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

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.

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.)

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)

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.)

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.

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.

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

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)

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

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.)

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

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

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

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

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.

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)

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.

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

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

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

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.

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)

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...

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

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

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)

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

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.

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

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)

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)

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)

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.)

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)

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)

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.

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

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.

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.

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

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.

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...

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.

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

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

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)

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.

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

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

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)

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

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.

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

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.

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)

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

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

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.)

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.

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.

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.

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.

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)

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

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

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.

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)

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

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

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.

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)

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

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

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)

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

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

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.

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 ...

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.

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.

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

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

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

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.

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.

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.

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

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

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.

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)

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.

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

The composition and character of oxycarbide phase in uranium metal

International Nuclear Information System (INIS)

Liu Kezhao; Lai Xinchun; Yu Yong; Ni Ranfu

1999-08-01

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

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.)

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

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...

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.

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.

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.

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.

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

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)

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.

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

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.

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

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

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)

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.

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

International Nuclear Information System (INIS)

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

1999-10-01

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

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.

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

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

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

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

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.

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

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

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

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.

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.

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

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.

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.

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)

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)

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.

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)

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

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

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

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...

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)

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

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.

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.

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

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.

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

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

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

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)

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.

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

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

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.

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

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.

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.

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

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

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

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.

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.

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

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.

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.

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.)

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

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

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

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

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

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.

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

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.

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)

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)

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

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

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.

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

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)

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)

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

Composition and microstructure of beryllium carbide films prepared by thermal MOCVD

Energy Technology Data Exchange (ETDEWEB)

He, Yu-dan; Luo, Jiang-shan; Li, Jia; Meng, Ling-biao; Luo, Bing-chi; Zhang, Ji-qiang; Zeng, Yong; Wu, Wei-dong, E-mail: wuweidongding@163.com

2016-02-15

Highlights: • Non-columnar-crystal Be{sub 2}C films were firstly prepared by thermal MOCVD. • Beryllium carbide was always the dominant phase in the films. • α-Be and carbon existed in films deposited below and beyond 400 °C, respectively. • Morphology evolved with temperatures and no columnar grains were characterized. • The preferred substrate temperature for depositing high quality Be{sub 2}C films was 400 °C. - Abstract: Beryllium carbide films without columnar-crystal microstructures were prepared on the Si (1 0 0) substrate by thermal metal organic chemical vapor deposition using diethylberyllium as precursor. The influence of the substrate temperature on composition and microstructure of beryllium carbide films was systematically studied. Crystalline beryllium carbide is always the dominant phase according to XRD analysis. Meanwhile, a small amount of α-Be phase exists in films when the substrate temperature is below 400 °C, and hydrocarbon or amorphous carbon exists when the temperature is beyond 400 °C. Surfaces morphology shows transition from domes to cylinders, to humps, and to tetraquetrous crystalline needles with the increase of substrate temperature. No columnar grains are characterized throughout the thickness as revealed from the cross-section views. The average densities of these films are determined to be 2.04–2.17 g/cm{sup 3}. The findings indicate the substrate temperature has great influences on the composition and microstructure of the Be{sub 2}C films grown by thermal MOCVD.

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.

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

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

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)

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.

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)

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

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

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

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.

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

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.

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

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)

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)

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.

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

Effects of silicon carbide on the phase developments in mullite-carbon ceramic composite

Directory of Open Access Journals (Sweden)

Fatai Olufemi ARAMIDE

2017-12-01

Full Text Available The effects of the addition of silicon carbide and sintering temperatures on the phases developed, in sintered ceramic composite produced from kaolin and graphite was investigated. The kaolin and graphite of known mineralogical composition were thoroughly blended with 4 and 8 vol % silicon carbide. From the homogeneous mixture of kaolin, graphite and silicon carbide, standard samples were prepared via uniaxial compaction. The test samples produced were subjected to firing (sintering at 1300°C, 1400°C and 1500°C. The sintered samples were characterized for the developed phases using x‐ray diffractometry analysis, microstructural morphology using ultra‐high resolution field emission scanning electron microscope (UHRFEGSEM. It was observed that microstructural morphology of the samples revealed the evolution of mullite, cristobalite and microcline. The kaolinite content of the raw kaolin undergoes transformation into mullite and excess silica, the mullite and the silica phases contents increased with increased sintering temperature. It is also generally observed that the graphite content progressively reduced linearly with increased sintering temperature. It is concluded that silicon carbide acts as anti-oxidant for the graphite, this anti-oxidant effect was more effective at 4 vol % silicon carbide.

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

International Nuclear Information System (INIS)

Radulescu, H.

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

H. radulescu

2001-09-28

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Tungsten carbide and tungsten-molybdenum carbides as automobile exhaust catalysts

International Nuclear Information System (INIS)

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

1987-01-01

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

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)

Iron Carbides in Fischer–Tropsch Synthesis: Theoretical and Experimental Understanding in Epsilon-Iron Carbide Phase Assignment

International Nuclear Information System (INIS)

Liu, Xing-Wu; Cao, Zhi; Zhao, Shu; Gao, Rui

2017-01-01

As active phases in low-temperature Fischer–Tropsch synthesis for liquid fuel production, epsilon iron carbides are critically important industrial materials. However, the precise atomic structure of epsilon iron carbides remains unclear, leading to a half-century of debate on the phase assignment of the ε-Fe 2 C and ε’-Fe 2.2 C. Here, we resolve this decades-long question by a combining theoretical and experimental investigation to assign the phases unambiguously. First, we have investigated the equilibrium structures and thermal stabilities of ε-Fe x C, (x = 1, 2, 2.2, 3, 4, 6, 8) by first-principles calculations. We have also acquired X-ray diffraction patterns and Mössbauer spectra for these epsilon iron carbides, and compared them with the simulated results. These analyses indicate that the unit cell of ε-Fe 2 C contains only one type of chemical environment for Fe atoms, while ε’-Fe 2.2 C has six sets of chemically distinct Fe atoms.

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.)

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.

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.

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)

Additive-assisted synthesis of boride, carbide, and nitride micro/nanocrystals

International Nuclear Information System (INIS)

Chen, Bo; Yang, Lishan; Heng, Hua; Chen, Jingzhong; Zhang, Linfei; Xu, Liqiang; Qian, Yitai; Yang, Jian

2012-01-01

General and simple methods for the syntheses of borides, carbides and nitrides are highly desirable, since those materials have unique physical properties and promising applications. Here, a series of boride (TiB 2 , ZrB 2 , NbB 2 , CeB 6 , PrB 6 , SmB 6 , EuB 6 , LaB 6 ), carbide (SiC, TiC, NbC, WC) and nitride (TiN, BN, AlN, MgSiN 2 , VN) micro/nanocrystals were prepared from related oxides and amorphous boron/active carbon/NaN 3 with the assistance of metallic Na and elemental S. In-situ temperature monitoring showed that the reaction temperature could increase quickly to ∼850 °C, once the autoclave was heated to 100 °C. Such a rapid temperature increase was attributed to the intense exothermic reaction between Na and S, which assisted the formation of borides, carbides and nitrides. The as-obtained products were characterized by XRD, SEM, TEM, and HRTEM techniques. Results in this report will greatly benefit the future extension of this approach to other compounds. - Graphical abstract: An additive-assisted approach is successfully developed for the syntheses of borides, carbides and nitrides micro/nanocrystals with the assistance of the exothermic reaction between Na and S. Highlights: ► An additive-assisted synthesis strategy is developed for a number of borides, carbides and nitrides. ► The reaction mechanism is demonstrated by the case of SiC nanowires. ► The formation of SiC nanowires is initiated by the exothermic reaction of Na and S.

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

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

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)

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

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.

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

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.

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)

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

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)

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

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)

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

Process of making titanium carbide (TiC) nano-fibrous felts

Science.gov (United States)

Fong, Hao; Zhang, Lifeng; Zhao, Yong; Zhu, Zhengtao

2015-01-13

A method of synthesizing mechanically resilient titanium carbide (TiC) nanofibrous felts comprising continuous nanofibers or nano-ribbons with TiC crystallites embedded in carbon matrix, comprising: (a) electrospinning a spin dope for making precursor nanofibers with diameters less than 0.5 J.Lm; (b) overlaying the nanofibers to produce a nanofibrous mat (felt); and then (c) heating the nano-felts first at a low temperature, and then at a high temperature for making electrospun continuous nanofibers or nano-ribbons with TiC crystallites embedded in carbon matrix; and (d) chlorinating the above electrospun nano-felts at an elevated temperature to remove titanium for producing carbide derived carbon (CDC) nano-fibrous felt with high specific surface areas.

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.)

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.)

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.)

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

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)

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

Science.gov (United States)

Dolce, Gregory Martin

1997-11-01

A series of gamma-Alsb2Osb3 supported molybdenum nitrides and carbides were prepared by the temperature programmed reaction of supported molybdates with ammonia and methane/hydrogen mixtures, respectively. In the first part of this research, the effects of synthesis heating rates and molybdenum loading on the catalytic properties of the materials were examined. A significant amount of excess carbon was deposited on the surface of the carbides during synthesis. The materials consisted of small particles which were very highly dispersed. Oxygen chemisorption indicated that the nitride particles may have been two-dimensional. The dispersion of the carbides, however, appeared to decrease as the loading increased. The catalysts were evaluated for hydrodenitrogenation (HDN), hydrodesulfurization (HDS), and hydrodeoxygenation (HDO). The molybdenum loading had the largest effect on the activity of the materials. For the nitrides, the HDN and HDS activities were inverse functions of the loading. This suggested that the most active HDN and HDS sites were located at the perimeter of the two-dimensional particles. The HDN and HDS activities of the carbides followed the same trend as the oxygen uptake. This result suggested that oxygen titrated the active sites on the supported carbides. Selected catalysts were evaluated for methylcarbazole HDN, dibenzothiophene HDS, and dibenzofuran HDO. The activity and selectivity of the nitrides and carbides were competitive with a presulfided commercial catalyst. In the second part of this work, a series of supported nitrides and carbides were prepared using a wider range of loadings (5-30 wt% Mo). Thermogravimetric analysis was used to determine the temperature at which excess carbon was deposited on the carbides. By modifying the synthesis parameters, the deposition of excess carbon was effectively inhibited. The dispersions of the supported nitrides and carbides were constant and suggested that the materials consisted of two

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)

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

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

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.

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

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.

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

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

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.

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.

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.

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

Tungsten carbide nanoparticles as efficient cocatalysts for photocatalytic overall water splitting

KAUST Repository

Garcia Esparza, Angel T.

2012-12-17

Tungsten carbide exhibits platinum-like behavior, which makes it an interesting potential substitute for noble metals in catalytic applications. Tungsten carbide nanocrystals (≈5 nm) are directly synthesized through the reaction of tungsten precursors with mesoporous graphitic C3N 4 (mpg-C3N4) as the reactive template in a flow of inert gas at high temperatures. Systematic experiments that vary the precursor compositions and temperatures used in the synthesis selectively generate different compositions and structures for the final nanocarbide (W 2C or WC) products. Electrochemical measurements demonstrate that the WC phase with a high surface area exhibits both high activity and stability in hydrogen evolution over a wide pH range. The WC sample also shows excellent hydrogen oxidation activity, whereas its activity in oxygen reduction is poor. These tungsten carbides are successful cocatalysts for overall water splitting and give H2 and O2 in a stoichiometric ratio from H 2O decomposition when supported on a Na-doped SrTiO3 photocatalyst. Herein, we present tungsten carbide (on a small scale) as a promising and durable catalyst substitute for platinum and other scarce noble-metal catalysts in catalytic reaction systems used for renewable energy generation. Platinum replacement: The phase-controlled synthesis of tungsten carbide nanoparticles from the nanoconfinement of a mesoporous graphite C 3N4 (mpg-C3N4) reactive template is shown. The nanomaterials catalyze hydrogen evolution/oxidation reactions, but are inactive in the oxygen reduction reaction. Tungsten carbide is an effective cocatalyst for photocatalytic overall water splitting (see picture). Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Tungsten carbide nanoparticles as efficient cocatalysts for photocatalytic overall water splitting

KAUST Repository

Garcia Esparza, Angel T.; Cha, Dong Kyu; Ou, Yiwei; Kubota, Jun; Domen, Kazunari; Takanabe, Kazuhiro

2012-01-01

Tungsten carbide exhibits platinum-like behavior, which makes it an interesting potential substitute for noble metals in catalytic applications. Tungsten carbide nanocrystals (≈5 nm) are directly synthesized through the reaction of tungsten precursors with mesoporous graphitic C3N 4 (mpg-C3N4) as the reactive template in a flow of inert gas at high temperatures. Systematic experiments that vary the precursor compositions and temperatures used in the synthesis selectively generate different compositions and structures for the final nanocarbide (W 2C or WC) products. Electrochemical measurements demonstrate that the WC phase with a high surface area exhibits both high activity and stability in hydrogen evolution over a wide pH range. The WC sample also shows excellent hydrogen oxidation activity, whereas its activity in oxygen reduction is poor. These tungsten carbides are successful cocatalysts for overall water splitting and give H2 and O2 in a stoichiometric ratio from H 2O decomposition when supported on a Na-doped SrTiO3 photocatalyst. Herein, we present tungsten carbide (on a small scale) as a promising and durable catalyst substitute for platinum and other scarce noble-metal catalysts in catalytic reaction systems used for renewable energy generation. Platinum replacement: The phase-controlled synthesis of tungsten carbide nanoparticles from the nanoconfinement of a mesoporous graphite C 3N4 (mpg-C3N4) reactive template is shown. The nanomaterials catalyze hydrogen evolution/oxidation reactions, but are inactive in the oxygen reduction reaction. Tungsten carbide is an effective cocatalyst for photocatalytic overall water splitting (see picture). Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Yalina booster subcritical assembly performance with low enriched uranium fuel

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Yousry

2011-01-01

The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

Yalina booster subcritical assembly performance with low enriched uranium fuel

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

2011-07-01

The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

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.

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.

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)

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)

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

International Nuclear Information System (INIS)

1993-07-01

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

The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

International Nuclear Information System (INIS)

Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W.

1998-01-01

While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study, the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation

The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

Energy Technology Data Exchange (ETDEWEB)

Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W. [Los Alamos National Lab., NM (United States). Materials Science and Technology Div.

1998-12-31

While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation.

The influence of surface morphology and oxide microstructure on the nucleation and growth of uranium hydride on alpha uranium

Energy Technology Data Exchange (ETDEWEB)

Hanrahan, R.J. Jr.; Hawley, M.E.; Brown, G.W.

1998-12-31

While the bulk kinetics of the uranium-hydrogen reaction are well understood, the mechanisms underlying the initial nucleation of uranium hydride on uranium remain controversial. In this study, the authors have employed environmental cell optical microscopy, Scanning Electron Microscopy (SEM) and Atomic Force Microscopy, (AFM) in an attempt to relate the structure of the surface and the microstructure of the substrate with the susceptibility and site of hydride nucleation. Samples have been investigated with varying grain size, inclusion (carbide) concentration, and thermal history. There is a clear correlation to heat treatment immediately prior to hydrogen exposure. Susceptibility to hydride formation also appears to be related to impurities in the uranium. The oxidized surface is very complex, exhibiting wide variations in thickness and topography between samples, between grains in the same sample, and within individual grains. It is, however, very difficult to relate this fine scale variability to the relatively sparse hydride initiation sites. Therefore, the surface oxide layer itself does not appear to control the sites where hydride attack is initiated, although it must play a role in the induction period prior to hydride initiation.

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.

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)

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.

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.

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

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

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

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)

Investigation of the fire at the Uranium Enrichment Laboratory. Analysis of samples and pressurization experiment/analysis of container

International Nuclear Information System (INIS)

Akabori, Mitsuo; Minato, Kazuo; Watanabe, Kazuo

1998-05-01

To investigate the cause of the fire at the Uranium Enrichment Laboratory of the Tokai Research Establishment on November 20, 1997, samples of uranium metal waste and scattered residues were analyzed. At the same time the container lid that had been blown off was closely inspected, and the pressurization effects of the container were tested and analyzed. It was found that 1) the uranium metal waste mainly consisted of uranium metal, carbides and oxides, whose relative amounts were dependent on the particle size, 2) the uranium metal waste hydrolyzed to produce combustible gases such as methane and hydrogen, and 3) the lid of the outer container could be blown off by an explosive rise of the inner pressure caused by combustion of inflammable gas mixture. (author)

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

Effect of Ion Beam Irradiation on Silicon Carbide with Different Microstructures

International Nuclear Information System (INIS)

Park, Kyeong Hwan; Park, Ji Yeon; Kim, Weon Ju; Jung, Choong Hwan; Ryu, Woo Seog

2006-01-01

SiC and SiC/SiC composites are one of promising candidates for structural materials of the next generation energy systems such as the gas-cooled reactors and fusion reactors. This anticipation yields many material issues, and radiation effects of silicon carbide are recognized as an important research subject. Silicon carbide has diverse crystal structures (called polytypes), such as α-SiC (hexagonal structure), β-SiC (cubic structure) and amorphous SiC. Among these polytypes, β-SiC has been studied as matrix material in SiC/SiC composites. Near-stoichiometric β-SiC with high crystallinity and purity is considered as suitable material in the next generation energy system and matrix material in SiC/SiC composites because of its excellent radiation resistance. Highly pure and crystalline β-SiC and SiC/SiC composites could be obtained by the chemical vapor deposition (CVD) and Infiltration (CVI) process using a gas mixture of methyltrichlorosilane (CH 3 SiCl 3 , MTS) and purified H 2 . SiC produced by the CVD method has different grain size and microstructural morphology depended on the process conditions such as temperature, pressure and the input gas ratio. In this work, irradiation effects of silicon carbide were investigated using ion beam irradiation with emphasis on the influence of grain size and grain boundary. MeV ion irradiation at low temperature makes amorphous phase in silicon carbide. The microstructures and mechanical property changes of silicon carbide with different structures were analyzed after ion beam irradiation

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

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

Science.gov (United States)

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

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

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

Science.gov (United States)

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

1982-01-01

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

On electronic structure of polymer-derived amorphous silicon carbide ceramics

Science.gov (United States)

Wang, Kewei; Li, Xuqin; Ma, Baisheng; Wang, Yiguang; Zhang, Ligong; An, Linan

2014-06-01

The electronic structure of polymer-derived amorphous silicon carbide ceramics was studied by combining measurements of temperature-dependent conductivity and optical absorption. By comparing the experimental results to theoretical models, electronic structure was constructed for a carbon-rich amorphous silicon carbide, which revealed several unique features, such as deep defect energy level, wide band-tail band, and overlap between the band-tail band and defect level. These unique features were discussed in terms of the microstructure of the material and used to explain the electric behavior.

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.

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.

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.

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

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

Energy Technology Data Exchange (ETDEWEB)

Zalkind, S., E-mail: shimonzl@nrcn.org.il; Rafailov, G.; Halevy, I.; Livneh, T.; Rubin, A.; Maimon, H.; Schweke, D.

2017-03-15

The oxidation kinetics of U-0.1 wt%Cr at oxygen pressures of 150 Torr and the temperature range of 90–150 °C was studied by means of in-situ X-ray diffraction (XRD). A “breakaway” in the oxidation kinetics is found at ∼0.25 μm, turning from a parabolic to a linear rate law. At the initial stage of oxidation the growth plane of UO{sub 2}(111) is the prominent one. As the oxide thickens, the growth rate of UO{sub 2}(220) plane increases and both planes grow concurrently. The activation energies obtained for the oxide growth are Q{sub parabolic} = 17.5 kcal/mol and Q{sub linear} = 19 kcal/mol. Enhanced oxidation around uranium carbide (UC) inclusions is clearly observed by scanning electron microscopy (SEM).

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)

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

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.

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.

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)

Experimental investigation and thermodynamic modeling of molybdenum and vanadium-containing carbide hardened iron-based alloys

International Nuclear Information System (INIS)

Cabrol, E.; Bellot, C.; Lamesle, P.; Delagnes, D.; Povoden-Karadeniz, E.

2013-01-01

Highlights: ► Improvement of a carbide selective extraction method. ► Determination of experimental data on the Fe–C–Cr–Mo–V system for carbides above 900 °C: crystallographic structures and compositions of precipitates, matrix composition. ► High molybdenum solubility in FCC carbides. ► Improvement of thermodynamic databases from experimental results. ► Validation of the optimized database with different compositions steels. -- Abstract: A technique for the microstructural study of steels, based on the use of matrix dissolution to collect the very low number density precipitates formed in martensitic steels, has been considerably improved. This technique was applied to two different grades of alloy, characterized by high nickel and cobalt contents and varying chromium, molybdenum and vanadium contents. The technique was implemented at temperatures ranging between 900 °C and 1000 °C, in order to accurately determine experimental data including the crystallographic structure and chemical composition of the carbides, the carbide solvus temperatures, and variations in the chemical composition of the matrix. These experimental investigations reveal that the solubility of molybdenum in FCC carbides can be very high. These results have been compared with the behavior predicted by computational thermodynamics, and used to evaluate and improve the thermodynamic Matcalc steel database. This upgraded database has been validated on three other steels with different chemical compositions, characterized by the same Fe–Cr–Mo–V–C system

Experimental investigation and thermodynamic modeling of molybdenum and vanadium-containing carbide hardened iron-based alloys

Energy Technology Data Exchange (ETDEWEB)

Cabrol, E., E-mail: ecabrol@mines-albi.fr [Institut Clément Ader, Mines Albi, Campus Jarlard, F-81013 Albi Cedex 09 (France); Aubert and Duval, BP1 F-63770 Les Ancizes (France); Bellot, C. [Institut Clément Ader, Mines Albi, Campus Jarlard, F-81013 Albi Cedex 09 (France); Aubert and Duval, BP1 F-63770 Les Ancizes (France); Lamesle, P.; Delagnes, D. [Institut Clément Ader, Mines Albi, Campus Jarlard, F-81013 Albi Cedex 09 (France); Povoden-Karadeniz, E. [Christian Doppler Laboratory for Early Stages of Precipitation, Vienna University of Technology, Favoritenstrasse 9-11, A-1040 Vienna (Austria)

2013-04-15

Highlights: ► Improvement of a carbide selective extraction method. ► Determination of experimental data on the Fe–C–Cr–Mo–V system for carbides above 900 °C: crystallographic structures and compositions of precipitates, matrix composition. ► High molybdenum solubility in FCC carbides. ► Improvement of thermodynamic databases from experimental results. ► Validation of the optimized database with different compositions steels. -- Abstract: A technique for the microstructural study of steels, based on the use of matrix dissolution to collect the very low number density precipitates formed in martensitic steels, has been considerably improved. This technique was applied to two different grades of alloy, characterized by high nickel and cobalt contents and varying chromium, molybdenum and vanadium contents. The technique was implemented at temperatures ranging between 900 °C and 1000 °C, in order to accurately determine experimental data including the crystallographic structure and chemical composition of the carbides, the carbide solvus temperatures, and variations in the chemical composition of the matrix. These experimental investigations reveal that the solubility of molybdenum in FCC carbides can be very high. These results have been compared with the behavior predicted by computational thermodynamics, and used to evaluate and improve the thermodynamic Matcalc steel database. This upgraded database has been validated on three other steels with different chemical compositions, characterized by the same Fe–Cr–Mo–V–C system.

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

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.

Carbides precipitated from the melt in a Zr-2.5 Nb alloy

International Nuclear Information System (INIS)

Piotrkowski, R.; Garcia, E.A.; Vigna, G.L.; Bermudez, S.E.

1993-01-01

An experimental method is presented which leads to the formation of carbides similar in size (3 to 8 microns) and composition to those observed in some pressure tubes of CANDU type reactors. The method is based on melting the Zr-2.5 Nb alloy in a graphite crucible, where isothermal C diffusion in the Zr-Nb melt took place. It can be inferred that the carbides observed in pressure tubes could be originated in high temperature stages of the manufacture process. Otherwise, they could have been incorporated in the Zr sponge. As a result of the diffusion couple Liquid Zr-2.5 Nb/Solid Graphite, a carbide layer, up to 100μm thick, grew attached to the crucible wall, together with carbide particles whose size was in the some microns range. The smallest particles were arranged in rows determined by the prior β phase grains. The main carbide phase detected was the cubic MC 1-x ; the hexagonal M 2 C was also detected; M for metal. (Author)

Carbide Coatings for Nickel Alloys, Graphite and Carbon/Carbon Composites to be used in Fluoride Salt Valves

Energy Technology Data Exchange (ETDEWEB)

Nagle, Denis [Johns Hopkins Univ., Baltimore, MD (United States); Zhang, Dajie [Johns Hopkins Univ., Baltimore, MD (United States)

2015-10-22

The focus of this research was concerned with developing materials technology that supports the evolution of Generation IV Advanced High Temperature Reactor (AHTR) concepts. Specifically, we investigate refractory carbide coatings for 1) nickel alloys, and 2) commercial carbon-carbon composites (CCCs). Numerous compelling reasons have driven us to focus on carbon and carbide materials. First, unlike metals, the strength and modulus of CCCs increase with rising temperature. Secondly, graphite and carbon composites have been proven effective for resisting highly corrosive fluoride melts such as molten cryolite [Na₃AlF₆] at ~1000°C in aluminum reduction cells. Thirdly, graphite and carbide materials exhibit extraordinary radiation damage tolerance and stability up to 2000°C. Finally, carbides are thermodynamically more stable in liquid fluoride salt than the corresponding metals (i.e. Cr and Zr) found in nickel based alloys.

The effect of fiber microstructure on evolution of residual stresses in silicon carbide/titanium aluminide composites

Science.gov (United States)

Pindera, Marek-Jerzy; Freed, Alan D.

1992-01-01

This paper examines the effect of the morphology of the SCS6 silicon carbide fiber on the evolution of residual stresses in SiC/Ti composites. A micromechanics model based on the concentric cylinder concept is presented which is used to calculate residual stresses in a SiC/Ti composite during axisymmetric cooling by a spatially uniform temperature change. The silicon carbide fiber is modeled as a layered material with five distinct transversely isotropic and orthotropic, elastic layers, whereas the titanium matrix is taken to be isotropic, with temperature-dependent elastoplastic properties. The results arc compared with those obtained based on the assumption that the silicon carbide fiber is isotropic and homogeneous.

Multilayered and composite PVD-CVD coatings in cemented carbides manufacture

International Nuclear Information System (INIS)

Glushkov, V.N.; Anikeev, A.I.; Anikin, V.N.; Vereshchaka, A.S.

2001-01-01

Carbide cutting tools with wear-resistant coatings deposited by CVD process are widely employed in mechanical engineering to ensure a substantially longer service life of tool systems. However, the relatively high temperature and long time of the process make the substrate decarburise and, as a result, the bend strength and performance characteristics of a tool decrease. The present study suggests the problem of deteriorated strength of CVD-coated carbide tools be solved by the development of a technology that combines arc-PVD and CVD processes to deposit multilayered coatings of titanium and aluminium compounds. (author)

Plasma spraying of zirconium carbide – hafnium carbide – tungsten cermets

Czech Academy of Sciences Publication Activity Database

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

2009-01-01

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

A study on the development of silicon carbide materials for nuclear application

International Nuclear Information System (INIS)

Won, Dong Yeon; Kim, Chan Jung; Lee, Jae Choon; Kim, Joon Hyung; Lim, Kyung Soo; Kim, Ki Baik

1987-12-01

Silicon carbide was synthesized by reaction sintering process from carbon and silicon powders as starting materials. The effects of two processing parameters, i.e., heat treatment time and temperature, were examined (to characterize the reaction sintering process) in terms of the degree of reaction and phase developed during heat treatment. The final products after reaction of silicon and carbon powders were identified as β-SiC having ZnS crystal structure. Sintering of cordierite ceramics which was used as an high temperature inorganic binder to fabricate ceramically bound silicon carbide, and phase identification of the sintered ceramics by X-ray powder diffraction techniques. (Author)

Three-point bending fatigue behavior of WC–Co cemented carbides

International Nuclear Information System (INIS)

Li, Anhai; Zhao, Jun; Wang, Dong; Gao, Xinliang; Tang, Hongwei

2013-01-01

Highlights: ► Mechanical fatigue tests were conducted on a specific designed jig. ► Three-point bending fatigue behavior of WC–Co cemented carbides was studied. ► Fatigue mechanisms of WC–Co cemented carbides with different WC grain sizes and Co binder contents were revealed. -- Abstract: WC–Co cemented carbides with different WC grain sizes and Co binder contents were sintered and fabricated. The three-point bending specimens with a single edge notch were prepared for tests. In the experiments, the mechanical properties of materials were investigated under static and cyclic loads (20 Hz) in air at room temperature. The fatigue behaviors of the materials under the same applied loading conditions are presented and discussed. Optical microscope and scanning electron microscopy were used to investigate the micro-mechanisms of damage during fatigue, and the results were used to correlate with the mechanical fatigue behavior of WC–Co cemented carbides. Experimental results indicated that the fatigue fracture surfaces exhibited more fracture origins and diversification of crack propagation paths than the static strength fracture surfaces. The fatigue fracture typically originates from inhomogeneities or defects such as micropores or aggregates of WC grains near the notch tip. Moreover, due to the diversity and complexity of the fatigue mechanisms, together with the evolution of the crack tip and the ductile deformation zone, the fatigue properties of WC–Co cemented carbides were largely relevant with the combination of transverse rupture strength and fracture toughness, rather than only one of them. Transverse rupture strength dominated the fatigue behavior of carbides with low Co content, whilst the fatigue behavior of carbides with high Co content was determined by fracture toughness.

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.

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.

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

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)

Study of reactions between uranium-plutonium mixed oxide and uranium nitride and between uranium oxide and uranium nitride; Etude des reactions entre l`oxyde mixte d`uranium-plutonium et le nitrure d`uranium et entre l`oxyde d`uranium et le nitrure d`uranium

Energy Technology Data Exchange (ETDEWEB)

Lecraz, C

1993-06-11

A new type of combustible elements which is a mixture of uranium nitride and uranium-plutonium oxide could be used for Quick Neutrons Reactors. Three different studies have been made on the one hand on the reactions between uranium nitride (UN) and uranium-plutonium mixed oxide (U,Pu)O{sub 2}, on the other hand on these between UN and uranium oxide UO{sub 2}. They show a sizeable reaction between nitride and oxide for the studied temperatures range (1573 K to 1973 K). This reaction forms a oxynitride compound, MO{sub x} N{sub y} with M=U or M=(U,Pu), whose crystalline structure is similar to oxide`s. Solubility of nitride in both oxides is studied, as the reaction kinetics. (TEC). 32 refs., 48 figs., 22 tabs.

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

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

Chlorination of uranium ore for extraction of uranium, thorium and radium and for pyrite removal

International Nuclear Information System (INIS)

Skeaf, J.M.

1979-01-01

The high-temperature chlorination of uranium ore was investigated. The objective was to develop a process which is both economically viable and environmentally acceptable. Test work was directed toward obtaining high extractions of uranium, thorium and radium-226, as well as iron, sulphur and the rare earths, and consists of chlorinating samples of an Elliot Lake uranium ore at elevated temperatures and repulping the resulting calcine in dilute hydrochloric acid. The effect of temperature and chlorine throughput on the extraction of the various metals was investigated. The best conditions yielded extractions of uranium, iron and sulphur (all as chlorides) greater than 95 percent. Chlorine consumption varied between 6 and 16 percent by weight of the ore charge. (author)

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

Cemented carbide cutting tool: Laser processing and thermal stress analysis

Energy Technology Data Exchange (ETDEWEB)

Yilbas, B.S. [Mechanical Engineering Department, KFUPM, Box 1913, Dhahran 31261 (Saudi Arabia)]. E-mail: bsyilbas@kfupm.edu.sa; Arif, A.F.M. [Mechanical Engineering Department, KFUPM, Box 1913, Dhahran 31261 (Saudi Arabia); Karatas, C. [Engineering Faculty, Hacettepe University, Ankara (Turkey); Ahsan, M. [Mechanical Engineering Department, KFUPM, Box 1913, Dhahran 31261 (Saudi Arabia)

2007-04-15

Laser treatment of cemented carbide tool surface consisting of W, C, TiC, TaC is examined and thermal stress developed due to temperature gradients in the laser treated region is predicted numerically. Temperature rise in the substrate material is computed numerically using the Fourier heating model. Experiment is carried out to treat the tool surfaces using a CO{sub 2} laser while SEM, XRD and EDS are carried out for morphological and structural characterization of the treated surface. Laser parameters were selected include the laser output power, duty cycle, assisting gas pressure, scanning speed, and nominal focus setting of the focusing lens. It is found that temperature gradient attains significantly high values below the surface particularly for titanium and tantalum carbides, which in turn, results in high thermal stress generation in this region. SEM examination of laser treated surface and its cross section reveals that crack initiation below the surface occurs and crack extends over the depth of the laser treated region.

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

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.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

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.)

Thermal evolution behavior of carbides and γ′ precipitates in FGH96 superalloy powder

International Nuclear Information System (INIS)

Zhang Lin; Liu Hengsan; He Xinbo; Rafi-ud-din; Qu Xuanhui; Qin Mingli; Li Zhou; Zhang Guoqing

2012-01-01

The characteristics of rapidly solidified FGH96 superalloy powder and the thermal evolution behavior of carbides and γ′ precipitates within powder particles were investigated. It was observed that the reduction of powder size and the increase of cooling rate had transformed the solidification morphologies of atomized powder from dendrite in major to cellular structure. The secondary dendritic spacing was measured to be 1.02–2.55 μm and the corresponding cooling rates were estimated to be in the range of 1.4 × 10 4 –4.7 × 10 5 K·s −1 . An increase in the annealing temperature had rendered the phase transformation of carbides evolving from non-equilibrium MC′ carbides to intermediate transition stage of M 23 C 6 carbides, and finally to thermodynamically stable MC carbides. The superfine γ′ precipitates were formed at the dendritic boundaries of rapidly solidified superalloy powder. The coalescence, growth, and homogenization of γ' precipitates occurred with increasing annealing temperature. With decreasing cooling rate from 650 °C·K −1 to 5 °C·K −1 , the morphological development of γ′ precipitates had been shown to proceed from spheroidal to cuboidal and finally to solid state dendrites. Meanwhile, a shift had been observed from dendritic morphology to recrystallized structure between 900 °C and 1050 °C. Moreover, accelerated evolution of carbides and γ' precipitates had been facilitated by the formation of new grain boundaries which provide fast diffusion path for atomic elements. - Highlights: ► Microstructural characteristic of FGH96 superalloy powder was investigated. ► The relation between microstructure, particle size, and cooling rate was studied. ► Thermal evolution behavior of γ′ and carbides in loose FGH96 powder was studied.

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.

Corrosion-electrochemical characteristics of oxide-carbide and oxide-nitride coatings formed by electrolytic plasma

International Nuclear Information System (INIS)

Tomashov, N.D.; Chukalovskaya, T.V.; Medova, I.L.; Duradzhi, V.N.; Plavnik, G.M.

1990-01-01

The composition, structure, microhardness and corrosion-electrochemical properties of oxide-carbide and oxide-nitride coatings on titanium in 5n H 2 SO 4 , 50 deg, produced by the method of chemical-heat treatment in electrolytic plasma, containing saturation components of nitrogen and carbon, were investigated. It is shown that the coatings produced have increased hardness, possess high corrosion resistance in sulfuric acid solution at increased temperature, as to their electrochemcial behaviour they are similar to titanium carbide and nitride respectively. It is shown that high corrosion resistance is ensured by electrochemical mechanism of the oxide-carbide and oxide-nitride coating protection

Elimination of eight interfering radioisotopes in the determination of uranium by activation analysis with epithermic neutrons

International Nuclear Information System (INIS)

Requejo, C.S.

1977-01-01

The total or parcial elimination interfering radioisotopes in activation analysis of uranium by epithermic neutrons, has been made. It was possible to determine uranium, after chemical separation, from samples of organic and mineral matrixes, which had mercury, selenium, bromine, antimony, gold, barium, molybden and tungsten. Mineral samples were analysed giving results between 0.2 to 5.0 ppm of uranium. The same mineral were ground in agate mortar and in tungsten carbide mill. In the first sample is has been found 0.2277 +- -+ 0.0474 ppm U. The second which had tungsten, at level of 150 ppm, after radiochemical separation, it has been found 0.2465+- -+0.0326 ppm U. These results are considered statistically the same [pt

Carbide Precipitation in 2.25 Cr-1 Mo Bainitic Steel: Effect of Heating and Isothermal Tempering Conditions

Science.gov (United States)

Dépinoy, Sylvain; Toffolon-Masclet, Caroline; Urvoy, Stéphane; Roubaud, Justine; Marini, Bernard; Roch, François; Kozeschnik, Ernst; Gourgues-Lorenzon, Anne-Françoise

2017-05-01

The effect of the tempering heat treatment, including heating prior to the isothermal step, on carbide precipitation has been determined in a 2.25 Cr-1 Mo bainitic steel for thick-walled applications. The carbides were identified using their amount of metallic elements, morphology, nucleation sites, and diffraction patterns. The evolution of carbide phase fraction, morphology, and composition was investigated using transmission electron microscopy, X-ray diffraction, as well as thermodynamic calculations. Upon heating, retained austenite into the as-quenched material decomposes into ferrite and cementite. M7C3 carbides then nucleate at the interface between the cementite and the matrix, triggering the dissolution of cementite. M2C carbides precipitate separately within the bainitic laths during slow heating. M23C6 carbides precipitate at the interfaces (lath boundaries or prior austenite grain boundaries) and grow by attracting nearby chromium atoms, which results in the dissolution of M7C3 and, depending on the temperature, coarsening, or dissolution of M2C carbides, respectively.

Role of intericosahedral chains on the hardness of sputtered boron carbide films

International Nuclear Information System (INIS)

Jacobsohn, L.G.; Averitt, R.D.; Wetteland, C.J.; Schulze, R.K.; Nastasi, M.; Daemen, L.L.; Jenei, Z.; Asoka-Kumar, P.

2004-01-01

The relationship between the structure and mechanical properties of sputter-deposited boron carbide films was investigated. Changes in the structure induced by annealing were characterized in terms of chemical composition, chemical bonding, and concentrations of defects and trapped impurities. The creation of intericosahedral chains for higher annealing temperatures was revealed by infrared and Raman measurements, and the intensity of the infrared band at 1500 cm-1 was found to be related to the hardness. The presence of residual trapped Ar atoms and of open-volume defects is insensitive to relatively high annealing temperatures and does not influence the recovery of the hardness. Our results suggest postdeposition annealing as a pathway to enhance the mechanical properties of boron carbide films

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

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....

Modification of NiAl intermetallic coatings processed by PTA with chromium carbides

International Nuclear Information System (INIS)

Yano, Diogo Henrique Sepel; Brunetti, Cristiano; Pintaude, Giuseppe; Oliveira, Ana Sofia Climaco Monteiro d'

2010-01-01

Equipment that operate under high-temperatures can be protected with NiAl intermetallic coatings mainly because of their metallurgical stability. This study as it evaluates the effect of chromium carbide added to Ni-Al intermetallic coatings processed by PTA. Three Ni-Al-Cr23C6 powder mixtures with different carbide fractions (15, 30 and 45 wt%) and another without carbides were deposited by PTA on an AISI 304 stainless steel plate, using two different current intensities (100 and 150A). Coatings were evaluated regarding the presence of welding defects, and resultant microstructures were characterized by X-ray diffraction and scanning electron microscopy. Vickers microhardness and EDS chemical composition were also determined. NiAl and Cr_7C_3 development was confirmed by X-ray diffraction analysis. A combination of NiAl/Cr-Fe-Ni phases was identified. The hardness was strongly related to the formed phases and their amounts. Besides presenting advances toward the development of coatings which can withstand severe operation conditions, the present study shows that PTA hardfacing is able to produce reinforced intermetallic coatings for high-temperature applications. (author)

Niobium carbide synthesis by solid-gas reaction using a rotating cylinder reactor

International Nuclear Information System (INIS)

Fontes, F.A.O.; Gomes, K.K.P.; Oliveira, S.A.; Souza, C.P.; Sousa, J.F.; Rio Grande do Norte Univ., Natal, RN

2004-01-01

A rotating cylinder reactor was designed for the synthesis of niobium carbide powders at 1173 K. Niobium carbide, NbC, was prepared by carbothermal reduction starting from commercial niobium pentoxide powders. The reactor was heated using a custom-made, two-part, hinged, electric furnace with programmable temperature control. The design and operational details of the reactor are presented. The longitudinal temperature gradient inside the reactor was determined. Total reaction time was monitored by a gas chromatograph equipped with an FID detector for determination of methane concentrations. The results show that time of reaction depended on rotation speed. NbC was also prepared in a static-bed alumina reactor using the same conditions as in the previous case. The niobium carbide powders were characterized by X-ray diffraction and compared with commercially available products. Morphological, particle size distribution and surface area analyses were obtained using SEM, LDPS and BET, respectively. Therefore, the present study offers a significant technological contribution to the synthesis of NbC powders in a rotating cylinder reactor. (author)

Development of casting techniques for uranium and uranium alloys

International Nuclear Information System (INIS)

Singh, S.P.

2003-01-01

The casting process concerning furnace set-up, mould temperatures, pouring temperatures, out gassing, post heating, casting recovery and crucible and mould clean-up is discussed. Some applications of casting theory can be made in practice, but experience in handling the metal is most valuable in the successful solution of a new problem. The casting of uranium alloys using induction stirring of the melt to promote homogeneity in the casting is described. A few remarks are made concerning safety aspects associated with the casting of uranium

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

Highly Efficient Optical Pumping of Spin Defects in Silicon Carbide for Stimulated Microwave Emission

Science.gov (United States)

Fischer, M.; Sperlich, A.; Kraus, H.; Ohshima, T.; Astakhov, G. V.; Dyakonov, V.

2018-05-01

We investigate the pump efficiency of silicon-vacancy-related spins in silicon carbide. For a crystal inserted into a microwave cavity with a resonance frequency of 9.4 GHz, the spin population inversion factor of 75 with the saturation optical pump power of about 350 mW is achieved at room temperature. At cryogenic temperature, the pump efficiency drastically increases, owing to an exceptionally long spin-lattice relaxation time exceeding one minute. Based on the experimental results, we find realistic conditions under which a silicon carbide maser can operate in continuous-wave mode and serve as a quantum microwave amplifier.

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.

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

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

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

Science.gov (United States)

Alconchel, Silvia; Sapiña, Fernando; Martínez, Eduardo

2004-08-21

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

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)

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

KAUST Repository

Wezendonk, Tim A.

2018-04-19

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

Irradiation and annealing of p-type silicon carbide

Energy Technology Data Exchange (ETDEWEB)

Lebedev, Alexander A.; Bogdanova, Elena V.; Grigor' eva, Maria V.; Lebedev, Sergey P. [A.F. Ioffe Physical-Technical Institute, St. Petersburg, 194021 (Russian Federation); Kozlovski, Vitaly V. [St. Petersburg State Polytechnic University, St. Petersburg, 195251 (Russian Federation)

2014-02-21

The development of the technology of semiconductor devices based on silicon carbide and the beginning of their industrial manufacture have made increasingly topical studies of the radiation hardness of this material on the one hand and of the proton irradiation to form high-receptivity regions on the other hand. This paper reports on a study of the carrier removal rate (V{sub d}) in p-6H-SiC under irradiation with 8 MeV protons and of the conductivity restoration in radiation- compensated epitaxial layers of various p-type silicon carbide polytypes. V{sub d} was determined by analysis of capacitance-voltage characteristics and from results of Hall effect measurements. It was found that the complete compensation of samples with the initial value of Na - Nd ≈ 1.5 × 10{sup 18} cm{sup −3} occurs at an irradiation dose of ∼1.1 × 10{sup 16} cm{sup −2}. It is shown that specific features of the sublimation layer SiC (compared to CVD layers) are clearly manifested upon the gamma and electron irradiation and are hardly noticeable under the proton and neutron irradiation. It was also found that the radiation-induced compensation of SiC is retained after its annealing at ≤1000°C. The conductivity is almost completely restored at T ≥ 1200°C. This character of annealing of the radiation compensation is independent of a silicon carbide polytype and the starting doping level of the epitaxial layer. The complete annealing temperatures considerably exceed the working temperatures of SiC-based devices. It is shown that the radiation compensation is a promising method in the technology of high-temperature devices based on SiC.

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

The development of the production process for the thorium/uranium dicarbide fuel kernels for the first charge of the Dragon Reactor

International Nuclear Information System (INIS)

Burnett, R.C.; Hankart, L.J.; Horsley, G.W.

1965-05-01

The development of methods of producing spheroidal sintered porous kernels of hyperstoichiometric thorium/uranium dicarbide solid solution from thorium/uranium monocarbide/carbon and thoria/urania/carbon powder mixes is described. The work has involved study of (i) Methods of preparing green kernels from UC/Th/C powder mixes using the rotary sieve technique. (ii) Methods of producing green kernels from UO2/Th02/C powder mixes using the planetary mill technique. (iii) The conversion by appropriate heat treatment of green kernels produced by both routes to sintered porous kernels of thorium/uranium carbide. (iv) The efficiency of the processes. (author)

Recovery and treatment of uranium from uranium-containing solution by liquid membrane emulsion technology

International Nuclear Information System (INIS)

Xia Liangshu; Zhou Yantong; Xiao Yiqun; Peng Anguo; Xiao Jingshui; Chen Wei

2014-01-01

The recovery and treatment of uranium from uranium-containing solution using liquid membrane emulsion (LME) technology were studied in this paper, which contained the best volume ratio of membrane materials, stirring speed during emulsion process, the conditions of extracting, such as temperature, pH, initial concentration of uranium. Moreover, the mechanism for extracting uranium was also discussed. The best experimental conditions of emulsifying were acquired. The volume fractions of P 204 and liquid paraffin are 0.1 and 0.05, the volume ratios of Span80 and sulphonated kerosene to P 204 are 0.06 and 0.79 respectively, stirring speed is controlled in 2 000 r/min, and the concentration of inner phase is 4 mol/L. The recovery rate of uranium is up to 99% through the LME extracted uranium for 0.5 h at pH 2.5 and room temperature when the initial concentration is less than 400 mg/L and the volume ratio is 5 between the uranium-containing waste water and LME. The calculation results of Gibbs free energy show that the reaction process is spontaneous. (authors)

Implementation Analysis of Cutting Tool Carbide with Cast Iron Material S45 C on Universal Lathe

Science.gov (United States)

Junaidi; hestukoro, Soni; yanie, Ahmad; Jumadi; Eddy

2017-12-01

Cutting tool is the tools lathe. Cutting process tool CARBIDE with Cast Iron Material Universal Lathe which is commonly found at Analysiscutting Process by some aspects numely Cutting force, Cutting Speed, Cutting Power, Cutting Indication Power, Temperature Zone 1 and Temperatur Zone 2. Purpose of this Study was to determine how big the cutting Speed, Cutting Power, electromotor Power,Temperatur Zone 1 and Temperatur Zone 2 that drives the chisel cutting CARBIDE in the Process of tur ning Cast Iron Material. Cutting force obtained from image analysis relationship between the recommended Component Cuting Force with plane of the cut and Cutting Speed obtained from image analysis of relationships between the recommended Cutting Speed Feed rate.

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...

Characterization and performances of cobalt-tungsten and molybdenum-tungsten carbides as anode catalyst for PEFC

International Nuclear Information System (INIS)

Izhar, Shamsul; Yoshida, Michiko; Nagai, Masatoshi

2009-01-01

The preparation of carbon-supported cobalt-tungsten and molybdenum-tungsten carbides and their activity as an anode catalyst for a polymer electrolyte fuel cell were investigated. The electrocatalytic activity for the hydrogen oxidation reaction over the catalysts was evaluated using a single-stack fuel cell and a rotating disk electrode. The characterization of the catalysts was performed by XRD, temperature-programmed carburization, temperature-programmed reduction and X-ray photoelectron spectroscopy. The maximum power densities of the 30 wt% 873 K-carburized cobalt-tungsten and molybdenum-tungsten mixed with Ketjen carbon (cobalt-tungsten carbide (CoWC)/Ketjen black (KB) and molybdenum-tungsten carbide (MoWC)/KB) were 15.7 and 12.0 mW cm -2 , respectively, which were 14 and 11%, compared to the in-house membrane electrode assembly (MEA) prepared from a 20 wt% Pt/C catalyst. The CoWC/KB catalyst exhibited the highest maximum power density compared to the MoWC/KB and WC/KB catalysts. The 873 K-carburized CoW/KB catalyst formed the oxycarbided and/or carbided CoW that are responsible for the excellent hydrogen oxygen reaction

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

Preparation of Autoclaved Foamed Concrete Block from Fly Ash and Carbide Slag

Directory of Open Access Journals (Sweden)

Tan Xing

2018-01-01

Full Text Available To achieve the comprehensive utilization of solid waste and reduce costs, fly ash, carbide slag, and low-clinker cement were used to produce lightweight foamed concrete block. Granulated blast-furnace slag (GBFS was used as composition correction material in the block. The effects of curing temperature and dosage of low-clinker cement on the performance of foamed concrete block were investigated. The optimal material proportioning is obtained: fly ash 58.5%, carbide slag 20%, GBFS 10%, gypsum 1.5% and low-clinker cement 10%. The proper curing regime is “temperature rising 4h-180°C constant temperature 4h-natural cooling”. The results indicate that the compressive strength of the block reaches 3.55 MPa while the density is 616.9 kg/m3. The performance of the product meets JC/T 1062-2007 (China professional standard of foamed concrete block.

Microstructure of reactive synthesis TiC/Cr18Ni8 stainless steel bonded carbides

Institute of Scientific and Technical Information of China (English)

Jiang Junsheng; Liu Junbo; Wang Limei

2008-01-01

TiC/Cr18Ni8 steel bonded carbides were synthesized by vacuum sintering with mixed powders of iron, ferrotitanium, ferrochromium, colloidal graphite and nickel as raw materials. The microstructure and microhardness of the steel bonded carbides were analyzed by scanning electron microscope (SEM),X-ray diffraction (XRD) and Rockwell hardometer. Results show that the phases of steel bonded carbides mainly consist of TiC and Fe-Cr-Ni solid solution. The synthesized TiC particles are fine. Most of them are not more than 1 μm With the increase of sintering temperature, the porosity of TiC/Cr18Ni8 steel bonded carbides decreases and the density and hardness increase, but the size of TiC panicles slightly increases. Under the same sintering conditions, the density and hardness of steel bonded carbides with C/Ti atomic ratio 0.9 are higher than those with C/Ti atomic ratio 1.0.The TiC particles with C/Ti atomic ratio 0.9 are much finer and more homogeneous.

Thermal stress relieving of dilute uranium alloys

International Nuclear Information System (INIS)

Eckelmeyer, K.H.

1981-01-01

The kinetics of thermal stress relieving of uranium - 2.3 wt % niobium, uranium - 2.0 wt % molybdenum, and uranium - 0.75 wt % titanium are reported and discussed. Two temperature regimes of stress relieving are observed. In the low temperature regime (T 0 C) the process appears to be controlled by an athermal microplasticity mechanism which can be completely suppressed by prior age hardening. In the high temperature regime (300 0 C 0 C) the process appears to be controlled by a classical diffusional creep mechanism which is strongly dependent on temperature and time. Stress relieving is accelerated in cases where it occurs simultaneously with age hardening. The potential danger of residual stress induced stress corrosion cracking of uranium alloys is discussed

Contribution to the thermodynamic study of the uranium-carbon system

International Nuclear Information System (INIS)

Jeanne, Francis

1972-01-01

This research thesis aims at illustrating, in the particular case of uranium carbides, the contribution which the elaboration of a thermodynamic model may have, as experiments provide reliable but diffuse results, and as the lack of quantitative knowledge delays the practical application of the studied materials. The objective is then to develop such models for the calculation, in any point of the phase diagram, of the value of activities of uranium and carbon. Thus, after the presentation of two models (pair energies independent on stoichiometry, varying pair energies with respect to stoichiometry), the author reports and discusses numerical applications, and compares the obtained results with experimental results and with those obtained by other theoretical models. A thermodynamic model of a U-C quadratic solid solution is then presented, applied and discussed

Irradiation damage in boron carbide: point defects, clusters and helium bubbles

International Nuclear Information System (INIS)

Stoto, T.; Zuppiroli, L.

1986-06-01

Boron carbide is a refractory hard and light material of interest in nuclear technology (fission and also fusion). Transmission electron microscopy was used to examine the properties of radiation induced damage. Firstly, the production of point defects and their clustering was studied in samples irradiated by 1 MeV electron in a high voltage electron microscope at selected temperatures from 12 K to 1000 K. Secondly, conventional transmission electron microscopy was used to understand the production of helium bubbles in neutron irradiated boron carbide and their role in the generation of microcracks. Finally, the interaction between point defects and bubbles was also examined

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)

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.

A Study On The Metal Carbide Composite Diffusion Bonding For Mechanical Seal

Directory of Open Access Journals (Sweden)

Kim D.-K.

2015-06-01

Full Text Available Mechanical Seal use highly efficient alternative water having a great quantity of an aqueous solution and has an advantage no corrosion brine. Metal Carbide composites have been investigated as potential materials for high temperature structural applications and for application in the processing industry. The existing Mechanical seal material is a highly expensive carbide alloy, and it is difficult to take a price advantage. Therefore the study of replacing body area with inexpensive steel material excluding O-ring and contact area which demands high characteristics is needed.

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)

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)

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...

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.

Equations of state for enriched uranium and uranium alloy to 3500 MPa

International Nuclear Information System (INIS)

Bai Chaomao; Hai Yuying; Liu Jenlong; Li Zhenrong

1990-04-01

The volume compressions of 6 kinds of cast materials including enriched uranium, poor uranium, U-0.57 wt% Ti, U-0.33 wt% Nb, U-2.85 wt% Nb and U-7.5 wt% Nb-3.3 wt% Zr have been determined by monitoring piston displacements in a piston cylinder apparatus with double strengthening rings to 3500 MPa at room temperature. The dilation of the cylinder vessel and the press deformation were corrected by some experiments. The calculational data free from using the standard sample closed with used standard sample. The volume compressions of enriched uranium and poor uranium are nearly coincident. Pure uranium is more compressible than uranium alloys. These values of enriched uranium are in close agreement with values of Bridgman's pure uranium. The fitting coefficients of Bridgman's polynomial and Anderson's equation of state and isothermal bulk modules for the above materials are given

Study of the recrystallisation of irradiated uranium; Etude sur l'uranium irradie

Energy Technology Data Exchange (ETDEWEB)

Bloch, J; Mustelier, J P; Bussy, P; Blin, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

1- Study of the recrystallisation of irradiated uranium. The recrystallisation of uranium irradiated to a burnup level of 220 MWj/t, at a temperature of the order of 350 deg. C, has been investigated. The observations were made chiefly by means of micrography an hardness measurements. If the irradiated metal is compared with a cold-drawn metal showing the same shearing of the twinned crystals, and therefore the same rate of plastic deformation, as the irradiated metal, it is noted that the restoring of the irradiated metal takes place at a considerably higher temperature than that of the cold-drawn metal. Pre-crystallisation is very much delayed. Only, a passage of the {alpha}-{beta} transformation point quickly wipes out irradiation effect. 2- Hardening of uranium by irradiation. Using hardness measurements we have studied more especially the effect of very weak irradiations on uranium (integrated flux < 10{sup 16} nvt). The hardness does not increase linearly with the flux, but a period of incubation is observed probably representing the time necessary for saturation of the dislocations. (author)Fren. [French] 1- Etude de la recristallisation de l'uranium irradie. On a etudie la recristallisation d'uranium irradie jusqu'a un taux de combustion de 220 MWj/t a une temperature de l'ordre de 350 deg. C. Les observations ont ete faites principalement a l'aide de la micrographie et de la durete. Si l'on compare le metal irradie avec un metal ecroui presentant le meme cisaillement des macles, donc le meme taux de deformation plastique que le metal irradie, on constate que la restauration du metal irradie se fait a une temperature notablement superieure a celle du metal ecroui. La recristallisation est tres retardee. Seul, un passage du point de transformation {alpha}-{beta} efface rapidement l'effet de l'irradiation. 2- Durcissement de l'uranium par irradiation. Nous avons, a l'aide de la durete, etudie plus particulierement l'effet de tres faibles irrtions sur l'uranium

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)

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

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)

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

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.

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)

Artificial Composites for High Temperature Applications; A Review (Composites Artificiels Destines a des Applications a Haute Temperature; Un Expose),

Science.gov (United States)

1987-01-01

ceramic fibres and whiskers including silicon carbide, aluminium oxide, zirconium oxide, hafnium oxide and uranium dioxide, were being considered as...3.1.3 Fabrication Processes.................................. 9 3.1.4 Wire/Matrix Interactions and diffusion barriers ........ 10 3.1.5 Mechanical...20 3.2.1 Fibres, Matrices and Fabrication Processes ............ 20 3.2.2 Fibre/Matrix Interaction and Properties............... 21 3.3 Ductile

Method of fabricating porous silicon carbide (SiC)

Science.gov (United States)

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

1995-01-01

Porous silicon carbide is fabricated according to techniques which result in a significant portion of nanocrystallites within the material in a sub 10 nanometer regime. There is described techniques for passivating porous silicon carbide which result in the fabrication of optoelectronic devices which exhibit brighter blue luminescence and exhibit improved qualities. Based on certain of the techniques described porous silicon carbide is used as a sacrificial layer for the patterning of silicon carbide. Porous silicon carbide is then removed from the bulk substrate by oxidation and other methods. The techniques described employ a two-step process which is used to pattern bulk silicon carbide where selected areas of the wafer are then made porous and then the porous layer is subsequently removed. The process to form porous silicon carbide exhibits dopant selectivity and a two-step etching procedure is implemented for silicon carbide multilayers.

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)

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

Uranium leaching by fungal metabolite

International Nuclear Information System (INIS)

Wang Yongdong; Li Guangyue; Ding Dexin; Hu Nan

2012-01-01

To explore new means of bioleaching, one strain of high-yielding fungi-Aspergillus niger which could produce organic acids was separated and purified from soil samples of uranium mine. The influence of cultural temperature, initial pH value, inoculum sizes on its growth characteristics were carried out. And the tests of uranium leaching of metabolin of Aspergillus niger were operated. On these tests, the effects of metabolin of Aspergillus niger with different pH value produced in the diverse culture temperature on uranium leaching were investigated. The results show that this strain of Aspergillus niger can grow best under the following conditions: the temperature is 37℃, the initial pH value is 7.0, the inoculum sizes is 2% (the OD value of the spores solution is 0.06). The uranium extraction effects relative to the final pH value of the cultures. and the maximum leaching rates is 83.05% when the pH value is 2.3. (authors)

Electron microscopy study of radiation effects in boron carbide

International Nuclear Information System (INIS)

Stoto, T.

1987-03-01

Boron carbide is a disordered non-stoechiometric material with a strongly microtwinned polycristallyne microstructure. This ceramic is among the candidate materials for the first wall coating in fusion reactor and is used as a neutron absorber in the control rods of fast breeder reactors. The present work deals with the nature of radiation damage in this solid. Because of helium internal production, neutron irradiated boron carbide is affected by swelling and by a strong microcracking which can break up a pellet in fine powder. These processes are rather intensitive to the irradiation parameters (temperature, flux and even neutron spectrum). Transmission electron microscopy of samples irradiated by the fast neutrons of a reactor, the electrons of a high voltage electron microscope and of samples implanted with helium ions was used to understand the respective roles of helium and point defects in the processes of swelling and microcracking. The design of an irradiation chamber for helium implantation at controlled temperature from 600 to 1700 0 C was an important technical part of this work [fr

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....

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Thermodynamic properties of α-uranium

International Nuclear Information System (INIS)

Ren, Zhiyong; Wu, Jun; Ma, Rong; Hu, Guichao; Luo, Chao

2016-01-01

The lattice constants and equilibrium atomic volume of α-uranium were calculated by Density Functional Theory (DFT). The first principles calculation results of the lattice for α-uranium are in agreement with the experimental results well. The thermodynamic properties of α-uranium from 0 to 900 K and 0–100 GPa were calculated with the quasi-harmonic Debye model. Volume, bulk modulus, entropy, Debye temperature, thermal expansion coefficient and the heat capacity of α-uranium were calculated. The calculated results show that the bulk modulus and Debye temperature increase with the increasing pressure at a given temperature while decreasing with the increasing temperature at a given pressure. Volume, entropy, thermal expansion coefficient and the heat capacity decrease with the increasing pressure while increasing with the increasing temperature. The theoretical results of entropy, Debye temperature, thermal expansion coefficient and the heat capacity show good agreement with the general trends of the experimental values. The constant-volume heat capacity shows typical Debye T"3 power-law behavior at low temperature limit and approaches to the classical asymptotic Dulong-Petit limit at high temperature limit. - Highlights: • Thermodynamic properties of α-U were predicted systematically with quasi-harmonic Debye model. • Summarizations of the corresponding experimental and theoretical results have been made for the EOS and other thermodynamic parameters. • The calculated thermodynamic properties show good agreement with the experimental results in general trends.

Thermodynamic properties of α-uranium

Energy Technology Data Exchange (ETDEWEB)

Ren, Zhiyong; Wu, Jun; Ma, Rong; Hu, Guichao; Luo, Chao, E-mail: luochaoboss@sohu.com

2016-11-15

The lattice constants and equilibrium atomic volume of α-uranium were calculated by Density Functional Theory (DFT). The first principles calculation results of the lattice for α-uranium are in agreement with the experimental results well. The thermodynamic properties of α-uranium from 0 to 900 K and 0–100 GPa were calculated with the quasi-harmonic Debye model. Volume, bulk modulus, entropy, Debye temperature, thermal expansion coefficient and the heat capacity of α-uranium were calculated. The calculated results show that the bulk modulus and Debye temperature increase with the increasing pressure at a given temperature while decreasing with the increasing temperature at a given pressure. Volume, entropy, thermal expansion coefficient and the heat capacity decrease with the increasing pressure while increasing with the increasing temperature. The theoretical results of entropy, Debye temperature, thermal expansion coefficient and the heat capacity show good agreement with the general trends of the experimental values. The constant-volume heat capacity shows typical Debye T{sup 3} power-law behavior at low temperature limit and approaches to the classical asymptotic Dulong-Petit limit at high temperature limit. - Highlights: • Thermodynamic properties of α-U were predicted systematically with quasi-harmonic Debye model. • Summarizations of the corresponding experimental and theoretical results have been made for the EOS and other thermodynamic parameters. • The calculated thermodynamic properties show good agreement with the experimental results in general trends.

Rirang uranium ore processing: continuous solvent extraction of uranium from Rirang ore acid digestion solution

International Nuclear Information System (INIS)

Riza, F.; Nuri, H. L.; Waluya, S.; Subijanto, A.; Sarono, B.

1998-01-01

Separation of uranium from Rirang ore acid digestion solution by means of continuous solvent extraction using mixer-settlers has been studied and a mixture of 0.3 M D2EHPA and 0.075 M TOPO extracting agent and kerosene diluent is employed to recover and separate uranium from Th, RE, phosphate containing solution. The experiments have been conducted batch-wise and several parameters have been studied including the aqueous to organic phase ratio, A/O, the extraction and the stripping times, and the operation temperature. The optimum conditions for extraction have been found to be A/O = 2 ratio, five minute extraction time per stage at room temperature. The uranium recovery of 99.07% has been achieved at those conditions whilst U can be stripped from the organic phase by 85% H 3 PO 4 solution with an O/A = 1 for 5 minutes stripping time per stage, and in a there stage operation at room temperature yielding a 100% uranium recovery from the stripping process

The determination of boron and carbon in reactor grade boron carbide

International Nuclear Information System (INIS)

Crossley, D.; Wood, A.J.; McInnes, C.A.J.; Jones, I.G.

1978-09-01

The sealed tube method of dissolution at high temperature and pressure has been successfully applied in the analysis of reactor grade boron carbide for the determination of boron. A 50 mg sample of boron carbide is completely dissolved by heating with concentrated nitric acid in a sealed tube at 300 0 C. The boron content of the resultant sample solution is determined by the mannitol potentiometric titration method. The precision of the method for the determination of 2.5 mg of boron using the Harwell automatic potentiometric titrator is 0.2% (coefficient of variation). The carbon content of a boron carbide sample is determined by combustion of the sample at 1050 0 C in a stream of oxygen using vanadium pentoxide to ensure the complete oxidation of the sample. The carbon dioxide produced from the sample is measured manometrically and the precision of the method for the determination of 4 mg of carbon is 0.4% (coefficient of variation). (author)

Reaction of uranium and the fluorocarbon FC-75

Science.gov (United States)

Young, R. H.

1985-04-01

Because of criticality concerns with water cooling in enriched uranium upgrading, a fluorocarbon has been evaluated as a replacement coolant for internal module components in the Plasma Separation Process (PSP). The interaction of bulk uranium and of powdered uranium with FC-75 has been investigated at temperatures between 200 and 700 C. The gas pressure and the metal temperature were monitored as a function of time. Modest temperature changes of 50 to 100 C were observed for the bulk uranium/fluorocarbon reaction. Much larger changes (up to 1000 C) were noted for the reaction involving high surface area uranium powder. These temperature transients, particularly for the powdered uranium reaction, were short-lived ( 10 seconds) and indicative of the formation of a protective layer of reaction products. Analysis of residual gas products by infrared spectroscopy indicated that one potentially serious hazard, UF6, was not present; however, several small toxic fluorocarbons were produced by thermolysis and/or reaction. X-ray diffraction analysis of the residual solids indicated UF4 and UO2 were the major solid products.

Determination of an instability temperature for alloys in the cooling gas of a high temperature reactor

International Nuclear Information System (INIS)

Grimmer, H.; Grman, D.; Krompholz, K.; Zimmermann, U.; Ullrich, G.

1985-05-01

High temperature alloys designed to be used for components in the primary circuit of a helium cooled high temperature nuclear reactor show massive CO production above a certain temperature, called the instability temperature T/sub i/, which increases with increasing partial pressure of CO in the cooling gas. At p/sub CO/ = 15 microbar, T/sub i/ lies between 900 and 950 degrees C for the four alloys under investigation: T/sub i/ is lowest for the iron base alloy Incoloy 800 H and increases for the nickel base alloys in the order Inconel 617, HDA 230 and Nimonic 86. Measurements of T/sub i/ made at 3 different laboratories were compared and shown to agree for p/sub CO/ 25 microbar, compatible with CO production by a reaction of Cr2O3 with carbides. Some measurements of T/sub i/ on HDA 230 and Nimonic 86 were performed in the course of simulated reactor disturbances. They showed that the oxide layer looses its protective properties above T/sub i/. A highlight of the examinations was the detection of eta-carbides (M6C) with unusual properties. M6C is the only type of carbide occuring in HDA 230. An eta-carbide with a lattice constant of 1088.8 pm had developed at the surface of Nimonic 86 during pre-oxidation before the disturbance simulation. Its composition is estimated at Ni3SiMo2C. Eta-carbides containing Si and especially eta-carbides with lattice constants as low as 1088.8 pm have been described only rarely until now. (author)

Effect of bond coat and preheat on the microstructure, hardness, and porosity of flame sprayed tungsten carbide coatings

Science.gov (United States)

Winarto, Winarto; Sofyan, Nofrijon; Rooscote, Didi

2017-06-01

Thermally sprayed coatings are used to improve the surface properties of tool steel materials. Bond coatings are commonly used as intermediate layers deposited on steel substrates (i.e. H13 tool steel) before the top coat is applied in order to enhance a number of critical performance criteria including adhesion of a barrier coating, limiting atomic migration of the base metal, and corrosion resistance. This paper presents the experimental results regarding the effect of nickel bond coat and preheats temperatures (i.e. 200°C, 300°C and 400°C) on microstructure, hardness, and porosity of tungsten carbide coatings sprayed by flame thermal coating. Micro-hardness, porosity and microstructure of tungsten carbide coatings are evaluated by using micro-hardness testing, optical microscopy, scanning electron microscopy, and X-ray diffraction. The results show that nickel bond coatings reduce the susceptibility of micro crack formation at the bonding area interfaces. The percentage of porosity level on the tungsten carbide coatings with nickel bond coat decreases from 5.36 % to 2.78% with the increase of preheat temperature of the steel substrate of H13 from 200°C to 400°C. The optimum hardness of tungsten carbide coatings is 1717 HVN in average resulted from the preheat temperature of 300°C.

Granite-related hypothermal uranium mineralization in South China

International Nuclear Information System (INIS)

Liu, X.; Wu, J.; Pan, J.; Zhu, M.

2014-01-01

As one of the important geological types, granite-related uranium deposits account for about 29% of the total discovered natural uranium resources in China. Most of the granite-related uranium deposits located in Taoshan - Zhuguang uranium metallogenic belt, South China. In addition to the typical pitchblende vein-type uranium mineralization of epithermal metallogenic system, a new type of granite-related uranium mineralization with characteristics of hypothermal matallogenic system was discovered in South China by current studies. However, hypothermal is contact thermal to epithermal mineralization, and not the conventional intrusive high temperature mineralization. Hypothermal uranium mineralization is presented by disseminated uraninite or pitchblende stockwork in fissures in granites normally with extensive alkaline alteration. The high temperature mineral assemblage of uraninite associate with scheelite and tourmaline was identified in hypothermal uranium mineralization. Fluid inclusion studies on this type mineralization indicated the middle to high temperature (>250℃) mineralization with the mixing evidence of ore forming solution derived from deep level, and the boiling and mixing of ore forming solution are regarded as the dominant mineralization mechanism for the precipitating of uranium. In contrast to the mineralization ages of 67 Ma to 87 Ma for typical pitchblende vein mineralization of epithermal metallogenic system, the mineralization age is older than 100 Ma for hypothermal uranium mineralization in granite. In the Shituling deposit, Xiazhuang uranium ore field, uraninite and pitchblende micro veins with extensive potassic alteration, chloritization and sericitization are hosted in fissures of Indo-Chinese epoch granites with the uranium mineralization age of 130 Ma to 138 Ma with a mineralization temperature of 290℃ to 330℃ indicated. Other examples sharing the similar characters of hypothermal uranium mineralization have been recognized in

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

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

The effects of the local fracture stress and carbides on the cleavage fracture characteristics of Mn-Mo-Ni low alloy steels in the transition region

International Nuclear Information System (INIS)

Yang, Won Jon; Huh, Moo Young; Roh, Sung Joo; Lee, Bong Sang; Oh, Yong Jun; Hong, Jun Hwa

2000-01-01

In the ductile-brittle transition temperature region of SA508 C1.3 Mn-Mo-Ni low alloy steels, the relationship of the local fracture stress and carbides influencing the cleavage fracture behavior was investigated. Based on the ASTM E1921-97 standard method, the reference transition temperatures were determined by three point bending fracture toughness tests. A local fracture stress σ f * , was determined from a theoretical stress distribution in front of crack tip using the cleavage initiation distance measured in each fractured specimen surface. The local fracture stress values showed a strong relationship with toughness characteristics of the materials and those were larger in the materials of smaller carbide size. Quantitative analysis of carbides showed that carbides larger than a certain size are mainly responsible for the cleavage fracture in the ductile-brittle transition temperature region. (author)

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

Kinetics and mechanism of oxidation of carbidized electrolytic chromium coatings

International Nuclear Information System (INIS)

Arkharov, V.I.; Yar-Mukhamedov, Sh.Kh.

1978-01-01

Thermal stability carbidized electrolytic chromium coatings has been studied depending on the conditions of their formation; the specific features of the mechanism of oxidation at 1200 deg in an air atmosphere have been elucidated. It has been established that kinetics of high temperature oxidation of the coatings depends essentially on the conditions of their formation and on the composition of steel to which the coating is applied. It has been shown that two oxidation mechanisms are possible: by diffusion of the residual chromium through a carbide layer along the carbide grain boundaries outwards or, when there is no residual chromium, by chemical reaction of carbon combustion and oxidation of the liberated chromium. The comparison of oxidation kinetic curves of the samples of 38KhMYuA, 35KhGSA, and DI-22 steels with and without coating has shown that the coatings under study have a better protective effect on 38KhMYuA steel than on 35KhGSA, although without coating oxidability of the first steel is higher than that of the second

Thermal stress relieving of dilute uranium alloys

International Nuclear Information System (INIS)

Eckelmeyer, K.H.

1980-01-01

The kinetics of thermal stress relieving of uranium - 2.3 wt. % niobium, uranium - 2.0 wt. % molybdenum, and uranium - 0.75 wt. % titanium are reported and discussed. Two temperature regimes of stress relieving are observed. In the low temperature regime (T 0 C) the process appears to be controlled by an athermal microplasticity mechanism which can be completely suppressed by prior age hardening. In the high temperature regime (300 0 C 0 C) the process appears to be controlled by a classical diffusional creep mechanism which is strongly dependent on temperature and time. Stress relieving is accelerated in cases where it occurs simultaneously with age hardening. The potential danger of residual stress induced stress corrosion cracking of uranium alloys is discussed. It is shown that the residual stress relief which accompanies age hardening of uranium - 0.75% titanium more than compensates for the reduction in K/sub ISCC/ caused by aging. As a result, age hardening actually decreases the susceptibility of this alloy to residual stress induced stress corrosion cracking

Disorder and defects are not intrinsic to boron carbide

Science.gov (United States)

Mondal, Swastik; Bykova, Elena; Dey, Somnath; Ali, Sk Imran; Dubrovinskaia, Natalia; Dubrovinsky, Leonid; Parakhonskiy, Gleb; van Smaalen, Sander

2016-01-01

A unique combination of useful properties in boron-carbide, such as extreme hardness, excellent fracture toughness, a low density, a high melting point, thermoelectricity, semi-conducting behavior, catalytic activity and a remarkably good chemical stability, makes it an ideal material for a wide range of technological applications. Explaining these properties in terms of chemical bonding has remained a major challenge in boron chemistry. Here we report the synthesis of fully ordered, stoichiometric boron-carbide B13C2 by high-pressure-high-temperature techniques. Our experimental electron-density study using high-resolution single-crystal synchrotron X-ray diffraction data conclusively demonstrates that disorder and defects are not intrinsic to boron carbide, contrary to what was hitherto supposed. A detailed analysis of the electron density distribution reveals charge transfer between structural units in B13C2 and a new type of electron-deficient bond with formally unpaired electrons on the C-B-C group in B13C2. Unprecedented bonding features contribute to the fundamental chemistry and materials science of boron compounds that is of great interest for understanding structure-property relationships and development of novel functional materials.

The uranium-carbon and plutonium-carbon systems. A thermochemical assessment

International Nuclear Information System (INIS)

1963-01-01

A fair amount of thermochemical data has been accumulated on the compounds in the uranium-carbon system. The main difficulties involved appear to be the sluggishness of the reaction of these carbides and the lack of information on the true equilibrium diagram. The information assessed in this report is accurate to, say ± 5 kcal on the average. This is in fact satisfactory for quite a number of calculations of equilibria involving uranium and carbon. It is not accurate enough for more ambitious calculations such as that of the equilibrium diagram. Present assessment has also made clear the gaps that still exist. It appears that it is mainly the non-stoichiometric parts of the diagram that need extensive further studies; this would also assist in increasing the accuracy of the known data. 66 refs, 6 figs, 15 tabs

Quantitative Analysis on Carbide Precipitation in V-Ti Microalloyed TRIP Steel Containing Aluminum

Directory of Open Access Journals (Sweden)

Fu Shiyu

2016-01-01

Full Text Available Introducing fine precipitates is an important way to enhance the properties of transformation-induced plasticity (TRIP steels. In present work, two V-Ti microalloyed TRIP steels containing aluminum with different content were compared. The average size, size distribution and numbers of vanadium-titanium carbides in samples cold rolled, quenched after being held at 800°C and quenched after intercritical annealing at 800°C and being held at bainitic isothermal transformation temperature of 400°C were investigated by using the technique of carbon extraction replica, twin jet chemical polishing thinning and transmission electron microscopy. The carbides were identified to be (Ti,VC precipitates in steel A and VC in steel B respectively, precipitated mainly from ferrites grains. The average equivalent radius was 3~6nm. Comparison of the experimental results in A and B steel revealed low carbon diffusion rate caused by aluminum inhibited the coarsening of vanadium-titanium carbides. The experimental results also showed that VC carbides dissolution occurred during the intercritical annealing at 800°C.

Electrochemical preparation of new uranium oxide phases

International Nuclear Information System (INIS)

Smolenskij, V.V.; Lyalyushkin, N.V.; Bove, A.L.; Komarov, V.K.; Kapshukov, I.I.

1992-01-01

Behaviour of uranium ions in oxidation states 3+ and 4+ in molten chlorides of alkali metals in the temperature range of 700-900 degC in the atmosphere of an inert gas was studied by the method of cyclic voltametry. It is shown that as a result of introduction of crystal uranium dioxide into the salt melt formation of uranium oxide ions of the composition UO + and UO 2+ occurs, the ions participating in electrode reactions and bringing about formation of the following uranium oxides on the cathode: UO and, presumably, U 3 O 4 . Oxides UO and U 3 O 4 are thermodynamically unstable at low temperatures and decompose into uranium oxide of the composition UO 2-x , where x varies from 0 to 0.05, and metal uranium

Uranium self-diffusion in uranium monocarbide; Determination du coefficient d'autodiffusion de l'uranium dans son monocarbure

Energy Technology Data Exchange (ETDEWEB)

Villaine, P [Commissariat a l' Energie Atomique, 38 - Grenoble (France). Centre d' Etudes Nucleaires

1967-10-01

Uranium self diffusion in near-stoichiometric stabilized uranium monocarbide has been investigated in the temperature range 1450-2000 deg. C. A thin layer of {sup 235}UC was deposited onto the samples and the diffusion profiles were analyzed by both sectioning and alpha-spectrometry techniques. The variation with temperature of the self-diffusion coefficient can be expressed by the equation: D = 7.5 x 10{sup -5} exp [-(81 {+-} 10) kcal/mole / RT] Cm{sup 2} s{sup -1} The coefficient D decreases with increasing carbon content. Autoradiographs and profile analysis have evidenced a preferential grain-boundary diffusion at all temperatures and compositions investigated. This phenomenon was used for a study of grain-boundary migration and for the evaluation of grain-boundary diffusion coefficients. The activation energy thus derived is close to the volume diffusion activation energy. (author) [French] L'autodiffusion de l'uranium dans le monocarbure d'uranium de composition voisine de la stoechiometrie et stabilise par recuit prealable, a ete etudiee entre 1450 et 2000 deg. C par la methode du depot mince de traceur, suivie des techniques d'abrasion comptage et de spectrometrie alpha. La variation avec la temperature du coefficient d'autodiffusion peut s'ecrire: D = 7.5 x 10{sup -5} exp [-(81 {+-} 10) kcal/mole / RT] Cm{sup 2} s{sup -1} Le coefficient D decroit avec une augmentation de la teneur en carbone. L'observation d'autoradiographies et l'analyse de profils de diffusion ont mis en evidence l'importance d'une diffusion intergranulaire preferentielle pour toutes les compositions etudiees et a toutes les temperatures. Cette diffusion a egalement ete utilisee pour l'etude de la migration des joints de grains et pour le calcul approche du coefficient de diffusion mtergranulaire. L'energie d'activation ainsi determinee est voisine de celle correspondant a la diffusion volumique. (auteur)

Characterisation of silicon carbide films deposited by plasma-enhanced chemical vapour deposition

International Nuclear Information System (INIS)

Iliescu, Ciprian; Chen Bangtao; Wei Jiashen; Pang, A.J.

2008-01-01

The paper presents a characterisation of amorphous silicon carbide films deposited in plasma-enhanced chemical vapour deposition (PECVD) reactors for MEMS applications. The main parameter was optimised in order to achieve a low stress and high deposition rate. We noticed that the high frequency mode (13.56 MHz) gives a low stress value which can be tuned from tensile to compressive by selecting the correct power. The low frequency mode (380 kHz) generates high compressive stress (around 500 MPa) due to ion bombardment and, as a result, densification of the layer achieved. Temperature can decrease the compressive value of the stress (due to annealing effect). A low etching rate of the amorphous silicon carbide layer was noticed for wet etching in KOH 30% at 80 o C (around 13 A/min) while in HF 49% the layer is practically inert. A very slow etching rate of amorphous silicon carbide layer in XeF 2 -7 A/min- was observed. The paper presents an example of this application: PECVD-amorphous silicon carbide cantilevers fabricated using surface micromachining by dry-released technique in XeF 2

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

Fluidized bed deposition and evaluation of silicon carbide coatings on microspheres

International Nuclear Information System (INIS)

Federer, J.I.

1977-01-01

The fuel element for the HTGR is an array of closely packed fuel microspheres in a carbonaceous matrix. A coating of dense silicon carbide (SiC), along with pyrocarbon layers, is deposited on the fueled microspheres to serve as a barrier against diffusion of fission products. The microspheres are coated with silicon carbide in a fluidized bed by reaction of methyltrichlorosilane (CH 3 SiCl 3 or MTS) and hydrogen at elevated temperatures. The principal variables of coating temperature and reactant gas composition (H 2 /MTS ratio) have been correlated with coating rate, morphology, stoichiometry, microstructure, and density. The optimum temperature for depositing highly dense coatings is in the range 1475 to 1675 0 C. Lower temperatures result in silicon-rich deposits, while higher temperatures may cause unacceptable porosity. The optimum H 2 /MTS ratio for highly dense coatings is 20 or more (approximately 5% MTS or less). The amount of grown-in porosity increases as the H 2 /MTS ratio decreases below 20. The requirement that the H 2 /MTS ratio be about 20 or more imposes a practical restraint on coating rate, since increasing the total flow rate would eventually expel microspheres from the coating tube. Evaluation of stoichiometry, morphology, and microstructure support the above mentioned optimum conditions of temperature and reactant gas composition. 18 figures, 3 tables

Production of technical silicon and silicon carbide from rice-husk

Directory of Open Access Journals (Sweden)

A. Z. Issagulov

2014-10-01

Full Text Available In the article there are studied physical and chemical properties of silicon-carbonic raw material – rice-husk, thermophysical characteristics of the process of rice-husk pyrolysis in nonreactive and oxidizing environment; structure and phase composition of products of the rice-husk pyrolysis in interval of temperatures 150 – 850 °С and high temperature pyrolysis in interval of temperatures 900 – 1 500 °С. There are defined the silicon-carbon production conditions, which meet the requirements applicable to charging materials at production of technical silicon and silicon carbide.

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.)

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.

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

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

Synthesis of Uranium nitride powders using metal uranium powders

International Nuclear Information System (INIS)

Yang, Jae Ho; Kim, Dong Joo; Oh, Jang Soo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik

2012-01-01

Uranium nitride (UN) is a potential fuel material for advanced nuclear reactors because of their high fuel density, high thermal conductivity, high melting temperature, and considerable breeding capability in LWRs. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. The carbothermic reduction has an advantage in the production of fine powders. However it has many drawbacks such as an inevitable engagement of impurities, process burden, and difficulties in reusing of expensive N 15 gas. Manufacturing concerns issued in the carbothermic reduction process can be solved by changing the starting materials from oxide powder to metals. However, in nitriding process of metal, it is difficult to obtain fine nitride powders because metal uranium is usually fabricated in the form of bulk ingots. In this study, a simple reaction method was tested to fabricate uranium nitride powders directly from uranium metal powders. We fabricated uranium metal spherical powder and flake using a centrifugal atomization method. The nitride powders were obtained by thermal treating those metal particles under nitrogen containing gas. We investigated the phase and morphology evolutions of powders during the nitriding process. A phase analysis of nitride powders was also a part of the present work

Dissolution of metallic uranium and its alloys. Part II. Screening study results: Identification of an effective non-thermal uranium dissolution method

International Nuclear Information System (INIS)

Laue, C.A.; Gates-Anderson, D.; Fitch, T.E.

2004-01-01

Screening experiments were performed to evaluate reagent systems that deactivate pyrophoric, metallic depleted uranium waste streams at ambient temperature. The results presented led to the selection of two systems, which would be investigated further, for the design of the LLNL onsite treatment process of metallic depleted uranium wastes. The two feasible systems are: (a) 7.5 mol/l H 2 SO 4 - 1 mol/l HNO 3 and (b) 3 mol/l HCl - 1 mol/l H 3 PO 4 . The sulfuric acid system dissolves uranium metal completely, while the hydrochloric-phosphoric acid system converts the metal completely into a solid, which might be suitable for direct disposal. Both systems combine oxidation of metallic uranium with complexation of the uranium ions formed to effectively deactivate uranium.s pyrophoricity at ambient temperature. (author)

Drawing of uranium in {gamma} phase; Filage de l'uranium en phase gamma

Energy Technology Data Exchange (ETDEWEB)

Stohr, J A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Chevigny, R [Conpagnie, Pechiney, 73 - Chambery (France)

1955-07-01

It describes the study of working and drawing of uranium in {gamma} phase. In a first part, the forgeable characteristics of uranium metal in the three different phases ({alpha}, {beta} and {gamma}) are compared by using BASTIEN-PORTEVIN method. The different experiments are testing the behaviour metal in each phase under different stresses and a temperature gradient as slow and shock traction, slow and shock compression, resilience, flexibility. Results show that optimum conditions for uranium drawing is uranium in phase {gamma}. In a second part, it described the drawing method and process. The uranium rods obtained by this technique are of very good quality. In addition, the material wear is very low which permits a low production cost. Finally, the uranium rod physical properties are studied. (M.P.)

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.

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.

Undercooling, nodule count and carbides in thin walled ductile cast iron

DEFF Research Database (Denmark)

Pedersen, Karl Martin; Tiedje, Niels Skat

2008-01-01

Ductile cast iron has been cast in plate thicknesses between 2 to 8 mm. The temperature has been measured during the solidification and the graphite nodule count and size distribution together with the type and amount of carbides have been analysed afterwards. Low nodule count gives higher...

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

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

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

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

Thermodynamic studies of thorium carbide fuel preparation and fuel-clad comptability

International Nuclear Information System (INIS)

Besmann, T.M.; Beahm, E.C.

1979-01-01

The carbothermic reduction of thorium and uranium-thorium dioxide to monocarbide has been assessed. Equilibrium calculations have yielded Th-C-O and U-Th-C-O phase equilibria and (CO) pressures generated during reduction. The (CO) pressures were found to be at least five orders of magnitude greater than any of the other 15 gaseous species considered. This confirms that the monocarbide can successfully be prepared by carbothermic reduction. The chemical compatibility of thorium carbides with the Cr-Fe-Ni content of clad alloys has been thermodynamically avaluated. Solid solutions of 5 > and 5 > and of 7 C 3 > and 7 C 3 > were the principal reaction products. The Cr-Fe-Ni content of 316 stainless steel showed much less reaction product than that for any of the other six alloys considered. (orig.) [de

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

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.

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

Directory of Open Access Journals (Sweden)

W. Tillmann

2017-06-01

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

Electrodeposition of uranium metal by reduction of uranium oxides in molten Lif-KF=NaF-CaF 2-UF4

International Nuclear Information System (INIS)

Pao, D.S.; Burris, L.; Steunenberg, R.K.; Tomczuk, Z.

1990-01-01

Although electrolytic reduction of uranium oxides was shown to be feasible in the early 1960's it is recognized that considerable improvement in the electrolytic reduction technology must be achieved for practical applications. This exploratory work on electrolytic reduction of uranium oxide was undertaken to investigate potential improvements in the technology. The approach taken was to deposit solid uranium metal directly on a solid cathode at temperatures below the melting point of uranium (1132 degrees C). The lower temperature electrolytic reduction process has several advantages over the existing chemical reduction processes. It lessens materials problems and special heating and insulating requirements associated with high-temperature operations. It removes most impurities. It does not produce the large quantities of byproduct oxides wastes typical of chemical reduction processes

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

International Nuclear Information System (INIS)

Malinge, A.

2011-01-01

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

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

Energy Technology Data Exchange (ETDEWEB)

Ammirabile, L.; Tsige-Tamirat, H. [European Commission, JRC, Inst. for Energy, Petten (Netherlands)

2011-07-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

International Nuclear Information System (INIS)

Ammirabile, L.; Tsige-Tamirat, H.

2011-01-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Experimental evaluation of chromium-carbide-based solid lubricant coatings for use to 760 C

Science.gov (United States)

Dellacorte, Christopher

1987-01-01

A research program is described which further developed and investigated chromium carbide based self-lubricating coatings for use to 760 C. A bonded chromium carbide was used as the base stock because of the known excellent wear resistance and the chemical stability of chromium carbide. Additives were silver and barium fluoride/calcium fluoride eutectic. The three coating components were blended in powder form, applied to stainless steel substrates by plasma spraying and then diamond ground to the desired coating thickness. A variety of coating compositions was tested to determine the coating composition which gave optimum tribological results. Coatings were tested in air, helium, and hydrogen at temperatures from 25 to 760 C. Several counterface materials were evaluated with the objective of discovering a satisfactory metal/coating sliding combination for potential applications, such as piston ring/cylinder liner couples for Stirling engines. In general, silver and fluoride additions to chromium carbide reduced the friction coefficient and increased the wear resistance relative to the unmodified coating. The lubricant additives acted synergistically in reducing friction and wear.

Carbide Nanoparticles Encapsulated in the Caves of Carbon Nanotubes by an In Situ Reduction-Carbonization Route

Directory of Open Access Journals (Sweden)

Chunli Guo

2011-01-01

Full Text Available Carbides (TiC, WC, and NbC nanoparticles fully encapsulated in the caves of carbon nanotubes (CNTs were synthesized via an in situ reduction-carbonization route at 600∘C in an autoclave. The structural features and morphologies of as-obtained products were investigated using by X-ray diffraction, scanning electron microscopy, and high-resolution transmission electron microscopy (HRTEM. HRTEM studies showed that the average diameter of CNTs encapsulated with carbide nanoparticles are in the range of 15–40 nm. The reaction temperature, the reaction time, and the metal catalyst are found to play crucial roles to the product morphology. The growth mechanism of carbide nanoparticles encapsulated in CNTs was discussed in detail.

The interdiffusion behaviour of uranium and aluminum

Energy Technology Data Exchange (ETDEWEB)

Le Claire, A D; Bear, I J

1952-03-05

Results are given for measurements made on the rate, as a function of temperature and pressure, of the increase in width of the diffusion zone formed by the interdiffusion of uranium and aluminium and of the relative penetration of uranium into aluminium and of aluminium into uranium. The experiments are unusual among diffusion meaurements in showing a marked effect of pressure on the rates of diffusion. Between 570 deg and 640 deg C (the eutectic temperature) the rate of growth of the diffusion zone increases regularly with increase of temperature or of pressure, at least up to the maximum pressure of 10 tons/sq.in. employed in the experiments. Below 570 deg C the rate of diffusion, rising with the rise in pressure at constant temperature increases discontinuously when the pressure reaches a value a little above 1 1/4 tons/sq.in. by a factor which increases, as the temperature is lowered, to a maximum value at 540 deg C. It then remains constant at this maximum value for temperatures below 540 deg C. The accelerating effect of pressure on the rate of interdiffusion is then most apparent when the temperature is below 540 deg C and the pressure raised from below to above about 1 1/4 tons/sq.in. The penetration of the uranium into the aluminium is about 2 1/4 times faster than that of the aluminium into the uranium. The results of a microscopical and X-ray examination of the diffusion zone are presented, and a discussion is given of the possible origins of the pressure effect and of the importance of the work in an understanding of the ''pimpling'' of aluminium clad uranium bars. (auth)

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