuel element fabrication; HTR operation, maintenance, and decommissioning; and reprocessing spent fuel where the waste is unique to the HTR fuel cycle

  3. High-temperature vaporization of thorium-uranium mixed monocarbide (Th1-y, Uy)C

    International Nuclear Information System (INIS)

    Koyama, Tadafumi; Yamawaki, Michio

    1989-01-01

    Vaporization thermodynamics of thorium-uranium mixed monocarbide phase (Th 1-y , U y )C was studied by mass spectrometric Knudsen effusion method for the compositions of (Th 0.9 , U 0.1 )C 0.855 , (Th 0.8 , U 0.2 )C 0.973 and (Th 0.6 , U 0.4 )C 0.973 . The partial vapor pressures of Th(g) and U(g) and activities of Th and U of these mixed monocarbides were determined at temperatures ranging from about 2000 to 2200 K. Further, the partial pressures of Th(g) and U(g) and activities of Th and U of the stoichiometric mixed monocarbides (Th 1-y , U y )C 1.00 were evaluated by compensating for the effect of carbon content. The Gibbs energies of formation of stoichiometric (Th 1-y , U y )C 1.00 were also evaluated. (orig.)

  4. Effect of low-temperature thermomechanical treatment on mechanical properties of low-alloying molybdenum alloys with carbide hardening

    International Nuclear Information System (INIS)

    Bernshtejn, L.M.; Zakharov, A.M.; Veller, M.V.

    1978-01-01

    Presented are results of testing low-temperature thermomechanical treatment of low-alloying molybdenum alloys, including quenching from 2100 deg C, 40% deformation by hydroextrusion and aging at the temperature of 1200-1400 deg C. Tensile tests at room temperature with the following processing of results have shown that low-temperature thermomechanical treatment of low-alloying molybdenum alloys of Mo-Zr-C and Mo-Zr-Nb-C systems leads to a significant increase in low-temperature mechanical properties (strength properties - by 30-35%, ductility - by 30-40%) as compared with conventional heat treatment (aging after quenching). The treatment proposed increases resistance to small, as well as large plastic deformations, and leads to a simultaneous rise of strength and plastic properties at all stages of tensile test. Alloying of the Mo-Zr-C system with niobium increases both strength and plastic characteristics as compared with alloys without niobium when testing samples, subjected to low temperature thermomechanical treatment and conventional heat treatment at room temperature

  5. 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

  6. 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)

  7. 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

  8. 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

  9. Transition metal carbides (WC, Mo2C, TaC, NbC) as potential electrocatalysts for the hydrogen evolution reaction (HER) at medium temperatures

    DEFF Research Database (Denmark)

    Meyer, Simon; Nikiforov, Aleksey V.; Petrushina, Irina M.

    2015-01-01

    One limitation for large scale water electrolysis is the high price of the Pt cathode catalyst. Transition metal carbides, which are considered as some of the most promising non-Pt catalysts, are less active than Pt at room temperature. The present work demonstrates that the situation is different......C > TaC. Copyright (C) 2014, Hydrogen Energy Publications, LLC. Published by Elsevier Ltd. All rights reserved....

  10. Rectification properties of n-type nanocrystalline diamond heterojunctions to p-type silicon carbide at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Goto, Masaki; Amano, Ryo; Shimoda, Naotaka [Graduate School of Automotive Science, Kyushu University, Nishiku, Fukuoka 819-0395 (Japan); Kato, Yoshimine, E-mail: yoshimine.kato@zaiko.kyushu-u.ac.jp [Department of Materials Science and Engineering, Kyushu University, Nishiku, Fukuoka 819-0395 (Japan); Teii, Kungen [Department of Applied Science for Electronics and Materials, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2014-04-14

    Highly rectifying heterojunctions of n-type nanocrystalline diamond (NCD) films to p-type 4H-SiC substrates are fabricated to develop p-n junction diodes operable at high temperatures. In reverse bias condition, a potential barrier for holes at the interface prevents the injection of reverse leakage current from the NCD into the SiC and achieves the high rectification ratios of the order of 10{sup 7} at room temperature and 10{sup 4} even at 570 K. The mechanism of the forward current injection is described with the upward shift of the defect energy levels in the NCD to the conduction band of the SiC by forward biasing. The forward current shows different behavior from typical SiC Schottky diodes at high temperatures.

  11. An all optical system for studying temperature induced changes in polycrystalline diamond deposited on a tungsten carbide substrate

    CSIR Research Space (South Africa)

    Masina, BN

    2010-09-01

    Full Text Available In this poster the authors discussed the ability to heat an industrial diamond sample by means of optical absorption of a CO2 laser beam, and then measure the resulting temperature on the surface of the diamond optically by means of radiometry...

  12. Comparative study of the oxidation of various qualities of uranium in carbon dioxide at high temperatures; Etude comparative de l'oxydation de diverses qualites d'uranium dans l'anhydride carbonique aux temperatures elevees

    Energy Technology Data Exchange (ETDEWEB)

    Desrues, R; Paidassi, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Uranium samples of six different qualities were subjected, in the temperature range 400 - 1000 C, to the action of carbon dioxide carefully purified to eliminate oxygen and water vapour; the resulting oxidation was followed micro-graphically and also (but only in the range 400 - 700 C) gravimetrically using an Ugine-Eyraud microbalance. A comparison of the results leads to the following 3 observations. First, the oxidation of the six uraniums studied obeys a linear law, (followed at 700 C by an accelerating law). The rates of reaction differ by a maximum of 100 per cent, the higher purity grades being oxidized more slowly except at 700 C when the reverse is true. Secondly, simultaneously with the growth, of an approximately uniform film of uranium dioxide on the metal, there occurs a localized attack in the form of blisters in the immediate neighbourhood of the monocarbide inclusions in the uranium. The relative importance of this attack is greater for lower oxidation temperatures and for a larger size, number and inequality of distribution of the inclusions, that is to say for higher carbon concentrations in the uranium (which have values from 7 to 1000 ppm in our tests). Thirdly, for oxidation temperatures above 600 C blistering is much less pronounced, but at 700 C the beginning of a general deformation of the sample occurs, which, above 750 C, becomes much greater; this leads to an acceleration of the reaction rate with respect to the linear law. In view of the over-heating, the sample must already be in the {gamma}-phase which is particularly easily deformed; furthermore this expansion phenomenon is more pronounced when the sample is more plastic and therefore purer. (authors) [French] Des echantillons de six qualites d'uranium ont ete soumis, dans l'intervalle 400-1000 C, a l'action de l'anhydride carbonique tres soigneusement purifie en oxygene et en vapeur d'eau, et leur oxydation a ete suivie par voie micrographique et egalement (mais seulement entre 400

  13. Simultaneous determination of nitric acid and uranium concentrations in aqueous solution from measurements of electrical conductivity, density, and temperature

    International Nuclear Information System (INIS)

    Spencer, B.B.

    1991-01-01

    Nuclear fuel reprocessing plants handle aqueous solutions of nitric acid and uranium in large quantities. Automatic control of process operations requires reliable measurements of these solutes concentration, but this is difficult to directly measure. Physical properties such as solution density and electrical conductivity vary with solute concentration and temperature. Conductivity, density and temperature can be measured accurately with relatively simple and inexpensive devices. These properties can be used to determine solute concentrations will good correlations. This paper provides the appropriate correlations for solutions containing 2 to 6 Molar (M) nitric acid and 0 to 300 g/L uranium metal at temperatures from 25--90 degrees C. The equations are most accurate below 5 M nitric acid, due to a broad maximum in the conductivity curve at 6 M. 12 refs., 9 figs., 6 tabs

  14. Refining of high-temperature uranium melt by filtration through foam-ceramic filters

    International Nuclear Information System (INIS)

    Antsiferov, V.N.; Porozova, S.E.; Filippov, V.B.; Shtutsa, M.G.; Il'enko, E.V.; Kolotygina, N.S.

    2004-01-01

    An opportunity of applying foam-ceramic filters of corundum-mullite composition has been studied in refining natural uranium melts. Uranium melting conditions were chosen depending on technical characteristics of the foam ceramic filters. When their using, a portion of nonmetallic inclusions decreases by 20-30% (as little as 2.0-3.5% ingot weight), their size is reduced and their distribution in the ingot volume is equalized, contamination of uranium by the filter material being failed to be noticed. The parameters of foam-ceramic filters are optimized for provision of stable characteristics of uranium melt filtration process [ru

  15. Integration and High-Temperature Characterization of Ferroelectric Vanadium-Doped Bismuth Titanate Thin Films on Silicon Carbide

    Science.gov (United States)

    Ekström, Mattias; Khartsev, Sergiy; Östling, Mikael; Zetterling, Carl-Mikael

    2017-07-01

    4H-SiC electronics can operate at high temperature (HT), e.g., 300°C to 500°C, for extended times. Systems using sensors and amplifiers that operate at HT would benefit from microcontrollers which can also operate at HT. Microcontrollers require nonvolatile memory (NVM) for computer programs. In this work, we demonstrate the possibility of integrating ferroelectric vanadium-doped bismuth titanate (BiTV) thin films on 4H-SiC for HT memory applications, with BiTV ferroelectric capacitors providing memory functionality. Film deposition was achieved by laser ablation on Pt (111)/TiO2/4H-SiC substrates, with magnetron-sputtered Pt used as bottom electrode and thermally evaporated Au as upper contacts. Film characterization by x-ray diffraction analysis revealed predominately (117) orientation. P- E hysteresis loops measured at room temperature showed maximum 2 P r of 48 μC/cm2, large enough for wide read margins. P- E loops were measurable up to 450°C, with losses limiting measurements above 450°C. The phase-transition temperature was determined to be about 660°C from the discontinuity in dielectric permittivity, close to what is achieved for ceramics. These BiTV ferroelectric capacitors demonstrate potential for use in HT NVM applications for SiC digital electronics.

  16. 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)

  17. Influence of pyrolytic temperature on uranium adsorption capability by biochar derived from macauba coconut residue

    International Nuclear Information System (INIS)

    Guilhen, Sabine Neusatz; Fungaro, Denise Alves; Coleti, Jorge; Tenório, Jorge Alberto Soares

    2017-01-01

    Biochar (BC) is a carbon-rich product obtained when biomass is thermally decomposed at relatively low temperatures (under 700°C) and limited supply of oxygen in a process called pyrolysis. The conversion of biomass into BC can not only result in renewable energy source of synthetic gas and bio-oil, but also decrease the content of CO 2 in the atmosphere, as well as improving soil fertility. Because of its porous structure, charged surface and surface functional groups, BC exhibits a great potential as an adsorbent. Brazilian agro energy chain involves tons of biomass waste, providing a wide range of biomass with different chemical and physical properties. BC characteristics strongly depend on the feedstock and the pyrolysis conditions, in which the temperature is the key parameter. The aim of this study was to evaluate the adsorption potential for the removal of uranium, U(VI), from aqueous solutions using BC obtained through the pyrolysis of the Macauba (Acrocomia aculeata) coconut endocarp as a function of the final pyrolytic temperature. BCs produced at higher temperatures are likely to present lower H/C and O/C ratios, indicating the loss of easily degradable carbon compounds such as volatile matter. In contrast, low-temperature pyrolysis produces not only a higher BC yield, but also richer in surface functional groups which will likely enable interactions with the U(VI) ions. The endocarp was subjected to six different pyrolytic temperatures, ranging from 250°C to 750 °C. The influence of parameters such as pH, sorbent dose and initial concentration on the adsorption of U(VI) was investigated. The maximum adsorption capacity (q) was achieved for the BC obtained at 250°C (BC250), which presented a removal percentage of approx. 86%, demonstrating the potential of the BC from macauba endocarp for treatment of wastewaters. Thus, submitting the endocarp to temperatures higher than 250°C becomes unnecessary, saving time and reducing operating costs. (author)

  18. Influence of pyrolytic temperature on uranium adsorption capability by biochar derived from macauba coconut residue

    Energy Technology Data Exchange (ETDEWEB)

    Guilhen, Sabine Neusatz; Fungaro, Denise Alves [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Coleti, Jorge; Tenório, Jorge Alberto Soares, E-mail: snguilhen@ipen.br, E-mail: dfungaro@ipen.br, E-mail: jorgecoleti@usp.br, E-mail: jtenorio@usp.br [Universidade de São Paulo (USP), São Paulo, SP (Brazil). Departamento de Engenharia Metalúrgica e de Materiais

    2017-07-01

    Biochar (BC) is a carbon-rich product obtained when biomass is thermally decomposed at relatively low temperatures (under 700°C) and limited supply of oxygen in a process called pyrolysis. The conversion of biomass into BC can not only result in renewable energy source of synthetic gas and bio-oil, but also decrease the content of CO{sub 2} in the atmosphere, as well as improving soil fertility. Because of its porous structure, charged surface and surface functional groups, BC exhibits a great potential as an adsorbent. Brazilian agro energy chain involves tons of biomass waste, providing a wide range of biomass with different chemical and physical properties. BC characteristics strongly depend on the feedstock and the pyrolysis conditions, in which the temperature is the key parameter. The aim of this study was to evaluate the adsorption potential for the removal of uranium, U(VI), from aqueous solutions using BC obtained through the pyrolysis of the Macauba (Acrocomia aculeata) coconut endocarp as a function of the final pyrolytic temperature. BCs produced at higher temperatures are likely to present lower H/C and O/C ratios, indicating the loss of easily degradable carbon compounds such as volatile matter. In contrast, low-temperature pyrolysis produces not only a higher BC yield, but also richer in surface functional groups which will likely enable interactions with the U(VI) ions. The endocarp was subjected to six different pyrolytic temperatures, ranging from 250°C to 750 °C. The influence of parameters such as pH, sorbent dose and initial concentration on the adsorption of U(VI) was investigated. The maximum adsorption capacity (q) was achieved for the BC obtained at 250°C (BC250), which presented a removal percentage of approx. 86%, demonstrating the potential of the BC from macauba endocarp for treatment of wastewaters. Thus, submitting the endocarp to temperatures higher than 250°C becomes unnecessary, saving time and reducing operating costs

  19. 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)

  20. 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

  1. Uranium geochemistry, mineralogy, geology, exploration and resources

    International Nuclear Information System (INIS)

    De Vivo, B.

    1984-01-01

    This book comprises papers on the following topics: history of radioactivity; uranium in mantle processes; transport and deposition of uranium in hydrothermal systems at temperatures up to 300 0 C: Geological implications; geochemical behaviour of uranium in the supergene environment; uranium exploration techniques; uranium mineralogy; time, crustal evolution and generation of uranium deposits; uranium exploration; geochemistry of uranium in the hydrographic network; uranium deposits of the world, excluding Europe; uranium deposits in Europe; uranium in the economics of energy; role of high heat production granites in uranium province formation; and uranium deposits

  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. 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)

  5. 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.

  6. 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

  7. 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

  8. 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

  9. Metal electrodeposition and electron transfer studies of uranium compounds in room temperature ionic liquids

    International Nuclear Information System (INIS)

    Stoll, M.E.; Oldham, W.J.; Costa, D.A.

    2004-01-01

    Room temperature ionic liquids (RTIL's) comprised of 1,3-dialkylimidazolium or quaternary ammonium cations and one of several anions such as PF 6 - , BF 4 - , or - N(SO 2 CF 3 ) 2 , represent a class of solvents that possess great potential for use in applications employing electrochemical procedures. Part of the intrigue with RTIL's stems from some of their inherent solvent properties including negligible vapor pressure, good conductivity, high chemical and thermal stability, and non-flammability. Additionally, a substantial number of RTIL's can be envisioned simply by combining different cation and anion pairs, thereby making them attractive for specific application needs. We are interested in learning more about the possible use of RTIL's within the nuclear industry. In this regard our research team has been exploring the electron transfer behavior of simple metal ions in addition to coordination and organometallic complexes in these novel solvents. Results from our research have also provided us with insight into the bonding interactions between our current anion of choice, bis(trifluoromethylsulfonyl)imide = NTf 2 , and open coordination sites on actinide and transition metal fragments. This presentation will focus on recent results in two areas: the electrodeposition of electropositive metal ions from RTIL solutions and the electron transfer behavior for several uranium complexes. Details concerning the cathodic electrodeposition and anodic stripping of alkali metals (Na, K) from various working electrode surfaces (Pt, Au, W, Glassy Carbon) will be discussed. Figure 1 displays typical behavior for the electrodeposition of potassium metal from an RTIL containing potassium ions produced through the reaction of KH with H[NTf 2 ]. Our efforts with other metal ions, including our results to date with uranium electrodeposition, will be covered during the presentation. The electron transfer behavior for a number of uranium complexes have been studied with various

  10. 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

  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. 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.)

  13. Nuclear fuels for very high temperature applications

    International Nuclear Information System (INIS)

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

    1992-01-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO 2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures

  14. Hydrogen evolution activity and electrochemical stability of selected transition metal carbides in concentrated phosphoric acid

    DEFF Research Database (Denmark)

    Tomás García, Antonio Luis; Jensen, Jens Oluf; Bjerrum, Niels J.

    2014-01-01

    phosphoric acid were investigated in a temperature range from 80 to 170°C. A significant dependence of the activities on temperature was observed for all five carbide samples. Through the entire temperature range Group 6 metal carbides showed higher activity than that of the Group 5 metal carbides......Alternative catalysts based on carbides of Group 5 (niobium and tantalum) and 6 (chromium, molybdenum and tungsten) metals were prepared as films on the metallic substrates. The electrochemical activities of these carbide electrodes towards the hydrogen evolution reaction (HER) in concentrated...

  15. 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

  16. 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)

  17. 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

  18. Enriched-uranium feed costs for the High-Temperature Gas-Cooled reactor: trends and comparison with other reactor concepts

    International Nuclear Information System (INIS)

    Thomas, W.E.

    1976-04-01

    This report discusses each of the components that affect the unit cost for enriched uranium; that is, ore costs, U 3 O 8 to UF 6 conversion cost, costs for enriching services, and changes in transaction tails assay. Historical trends and announced changes are included. Unit costs for highly enriched uranium (93.15 percent 235 U) and for low-enrichment uranium (3.0, 3.2, and 3.5 percent 235 U) are displayed as a function of changes in the above components and compared. It is demonstrated that the trends in these cost components will probably result in significantly less cost increase for highly enriched uranium than for low-enrichment uranium--hence favoring the High-Temperature Gas-Cooled Reactor

  19. 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

  20. 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.

  1. Recovery of pure slaked lime from carbide sludge: Case study of ...

    African Journals Online (AJOL)

    Adaobi

    Carbide sludge is the by-product of reaction between calcium carbide and water in the production of ... soluble in water. The optimum percentage yield was 78.2% at a ratio of 1:1000(w/v) of sludge to water held for 24 h at room temperature. Key words: Carbide, recovery, ..... calcium carbonate and other calcium products.

  2. 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.)

  3. 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)

  4. The Influences of Uranium Concentration and Polyvinyl Alcohol on the Quality UO2 Microsphere for Fuel of High Temperature Reactor

    International Nuclear Information System (INIS)

    Damunir; Sukarsono; Bangun-Wasito; Endang Nawangsih

    2000-01-01

    The influences of uranium concentration and PVA on the quality of UO 2 microspheres for fuel of high temperature reactor have been investigated. The UO 2 particles were prepared by gel precipitation using internal gelation process. Uranyl nitrate solution containing uranium of 100 g/l was neutralized using NH 4 OH 1 M. The solution was changed into sol by adding 60 g PVA/l solution while stirred and heated up to 80 o C for 20 minutes. In order to find gels in spherical shape, the sol solution was dropped into 5 M NH 4 OH medium. The formed gels were small spheres, was washed, screened and heated up to 120 o C. After that, the gels were calcined at 800 o C for 4 hours, resulting in U 3 O 8 spheres. The U 3 O 8 particles were reduced using H 2 gas in a N 2 media at 800 o C for 4 hours, yielded in UO 2 spheres. Using a similar procedure, the influence of uranium concentration of 150-250 g/l and PVA 40-80 g/l were studied. The qualities of UO 2 particles were obtained by their physical properties, i.e. density, specific surface area, total volume of pores and pore radius using surface area meter and N 2 gas used as absorbent, and the particle size was observed using optical microscope. The result showed that the changing of uranium and PVA concentrations on the internal gelation affected the density, specific surface area, total volume of pores and pore radius of UO 2 particles. (author)

  5. Effect of small additions of silicon, iron, and aluminum on the room-temperature tensile properties of high-purity uranium

    International Nuclear Information System (INIS)

    Ludwig, R.L.

    1983-01-01

    Eleven binary and ternary alloys of uranium and very low concentrations of iron, silicon, and aluminum were prepared and tested for room-temperature tensile properties after various heat treatments. A yield strength approximately double that of high-purity derby uranium was obtained from a U-400 ppM Si-200 ppM Fe alloy after beta solution treatment and alpha aging. Higher silicon plus iron alloy contents resulted in increased yield strength, but showed an unacceptable loss of ductility

  6. Liquid phase sintering of carbides using a nickel-molybdenum alloy

    International Nuclear Information System (INIS)

    Barranco, J.M.; Warenchak, R.A.

    1987-01-01

    Liquid phase vacuum sintering was used to densify four carbide groups. These were titanium carbide, tungsten carbide, vanadium carbide, and zirconium carbide. The liquid phase consisted of nickel with additions of molybdenum of from 6.25 to 50.0 weight percent at doubling increments. The liquid phase or binder comprised 10, 20, and 40 percent by weight of the pressed powders. The specimens were tested using 3 point bending. Tungsten carbide showed the greatest improvement in bend rupture strength, flexural modulus, fracture energy and hardness using 20 percent binder with lesser amounts of molybdenum (6.25 or 12.5 wt %) added to nickel compared to pure nickel. A refinement in the carbide microstructure and/or a reduction in porosity was seen for both the titanium and tungsten carbides when the alloy binder was used compared to using the nickel alone. Curves depicting the above properties are shown for increasing amounts of molybdenum in nickel for each carbide examined. Loss of binder phase due to evaporation was experienced during heating in vacuum at sintering temperatures. In an effort to reduce porosity, identical specimens were HIP processed at 15 ksi and temperatures averaging 110 C below the sintering g temperature. The tungsten carbide and titanium carbide series containing 80 and 90 weight percent carbide phase respectively showed improvement properties after HIP while properties decreased for most other compositions

  7. Bacteriological lixiviation of low-grade uranium ores at low temperatures, by phiobacillus ferrooxidaus

    International Nuclear Information System (INIS)

    Lobato Filho, A.N.S.

    1976-12-01

    Laboratory experiments are described that, using selective and mutagenic agents, allowed the isolation of a strain of thiobacillus ferrooxidams capable of developing at 8 0 C, and keeping its oxidesing characteristics tests showed that the isoled sample is capable of solubilizing 95% of the uranium content in samples with U 3 O 8 content below 1000ppm [pt

  8. The effect of temperature on the sorption of technetium, uranium, neptunium and curium on bentonite, tuff and granodiorite

    International Nuclear Information System (INIS)

    Baston, G.M.N.; Berry, J.A.; Brownsword, M.; Heath, T.G.; Ilett, D.J.; Tweed, C.J.; Yui, M.

    1997-01-01

    A study of the sorption of the radioelements technetium; uranium; neptunium; and curium onto geological materials has been carried out as part of the PNC program to increase confidence in the performance assessment for a high-level radioactive waste repository in Japan. Batch sorption experiments have been performed in order to study the sorption of the radioelements onto bentonite, tuff and granodiorite from equilibrated de-ionized water under strongly-reducing conditions at both room temperature and at 60 C. Mathematical modelling using the geochemical speciation program HARPHRQ in conjunction with the HATCHES database has been undertaken in order to interpret the experimental results

  9. Corrosion of metallic materials by uranium hexafluoride at high temperatures (1963); Corrosion de materiaux metalliques par l'hexafluorure d'uranium a haute temperature (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-01

    The corrosion of the following metals or alloys by UF{sub 6}: nickel, monel, Inconel, gold, platinum, stainless steel, is studied in the temperature range from 300 to 1000 deg. C. The test method, designed to avoid heating the apparatus containing the corrosive fluid to a high temperature, consists in using threadlike samples heated by the Joule effect, the rest of the apparatus being maintained close to room temperature. This technique makes it possible also to determine continuously the penetration of the corrosion by measuring the electrical resistance of the sample with a double Thomson bridge. A series of rapid comparison tests shows that stainless steel, precious metals and Inconel are attacked far too rapidly to be used above 500 deg. C; only monel and especially nickel appear capable of resisting at high temperatures. The detailed examination of the behaviour of nickel shows that the metallic fluoride is volatilized and that this influences the corrosion rate. It shows also the existence of a temperature zone situated between 550 and 700 deg. C in which occurs A strong intergranular corrosion the cause of which appears to be the presence of impurities in the metal. (author) [French] La corrosion par l'UF{sub 6} des metaux ou alliages suivants: lickel, monel, inconel, or, platine, acier inoxydable, est etudiee dans le un domaine de temperature compris entre 300 et 1000 deg. C. La methode d'essai, destinee a eviter le chauffage de l'enceinte contenant le fluide corrosif a temperature elevee, consiste a utiliser des eprouvettes filiformes, echauffees par effet Joule, le reste de l'appareillage etant maintenu a une temperature proche de l'ambiance. Cette technique permet en outre de determiner en continu la penetration de la corrosion, par mesure de la resistance electrique de l'eprouvette, au moyen d'un pont double de Thomson. Une serie d'essais comparatifs, assez sommaires, montre que l'acier inoxydable, les metaux precieux et l'inconel sont attaques beaucoup

  10. Corrosion of metallic materials by uranium hexafluoride at high temperatures (1963); Corrosion de materiaux metalliques par l'hexafluorure d'uranium a haute temperature (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-01

    The corrosion of the following metals or alloys by UF{sub 6}: nickel, monel, Inconel, gold, platinum, stainless steel, is studied in the temperature range from 300 to 1000 deg. C. The test method, designed to avoid heating the apparatus containing the corrosive fluid to a high temperature, consists in using threadlike samples heated by the Joule effect, the rest of the apparatus being maintained close to room temperature. This technique makes it possible also to determine continuously the penetration of the corrosion by measuring the electrical resistance of the sample with a double Thomson bridge. A series of rapid comparison tests shows that stainless steel, precious metals and Inconel are attacked far too rapidly to be used above 500 deg. C; only monel and especially nickel appear capable of resisting at high temperatures. The detailed examination of the behaviour of nickel shows that the metallic fluoride is volatilized and that this influences the corrosion rate. It shows also the existence of a temperature zone situated between 550 and 700 deg. C in which occurs A strong intergranular corrosion the cause of which appears to be the presence of impurities in the metal. (author) [French] La corrosion par l'UF{sub 6} des metaux ou alliages suivants: lickel, monel, inconel, or, platine, acier inoxydable, est etudiee dans le un domaine de temperature compris entre 300 et 1000 deg. C. La methode d'essai, destinee a eviter le chauffage de l'enceinte contenant le fluide corrosif a temperature elevee, consiste a utiliser des eprouvettes filiformes, echauffees par effet Joule, le reste de l'appareillage etant maintenu a une temperature proche de l'ambiance. Cette technique permet en outre de determiner en continu la penetration de la corrosion, par mesure de la resistance electrique de l'eprouvette, au moyen d'un pont double de Thomson. Une serie d'essais comparatifs, assez sommaires, montre que l'acier inoxydable, les metaux

  11. 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.)

  12. Plasma spheroidization and high temperature stability of lanthanum phosphate and its compatibility with molten uranium

    Energy Technology Data Exchange (ETDEWEB)

    Ananthapadmanabhan, P.V. [Laser and Plasma Technology Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)], E-mail: pvananth@barc.gov.in; Sreekumar, K.P.; Thiyagarajan, T.K.; Satpute, R.U. [Laser and Plasma Technology Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Krishnan, K.; Kulkarni, N.K. [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kutty, T.R.G. [Radiometallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2009-01-15

    Lanthanum phosphate has excellent thermal stability and corrosion resistance against many molten metals and other chemically corrosive environments. Lanthanum phosphate (LaPO{sub 4}) was synthesized from lanthanum oxalate by thermal dissociation of the oxalate to the oxide, followed by conversion to hydrated lanthanum phosphate (LaPO{sub 4}.0.5H{sub 2}O). Thermal treatment of LaPO{sub 4}.0.5H{sub 2}O above 773 K resulted in the irreversible transformation of the hydrated phase to the stable monazite phase. Thermal and chemical stability of monazite was studied by plasma spheroidization experiments using a DC thermal plasma reactor set up. Compatibility of monazite with molten uranium was studied by thermal analysis. Results showed that monazite is thermally stable up to its melting point and also is resistant towards attack by molten uranium. Adherent coatings of LaPO{sub 4} could be deposited onto various substrates by atmospheric plasma spray technique.

  13. Plasma spheroidization and high temperature stability of lanthanum phosphate and its compatibility with molten uranium

    International Nuclear Information System (INIS)

    Ananthapadmanabhan, P.V.; Sreekumar, K.P.; Thiyagarajan, T.K.; Satpute, R.U.; Krishnan, K.; Kulkarni, N.K.; Kutty, T.R.G.

    2009-01-01

    Lanthanum phosphate has excellent thermal stability and corrosion resistance against many molten metals and other chemically corrosive environments. Lanthanum phosphate (LaPO 4 ) was synthesized from lanthanum oxalate by thermal dissociation of the oxalate to the oxide, followed by conversion to hydrated lanthanum phosphate (LaPO 4 .0.5H 2 O). Thermal treatment of LaPO 4 .0.5H 2 O above 773 K resulted in the irreversible transformation of the hydrated phase to the stable monazite phase. Thermal and chemical stability of monazite was studied by plasma spheroidization experiments using a DC thermal plasma reactor set up. Compatibility of monazite with molten uranium was studied by thermal analysis. Results showed that monazite is thermally stable up to its melting point and also is resistant towards attack by molten uranium. Adherent coatings of LaPO 4 could be deposited onto various substrates by atmospheric plasma spray technique

  14. 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.

  15. 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...

  16. Influence of temperature, strain rate and thermal aging on the structure/property behavior of uranium 6 wt% Nb

    Energy Technology Data Exchange (ETDEWEB)

    Cady, C.M.; Gray, G.T.; Chen, S.R.; Lopez, M.F. [Los Alamos National Lab., MST-8, MS G-755, NM (United States); Field, R.D.; Korzekwa, D.R. [Los Alamos National Lab., MST-6, MS G-770, NM (United States); Hixson, R.S. [Los Alamos National Lab, DX-9, MS P-952, NM (United States)

    2006-08-15

    A rigorous experimentation and validation program is being undertaken to create constitutive models that elucidate the fundamental mechanisms controlling plasticity in uranium-6 wt% niobium alloys (U-6Nb). These models should accurately predict high-strain-rate large-strain plasticity, damage evolution and failure. The goal is a physically-based constitutive model that captures 1) an understanding of how strain rate, temperature, and aging affects the mechanical response of a material, and 2) an understanding of the operative deformation mechanisms. The stress-strain response of U-6Nb has been studied as a function of temperature, strain-rate, and thermal aging. U-6Nb specimens in a solution-treated and quenched condition and after subsequent aging at 473 K for 2 hours were studied. The constitutive behavior was evaluated over the range of strain rates from quasi-static (0.001 s{sup -1}) to dynamic ({approx} 2000 s{sup -1}) and temperatures ranging from 77 to 773 K. The yield stress of U-6Nb was exhibited pronounced temperature sensitivity. The strain hardening rate is seen to be less sensitive to strain rate and temperature beyond plastic strains of 0.10. The yield strength of the aged material is less significantly affected by temperature and the work hardening rate shows adiabatic heating at lower strains rates (1/s). (authors)

  17. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures

    International Nuclear Information System (INIS)

    Oliveira, Fabio Branco Vaz de

    2008-01-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and

  18. Buckling and reaction rate experiments in plutonium/uranium metal fuelled, graphite moderated lattices at temperatures up to 400 deg. C. Part I: Experimental techniques and results

    Energy Technology Data Exchange (ETDEWEB)

    Carter, D H; Clarke, W G; Gibson, M; Hobday, R; Hunt, C; Marshall, J; Puckett, B J; Symons, C R; Wass, T [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1964-07-15

    This report presents experimental measurements of bucklings, flux fine structure and fission rate distributions in graphite moderated lattices fuelled with plutonium/uranium metal at temperatures up to 400 deg. C in the sub-critical assemblies SCORPIO I and SCORPIO II. The experimental techniques employed are described in some detail. The accuracy of the experimental measurements appears to be adequate for testing methods of calculation being developed for the calculation of reactivity and temperature coefficient of reactivity for power reactors containing plutonium and uranium. (author) 26 refs, 17 tabs, 17 figs

  19. 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

  20. 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

  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. 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

  3. Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

    International Nuclear Information System (INIS)

    Zakova, Jitka; Talamo, Alberto

    2008-01-01

    The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF 2 , LiF, ZrF 4 and Li 2 BeF 4 eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large

  4. Simulation of uranium transport with variable temperature and oxidation potential: The computer program THCC [Thermo-Hydro-Chemical Coupling

    International Nuclear Information System (INIS)

    Carnahan, C.L.

    1986-12-01

    A simulator of reactive chemical transport has been constructed with the capabilities of treating variable temperatures and variable oxidation potentials within a single simulation. Homogeneous and heterogeneous chemical reactions are simulated at temperature-dependent equilibrium, and changes of oxidation states of multivalent elements can be simulated during transport. Chemical mass action relations for formation of complexes in the fluid phase are included explicitly within the partial differential equations of transport, and a special algorithm greatly simplifies treatment of reversible precipitation of solid phases. This approach allows direct solution of the complete set of governing equations for concentrations of all aqueous species and solids affected simultaneously by chemical and physical processes. Results of example simulations of transport, along a temperature gradient, of uranium solution species under conditions of varying pH and oxidation potential and with reversible precipitation of uraninite and coffinite are presented. The examples illustrate how inclusion of variable temperature and oxidation potential in numerical simulators can enhance understanding of the chemical mechanisms affecting migration of multivalent waste elements

  5. Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Zakova, Jitka [Department of Nuclear and Reactor Physics, Royal Institute of Technology, KTH, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)], E-mail: jitka.zakova@neutron.kth.se; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, ANL, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

    2008-05-15

    The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF{sub 2}, LiF, ZrF{sub 4} and Li{sub 2}BeF{sub 4} eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.

  6. 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.

  7. Compositional changes at the interface between thorium-doped uranium dioxide and zirconium due to high-temperature annealing

    Science.gov (United States)

    Youn, Young-Sang; Lee, Jeongmook; Kim, Jandee; Kim, Jong-Yun

    2018-06-01

    Compositional changes at the interface between thorium-doped uranium dioxide (U0.97Th0.03O2) and Zr before and after annealing at 1700 °C for 18 h were studied by X-ray photoelectron spectroscopy, X-ray diffraction, and Raman spectroscopy. At room temperature, the U0.97Th0.03O2 pellet consisted of hyperstoichiometric UO2+x with UO2 and ThO2, and the Zr sample contained Zr with ZrO2. After annealing, the former contained stoichiometric UO2 with ThO2 and the latter consisted of ZrO2 along with ZrO2·2H2O.

  8. Deformation of depleted uranium - 0.78 Ti under shock compression to 11.0 GPa at room temperature

    International Nuclear Information System (INIS)

    Dandekar, D.P.; Martin, A.G.; Kelley, J.V.

    1980-01-01

    The present work on depleted uranium alloyed with 0.78% titanium by weight (i.e., U-0.8 Ti) describes the nature of deformation it undergoes when subjected to shock compression at room temperature. The principal results emerging out of the present work are: (1) The stress limits of elastic deformation are dependent on the thickness of U-0.8Ti. The stress limit decreases from over 3.0 GPa at the impact surface to 1.2 GPa at a depth of 9 mm in U-0.8 Ti; (2) The lower limit of the stress agrees with the static yield stress in U-0.8 Ti; (3) Above the elastic stress limit, the deformation of U-0.8 Ti proceeds in a manner of the ideal plastic solid; and (4) The pressure derivative of Lame's parameter of U-0.8 Ti is estimated to be 3.8

  9. Oxide film assisted dopant diffusion in silicon carbide

    Energy Technology Data Exchange (ETDEWEB)

    Tin, Chin-Che, E-mail: cctin@physics.auburn.ed [Department of Physics, Auburn University, Alabama 36849 (United States); Mendis, Suwan [Department of Physics, Auburn University, Alabama 36849 (United States); Chew, Kerlit [Department of Electrical and Electronic Engineering, Faculty of Engineering and Science, Universiti Tunku Abdul Rahman, Kuala Lumpur (Malaysia); Atabaev, Ilkham; Saliev, Tojiddin; Bakhranov, Erkin [Physical Technical Institute, Uzbek Academy of Sciences, 700084 Tashkent (Uzbekistan); Atabaev, Bakhtiyar [Institute of Electronics, Uzbek Academy of Sciences, 700125 Tashkent (Uzbekistan); Adedeji, Victor [Department of Chemistry, Geology and Physics, Elizabeth City State University, North Carolina 27909 (United States); Rusli [School of Electrical and Electronic Engineering, Nanyang Technological University (Singapore)

    2010-10-01

    A process is described to enhance the diffusion rate of impurities in silicon carbide so that doping by thermal diffusion can be done at lower temperatures. This process involves depositing a thin film consisting of an oxide of the impurity followed by annealing in an oxidizing ambient. The process uses the lower formation energy of silicon dioxide relative to that of the impurity-oxide to create vacancies in silicon carbide and to promote dissociation of the impurity-oxide. The impurity atoms then diffuse from the thin film into the near-surface region of silicon carbide.

  10. Oxide film assisted dopant diffusion in silicon carbide

    International Nuclear Information System (INIS)

    Tin, Chin-Che; Mendis, Suwan; Chew, Kerlit; Atabaev, Ilkham; Saliev, Tojiddin; Bakhranov, Erkin; Atabaev, Bakhtiyar; Adedeji, Victor; Rusli

    2010-01-01

    A process is described to enhance the diffusion rate of impurities in silicon carbide so that doping by thermal diffusion can be done at lower temperatures. This process involves depositing a thin film consisting of an oxide of the impurity followed by annealing in an oxidizing ambient. The process uses the lower formation energy of silicon dioxide relative to that of the impurity-oxide to create vacancies in silicon carbide and to promote dissociation of the impurity-oxide. The impurity atoms then diffuse from the thin film into the near-surface region of silicon carbide.

  11. Study of the transformation of uranium-niobium alloys with low niobium concentrations, tempered from the gamma and beta + gamma 1 regions and then annealed at different temperatures. Comparison with uranium-molybdenum alloys (1963); Etude des transformations des alliages uranium-niobium a faible teneur en niobium trempes depuis les domaines gamma et beta + gamma 1 puis revenus a differentes temperatures. Comparaison avec les alliages uranium-molybdene (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-09-15

    The author shows that uranium-niobium alloys, like uranium-molybdenum alloys, tempered from the gamma region, give a martensitic phase with a structure deriving from that of alpha uranium by a slight contraction parallel to the axis [001], The critical cooling rate allowing the formation of this martensite is 80 deg. C/s at 750 deg. C. Retention of the beta phase of uranium-niobium alloys is particularly difficult, the critical retention rate being 700 deg. C/s at 668 deg. C for an alloy containing 2.5 at. per cent of Nb. This beta phase is completely converted to the alpha phase at room temperature in about 6 hours. The TTT curves of this beta alloy are effectively reduced to the lower branch of the lower 'C'. The beta phase conversion law is expressed as: 1-x = exp. (kt){sup n} x being the degree of progression of the conversion, t the time, n an exponent no-varying with temperature and having approximately the value 2 for the alloy considered, k an increasing function of temperature. The activation energy of conversion is of the order of 14,600 cal/mole. Niobium is much less active than molybdenum as a stabiliser of beta uranium. (author) [French] Dans ce travail l'auteur montre que les alliages uranium-niobium, comme d'ailleurs les alliages uranium-molybdene, trempes depuis le domaine gamma, donnent une phase martensitique dont la structure derive de celle de l'uranium alpha par une legere contraction parallele de l'axe [001]. La vitesse critique de refroidissement permettant la formation de cette martensite est de 80 deg. C/s a 750 deg. C. La retention de la phase beta des alliages uranium-niobium est particulierement delicate car la vitesse critique de retention est de 700 deg. C/s a 668 deg. C pour l'alliage a 2,5 at. pour cent de Nb. Cette phase beta se transforme completement en phase alpha a la temperature ordinaire en 6 heures environ. Les courbes TTT de cet alliage de structure beta se reduisent pratiquement a la branche inferieure du 'C' inferieur. La

  12. Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium-thorium fuels

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2007-01-01

    We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in 235 U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides 240 Pu, 238 U and 232 Th and it requires the esteem of the Dancoff-Ginsburg factor for a pebble bed or prismatic core. The Dancoff-Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057-5692, 6.674-14485, 21.78-3472 eV for 240 Pu, 238 U and 232 Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 μm and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core

  13. Structure and thermal expansion of NbC complex carbides

    International Nuclear Information System (INIS)

    Khatsinskaya, I.M.; Chaporova, I.N.; Cheburaeva, R.F.; Samojlov, A.I.; Logunov, A.V.; Ignatova, I.A.; Dodonova, L.P.

    1983-01-01

    Alloying dependences of the crystal lattice parameters at indoor temperature and coefficient of thermal linear exspansion within a 373-1273 K range are determined for complex NbC-base carbides by the method of mathematical expemental design. It is shown that temperature changes in the linear expansion coefficient of certain complex carbides as distinct from NbC have an anomaly (minimum) within 773-973 K caused by occurring reversible phase transformations. An increase in the coefficient of thermal linear expansion and a decrease in hardness of NbC-base tungsten-, molybdenum-, vanadium- and hafnium-alloyed carbides show a weakening of a total chemical bond in the complex carbides during alloying

  14. Study of the low temperature oxidation of uranium powders and its application to the sintering of uranium oxide powders; Etude de l'oxydation des poudres dtranium a basse temperature et son application au frittage de poudres d'uranium oxyde

    Energy Technology Data Exchange (ETDEWEB)

    Conte-Albert, M. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-06-01

    The uranium oxygen reaction has been studied with a view to obtaining U-UO{sub 2} samples containing about 20 per cent by weight of UO{sub 2} starting from spherical grain uranium powder (36 {mu} < {phi} < 50 {mu}). The techniques used are micrography, thermogravimetry, sintering under pressure, radio-crystallography. At 170 deg. C in air or argon + oxygen mixtures, the uranium oxide formed is always UO{sub 2} and it is uniformly distributed around the initial uranium spheres. These mixed powders can easily be sintered under pressure in the {gamma}-phase. The density of the samples obtained is 85 to 90 per cent of the theoretical density. The influence of UO{sub 2} on the properties of uranium has been shown by the use of dilatometry and thermal cycling in the {alpha} phase. The temperatures at which the phase changes {alpha} {r_reversible} {beta} and {beta} {r_reversible} {gamma} occur are lowered, the remnant expansion is decreased. High density samples resist well to thermal cycling; the characteristic defects of uranium: high distortion, wrinkled surface, have almost disappeared. Heat treatments in a secondary vacuum at 1050 deg. C cause crystallization of UO{sub 2} in a geometrical form and the appearance of a phase of the F.C.C. crystalline type having the composition U{sub W}C{sub X}O{sub Y}N{sub Z}. This phase causes a new decrease in the {alpha} {r_reversible} {beta}, {beta} {r_reversible} {gamma} transformation temperatures for the uranium. After ten dilatometric cycles the remanent expansion of the sample is about 0.5 per cent. The resistance to thermal cycling of a low density sample which has been heat-treated is similar to that of a high density sample which has not undergone a heat treatment. (author) [French] La reaction uranium-oxygene a ete etudiee pour permettre l'obtention d'echantillons U-UO{sub 2} a 20 pour cent en poids environ d'UO{sub 2}, a partir de billes d'uranium pulverulent (36 {mu} < {phi} < 50 {mu}). Les

  15. Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon [KAERI, Daejeon (Korea, Republic of)

    2016-09-15

    The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr{sub 2}O{sub 3}, and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged.

  16. Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

    International Nuclear Information System (INIS)

    Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon

    2016-01-01

    The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr 2 O 3 , and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged

  17. The effect of strain rate and temperature on the tensile behaviour of uranium - 2sup(w)/o molybdenum

    International Nuclear Information System (INIS)

    Harding, J.; Boyd, G.A.C.

    1983-01-01

    This report describes the uniaxial tensile behaviour of uranium 2 w/o molybdenum alloy over a wide range of temperature and strain rate. Specimen blanks taken from co-reduced and extruded U2 w/o Mo rods were given one of two heat treatments. Longitudinal tensile test pieces, taken from these blanks at near surface locations were tested in the temperature range -150 deg C to +100 deg C at strain rates from quasistatic (10 -4 s -1 ) to 10 3 s -1 . To achieve this range of testing rates three machines were required: an Instron screw driven machine for rates up to 0.1 s -1 , a second specially constructed hydraulic machine for the range 0.1 s -1 to 50 s -1 and a drop weight machine for the highest strain rates. The ways in which the mechanical properties - elongation to fracture, flow stresses and ultimate tensile stress - vary with both temperature and strain rate are presented and discussed for material in both heat treatment conditions. (author)

  18. The effect of strain rate and temperature on the tensile behaviour of uranium 2 w/o molybdenum

    International Nuclear Information System (INIS)

    Harding, J.; Boyd, G.A.C.

    1983-01-01

    This report describes the uniaxial tensile behaviour of uranium 2 w/o molybdenum alloy over a wide range of temperature and strain rate. Specimen blanks taken from co-reduced and extruded U 2 w/o Mo rods were given one of two heat treatments. Longitudinal tensile test pieces, taken from these blanks at near surface locations were tested in the temperature range -150 deg C to +100 deg C at strain rates from quasistatic (10 -4 s -1 ) to 10 3 s -1 . To achieve this range of testing rates three machines were required: an Instron screw driven machine for rates up to 0.1 s -1 , a second specially constructed hydraulic machine for the range 0.1 s -1 to 50 s -1 and a drop weight machine for the highest strain rates. The ways in which the mechanical properties - elongation to fracture, flow stresses and ultimate tensile stress - vary with both temperature and strain rate are presented and discussed for material in both heat treatment conditions. (author)

  19. Effects of temperature, concentration, and uranium chloride mixture on zirconium electrochemical studies in LiCl−KCl eutectic salt

    Energy Technology Data Exchange (ETDEWEB)

    Hoover, Robert O. [Department of Chemical and Materials Engineering and Nuclear Engineering Program, University of Idaho, Center for Advanced Energy Studies, 995 University Blvd, Idaho Falls, ID 8340 (United States); Yoon, Dalsung [Department of Mechanical & Nuclear Engineering, Virginia Commonwealth University, 401 West Main St., Richmond, VA 23284 (United States); Phongikaroon, Supathorn, E-mail: sphongikaroon@vcu.edu [Department of Mechanical & Nuclear Engineering, Virginia Commonwealth University, 401 West Main St., Richmond, VA 23284 (United States)

    2016-08-01

    Experimental studies were performed to provide measurement and analysis of zirconium (Zr) electrochemistry in LiCl−KCl eutectic salt at different temperatures and concentrations using cyclic voltammetry (CV). An additional experimental set with uranium chloride added into the system forming UCl{sub 3}−ZrCl{sub 4}−LiCl−KCl was performed to explore the general behavior of these two species together. Results of CV experiments with ZrCl{sub 4} show complicated cathodic and anodic peaks, which were identified along with the Zr reactions. The CV results reveal that diffusion coefficients (D) of ZrCl{sub 4} and ZrCl{sub 2} as the function of temperature can be expressed as D{sub Zr(IV)} = 0.00046exp(−3716/T) and D{sub Zr(II)} = 0.027exp(−5617/T), respectively. The standard rate constants and apparent standard potentials of ZrCl{sub 4} at different temperatures were calculated. Furthermore, the results from the mixture of UCl{sub 3} and ZrCl{sub 4} indicate that high concentrations of UCl{sub 3} hide the features of the smaller concentration of ZrCl{sub 4} while Zr peaks become prominent as the concentration of ZrCl{sub 4} increases.

  20. Influence of the temperature in the uranium (Vi) sorption in zirconium diphosphate; Influencia de la temperatura en la sorcion de uranio (VI) en difosfato de circonio

    Energy Technology Data Exchange (ETDEWEB)

    Garcia G, N.; Solis, D. [Universidad Autonoma del Estado de Mexico, Facultad de Quimica, Paseo Colon y Paseo Tollocan, 50120 Toluca, Estado de Mexico (Mexico); Ordonez R, E., E-mail: nidgg@yahoo.com.mx [ININ, Departamento de Quimica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In the present work was evaluated the uranium (Vi) sorption at 10, 20, 30, 40 and 60 C on the zirconium diphosphate (ZrP{sub 2}O{sub 7}). They were carried out kinetic and isotherms using the method by lots, these will allow to fix the sorption time (kinetic) and to explain the behavior of this sorption in different ph conditions and temperature (isotherm). The quantity of retained uranium in the surface was quantified by means of the fluorescence technique. (Author)

  1. The reaction of sintered aluminium products with uranium dioxide and monocarbide

    DEFF Research Database (Denmark)

    Lauritzen, T.; Knudsen, Per

    1965-01-01

    The compatibility of SAP 930 with uranium dioxide and uranium monocarbide was investigated in the temperature range 450–600° C. The results indicate that a severe reaction occurs between SAP 930 and UO2 within 8000 hours at 600° C, a slight reaction at 600° C for 1000 hours and after 11 900 hours...... at 525° C, and no reaction in 14 300 hours at 450° C. Of the three grades of UC tested (hot pressed, arc cast, cold pressed and sintered) the slightly substoichiometric, hot-pressed UC is judged to be least compatible with SAP 930, reaction occurring after 7300 hours at 450° C. No reaction was observed...... between SAP 930 and the other carbides at this temperature. All SAP−UC combinations are incompatible at 600° C for as little as 100 hours of heat treatment. Tests designed to study the effect of a diffusion barrier on the SAP−UC reaction have shown that anodized SAP 930 and the three uranium carbides...

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

    Science.gov (United States)

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

    2017-12-01

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

  3. 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

  4. 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)

  5. METHOD OF ROLLING URANIUM

    Science.gov (United States)

    Smith, C.S.

    1959-08-01

    A method is described for rolling uranium metal at relatively low temperatures and under non-oxidizing conditions. The method involves the steps of heating the uranium to 200 deg C in an oil bath, withdrawing the uranium and permitting the oil to drain so that only a thin protective coating remains and rolling the oil coated uranium at a temperature of 200 deg C to give about a 15% reduction in thickness at each pass. The operation may be repeated to accomplish about a 90% reduction without edge cracking, checking or any appreciable increase in brittleness.

  6. 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)

  7. 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.

  8. On change of vanadium carbide state during 20Kh3MVF steel heat treatment

    International Nuclear Information System (INIS)

    Gitgarts, M.I.; Maksimenko, V.N.

    1975-01-01

    The Xray diffraction study of vanadium carbide MC has been made in the steel-20KH3MVF quenched from 970 and 1040 deg and tempered at 660 deg for 210 hrs. It has been found that the constant of the MC crystal lattice regularly varies with the temperature of isothermal hold-up. In the steel tempered after quenching two vanadium carbides of different content could co-exist simultaneously: carbide formed in the quenching process and carbide formed during tempering. The discovered effect of the temperature dependence of the MC content is, evidently, inherent also to other steels containing vanadium

  9. Corrosion behavior of porous chromium carbide in supercritical water

    International Nuclear Information System (INIS)

    Dong Ziqiang; Chen Weixing; Zheng Wenyue; Guzonas, Dave

    2012-01-01

    Highlights: ► Corrosion behavior of porous Cr 3 C 2 in various SCW conditions was investigated. ► Cr 3 C 2 is stable in SCW at temperature below 420–430 °C. ► Cracks and disintegration were observed at elevated testing temperatures. ► Degradation of Cr 3 C 2 is related to the intermediate product CrOOH. - Abstract: The corrosion behavior of highly porous chromium carbide (Cr 3 C 2 ) prepared by a reactive sintering process was characterized at temperatures ranging from 375 °C to 625 °C in a supercritical water environment with a pressure of 25–30 MPa. The test results show that porous chromium carbide is stable in SCW environments at temperatures under 425 °C, above which disintegration occurred. The porous carbide was also tested under hydrothermal conditions of pressures between 12 MPa and 50 MPa at constant temperatures of 400 °C and 415 °C, respectively. The pressure showed little effect on the stability of chromium carbide in the tests at those temperatures. The mechanism of disintegration of chromium carbide in SCW environments is discussed.

  10. Design and installation of high-temperature ultrasonic measuring system and grinder for nuclear fuel containing trans-uranium elements

    International Nuclear Information System (INIS)

    Serizawa, Hiroyuki; Kikuchi, Hironobu; Iwai, Takashi; Arai, Yasuo; Kurosawa, Makoto; Mimura, Hideaki; Abe, Jiro

    2005-07-01

    A high-temperature ultrasonic measuring system had been designed and installed in a glovebox (711-DGB) to study a mechanical property of nuclear fuel containing trans-uranium (TRU) elements. A figuration apparatus for the cylinder-type sample preparation had also been modified and installed in an established glovebox (142-D). The system consists of an ultrasonic probe, a heating furnace, cooling water-circulating system, a cooling air compressor, vacuum system, gas supplying system and control system. An A/D converter board and an pulsar/receiver board for the measurement of wave velocity were installed in a personal computer. The apparatus was modified to install into the glovebox. Some safety functions were supplied to the control system. The shape and size of the sample was revised to minimize the amount of TRU elements for the use of the measurement. The maximum sample temperature is 1500degC. The performance of the installed apparatuses and the glovebox were confirmed through a series of tests. (author)

  11. Equation of state, phase stability, and phase transformations of uranium-6 wt. % niobium under high pressure and temperature

    Science.gov (United States)

    Zhang, Jianzhong; Vogel, Sven; Brown, Donald; Clausen, Bjorn; Hackenberg, Robert

    2018-05-01

    In-situ time-of-flight neutron diffraction experiments were conducted on the uranium-niobium alloy with 6 wt. % Nb (U-6Nb) at pressures up to 4.7 GPa and temperatures up to 1073 K. Upon static compression at room temperature, the monoclinic structure of U-6Nb (α″ U-6Nb) remains stable up to the highest experimental pressure. Based on the pressure-volume measurements at room temperature, the least-squares fit using the finite-strain equation of state (EOS) yields an isothermal bulk modulus of B0 = 127 ± 2 GPa for the α″-phase of U-6Nb. The calculated zero-pressure bulk sound speed from this EOS is 2.706 ± 0.022 km/s, which is in good agreement with the linear extrapolation of the previous Hugoniot data above 12 GPa for α″ U-6Nb, indicating that the dynamic response under those shock-loading conditions is consistent with the stabilization of the initial monoclinic phase of U-6Nb. Upon heating at ambient and high pressures, the metastable α″ U-6Nb exhibits complex transformation paths leading to the diffusional phase decomposition, which are sensitive to applied pressure, stress state, and temperature-time path. These findings provide new insight into the behavior of atypical systems such as U-Nb and suggest that the different U-Nb phases are separated by rather small energies and hence highly sensitive to compositional, thermal, and mechanical perturbations.

  12. Is uranium dioxide a glass at high temperature: the reason for its irradiation resistance?

    International Nuclear Information System (INIS)

    Desgranges, Lionel

    2008-01-01

    Electronic intrinsic carriers are shown to have some influence on UO 2 high temperature properties. The physical nature of these carriers, called polarons, is discussed and it is proposed that they could correspond to quasi-broken bonds, in a similar way to intrinsic electronic defects in SiO 2 . It is shown that this hypothesis provides an explanation, at least qualitative, for UO 2 specific behavior at high temperature and under irradiation. (author)

  13. Electronic Effects on Room-Temperature, Gas-Phase C-H Bond Activations by Cluster Oxides and Metal Carbides: The Methane Challenge.

    Science.gov (United States)

    Schwarz, Helmut; Shaik, Sason; Li, Jilai

    2017-12-06

    This Perspective discusses a story of one molecule (methane), a few metal-oxide cationic clusters (MOCCs), dopants, metal-carbide cations, oriented-electric fields (OEFs), and a dizzying mechanistic landscape of methane activation! One mechanism is hydrogen atom transfer (HAT), which occurs whenever the MOCC possesses a localized oxyl radical (M-O • ). Whenever the radical is delocalized, e.g., in [MgO] n •+ the HAT barrier increases due to the penalty of radical localization. Adding a dopant (Ga 2 O 3 ) to [MgO] 2 •+ localizes the radical and HAT transpires. Whenever the radical is located on the metal centers as in [Al 2 O 2 ] •+ the mechanism crosses over to proton-coupled electron transfer (PCET), wherein the positive Al center acts as a Lewis acid that coordinates the methane molecule, while one of the bridging oxygen atoms abstracts a proton, and the negatively charged CH 3 moiety relocates to the metal fragment. We provide a diagnostic plot of barriers vs reactants' distortion energies, which allows the chemist to distinguish HAT from PCET. Thus, doping of [MgO] 2 •+ by Al 2 O 3 enables HAT and PCET to compete. Similarly, [ZnO] •+ activates methane by PCET generating many products. Adding a CH 3 CN ligand to form [(CH 3 CN)ZnO] •+ leads to a single HAT product. The CH 3 CN dipole acts as an OEF that switches off PCET. [MC] + cations (M = Au, Cu) act by different mechanisms, dictated by the M + -C bond covalence. For example, Cu + , which bonds the carbon atom mostly electrostatically, performs coupling of C to methane to yield ethylene, in a single almost barrier-free step, with an unprecedented atomic choreography catalyzed by the OEF of Cu + .

  14. Study of the temperature influence during the uranium (Vi) sorption on surface of ZrP2O7 in presence of oxalic and salicylic acid

    International Nuclear Information System (INIS)

    Garcia G, N.

    2013-01-01

    This work studies the effect of temperature on the uranium (Vi) sorption onto zirconium diphosphate in the presence of organic acids (oxalic and salicylic acids). Zirconium diphosphate was synthesized by a chemical condensation reaction and characterized using several analytical techniques, in order to check its purity. This point is very important because the presence of any impurities or secondary phases may interfere with the hydration and sorption process. Prior to the sorption experiments, three batches of zirconium diphosphate were pre-equilibrated with NaClO 4 , oxalic acid or salicylic acid solutions. The hydrated solids were washed and dried and then again characterized in order to study the interactions between organic acids and zirconium diphosphate surface. Uranium sorption onto zirconium diphosphate (pre-equilibrated with NaClO 4 , oxalic acid and salicylic acid solutions) was investigated as a function of ph, organic acid and temperature (20, 40 y 60 grades C). Thermodynamic parameters for the sorption reactions (enthalpy change, entropy change and Gibbs free energy change) were determined from temperature dependence of distribution coefficient by using the Vant Hoff equation. Solids characterization after hydration shows that exist an interaction between organic acids and ZrP 2 O 7 . This fact was confirmed with the microcalorimetry study, the reaction heat for hydration of zirconium diphosphate in NaClO 4 solution was exothermic (-269.59 mJ) and for hydration of zirconium diphosphate in oxalic acid solution was endothermic (53.64 mJ). The experimental results showed important differences in the sorption mechanisms for the reaction of Uranium with ZrP 2 O 7 in the presence and absence of organic acids. For the zirconium diphosphate hydrated with oxalic acid, the sorption percentage was 50% from lowest ph values. For the zirconium diphosphate hydrated with salicylic acid, the initial concentration of uranium was 6 x 10 -4 M and a percentage of 10% was

  15. Study of elementary mechanisms of creep in uranium as a function of temperature (150 deg. to 760 deg. C) by activation energy measurements

    International Nuclear Information System (INIS)

    Grenier, P.

    1966-06-01

    Creep tests were carried out on single crystals and polycrystalline specimens of uranium in both the α and β phases over the temperature range 150 - 760 deg. C. The determination of the activation energy for creep and the study of its variation with temperature made it possible to distinguish various temperature ranges in which one or more elementary mechanisms govern deformation. Micrographic observations after creep and the study of the variation of creep-rate with load support the conclusions. The creep behavior of single crystals is identical with that of polycrystalline material below 325 deg. C. From 325 deg. C to one upper limiting temperature whose value depends on the purity and previous history of the metal, the creep deformation of uranium is controlled by cross-slip. From this limiting temperature up to 520 deg. C, the creep of uranium involves two independent mechanisms operating simultaneously, the movement of screw dislocation by cross-slip and the climbing of edge dislocations out of their slip plane. Between 520 deg. C and the α - β transformation temperature creep in polycrystals is governed by the climb of edge dislocations out of their slip planes, by a pile up mechanism in the case of primary creep and by dipole annihilation in the case of secondary creep. In single crystals creep is dependent on the climb of edge dislocations into pre-existent sub-boundaries and their subsequent rearrangement within these boundaries. In the β phase the creep of polycrystals is governed by the diffusional climb of edge dislocations. Between 450 and 630 deg. C small alloy additions of molybdenum modify the creep characteristics of uranium although the deformation mechanisms involved are analogous to those in the pure metal. (author) [fr

  16. Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium-thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology - KTH, Roslagstullsbacken 21, S-10691 Stockholm (Sweden)]. E-mail: alby@anl.gov

    2007-01-15

    We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in {sup 235}U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides {sup 240}Pu, {sup 238}U and {sup 232}Th and it requires the esteem of the Dancoff-Ginsburg factor for a pebble bed or prismatic core. The Dancoff-Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057-5692, 6.674-14485, 21.78-3472 eV for {sup 240}Pu, {sup 238}U and {sup 232}Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 {mu}m and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core.

  17. A review of the oxidation of uranium dioxide at temperatures below 400oC

    International Nuclear Information System (INIS)

    McEachern, R.J.; Taylor, P.

    1997-01-01

    A critical review of the extensive literature on the air oxidation Of U0 2 at temperatures below 400 o C is presented. The key parameters that affect the rate Of U0 2 oxidation are examined systematically, and their importance to the reaction rate is evaluated. The formation of U 30 7/U 4 0 9 on unirradiated U0 2 powders follows the discrete-layer mechanism and displays diffusion-controlled kinetics. In contrast, U 3 0 8 formation on unirradiated U0 2 displays sigmoidal 'nucleation-and-growth' kinetics. Low-temperature oxidation of used fuel tends to proceed by rapid grain-boundary oxidation followed by simultaneous intragranular oxidation throughout the sample. The activation energy for the formation Of U 3 0 7 /U 4 0 9 is 96 kJ mol -1 for U0 2 powders, 99 kJ mol -1 for sintered pellets and 106 kJ mol -1 for used fuel. The activation energy for the formation Of U 3 0 8 is temperature dependent. The best estimate of the activation energy below ∼325 o C is 154 kJ mol -1 , but all the kinetic data incorporate substantial approximations so that further study is required to properly predict the behaviour of used fuel under low-temperature ( o C) dry-air storage conditions, based on high-temperature (200 to 350 o C) laboratory data. (author). 204 refs., 5 tabs., 4 figs

  18. Corrosion of metallic materials by uranium hexafluoride at high temperatures (1963)

    International Nuclear Information System (INIS)

    Langlois, G.

    1963-01-01

    The corrosion of the following metals or alloys by UF 6 : nickel, monel, Inconel, gold, platinum, stainless steel, is studied in the temperature range from 300 to 1000 deg. C. The test method, designed to avoid heating the apparatus containing the corrosive fluid to a high temperature, consists in using threadlike samples heated by the Joule effect, the rest of the apparatus being maintained close to room temperature. This technique makes it possible also to determine continuously the penetration of the corrosion by measuring the electrical resistance of the sample with a double Thomson bridge. A series of rapid comparison tests shows that stainless steel, precious metals and Inconel are attacked far too rapidly to be used above 500 deg. C; only monel and especially nickel appear capable of resisting at high temperatures. The detailed examination of the behaviour of nickel shows that the metallic fluoride is volatilized and that this influences the corrosion rate. It shows also the existence of a temperature zone situated between 550 and 700 deg. C in which occurs A strong intergranular corrosion the cause of which appears to be the presence of impurities in the metal. (author) [fr

  19. Extreme-Environment Silicon-Carbide (SiC) Wireless Sensor Suite

    Science.gov (United States)

    Yang, Jie

    2015-01-01

    Phase II objectives: Develop an integrated silicon-carbide wireless sensor suite capable of in situ measurements of critical characteristics of NTP engine; Compose silicon-carbide wireless sensor suite of: Extreme-environment sensors center, Dedicated high-temperature (450 deg C) silicon-carbide electronics that provide power and signal conditioning capabilities as well as radio frequency modulation and wireless data transmission capabilities center, An onboard energy harvesting system as a power source.

  20. Temperature dependence of diffusion coefficients of trivalent uranium ions in chloride and chloride-fluoride melts

    International Nuclear Information System (INIS)

    Komarov, V.E.; Borodina, N.P.

    1981-01-01

    Diffusion coefficients of U 3+ ions are measured by chronopotentiometric method in chloride 3LiCl-2KCl and in mixed chloride fluoride 3LiCl(LiF)-2KCl melts in the temperature range 633-1235 K. It is shown It is shown that experimental values of diffusion-coefficients are approximated in a direct line in lg D-1/T coordinate in chloride melt in the whole temperature range and in chloride-fluoride melt in the range of 644-1040 K. Experimental values of diffusion coefficients diviate from Arrhenius equation in the direction of large values in chloride-fluoride melt at further increase of temperature up to 1235 K. Possible causes of such a diviation are considered [ru

  1. Thermodynamic Calculation of Carbide Precipitate in Niobium Microalloyed Steels

    Institute of Scientific and Technical Information of China (English)

    XU Yun-bo; YU Yong-mei; LIU Xiang-hua; WANG Guo-dong

    2006-01-01

    On the basis of regular solution sublattice model, thermodynamic equilibrium of austenite/carbide in Fe-Nb-C ternary system was investigated. The equilibrium volume fraction, chemical driving force of carbide precipitates and molar fraction of niobium and carbon in solution at different temperatures were evaluated respectively. The volume fraction of precipitates increases, molar fraction of niobium dissolved in austenite decreases and molar fraction of carbon increases with decreasing the niobium content. The driving force increases with the decrease of temperature, and then comes to be stable at relatively low temperatures. The predicted ratio of carbon in precipitates is in good agreement with the measured one.

  2. 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

  3. Enhancing the activation of silicon carbide tracer particles for PEPT applications using gas-phase deposition of alumina at room temperature and atmospheric pressure

    Energy Technology Data Exchange (ETDEWEB)

    Valdesueiro, D. [Delft University of Technology, Department of Chemical Engineering, 2628 BL Delft (Netherlands); Garcia-Triñanes, P., E-mail: p.garcia@surrey.ac.uk [Department of Chemical and Process Engineering, Faculty of Engineering and Physical Sciences, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom); Meesters, G.M.H.; Kreutzer, M.T. [Delft University of Technology, Department of Chemical Engineering, 2628 BL Delft (Netherlands); Gargiuli, J.; Leadbeater, T.W.; Parker, D.J. [Positron Imaging Centre, School of Physics and Astronomy, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); Seville, J.P.K. [Department of Chemical and Process Engineering, Faculty of Engineering and Physical Sciences, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom); Ommen, J.R. van, E-mail: j.r.vanommen@tudelft.nl [Delft University of Technology, Department of Chemical Engineering, 2628 BL Delft (Netherlands)

    2016-01-21

    We have enhanced the radio-activation efficiency of SiC (silicon carbide) particles, which by nature have a poor affinity towards {sup 18}F ions, to be employed as tracers in studies using PEPT (Positron Emission Particle Tracking). The resulting SiC–Al{sub 2}O{sub 3} core–shell structure shows a good labelling efficiency, comparable to γ-Al{sub 2}O{sub 3} tracer particles, which are commonly used in PEPT. The coating of the SiC particles was carried at 27±3 °C and 1 bar in a fluidized bed reactor, using trimethylaluminium and water as precursors, by a gas phase technique similar to atomic layer deposition. The thickness of the alumina films, which ranged from 5 to 500 nm, was measured by elemental analysis and confirmed with FIB-TEM (focused ion beam – transmission electron microscope), obtaining consistent results from both techniques. By depositing such a thin film of alumina, properties that influence the hydrodynamic behaviour of the SiC particles, such as size, shape and density, are hardly altered, ensuring that the tracer particle shows the same flow behaviour as the other particles. The paper describes a general method to improve the activation efficiency of materials, which can be applied for the production of tracer particles for many other applications too. - Highlights: • We deposited Al{sub 2}O{sub 3} films on SiC particles at ambient conditions in a fluidized bed. • The affinity of {sup 18}F ions towards Al{sub 2}O{sub 3}–SiC particle was improved compared to SiC. • We used the Al{sub 2}O{sub 3}–SiC activated particle as tracer in a PEPT experiment. • Tracer particles have suitable activity for accurate tracking. • The Al{sub 2}O{sub 3} film is thin enough not to alter the particle size, shape and density.

  4. 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)

  5. 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)

  6. Metallographic detection of carbides in the steel X 41 CrMoV 51 after different austenizing processes

    International Nuclear Information System (INIS)

    Fleer, R.; Rickel, J.; Draugelates, U.

    1979-01-01

    The etchant most suitable for clearly revealing the carbide particles in the developed hardened structure was determined by comparative structural investigations with several etchants in order to be able to undertake the metallographic detection of finely distributed carbides in the structure of the high alloy ultra-high strength steel X 41 CrMoV 51. The characteristic distribution and number of carbides could be revealed as well as the ferrite pearlite matrix. The picric-hydrochloric acid solution which, on a comparative basis, was the most effective, revealed the dependence of the carbide dissolution and structural formation on the temperature. The carbide components of the structure dissolved to an increasing extent at temperatures above 1100 0 C. All carbides up to the large volume mixed carbides appeared to dissolve in the segregation zone after annealing for one hour at 1200 0 C. Considerable grain growth also occurred. (orig./RW) [de

  7. Crystallographic and oxidation kinetic study of uranium dioxide by high temperature X-ray diffractometry

    International Nuclear Information System (INIS)

    Teixeira, S.R.

    1981-01-01

    The structural behavior of UO 2 sintered plates was studied as a function of temperature by X-ray diffractometry. All the experiments were carried out under an inert atmosphere with low oxygen content (approximated 140 ppm). The thermal expansion coefficient of UO 2 05 was found to be 10,5 x 10 - 6 0 C - 1 for temperatures above 165 0 C. Structural transformations during oxidation were observed at 170,235 and 275 0 C. The isothermal oxidation of UO 2 to U 3 O 7 follows a parabolic form and the diffusion of oxygen through the product layer U 4 O 9 is the mechanism controlling the oxidation rate. The phases observed were UO 2 (cubic) - U 4 O 9 (cubic) - U 3 O 7 (tetragonal). Activation energies of oxidation were found for different crystallographic planes (hkl). From this one can conclude that there is a preferential occupation of interstitial oxygen within the UO 2 structure. (Author) [pt

  8. Mechanisms of the plastic deformation of uranium alloys at low temperature

    International Nuclear Information System (INIS)

    Le Poac, P.; Nomine, A.M.; Miannay, D.

    1976-01-01

    The mechanical characteristics of the bcc binary alloys U-6Mo, U-8Mo, U-10Mo, U-12Mo and bcc ternary alloys U-8Mo-1Ti, U-10Mo-1Ti, U-10Mo-1Zr, stressed in compression, were determined between -196 deg C and + 450 deg C. The plastic flow shear stress in non-dependent on temperature above 300 deg C. At lower temperature shear stress is highly activated, except for the alloy U-6Mo and U-12Mo. Athermal shear stress above 300 deg C is due to the hardening of the solid solution described by Mott and Nabarro. In the thermal range, the recombination of the dissociated dislocations controls the plastic deformation [fr

  9. Mechanical properties of aluminium-uranium alloy and aluminium commercially pure at several temperatures

    International Nuclear Information System (INIS)

    Quadros, N.F. de.

    1976-01-01

    The mechanical properties of Ai-U (18,4 wt %) alloy with and without heat treatment were determined, and they were compared with the mechanical properties of aluminum alloy of commercial purity, AI-1100, at tempiratures of 25, 500, 550 and 600 0 C, the changes of both the yield point stress and the ultimate tensile strength as a function of temperature may be described through two emperical relationships. A fractography study was also made [pt

  10. Assessment of the thorium and uranium fuel cycle in the fast breeder and the high temperature reactor

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    1977-01-01

    This report assesses the fissile fuel economy of the uranium and thorium cycle in the advanced reactors currently under development, the fast breeder reactor (FBR) and the high temperature reactor (HTR). It is shown by means of detailed burnup calculations that replacing UO 2 with ThO 2 or Th-metal as the radial blanket breeding material will not have any significant imapct on the breeding and burnup properties of the FBR. A global, analytical investigation is performed to study the fissile fuel economy of the many fissile fuel cycles possible in the HTR. Here it is demonstrated that the optimum conversion ratio of CR 3 O 8 ) demands are evaluated for a country such as the FRG under the assumptions of different future reactor strategy scenarios. Here it is demonstrated that the employement of both HTRs and FBRs can lead to a practically resource independent energy supply system within the next 40 to 60 years. However only through the large scale employement of the fast breeder can the future nuclear resource requirements be assured. (orig.) [de

  11. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  12. 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)

  13. Spectroscopic evidence for 5f bands at room temperature in uranium-based heavy fermions

    International Nuclear Information System (INIS)

    Arko, A.J.; Koelling, D.D.; Dunlap, B.D.; Capasso, C.; del Giudice, M.

    1988-01-01

    We present data on the alloy system UPd/sub 3-x/Pt/sub x/ and show that in the double hexagonal phase (x 2.4) as well, except that the low-binding energy feature is locked in at E/sub F/ and shows evidence of energy dispersion at room temperature/endash/consistent with well-defined bands. Conversely, we show that even in well-behaved narrow band systems (USn 3 there is evidence for satellite formation. 44 refs., 8 figs

  14. Proposal of new 235U nuclear data to improve keff biases on 235U enrichment and temperature for low enriched uranium fueled lattices moderated by light water

    International Nuclear Information System (INIS)

    Wu, Haicheng; Okumura, Keisuke; Shibata, Keiichi

    2005-06-01

    The under prediction of k eff depending on 235 U enrichment in low enriched uranium fueled systems, which had been a long-standing puzzle especially for slightly enriched ones, was studied in this report. Benchmark testing was carried out with several evaluated nuclear data files, including the new uranium evaluations from preliminary ENDF/B-VII and CENDL-3.1. Another problem reviewed here was k eff underestimation vs. temperature increase, which was observed in the sightly enriched system with recent JENDL and ENDF/B uranium evaluations. Through the substitute analysis of nuclear data of 235 U and 238 U, we propose a new evaluation of 235 U data to solve both of the problems. The new evaluation was tested for various uranium fueled systems including low or highly enriched metal and solution benchmarks in the ICSBEP handbook. As a result, it was found that the combination of the new evaluation of 235 U and the 238 U data from the preliminary ENDF/B-VII gives quite good results for most of benchmark problems. (author)

  15. Tantalum and niobium carbides obtention by carbothermic reduction of columbotantalite ores

    International Nuclear Information System (INIS)

    Gordo, E.; Garcia-Carcedo, F.; Torralba, J.M.

    1998-01-01

    Tantalum and niobium carbides are characterized by its high hardness and chemical corrosion resistance. Both carbides, but mainly TaC, are used in hard metals (sintered carbides), together with their carbides, to manufacture cutting tools and dies in special machining applications involving mechanical shock at high temperature. Its use as reinforcement of wear resistant materials through powder metallurgy techniques are being investigated. However, the use of TaC is usually limited because of its high cost. Therefore tantalum carbide with niobium content, which is cheaper, is used. In this work the obtention of complex tantalum and niobium carbides from a Spanish columbotantalite ore is studied through relatively cheap and simple process as it is carbothermic reduction. Concentration of the ore, its reduction and the characterization of products are described. (Author) 11 refs

  16. Highly thermal conductive carbon fiber/boron carbide composite material

    International Nuclear Information System (INIS)

    Chiba, Akio; Suzuki, Yasutaka; Goto, Sumitaka; Saito, Yukio; Jinbo, Ryutaro; Ogiwara, Norio; Saido, Masahiro.

    1996-01-01

    In a composite member for use in walls of a thermonuclear reactor, if carbon fibers and boron carbide are mixed, since they are brought into contact with each other directly, boron is reacted with the carbon fibers to form boron carbide to lower thermal conductivity of the carbon fibers. Then, in the present invention, graphite or amorphous carbon is filled between the carbon fibers to provide a fiber bundle of not less than 500 carbon fibers. Further, the surface of the fiber bundle is coated with graphite or amorphous carbon to suppress diffusion or solid solubilization of boron to carbon fibers or reaction of them. Then, lowering of thermal conductivity of the carbon fibers is prevented, as well as the mixing amount of the carbon fiber bundles with boron carbide, a sintering temperature and orientation of carbon fiber bundles are optimized to provide a highly thermal conductive carbon fiber/boron carbide composite material. In addition, carbide or boride type short fibers, spherical graphite, and amorphous carbon are mixed in the boron carbide to prevent development of cracks. Diffusion or solid solubilization of boron to carbon fibers is reduced or reaction of them if the carbon fibers are bundled. (N.H.)

  17. Fuel element for high-temperature nuclear power reactors

    International Nuclear Information System (INIS)

    Schloesser, J.

    1974-01-01

    The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

  18. Uranium extraction history using pressure leaching

    International Nuclear Information System (INIS)

    Fraser, K.S.; Thomas, K.G.

    2010-01-01

    Over the past 60 years of uranium process development only a few commercial uranium plants have adopted a pressure leaching process in their flowsheet. The selection of acid versus alkaline pressure leaching is related to the uranium and gangue mineralogy. Tetravalent (U"+"4) uranium has to be oxidized to hexavalent (U"+"6) uranium to be soluble. Refractory tetravalent uranium requires higher temperature and pressure, as practised in pressure leaching, for conversation to soluble hexavalent uranium. This paper chronicles the history of these uranium pressure leaching facilities over the past 60 years, with specific details of each design and operation. (author)

  19. 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.

  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. 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

  2. Features of order-disorder phase transformation in nonstoichiometric transition metals carbides

    International Nuclear Information System (INIS)

    Emel'yanov, A.N.

    1996-01-01

    Measurements of temperature and electric conductivity of nonstoichiometric transition metals carbides TiC χ and NbC χ in the area of order-disorder phase transformation are carried out. There are certain peculiarities on the temperature and electric conductivity curves of the carbides, connected with the carbon sublattice disordering. On the basis of the anomalies observed on the curves of the temperature conductivity of nonstoichiometric carbides of transition metals above the temperature of the order-disorder transition the existence of the second structural transition is supposed

  3. Swelling of uranium dioxide and deformation behavior of the fuel element at high temperature irradiation

    International Nuclear Information System (INIS)

    Gontar, A.S.; Gutnik, V.S.; Nelidov, M.V.; Nikolaev, Yu.V.

    1993-01-01

    As post-reactor investigations showed, significant difference of swelling rates is connected with the fact that swelling of UO 2 with the equiaxial structure is mainly the result of fission gas bubbles accumulation along grain boundaries, and, in the case of the column structure, with formation of fine bubbles inside grains. The data given testify to usefulness of such investigations to predict TFE lifetime. As proven in this study one can see rates of radial deformation of fuel element cladding of a multi-cell TFE as a result of UO 2 swelling. They were calculated using the code SDS. Typical sizes were taken for calculation: cladding diameter--20 mm, cladding temperature at the central section of the fuel element--1,900 K, energy generation rate--145 W/cm 3 . These parameters provide output electric power of the TFE 600 W at the active zone length--400 mm

  4. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    International Nuclear Information System (INIS)

    Zhu, G.; Zou, Y.; Xu, H.

    2016-01-01

    Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PBFHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF_2) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tri-structural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern) and two kinds of reflector materials (SiC and graphite). This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong "9Be(n,2n) reaction and low neutron absorption of "6Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows a good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel. (A.C)

  5. TEM investigation of aluminium containing precipitates in high aluminium doped silicon carbide

    International Nuclear Information System (INIS)

    Wong-Leung, J.; FitzGerald, J.D.

    2002-01-01

    Full text: Silicon carbide is a promising semiconductor material for applications in high temperature and high power devices. The successful growth of good quality epilayers in this material has enhanced its potential for device applications. As a novel semiconductor material, there is a need for studying its basic physical properties and the role of dopants in this material. In this study, silicon carbide epilayers were grown on 4H-SiC wafers of (0001) orientation with a miscut angle of 8 deg at a temperature of 1550 deg C. The epilayers contained regions of high aluminium doping well above the solubility of aluminium in silicon carbide. High temperature annealing of this material resulted in the precipitation of aluminium in the wafers. The samples were analysed by secondary ion mass spectrometry and transmission electron microscopy. Selected area diffraction studies show the presence of aluminium carbide and aluminium silicon carbide phases. Copyright (2002) Australian Society for Electron Microscopy Inc

  6. High-temperature, Knudsen cell-mass spectroscopic studies on lanthanum oxide/uranium dioxide solid solutions

    International Nuclear Information System (INIS)

    Sunder, S.; McEachern, R.; LeBlanc, J.C.

    2001-01-01

    Knudsen cell-mass spectroscopic experiments were carried out with lanthanum oxide/uranium oxide solid solutions (1%, 2% and 5% (metal at.% basis)) to assess the volatilization characteristics of rare earths present in irradiated nuclear fuel. The oxidation state of each sample used was conditioned to the 'uranium dioxide stage' by heating in the Knudsen cell under an atmosphere of 10% CO 2 in CO. The mass spectra were analyzed to obtain the vapour pressures of the lanthanum and uranium species. It was found that the vapour pressure of lanthanum oxide follows Henry's law, i.e., its value is directly proportional to its concentration in the solid phase. Also, the vapour pressure of lanthanum oxide over the solid solution, after correction for its concentration in the solid phase, is similar to that of uranium dioxide. (authors)

  7. Effect of carbide precipitation on the corrosion behavior of Inconel alloy 690

    International Nuclear Information System (INIS)

    Sarver, J.M.; Crum, J.R.; Mankins, W.L.

    1987-01-01

    Intergranular carbide precipitation reactions have been shown to affect the stress corrosion cracking (SCC) resistance of nickel-chromium-iron alloys in environments relative to nuclear steam generators. Carbon solubility curves, time-temperature-sensitization plots and other carbide precipitation data are presented for alloy 690 as an aid in developing heat treatments for improved SCC resistance

  8. 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)

  9. 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.)

  10. Study of elementary mechanisms of creep in uranium as a function of temperature (150 deg. to 760 deg. C) by activation energy measurements; Etude des mecanismes elementaires de deformation par fluage de l'uranium en fonction de la temperature (de 150 deg. a 760 deg. C) par la mesure des energies d'activation

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-07-01

    Creep tests were carried out on single crystals and polycrystalline specimens of uranium in both the {alpha} and {beta} phases over the temperature range 150 - 760 deg. C. The determination of the activation energy for creep and the study of its variation with temperature made it possible to distinguish various temperature ranges in which one or more elementary mechanisms govern deformation. Micrographic observations after creep and the study of the variation of creep-rate with load support the conclusions. The creep behavior of single crystals is identical with that of polycrystalline material below 325 deg. C. From 325 deg. C to one upper limiting temperature whose value depends on the purity and previous history of the metal, the creep deformation of uranium is controlled by cross-slip. From this limiting temperature up to 520 deg. C, the creep of uranium involves two independent mechanisms operating simultaneously, the movement of screw dislocation by cross-slip and the climbing of edge dislocations out of their slip plane. Between 520 deg. C and the {alpha} - {beta} transformation temperature creep in polycrystals is governed by the climb of edge dislocations out of their slip planes, by a pile up mechanism in the case of primary creep and by dipole annihilation in the case of secondary creep. In single crystals creep is dependent on the climb of edge dislocations into pre-existent sub-boundaries and their subsequent rearrangement within these boundaries. In the {beta} phase the creep of polycrystals is governed by the diffusional climb of edge dislocations. Between 450 and 630 deg. C small alloy additions of molybdenum modify the creep characteristics of uranium although the deformation mechanisms involved are analogous to those in the pure metal. (author) [French] Des essais de fluage a diverses temperatures comprises entre 150 et 760 deg. C ont ete effectues sur des polycristaux et des monocristaux d'uranium, en phase {alpha} et en phase {beta}. La

  11. 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....

  12. Effect of HIP temperatures on the microstructure and mechanical properties of carbide dispersed Ti-48Al-1Mn mechanically alloyed compacts

    International Nuclear Information System (INIS)

    Ameyama, Kei; Hashii, Mitsuya; Imai, Nobuyuki; Fujii, Toshinori; Sasaki, Nobuyuki.

    1996-01-01

    The effect of hot isostatic pressing (HIP) temperature on the microstructure and mechanical properties of Ti-48 mol%Al-1 mol%Mn compacts fabricated by mechanical alloying was investigated. N-heptane was used as a process control agent for the mechanical alloying. The compacts HIP treated at 1173, 1373 or 1573 K showed an ultra-fine equiaxed grain structure, i.e., a microduplex structure, consisting of TiAl (γ) and Ti 2 AlC phases, and their average grain sizes were 185 nm, 510 nm and 1.5 μm, respectively. The γ phase was considered to be formed by an α → γ massive transformation during heating. On the other hand, the compacts HIP treated at 1623 or 1673 K showed quite different microstructures from the above HIP compacts. The 1623 K-HIP compact was composed of equiaxed γ grains, whose size was approximately 11.5 μm, rectangular shaped Ti 2 AlC particles, and a small amount of the grain boundary nucleated α phase. Although the 1673 K-HIP compact showed a microstructure similar to the 1623 K-HIP compact, the γ grains were coarsened to be approximately 27.8 μm in diameter and the Ti 2 AlC particles were more elongated rectangles. Furthermore, the amount of the grain boundary nucleated α phase was increased and the lamella α phase nucleated at γ twin boundaries was observed in the 1673 K-HIP compact. Mechanical properties determined by compressive testing at various temperatures made clear that the compacts HIP treated at 1173, 1373 or 1573 K have good workability at elevated temperatures and those HIP treated at 1623 or 1673 K have good high temperature strength. These mechanical properties were influenced significantly by the microstructure, especially by the grain size and morphology of the Ti 2 AlC phase. (author)

  13. Effect of carbides on the creep properties of a Ni-base superalloy M963

    International Nuclear Information System (INIS)

    He, L.Z.; Zheng, Q.; Sun, X.F.; Guan, H.R.; Hu, Z.Q.; Tieu, A.K.; Lu, C.; Zhu, H.T.

    2005-01-01

    Effect of carbides on the creep properties of a cast Ni-base superalloy M963 tested at 800 and 900 deg. C over a broad stress range has been investigated. Correlation between the carbides and creep properties of the alloy is enabled through scanning electron microscopy (SEM) and transmission electron microscopy (TEM). During high temperature creep tests, the primary MC carbide decomposes sluggishly and a large amount of secondary carbides precipitate. The cubic and acicular M 6 C carbide precipitates at the dendritic core region. Extremely fine chromium-rich M 23 C 6 carbide precipitates preferentially at grain boundaries. The M 6 C and M 23 C 6 carbides are found to be beneficial to the creep properties of the alloy. At lower temperature (800 deg. C), the interface of MC carbide with matrix is one of the principal sites for crack initiation. At higher temperature (900 deg. C), the oxidation and the precipitation of μ phase are the main factors for significant loss in creep strength of the alloy

  14. Calculation and measurement of the uranium temperature during irradiation in the experimental channel in the reflector of the RA reactor - Annex 15; Prilog 15 - Proracun i merenje temperature urana pri ozracivanju u eksperimentalnom kanalu reflektora na reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M; Strugar, P; Mitrovic, S [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    Upon demand of the Laboratory for fuel reprocessing, six domestic metal uranium pellets were exposed to neutron flux ( 4 - 5 10{sup 12} n cm{sup -2} sec {sup -1}) in the RA reactor. Irradiation of fuel demanded special analyses for safety reasons. Weight of the fuel pellets was 13 - 20 g, having diameter 20 mm. pellets were placed in leak tight aluminium capsules with helium. The irradiation was dome in the aluminium experimental channel in the graphite reflector. Theoretical study has shown that the expected fuel temperature in the core could be up to 300 deg C at nominal power. For that reason temperature of the capsule with the uranium sample was measured during irradiation by using thermocouples. Results showed the discrepancy between measure and calculated values to be about 30%.

  15. Preparation and thermochemical stability of uranium-zirconium-carbonitrides

    International Nuclear Information System (INIS)

    Kouhsen, C.

    1975-08-01

    This investigation deals with the preparation and the thermochemical stability of uranium-zirconium-carbonitrides as well as with the mechanism of (U,Zr) (C,N)-preparation by carbothermic reduction of uranium-zirconium-oxide. Single-phase (U,Zr) (C,N)-solid solutions with U:Zr-propertions of 3:1, 1:1, and 1:3 were prepared from oxide powder. The thermochemical stability of the (U,Zr) (C,N)-solid solutions against carbon was measured for varying Zr- and N-contents and for several temperatures; the results indicate an increase of the uranium carbide stability potential by the formation of (U,Zr) (C,N)-solid solutions. The thermodynamic properties ΔG 0 , ΔH 0 , and ΔS 0 were calculated and the correlation between the M(C,N)-lattice constant and the N-content was evaluated. Through an intensive investigation of the reaction mechanism, several different reaction paths were found; for each of them the characteristical diffusion of matter was explained by means of the microsections. It was shown that the Zr-concentration of the oxide reactant and the heating rate during the carbothermic reduction influence the species of the reaction product, especially the homogeneity of the (U,Zr) (C,N)-solid solution. (orig.) [de

  16. MC Carbide Characterization in High Refractory Content Powder-Processed Ni-Based Superalloys

    Science.gov (United States)

    Antonov, Stoichko; Chen, Wei; Huo, Jiajie; Feng, Qiang; Isheim, Dieter; Seidman, David N.; Sun, Eugene; Tin, Sammy

    2018-04-01

    Carbide precipitates in Ni-based superalloys are considered to be desirable phases that can contribute to improving high-temperature properties as well as aid in microstructural refinement of the material; however, they can also serve as crack initiation sites during fatigue. To date, most of the knowledge pertaining to carbide formation has originated from assessments of cast and wrought Ni-based superalloys. As powder-processed Ni-based superalloys are becoming increasingly widespread, understanding the different mechanisms by which they form becomes increasingly important. Detailed characterization of MC carbides present in two experimental high Nb-content powder-processed Ni-based superalloys revealed that Hf additions affect the resultant carbide morphologies. This morphology difference was attributed to a higher magnitude of elastic strain energy along the interface associated with Hf being soluble in the MC carbide lattice. The composition of the MC carbides was studied through atom probe tomography and consisted of a complex carbonitride core, which was rich in Nb and with slight Hf segregation, surrounded by an Nb carbide shell. The characterization results of the segregation behavior of Hf in the MC carbides and the subsequent influence on their morphology were compared to density functional theory calculations and found to be in good agreement, suggesting that computational modeling can successfully be used to tailor carbide features.

  17. Titanium carbide nanocube core induced interfacial growth of crystalline polypyrrole/polyvinyl alcohol lamellar shell for wide-temperature range supercapacitors

    Science.gov (United States)

    Weng, Yu-Ting; Pan, Hsiao-An; Wu, Nae-Lih; Chen, Geroge Zheng

    2015-01-01

    This is the first investigation on electrically conducting polymers-based supercapacitor electrodes over a wide temperature range, from -18 °C to 60 °C. A high-performance supercapacitor electrode material consisting of TiC nanocube core and conformal crystalline polypyrrole (PPy)/poly-vinyl-alcohol (PVA) lamellar shell has been synthesized by heterogeneous nucleation-induced interfacial crystallization. PPy is induced to crystallize on the negatively charged TiC nanocube surfaces via strong interfacial interactions. In this organic-inorganic hybrid nanocomposite, the long chain PVA enables enhanced cycle life due to improved mechanical properties, and the TiC nanocube not only contributes to electron conduction, but also dictates the PPy morphology/crystallinity for maximizing the charging-discharging performance. The crystalline PPy/PAV layer on the TiC nanocube offers unprecedented high capacity (>350 F g-1-PPy at 300 mV s-1 with ΔV = 1.6 V) and cycling stability in a temperature range from -18 °C to 60 °C. The presented hybrid-filler and interfacial crystallization strategies can be applied to the exploration of new-generation high-power conducting polymer-based supercapacitor materials.

  18. 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

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

    Science.gov (United States)

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

    2018-01-01

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

  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. 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

  2. Mechanical behaviour of uranium; Comportement mecanique de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J L; Coureau, G [Commissariat a l' Energie Atomique, Dir. Industrielle, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    The chief mechanical properties of uranium, taken at room and at different temperatures, are presented in this report. (author) [French] Dans ce rapport sont presentees les principales caracteristiques mecaniques de l'uranium, relevees a l'ambiante et a differentes temperatures. (auteur)

  3. Uranium oxide recovering method

    International Nuclear Information System (INIS)

    Ota, Kazuaki; Takazawa, Hiroshi; Teramae, Naoki; Onoue, Takeshi.

    1997-01-01

    Nitrates containing uranium nitrate are charged in a molten salt electrolytic vessel, and a heat treatment is applied to prepare molten salts. An anode and a cathode each made of a graphite rod are disposed in the molten salts. AC voltage is applied between the anode and the cathode to conduct electrolysis of the molten salts. Uranium oxides are deposited as a recovered product of uranium, on the surface of the anode. The nitrates containing uranium nitrate are preferably a mixture of one or more nitrates selected from sodium nitrate, potassium nitrate, calcium nitrate and magnesium nitrate with uranium nitrate. The nitrates may be liquid wastes of nitrates. The temperature for the electrolysis of the molten salts is preferably from 150 to 300degC. The voltage for the electrolysis of the molten salts is preferably an AC voltage of from 2 to 6V, more preferably from 4 to 6V. (I.N.)

  4. The Apparent Contact Angle and Wetted Area of Active Alloys on Silicon Carbide as a Function of the Temperature and the Surface Roughness: A Multivariate Approach

    Science.gov (United States)

    Tillmann, Wolfgang; Pfeiffer, Jan; Wojarski, Lukas

    2015-08-01

    Despite the broad field of applications for active filler alloys for brazing ceramics, as well as intense research work on the wetting and spreading behavior of these alloys on ceramic surfaces within the last decades, the manufactured joints still exhibit significant variations in their properties due to the high sensitivity of the alloys to changing brazing conditions. This increases the need for investigations of the wetting and spreading behavior of filler alloys with regard to the dominating influences combined with their interdependencies, instead of solely focusing on single parameter investigations. In this regard, measurements of the wetting angle and area were conducted at solidified AgCuTi and CuSnTi alloys on SiC substrates. Based on these measurements, a regression model was generated, illustrating the influence of the brazing temperature, the roughness of the faying surfaces, the furnace atmosphere, and their interdependencies on the wetting and spreading behavior of the filler alloys. It was revealed that the behavior of the melts was significantly influenced by the varied brazing parameters, as well as by their interdependencies. This result was also predicted by the developed model and showed a high accuracy.

  5. Microstructural evaluation of the NbC-20Ni cemented carbides during sintering

    International Nuclear Information System (INIS)

    Rodrigues, D.; Cannizza, E.

    2016-01-01

    Full text: Fine carbides in a metallic matrix (binder) form the microstructure of the cemented carbides. Grain size and binder content are the main variables to adjust hardness and toughness. These products are produced by Powder Metallurgy, and traditional route involves mixing carbides with binder by high energy milling, pressing and sintering. During sintering, a liquid phase promotes densification, and a final relative density higher than 99% is expected. Sintering is carried out at high temperatures, and dissolution of the carbides changes the chemical composition of the binder. To control grain growth of the main carbide, which reduces hardness, small quantities of secondary carbides are used. These additives limit dissolution and precipitation of the main carbides reducing the final grain size. This paper focused the structural and chemical evolution during sintering using NbC-20Ni cermets. Mixtures of very fine NbC carbides and carbonyl Ni powders were produce by intense milling. These mixtures were pressed using uniaxial pressures from 50 to 200MPa. Shrinkage was evaluated using dilatometric measurements under an atmosphere of dynamic argon. Samples were also sintered under vacuum in high temperature industrial furnace. The sintered samples were characterized in terms of density hardness, toughness and microstructure. DRX was the main tool used to evaluate the structural evolution of the binder. In situ chemical analysis helped to understand the dissolution mechanisms. (author)

  6. Influence of Heat Treatment on Content of the Carbide Phases in the Microstructure of High-Speed Steel

    Directory of Open Access Journals (Sweden)

    Jaworski J.

    2017-09-01

    Full Text Available This article presents the results of investigations of the effect of heat treatment temperature on the content of the carbide phase of HS3-1-2 and HS6-5-2 low-alloy high-speed steel. Analysis of the phase composition of carbides is carried out using the diffraction method. It is determined that with increasing austenitising temperature, the intensification of dissolution of M6C carbide increases. As a result, an increase in the grain size of the austenite and the amount of retained austenite causes a significant reduction in the hardness of hardened steel HS3-1-2 to be observed. The results of diffraction investigations showed that M7C3 carbides containing mainly Cr and Fe carbides and M6C carbides containing mainly Mo and W carbides are dissolved during austenitisation. During austenitisation of HS3-1-2 steel, the silicon is transferred from the matrix to carbides, thus replacing carbide-forming elements. An increase in a degree of tempering leads to intensification of carbide separation and this process reduce the grindability of tested steels.

  7. Electron microscopy of boron carbide before and after electron irradiation

    International Nuclear Information System (INIS)

    Stoto, T.; Zuppiroli, L.; Beauvy, M.; Athanassiadis, T.

    1984-06-01

    The microstructure of boron carbide has been studied by electron microscopy and related to the composition of the material. After electron irradiations in an usual transmission electron microscope and in a high voltage electron microscope at different temperatures and fluxes no change of these microstructures have been observed but a sputtering of the surface of the samples, which has been studied quantitatively [fr

  8. 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

  9. Comparative sinterability of combustion synthesized and commercial titanium carbides

    International Nuclear Information System (INIS)

    Manley, B.W.

    1984-11-01

    The influence of various parameters on the sinterability of combustion synthesized titanium carbide was investigaged. Titanium carbide powders, prepared by the combustion synthesis process, were sintered in the temperature range 1150 to 1600 0 C. Incomplete combustion and high oxygen contents were found to be the cause of reduced shrinkage during sintering of the combustion syntheized powders when compared to the shrinkage of commercial TiC. Free carbon was shown to inhibit shrinkage. The activation energy for sintering was found to depend on stoichiometry (C/Ti). With decreasing C/Ti, the rate of sintering increased. 29 references, 16 figures, 13 tables

  10. Influence of uranium hydride oxidation on uranium metal behaviour

    International Nuclear Information System (INIS)

    Patel, N.; Hambley, D.; Clarke, S.A.; Simpson, K.

    2013-01-01

    This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

  11. Influence of uranium hydride oxidation on uranium metal behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Patel, N.; Hambley, D. [National Nuclear Laboratory (United Kingdom); Clarke, S.A. [Sellafield Ltd (United Kingdom); Simpson, K.

    2013-07-01

    This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

  12. Preparation of hafnium carbide by chemical vapor deposition

    International Nuclear Information System (INIS)

    Hertz, Dominique.

    1974-01-01

    Hard, adhesive coatings of single-phase hafnium carbide were obtained by chemical vapor reaction in an atmosphere containing hafnium tetrachloride, methane and a large excess of hydrogen. By varying the gas phase composition and temperature the zones of formation of the different solid phases were studied and the growth of elementary hafnium and carbon deposits evaluated separately. The results show that the mechanism of hafnium carbide deposition does not hardly involve phenomene of homogeneous-phase methane decomposition or tetrachloride reduction by hydrogen unless the atmosphere is very rich or very poor in methane with respect to tetrachloride. However, hydrogen acting inversely on these two reactions, affects the stoichiometry of the substance deposited. The methane decomposition reaction is fairly slow, the reaction leading to hafnium carbide deposition is faster and that of tetrachloride reduction by hydrogen is quite fast [fr

  13. Synthesis of carbides of refractory metals in salt melts

    International Nuclear Information System (INIS)

    Ilyushchenko, N.G.; Anfinogenov, A.I.; Chebykin, V.V.; Chernov, Ya.B.; Shurov, N.I.; Ryaposov, Yu.A.; Dobrynin, A.I.; Gorshkov, A.V.; Chub, A.V.

    2003-01-01

    The ion-electron melts, obtained through dissolving the alkali and alkali-earth metals in the molten chlorides above the chloride melting temperature, were used for manufacturing the high-melting metal carbides as the transport melt. The lithium, calcium and magnesium chlorides and the mixture of the lithium chloride with the potassium or calcium chloride were used from the alkali or alkali-earth metals. The metallic lithium, calcium, magnesium or the calcium-magnesium mixtures were used as the alkali or alkali-earth metals. The carbon black or sugar was used as carbon. It is shown, that lithium, magnesium or calcium in the molten salts transfer the carbon on the niobium, tantalum, titanium, forming the carbides of the above metals. The high-melting metal carbides are obtained both from the metal pure powders and from the oxides and chlorides [ru

  14. Temperature-dependent chemical changes of metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Youn, Young Sang; Lee, Jeong Mook; KimJong Hwan; Song, Hoon; Kim, Jong Yun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    We observed the temperature-dependent variations of UZr alloy using surface analysis techniques such as X-ray photoelectron spectroscopy (XPS), Raman spectroscopy, X-ray diffraction (XRD), and scanning electron microscope (SEM) equipped with energy-dispersive Xray spectroscope (EDS). In this work, we exhibited the results of XPS, Raman, XRD, and SEM-EDS for U-10wt%Zr alloy at room temperature, 610 and 1130 .deg. C. In SEM-EDS data, we observed that uranium and zirconium elements uniformly exist. After the annealing of U-10Zr sample at 1130 .deg. C, the formation of zirconium carbide is verified through Raman spectroscopy and XRD results. Additionally, the change of valence state for uranium element is also confirmed by XPS analysis.

  15. Thermodynamic analysis of thermal plasma process of composite zirconium carbide and silicon carbide production from zircon concentrates

    International Nuclear Information System (INIS)

    Kostic, Z.G.; Stefanovic, P.Lj.; Pavlovic; Pavlovic, Z.N.; Zivkovic, N.V.

    2000-01-01

    Improved zirconium ceramics and composites have been invented in an effort to obtain better resistance to ablation at high temperature. These ceramics are suitable for use as thermal protection materials on the exterior surfaces of spacecraft, and in laboratory and industrial environments that include flows of hot oxidizing gases. Results of thermodynamic consideration of the process for composite zirconium carbide and silicon carbide ultrafine powder production from ZrSiO 4 in argon thermal plasma and propane-butane gas as reactive quenching reagents are presented in the paper. (author)

  16. High-temperature gas-cooled reactors and process heat

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1980-01-01

    High-Temperature Gas-Cooled Reactors (HTGRs) are fueled with ceramic-coated microspheres of uranium and thorium oxides/carbides embedded in graphite blocks which are cooled with helium. Promising areas of HTGR application are in cogeneration, energy transport using Heat Transfer Salt, recovery of oils from oil shale, steam reforming of methane for chemical production, coal gasification, and in energy transfer using chemical heat jpipes in the long term. Further, HTGRs could be used as the energy source for hydrogen production through thermochemical water splitting in the long term. The potential market for Process Heat HTGRs is 100-200 large units by about the year 2020

  17. RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

    Science.gov (United States)

    Gens, T.A.

    1962-07-10

    An improvement was made in a process of recovering uranium from a uranium-zirconium composition which was hydrochlorinated with gsseous hydrogen chloride at a temperature of from 350 to 800 deg C resulting in volatilization of the zirconium, as zirconium tetrachloride, and the formation of a uranium containing nitric acid insoluble residue. The improvement consists of reacting the nitric acid insoluble hydrochlorination residue with gaseous carbon tetrachloride at a temperature in the range 550 to 600 deg C, and thereafter recovering the resulting uranium chloride vapors. (AEC)

  18. Production of uranium dioxide

    International Nuclear Information System (INIS)

    Hart, J.E.; Shuck, D.L.; Lyon, W.L.

    1977-01-01

    A continuous, four stage fluidized bed process for converting uranium hexafluoride (UF 6 ) to ceramic-grade uranium dioxide (UO 2 ) powder suitable for use in the manufacture of fuel pellets for nuclear reactors is disclosed. The process comprises the steps of first reacting UF 6 with steam in a first fluidized bed, preferably at about 550 0 C, to form solid intermediate reaction products UO 2 F 2 , U 3 O 8 and an off-gas including hydrogen fluoride (HF). The solid intermediate reaction products are conveyed to a second fluidized bed reactor at which the mol fraction of HF is controlled at low levels in order to prevent the formation of uranium tetrafluoride (UF 4 ). The first intermediate reaction products are reacted in the second fluidized bed with steam and hydrogen at a temperature of about 630 0 C. The second intermediate reaction product including uranium dioxide (UO 2 ) is conveyed to a third fluidized bed reactor and reacted with additional steam and hydrogen at a temperature of about 650 0 C producing a reaction product consisting essentially of uranium dioxide having an oxygen-uranium ratio of about 2 and a low residual fluoride content. This product is then conveyed to a fourth fluidized bed wherein a mixture of air and preheated nitrogen is introduced in order to further reduce the fluoride content of the UO 2 and increase the oxygen-uranium ratio to about 2.25

  19. Boron-carbide-aluminum and boron-carbide-reactive metal cermets. [B/sub 4/C-Al

    Science.gov (United States)

    Halverson, D.C.; Pyzik, A.J.; Aksay, I.A.

    1985-05-06

    Hard, tough, lighweight boron-carbide-reactive metal composites, particularly boron-carbide-aluminum composites, are produced. These composites have compositions with a plurality of phases. A method is provided, including the steps of wetting and reacting the starting materials, by which the microstructures in the resulting composites can be controllably selected. Starting compositions, reaction temperatures, reaction times, and reaction atmospheres are parameters for controlling the process and resulting compositions. The ceramic phases are homogeneously distributed in the metal phases and adhesive forces at ceramic-metal interfaces are maximized. An initial consolidated step is used to achieve fully dense composites. Microstructures of boron-carbide-aluminum cermets have been produced with modules of rupture exceeding 110 ksi and fracture toughness exceeding 12 ksi..sqrt..in. These composites and methods can be used to form a variety of structural elements.

  20. 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

  1. 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.

  2. Uranium dioxide pellets

    International Nuclear Information System (INIS)

    Zawidzki, T.W.

    1982-01-01

    A process for the preparation of a sintered, high density, large crystal grain size uranium dioxide pellet is described which involves: (i) reacting a uranyl nitrate of formula UO 2 (NO 3 ) 2 .6H 2 O with a sulphur source, at a temperature of from about 300 deg. C to provide a sulphur-containing uranium trioxide; (ii) reacting the thus-obtained modified uranium trioxide with ammonium nitrate to form an insoluble sulphur-containing ammonium uranate; (iii) neutralizing the thus-formed slurry with ammonium hydroxide to precipitate out as an insoluble ammonium uranate the remaining dissolved uranium; (iv) recovering the thus-formed precipitates in a dry state; (v) reducing the dry precipitate to UO 2 , and forming it into 'green' pellets; and (vi) sintering the pellets in a hydrogen atmosphere at an elevated temperature

  3. Influent of Carbonization of Sol Solution at the External Gelation Process on the Quality of Uranium Oxide Kernel

    International Nuclear Information System (INIS)

    Damunir; Sukarsono

    2007-01-01

    The influent of carbonization of sol solution at the external gelation process on the quality of uranium oxide kernel was done. Variables observed are the influent of carbon, temperature and time of reduction process of U 3 O 8 kernel resulted from carbonization of sol solution. First of all, uranyl nitrate was reacted with 1 M NH 4 OH solution, producing the colloid of UO 3 . Then by mixing and heating up to the temperature of 60-80 °C, the colloid solution was reacted with PVA, mono sorbitol oleate and paraffin producing of uranium-PVA sol. Then sol solution was carbonized with carbon black of mol ratio of carbon to uranium =2.32-6.62, produce of carbide gel. Gel then washed, dried and calcined at 800 °C for 4 hours to produce of U 3 O 8 kernel containing carbon. Then the kernel was reduced by H 2 gas in the medium of N 2 gas at 500-800 °C, 50 mmHg pressure for 3 hours. The process was repeated at 700 °C, 50 mmHg pressure for 1-4 hours. The characterization of chemical properties of the gel grains and uranium oxide kernel using FTIR covering the analysis of absorption band of infra red spectrum of UO 3 , C-OH, NH 3 , C-C, C-H and OH functional group. The physical properties of uranium oxide covering specific surface area, void volume, mean diameter using surface area meter Nova-1000 and as N 2 gas an absorbent. And O/U ratio of uranium dioxide kernel by gravimetry method. The result of experiment showed that carbonization of sol solution at the external gelation process give influencing the quality of uranium oxide kernel. (author)

  4. Ordering effects on structure and specific heat of nonstoichiometric titanium carbide

    International Nuclear Information System (INIS)

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

    1999-01-01

    The experimental results on the change in the crystal structure and specific heat of the nonstoichiometric titanium carbide TiC y (0.5 2 C phases with cubic and trigonal symmetry and the rhombic ordered Ti 3 C 2 phase are formed in the titanium carbide at the temperature below 1000 K by the phase transitions mechanism. The temperatures and heats of the order-disorder phase transitions are determined [ru

  5. Preparation of fiber reinforced titanium diboride and boron carbide composite bodies

    International Nuclear Information System (INIS)

    Newkirk, L.R.; Riley, R.E.; Sheinberg, H.; Valencia, F.A.; Wallace, T.C.

    1979-01-01

    A process is described for uniformly infiltrating woven carbon cloth with either titanium diboride or boron carbide at reduced pressure (15 to 25 torr). The effects of deposition temperature on the uniformity of penetration and on coating rate are described for temperatures from 750 to 1000 0 C and deposit loadings from 20 to 43 vol. %. For the boron carbides, boron composition is discussed and evidence is presented suggesting that propene is the dominant rate controlling reactant

  6. 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.

  7. Formation of carbides and their effects on stress rupture of a Ni-base single crystal superalloy

    International Nuclear Information System (INIS)

    Liu, L.R.; Jin, T.; Zhao, N.R.; Sun, X.F.; Guan, H.R.; Hu, Z.Q.

    2003-01-01

    Creep tests of a nickel-base single crystal superalloy with minor C addition and non-carbon were carried out at different temperatures and stresses. Correlations between microstructural change and testing temperature and stress were enabled through scanning electron microscopy (SEM) and transmission electron microscopy (TEM), detailing the rafting microstucture and carbides precipitation. The results showed that minor carbon addition prolonged the second stage of creep strain curves and improved creep properties. Some carbide was precipitated during creep tests in modified alloy. M 23 C 6 carbide precipitated at lower temperature (871-982 deg. C), while (M 6 C) 2 carbide precipitated at higher temperature (>1000 deg. C), all of which was considered to be beneficial to creep properties. A small amount of MC carbide formed during solidification and its decomposition product (M 6 C) 1 were detrimental to mechanical properties, which together with micropores provided the site of initiation of cracks and led to the final fracture

  8. Natural precursor based hydrothermal synthesis of sodium carbide for reactor applications

    Science.gov (United States)

    Swapna, M. S.; Saritha Devi, H. V.; Sebastian, Riya; Ambadas, G.; Sankararaman, S.

    2017-12-01

    Carbides are a class of materials with high mechanical strength and refractory nature which finds a wide range of applications in industries and nuclear reactors. The existing synthesis methods of all types of carbides have problems in terms of use of toxic chemical precursors, high-cost, etc. Sodium carbide (Na2C2) which is an alkali metal carbide is the least explored one and also that there is no report of low-cost and low-temperature synthesis of sodium carbide using the eco-friendly, easily available natural precursors. In the present work, we report a simple low-cost, non-toxic hydrothermal synthesis of refractory sodium carbide using the natural precursor—Pandanus. The formation of sodium carbide along with boron carbide is evidenced by the structural and morphological characterizations. The sample thus synthesized is subjected to field emission scanning electron microscopy (FESEM), x-ray powder diffraction (XRD), ultraviolet (UV)—visible spectroscopy, Fourier transform infrared spectroscopy (FTIR), Raman, and photoluminescent (PL) spectroscopic techniques.

  9. 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)

  10. 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)

  11. 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

  12. The use of room temperature phosphorescence for the determination of uranium in tin-tailings mineral samples

    International Nuclear Information System (INIS)

    Meor Yusof bin Meor Sulaiman

    1988-01-01

    The possibility of using phosphorescence technique in determining uranium in mineral samples and its comparison with that of fluorescence using high carbonate flux is presented. Samples used are tin-tailings mineral such as monazite, xenotime, ilmenite and zircon. The calibration graph obtained shows a linear relationship between the concentration range of 0-55 ppm U. From here, analysis of the standard showed that the result obtained and that of the certified value are consistent. HN0 3 :H 2 SO 4 (1:3) and phosphoric acid leaching methods are tried and the results show that phosphoric acid is the better method for phosphate mineral. Comparison of the results obtained from this technique and that of the direct and extraction methods of fluorimetry are also made. Phosphorescence is found to be a better method in determining uranium in this type of samples. (author)

  13. Study on complexation behaviour of uranium and thorium with amino acids at different temperatures in aqueous media

    International Nuclear Information System (INIS)

    Joshi, J.D.; Patel, M.R.; Patel, A.D.

    1992-01-01

    The complexation behaviour of uranium and thorium with important amino acids have been studied using Irving-Rossotti titration technique at 25deg, 35deg and 45degC in inert atmosphere of nitrogen and 0.1M ionic strength using NaClO 4 . The thermodynamic parameters ΔG, ΔH and ΔS have been calculated. Results indicate that thorium (IV) is forming more stable complexes than UO 2 2+ . (author). 3 refs., 2 tab

  14. Thermodynamic study contribution of U-Fe and U-Ga alloys by high temperature mass spectroscopy, and of the wetting of yttrium oxide by uranium

    International Nuclear Information System (INIS)

    Gardie, P.

    1992-01-01

    High temperature thermodynamic properties study of U-Fe and U-Ga alloys, and wetting study of yttrium oxide by uranium are presented. High temperature mass spectrometry coupled to a Knudsen effusion multi-cell allows to measure iron activity in U-Fe alloys and of gallium in U-Ga alloys, the U activity is deduced from Gibbs-Duhem equation. Wetting of the system U/Y_2O_3_-_x is studied between 1413 K and 1973 K by the put drop method visualized by X-rays. This technique also furnishes density, surface tension of U and of U-Fe alloys put on Y_2O_3_-_x. A new model of the interfacial oxygen action on wetting is done for the system U/Y_2O_3_-_x. (A.B.)

  15. Boron carbide in pile behaviour Rapsodie experience

    International Nuclear Information System (INIS)

    Kryger, B.; Colin, M.

    1983-04-01

    Results concerning boron carbide irradiation experiments performed in RAPSODIE up to 10 22 .cm - 3 capture density in the temperature range 600-1100 0 lead to the following main conclusions: initial density and grain size lowering contribute to swelling decrease but density is the major parameter for swelling limitation; swelling rate can vary in a wide range (ratio 1 to 3) according to combinations of density (1.8 to 2.3) and grain size (10 to 50 μm) values; a swelling balance reveals that the most important contribution to swelling should be a high density of helium small bubbles (<400 A); helium retention increases with density and grain size and decreases with temperature elevation. A diffusion law is proposed to describe the rate of helium release

  16. Reactivation properties of carbide slag as a CO{sub 2} sorbent during calcination/carbonation cycles

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yingjie; Sun, Rongyue; Liu, Hongling; Lu, Chunmei [Shandong Univ., Jinan (China). School of Energy and Power Engineering

    2013-07-01

    The carbide slag from polyvinyl chloride production as industry hazardous wastes was proposed as CO{sub 2} sorbent at high temperature in calcium looping cycle. The cyclic CO{sub 2} capture behavior and the microstructure characteristics of the carbide slag as one of the typical calcium-based industrial wastes during the multiple calcination/carbonation cycles. Also, the comparisons between the carbide slag and the natural limestone in cyclic CO{sub 2} capture behavior were made. XRD analysis demonstrates that the predominating constituent of the carbide slag is Ca(OH){sub 2}. The carbonation temperature ranging from 650 to 700 C is favourable to cyclic carbonation of the carbide slag. The cyclic carbonation conversions of the carbide slag is lower than that of the limestone before a certain time, but the situation is converse after that time in a thermogravimetric analyzer. The carbide slag has better cyclic CO{sub 2} capture capacity. The carbonation conversion of the carbide slag retains 0.28 after 100 calcination/carbonation cycles, while the two limestones achieve 0.08 and 0.14 respectively at the same reaction conditions in a dual fixed-bed reactor. The microstructure of the carbide slag by SEM reveals the reason why it possesses better CO{sub 2} capture capacity.

  17. Precise coulometric titration of uranium in a high-purity uranium metal and in uranium compounds

    International Nuclear Information System (INIS)

    Tanaka, Tatsuhiko; Yoshimori, Takayoshi

    1975-01-01

    Uranium in uranyl nitrate, uranium trioxide and a high-purity uranium metal was assayed by the coulometric titration with biamperometric end-point detection. Uranium (VI) was reduced to uranium (IV) by solid bismuth amalgam in 5M sulfuric acid solution. The reduced uranium was reoxidized to uranium (VI) with a large excess of ferric ion at a room temperature, and the ferrous ion produced was titrated with the electrogenerated manganese(III) fluoride. In the analyses of uranium nitrate and uranium trioxide, the results were precise enough when the error from uncertainty in water content in the samples was considered. The standard sample of pure uranium metal (JAERI-U4) was assayed by the proposed method. The sample was cut into small chips of about 0.2g. Oxides on the metal surface were removed by the procedure shown by National Bureau of Standards just before weighing. The mean assay value of eleven determinations corrected for 3ppm of iron was (99.998+-0.012) % (the 95% confidence interval for the mean), with a standard deviation of 0.018%. The proposed coulometric method is simple and permits accurate and precise determination of uranium which is matrix constituent in a sample. (auth.)

  18. Development of a Robust Tri-Carbide Fueled Reactor for Multi-Megawatt Space Power and Propulsion Applications

    International Nuclear Information System (INIS)

    Samim Anghaie; Knight, Travis W.; Plancher, Johann; Gouw, Reza

    2004-01-01

    An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors

  19. Temperature effect on the physico-chemical properties of silica based bio-hybrid composite for uranium uptake

    International Nuclear Information System (INIS)

    Mishra, Archana; Melo, Jose Savio

    2013-01-01

    In the present work, silica based bio-hybrid composite has been prepared using Streptococcus lactis cells and silica nanoparticles through one step single process of spray drying. Bio-hybrids have many desired characteristics, and are thus used in a wide range of applications for example environmental cleanup which is of increasing importance. Thermogravimetric and thermodynamic analysis have been employed to understand the binding of uranium to the synthesized bio-hybrid material. Analysis of the thermodynamic parameters (ΔG 0 , ΔS 0 and ΔH 0 ) provides information regarding the inherent energy and feasibility of the sorption process. (author)

  20. Examination of temperature-induced shape memory of uranium--5.3-to 6.9 weight percent niobium alloys

    International Nuclear Information System (INIS)

    Hemperly, V.C.

    1976-01-01

    The uranium-niobium alloy system was examined in the range of 5.3-to-6.9 weight percent niobium with respect to shape memory, mechanical properties, metallography, Coefficients of linear thermal expansion, and differential thermal analysis. Shape memory increased with increasing niobium levels in the study range. There were no useful correlations found between shape memory and the other tests. Coefficients of linear thermal expansion tests of as-quenched 5.8 and 6.2 weight percent niobium specimens, but not 5.3 and 6.9 weight percent niobium specimens, had a contraction component on heating, but the phenomenon was not a contributor to shape memory

  1. Internal friction in uranium

    International Nuclear Information System (INIS)

    Selle, J.E.

    1975-01-01

    Results are presented of studies conducted to relate internal friction measurements in U to allotropic transformations. It was found that several internal friction peaks occur in α-uranium whose magnitude changed drastically after annealing in the β phase. All of the allotropic transformations in uranium are diffusional in nature under slow heating and cooling conditions. Creep at regions of high stress concentration appears to be responsible for high temperature internal friction in α-uranium. The activation energy for grain boundary relaxation in α-uranium was found to be 65.1 +- 4 kcal/mole. Impurity atoms interfere with the basic mechanism for grain boundary relaxation resulting in a distribution in activation energies. A considerable distribution in ln tau 0 was also found which is a measure of the distribution in local order and in the Debye frequency around a grain boundary

  2. METHOD OF APPLYING NICKEL COATINGS ON URANIUM

    Science.gov (United States)

    Gray, A.G.

    1959-07-14

    A method is presented for protectively coating uranium which comprises etching the uranium in an aqueous etching solution containing chloride ions, electroplating a coating of nickel on the etched uranium and heating the nickel plated uranium by immersion thereof in a molten bath composed of a material selected from the group consisting of sodium chloride, potassium chloride, lithium chloride, and mixtures thereof, maintained at a temperature of between 700 and 800 deg C, for a time sufficient to alloy the nickel and uranium and form an integral protective coating of corrosion-resistant uranium-nickel alloy.

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

    International Nuclear Information System (INIS)

    Tench, R.J.

    1992-11-01

    For the first time, nanometer scale uranium clusters were created on the basal plane of highly oriented pyrolytic graphite by laser ablation under ultra-high vacuum conditions. The physical and chemical properties of these clusters were investigated by scanning tunneling microscopy (STM) as well as standard surface science techniques. Auger electron and X-ray photoelectron spectroscopies found the uranium deposit to be free of contamination and showed that no carbide had formed with the underlying graphite. Clusters with sizes ranging from 42 Angstrom 2 to 630 Angstrom 2 were observed upon initial room temperature deposition. Surface diffusion of uranium was observed after annealing the substrate above 800 K, as evidenced by the decreased number density and the increased size of the clusters. Preferential depletion of clusters on terraces near step edges as a result of annealing was observed. The activation energy for diffusion deduced from these measurements was found to be 15 Kcal/mole. Novel formation of ordered uranium thin films was observed for coverages greater than two monolayers after annealing above 900 K. These ordered films displayed islands with hexagonally faceted edges rising in uniform step heights characteristic of the unit cell of the P-phase of uranium. In addition, atomic resolution STM images of these ordered films indicated the formation of the β-phase of uranium. The chemical properties of these surfaces were investigated and it was shown that these uranium films had a reduced oxidation rate in air as compared to bulk metal and that STM imaging in air induced a polarity-dependent enhancement of the oxidation rate

  4. Atmospheric corrosion of uranium-carbon alloys; Corrosion atmospherique des alliages uranium-carbone

    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 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) [French] Les auteurs etudient la corrosion des alliages uranium-carbone de composition voisine du monocarbure; ils montrent que l'importance des effets de la corrosion observee augmente avec la teneur en vapeur d'eau du milieu gazeux ambiant et concluent que la corrosion atmospherique de ces alliages est due essentiellement a l'humidite de l'air, l'action de l'oxygene de l'air etant tres faible a la temperature ambiante. Ils indiquent que les conditions optimales de conservation des alliages U-C sont le vide ou une atmosphere d'argon parfaitement desseches. D'autre part, les auteurs etablissent que le type de corrosion mis en jeu est une corrosion 'fissurante sous contrainte', transgranulaire (pouvant egalement etre intergranulaire dans le cas d'alliages sous-stoechiometriques). Ils proposent enfin deux hypotheses pour rendre compte de ce mecanisme, dont l'une est illustree par la mise en evidence, a l'interface des fissures, de produits de corrosion pouvant jouer le role de 'coins' dans les grains de

  5. Anticorrosion protection of uranium

    Energy Technology Data Exchange (ETDEWEB)

    Goncharov, Ivan D.; Kazakovskaya, Tatiana; Tukmakov, Victor; Shapovalov, Vyacheslav [Russian Federal Nuclear Center-VNIIEF, 37, Mira Ave., RU-607190 Sarov (Nizhnii Gorod), 010450 (Russian Federation)

    2004-07-01

    Uranium in atmospheric conditions is non-stable. Sloughing products are being generated on its surface during storage or use. These corrosion products make many difficulties because of necessity to provide personnel safety. Besides, uranium corrosion may cause damage in parts. The first works devoted to uranium corrosion were performed in the framework of the USA Manhattan Project in the early forties of last century. Various methods of uranium protection were investigated, among them the galvanic one was the most studied. Later on the galvanic technology was patented. The works on this problem remains urgent up to the present time. In Russia, many methods of uranium corrosion protection, mainly against atmospheric corrosion, were tried on. In particular, such methods as diffusion zinc and paint coating were investigated. In the first case, a complex intermetallic U-Zn compound was formed but its protection was not reliable enough, this protection system was inconvenient and uncertain and that is why an additional paint coating was necessary. In the case of paint coatings another problem appeared. It was necessary to find such a coating where gas-permeability would prevail over water-permeability. Otherwise significant uranium corrosion occurs. This circumstance together with low mechanical resistance of paint coatings does not allow to use paint coating for long-term protection of uranium. Currently, there are following methods of uranium protection: ion-plasma, galvanic and thermo-vacuum annealing. These are described in this paper. In the end the issue of corrosion protection in reactor core zones is addressed. Here the greatest difficulties are caused when enriched uranium heated up to 500 deg. C needs anticorrosion protection. In this case various metal coatings are not reliable because of brittle inter-metallide formation. The reliable protection may be provided only up to the temperature plus 400 - 500 deg. C with the help of galvanic copper coating since

  6. Anticorrosion protection of uranium

    International Nuclear Information System (INIS)

    Goncharov, Ivan D.; Kazakovskaya, Tatiana; Tukmakov, Victor; Shapovalov, Vyacheslav

    2004-01-01

    Uranium in atmospheric conditions is non-stable. Sloughing products are being generated on its surface during storage or use. These corrosion products make many difficulties because of necessity to provide personnel safety. Besides, uranium corrosion may cause damage in parts. The first works devoted to uranium corrosion were performed in the framework of the USA Manhattan Project in the early forties of last century. Various methods of uranium protection were investigated, among them the galvanic one was the most studied. Later on the galvanic technology was patented. The works on this problem remains urgent up to the present time. In Russia, many methods of uranium corrosion protection, mainly against atmospheric corrosion, were tried on. In particular, such methods as diffusion zinc and paint coating were investigated. In the first case, a complex intermetallic U-Zn compound was formed but its protection was not reliable enough, this protection system was inconvenient and uncertain and that is why an additional paint coating was necessary. In the case of paint coatings another problem appeared. It was necessary to find such a coating where gas-permeability would prevail over water-permeability. Otherwise significant uranium corrosion occurs. This circumstance together with low mechanical resistance of paint coatings does not allow to use paint coating for long-term protection of uranium. Currently, there are following methods of uranium protection: ion-plasma, galvanic and thermo-vacuum annealing. These are described in this paper. In the end the issue of corrosion protection in reactor core zones is addressed. Here the greatest difficulties are caused when enriched uranium heated up to 500 deg. C needs anticorrosion protection. In this case various metal coatings are not reliable because of brittle inter-metallide formation. The reliable protection may be provided only up to the temperature plus 400 - 500 deg. C with the help of galvanic copper coating since

  7. ELECTRODEPOSITION OF NICKEL ON URANIUM

    Science.gov (United States)

    Gray, A.G.

    1958-08-26

    A method is described for preparing uranium objects prior to nickel electroplating. The process consiats in treating the surface of the uranium with molten ferric chloride hexahydrate, at a slightiy elevated temperature. This treatment etches the metal surface providing a structure suitable for the application of adherent electrodeposits and at the same time plates the surface with a thin protective film of iron.

  8. Method for preparation of uranium hydride

    International Nuclear Information System (INIS)

    Gorski, M.S.; Goncalves, Miriam; Mirage, A.; Lima, W. de.

    1985-01-01

    A method for preparation of Uranium Hydride starting from Hidrogen and Uranium is described. In the temperature range of 250 0 up to 350 0 C, and pressures above 10torr, Hydrogen reacts smoothly with Uranium turnings forming a fine black or dark gray powder (UH 3 ). Samples containing a significant amount of oxides show a delay before the reaction begging. (Author) [pt

  9. Yellowcake processing in uranium recovery

    International Nuclear Information System (INIS)

    Paul, J.M.

    1981-01-01

    This information relates to the recovery of uranium from uranium peroxide yellowcake produced by precipitation with hydrogen peroxide. The yellowcake is calcined at an elevated temperature to effect decomposition of the yellowcake to uranium oxide with the attendant evolution of free oxygen. The calcination step is carried out in the presence of a reducing agent which reacts with the free oxygen, thus retarding the evolution of chlorine gas from sodium chloride in the yellowcake. Suitable reducing agents include ammonia producing compounds such as ammonium carbonate and ammonium bicarbonate. Ammonium carbonate and/or ammonium bicarbonate may be provided in the eluant used to desorb the uranium from an ion exchange column

  10. Yellowcake processing in uranium recovery

    Energy Technology Data Exchange (ETDEWEB)

    Paul, J.M.

    1981-10-06

    This information relates to the recovery of uranium from uranium peroxide yellowcake produced by precipitation with hydrogen peroxide. The yellowcake is calcined at an elevated temperature to effect decomposition of the yellowcake to uranium oxide with the attendant evolution of free oxygen. The calcination step is carried out in the presence of a reducing agent which reacts with the free oxygen, thus retarding the evolution of chlorine gas from sodium chloride in the yellowcake. Suitable reducing agents include ammonia producing compounds such as ammonium carbonate and ammonium bicarbonate. Ammonium carbonate and/or ammonium bicarbonate may be provided in the eluant used to desorb the uranium from an ion exchange column.

  11. Hafnium carbide formation in oxygen deficient hafnium oxide thin films

    Energy Technology Data Exchange (ETDEWEB)

    Rodenbücher, C. [Forschungszentrum Jülich GmbH, Peter Grünberg Institute (PGI-7), JARA-FIT, 52425 Jülich (Germany); Hildebrandt, E.; Sharath, S. U.; Kurian, J.; Komissinskiy, P.; Alff, L. [Technische Universität Darmstadt, Institute of Materials Science, 64287 Darmstadt (Germany); Szot, K. [Forschungszentrum Jülich GmbH, Peter Grünberg Institute (PGI-7), JARA-FIT, 52425 Jülich (Germany); University of Silesia, A. Chełkowski Institute of Physics, 40-007 Katowice (Poland); Breuer, U. [Forschungszentrum Jülich GmbH, Central Institute for Engineering, Electronics and Analytics (ZEA-3), 52425 Jülich (Germany); Waser, R. [Forschungszentrum Jülich GmbH, Peter Grünberg Institute (PGI-7), JARA-FIT, 52425 Jülich (Germany); RWTH Aachen, Institute of Electronic Materials (IWE 2), 52056 Aachen (Germany)

    2016-06-20

    On highly oxygen deficient thin films of hafnium oxide (hafnia, HfO{sub 2−x}) contaminated with adsorbates of carbon oxides, the formation of hafnium carbide (HfC{sub x}) at the surface during vacuum annealing at temperatures as low as 600 °C is reported. Using X-ray photoelectron spectroscopy the evolution of the HfC{sub x} surface layer related to a transformation from insulating into metallic state is monitored in situ. In contrast, for fully stoichiometric HfO{sub 2} thin films prepared and measured under identical conditions, the formation of HfC{sub x} was not detectable suggesting that the enhanced adsorption of carbon oxides on oxygen deficient films provides a carbon source for the carbide formation. This shows that a high concentration of oxygen vacancies in carbon contaminated hafnia lowers considerably the formation energy of hafnium carbide. Thus, the presence of a sufficient amount of residual carbon in resistive random access memory devices might lead to a similar carbide formation within the conducting filaments due to Joule heating.

  12. Hydrotreatment activities of supported molybdenum nitrides and carbides

    Energy Technology Data Exchange (ETDEWEB)

    Dolce, G.M.; Savage, P.E.; Thompson, L.T. [University of Michigan, Ann Arbor, MI (United States). Dept. of Chemical Engineering

    1997-05-01

    The growing need for alternative sources of transportation fuels encourages the development of new hydrotreatment catalysts. These catalysts must be active and more hydrogen efficient than the current commercial hydrotreatment catalysts. Molybdenum nitrides and carbides are attractive candidate materials possessing properties that are comparable or superior to those of commercial sulfide catalysts. This research investigated the catalytic properties of {gamma}-Al{sub 2}O{sub 3}-supported molybdenum nitrides and carbides. These catalysts were synthesized via temperature-programmed reaction of supported molybdenum oxides with ammonia or methane/hydrogen mixtures. Phase constituents and compositions were determined by X-ray diffraction, elemental analysis, and neutral activation analysis. Oxygen chemisorption was used to probe the surface properties of the catalysts. Specific activities of the molybdenum nitrides and carbides were competitive with those of a commercial sulfide catalyst for hydrodenitrogenation (HDN), hydrodesulfurization (HDS), and hydrodeoxygenation (HDO). For HDN and HDS, the catalytic activity on a molybdenum basis was a strong inverse function of the molybdenum loading. Product distributions of the HDN, HDO and HDS of a variety of heteroatom compounds indicated that several of the nitrides and carbides were more hydrogen efficient than the sulfide catalyst. 35 refs., 8 figs., 7 tabs.

  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. Dependence of silicon carbide coating properties on deposition parameters: preliminary report

    International Nuclear Information System (INIS)

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

    1980-05-01

    Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain a layer of pyrolytic silicon carbide, which acts as a pressure vessel and provides containment of metallic fission products. The silicon carbide (SiC) is deposited by the thermal decomposition of methyltrichlorosilane (CH 3 SiCl 3 or MTS) in an excess of hydrogen. The purpose of the current study is to determine how the deposition variables affect the structure and properties of the SiC layer

  16. DC characteristics and parameters of silicon carbide high-voltage power BJTs

    International Nuclear Information System (INIS)

    Patrzyk, Joanna; Zarębski, Janusz; Bisewski, Damian

    2016-01-01

    The paper shows the static characteristics and operating parameters of the bipolar power transistors made of silicon carbide and for comparison their equivalents made of classical silicon technology. The characteristics and values of selected operating parameters with special emphasis on the effect of temperature and operating point of considered devices are discussed. Quantitative as well as qualitative differences between the characteristics of the transistor made of silicon and silicon carbide are indicated as well

  17. Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

    International Nuclear Information System (INIS)

    Carter, D.H.; Gibson, M.; King, D.C.; Marshall, J.; Puckett, B.J.; Richards, A.E.; Wass, T.; Wilson, D.J.

    1965-07-01

    The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO 2 /UO 2 fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author)

  18. Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Carter, D H; Gibson, M; King, D C; Marshall, J; Puckett, B J; Richards, A E; Wass, T; Wilson, D J [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1965-07-15

    The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO{sub 2}/UO{sub 2} fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author) 14 refs, 30 figs, 18 tabs

  19. Stored energy in irradiated silicon carbide

    Energy Technology Data Exchange (ETDEWEB)

    Snead, L.L.; Burchell, T.D. [Oak Ridge National Lab., TN (United States)

    1997-04-01

    This report presents a short review of the phenomenon of Wigner stored energy release from irradiated graphite and discusses it in relation to neutron irradiation of silicon carbide. A single published work in the area of stored energy release in SiC is reviewed and the results are discussed. It appears from this previous work that because the combination of the comparatively high specific heat of SiC and distribution in activation energies for recombining defects, the stored energy release of SiC should only be a problem at temperatures lower than those considered for fusion devices. The conclusion of this preliminary review is that the stored energy release in SiC will not be sufficient to cause catastrophic heating in fusion reactor components, though further study would be desirable.

  20. Neutron irradiation induced amorphization of silicon carbide

    International Nuclear Information System (INIS)

    Snead, L.L.; Hay, J.C.

    1998-01-01

    This paper provides the first known observation of silicon carbide fully amorphized under neutron irradiation. Both high purity single crystal hcp and high purity, highly faulted (cubic) chemically vapor deposited (CVD) SiC were irradiated at approximately 60 C to a total fast neutron fluence of 2.6 x 10 25 n/m 2 . Amorphization was seen in both materials, as evidenced by TEM, electron diffraction, and x-ray diffraction techniques. Physical properties for the amorphized single crystal material are reported including large changes in density (-10.8%), elastic modulus as measured using a nanoindentation technique (-45%), hardness as measured by nanoindentation (-45%), and standard Vickers hardness (-24%). Similar property changes are observed for the critical temperature for amorphization at this neutron dose and flux, above which amorphization is not possible, is estimated to be greater than 130 C

  1. Ordering effects in nonstoichiometric titanium carbide

    International Nuclear Information System (INIS)

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

    2000-01-01

    The effect of nonstoichiometry and ordering on crystalline structure and specific electric resistance (ρ) of TiC y (0.52≤y≤0.98) is studied within the temperature range of 300-1100 K. It is shown that the titanium carbide ordering in the areas 0.52≤y≤0.55, 0.56≤y≤0.58 and 0.62≤y≤0.68 leads to formation of the Ti 2 C cubic and trigonal ordered phase and the Ti 3 C 2 rhombic ordered phase correspondingly. Availability of hysteresis on the ρ(T) dependences in the area of the disorder-order reversible equilibrium transition points out to the fact that the TiC y ↔Ti 2 C and TiC y ↔Ti 3 C 2 transformations are the first order phase transitions [ru

  2. 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.

  3. Single Photon Sources in Silicon Carbide

    International Nuclear Information System (INIS)

    Brett Johnson

    2014-01-01

    Single photon sources in semiconductors are highly sought after as they constitute the building blocks of a diverse range of emerging technologies such as integrated quantum information processing, quantum metrology and quantum photonics. In this presentation, we show the first observation of single photon emission from deep level defects in silicon carbide (SiC). The single photon emission is photo-stable at room temperature and surprisingly bright. This represents an exciting alternative to diamond color centers since SiC possesses well-established growth and device engineering protocols. The defect is assigned to the carbon vacancy-antisite pair which gives rise to the AB photoluminescence lines. We discuss its photo-physical properties and their fabrication via electron irradiation. Preliminary measurements on 3C SiC nano-structures will also be discussed. (author)

  4. Effect of temperature on the expansion and microstructure Of U3 Si2-AI mini plate fuel of 3.6 g/cm3 uranium loading

    International Nuclear Information System (INIS)

    Ginting, A. Br.; Samosir, N.; Suparjo; Nasution, H.

    2000-01-01

    Expansion analysis has been conducted to 50 x 20-mm U 3 Si 2 -AI mini plate of 3.6 g/cm 3 uranium loading using dilatometer. The analysis was carried out at various temperatures of 170 o C, 350 o C and 550 o C in Argon medium with delay time 4 days. The result showed that the fuel plate was relatively stable with increasing of heating time but underwent significant expansion. Heating at 170 o C, 350 o C and 550 o C resulted in the expansion of the U 3 Si 2 -AI fuel plate of to 83-212 mum, 333-475 mum, and 433-724 mum with coefficient expansion of 24.2x10 -6 / o C - 24.3x10 -6 / o C, 25.5x10 -6 / o C - 26.2x10 -6 /'oC and 26.6 x 10 -6 / o C - 28.2 x 10 -6 / o C respectively. Microanalysis of the U 3 Si 2 -AI mini plate fuel with SEM-EDS upon heating at those temperature variation showed that microstructure change didn't occur at 170 o C, mean while interaction between AIMg2 cladding and the fuel meat appeared to take place at 350 o C and 550 o C. Data on the expansion and microstructure change of U 3 Si 2 -AI fuel plate upon heating are of great important for the manufacture/fabrication of research fuel plate to produce silicide fuel element for higher uranium loading. (author)

  5. Helium behaviour in implanted boron carbide

    Directory of Open Access Journals (Sweden)

    Motte Vianney

    2015-01-01

    Full Text Available When boron carbide is used as a neutron absorber in nuclear power plants, large quantities of helium are produced. To simulate the gas behaviour, helium implantations were carried out in boron carbide. The samples were then annealed up to 1500 °C in order to observe the influence of temperature and duration of annealing. The determination of the helium diffusion coefficient was carried out using the 3He(d,p4He nuclear reaction (NRA method. From the evolution of the width of implanted 3He helium profiles (fluence 1 × 1015/cm2, 3 MeV corresponding to a maximum helium concentration of about 1020/cm3 as a function of annealing temperatures, an Arrhenius diagram was plotted and an apparent diffusion coefficient was deduced (Ea = 0.52 ± 0.11 eV/atom. The dynamic of helium clusters was observed by transmission electron microscopy (TEM of samples implanted with 1.5 × 1016/cm2, 2.8 to 3 MeV 4He ions, leading to an implanted slab about 1 μm wide with a maximum helium concentration of about 1021/cm3. After annealing at 900 °C and 1100 °C, small (5–20 nm flat oriented bubbles appeared in the grain, then at the grain boundaries. At 1500 °C, due to long-range diffusion, intra-granular bubbles were no longer observed; helium segregates at the grain boundaries, either as bubbles or inducing grain boundaries opening.

  6. Rare-earth elements and uranium in high-temperature solutions from East Pacific Rise hydrothermal vent field (130N)

    International Nuclear Information System (INIS)

    Michard, A.; Albarede, F.; Michard, G.; Minster, J.F.; Charlou, J.L.

    1983-01-01

    The mobility of rare-earth elements (REE) and U during hydrothermal alteration of the basalts at spreading centres has long been a matter of concern because of its bearing on the evolution and recycling of the oceanic crust. Previous approaches to this problem have been indirect, through studies on altered dredged basalts or ophiolites. Sampling of hydrothermal vent waters from the East Pacific Rise (EPR) at 13 0 N is reported. It provides the first direct evidence of REE-enriched solutions which, however, leave the budget of these elements in the crust and the ocean rather unmodified. In constrast, uranium, like magnesium, is quantitatively taken up from the seawater during the hydrothermal process. (author)

  7. Novel fabrication of silicon carbide based ceramics for nuclear applications

    Science.gov (United States)

    Singh, Abhishek Kumar

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; carbon--carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous

  8. Process for recovering uranium

    Science.gov (United States)

    MacWood, G. E.; Wilder, C. D.; Altman, D.

    1959-03-24

    A process useful in recovering uranium from deposits on stainless steel liner surfaces of calutrons is presented. The deposit is removed from the stainless steel surface by washing with aqueous nitric acid. The solution obtained containing uranium, chromium, nickel, copper, and iron is treated with an excess of ammonium hydroxide to precipitnte the uranium, iron, and chromium and convert the nickel and copper to soluble ammonio complexions. The precipitated material is removed, dried and treated with carbon tetrachloride at an elevated temperature of about 500 to 600 deg C to form a vapor mixture of UCl/ sub 4/, UCl/sub 5/, FeCl/sub 3/, and CrCl/sub 4/. The UCl/sub 4/ is separated from this vapor mixture by selective fractional condensation at a temperature of about 500 to 400 deg C.

  9. Uranium dioxide calcining apparatus

    International Nuclear Information System (INIS)

    Cole, E.A.; Peterson, R.S.

    1978-01-01

    This invention relates to an improved continuous calcining apparatus for consistently and controllably producing from calcinable reactive solid compounds of uranium, such as ammonium diuranate, uranium dioxide (UO 2 ) having an oxygen to uranium ratio of less than 2.2. The apparatus comprises means at the outlet end of a calciner kiln for receiving hot UO 2 , means for cooling the UO 2 to a temperature of below 100 deg C and conveying the cooled UO 2 to storage or to subsequent UO 2 processing apparatus where it finally comes into contact with air, the means for receiving cooling and conveying being sealed to the outlet end of the calciner and being maintained full of UO 2 and so operable as to exclude atmospheric oxygen from coming into contact with any UO 2 which is at elevated temperatures where it would readily oxidize, without the use of extra hydrogen gas in said means. (author)

  10. Understanding the Irradiation Behavior of Zirconium Carbide

    International Nuclear Information System (INIS)

    Motta, Arthur; Sridharan, Kumar; Morgan, Dane; Szlufarska, Izabela

    2013-01-01

    Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450ee)C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC-based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory, (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800ee)C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation

  11. 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)

  12. Oxidation of mullite-zirconia-alumina-silicon carbide composites

    International Nuclear Information System (INIS)

    Baudin, C.; Moya, J.S.

    1990-01-01

    This paper reports the isothermal oxidation of mullite-alumina-zirconia-silicon carbide composites obtained by reaction sintering studied in the temperature interval 800 degrees to 1400 degrees C. The kinetics of the oxidation process was related to the viscosity of the surface glassy layer as well as to the crystallization of the surface film. The oxidation kinetics was halted to T ≤ 1300 degrees C, presumably because of crystallization

  13. Correlation for boron carbide helium release in fast reactors

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Pitner, A.L.

    1977-04-01

    An empirical helium correlation for the helium release from boron carbide has been developed. The correlation provides a good fit to the experimental data in the temperature range from 800 to 1350 0 K, and burnup levels up to 80 x 10 20 captures/cm 3 . The correlation has the capability of extrapolation to 2200 0 K (3500 0 F) and 200 x 10 20 captures/cm 3 . In this range the helium release rate will not exceed the generation rate

  14. An electrochemical process for the recycling of tungsten carbide scrap

    International Nuclear Information System (INIS)

    Johns, M.W.

    1984-01-01

    An account is given of the development of a number of designs for electrochemical cells, and the subsequent construction and operation of a vibrating-plate cell capable of oxidizing 15 kilograms of tungsten carbide a day to a crude tungstic acid precipitate, with similtaneous recovery of cobalt metal on the cathode. The effects on the process of the reagent concentration, temperature, current density, and cathode material are discussed

  15. The effect of low-temperature aging on the microstructure and deformation of uranium- 6 wt% niobium: An in-situ neutron diffraction study

    Energy Technology Data Exchange (ETDEWEB)

    Brown, D.W., E-mail: dbrown@lanl.gov [Material Science and Technology Division, Los Alamos National Laboratory, Los Alamos, NM, 87545 (United States); Bourke, M.A.M. [Material Science and Technology Division, Los Alamos National Laboratory, Los Alamos, NM, 87545 (United States); Clarke, A.J. [Department of Metallurgical and Materials Engineering, Colorado School of Mines, 1500 Illinois Street, Golden, CO, 80401 (United States); Field, R.D.; Hackenberg, R.E.; Hults, W.L. [Material Science and Technology Division, Los Alamos National Laboratory, Los Alamos, NM, 87545 (United States); Thoma, D.J. [Department of Materials Science and Engineering, University of Wisconsin Madison, Madison, WI, 3706 (United States)

    2016-12-01

    The mechanical properties of uranium-niobium alloys evolve with aging at relatively low temperatures due to subtle microstructural changes. In-situ neutron diffraction measurements during aging of a monoclinic U-6Nb alloy at temperatures to 573 K were performed to monitor these changes. Further, in-situ neutron diffraction studies during deformation of U-6Nb in the as-quenched state and after aging for two and eight hours at 473 K were completed to assess the influence of microstructural evolution on mechanical properties. With heating, large anisotropic changes in lattice parameter were observed followed by relaxation with time at the aging temperature. The lattice parameters return to nearly their initial values with cooling. The active plastic deformation mechanisms including, in order of occurrence, shape-memory de-twinning, mechanical twinning, and slip-mediated deformation do not change with prior aging. However, the resistance to motion of the as-quenched martensitic twin boundaries increases following aging, resulting in the observed increase in initial yield strength.

  16. 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.

  17. 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.)

  18. 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)

  19. 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?

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    Our current state-of-the-art X-ray diffraction experiments are primarily sensitive to the position of the uranium atom. While the uranium - low-Z element bond (such as U-H or U-F) changes under pressure and temperature the X-ray diffraction investigations do not reveal information about the bonding or the stoichiometry. Questions that can be answered by Raman spectroscopy are (i) whether the bonding strength changes under pressure, as observed by either blue- or red-shifted peaks of the Raman active bands in the spectrum and (ii) whether the low-Z element will eventually be liberated and leave the host lattice, i.e. do the fluorine, oxygen, or hydrogen atoms form dimers after breaking the bond to the uranium atom. Therefore Raman spectra were also collected in the range where those decomposition products would appear. Raman is particularly well suited to these types of investigations due to its sensitivity to trace amounts of materials. One challenge for Raman investigations of the uranium compounds is that they are opaque to visible light. They absorb the incoming radiation and quickly heat up to the point of decomposition. This has been dealt with in the past by keeping the incoming laser power to very low levels on the tens of milliWatt range consequently affecting signal to noise. Recent modern investigations also used very small laser spot sizes (micrometer range) but ran again into the problem of heating and chemical sensitivity to the environment. In the studies presented here (in contrast to all other studies that were performed at ambient conditions only) we employ micro-Raman spectroscopy of samples situated in a diamond anvil cell. This increases the trustworthiness of the obtained data in several key-aspects: (a) We surrounded the samples in the DAC with neon as a pressure transmitting medium, a noble gas that is absolutely chemically inert. (b) Through the medium the sample is thermally heat sunk to the diamond anvils, diamond of course possessing the

  1. Creep behaviour of a casting titanium carbide reinforced AlSi12CuNiMg piston alloy at elevated temperatures; Hochtemperaturkriechverhalten der schmelzmetallurgisch hergestellten dispersionsverstaerkten Kolbenlegierung AlSi12CuNiMg

    Energy Technology Data Exchange (ETDEWEB)

    Michel, S.; Scholz, A. [Zentrum fuer Konstruktionswerkstoffe, TU Darmstadt (Germany); Tonn, B. [Institut fuer Metallurgie, TU Clausthal (Germany); Zak, H.

    2012-03-15

    This paper deals with the creep behaviour of the titanium carbide reinforced AlSi12CuNiMg piston alloy at 350 C and its comparison to the conventional AlSi12Cu4Ni2MgTiZr piston alloy. With only 0,02 vol-% TiC reinforcement the creep strength and creep rupture strength of the AlSi12CuNiMg piston alloy are significantly improved and reach the level of the expensive AlSi12Cu4Ni2MgTiZr alloy. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  2. Advances in carbide fuel element development for fast reactor application

    International Nuclear Information System (INIS)

    Dienst, W.; Kleykamp, H.; Muehling, G.; Reiser, H.; Steiner, H.; Thuemmler, F.; Wedermeyer, H.; Weimar, P.

    1977-01-01

    The features of the carbide fuel development programme are reviewed and evaluated. Single pin and bundle irradiations are carried out under thermal, epithermal and fast flux conditions, the latter in the DFR and KNK-II reactors. Several fuel concepts in the region of representative SNR clad temperatures are compared by parameter and performance tests. A conservative concept is based on He-bonded 8 mm pins with (U,Pu)C pellets and a smear density of 75% TD, operating at 800 W/cm rod power and burnup to 70 MWd/kg. The preparation of mixed carbide fuels is carried out by carbothermic reduction of the oxides in different methods supported by equivalent carbon content, grain size and phase distribution analysis. The fuel for subassembly performance tests is produced in a pilot plant of 0,5 t/year capacity. Compatibility studies reveal that cladding carburization is the only chemical interaction with carbide fuels. This effect leads to a reduction in ductility of the stainless steel. Fission products apparently play no role in the compatibility behaviour. Comprehensive studies lead to reliable information on the chemical and thermodynamic state of the fuel under irradiation. The swelling of carbide fuels and the fission gas release are examined and analysed. Cladding plastic strain by fuel swelling occurs during steady-state operation because the irradiation creep is rather slow compared to oxide fuels. The cladding strain observed depends on the fuel porosity and the cladding strength. The development of carbide fuel pins is complemented by the application of comprehensive computer models. In addition to the steady-state tests power cycling and safety tests are under performance. Up to 1980 the results are summarized for the final design and specification. The development target of the present program is to fabricate several subassemblies for test operation in the SNR 300 by 1981

  3. Manganese in silicon carbide

    Energy Technology Data Exchange (ETDEWEB)

    Linnarsson, M.K., E-mail: marga@kth.se [Royal Institute of Technology, School of Information and Communication Technology, P.O. Box E229, SE-16440 Kista-Stockhom (Sweden); Hallen, A. [Royal Institute of Technology, School of Information and Communication Technology, P.O. Box E229, SE-16440 Kista-Stockhom (Sweden)

    2012-02-15

    Structural disorder and relocation of implanted Mn in semi-insulating 4H-SiC has been studied. Subsequent heat treatment of Mn implanted samples has been performed in the temperature range 1400-2000 Degree-Sign C. The depth distribution of manganese is recorded by secondary ion mass spectrometry. Rutherford backscattering spectrometry has been employed for characterization of crystal disorder. Ocular inspection of color changes of heat-treated samples indicates that a large portion of the damage has been annealed. However, Rutherford backscattering shows that after heat treatment, most disorder from the implantation remains. Less disorder is observed in the [0 0 0 1] channel direction compared to [112{sup Macron }3] channel direction. A substantial rearrangement of manganese is observed in the implanted region. No pronounced manganese diffusion deeper into the sample is recorded.

  4. Manganese in silicon carbide

    International Nuclear Information System (INIS)

    Linnarsson, M.K.; Hallén, A.

    2012-01-01

    Structural disorder and relocation of implanted Mn in semi-insulating 4H–SiC has been studied. Subsequent heat treatment of Mn implanted samples has been performed in the temperature range 1400–2000 °C. The depth distribution of manganese is recorded by secondary ion mass spectrometry. Rutherford backscattering spectrometry has been employed for characterization of crystal disorder. Ocular inspection of color changes of heat-treated samples indicates that a large portion of the damage has been annealed. However, Rutherford backscattering shows that after heat treatment, most disorder from the implantation remains. Less disorder is observed in the [0 0 0 1] channel direction compared to [112 ¯ 3] channel direction. A substantial rearrangement of manganese is observed in the implanted region. No pronounced manganese diffusion deeper into the sample is recorded.

  5. Study of elastic and thermodynamic properties of uranium dioxide under high temperature and pressure with density functional theory

    International Nuclear Information System (INIS)

    Zhou Mu; Wang Feng; Zheng Zhou; Liu Xiankun; Jiang Tao

    2013-01-01

    The elastic and thermodynamic properties of UO 2 under extreme physical condition are studied by using the density functional theory and quasi-harmonic Debye model. Results show that UO 2 is still stable ionic crystal under high temperatures, and pressures. Tetragonal shear constant is steady under high pressures and temperatures, while elastic constant C 44 is stable under high temperatures, but rises with pressure sharply. Bulk modulus, shear modulus and Young's modulus increase with pressure rapidly, but temperature would not cause evident debasement of the moduli, all of which indicate that UO 2 has excellent mechanical properties. Heat capacity of different pressures increases with temperature and is close to the Dulong-Petit limit near 1000 K. Debye temperature decreases with temperature, and increases with pressure. Under low pressure, thermal expansion coefficient raises with temperature rapidly, and then gets slow at higher pressure and temperature. Besides, the thermal expansion coefficient of UO 2 is much lower than that of other nuclear materials. (authors)

  6. Separation of uranium isotopes

    International Nuclear Information System (INIS)

    Porter, J.T.

    1980-01-01

    Methods and apparatus are disclosed for separation of uranium isotopes by selective isotopic excitation of photochemically reactive uranyl salt source material at cryogenic temperatures, followed by chemical separation of selectively photochemically reduced U+4 thereby produced from remaining uranyl source material

  7. Investigations on the conditions for obtaining high density boron carbide by sintering

    International Nuclear Information System (INIS)

    Kislyj, P.S.; Grabtschuk, B.L.

    1975-01-01

    The results of investigations on kinetics of condensation and mechanisms of mass transfer in the process of sintering of technical, chemically pure and synthesized boron carbide are generalized. Laws on boron carbide densification depending upon temperature, time of isothermic endurance, thermal speed, size of powder particles and variable composition in homogeneity are determined. From the results obtained on condensation kinetics and special experiments on studying the changes in properties after heating under different conditions, the role of dislocation and diffusion processes in mass transfer during boron carbide sintering is exposed. The properties of sintered boron carbide are 15-20% lower than the properties of high-pressed one, that is conditioned by intercrystallite distortion of the first one and transcrystallite of the second one

  8. Atom-vacancy ordering and magnetic susceptibility of nonstoichiometric hafnium carbide

    International Nuclear Information System (INIS)

    Gusev, A.I.; Zyryanova, A.N.

    1999-01-01

    Experimental results on magnetic susceptibility of nonstoichiometric hafnium carbide HfC y (0.6 0.71 , HfC 0.78 and HfC 0.83 in the range of 870-930 K the anomalies are revealed which are associated with superstructure short-range ordering in a non-metallics sublattice. It is shown that a short-range order in HfC 0.71 and HfC 0.78 carbides corresponds to Hf 3 C 2 ordered phase, and in HfC 0.83 carbide - to Hf 6 C 5 ordered phase. HfC 0.78 carbide is found to possesses zero magnetic susceptibility in temperature range 910-980 K [ru

  9. Silicon-Carbide Power MOSFET Performance in High Efficiency Boost Power Processing Unit for Extreme Environments

    Science.gov (United States)

    Ikpe, Stanley A.; Lauenstein, Jean-Marie; Carr, Gregory A.; Hunter, Don; Ludwig, Lawrence L.; Wood, William; Del Castillo, Linda Y.; Fitzpatrick, Fred; Chen, Yuan

    2016-01-01

    Silicon-Carbide device technology has generated much interest in recent years. With superior thermal performance, power ratings and potential switching frequencies over its Silicon counterpart, Silicon-Carbide offers a greater possibility for high powered switching applications in extreme environment. In particular, Silicon-Carbide Metal-Oxide- Semiconductor Field-Effect Transistors' (MOSFETs) maturing process technology has produced a plethora of commercially available power dense, low on-state resistance devices capable of switching at high frequencies. A novel hard-switched power processing unit (PPU) is implemented utilizing Silicon-Carbide power devices. Accelerated life data is captured and assessed in conjunction with a damage accumulation model of gate oxide and drain-source junction lifetime to evaluate potential system performance at high temperature environments.

  10. Corrosion behaviour of porous chromium carbide/oxide based ceramics in supercritical water

    International Nuclear Information System (INIS)

    Dong, Z.; Xin, T.; Chen, W.; Zheng, W.; Guzonas, D.

    2011-01-01

    Porous chromium carbide with a high density of open pores was fabricated by a reactive sintering method. Chromium oxide ceramics were obtained by re-oxidizing the porous chromium carbides formed. Some samples were added with yttria at 5 wt. %, prior to reactive sintering to form porous structures. Corrosion tests in SCW were performed at temperatures ranging from 375 o C to 625 o C with a fixed pressure at around 25∼30 MPa. The results show that chromium carbide is stable in SCW environments at temperatures up to 425 o C, above which disintegration of carbides through oxidation occurs. Porous chromium oxide samples show better corrosion resistance than porous chromium carbide, but disintegrate in SCW at around 625 o C. Among all the samples tested, chromium oxide ceramics with added yttria exhibited much better corrosion resistance compared with the pure chromium carbide/oxides. No evidence of weight change or disintegration of porous chromium oxides with 5 wt % added yttria was observed after exposure at 625 o C in SCW for 600 hours. (author)

  11. Mechanical-thermal synthesis of chromium carbides

    International Nuclear Information System (INIS)

    Cintho, Osvaldo Mitsuyuki; Favilla, Eliane Aparecida Peixoto; Capocchi, Jose Deodoro Trani

    2007-01-01

    The present investigation deals with the synthesis of chromium carbides (Cr 3 C 2 and Cr 7 C 3 ), starting from metallic chromium (obtained from the reduction of Cr 2 O 3 with Al) and carbon (graphite). The synthesis was carried out via high energy milling, followed by heat-treating of pellets made of different milled mixtures at 800 o C, for 2 h, under an atmosphere of argon. A SPEX CertPrep 8000 Mixer/Mill was used for milling under argon atmosphere. A tool steel vat and two 12.7 mm diameter chromium steel balls were used. The raw materials used and the products were characterized by differential thermal analysis, thermo gravimetric analysis, X-ray diffraction, electronic microscopy and X-ray fluorescence chemical analysis. The following variables were investigated: the quantity of carbon in the mixture, the milling time and the milling power. Mechanical activation of the reactant mixture depends upon the milling power ratio used for processing. The energy liberated by the reduction of the chromium oxide with aluminium exhibits a maximum for milling power ratio between 5:1 and 7.5:1. Self-propagating reaction occurred for all heat-treated samples whatever the carbon content of the sample and the milling power ratio used. Bearing carbon samples exhibited hollow shell structures after the reaction. The level of iron contamination of the milled samples was kept below 0.3% Fe. The self-propagated reaction caused high temperatures inside the samples as it may be seen by the occurrence of spherules, dendrites and whiskers. The carbon content determines the type of chromium carbide formed

  12. 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

  13. Stereological analysis of structure formation for solid WC-Co alloys in the process of carbide powder consolidation

    Energy Technology Data Exchange (ETDEWEB)

    Chernyavskij, K S

    1986-03-01

    Evolution of particle size distribution in carbide powders of different technological prehistory is studied in the process of their consolidation as a hard alloy. A successive estimate on identical preparations is used to study a structural powder->alloy transition. Temperature dependences of integral measures of the consolidated structure and characteristics of its heterogeneity are studied. It is shown that all studied structural rearrangements: formation of regular alternation of carbide and binding phases, development of particle-phase interfaces, change in size distribution - more intensely proceed in the high-temperature carbide base alloy.

  14. 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

  15. 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.)

  16. PRODUCTION OF URANIUM TUBING

    Science.gov (United States)

    Creutz, E.C.

    1958-04-15

    The manufacture of thin-walled uranium tubing by the hot-piercing techique is described. Uranium billets are preheated to a temperature above 780 d C. The heated billet is fed to a station where it is engaged on its external surface by three convex-surfaced rotating rollers which are set at an angle to the axis of the billet to produce a surface friction force in one direction to force the billet over a piercing mandrel. While being formed around the mandrel and before losing the desired shape, the tube thus formed is cooled by a water spray.

  17. 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.)

  18. 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)

  19. Synthesis of transfer-free graphene on cemented carbide surface.

    Science.gov (United States)

    Yu, Xiang; Zhang, Zhen; Liu, Fei; Ren, Yi

    2018-03-19

    Direct growth of spherical graphene with large surface area is important for various applications in sensor technology. However, the preparation of transfer-free graphene on different substrates is still a challenge. This study presents a novel approach for the transfer-free graphene growth directly on cemented carbide. The used simple thermal annealing induces an in-situ transformation of magnetron-sputtered amorphous silicon carbide films into the graphene matrix. The study reveals the role of Co, a binding phase in cemented carbides, in Si sublimation process, and its interplay with the annealing temperature in development of the graphene matrix. A detailed physico-chemical characterisation was performed by structural (XRD analysis and Raman spectroscopy with mapping studies), morphological (SEM) and chemical (EDS) analyses. The optimal bilayer graphene matrix with hollow graphene spheres on top readily grows at 1000 °C. Higher annealing temperature critically decreases the amount of Si, which yields an increased number of the graphene layers and formation of multi-layer graphene (MLG). The proposed action mechanism involves silicidation of Co during thermal treatment, which influences the existing chemical form of Co, and thus, the graphene formation and variations in a number of the formed graphene layers.

  20. Carbide coated fibers in graphite-aluminum composites

    Science.gov (United States)

    Imprescia, R. J.; Levinson, L. S.; Reiswig, R. D.; Wallace, T. C.; Williams, J. M.

    1975-01-01

    The NASA-supported program at the Los Alamos Scientific Laboratory (LASL) to develop carbon fiber-aluminum matrix composites is described. Chemical vapor deposition (CVD) was used to uniformly deposit thin, smooth, continuous coats of TiC on the fibers of graphite tows. Wet chemical coating of fibers, followed by high-temperature treatment, was also used, but showed little promise as an alternative coating method. Strength measurements on CVD coated fiber tows showed that thin carbide coats can add to fiber strength. The ability of aluminum alloys to wet TiC was successfully demonstrated using TiC-coated graphite surfaces. Pressure-infiltration of TiC- and ZrC-coated fiber tows with aluminum alloys was only partially successful. Experiments were performed to evaluate the effectiveness of carbide coats on carbon as barriers to prevent reaction between alluminum alloys and carbon. Initial results indicate that composites of aluminum and carbide-coated graphite are stable for long periods of time at temperatures near the alloy solidus.

  1. Lattice dynamics of α boron and of boron carbide

    International Nuclear Information System (INIS)

    Vast, N.

    1999-01-01

    The atomic structure and the lattice dynamics of α boron and of B 4 C boron carbide have been studied by Density Functional Theory (D.F.T.) and Density Functional Perturbation Theory (D.F.P.T.). The bulk moduli of the unit-cell and of the icosahedron have been investigated, and the equation of state at zero temperature has been determined. In α boron, Raman diffusion and infrared absorption have been studied under pressure, and the theoretical and experimental Grueneisen coefficients have been compared. In boron carbide, inspection of the theoretical and experimental vibrational spectra has led to the determination of the atomic structure of B 4 C. Finally, the effects of isotopic disorder have been modeled by an exact method beyond the mean-field approximation, and the effects onto the Raman lines has been investigated. The method has been applied to isotopic alloys of diamond and germanium. (author)

  2. Silicon carbide layer structure recovery after ion implantation

    International Nuclear Information System (INIS)

    Violin, Eh.E.; Demakov, K.D.; Kal'nin, A.A.; Nojbert, F.; Potapov, E.N.; Tairov, Yu.M.

    1984-01-01

    The process of recovery of polytype structure of SiC surface layers in the course of thermal annealing (TA) and laser annealing (LA) upon boron and aluminium implantation is studied. The 6H polytype silicon carbide C face (0001) has been exposed to ion radiation. The ion energies ranged from 80 to 100 keV, doses varied from 5x10 14 to 5x10 16 cm -2 . TA was performed in the 800-2000 K temperature range. It is shown that the recovery of the structure of silicon carbide layers after ion implantation takes place in several stages. Considerable effect on the structure of the annealed layers is exerted by the implantation dose and the type of implanted impurity. The recovery of polytype structure is possible only under the effect of laser pulses with duration not less than the time for the ordering of the polytype in question

  3. Electronic and vibrational hopping transport in boron carbides

    International Nuclear Information System (INIS)

    Emin, D.

    1991-01-01

    General concepts of hopping-type transport and localization are reviewed. Disorder, electronic correlations and atomic displacements, effects ignored in electronic band structure calculations, foster localization of electronic charge carriers. Examples are given that illustrate the efficacy of these effects in producing localization. This introduction is followed by a brief discussion of the relation between hopping-type transport and localization. The fundamentals of the formation, localization, and hopping transport of small polarons and/or bipolarons is then described. Electronic transport in boron carbides is presented as an example of the adiabatic hopping of small bipolarons. Finally, the notion of vibrational hopping is introduced. The high-temperature thermal diffusion in boron carbides is presented as a potential application of this idea

  4. Mechanism of protective action of surface carbide layers on titanium

    International Nuclear Information System (INIS)

    Chukalovskaya, T.V.; Chebotareva, N.P.; Tomashov, N.D.

    1990-01-01

    The protective action of surface carbide layer on titanium produced in methane atmosphere at 1000 deg C and under 6.7 kPa pressure in H 2 SO 4 solutions is studied through comparison of microsection metallographic specimens prior to and after corrosion testing (after specimen activation); through comparison of anodic characteristics after partial stripping of the layer up to its complete stripping; through analysis of the behaviour of Ti-TiC galvanic couple, and through investigation of corresponding corrosion diagrams under test conditions. It is shown that screening protective mechanism is primarily got involved in highly agressive media (high temperature and concentration of solution), and in less agressive environment the protection of titanium with carbide layer is primarily ensured by electrochemical mechanism

  5. METHOD OF SEPARATING URANIUM VALUES, PLUTONIUM VALUES AND FISSION PRODUCTS BY CHLORINATION

    Science.gov (United States)

    Brown, H.S.; Seaborg, G.T.

    1959-02-24

    The separation of plutonium and uranium from each other and from other substances is described. In general, the method comprises the steps of contacting the uranium with chlorine in the presence of a holdback material selected from the group consisting of lanthanum oxide and thorium oxide to form a uranium chloride higher than uranium tetrachloride, and thereafter heating the uranium chloride thus formed to a temperature at which the uranium chloride is volatilized off but below the volatilizalion temperature of plutonium chloride.

  6. Method for producing uranium atomic beam source

    International Nuclear Information System (INIS)

    Krikorian, O.H.

    1976-01-01

    A method is described for producing a beam of neutral uranium atoms by vaporizing uranium from a compound UM/sub x/ heated to produce U vapor from an M boat or from some other suitable refractory container such as a tungsten boat, where M is a metal whose vapor pressure is negligible compared with that of uranium at the vaporization temperature. The compound, for example, may be the uranium-rhenium compound, URe 2 . An evaporation rate in excess of about 10 times that of conventional uranium beam sources is produced

  7. 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...

  8. 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.

  9. 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.

  10. 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)

  11. 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

  12. Use of electrical imaging and distributed temperature sensing methods to characterize surface water–groundwater exchange regulating uranium transport at the Hanford 300 Area, Washington

    Science.gov (United States)

    Slater, Lee D.; Ntarlagiannis, Dimitrios; Day-Lewis, Frederick D.; Mwakanyamale, Kisa; Versteeg, Roelof J.; Ward, Andy; Strickland, Christopher; Johnson, Carole D.; Lane, John W.

    2010-01-01

    We explored the use of continuous waterborne electrical imaging (CWEI), in conjunction with fiber‐optic distributed temperature sensor (FO‐DTS) monitoring, to improve the conceptual model for uranium transport within the Columbia River corridor at the Hanford 300 Area, Washington. We first inverted resistivity and induced polarization CWEI data sets for distributions of electrical resistivity and polarizability, from which the spatial complexity of the primary hydrogeologic units was reconstructed. Variations in the depth to the interface between the overlying coarse‐grained, high‐permeability Hanford Formation and the underlying finer‐grained, less permeable Ringold Formation, an important contact that limits vertical migration of contaminants, were resolved along ∼3 km of the river corridor centered on the 300 Area. Polarizability images were translated into lithologic images using established relationships between polarizability and surface area normalized to pore volume (Spor). The FO‐DTS data recorded along 1.5 km of cable with a 1 m spatial resolution and 5 min sampling interval revealed subreaches showing (1) temperature anomalies (relatively warm in winter and cool in summer) and (2) a strong correlation between temperature and river stage (negative in winter and positive in summer), both indicative of reaches of enhanced surface water–groundwater exchange. The FO‐DTS data sets confirm the hydrologic significance of the variability identified in the CWEI and reveal a pattern of highly focused exchange, concentrated at springs where the Hanford Formation is thickest. Our findings illustrate how the combination of CWEI and FO‐DTS technologies can characterize surface water–groundwater exchange in a complex, coupled river‐aquifer system.

  13. Influence of quenching parameters in the carbides presence in the AISI M2 high speed steel

    International Nuclear Information System (INIS)

    Magalhaes, A.S.; Maria, G.G.B; Martins, S.C.S.; Lopes, W.; Correa, E.C.S.; Bezerra, A.C.S.

    2014-01-01

    The main characteristic of high speed steels, besides maintaining high hardness at room temperature, is the ability of retain hardness when subjected to high temperatures and high cutting speeds. The high percentage of alloying elements in these steels allows the development of complex carbides, acquiring a high hardness by heat treatment. The aim of this study is to evaluate the effects of quenching parameters in the volumetric fraction of carbides by semi-quantitative metallography and of retained austenite by X-ray diffraction. It has been observed that, in general, the increase in the soaking time and in the austenitizing temperature resulted in the reduction of the amount of carbides and in an increase in the amount of retained austenite in the martensitic matrix. (author)

  14. Thermodynamic Studies at Higher Temperatures of the Phase Relationships of Substoichiometric Plutonium and Uranium/Plutonium Oxides

    DEFF Research Database (Denmark)

    Sørensen, Ole Toft

    1976-01-01

    Partial molar thermodynamic quantities for oxygen in non-stoichiometric Pu and U/Pu oxides were determined by thermogravimetric measurements in CO/CO2 mixtures in the temperature range 900-1450°C. A detailed analysis of the thermodynamic data obtained, as well as data previously published...

  15. Vapor pressure and thermodynamics of beryllium carbide

    International Nuclear Information System (INIS)

    Rinehart, G.H.; Behrens, R.G.

    1980-01-01

    The vapor pressure of beryllium carbide has been measured over the temperature range 1388 to 1763 K using Knudsen-effusion mass spectrometry. Vaporization occurs incongruently according to the reaction Be 2 C(s) = 2Be(g) + C(s). The equilibrium vapor pressure above the mixture of Be 2 C and C over the experimental temperature range is (R/J K -1 mol -1 )ln(p/Pa) = -(3.610 +- 0.009) x 10 5 (K/T) + (221.43 +- 1.06). The third-law enthalpy change for the above reaction obtained from the present vapor pressures is ΔH 0 (298.15 K) = (740.5 +- 0.1) kJ mol -1 . The corresponding second-law result is ΔH 0 (298.15 K) = (732.0 +- 1.8) kJ mol -1 . The enthalpy of formation for Be 2 C(s) calculated from the present third-law vaporization enthalpy and the enthalpy of formation of Be(g) is ΔH 0 sub(f)(298.15 K) = -(92.5 +- 15.7) kJ mol -1 . (author)

  16. 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.

  17. Carbide precipitation kinetics in austenite of a Nb-Ti-V microalloyed steel

    International Nuclear Information System (INIS)

    Jung, Jae-Gil; Park, June-Soo; Kim, Jiyoung; Lee, Young-Kook

    2011-01-01

    Highlights: → Carbide precipitation kinetic was fastest at 950 deg. C and accelerated by strain. → Nucleation sites for (Nb,Ti)C above 950 deg. C were mainly undissolved (Ti,Nb)(C,N). → Strain enabled (Nb,Ti)C to nucleate on all sides of (Ti,Nb)(C,N) above 950 deg. C. → Strain changed nucleation sites from (Ti,Nb)(C,N) to dislocations below 900 deg. C. → Strain also accelerated the change in particle composition to equilibrium one. - Abstract: The isothermal precipitation kinetics of carbides in both strain-free and strained austenite (γ) of a microalloyed steel were quantitatively investigated through the electrical resistivity and transmission electron microscopy. The (Nb,Ti)C carbides at the interfaces of the undissolved (Ti,Nb)(C,N) carbonitrides were observed at all temperatures in strain-free γ. However, for strain-induced precipitation, above 950 deg. C, the precipitation of (Nb,Ti)C carbides near the undissolved (Ti,Nb)(C,N) carbonitrides was predominant due to the recrystallization of strained γ. Meanwhile, the fine (Nb,Ti,V)C carbides were homogeneously precipitated in non-recrystallized γ at 850 deg. C and 900 deg. C, as well as near the undissolved (Ti,Nb)(C,N) particles. The electrical resistivity method was successfully used to quantitatively measure the isothermal precipitation kinetics of carbides in both strain-free and strained γ. The precipitation-time-temperature diagrams of the carbide in strain-free and strained γ, with nose temperatures of 950 deg. C, were generated and the precipitation kinetics were greatly accelerated by the applied strain.

  18. 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

  19. Literature review of thermal and radiation performance parameters for high-temperature, uranium dioxide fueled cermet materials

    International Nuclear Information System (INIS)

    Haertling, C.; Hanrahan, R.J.

    2007-01-01

    High-temperature fissile-fueled cermet literature was reviewed. Data are presented primarily for the W-UO 2 as this was the system most frequently studied; other reviewed systems include cermets with Mo, Re, or alloys as a matrix. Failure mechanisms for the cermets are typically degradation of mechanical integrity and loss of fuel. Mechanical failure can occur through stresses produced from dissimilar expansion coefficients, voids created from diffusion of dissimilar materials or formation of metal hydride and subsequent volume expansion. Fuel loss failure can occur by high temperature surface vaporization or by vaporization after loss of mechanical integrity. Techniques found to aid in retaining fuel include the use of coatings around UO 2 fuel particles, use of oxide stabilizers in the UO 2 , minimizing grain sizes in the metal matrix, minimizing impurities, controlling the cermet sintering atmosphere, and cladding around the cermet

  20. Construction of an apparatus for measuring the low-temperature thermal conductivity before and after neutron irradiation. Application to uranium dioxide (1963); Realisation d'un appareil pour la mesure de la conductibilite thermique a basse temperature avant et apres irradiation neutronique. Application au dioxyde d'uranium (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-09-15

    An apparatus has been studied and built which makes it possible to alternatively irradiate a sample at room temperature in the reactor 'Melusine' at the Grenoble Nuclear Research Centre, and to measure its thermal conductivity between 20 and 100 deg. K in perfect safety. The results obtained on UO{sub 2} have made it possible on the one hand to check experimentally that the spin-phonon diffusion leads to a thermal resistance independent of temperature above 30 deg. K, and on the other hand to propose a simple theory which takes into count the role played by the damage due to U-235 fission products in the decrease of thermal conductivity after irradiation. (author) [French] Un appareil permettant alternativement d'irradier un echantillon a temperature ambiante dans le reacteur ''Melusine'' du C.E.N.G., et de mesurer sa conductibilite thermique entre 20 et 100 deg. K en toute securite, a ete etudie et construit Les resultats obtenus sur UO{sub 2} ont permis, d'une part, de verifier experimentalement que la diffusion spin-phonon conduit a une resistance thermique independante de la temperature au-dessus de 30 deg. K, et, d'autre part, de proposer une theorie simple tenant compte du role joue par les degats dus aux produits de fission de l'uranium 235, dans la deterioration de la conductibilite thermique apres irradiation. (auteur)