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Sample records for uranium graphite gas

  1. Uranium Oxide Aerosol Transport in Porous Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

    2012-01-23

    The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

  2. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France

    International Nuclear Information System (INIS)

    Gaussens, J.; Tanguy, P.

    1964-01-01

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  3. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France; Quelques aspects economiques de la filiere uranium naturel - Graphite - gaz. Etat actuel et tendance des couts en France

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J; Tanguy, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leo, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  4. Channel uranium-graphite reactor mounting

    International Nuclear Information System (INIS)

    Polushkin, K.K.; Kuznetsov, A.G.; Zheleznyakov, B.N.

    1981-01-01

    According to theoretical principles of general engineering technology the engineering experience of construction-mounting works at the NPP with channel uranium-graphite reactors is systematized. Main parameters and structural features of the 1000 MW channel uranium-graphite reactors are considered. The succession of mounting operations, premounting equipment and pipelines preparation and mounting works technique are described. The most efficient methods of fitting, welding and machining of reactor elements are recommended. Main problems of technical control service are discussed. A typical netted diagram of main equipment of channel uranium-graphite reactors mounting is given

  5. Developments in natural uranium - graphite reactors

    International Nuclear Information System (INIS)

    Bourgeois, J.

    1964-01-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  6. Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  7. Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B. [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  8. Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

    1980-11-01

    As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

  9. Study of the thermal drop at the uranium-can interface for fuel elements in gas-graphite reactors; Etude de la chute thermique au contact uranium-gaine pour des elements combustibles de reacteur de la filiere graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    Faussat, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Levenes, G; Michel, M [Societe Industrielle de Combustible Nucleaire (France)

    1964-07-01

    The report reviews the tests now under way at the CEA, for determining the thermal contact resistance at the uranium-can interface for fuel elements used in gas-graphite type reactors. These are laboratory tests carried out with equipment based on the principle of a heat flow across a stack of test pieces having planar contact surfaces. The following points emerge from this work: - for a metallic uranium element canned in magnesium, of the type G-2 or EDF-2, a value of 0.2 deg C/W/cm{sup 2} seems reasonable for can temperatures of 400 deg C and above. - this value is independent of the micro-geometric state of the uranium surface in a range of roughness which easily includes those observed on tubes and rods produced industrially. - for the internal cans of elements cooled internally and externally, the value of the contact resistance for temperatures of under 400 deg C as a function of the stresses in the can has not yet been measured exactly. (authors) [French] Le rapport fait le point des essais actuellement en cours au CEA pour determiner la resistance thermique de contact uranium-gaine pour des reacteurs de la filiere graphite-gaz. Ces essais sont effectues en laboratoire sur des appareils bases sur le principe d'une circulation de flux de chaleur a travers un empilement d'eprouvettes dont les faces en contact sont planes. De l'etude, il ressort essentiellement que: - pour un element a uranium metallique et gaine de magnesium type G-2 ou EdF-2, on peut admettre la valeur de 0,2 deg C/W/cm{sup 2} pour des temperatures de gaines de 400 deg C et plus. - cette valeur ne depend pas de l'etat de surface microgeometrique de l'uranium pour un domaine de rugosites couvrant largement celles que l'on observe sur des tubes et barreaux fabriques en serie. - pour les gaines internes d'elements a refroidissement interne et externe la valeur de la resistance de contact reste a preciser pour les temperatures inferieures a 400 deg C, en fonction des contraintes existant dans les

  10. Gas transport in graphitic materials

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1995-02-01

    The characterization of the gas transport properties of porous solids is of interest in several fields of science and technology. Many catalysts, adsorbents, soils, graphites and carbons are porous. The gas transport through most porous solids can be well described by the dusty gas model invented by Evans, Watson and Mason. This model includes all modes of gas tranport under steady-state conditions, which are Knudsen diffusion, combined Knudsen/continuum diffusion and continuum diffusion, both for gas pairs with equal and different molecular weights. In the absence of a pressure difference gas transport in a pore system can be described by the combined Knudsen/continuum diffusion coefficient D 1 for component 1 in the pores, the Knudsen diffusion coefficient D 1K in the pores, and the continuum diffusion coefficient D 12 for a binary mixture in the pores. The resistance to stationary continuum diffusion of the pores is characterized by a geometrical factor (ε/τ) 12 = (ε/τ)D 12 , were D 12 is the continuum diffusion coefficient for a binary mixture in free space. The Wicke-Kallenbach method was often used to measure D 1 as function of pressure. D 12 and D 1K can be derived from a plot 1/D 1 νs P, and ε/τcan be calculated since D 12 is known. D 1K and the volume of dead end pores can be derived from transient measurements of the diffusional flux at low pressures. From D 1K the expression (ε/τ c ) anti l por may be calculated, which characterizes the pore system for molecular diffusion, where collisions with the pore walls are predominant. (orig.)

  11. Diffusion of uranium in H-451 graphite at 900 to 14000C

    International Nuclear Information System (INIS)

    Tallent, O.K.; Wichner, R.P.; Towns, R.L.

    1983-03-01

    In this study, the diffusion of uranium (as a stand-in for plutonium) was investigated under conditions approximating those of the primary coolant loop in a High Temperature Gas-Cooled Reactor (HTGR). Profiles were obtained for uranium penetration in H-451 graphite (from the Great Lakes Carbon Company) at temperatures ranging from 900 to 1400 0 C. Diffusion coefficients are established for UO 2 and UC 2

  12. Exponential experiment on a uranium-graphite lattice; Experience exponentielle sur un reseau uranium-graphite

    Energy Technology Data Exchange (ETDEWEB)

    Leroy, J; Martelly, J

    1958-12-01

    An exponential experiment on a natural uranium and graphite lattice is described. The critical buckling for a cubic pile made with this lattice is B{sup 2} = 0.726 {+-} 0.011 m{sup -2} and the anisotropy {alpha} = (B{sub L}{sup 2}/B{sub T}{sup 2}) is 0.987 {+-} 0.006. The behavior of the neutron density in the lattice near the reflector and sources is discussed in detail. (authors) [French] Une experience exponentielle sur reseau a uranium naturel et graphite est decrite. Le Laplacien de la pile cubique nue, critique, constituee avec un tel reseau serait: B{sup 2} = 0.726 {+-} 0.011 m{sup -2} et l'anisotropie {alpha} (B{sub L}{sup 2}/B{sub T}{sup 2}) est egale a 0.987 {+-} 0.006. La perturbation apportee a la densite de neutrons dans le reseau par le voisinage du reflecteur est discutee en detail. (auteurs)

  13. Experience of on-site disposal of production uranium-graphite nuclear reactor.

    Science.gov (United States)

    Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

    2018-04-01

    The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

  14. Features of spherical uranium-graphite HTGR fuel elements control

    International Nuclear Information System (INIS)

    Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

    1985-01-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

  15. Features of spherical uranium-graphite HTGR fuel elements control

    Energy Technology Data Exchange (ETDEWEB)

    Kreindlin, I I; Oleynikov, P P; Shtan, A S

    1985-07-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

  16. Critical experiments on enriched uranium graphite moderated cores

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

    1978-07-01

    A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)

  17. Monte Carlo calculations of fast effects in uranium graphite lattices

    International Nuclear Information System (INIS)

    Beardwood, J.E.; Tyror, J.G.

    1962-12-01

    Details are given of the results of a series of computations of fast neutron effects in natural uranium metal/graphite cells. The computations were performed using the Monte Carlo code SPEC. It is shown that neutron capture in U238 is conveniently discussed in terms of a capture escape probability ζ as well as the conventional probability p. The latter is associated with the slowing down flux and has the classical exponential dependence on fuel-to-moderator volume ratio whilst the former is identified with the component of neutron flux above 1/E. (author)

  18. Organic tissues, graphite, and hydrocarbons in host rocks of the Rum Jungle Uranium Field, northern Australia

    Science.gov (United States)

    Foster, C.B.; Robbins, E.I.; Bone, Y.

    1990-01-01

    The Rum Jungle Uranium field consists of at least six early Proterozoic deposits that have been mined either for uranium and/or the associated base and precious metals. Organic matter in the host rocks of the Whites Formation and Coomalie Dolomite is now predominantly graphite, consistent with the metamorphic history of these rocks. For nine samples, the mean total organic carbon content is high (3.9 wt%) and ranged from 0.33 to 10.44 wt%. Palynological extracts from the host rocks include black, filamentous, stellate (Eoastrion-like), and spherical morphotypes, which are typical of early Proterozoic microbiota. The colour, abundance, and shapes of these morphotypes reflect the thermal history, organic richness, and probable lacustrine biofacies of the host rocks. Routine analysis of rock thin sections and of palynological residues shows that mineral grains in some of the host rocks are coated with graphitized organic matter. The grain coating is presumed to result from ultimate thermal degradation of a petroleum phase that existed prior to metamorphism. Hydrocarbons are, however, still present in fluid inclusions within carbonates of the Coomalie Dolomite and lower Whites Formation. The fluid inclusions fluoresce dull orange in blue-light excitation and their hydrocarbon content is confirmed by gas chromatography of whole-rock extracts. Preliminary analysis of the oil suggests that it is migrated, and because it has escaped graphitization through metamorphism it is probably not of early Proterozoic age. The presence of live oil is consistent with fluid inclusion data that suggest subsequent, low-temperature brine migration through the rocks. The present observations support earlier suggestions that organic matter in the host formations trapped uranium to form protore. Subsequent fluid migrations probably brought additional uranium and other metals to these formations, and the organic matter provided a reducing environment for entrapment. ?? 1990.

  19. Forming gas treatment of lithium ion battery anode graphite powders

    Science.gov (United States)

    Contescu, Cristian Ion; Gallego, Nidia C; Howe, Jane Y; Meyer, III, Harry M; Payzant, Edward Andrew; Wood, III, David L; Yoon, Sang Young

    2014-09-16

    The invention provides a method of making a battery anode in which a quantity of graphite powder is provided. The temperature of the graphite powder is raised from a starting temperature to a first temperature between 1000 and 2000.degree. C. during a first heating period. The graphite powder is then cooled to a final temperature during a cool down period. The graphite powder is contacted with a forming gas during at least one of the first heating period and the cool down period. The forming gas includes H.sub.2 and an inert gas.

  20. Dosage of boron traces in graphite, uranium and beryllium oxide

    International Nuclear Information System (INIS)

    Coursier, J.; Hure, J.; Platzer, R.

    1955-01-01

    The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [fr

  1. Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

    International Nuclear Information System (INIS)

    Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

    1999-01-01

    This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept

  2. Uranium and thorium abundances in some graphite-bearing precambrian rocks of India and implications

    International Nuclear Information System (INIS)

    Pandey, U.K.; Krishnamurthy, P.

    1995-01-01

    Graphite schists from parts of Gujarat in the Aravalli supergroup show maximum contents of uranium (70-95 ppm), hosted mainly in the graphites, whereas such schists from the Tamil Nadu granulite terrain contain distinctly lower amounts of uranium (7-9 ppm). Graphite-bearing hornblende gneiss and calc-granulites from Madurai, Tamil Nadu, contain higher amounts of uranium (12-28 ppm) than the schists, and uranium is mainly hosted by the magnetite and allanite occurring as independent grains with flaky graphite and also as inclusions within quartz. Khondalites from Andhra Pradesh are depleted in uranium (0.9-1.3 ppm) compared to Th (17.5-20.2 ppm). Except for the khondalites, which have high Th/U ratio (13.5-22.4), all the other samples have very low Th/U ratios (0.10-0.80) compared to the crustal average (3-4). Such variations among similar rock types, may in part be related to uranium and thorium abundances inherited from parental rocks, modified later by hydrothermal and/or metasomatic processes. Graphites from such rock types can provide both in situ and migrant reductants for hosting a variety of uranium and other metallic deposits. (author). 12 refs., 1 tab., 1 fig

  3. Calculation of the fissile mass of a graphite moderated critical assembly using 93% enriched uranium

    International Nuclear Information System (INIS)

    Correa, F.; Marzo, M.A.S.; Collussi, I.; Ferreira, A.C.A.

    1976-01-01

    The critical mass of uranium has been calculated for a graphite moderated set fueled with 93% enriched uranium to be mounted on the Instituto de Energia Atomica split table Zero Power Reactor. The core composition was optimized to permit the maximum number of configurations to be studied. Analysis of three core compositions shows that 8 Kg of uranium enriched to 93% - U-235 (by weight) and 100 Kg of thorium would be sufficient for criticality experiments [pt

  4. Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors

    International Nuclear Information System (INIS)

    Boudouresque, B.; Courcon, P.; Lestiboubois, G.

    1964-01-01

    The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm 2 gas pressure, should remain in contact with the fuel. (authors) [fr

  5. Comparative analysis of calculations and experiment for uranium-graphite lattices with natural and slightly-enriched uranium

    International Nuclear Information System (INIS)

    Khrennikov, N.N.; Shchukin, A.V.

    1988-01-01

    Three sets of experiments carried out at different times and in different laboratories on measuring the material parameter for uranium-graphite lattices using natural and slightly enriched uranium are analyzed. Comparison with the calculations by the TRIFOGR and MCU (the Monte Carlo method) codes reveals resonable agreement between the calculation and experiment (of the order of 0.4% in K eff ). 17 refs.; 3 tabs

  6. Graphite development for gas-cooled reactors in the USA

    International Nuclear Information System (INIS)

    Burchell, T.D.

    1991-01-01

    This document discusses Modular High-Temperature Gas-Cooled Reactor (MHTGR) graphite activities in the USA which currently include the following research and development tasks: coke examination; effects of irradiation; variability of physical properties (mechanical, thermal-physical, and fracture); fatigue behavior, oxidation behavior; NDE techniques; structural design criteria; and carbon-carbon composite control rod clad materials. These tasks support nuclear grade graphite manufacturing technology including nondestructive examination of billets and components. Moreover, data shall be furnished to support design and licensing of graphite components for the MHTGR

  7. Criticality calculations for a critical assembly, graphite moderate, using 20% enriched uranium

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The construction of a Zero Power Reactor (ZPR) at the Instituto de Energia Atomica in order to measure the neutron characteristics (parameters) of HTGR reactors is proposed. The necessary quantity fissile uranium for these measurements has been calculed. Criticality studies of graphite moderated critical assemblies containing thorium have been made and the critical mass of each of several typical commercial HTGR compositions has been calculated using computer codes HAMMER and CITATION. Assemblies investigated contained a central cylindrical core region, simulating a typical commercial HTGR composition, a uranium-graphite driver region and a outer pure graphite reflector region. It is concluded that a 10Kg inventory of fissile uranium will be required for a program of measurements utilizing each of the several calculated assemblies

  8. Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

    International Nuclear Information System (INIS)

    Le-Guillou, Mael

    2014-01-01

    In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

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

    International Nuclear Information System (INIS)

    Muenzer, W.A.

    1978-01-01

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

  10. 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.

  11. Chapter 8: Exponential experiments on graphite moderated lattices fuelled by natural uranium tubes containing cylindrical graphite cores

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Hoskins, T.A.

    1963-01-01

    Experiments have been carried out using a fuel element comprising a 2.75 in. o.d./2.40 in. i.d. natural uranium tube containing a graphite core of diameter 2.0 in. Values of material buckling and migration area asymmetry for lattices at 7 in., 8 in. and 8/2 in. pitch have been obtained, and correlated with the theory of Syrett (1961) to derive an effective resonance integral for the cored element. By comparison with the resonance integral for the same fuel tube without a core, a value for the constant 'γ' of the theory of Stace (1959) is obtained. (author)

  12. Erosion of graphite surface exposed to hot supersonic hydrogen gas

    Science.gov (United States)

    Sharma, O. P.

    1972-01-01

    A theoretical model based on laminar boundary layer flow equations was developed to predict the erosion rate of a graphite (AGCarb-101) surface exposed to a hot supersonic stream of hydrogen gas. The supersonic flow in the nozzle outside the boundary layer formed over the surface of the specimen was determined by assuming one-dimensional isentropic conditions. An overall surface reaction rate expression based on experimental studies was used to describe the interaction of hydrogen with graphite. A satisfactory agreement was found between the results of the computation, and the available experimental data. Some shortcomings of the model and further possible improvements are discussed.

  13. Graphite-MicroMégas, a tool for DNA modeling

    OpenAIRE

    Hornus , Samuel; Larivière , Damien

    2011-01-01

    National audience; MicroMégas is the current state of an ongoing effort to develop tools for modeling biological assembly of molecules. We here present its DNA modeling part. MicroMégas is implemented as a plug-in to Graphite, which is a research plat- form for computer graphics, 3D modeling and numerical geometry that is developed by members of the ALICE team of INRIA.; Nous décrivons l'outils MicroMégas et les techniques qu'il met en jeu pour la modélisation d'assemblage de molécule, en par...

  14. The effect of radiolytically induced gas pressure on the CO2/CO/graphite system

    International Nuclear Information System (INIS)

    Curtis, A.R.; Faircloth, R.L.; Norwood, K.S.

    1980-03-01

    When radiolysed, carbon dioxide corrodes graphite to produce carbon monoxide, which can blow gas out of the graphite pores. This memorandum demonstrates how to calculate this effect, both analytically and numerically with FACSIMILE, and shows how it depends on dose rate, gas composition, specimen size and graphite diffusivity, for cylindrical geometry. The effect is very small for all cases of interest. (author)

  15. Reference materials for nondestructive assay of special nuclear material. Volume 1. Uranium oxide plus graphite powder

    International Nuclear Information System (INIS)

    Sprinkle, J.K.; Likes, R.N.; Parker, J.L.; Smith, H.A.

    1983-10-01

    This manual describes the fabrication of reference materials for use in gamma-ray-based nondestructive assay of low-density uranium-bearing samples. The sample containers are 2-l bottles. The reference materials consist of small amounts of UO 2 spread throughout a graphite matrix. The 235 U content ranges from 0 to 100 g. The manual also describes the far-field assay procedure used with low-resolution detectors

  16. Synthesis of graphite intercalation compound of group VI metals and uranium hexafluorides

    International Nuclear Information System (INIS)

    Fukui, Toshihiro; Hagiwara, Rika; Ema, Keiko; Ito, Yasuhiko

    1993-01-01

    Systematic investigations were made on the synthesis of graphite intercalation compounds of group VI transition metals (W and Mo) and uranium hexafluorides. The reactions were performed by interacting liquid or gaseous metal hexafluorides with or without elemental fluorine at ambient temperature. The degree of intercalation of these metal fluorides depends on the formation enthalpy of fluorometallate anion from the original metal hexafluoride, as has been found for other intercalation reactions of metal fluorides. (author)

  17. Analysis of Off Gas From Disintegration Process of Graphite Matrix by Electrochemical Method

    International Nuclear Information System (INIS)

    Tian Lifang; Wen Mingfen; Chen Jing

    2010-01-01

    Using electrochemical method with salt solutions as electrolyte, some gaseous substances (off gas) would be generated during the disintegration of graphite from high-temperature gas-cooled reactor fuel elements. The off gas is determined to be composed of H 2 , O 2 , N 2 , CO 2 and NO x by gas chromatography. Only about 1.5% graphite matrix is oxidized to CO 2 . Compared to the direct burning-graphite method, less off gas,especially CO 2 , is generated in the disintegration process of graphite by electrochemical method and the treatment of off gas becomes much easier. (authors)

  18. Dosage of boron traces in graphite, uranium and beryllium oxide; Dosage de traces de bore dans le graphite, l'uranium et l'oxyde de beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Coursier, J [Ecole Nationale Superieure de Physique et Chimie Industrielles, 75 - Paris (France); Hure, J; Platzer, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [French] Le probleme du dosage du bore dans les materiaux servant a la construction de reacteurs nucleaires se pose de la facon suivante: determiner a environ 0,1 ppm pres des quantites de bore de l'ordre de quelques dixiemes de ppm. Nous avons choisit la colorimetrie a la curcumine comme methode de dosage. Pour atteindre les teneurs indiquees, il est necessaire d'effectuer une separation prealable du bore et des materiaux de base, soit par extraction du fluoborate de tetraphenylarsonium dans le cas du dosage de bore dans l'uranium et l'oxyde de beryllium, soit par l'utilisation d'une resine echangeuse de cations dans le cas du graphite. (M.B.)

  19. Graphite

    Science.gov (United States)

    Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

  20. Production yields of noble-gas isotopes from ISOLDE UC$_{x}$/graphite targets

    CERN Document Server

    Bergmann, U C; Catherall, R; Cederkäll, J; Diget, C A; Fraile-Prieto, L M; Franchoo, S; Fynbo, H O U; Gausemel, H; Georg, U; Giles, T; Hagebø, E; Jeppesen, H B; Jonsson, O C; Köster, U; Lettry, Jacques; Nilsson, T; Peräjärvi, K; Ravn, H L; Riisager, K; Weissman, L; Äystö, J

    2003-01-01

    Yields of He, Ne, Ar, Kr and Xe isotopic chains were measured from UC$_{x}$/graphite and ThC$_{x}$/graphite targets at the PSB-ISOLDE facility at CERN using isobaric selectivity achieved by the combination of a plasma-discharge ion source with a water-cooled transfer line. %The measured half-lives allowed %to calculate the decay losses of neutron-rich isotopes in the %target and ion-source system, and thus to obtain information on the in-target %productions from the measured yields. The delay times measured for a UC$_x$/graphite target allow for an extrapolation to the expected yields of very neutron-rich noble gas isotopes, in particular for the ``NuPECC reference elements'' Ar and Kr, at the next-generation radioactive ion-beam facility EURISOL. \\end{abstract} \\begin{keyword} % keywords here, in the form: keyword \\sep keyword radioactive ion beams \\sep release \\sep ion yields \\sep ISOL (Isotope Separation On-Line) \\sep uranium and thorium carbide targets. % PACS codes here, in the form: \\PACS code \\sep code...

  1. Uranium enrichment by gas centrifuge

    International Nuclear Information System (INIS)

    Heriot, I.D.

    1988-01-01

    After recalling the physical principles and the techniques of centrifuge enrichment the report describes the centrifuge enrichment programmes of the various countries concerned and compares this technology with other enrichment technologies like gaseous diffusion, laser, aerodynamic devices and chemical processes. The centrifuge enrichment process is said to be able to replace with advantage the existing enrichment facilities in the short and medium term. Future prospects of the process are also described, like recycled uranium enrichment and economic improvements; research and development needs to achieve the economic prospects are also indicated. Finally the report takes note of the positive aspect of centrifuge enrichment as far as safeguards and nuclear safety are concerned. 27 figs, 113 refs

  2. From USA operation experience of industrial uranium-graphite reactors

    International Nuclear Information System (INIS)

    Burdakov, N.S.

    1996-01-01

    The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

  3. Separation of uranium isotopes by gas centrifugation

    International Nuclear Information System (INIS)

    Jordan, I.

    1980-05-01

    The uranium isotope enrichment is studied by means of the countercurrent gas centrifuge driven by thermal convection. A description is given of (a) the transfer and purification of the uranium hexafluoride used as process gas in the present investigation; (b) the countercurrent centrifuge ZG3; (c) the system designed for the introduction and extraction of the process gas from the centrifuge; (d) the measurement of the process gas flow rate through the centrifuge; (e) the determination of the uranium isotopic abundance by mass spectrometry; (f) the operation and mechanical behavior of the centrifuge and (g) the isotope separation experiments, performed, respectively, at total reflux and with production of enriched material. The results from the separation experiments at total reflux are discussed in terms of the enrichment factor variation with the magnitude and flow profile of the countercurrent given by the temperature difference between the rotor covers. As far as the separation experiments with production are concerned, the discussion of their results is presented through the variation of the enrichment factor as a function of the flow rate, the observed asymmetry of the process and the calculated separative power of the centrifuge. (Author) [pt

  4. Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

    OpenAIRE

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

    2015-01-01

    Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case.

  5. Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

    Energy Technology Data Exchange (ETDEWEB)

    Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

    2013-07-01

    About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

  6. Integral measurements of lattice properties in the natural uranium-graphite critical facility Marius; Mesures globales de reseaux a graphite dans l'empilement critique marius

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    A systematic study of natural uranium-graphite lattices has been undertaken in the critical facility MARIUS, which was built in 1959 in Marcoule. Integral measurement of lattice properties are carried out by the progressive replacement method. This report describes the experimental methods, the analysis of the experiments and the results obtained for lattices with pitches ranging from 192 to 317 mm and fuel elements with cross sections ranging from 6 to 20 cm{sup 2}. The principles of correlation of the results are also outlined. Additional experimental results are also given, pertaining to the determination of the anisotropy, of both the axial and the radial migration areas, and of the age in graphite. (author) [French] L'empilement critique MARIUS, construit en 1959 a Marcoule, a ete utilise pour l'etude systematique des reseaux a graphite-uranium naturel. Les mesures globales de reseaux sont faites par la methode de remplacement progressif. On decrit ici les methodes experimentales utilisees pour ces mesures globales, les principes du depouillement et les resultats obtenus pour des pas de 192 a 317 mm et des combustibles de 6 a 20 cm{sup 2} d'uranium naturel. On donne d'autre part le principe de correlation des mesures. Un certain nombre de resultats experimentaux complementaires sont donnes, en permettant de determiner l'anisotropie, les aires de migration axiale et radiale, l'age dans le graphite. (auteur)

  7. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  8. Gas storage cylinder formed from a composition containing thermally exfoliated graphite

    Science.gov (United States)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

    2012-01-01

    A gas storage cylinder or gas storage cylinder liner, formed from a polymer composite, containing at least one polymer and a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m(exp 2)/g to 2600 m(exp 2)2/g.

  9. Fine structure and spectral index measurements in natural uranium - graphite lattices; Mesures fines dans des reseaux a graphite

    Energy Technology Data Exchange (ETDEWEB)

    Cogne, F; Journet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The experiments described in this report have been carried out for the most part in the critical facility MARIUS, and a few during the start up of the EDF-1 power reactor. The first part deals with the fine structure measurements made in various lattices and with their analysis. Integration over the neutron spectrum of the mono-kinetic disadvantage factor derived by the A.B.H method yields results in good agreement with the experiments. The second part deals with spectral indexes measurements (Pu/U, In/Mn) made at room temperature in MARIUS. Comparison are made of experiments with calculations using various thermalization models. Experiments carried out at higher temperatures in EDF-1 are also described. (authors) [French] Les mesures decrites dans ce rapport ont ete faites pour la plupart dans l'empilement critique MARIUS sur des reseaux a graphite-uranium naturel. Une premiere partie traite des mesures de structure fine faites dans differents reseaux et de leur interpretation. On montre en particulier qu'une integration sur le spectre d'un calcul monocinetique type A.B.H. rend bien compte des experiences. Dans une deuxieme partie, on donne les resultats de mesures d'indices de spectre Pu/U et In/Mn faites sur des reseaux froids a MARIUS et leur comparaison avec les differents modeles de calculs de thermalisation. On donne egalement les resultats de quelques mesures en temperature effectuees lors du demarrage du reacteur EDF-1. (auteurs)

  10. Radioactively induced noise in gas-sampling uranium calorimeters

    International Nuclear Information System (INIS)

    Gordon, H.A.; Rehak, P.

    1982-01-01

    The signal induced by radioactivity of a U 238 absorber in a cell of a gas-sampling uranium calorimeter was studied. By means of Campbell's theorem, the levels of the radioactively induced noise in uranium gas-sampling calorimeters was calculated. It was shown that in order to obtain similar radioactive noise performance as U-liquid argon or U-scintillator combinations, the α-particles from the uranium must be stopped before entering the sensing volume of gas-uranium calorimeters

  11. Chapter 5: Exponential experiments on natural uranium graphite moderated systems. II: Correlation of results with the method of Syrett (1961)

    International Nuclear Information System (INIS)

    Brown, G.; Moore, P.G.F.; Richmond, R.

    1963-01-01

    The results are given of exponential experiments on graphite moderated systems with fuel elements consisting of single rods and tubes of natural uranium metal. A correlation is given with the method of calculation proposed by Syrett (1961) and new consistent values of neutron yield and effective resonance integral are derived. (author)

  12. On the separation of so-called non-volatile uranium fission products of uranium using the conversion of neutron-irradiated uranium dioxide and graphite

    International Nuclear Information System (INIS)

    Elhardt, W.

    1979-01-01

    The investigations are continued in the following work which arose from the concept of separating uranium fission products from uranium. This is achieved in that due to the lattice conversions occurring during the course of solid chemical reactions, fission products can easily pass from the uranium-contained solid to a second solid. The investigations carried out primarily concern the release behaviour of cerium and neodymium in the temperature region of 1200 to 1700 0 C. UO 2 + graphite, both in powder form, are selected as suitable reaction system having the preconditions needed for the lattice conversion for the release effect. The target aimed at from the practical aspect for the improved release of lanthanoids is achieved by an isobar test course - changing temperature from 1200 to 1500 0 C at constant pressure, with a cerium release of 75-80% and a neodynium release of 80-90% (maximum at 1400 0 C). The concepts on the mechanism of the fission product release are related to transport processes in crystal lattices, as well as chemical solid reactions and evaporation processes on the surface of UC 2 grains. (orig./RB) [de

  13. Special graphites; Graphites speciaux

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

  14. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  15. Radioactive Contamination Near Natural Uranium - Graphite - Gas Reactors

    International Nuclear Information System (INIS)

    Chassany, J.; Pouthier, J.

    1967-01-01

    The authors give the results of numerous assessments of contamination in connection with reactors in operation during maintenance; reactors shut down during overhaul and repair work (coolants, exchangers, interior of the tank, etc.) ; and accidents in the cooling circuit and ruptured cladding. They show that, except in special cases, it is mainly activation products that predominate. Moreover, after eight years of operation the points where contamination likely to give considerable dose rates accumulates remain very localized, and there has been no need to reinforce personnel protection measures. (author) [fr

  16. Coating of graphite flakes with MgO/carbon nanocomposite via gas state reaction

    Energy Technology Data Exchange (ETDEWEB)

    Sharif, M., E-mail: Sharif_m@metaleng.iust.ac.i [Iran University of Science and Technology, Tehran (Iran, Islamic Republic of); Faghihi-Sani, M.A. [Sharif University of Technology, Tehran (Iran, Islamic Republic of); Golestani-Fard, F. [Iran University of Science and Technology, Tehran (Iran, Islamic Republic of); Saberi, A. [Tabriz University (Iran, Islamic Republic of); Soltani, Ali Khalife [Iran University of Science and Technology, Tehran (Iran, Islamic Republic of)

    2010-06-18

    Coating of graphite flakes with MgO/carbon nanocomposite was carried out via gaseous state reaction between mixture of Mg metal, CO gas and graphite flakes at 1000 {sup o}C. XRD and FE-SEM analysis of coating showed that the coating was comprised of MgO nano particles and amorphous carbon distributed smoothly and covered the graphite surface evenly. Thermodynamic calculations were employed to predict the reaction sequences as well as phase stability. The effect of coating on water wettability and oxidation resistance of graphite was studied using contact angle measurement and TG analysis, respectively. It was demonstrated that the reaction between Mg and CO could result in MgO/C nanocomposite deposition. The coating improved water wettability of graphite and also enhanced the oxidation resistance of graphite flakes significantly. Also the graphite coating showed significant phenolic resin-wettabilty owing to high surface area of such hydrophilic nano composite coating. The importance of graphite coating is explained with emphasis on its potential application in graphite containing refractories.

  17. Coating of graphite flakes with MgO/carbon nanocomposite via gas state reaction

    International Nuclear Information System (INIS)

    Sharif, M.; Faghihi-Sani, M.A.; Golestani-Fard, F.; Saberi, A.; Soltani, Ali Khalife

    2010-01-01

    Coating of graphite flakes with MgO/carbon nanocomposite was carried out via gaseous state reaction between mixture of Mg metal, CO gas and graphite flakes at 1000 o C. XRD and FE-SEM analysis of coating showed that the coating was comprised of MgO nano particles and amorphous carbon distributed smoothly and covered the graphite surface evenly. Thermodynamic calculations were employed to predict the reaction sequences as well as phase stability. The effect of coating on water wettability and oxidation resistance of graphite was studied using contact angle measurement and TG analysis, respectively. It was demonstrated that the reaction between Mg and CO could result in MgO/C nanocomposite deposition. The coating improved water wettability of graphite and also enhanced the oxidation resistance of graphite flakes significantly. Also the graphite coating showed significant phenolic resin-wettabilty owing to high surface area of such hydrophilic nano composite coating. The importance of graphite coating is explained with emphasis on its potential application in graphite containing refractories.

  18. Management of graphite material: a key issue for High Temperature Gas Reactor system (HTGR)

    International Nuclear Information System (INIS)

    Bourdeloie, C.; Marimbeau, P.; Robin, J.C.; Cellier, F.

    2005-01-01

    Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR, Fig.1) as moderator, thermal absorber and also as structural components of the core (Fig.2). This type of reactor was selected by the Generation IV forum as a potential high temperature provider for supplying hydrogen production plants and is under development in France in the frame of the AREVA ANTARES program. In order to select graphite grades to be used in these future reactors, the requirements for mechanical, thermal, physical-chemical properties must match the internal environment of the nuclear core, especially with regard to irradiation effect. Another important aspect that must be addressed early in design is the waste issue. Indeed, it is necessary to reduce the amount of nuclear waste produced by operation of the reactor during its lifetime. Preliminary assessment of the nuclear waste output for an ANTARES type 280 MWe HTGR over 60 year-lifetime gives an estimated 6000 m 3 of activated graphite waste. Thus, reducing the graphite waste production is an important issue for any HTGR system. First, this paper presents a preliminary inventory of graphite waste fluxes coming from a HTGR, in mass and volume, with magnitudes of radiological activities based on activation calculations of graphite during its stay in the core of the reactor. Normalized data corresponding to an output of 1 GWe.year electricity allows comparison of the waste production with other nuclear reactor systems. Second, possible routes to manage irradiated graphite waste are addressed in both the context of French nuclear waste management rules and by comparison to other national regulations. Routes for graphite waste disposal studied in different countries (concerning existing irradiated graphite waste) will be discussed with regard to new issues of large graphite waste from HTGR. Alternative or complementary solutions aiming at lowering volume of graphite waste to be managed will be presented. For example

  19. The gas centrifuge, uranium enrichment and nuclear proliferation

    International Nuclear Information System (INIS)

    Chapman, A.

    1988-01-01

    The author considers the consequences for controlling nuclear proliferation of the emergence of the gas centrifuge method for enriching uranium and succeeds in the difficult and delicate task of saying enough about gas centrifuge techniques for readers to judge, what may be involved in fully embracing gas centrifuge enrichment within the constraints of an anti-proliferation strategy, whilst at the same time saying nothing that could be construed as encouraging an interest in the gas centrifuge route to highly enriched uranium where none had before existed. (author)

  20. Chapter 3: Exponential experiments on graphite-moderated lattices fuelled with near-natural uranium metal rods

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Clarke, W.G.; Ashworth, F.P.O.; Hoskins, T.A.

    1963-01-01

    Exponential experiments have been carried out on graphite lattices fuelled by 1.2 in. diameter uranium metal rods at three near-natural U 235 compositions, 0.6 Co, 1.3 Co and 1.6 Co. The results, together with those already existing from earlier exponential or critical measurements on these and similar natural uranium rods, have been correlated with the theory of Syrett (1961) and also with the modified form of this theory given in Vol.1, Ch. 7. (author)

  1. Study of new structures adapted to gas-graphite and gas-heavy water reactors

    International Nuclear Information System (INIS)

    Martin, R.; Roche, R.

    1964-01-01

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [fr

  2. Change of properties after oxidation of IG-11 graphite by air and CO2 gas

    International Nuclear Information System (INIS)

    Lim, Yun-Soo; Chi, Se-Hwan; Cho, Kwang-Yun

    2008-01-01

    Artificial graphite is typically manufactured by carbonization of a shaped body of a kneaded mixture using granular cokes as a filler and pitch as a binder. It undergoes a pitch impregnation process if necessary and finally applying graphitization heat treatment. The effect of thermal oxidation in air or a CO 2 atmosphere on IG-11 graphite samples is investigated in this study. The results show a localized oxidation process that progressively reveals the large coke particles with increasing level of overall weight loss in air. The surface of the graphite was peeled off and no change was found in the specific gravity after air oxidation. However, the specific gravity of graphite was continuously decreased by CO 2 oxidation. The decrease in the specific gravity by CO 2 oxidation was due to CO 2 gas that progressed from the surface to the interior. The pore shape after CO 2 oxidation differed from that under air oxidation

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

    International Nuclear Information System (INIS)

    Tench, R.J.

    1992-11-01

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

  4. On residual gas analysis during high temperature baking of graphite tiles

    International Nuclear Information System (INIS)

    Prakash, A A; Chaudhuri, P; Khirwadkar, S; Reddy, D Chenna; Saxena, Y C; Chauhan, N; Raole, P M

    2008-01-01

    Steady-state Super-conducting Tokamak-1 (SST-1) is a medium size tokamak with major radius of 1.1 m and minor radius of 0.20 m. It is designed for plasma discharge duration of 1000 seconds to obtain fully steady-state plasma operation. Plasma Facing Components (PFC), consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be UHV compatible for steady state operation. All PFC are made up of graphite tiles mechanically attached to the copper alloy substrate. Graphite is one of the preferred first wall armour material in present day tokamaks. High thermal shock resistance and low atomic number of carbon are the most important properties of graphite for this application. High temperature vacuum baking of graphite tiles is the standard process to remove the impurities. Residual Gas Analyzer (RGA) has been used for qualitative and quantitative measurements of released gases from graphite tiles during baking. Surface Analysis of graphite tiles has also been done before and after baking. This paper describes the residual gas analysis during baking and surface analysis of graphite tiles

  5. On residual gas analysis during high temperature baking of graphite tiles

    Energy Technology Data Exchange (ETDEWEB)

    Prakash, A A; Chaudhuri, P; Khirwadkar, S; Reddy, D Chenna; Saxena, Y C [Institute for Plasma Research, Bhat, Gandhinagar - 382 428 (India); Chauhan, N; Raole, P M [Facilitation Center for Industrial Plasma Technologies, IPR, Gandhinagar (India)], E-mail: arun@ipr.res.in

    2008-05-01

    Steady-state Super-conducting Tokamak-1 (SST-1) is a medium size tokamak with major radius of 1.1 m and minor radius of 0.20 m. It is designed for plasma discharge duration of 1000 seconds to obtain fully steady-state plasma operation. Plasma Facing Components (PFC), consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be UHV compatible for steady state operation. All PFC are made up of graphite tiles mechanically attached to the copper alloy substrate. Graphite is one of the preferred first wall armour material in present day tokamaks. High thermal shock resistance and low atomic number of carbon are the most important properties of graphite for this application. High temperature vacuum baking of graphite tiles is the standard process to remove the impurities. Residual Gas Analyzer (RGA) has been used for qualitative and quantitative measurements of released gases from graphite tiles during baking. Surface Analysis of graphite tiles has also been done before and after baking. This paper describes the residual gas analysis during baking and surface analysis of graphite tiles.

  6. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  7. Lead determination in uranium mineralization soils by atomic absorption spectrometry with graphite oven

    International Nuclear Information System (INIS)

    Teixeira, Gleber Tacio

    2001-01-01

    The contamination of soils by lead has a great environmental importance due to its toxicity to vegetables, animals and humans. In general, the mobility of the lead is small due to its low solubility and strong adsorption in the soil. However, its solubility can be altered by several conditions (pH, redox potential and ionic stronger). Consequently, lead can migrate through the soil and can contaminate superficial and underground waters. The objective of this work was to determine the concentration of total lead in soil samples with uranium mineralization, in an area at Ipora/GO, having been evaluated as economically insuitable the extraction of that mineral. The radiogenic lead appears as a product of natural radioactive elements decay. In the decay series of uranium-238 we found the isotope lead-214 (half-life of 26,8 min), lead-210 (half-life of 22,3 min), and lead-206 that is stable. The sampling was done in profiles around north, south, east and west directions, starting from a reference point (FT), chosen by presenting the largest radiation of that place (4800 cps). A mass of 1 Kg of superficial soil was collected to each 20 m, in each profile, until 150 m of FT. Approximately, 1 g of dry soil, fraction 2 mm, was digested with a mixture of acids HNO 3 /HClO 4 2:1 (v/v), and the resulting solution was analyzed by atomic absorption. An atomic absorption spectrometer was used with graphite furnace, with deuterium arc to background correction and pyrolytic coated tube. Phosphoric acid was used as chemical modifier. The obtained results, using the standard additions method, presented a decrease of the lead concentration, in all profiles, when the distance of FT was increased. It was also made a radiometric screening in each sampling point. The lead concentration variate from 115,1 μg.g -1 in FT, to less than 40 μg.g -1 at 150 m of distance of FT ( 3 ) 2 was used. The method was applied to a certified sample, showing a good agreement between certified and

  8. Melting of contaminated steel scrap from the dismantling of the CO2 systems of gas cooled, graphite moderated nuclear reactors

    International Nuclear Information System (INIS)

    Feaugas, J.; Jeanjacques, M.; Peulve, J.

    1994-01-01

    G2 and G3 are the natural Uranium cooled reactors Graphite/Gas. The two reactors were designed for both plutonium and electricity production (45 MWe). The dismantling of the reactors at stage 2 has produced more than 4 000 tonnes of contaminated scrap. Because of their large mass and low residual contamination level, the French Atomic Energy Commission (CEA) considered various possibilities for the processing of these metallic products in order to reduce the volume of waste going to be stored. After different studies and tests of several processes and the evaluation of their results, the choice to melt the dismantled pipeworks was taken. It was decided to build the Nuclear Steel Melting Facility known as INFANTE, in cooperation with a steelmaker (AHL). The realization time schedule for the INFANTE lasted 20 months. It included studies, construction and the licensing procedure. (authors). 2 tabs., 3 figs

  9. Special graphites

    International Nuclear Information System (INIS)

    Leveque, P.

    1964-01-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

  10. Graphite and carbon/carbon components for hot gas ducts

    International Nuclear Information System (INIS)

    Popp, G.; Gruber, U.; Boeder, H.; Janssen, K.

    1984-01-01

    The large coal reserves in the Federal Republic of Germany and the uncertainty of the future energy situation on the world market make it appear sound policy to devote some thought to the gasification of coal. For certain chemical processes, moreover, it would be advantageous to have a reasonably priced source of process heat available. In the Federal Republic of Germany this process heat shall be produced in a high-temperature nuclear reactor (HTR), the primary heating temperatures being in the range between 950 deg. C and 1050 deg. C. One serious problem in utilisation of high temperature heat is the resistance of the construction materials. Ceramic materials with high temperature resistance are considered. The material includes graphite and CC carbon fibre reinforced carbon. As a result of the project promoted by Ministerium fur Wirtschaft (Federal Republic of Germany) it has been demonstrated that both CC and graphite manufactured by SIGRI GmbH are well suited for use in high temperature reactors

  11. Graphite fuels combustion off-gas treatment options

    International Nuclear Information System (INIS)

    Kirkham, R.J.; Lords, R.E.

    1993-03-01

    Scenarios for burning bulk graphite and for burning crushed fuel particles from graphite spent nuclear fuels have been considered. Particulates can be removed with sintered metal filters. Subsequent cooling would then condense semi-volatile fission products into or onto a particulate. These particulates would be trapped by a second sintered metal filter or downstream packed bed. A packed bed scrub column can be used to eliminate most of the iodine-129 and tritium. A molecular sieve bed is proposed to collect the residual 129 I and other tramp radionuclides downstream (Ruthenium, etc.). Krypton-85 can be recovered, if need be, either by cryogenics or by the KALC process (Krypton Adsorption in Liquid Carbon dioxide). Likewise carbon-14 in the form of carbon dioxide could be collected with a caustic or lime scrub solution and incorporated into a grout. Sulfur dioxide present will be well below regulatory concern level of 4.0 tons per year and most of it would be removed by the scrubber. Carbon monoxide emissions will depend on the choice of burner and start-up conditions. Should the system exceed the regulatory concern level, a catalytic converter in the final packed bed will be provided. Radon and its daughters have sufficiently short half-lives (less than two minutes). If necessary, an additional holdup bed can be added before the final HEPA filters or additional volume can be added to the molecular sieve bed to limit radon emissions. The calculated total effective dose equivalent at the Idaho National Engineering Laboratory boundary from a single release of all the 3 , 14 C, 85 Kr, and 129 I in the total fuel mass if 0.43 mrem/year

  12. Gas chromatographic method fr determination of carbon in metallic uranium

    International Nuclear Information System (INIS)

    Nikol'skij, V.A.; Markov, V.K.; Evseeva, T.I.; Cherstvenkova, E.P.

    1983-01-01

    Gas chromatographic device to determine carbon in metal uranium is developed. Burnout unite, permitting to load in the burnout tube simultaneously quite a few (up to 20) weight amounts of materials to be burned is a characteristic feature of the device. As a result amendments for control experiment and determination limit are decreased. The time of a single determination is also reduced. Conditions of carbon burn out from metal uranium are studied and temperature and time of complete extraction of carbon in the form of dioxide from weight amount into gaseous phase are established

  13. Investigation of Pristine Graphite Oxide as Room-Temperature Chemiresistive Ammonia Gas Sensing Material

    Directory of Open Access Journals (Sweden)

    Alexander G. Bannov

    2017-02-01

    Full Text Available Graphite oxide has been investigated as a possible room-temperature chemiresistive sensor of ammonia in a gas phase. Graphite oxide was synthesized from high purity graphite using the modified Hummers method. The graphite oxide sample was investigated using scanning electron microscopy, energy dispersive X-ray spectroscopy, X-ray diffraction, thermogravimetry and differential scanning calorimetry. Sensing properties were tested in a wide range of ammonia concentrations in air (10–1000 ppm and under different relative humidity levels (3%–65%. It was concluded that the graphite oxide–based sensor possessed a good response to NH3 in dry synthetic air (ΔR/R0 ranged from 2.5% to 7.4% for concentrations of 100–500 ppm and 3% relative humidity with negligible cross-sensitivity towards H2 and CH4. It was determined that the sensor recovery rate was improved with ammonia concentration growth. Increasing the ambient relative humidity led to an increase of the sensor response. The highest response of 22.2% for 100 ppm of ammonia was achieved at a 65% relative humidity level.

  14. Determination of gas residues in uranium dioxide pellets

    International Nuclear Information System (INIS)

    Riella, H.G.

    1978-01-01

    The measurement of low amounts of residual gases, excluding water, in ceramic grade uranium dioxide pellets, using high temperature vacuum extraction technique, is dealt with. The high temperature extraction gas analysis apparatus was designed and assembled for sequential analysis of up to eight uranium dioxide pellets by run. The system consists of three major units, namely outgassing unit, transfer unit and analytical unit. The whole system is evacuated to a final pressure of less then 10 -5 torr. A weighed pellet is transfered into the outgassing unit for subsequent dropping into a Platinum-Rhodium crucible which is heated inductively up to 1600 0 C during 30 minutes. The released gases are imediately transfered from the outgassing to analytical unit passing through a cold trap at -95 0 C to remove water vapor. The gases are transfered to previously calibrated volumetric bulb where the total pressure and temperature are determined. An estimate of the gas content in the pellets at STP condition is obtained from the measured volume, pressure and temperature of the gas mixture by applying ideal gases equation. Analysis to two lots (fourteen samples) of uranium dioxide pellets by the method described here indicated a mean gas content of 0,060cm 3 /g UO 2 . The lower limit of this technique is 0,03cm 3 /g UO 2 (STP). The time required for the analysis of eight pellets is about 9 hours [pt

  15. The status of graphite development for gas cooled reactors

    International Nuclear Information System (INIS)

    1993-02-01

    The meeting was convened by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors. It was attended by 61 participants from 6 countries. The meeting covered the following subjects: overview of national programs; design criteria, fracture mechanisms and component test; materials development and properties; non-destructive examination, inspection and surveillance. The participants presented 33 papers on behalf of their countries. A separate abstract was prepared for each of these papers. Refs, figs, tabs, photos and diagrams

  16. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    International Nuclear Information System (INIS)

    Dubourg, M.

    1993-01-01

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

  17. Disintegration of graphite matrix from the simulative high temperature gas-cooled reactor fuel element by electrochemical method

    International Nuclear Information System (INIS)

    Tian Lifang; Wen Mingfen; Li Linyan; Chen Jing

    2009-01-01

    Electrochemical method with salt as electrolyte has been studied to disintegrate the graphite matrix from the simulative high temperature gas-cooled reactor fuel elements. Ammonium nitrate was experimentally chosen as the appropriate electrolyte. The volume average diameter of disintegrated graphite fragments is about 100 μm and the maximal value is less than 900 μm. After disintegration, the weight of graphite is found to increase by about 20% without the release of a large amount of CO 2 probably owing to the partial oxidation to graphite in electrochemical process. The present work indicates that the improved electrochemical method has the potential to reduce the secondary nuclear waste and is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end of reprocessing.

  18. Electrochemically induced chemical sensor properties in graphite screen-printed electrodes: The case of a chemical sensor for uranium

    International Nuclear Information System (INIS)

    Kostaki, Vasiliki T.; Florou, Ageliki B.; Prodromidis, Mamas I.

    2011-01-01

    Highlights: → Electrochemical treatment endows analytical characteristics to SPEs. → A sensitive chemical sensor for uranium is described. → Performance is due to a synergy between electrochemical treatment and ink's solvents. → The amount of the solvent controls the achievable sensitivity. - Abstract: We report for the first time on the possibility to develop chemical sensors based on electrochemically treated, non-modified, graphite screen-printed electrodes (SPEs). The applied galvanostatic treatment (5 μA for 6 min in 0.1 M H 2 SO 4 ) is demonstrated to be effective for the development of chemical sensors for the determination of uranium in aqueous solutions. A detailed study of the effect of various parameters related to the fabrication of SPEs on the performance of the resulting sensors along with some diagnostic experiments on conventional graphite electrodes showed that the inducible analytical characteristics are due to a synergy between electrochemical treatment and ink's solvents. Indeed, the amount of the latter onto the printed working layer controls the achievable sensitivity. The preconcentration of the analyte was performed in an electroless mode in an aqueous solutions of U(VI), pH 4.6, and then, the accumulated species was reduced by means of a differential pulse voltammetry scan in 0.1 M H 3 BO 3 , pH 3. Under selected experimental conditions, a linear calibration curve over the range 5 x 10 -9 to 10 -7 M U(VI) was constructed. The 3σ limit of detection at a preconcentration time of 30 min, and the relative standard deviation of the method were 4.5 x 10 -9 M U(VI) and >12% (n = 5, 5 x 10 -8 M U(VI)), respectively. The effect of potential interferences was also examined.

  19. Uranium fluoride chemistry. Part 1. The gas phase reaction of uranium hexafluoride with alcohols

    International Nuclear Information System (INIS)

    Schnautz, N.G.; Venter, P.J.

    1992-01-01

    The reaction between uranium hexafluoride (UF 6 ) and simple alcohols in the gas phase was observed to proceed by way of three possible reaction pathways involving dehydration, deoxygenative fluorination, and ether formation. These reactions can best be explained by assuming that alcohols first react with UF 6 to afford the alkoxy uranium pentafluoride intermediate ROUF 5 , which reacts further to give the dehydration, deoxygenative fluorination, and ether products. In each of the above three reaction pathways, UF 6 is transformed to UOF 4 , which being as reactive toward alcohols as UF 6 , reacts further with the alcohol in question to finally afford the unreactive uranyl fluoride (UO 2 F 2 ). 6 refs., 2 tabs

  20. Uranium enrichment in Europe by the gas centrifuge process

    International Nuclear Information System (INIS)

    Severin, D.J.E.

    1975-01-01

    To begin with, this lesson gives an outline of the expected energy demand of the Western World and the concentration of the European companies participating in uranium enrichment by the gas centrifuge method. Next, a) the principles of the gas centrifuge method are outlined, b) its advantages over other industrial processes are stressed, and c) the characteristic data of complete plants are given. The existing German, Dutch, and British pilot plants are mentioned as examples for the perfected state of the process. The Capenhurst (UK) and Almedo (NL) demonstration plants, each with a capacity of 200 t SW/a, will have been extended to 2 x 1.000 t SW/a by 1982. Finally, economic data of the gas centrifuge process are given. The term 'separative work' is explained in an annex. (GG) [de

  1. Projected uranium measurement uncertainties for the Gas Centrifuge Enrichment Plant

    International Nuclear Information System (INIS)

    Younkin, J.M.

    1979-02-01

    An analysis was made of the uncertainties associated with the measurements of the declared uranium streams in the Portsmouth Gas Centrifuge Enrichment Plant (GCEP). The total uncertainty for the GCEP is projected to be from 54 to 108 kg 235 U/year out of a measured total of 200,000 kg 235 U/year. The systematic component of uncertainty of the UF 6 streams is the largest and the dominant contributor to the total uncertainty. A possible scheme for reducing the total uncertainty is given

  2. Recent developments concerning French fuel elements used in natural uranium - graphite - CO{sub 2} reactor systems; Developpements recents des elements combustibles francais de la filiere uranium naturel - graphite - CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Salesse, M; Stohr, J A; Jeanpierre, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The policy followed in France for the development of fuel elements for reactors belonging to the Electricite de France has been to benefit as much as possible, for each new pile from the most recent technical progress by developing in each case a fuel element allowing the maximum power per channel. The two latest fuel elements thus studied by the French Atomic Energy Commission are of two different types: a tubular uranium element closed at both ends and cooled externally. (This type of element, chosen for the reactors EDF 2, EDF 3 and EDF 4 makes it possible to attain maximum specific powers of the order of 6 MW/metric ton.); an open tubular uranium element cooled both internally and externally, called an annular element which in being studied as a possibility for EDF 5. Such an element makes it possible to attain specific powers of over 12 MW/metric ton. The two types of element have the following common characteristics: - the can, for external cooling, has herning-bone type fins. This type of profile which has been vastly improved recently thereby increasing its thermal efficiency, has the important advantage of avoiding vibration of the element, but has posed problems of resistance to thermal cycling necessitating much research. - the fuel rods are placed inside graphite jackets, this limiting the vertical forces to which they are subjected and protecting them during charging and discharging. On the other hand, these elements present very different problems as for as the following points are concerned: - the characteristics required of the uranium tubes apart of course from a good dimensional stability during irradiation in the two cases are in the case of the closed tubes a very high resistance to external pressure, and in the case of the annular elements a low neutron absorption. Thus for each of these two cases it has been necessary to develop a suitable type of alloy. - a possible loosening of the can during thermal cycling, which is peculiar to the

  3. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

  4. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  5. Graphite-supported platinum catalysts: Effects of gas and aqueous phase treatments

    Energy Technology Data Exchange (ETDEWEB)

    Vleeming, J.H.; Kuster, B.F.M.; Marin, G.B. [Eindhoven Univ. of Technology (Netherlands)] [and others

    1997-03-01

    The effects on the platinum particle diameter and the available platinum surface area of a graphite-supported platinum catalyst resulting from pretreatments and from performing a selective oxidation reaction are investigated. In the gas phase considerable catalyst sintering occurs only in the presence of oxygen at 773 K due to extensive carbon burn-off, whereas in an aqueous phase platinum particle growth is limited upon oxidative treatment. A hydrogen treatment in aqueous phase at 363 K causes platinum particle growth, aggregate formation, and covering of metal sites. These phenomena become more important with increasing pH. Platinum particle growth and aggregate formation are attributed to platinum particle rather than platinum adatom mobility and is caused by the destruction of the oxygen-containing surface groups on the graphite support, which serve as anchorage sites for the platinum particles. Site covering is caused by products originating from the graphite support, which are formed as a result of the reductive treatments. When performing the aqueous phase oxidation of methyl {alpha}-D-glucopyranoside at 323 K and a pH of 9, catalyst modifications are small under oxidative conditions. Exposure of the catalyst for several hours to methyl {alpha}-D-glucopyranoside under the same conditions but in the absence of oxygen causes site covering. 50 refs., 9 figs., 1 tab.

  6. Analysis of graphite dust deposition in hot gas duct of HTGR

    International Nuclear Information System (INIS)

    Peng Wei; Zhen Ya'nan; Yang Xiaoyong; Ye Ping

    2013-01-01

    The behavior of the graphite dust is important to the safety of high-temperature gas-cooled reactor (HTGR). The temperature field in hot gas duct was obtained using computational fluid dynamics (CFD) method. Further analysis to the thermo-phoretic deposition and turbulent deposition shows that as the dust particle diameter increases, the thermo-phoretic deposition efficiency decreases, and the turbulent deposition efficiency initially decreases and then increases. The comparisons of calculation results for two reactor powers, namely 30% FP (full power) and 100 % FP, indicate that the thermo-phoretic deposition efficiency is higher at 30% FP than that at 100% FP. while the turbulent deposition efficiency grows more rapidly at 100% FP. Besides, the results also demonstrate that the thermo-phoretic deposition and the turbulent deposition are nearly equivalent when particle sizes are small, while the turbulent deposition becomes dominant when particle sizes are fairly large. The calculation results by using the most probable distribution of particle size show that the total deposition of graphite dusts in hot gas duct is limited. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-15

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

  8. Designing and analysis study of uranium enrichment with gas centrifuge

    International Nuclear Information System (INIS)

    Tsunetoshi Kai

    2006-01-01

    This note concerns a designing and analysis study of uranium enrichment with a gas centrifuge. At first, one dimensional model is presented and a conventional analytical method is applied to grasp the general idea of a centrifuge performance. Secondly, two-dimensional numerical method is adopted to describe the diffusion phenomena with assumption of simple flow patterns. Parametric surveys are made on the dimension of a centrifuge rotor, the gas feed, withdrawal and circulation system, and operation variables such as feed flow rate, cut and so on. Thirdly, full numerical solutions are obtained for the flow and diffusion equations in static state, using a modified version of the Newton method without neglect of any non-linear term. The numerical results are compared with the experimental data made by Beams et al. and Zippe, and found to be in good agreement. Further, the theoretical pressure and separative power are compared respectively with experimental ones on a comparatively recent centrifuge. The results reveal that the characteristics of separation performance of a centrifuge can be fully described by the present method. Some of inevitable problems are tackled regarding UF 6 gas isotope separation by centrifugation. To examine the influence of the extraneous light gas, the diffusion equations for ternary mixture are solved and also the flow field of binary mixture with large mass difference is obtained to simultaneously solve the Navier-Stokes equations and the diffusion equation.for binary case. Since the gas in the interior region of the rotor is so rarefied that the Navier-Stokes equations cease to be valid, the Burnett equations are solved.for gas flow in a rotating cylinder. Considering that the uranium recovered at a reprocessing plant includes 236 U besides 235 U and 238 U, the concentration distributions of the ternary gas isotopes are determined and a value function is defined for the evaluation of separative work for the multi-component mixture

  9. Chapter 4: Exponential experiments on natural uranium graphite moderated systems. I: Performance and analysis

    International Nuclear Information System (INIS)

    Brown, G.; Moore, P.G.F.; Richmond, R.

    1963-01-01

    A description is given of the methods used on the first BICEP stack for the performance and analysis of graphite exponential experiments. These differ in many respects from the methods formerly employed at A.E.R.E. The accuracy of the measurements has been increased, and the time taken to carry out and analyse an experiment has been reduced by approximately a factor of four. The following have contributed to the experimental work reported in this paper: J. R. Dyson, W. M. Holderay, R. M. Turner, S. D. R. Summers. (author)

  10. Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors; Etude de la tenue de la gaine interne pour-element combustible a refroidissement interne et externe d'un reacteur graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    Boudouresque, B; Courcon, P; Lestiboubois, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm{sup 2} gas pressure, should remain in contact with the fuel. (authors) [French] La cartouche d'un element combustible annulaire, a refroidissement interne et externe pour reacteur graphite-gaz, est composee d'un tube combustible en uranium, d'une gaine externe et d'une gaine interne en alliage de magnesium. Pour que l'echange thermique entre la gaine interne et le combustible soit bon, il faut que la gaine reste appliquee sur l'uranium quel que soit le regime de temperature. Cette note a pour but de montrer comment, d'apres une etude theorique, le jeu combustible-gaine interne varie au cours des operations de gainage, de chargement dans le reacteur, et des cyclages thermiques. Les parametres suivants sont etudies: diametres de tube, pression du gaz caloporteur, temperature d'entree du gaz, plasticite de l'alliage de gaine. Il est montre que, quel que soit le regime de fonctionnement, la gaine interne d'un element 77 x 95, en projet pour un reacteur graphite-gaz sous pression de 40 kg/cm{sup 2}, doit rester appliquee sur le combustible. (auteurs)

  11. Individual hollow and mesoporous aero-graphitic microtube based devices for gas sensing applications

    Science.gov (United States)

    Lupan, Oleg; Postica, Vasile; Marx, Janik; Mecklenburg, Matthias; Mishra, Yogendra K.; Schulte, Karl; Fiedler, Bodo; Adelung, Rainer

    2017-06-01

    In this work, individual hollow and mesoporous graphitic microtubes were integrated into electronic devices using a FIB/SEM system and were investigated as gas and vapor sensors by applying different bias voltages (in the range of 10 mV-1 V). By increasing the bias voltage, a slight current enhancement is observed, which is mainly attributed to the self-heating effect. A different behavior of ammonia NH3 vapor sensing by increasing the applied bias voltage for hollow and mesoporous microtubes with diameters down to 300 nm is reported. In the case of the hollow microtube, an increase in the response was observed, while a reverse effect has been noticed for the mesoporous microtube. It might be explained on the basis of the higher specific surface area (SSA) of the mesoporous microtube compared to the hollow one. Thus, at room temperature when the surface chemical reaction rate (k) prevails on the gas diffusion rate (DK) the structures with a larger SSA possess a higher response. By increasing the bias voltage, i.e., the overall temperature of the structure, DK becomes a limiting step in the gas response. Therefore, at higher bias voltages the larger pores will facilitate an enhanced gas diffusion, i.e., a higher gas response. The present study demonstrates the importance of the material porosity towards gas sensing applications.

  12. Determination of hydrogen in zirconium hydride and uranium-zirconium hydride by inert gas exraction-gravimetric method

    International Nuclear Information System (INIS)

    Hoshino, Akira; Iso, Shuichi

    1976-01-01

    An inert gas extraction-gravimetric method has been applied to the determination of hydrogen in zirconium hydride and uranium-zirconium hydride which are used as neutron moderator and fuel of nuclear safety research reactor (NSRR), respectively. The sample in a graphite-enclosed quartz crucible is heated inductively to 1200 0 C for 20 min in a helium stream. Hydrogen liberated from the sample is oxidized to water by copper(I) oxide-copper(II) oxide at 400 0 C, and the water is determined gravimetrically by absorption in anhydrone. The extraction curves of hydrogen for zirconium hydride and uranium-zirconium hydride samples are shown in Figs. 2 and 3. Hydrogen in the samples is extracted quantitatively by heating at (1000 -- 1250) 0 C for (10 -- 40) min. Recoveries of hydrogen in the case of zirconium hydride were examined as follows: a weighed zirconium rod (5 phi x 6 mm, hydrogen -5 Torr. After the chamber was filled with purified hydrogen to 200 Torr, the rod was heated to 400 0 C for 15 h, and again weighed to determine the increase in weight. Hydrogen in the rod was then determined by the proposed method. The results are in excellent agreement with the increase in weight as shown in Table 1. Analytical results of hydrogen in zirconium hydride samples and an uranium-zirconium hydride sample are shown in Table 2. (auth.)

  13. Oxidation damage evaluation by non-destructive method for graphite components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Tada, Tatsuya; Sumita, Junya; Sawa, Kazuhiro

    2008-01-01

    To develop non-destructive evaluation methods for oxidation damage on graphite components in High Temperature Gas-cooled Reactors (HTGRs), the applicability of ultrasonic wave and micro-indentation methods were investigated. Candidate graphites, IG-110 and IG-430, for core components of Very High Temperature Reactor (VHTR) were used in this study. These graphites were oxidized uniformly by air at 500degC. The following results were obtained from this study. (1) Ultrasonic wave velocities with 1 MHz can be expressed empirically by exponential formulas to burn-off, oxidation weight loss. (2) The porous condition of the oxidized graphite could be evaluated with wave propagation analysis with a wave-pore interaction model. It is important to consider the non-uniformity of oxidized porous condition. (3) Micro-indentation method is expected to determine the local oxidation damage. It is necessary to assess the variation of the test data. (author)

  14. Influence of Ambient Gas on Laser-Induced Breakdown Spectroscopy of Uranium Metal

    International Nuclear Information System (INIS)

    Zhang Dacheng; Ma Xinwen; Wang Shulong; Zhu Xiaolong

    2015-01-01

    Laser-induced breakdown spectroscopy (LIBS) is regarded as a suitable method for the remote analysis of materials in any phase, even in an environment with high radiation levels. In the present work we used the third harmonic pulse of a Nd:YAG laser for ablation of uranium metal and measured the plasma emission with a fiber-optic spectrometer. The LIBS spectra of uranium metal and their features in different ambient gases (i.e., argon, neon, oxygen, and nitrogen) at atmospheric pressure were studied. Strong continuum spectrum and several hundreds of emission lines from UI and UII were observed. It is found that the continuum spectrum observed in uranium not only comes from bremsstrahlung emission but is also due to the complex spectrum of uranium. The influence of ambient gas and the gas flow rate for ablation of uranium metal was investigated. The experimental results indicate that the intensity of the uranium lines was enhanced in argon and nitrogen. However, the intensity of uranium lines was decreased in oxygen due to the generation of UO and other oxides. The results also showed that the highest intensity of uranium lines were obtained in argon gas with a gas flow rate above 2.5 L/min. The enhanced mechanism in ambient gas and the influence of the gas flow rate were analyzed in this work. (paper)

  15. Development of uranium enrichment technology by gas centrifugation

    International Nuclear Information System (INIS)

    Sibata, Tomofumi; Kai, Tsunetoshi

    1996-01-01

    The development of a gas-centrifuge for uranium enrichment has been conducted by Power Reactor and Nuclear Fuel Development Corporation in Japan after the first several years' fruitless works, the R and D works came to the point and continuing rapid improvements of centrifuges have started, Cascade tests were given with C-1 and C-2 cascade experimental facilities. Life, reliability and feasibility tests were given with the pilot plant and the demonstration plant. As a result of these works, the private commercial plant has started the operation. Although the main efforts were devoted to the development of metal rotor centrifuges in the course mentioned above, composite material rotor centrifuges have also been developed in parallel to achieve higher performance. Promising results have been being obtained with cascade test facilities on the pilot plant scale. Furthermore, R and D works are being proceeded on more excellent and advanced centrifuges. (author)

  16. Spatial flux instabilities, and their control in the graphite gas power reactors; Les instabilites spatiales du flux et leur controle dans les reacteurs de puissance graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    Radial-azimuthal and axial spatial flux instabilities in graphite-gas reactors are studied by means of an analytical approach. Results are checked with those which are given by two dimensional (r, z and r, {theta}) kinetic models programmed for an IBM 7094 computer. At least, conclusions on the control of instabilities obtained from these models are reported. (author) [French] Les instabilites spatiales du flux dans les reacteurs graphite-gaz, radiales et azimutales d'une part, axiales d'autre part, sont etudiees au moyen d'une formulation analytique. Les resultats sont confrontes avec ceux que fournissent des modeles cinetiques a deux dimensions (r, z et r, {theta}) programmes sur IBM 7094. On donne enfin les conclusions relatives au controle de ces instabilites que ces modeles ont permis de degager. (auteur)

  17. A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

    International Nuclear Information System (INIS)

    Baltazar, O.

    1978-01-01

    A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt

  18. The behaviour of CAGR moderator and sleeve graphites radiolytically oxidised to high weight loss in inhibited coolant gas compositions

    International Nuclear Information System (INIS)

    Schofield, P.; Fitzgerald, B.; Ketchen, J.

    1987-01-01

    Gilsocarbon graphites were irradiated to high weight losses in three different CO 2 based coolants. The experimental data is tested against a model which interprets the gas phase chemistry and pore geometry and allows weight loss and gas flow properties to be calculated. The observed changes of oxidation rate with dose were successfully predicted from the model. An empirical relationship was also derived which was shown to fit data for moderator, sleeve and special pore structure graphites. Changes in graphite permeability and diffusivity were predicted by the model, and also by other simplified, more approximate methods. The model based upon the measured transport pore spectrum was shown to be the best with other methods proving adequate to moderate doses. (author)

  19. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

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

  20. International safeguards at the feed and withdrawal area of a gas centrifuge uranium enrichment plant

    International Nuclear Information System (INIS)

    Gordon, D.M.; Sanborn, J.B.

    1980-01-01

    This paper discusses the application of International Atomic Energy Agency (IAEA) safeguards at a model gas centrifuge uranium enrichment plant designed for the production of low-enriched uranium; particular emphasis is placed upon the verification by the IAEA of the facility material balance accounting. 13 refs

  1. Sulfur-doped graphene via thermal exfoliation of graphite oxide in H2S, SO2, or CS2 gas.

    Science.gov (United States)

    Poh, Hwee Ling; Šimek, Petr; Sofer, Zdeněk; Pumera, Martin

    2013-06-25

    Doping of graphene with heteroatoms is an effective way to tailor its properties. Here we describe a simple and scalable method of doping graphene lattice with sulfur atoms during the thermal exfoliation process of graphite oxides. The graphite oxides were first prepared by Staudenmaier, Hofmann, and Hummers methods followed by treatments in hydrogen sulfide, sulfur dioxide, or carbon disulfide. The doped materials were characterized by scanning electron microscopy, high-resolution X-ray photoelectron spectroscopy, combustible elemental analysis, and Raman spectroscopy. The ζ-potential and conductivity of sulfur-doped graphenes were also investigated in this paper. It was found that the level of doping is more dramatically influenced by the type of graphite oxide used rather than the type of sulfur-containing gas used during exfoliation. Resulting sulfur-doped graphenes act as metal-free electrocatalysts for an oxygen reduction reaction.

  2. Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2003-02-01

    This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

  3. Neutron study of fast neutron reactor systems by exponential experiments on Harmonie - Graphite program HUG-PHUG - Oxide program PHRIXOS - Uranium program UK

    International Nuclear Information System (INIS)

    Desprets, Alain.

    1977-12-01

    Exponential experiments allow to obtain the fundamental characteristics of a lattice (material buckling, reaction rate ratios) more economically than critical experiments. This report describes the experimental techniques and the methods of analysis used for this type of experiments. The results obtained with three programs performed with the source reactor HARMONIE are given: graphite-lattices program (3 U-fueled and 3 Pu-fueled lattices); oxide-fuel program (4 PuO 2 -UO 2 lattices); pure uranium program (one lattice). Some of these lattices were also studied in critical experiments. The coherence of the results obtained by the two types of experiments is established [fr

  4. Preconcentration of trace elements from high-purity thorium and uranium on Chelex-100 and determination by graphite furnace atomic absorption spectrometry with Zeeman-effect background correction

    International Nuclear Information System (INIS)

    Raje, Naina; Kayasth, Satish; Asari, T.P.S.; Gangadharan, S.

    1994-01-01

    Preconcentration of trace impurities from large-sized samples of uranium metal and thorium oxide using a small column of Chelex-100 followed by their determination using graphite furnace atomic absorption spectrometry (GFAAS) is reported. A 0.5-10-g amount of the sample (uranium metal or thorium oxide) was dissolved, complexed with ammonium carbonate and subjected to the ion-exchange procedure. The retained analytes were eluted with 2-4 M nitric acid and brought to a small volume for a final dilution to 10-25 ml for their determination using GFAAS. The validity of the separation procedure and recoveries at μg kg -1 levels was checked by standard addition; the recoveries were >95%

  5. Uranium

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  6. Impact of radiolysis and radiolytic corrosion on the release of {sup 13}C and {sup 37}Cl implanted into nuclear graphite: Consequences for the behaviour of {sup 14}C and {sup 36}Cl in gas cooled graphite moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moncoffre, N., E-mail: nathalie.moncoffre@ipnl.in2p3.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Toulhoat, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Silbermann, G. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); and others

    2016-04-15

    Graphite finds widespread use in many areas of nuclear technology based on its excellent moderator and reflector qualities as well as its strength and high temperature stability. Thus, it has been used as moderator or reflector in CO{sub 2} cooled nuclear reactors such as UNGG, MAGNOX, and AGR. However, neutron irradiation of graphite results in the production of {sup 14}C (dose determining radionuclide) and {sup 36}Cl (long lived radionuclide), these radionuclides being a key issue regarding the management of the irradiated waste. Whatever the management option (purification, storage, and geological disposal), a previous assessment of the radioactive inventory and the radionuclide's location and speciation has to be made. During reactor operation, the effects of radiolysis are likely to promote the radionuclide release especially at the gas/graphite interface. Radiolysis of the coolant is mainly initiated through γ irradiation as well as through Compton electrons in the graphite pores. Radiolysis can be simulated in laboratory using γ irradiation or ion irradiation. In this paper, {sup 13}C, {sup 37}Cl and {sup 14}N are implanted into virgin nuclear graphite in order to simulate respectively the presence of {sup 14}C, {sup 36}Cl and nitrogen, a {sup 14}C precursor. Different irradiation experiments were carried out using different irradiation devices on implanted graphite brought into contact with a gas simulating the coolant. The aim was to assess the effects of gas radiolysis and radiolytic corrosion induced by γ or He{sup 2+} irradiation at the gas/graphite interface in order to evaluate their role on the radionuclide release. Our results allow inferring that radiolytic corrosion has clearly promoted the release of {sup 14}C, {sup 36}Cl and {sup 14}N located at the graphite brick/gas interfaces and open pores.

  7. An investigation of heat exchanger fouling in dust suspension cooling systems using graphite powder and carbon dioxide gas

    International Nuclear Information System (INIS)

    Garton, D.A.; Hawes, R.I.; Rose, P.W.

    1966-01-01

    Some experiments have been performed to study the fouling of heat exchanger surfaces where heat is being transferred from a heated fluid to a cooled surface. The fluid studied was a suspension of 4-5 microns mean diameter graphite powder in carbon dioxide gas at near atmospheric pressures. The solids loading range covered was from 5 to 30 lb. graphite/lb. carbon dioxide, and gas Reynolds numbers from 6000 to 16000. Temperature gradients across the cooler of from 20 to 120 deg. C were obtained. The heat transfer ratio is correlated to show the dependence upon the solids loading ratio of the suspension, the gas Reynolds number and the temperature gradient across the cooler. The results have demonstrated that stringent precautions are necessary to ensure complete dryness of the graphite powder and the loop flow surfaces before any quantitative fouling data can be obtained, as the presence of entrained moisture will accelerate the deposition of material on the cold walls of the heat exchanger and can result in plugging. The heat transfer coefficient showed no obvious dependency upon either the gas Reynolds number or the temperature gradient across the cooler over the range investigated. The measured heat transfer coefficient was considerably lower than that obtained when the heat is transferred from a hot wall to a cooler fluid. At a solids loading of 30 lb, graphite/lb. carbon dioxide, the heat transfer coefficient was only 50% of that for heat transfer from a heated wall. At solids loadings below 7 lb/lb., the heat transfer was less than that for a gas alone. (author)

  8. The feasibility and prospect of uranium-gas in black rock series of joint exploration and development

    International Nuclear Information System (INIS)

    Xu Guochang; Zhang Dehua; Zhang Hongjian

    2014-01-01

    By the analysis and contrast of existing form of gas-uranium, correlation between gas-uranium and organic matter, distribution characteristics and control factors of mineralization (bosom) in the sedimentary formation of shale gas and black uranium bearing rock series, the authors come to the conclusion that: in the carbonate-siliceous-pelitic of black rock series the uranium and gas (oil) is essentially equipped coenosarc of the same homology, syngenetic, reservoir. They are ore source beds of carbonate-siliceous-pelitic rock uranium deposit, and also the hydrocarbon source beds in which the shale gas form. In black shales, uranium largely exist in the form of the ion adsorption (acetyl ion/uranyl ion). Under fracturing conditions, we can realize desorption mode of chemical solvents of adding acid or alkali, and extract uranium by concentrating liquid (the same as in-situ mimng technology). Therefore, the fracturing technology (clear water fracturing techniques, repeat fracturing techniques, synchronization fracturing techniques, multistage fracturing techniques, network fracturing techniques and so on) of shale gas exploitation lay a foundation for black shale uranium-gas joint development. The mature and corollary use of fracturing techniques and in-situ mining technology of low grade uranium will undoubtedly further increase the industrial resource extent of uranium and gas, improve guaranteeing degree of resource, reform of promote energy production structure and provide a large number of economical and effective clean energy. (authors)

  9. Treatment and Disposal of the Radioactive Graphite Waste of High-Temperature Gas-Cooled Reactor Spent Fuel

    International Nuclear Information System (INIS)

    Li Junfeng

    2016-01-01

    High-temperature gas-cooled reactors (HTGRs) represent one of the Gen IV reactors in the future market, with efficient generation of energy and the supply of process heat at high temperature utilised in many industrial processes. HTGR development has been carried out within China’s National High Technology Research and Development Program. The first industrial demonstration HTGR of 200 MWe is under construction in Shandong Province China. HTGRs use ceramic-coated fuel particles that are strong and highly resistant to irradiation. Graphite is used as moderator and helium is used as coolant. The fuel particles and the graphite block in which they are imbedded can withstand very high temperature (up to ~1600℃). Graphite waste presents as the fuel element components of HTGR with up to 95% of the whole element beside the graphite blocks in the core. For example, a 200 MWe reactor could discharge about 90,000 fuel elements with 17 tonnes irradiated graphite included each year. The core of the HTGR in China consists of a pebble bed with spherical fuel elements. The UO 2 fuel kernel particles (0.5mm diameter) (triple-coated isotropic fuel particles) are coated by several layers including inner buffer layer with less dense pyrocarbon, dense pyro-carbon, SiC layer and outer layer of dense pyro-carbon, which can prevent the leaking of fission products (Fig. 1). Spherical fuel elements (60mm diameter) consist of a 50mm diameter inner zone and 5mm thick shell of fuel free zone [3]. The inner zone contains about 8300 triple-coated isotropic fuel particles of 0.92mm in diameter dispersed in the graphite matrix

  10. Brazing graphite to graphite

    International Nuclear Information System (INIS)

    Peterson, G.R.

    1976-01-01

    Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

  11. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

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

  12. Uranium

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  13. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

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

  14. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

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

  15. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

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

  16. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

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

  17. Development and industrial application of gas centrifuges to uranium enrichment in the USSR

    International Nuclear Information System (INIS)

    Abbakumov, E.I.; Bazhenov, V.A.; Verbin, Yu.V.

    1989-01-01

    Review of state and studies in the field of gaseous diffusion technology and centrifugal method of uranium enrichment in the USSR is given. Domestic industrial gas centrifuges, forming to-day the main part of separation capacities in the USSR, are noted for low specific energy consumption and high reliability. Centrifugal technology in the USSR is applied both to uranium enrichment (including one for export) and to separation of isotopes of other chemical elements

  18. Numerical analysis of mass transfer with graphite oxidation in a laminar flow of multi-component gas mixture through a circular tube

    International Nuclear Information System (INIS)

    Ogawa, Masuro

    1992-10-01

    In the present paper, mass transfer has been numerically studied in a laminar flow through a circular graphite tube to evaluate graphite corrosion rate and generation rate of carbon monoxide during a pipe rupture accident in a high temperature gas cooled reactor. In the analysis, heterogeneous (graphite oxidation and graphite/carbon dioxide reaction) and homogeneous (carbon monoxide combustion) chemical reactions were dealt in the multi-component gas mixture; helium, oxygen, carbon monoxide and carbon dioxide. Multi-component diffusion coefficients were used in a diffusion term. Mass conservation equations of each gas component, mass conservation equation and momentum conservation equations of the gas mixture were solved by using SIMPLE algorism. Chemical reactions between graphite and oxygen, graphite and carbon dioxide, and carbon monoxide combustion were taken into account in the present numerical analysis. An energy equation for the gas mixture was not solved and temperature was held to be constant in order to understand basic mass transfer characteristics without heat transfer. But, an energy conservation equation for single component gas was added to know heat transfer characteristics without mass transfer. The effects of these chemical reactions on the mass transfer coefficients were quantitatively and qualitatively clarified in the range of 50 to 1000 of inlet Reynolds numbers, 0 to 0.5 of inlet oxygen mass fraction and 800 to 1600degC of temperature. (author)

  19. The problem of gas gap between graphite - fuel channel reduction impact at Ignalina NPP

    International Nuclear Information System (INIS)

    1999-01-01

    Safety analysis of Ignalina NPP operation in the case when gap closure between graphite - fuel channel occur was performed. The main results of this analysis as well as data of gap measurements during the year 1996 - 1998 are provided

  20. Trapping and detrapping of hydrogen in graphite materials exposed to hydrogen gas

    International Nuclear Information System (INIS)

    Atsumi, Hisao; Iseki, Michio; Shikama, Tatsuo.

    1994-01-01

    Measurements of hydrogen solubility have been performed for several unirradiated and neutron-irradiated graphite (and CFC) samples at temperatures between 973 and 1323 K under a ∼10 kPa hydrogen atmosphere. The hydrogen dissolution process has been studied and it is discussed here. The values of hydrogen solubility vary substantially among the samples up to about a factor of 16. A strong correlation has been observed between the values of hydrogen solubility and the degrees of graphitization determined by X-ray diffraction technique. The relation can be extended even for the neutron irradiated samples. Hydrogen dissolution into graphite can be explained with the trapping of hydrogen at defect sites (e.g. dangling carbon bonds) considering an equilibrium reaction between hydrogen molecules and the trapping sites. The migration of hydrogen in graphite is speculated to result from a sequence of detrapping and retrapping events with high energy activation processes. (author)

  1. Tables of formulae for calculating the mechanics of stacks in gas-graphite reactors

    International Nuclear Information System (INIS)

    1968-01-01

    This collection of formulae only gives, for nuclear graphite stacks. The mechanical effects due to the strains, thermal or not, of steel structures supporting or surrounding graphite blocks. Equations have been established by mean of experiments made at Chinon with large pile models. Thus, it is possible to calculate displacement, strain and stress in the EDF type stacks of horizontal triangular block lattice. (authors) [fr

  2. Study on efficient methods for removal and treatment of graphite blocks in a gas cooled reactor

    International Nuclear Information System (INIS)

    Fujii, S.; Shirakawa, M.; Murakami, T.

    2001-01-01

    Tokai Power Station (GCR, 166 MWe) started its commercial operation on July 1966 and ceased activities at the end of March 1998 after 32 years of operation. The decommissioning plans are being developed, to prepare for near future dismantling. In the study, the methods for removal of the graphite blocks of about 1,600 ton have been developed to carrying it out safely and in a short period of time, and the methods of treatment of graphite have also been developed. All technological items have been identified for which R and D work will be required for removal from the core and treatment for disposal. (1) In order to reduce the programme required for the dismantling of reactor internals, an efficient method for removal of the graphite blocks is necessary. For this purpose the design of a dismantling machine has been investigated which can extract several blocks at a time. The conceptual design has being developed and the model has been manufactured and tested in a mock-up facility. (2) In order to reduce disposal costs, it will be necessary to segment the graphite blocks, maximising the packing density available in the disposal containers. Some of the graphite blocks will be cut into pieces longitudinally by a remote machine. Relevant technical matters have been identified, such as graphite cutting methods, the nature of fine particles arising from the cutting operation, the treatment of fine particles for disposal, and the method of mortar filling inside the waste container. (author)

  3. {sup 36}Cl and {sup 14}C behaviour in UNGG graphite during leaching experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pichon, C.; Guy, C.; Comte, J. [Commissariat a l' Energie Atomique - C.E.A., Laboratoire d' Analyses Radiochimiques et Chimiques (L.A.R.C.) 13108 Saint Paul lez Durance (France)

    2008-07-01

    Graphite has been used as a moderator in Natural Uranium Graphite Gas reactors. Among the radionuclides, the long-lived activation product {sup 36}Cl and {sup 14}C, which are abundant in graphite after irradiation can be the main contributors to the dose during disposal. This paper deals with the first results obtained on irradiated graphite from French G2 reactor. Both leaching and diffusion experiments have been performed in order to understand and quantify the radionuclides behaviour. Chlorine leaching seems to be controlled by diffusion transport through graphite matrix. On the contrary {sup 14}C leaching is very low, probably because after irradiation, the remaining {sup 14}C was produced from {sup 13}C activation in the crystalline structure of graphite. (authors)

  4. Effects of reaction temperature and inlet oxidizing gas flow rate on IG-110 graphite oxidation used in HTR-PM

    International Nuclear Information System (INIS)

    Sun Ximing; Dong Yujie; Zhou Yangping; Shi Lei; Sun Yuliang; Zhang Zuoyi; Li Zhengcao

    2017-01-01

    The oxidation behavior of a selected nuclear graphite (IG-110) used in Pebble-bed Module High Temperature gas-cooled Reactor was investigated under the condition of air ingress accident. The oblate rectangular specimen was oxidized by oxidant gas with oxygen mole fraction of 20% and flow rates of 125–500 ml/min at temperature of 400–1200°C. Experiment results indicate that the oxidation behavior can also be classified into three regimes according to temperature. The regime I at 400–550°C has lower apparent activation energies of 75.57–138.59 kJ/mol when the gas flow rate is 125–500 ml/min. In the regime II at 600–900°C, the oxidation rate restricted by the oxygen supply to graphite is almost stable with the increase of temperature. In the regime III above 900°C, the oxidation rate increases obviously with the increase of temperature.With the increase of inlet gas flow from 125 to 500 ml/min, the apparent activation energy in regime I is increased and the stableness of oxidation rate in regime II is reduced. (author)

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

    International Nuclear Information System (INIS)

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

    1965-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-15

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

  7. Influence of uncertainties of isotopic composition of the reprocessed uranium on effectiveness of its enrichment in gas centrifuge cascades

    Science.gov (United States)

    Smirnov, A. Yu; Mustafin, A. R.; Nevinitsa, V. A.; Sulaberidze, G. A.; Dudnikov, A. A.; Gusev, V. E.

    2017-01-01

    The effect of the uncertainties of the isotopic composition of the reprocessed uranium on its enrichment process in gas centrifuge cascades while diluting it by adding low-enriched uranium (LEU) and waste uranium. It is shown that changing the content of 232U and 236U isotopes in the initial reprocessed uranium within 15% (rel.) can significantly change natural uranium consumption and separative work (up to 2-3%). However, even in case of increase of these parameters is possible to find the ratio of diluents, where the cascade with three feed flows (depleted uranium, LEU and reprocessed uranium) will be more effective than ordinary separation cascade with one feed point for producing LEU from natural uranium.

  8. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

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

  9. Natural uranium metallic fuel elements: fabrication and operating experience

    International Nuclear Information System (INIS)

    Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

    1980-01-01

    The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

  10. Interpretation of measurements made by oscillations of irradiated fuels in natural uranium, graphite-gas piles

    International Nuclear Information System (INIS)

    Laponche, Bernard; Luffin, Jean; Brunet, Max; Guerange, Jacques; Tonolli, Jacky

    1969-06-01

    When considering a pile operation, it is interesting to know the evolution of fuel quality with respect to irradiation, i.e. the variation of its fission rate and of its absorption rate. In order to experimentally obtain these features, a method is to introduce an irradiated cartridge into a critical reactor and to measure the induced effect on its reactivity and on the neutron density at the vicinity of the cartridge. An oscillation method presented in another document and based on a periodic introduction of fuel sample into a critical reactor allows, from the measurement of reactivity variation (global signal), and of the neutron density (local signal), effective macroscopic fission and absorption cross sections of this sample to be obtained. As previous studies revealed that the interpretation of the local signal was notably delicate, this information has been replaced by computed information, the fission rate, which is determined by means of the COREGRAF1 code. Thus, the remaining quantity to be obtained is the fuel absorption rate. The authors report studies performed on several sets of cartridges from different reactors, and with an irradiation range from about 700 to 4000 MWJ/T. In a first part, they describe the characteristics of the studied cartridges, their irradiation and measurement conditions, and the use of the evolution code. In a second part, they try to define the interpretation of oscillation-based measurements by using two methods, a first and fast one which gives an approximation of results, and a more elaborated second one which complies with measurement conditions. The last part presents and discusses the obtained results [fr

  11. Radio-active pollution near natural uranium-graphite-gas reactors

    International Nuclear Information System (INIS)

    Chassany, J.; Pouthier, J.; Delmar, J.

    1967-01-01

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors) [fr

  12. Study of new structures adapted to gas-graphite and gas-heavy water reactors; Etude de structures nouvelles adaptees aux reacteurs graphite-gaz et eau lourde-gaz

    Energy Technology Data Exchange (ETDEWEB)

    Martin, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [French] L'experience acquise par l'exploitation des reacteurs de MARCOULE, la construction et le demarrage des reacteurs d

  13. Starting up a programme of atomic piles using compressed gas

    International Nuclear Information System (INIS)

    Horowitz, J.; Yvon, J.

    1959-01-01

    1) An examination of the intellectual and material resources which have directed the French programme towards: a) the natural uranium and plutonium system, b) the use of compressed gas as heat transfer fluid (primary fluid). 2) The parts played in exploring the field by the pile EL2 and G1, EL2 a natural uranium, heavy water and compressed gas pile, G1 a natural uranium, graphite and atmospheric air pile. 3) Development of the neutronics of graphite piles: physical study of G1. 4) The examination of certain problem posed by centres equipped with natural uranium, graphite and compressed carbon dioxide piles: structure, special materials, fluid circuits, maximum efficiency. Economic aspects. 5) Aids to progress: a) piles for testing materials and for tests on canned fuel elements, b) laboratory and calculation facilities. 6) Possible new orientations of compressed gas piles: a) raising of the pressure, b) enriched fuel, c) higher temperatures, d) use of heavy water. (author) [fr

  14. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

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

  15. Radio-active pollution near natural uranium-graphite-gas reactors; La pollution radioactive aupres des piles uranium naturel - graphite - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J.; Pouthier, J.; Delmar, J. [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

    1967-07-01

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors) [French] Les resultats de nombreuses evaluations de la contamination sont donnes: - Piles en marche pendant les operations d'entretien - Piles a l'arret au cours des chantiers caracteristiques (refrigerants, echangeurs, interieur du caisson, etc.) - A la suite d'incidents sur le circuit de refroidissement et de rupture de gaine. Ils montrent que, sauf cas particulier, ce sont essentiellement les produits d'activation qui dominent. Par ailleurs apres 10 ans de fonctionnement, les points d'accumulation de la contamination susceptibles de delivrer des debits de dose importants restent tres localises et les moyens de protection individuels utilises n'ont pas du etre renforces. (auteurs)

  16. Radio-active pollution near natural uranium-graphite-gas reactors; La pollution radioactive aupres des piles uranium naturel - graphite - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J; Pouthier, J; Delmar, J [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

    1967-07-01

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors) [French] Les resultats de nombreuses evaluations de la contamination sont donnes: - Piles en marche pendant les operations d'entretien - Piles a l'arret au cours des chantiers caracteristiques (refrigerants, echangeurs, interieur du caisson, etc.) - A la suite d'incidents sur le circuit de refroidissement et de rupture de gaine. Ils montrent que, sauf cas particulier, ce sont essentiellement les produits d'activation qui dominent. Par ailleurs apres 10 ans de fonctionnement, les points d'accumulation de la contamination susceptibles de delivrer des debits de dose importants restent tres localises et les moyens de protection individuels utilises n'ont pas du etre renforces. (auteurs)

  17. A new mechanistic and engineering fission gas release model for a uranium dioxide fuel

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Yang, Yong Sik; Kim, Dae Ho; Kim, Sun Ki; Bang, Je Geun

    2008-01-01

    A mechanistic and engineering fission gas release model (MEGA) for uranium dioxide (UO 2 ) fuel was developed. It was based upon the diffusional release of fission gases from inside the grain to the grain boundary and the release of fission gases from the grain boundary to the external surface by the interconnection of the fission gas bubbles in the grain boundary. The capability of the MEGA model was validated by a comparison with the fission gas release data base and the sensitivity analyses of the parameters. It was found that the MEGA model correctly predicts the fission gas release in the broad range of fuel burnups up to 98 MWd/kgU. Especially, the enhancement of fission gas release in a high-burnup fuel, and the reduction of fission gas release at a high burnup by increasing the UO 2 grain size were found to be correctly predicted by the MEGA model without using any artificial factor. (author)

  18. Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

    International Nuclear Information System (INIS)

    Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

    2003-01-01

    The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

  19. Treatment of back flow fluids from shale gas exploration with recovery of uranium

    International Nuclear Information System (INIS)

    Gajda, D.; Zakrzewska-Koltuniewicz, G.; Abramowska, A.; Kiegiel, K.; Niescior-Borowinska, P.; Miskiewicz, A.; Olszewska, W.; Kulisa, K.; Samszynski, Z.; Drzewicz, P.; Konieczynska, M.

    2015-01-01

    Shale gas exploitation is the cause of many social protests. According to the protesters gas extraction technology threatens the environment: it consumes huge amounts of water, creates danger of poisoning drinking water, the formation of toxic wastewater, air contamination, noise, etc. Hydro-fracturing fluids could also leach radioactive isotopes e.g. uranium from the rock. The upper content of the main elements found in examined back flow fluids in Poland are the following: chlorine: 100.00 Kg/m 3 , sodium: 40.00 kg/m 3 , potassium: 0.90 kg/m 3 , lithium: 0.15 kg/m 3 , magnesium: 2.00 kg/m 3 , calcium: 20.00 kg/m 3 , strontium: 0.80 kg/m 3 and cesium: 0.06 kg/m 3 while the upper content of trace elements are the following: uranium: 3.5 g/m 3 , lanthanum: 12.4 g/m 3 , vanadium: 1.3 g/m 3 , yttrium: 1.3 g/m 3 , molybdenum: 2.0 g/m 3 and manganese: 9.7 g/m 3 . The recovery of uranium, and other valuable metals, from back flow fluids will reduce an environmental impact of hydro-fracturing process. This poster details the treatment of back flow fluids in Poland allowing rare earth elements and uranium recovery

  20. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

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

  1. Uranium

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  2. Safeguards considerations for uranium enrichment facilities, as applied to gas centrifuge and gaseous diffusion facilities

    International Nuclear Information System (INIS)

    1979-03-01

    The goals and objectives of IAEA safeguards as they are understood by the authors based on published documents are reviewed. These goals are then used to derive safeguards concerns, diversion strategies, and potential safeguards measures for four base cases, the production of highly enriched uranium (HEU) at a diffusion plant, the diversion of low enriched uranium (LEU) at a diffusion plant, the diversion of HEU at a gas centrifuge plant, and the diversion of LEU at a gas centrifuge plant. Tables of estimated capabilities are given for each case, under the assumption that the inspector would have access: to the cascade perimeter at or after the start of operations, to the cascade perimeter throughout construction and operation, to the cascade perimeter during operation plus a one-time access to the cascade itself, to the cascade during construction but only its perimeter during operation, or to the cascade itself during construction and operation

  3. Remediation of Uranium in the Hanford Vadose Zone Using Ammonia Gas: FY 2010 Laboratory-Scale Experiments

    International Nuclear Information System (INIS)

    Szecsody, James E.; Truex, Michael J.; Zhong, Lirong; Qafoku, Nikolla; Williams, Mark D.; McKinley, James P.; Wang, Zheming; Bargar, John; Faurie, Danielle K.; Resch, Charles T.; Phillips, Jerry L.

    2010-01-01

    This investigation is focused on refining an in situ technology for vadose zone remediation of uranium by the addition of ammonia (NH3) gas. Objectives are to: (a) refine the technique of ammonia gas treatment of low water content sediments to minimize uranium mobility by changing uranium surface phases (or coat surface phases), (b) identify the geochemical changes in uranium surface phases during ammonia gas treatment, (c) identify broader geochemical changes that occur in sediment during ammonia gas treatment, and (d) predict and test injection of ammonia gas for intermediate-scale systems to identify process interactions that occur at a larger scale and could impact field scale implementation. Overall, NH3 gas treatment of low-water content sediments appears quite effective at decreasing aqueous, adsorbed uranium concentrations. The NH3 gas treatment is also fairly effective for decreasing the mobility of U-carbonate coprecipitates, but shows mixed success for U present in Na-boltwoodite. There are some changes in U-carbonate surface phases that were identified by surface phase analysis, but no changes observed for Na-boltwoodite. It is likely that dissolution of sediment minerals (predominantly montmorillonite, muscovite, kaolinite) under the alkaline conditions created and subsequent precipitation as the pH returns to natural conditions coat some of the uranium surface phases, although a greater understanding of these processes is needed to predict the long term impact on uranium mobility. Injection of NH3 gas into sediments at low water content (1% to 16% water content) can effectively treat a large area without water addition, so there is little uranium mobilization (i.e., transport over cm or larger scale) during the injection phase.

  4. Remediation of Uranium in the Hanford Vadose Zone Using Ammonia Gas: FY 2010 Laboratory-Scale Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Szecsody, James E.; Truex, Michael J.; Zhong, Lirong; Qafoku, Nikolla; Williams, Mark D.; McKinley, James P.; Wang, Zheming; Bargar, John; Faurie, Danielle K.; Resch, Charles T.; Phillips, Jerry L.

    2010-12-01

    This investigation is focused on refining an in situ technology for vadose zone remediation of uranium by the addition of ammonia (NH3) gas. Objectives are to: a) refine the technique of ammonia gas treatment of low water content sediments to minimize uranium mobility by changing uranium surface phases (or coat surface phases), b) identify the geochemical changes in uranium surface phases during ammonia gas treatment, c) identify broader geochemical changes that occur in sediment during ammonia gas treatment, and d) predict and test injection of ammonia gas for intermediate-scale systems to identify process interactions that occur at a larger scale and could impact field scale implementation.Overall, NH3 gas treatment of low-water content sediments appears quite effective at decreasing aqueous, adsorbed uranium concentrations. The NH3 gas treatment is also fairly effective for decreasing the mobility of U-carbonate coprecipitates, but shows mixed success for U present in Na-boltwoodite. There are some changes in U-carbonate surface phases that were identified by surface phase analysis, but no changes observed for Na-boltwoodite. It is likely that dissolution of sediment minerals (predominantly montmorillonite, muscovite, kaolinite) under the alkaline conditions created and subsequent precipitation as the pH returns to natural conditions coat some of the uranium surface phases, although a greater understanding of these processes is needed to predict the long term impact on uranium mobility. Injection of NH3 gas into sediments at low water content (1% to 16% water content) can effectively treat a large area without water addition, so there is little uranium mobilization (i.e., transport over cm or larger scale) during the injection phase.

  5. Analysis and optimization of gas-centrifugal separation of uranium isotopes by neural networks

    Directory of Open Access Journals (Sweden)

    Migliavacca S.C.P.

    2002-01-01

    Full Text Available Neural networks are an attractive alternative for modeling complex problems with too many difficulties to be solved by a phenomenological model. A feed-forward neural network was used to model a gas-centrifugal separation of uranium isotopes. The prediction showed good agreement with the experimental data. An optimization study was carried out. The optimal operational condition was tested by a new experiment and a difference of less than 1% was found.

  6. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

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

  7. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

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

  8. Theory of uranium enrichment by the gas centrifuge

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D R [California Univ., Berkeley (USA). Lawrence Berkeley Lab.; California Univ., Berkeley (USA). Dept. of Nuclear Engineering)

    1981-01-01

    Onsager's analysis of the hydrodynamics of fluid circulation in the boundary layer on the rotor wall of a gas centrifuge is reviewed. The description of the flow in the boundary layers on the top and bottom end caps due to Carrier and Maslen is summarized. The method developed by Wood and Morton of coupling the flow models in the rotor wall and end cap boundary layers to complete the hydrodynamic analysis of the centrifuge is presented. Mechanical and thermal methods of driving the internal gas circulation are described. The isotope enrichment which results from the superposition of the elementary separation effect due to the centrifugal field in the gas and its internal circulation is analyzed by the Onsager-Cohen theory. The performance function representing the optimized separative power of a centrifuge as a function of throughput and cut is calculated for several simplified internal flow models. The use of asymmetric ideal cascades to exploit the distinctive features of centrifuge performance functions is illustrated.

  9. Electricity generation with natural gas or with uranium?

    International Nuclear Information System (INIS)

    Villanueva M, C.

    2009-10-01

    The program of works and investments of electric sector that actualize each year the Federal Commission of Electricity, include to the projects of electric power generating stations that will begin its commercial operation inside the horizon of the next ten years, in order to satisfy opportunely with appropriate reservation margins the demand of power and energy in the national interconnected system that grows year to year. In spite of its inherent advantages, in the electric sector prospective 2008-2017 are not considered explicitly to the nuclear power plants, except for the small amplification of capacity of nuclear power plant of Laguna Verde, that already is executing. In this context, the objective of this work is to present and to discuss arguments to favor and against the combined cycle and nuclear technologies, to indicate the risks and disadvantages in that it incurs the electric sector when leaning on so disproportionately on the fossil fuels for the electricity generation, in particular the natural gas, deferring to an indefinite future the installation of nuclear plants whose proven technology is economic, sure, clean and reliable and it contributes decisively to the national energy security. To mitigate the harmful effects of excessive dependence on natural gas to generate electric power, was propose alternatives to the expansion program of electric sector to year 2017, which would have as benefits the decrease of the annual total cost of electric power supply for public service, the significant reduction of natural gas imports and emissions reduction of CO 2 to the atmosphere. (Author)

  10. A review of selected aspects of the effect of water vapor on fission gas release from uranium oxycarbide

    International Nuclear Information System (INIS)

    Myers, B.F.

    1994-04-01

    A selective review is presented of previous measurements and the analysis of experiments on the effect of water vapor on fission gas release from uranium oxycarbide. Evidence for the time-dependent composition of the uranium oxycarbide fuel; the diffusional release of fission gas; and the initial, rapid and limited release of stored fission gas is discussed. In regard to the initial, rapid release of fission gas, clear restrictions on mechanistic hypotheses can be deduced from the experimental data. However, more fundamental experiments may be required to establish the mechanism of the rapid release

  11. Optimization of the cascade with gas centrifuges for uranium enrichment

    International Nuclear Information System (INIS)

    Ozaki, N.; Harada, I.

    1976-01-01

    Computer programs to optimize the step and tapered-step cascades with gas centrifuges are developed. The 'Complex Method', one of the direct search method, is employed to find the optimum of the nonlinear function of several variables within a constrained region. The separation characteristics of the optimized step and tapered-step cascades are discussed in comparison with that of the ideal cascade. The local optima of the cascade profile, the convergence of the object function, and the stopping criterion for the optimization trial are also discussed. (author)

  12. The Influences of Time and Velocity of Inert Gas on the Quality of theProcessing Product of Graphite Matrix on the Baking Step

    International Nuclear Information System (INIS)

    Imam-Dahroni; Dwi-Herwidhi; NS, Kasilani

    2000-01-01

    The research of the synthesis of matrix graphite on the step of bakingprocess was conducted, by focusing on the influence of time and velocityvariables of the inert gas. The investigation on baking times ranging from 5minutes to 55 minutes and by varying the velocity of inert gas from 0.30l/minute to 3.60 l/minute, resulted the product of different matrix.Optimizing at the time of operation and the flow rate of argon gas indicatedthat the baking time for 30 minutes and by the flow rate of argon gas of 2.60l/minute resulted best matrix graphite that has a hardness value of 11kg/mm 2 of hardness and the ductility of 1800 Newton. (author)

  13. The migration of intra-granular fission gas bubbles in irradiated uranium dioxide

    International Nuclear Information System (INIS)

    Baker, C.

    1977-05-01

    The mobility of intragranular fission gas bubbles in uranium dioxide irradiated at 1600-1800 0 C has been studied following isothermal annealing at temperatures below 1600 0 C. The intragranular fission gas bubbles, average diameter approximately 2nm, are virtually immobile at temperatures below 1500 0 C. The bubbles have clean surfaces with no solid fission product contamination and are faceted to the highest observed irradiation temperature of 1800 0 C. This bubble faceting is believed to be a major cause of bubble immobility. In fuel operating below 1500 0 C the predominant mechanism allowing the growth of intragranular bubbles and the subsequent gas release must be the diffusion of dissolved gas atoms rather than the movement of entire intragranular bubbles. (author)

  14. Chloride pyrometallurgy of uranium ore. 2. Selective extraction of uranium using the mixed gas of chlorine and oxygen in the presence of activated carbon

    International Nuclear Information System (INIS)

    Taki, Tomihiro; Komoto, Shigetoshi; Otomura, Keiichiro; Takenaka, Toshihide; Sato, Nobuaki; Fujino, Takeo.

    1996-01-01

    Selective extraction of uranium from a phosphate ore was studied by using the mixed gas of Cl 2 and O 2 . On heating the ore and carbon mixture in Cl 2 , the volatilized chloride of uranium is accompanied by iron, aluminum, phosphorus and silicon chlorides. The addition of O 2 gas effectively lowered the volatilization ratios of aluminum, phosphorus and silicon. The ratio decreased with increasing oxygen flow rate up to 50 ml/min at 1,223 K (Cl 2 : 100 ml/min, O 2 +N 2 : 400 ml/min). The volatilization ratio of uranium was almost unchanged at 90% up to 50 ml/min O 2 (carbon amount: 15 wt%). The effect of the other parameters, i.e. Cl 2 flow rate, carbon amount, reaction temperature and time was examined. (author)

  15. Japan-IAEA sefeguards demonstration programme in the gas centrifuge uranium enrichment facility

    International Nuclear Information System (INIS)

    Akiba, Mitsunori; Iwamoto, Tomonori; Omae, Masayoshi

    1985-01-01

    The Hexa-partite Safequard Project was started for the purpose of examining the effective techniques of safeguards for gas centrifuge uranium enrichment facilities. By the proposal of respective participating countries, it was decided to carry out the verifying test of various safeguard techniques at the actual plants. Japan carried out the verifying test of safeguard techniques at the Ningyotoge uranium enrichment pilot plant in June, 1982, and from November, 1983, to August, 1984. The contents of this test is reported. In Japan, this verifying test was positioned as a part of JASPAS (Japanese project of supporting IAEA safeguards). The verifying test of realtime and in-operation inventories, the verifying test of IAEA load cell type weighing machines for UF 6 cylinders, the verifying test of the measurement of the degree of enrichment in UF 6 cylinders by nondestructive test, the verifying test of confinement/watch system, and the verifying test of IAEA gas phase uranium enrichment monitors were carried out. The results were presented as the data for examination in the HSP, and evaluated as useful, informative and well compiled. It is necessary to pursue more cost-effective approaches. (Kako, I.)

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

    International Nuclear Information System (INIS)

    1979-01-01

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

  17. Mitigation of Hydrogen Gas Generation from the Reaction of Water with Uranium Metal in K Basins Sludge

    International Nuclear Information System (INIS)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2010-01-01

    Means to decrease the rate of hydrogen gas generation from the chemical reaction of uranium metal with water were identified by surveying the technical literature. The underlying chemistry and potential side reactions were explored by conducting 61 principal experiments. Several methods achieved significant hydrogen gas generation rate mitigation. Gas-generating side reactions from interactions of organics or sludge constituents with mitigating agents were observed. Further testing is recommended to develop deeper knowledge of the underlying chemistry and to advance the technology aturation level. Uranium metal reacts with water in K Basin sludge to form uranium hydride (UH3), uranium dioxide or uraninite (UO2), and diatomic hydrogen (H2). Mechanistic studies show that hydrogen radicals (H·) and UH3 serve as intermediates in the reaction of uranium metal with water to produce H2 and UO2. Because H2 is flammable, its release into the gas phase above K Basin sludge during sludge storage, processing, immobilization, shipment, and disposal is a concern to the safety of those operations. Findings from the technical literature and from experimental investigations with simple chemical systems (including uranium metal in water), in the presence of individual sludge simulant components, with complete sludge simulants, and with actual K Basin sludge are presented in this report. Based on the literature review and intermediate lab test results, sodium nitrate, sodium nitrite, Nochar Acid Bond N960, disodium hydrogen phosphate, and hexavalent uranium [U(VI)] were tested for their effects in decreasing the rate of hydrogen generation from the reaction of uranium metal with water. Nitrate and nitrite each were effective, decreasing hydrogen generation rates in actual sludge by factors of about 100 to 1000 when used at 0.5 molar (M) concentrations. Higher attenuation factors were achieved in tests with aqueous solutions alone. Nochar N960, a water sorbent, decreased hydrogen

  18. Mitigation of Hydrogen Gas Generation from the Reaction of Water with Uranium Metal in K Basins Sludge

    Energy Technology Data Exchange (ETDEWEB)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2010-01-29

    Means to decrease the rate of hydrogen gas generation from the chemical reaction of uranium metal with water were identified by surveying the technical literature. The underlying chemistry and potential side reactions were explored by conducting 61 principal experiments. Several methods achieved significant hydrogen gas generation rate mitigation. Gas-generating side reactions from interactions of organics or sludge constituents with mitigating agents were observed. Further testing is recommended to develop deeper knowledge of the underlying chemistry and to advance the technology aturation level. Uranium metal reacts with water in K Basin sludge to form uranium hydride (UH3), uranium dioxide or uraninite (UO2), and diatomic hydrogen (H2). Mechanistic studies show that hydrogen radicals (H·) and UH3 serve as intermediates in the reaction of uranium metal with water to produce H2 and UO2. Because H2 is flammable, its release into the gas phase above K Basin sludge during sludge storage, processing, immobilization, shipment, and disposal is a concern to the safety of those operations. Findings from the technical literature and from experimental investigations with simple chemical systems (including uranium metal in water), in the presence of individual sludge simulant components, with complete sludge simulants, and with actual K Basin sludge are presented in this report. Based on the literature review and intermediate lab test results, sodium nitrate, sodium nitrite, Nochar Acid Bond N960, disodium hydrogen phosphate, and hexavalent uranium [U(VI)] were tested for their effects in decreasing the rate of hydrogen generation from the reaction of uranium metal with water. Nitrate and nitrite each were effective, decreasing hydrogen generation rates in actual sludge by factors of about 100 to 1000 when used at 0.5 molar (M) concentrations. Higher attenuation factors were achieved in tests with aqueous solutions alone. Nochar N960, a water sorbent, decreased hydrogen

  19. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    International Nuclear Information System (INIS)

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  20. Gas centrifuge uranium enrichment programme in the United States of America

    International Nuclear Information System (INIS)

    Gestson, D.K.

    1983-01-01

    The technology of uranium enrichment using the gas centrifuge is fully proven as a result of over twenty years of research. The high performance of the centrifuge has been confirmed, and its reliability established, through detailed evaluation of a series of centrifuge designs. The baseline centrifuge for the Gas Centrifuge Enrichment Plant (GCEP) is now in commercial production by three qualified manufacturers. It will be ready for installation in GCEP on schedule. The GCEP construction is also on schedule, with two process buildings expected to start operation in 1988 and 1989. Development and demonstration of the Set IV advanced gas centrifuge is under way and it is expected to be ready for installation in Process Building 3 in early 1989. (author)

  1. Tables of formulae for calculating the mechanics of stacks in gas-graphite reactors; Formulaire pour le calcul de la mecanique des empilements des reacteurs graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-07-01

    This collection of formulae only gives, for nuclear graphite stacks. The mechanical effects due to the strains, thermal or not, of steel structures supporting or surrounding graphite blocks. Equations have been established by mean of experiments made at Chinon with large pile models. Thus, it is possible to calculate displacement, strain and stress in the EDF type stacks of horizontal triangular block lattice. (authors) [French] Le domaine de ce formulaire est strictement limite aux effets mecaniques, pour les empilements, des deformations, thermiques ou autres, des structures metalliques de soutien (aire - support et corset). On propose un ensemble de relations qui ont ete etablies a la suite des essais de CHINON sur des maquettes de grande taille. Ces relations permettent le calcul des mouvements, des deformations et des contraintes dans les empilements du type EDF, a reseau horizontal triangulaire regulier. (auteurs)

  2. Radon as a natural tracer for gas transport within uranium waste rock piles

    International Nuclear Information System (INIS)

    Silva, N.C.; Chagas, E.G.L.; Dias, D.C.S.; Guerreiro, E.T.Z.; Alberti, H.L.C.; Braz, M.L.; Abreu, C.B.; Lopez, D.; Branco, O.; Fleming, P.

    2014-01-01

    Acid mine drainage (AMD) has been identified as the main cause for outflow of acid water and radioactive/non-radioactive contaminants. AMD encompasses pyrites oxidation when water and oxygen are available. AMD was identified in uranium waste rock piles (WRPs) of Industrias Nucleares do Brasil-Caldas facility (Brazilian uranium mine), resulting in high costs for water treatment. AMD reduction is the main challenge, and scientific investigation has been conducted to understand oxygen and water transportation within WRPs, where 222 Rn is used as natural tracer for oxygen transportation. The study consists of soil radon gas mapping in the top layer of WRP4 using active soil gas pumping, radon adsorption in active charcoal and 222 Rn determination using high-resolution gamma-ray spectrometry. A sampling network of 71 points was built where samples were collected at a depth of 40 cm. Soil radon gas concentration ranged from 33.7 to 1484.2 kBq m -3 with mean concentration of 320.7±263.3 kBq m -3 . (authors)

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

    Science.gov (United States)

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

    2016-04-01

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

  4. Uranium mineralization distribution sounding at ANO CRN.1-ANO CG.6 Mentawa sector using radon gas measurement

    International Nuclear Information System (INIS)

    Paimin; Sartapa; Darmono, S.

    2000-01-01

    The investigation was based on Systematic Prospection (1996,1997) at Mentawa River and Ampola up stream which were found radiometry anomalous about 750-15.000 c/s on the metasilt outcrop. Form of uranium mineralization is uraninite which associate with tourmaline, quartz, and sulphide and fills WNW-ESE fracture. The aims of investigation were to know uranium mineralization in sub surface by radon gas measurement, surface radiometry, and topographical mapping. (author)

  5. Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se; Gudowski, Waclaw [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)

    2005-11-15

    In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: {sup 235}U, which represents the 20% of the fresh uranium, {sup 233}U, which is produced by the transmutation of fertile {sup 232}Th, and {sup 239}Pu, which is produced by the transmutation of fertile {sup 238}U. In order to compensate the depletion of {sup 235}U with the breeding of {sup 233}U and {sup 239}Pu, the quantity of fertile nuclides must be much larger than that one of {sup 235}U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of {sup 235}U. At the same time, the amount of {sup 235}U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k {sub eff} and mass

  6. Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw

    2005-01-01

    In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: 235 U, which represents the 20% of the fresh uranium, 233 U, which is produced by the transmutation of fertile 232 Th, and 239 Pu, which is produced by the transmutation of fertile 238 U. In order to compensate the depletion of 235 U with the breeding of 233 U and 239 Pu, the quantity of fertile nuclides must be much larger than that one of 235 U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of 235 U. At the same time, the amount of 235 U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k eff and mass evolution, reaction rates, neutron flux and spectrum at the

  7. Determination of hydrogen in uranium-niobium-zirconium alloy by inert-gas fusion

    International Nuclear Information System (INIS)

    Carden, W.F.

    1979-12-01

    An improved method has been developed using inert-gas fusion for determining the hydrogen content in uranium-niobium-zirconium (U-7.5Nb-2.5Zr) alloy. The method is applicable to concentrations of hydrogen ranging from 1 to 250 micrograms per gram and may be adjusted for analysis of greater hydrogen concentrations. Hydrogen is determined using a hydrogen determinator. The limit of error for a single determination at the 95%-confidence level (at the 3.7-μg/g-hydrogen level) is +-1.4 micrograms per gram hydrogen

  8. Uranium

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  9. Screen-Printed Graphite Electrodes as Low-Cost Devices for Oxygen Gas Detection in Room-Temperature Ionic Liquids.

    Science.gov (United States)

    Lee, Junqiao; Hussain, Ghulam; Banks, Craig E; Silvester, Debbie S

    2017-11-26

    Screen-printed graphite electrodes (SPGEs) have been used for the first time as platforms to detect oxygen gas in room-temperature ionic liquids (RTILs). Up until now, carbon-based SPEs have shown inferior behaviour compared to platinum and gold SPEs for gas sensing with RTIL solvents. The electrochemical reduction of oxygen (O₂) in a range of RTILs has therefore been explored on home-made SPGEs, and is compared to the behaviour on commercially-available carbon SPEs (C-SPEs). Six common RTILs are initially employed for O₂ detection using cyclic voltammetry (CV), and two RTILs ([C₂mim][NTf₂] and [C₄mim][PF₆]) chosen for further detailed analytical studies. Long-term chronoamperometry (LTCA) was also performed to test the ability of the sensor surface for real-time gas monitoring. Both CV and LTCA gave linear calibration graphs-for CV in the 10-100% vol. range, and for LTCA in the 0.1-20% vol. range-on the SPGE. The responses on the SPGE were far superior to the commercial C-SPEs; more instability in the electrochemical responses were observed on the C-SPEs, together with some breaking-up or dissolution of the electrode surface materials. This study highlights that not all screen-printed ink formulations are compatible with RTIL solvents for longer-term electrochemical experiments, and that the choice of RTIL is also important. Overall, the low-cost SPGEs appear to be promising platforms for the detection of O₂, particularly in [C₄mim][PF₆].

  10. Screen-Printed Graphite Electrodes as Low-Cost Devices for Oxygen Gas Detection in Room-Temperature Ionic Liquids

    Directory of Open Access Journals (Sweden)

    Junqiao Lee

    2017-11-01

    Full Text Available Screen-printed graphite electrodes (SPGEs have been used for the first time as platforms to detect oxygen gas in room-temperature ionic liquids (RTILs. Up until now, carbon-based SPEs have shown inferior behaviour compared to platinum and gold SPEs for gas sensing with RTIL solvents. The electrochemical reduction of oxygen (O2 in a range of RTILs has therefore been explored on home-made SPGEs, and is compared to the behaviour on commercially-available carbon SPEs (C-SPEs. Six common RTILs are initially employed for O2 detection using cyclic voltammetry (CV, and two RTILs ([C2mim][NTf2] and [C4mim][PF6] chosen for further detailed analytical studies. Long-term chronoamperometry (LTCA was also performed to test the ability of the sensor surface for real-time gas monitoring. Both CV and LTCA gave linear calibration graphs—for CV in the 10–100% vol. range, and for LTCA in the 0.1–20% vol. range—on the SPGE. The responses on the SPGE were far superior to the commercial C-SPEs; more instability in the electrochemical responses were observed on the C-SPEs, together with some breaking-up or dissolution of the electrode surface materials. This study highlights that not all screen-printed ink formulations are compatible with RTIL solvents for longer-term electrochemical experiments, and that the choice of RTIL is also important. Overall, the low-cost SPGEs appear to be promising platforms for the detection of O2, particularly in [C4mim][PF6].

  11. Forecasting sandstone uranium deposits in oil-and-gas bearing basins

    International Nuclear Information System (INIS)

    Pechenkin, I.

    2014-01-01

    The interrelation between oxidation and reduction processes in the carbonaceous strata of Paleogene age was first studied in the 1950s in deposit of the Fergana depression. The presence of pre-ore and post-ore epigenesis of petroleum series was established. Part of uranium mineralization was found to be covered with fluid oil. In the middle of the 1960s in the Sabirsay deposit (Uzbekistan) in primary red-coloured continental sediment of Cretaceous age were studied pre-ore reduction changes, which caused economic uranium mineralization in contrasting geochemical barrier. Further research showed that multidirectional epigenetic processes had changed repeatedly. Later, in the 1970s, American geologists studying uranium deposits in the oil-and-gas bearing Texas Plain reached similar conclusions. From their point of view, in the Benevides deposit the main zones of mineralization tend to be located near the boundary where the zones of oxidation in the strata wedge in, developing in epigenetically reduced formations. A second post-mineral reduction was registered in a number of rock bodies. The complexity of the processes is determined by the double role of hydrocarbon fluids and the products of their dissolution. On the one hand, bituminization of permeable strata as well as pyritization, chloritization, dolomitization and other alterations associated with it create favourable geochemical conditions of a reducing character for a subsequent concentration of ore and nonmetal raw materials. On the other hand, intrusion of bitumen and its dissolution in the aeration zone leads to the burial of the mineralization which formed earlier and disappearance of all traces of its formation (epigenetic oxidation zoning). Thus forecasting and subsequent prospecting become impeded. The established sequence of epigenetic alterations allows us to carry out specialized mapping in productive regions, uncovering hidden parts of epigenetic oxidation zoning and “buried” mineralization

  12. Reprocessing of gas-cooled reactor particulate graphite fuel in a multi-strata transmutation system

    International Nuclear Information System (INIS)

    Laidler, J.J.

    2001-01-01

    Spent nuclear fuel discharged for light water reactors (LWRs) contains significant quantities of plutonium and other transuranic elements. Recent practice in Europe and Japan has been to recover the plutonium from spent fuel and recycle it to LWRs in the form of mixed uranium-plutonium oxide (MOX) fuel. Irradiation of the recycle fuel results in the generation of further plutonium and an increase in the isotopic concentration of the higher isotopes of plutonium, those having much lover fission cross sections than 239 Pu. This restricts plutonium recycle to one or two cycles, after which use of the plutonium becomes economically unfavorable. Recycle of the highly-transmuted plutonium in fast spectrum reactors can be an efficient method of fissioning this plutonium as well as other minor transuranics such as neptunium, americium and perhaps even curium. Those minor transuranics that are not conveniently burned in a fast reactor can be sent to an accelerator driven subcritical transmutation device for ultimate destruction. The preceding describes what has become known as a 'dual strata' or 'multi-strata' system. It is driven by the incentives to realize the maximum amount of energy from nuclear fuel and to eliminate the discharge of radio-toxic transuranic elements to the environment. Its implementation will be dependent in the long run upon the economic viability of the system and on the value placed by society on the elimination of radio-toxic materials that can conceivably be used in the manufacture of weapons of mass destruction. (author)

  13. A pore structure model for the gas transport property changes, initial oxidation rates and cumulative weight loss of AGR moderator graphite

    International Nuclear Information System (INIS)

    Johnson, P.A.V.

    1985-09-01

    A quantitative model has been developed for the gas transport property variation, cumulative weight loss and initial oxidation rates of AGR moderator graphite. The model utilises the theory of dynamic moments of the pore structure to calculate the changes in physical properties brought about by radiolytic corrosion taking place within the graphite porosity. In order to account for the behaviour of the initial rate curves, and the weight loss data obtained it is necessary to invoke the presence of a group of cylindrical pore and a group of small slab-shaped pores. The latter are methane depleted. This is in addition to the pore group involved in gas transport which is best represented by cylinders of mean radius 2.13 μm. The model satisfactorily predicts the experimental weight loss data obtained from experiments in the DIDO 6V3 and BFB loops. (author)

  14. Antibody-integrated and functionalized graphite-encapsulated magnetic beads, produced using ammonia gas plasma technology, for capturing Salmonella.

    Science.gov (United States)

    Sakudo, Akikazu; Chou, Han; Nagatsu, Masaaki

    2015-03-01

    Salmonella spp. is the single and most important causative agent of foodborne infections, especially involving foods such as eggs, milk and meat. To prevent infection, a reliable surveillance system is required that can quickly and sensitively detect Salmonella. Here, we describe the development of antibody-integrated magnetic beads that are functionalized by a novel strategy using ammonia gas plasma. Ammonia plasma, produced by a radio frequency (RF) power supply, was allowed to react with the surface of graphite-encapsulated magnetic beads, resulting in the introduction of amino groups. An anti-Salmonella antibody was then anchored by sulfide groups present on the protein surface to the amino groups of the magnetic beads via N-succinimidyl 3-(2-pyridyldithio) propionate (SPDP). The potential usefulness of these magnetic beads for capturing Salmonella was examined as follows. The beads were incubated with Salmonella in liquid medium and then separated from the supernatant by applying a magnetic field. After thorough washing, adsorption of Salmonella to the beads was confirmed by immunochromatography, polymerase chain reaction and a direct culture assay. Our findings indicate that the capture and concentration of Salmonella using the antibody-integrated magnetic beads was more efficient than commercial Dynabeads® anti-Salmonella, which are conventionally used for concentrating Salmonella from liquid cultures. We believe this novel bead technology will contribute to the enhanced detection of Salmonella. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. Use of Gas Transported Reactants for Uranium Remediation in Vadose Zone Sediments

    International Nuclear Information System (INIS)

    Szecsody, James E.; Zhong, Lirong; Truex, Michael J.; Resch, Charles T.; Williams, Mark D.

    2010-01-01

    This laboratory-scale investigation is focused on decreasing mobility of uranium in subsurface contaminated sediments in the vadose zone by in situ geochemical manipulation at low water content. This geochemical manipulation of the sediment surface phases included reduction, pH change (acidic and alkaline), and additions of chemicals (phosphate, ferric iron) to form specific precipitates. Reactants were advected into 1-D columns packed with Hanford 200 area U-contaminated sediment as a reactive gas (for CO2, NH3, H2S, SO2), with a 0.1% water content mist (for NaOH, Fe(III), HCl, PO4) and with a 1% water content foam (for PO4). Because uranium is present in the sediment in multiple phases, changes in U surface phases were evaluated with a series of liquid extractions that dissolve progressively less soluble phases and electron microbe identification of mineral phases. In terms of the short-term decrease in U mobility (in decreasing order), NH3, NaOH mist, CO2, HCl mist, and Fe(III) mist showed 20% to 35% change in U surface phases. The two reductive gas treatments (H2S and SO2) showed little change. For long-term decrease in U transport, mineral phases created that had low solubility (phosphates, silicates) were desired, so NH3, phosphates (mist and foam delivered), and NaOH mist showed the greatest formation of these minerals.

  16. Characterization of Ignalina NPP RBMK Reactors Graphite

    International Nuclear Information System (INIS)

    Hacker, P.J.; Neighbour, G.B.; Levinskas, R.; Milcius, D.

    2001-01-01

    The paper concentrates on the investigations of the initial physical properties of graphite used in production of graphite bricks of Ignalina NPP. These graphite bricks are used as nuclear moderator and major core structural components. Graphite bulk density is calculated by mensuration, pore volumes are measured by investigation of helium gas penetration in graphite pore network, the Young's modulus is determined using an ultrasonic time of flight method, the coefficient of thermal expansion is determined using a Netzsch dilatometer 402C, the fractured and machined graphite surfaces are studied using SEM, impurities are investigated qualitatively by EDAX, the degree of graphitization of the material is tested using X-ray diffraction. (author)

  17. Erosion of common structural materials and the degradation of suspended particles in flowing suspension of graphite powder in carbon dioxide gas

    International Nuclear Information System (INIS)

    Garton, D.A.; Hawes, R.I.; Rose, P.W.

    1968-06-01

    Experiments have been performed to examine the erosion of common materials of construction by a flowing suspension of graphite powder in carbon dioxide gas and the degradation of the graphite powder in the suspension. The suspension was circulated through a stainless steel loop at a pressure of 200 p.s.i.g. and bulk fluid temperature of 100-150 deg. C. No change in the weight of pins of mild steel, stainless steel and zircaloy, which were placed across the flow stream in a region where the velocity approached 100 ft./sec, could be detected after 350 hours of circulation. Examination of micro-photographs of the cross sections of the specimens showed no change in the structure of the metals. Considerable erosion of graphite pins producing a 6% decrease in the weight was observed under similar conditions. Detailed spectrographic analysis of the suspended powder taken at various times during the experiment showed no noticeable increase in the impurity content which could be attributed to erosion of the test specimens. A considerable increase in the tungsten, tin and cobalt concentration was observed and this is attributed to wear of the pump seal surfaces. The mean particle size of the suspended graphite powder was observed to decrease rapidly from 5 microns to 3 microns after only a few hours of circulation in the loop. After this initial period there was little further change in the particle size, the mean diameter being 2.85 microns after 167 hours of circulation. (author)

  18. Uranium enrichment using gas centrifugation. An analysis focusing export control; Urananrikning med gascentrifugering. En analys med fokus paa exportkontroll

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina [Swedish Defense Research Inst., Stockholm (Sweden)

    2005-08-01

    The Swedish Defence Research Agency, FOI, has performed a study on uranium enrichment by gas centrifugation. The theory and principles of gas centrifugation is described in this report and relevant equipment used in the process has been identified. Different aspects of operating a gas centrifuge facility - and its indicators - are also presented. The separation efficiency and the flow of material through a centrifuge are very small, and therefore, a large number of centrifuges in cascades is needed to produce a larger amount of enriched uranium within a reasonable time. Countries with nuclear weapons ambitions often show an interest in gas centrifuges to produce weapons grade uranium - if they have managed to acquire the technology - because of the efficiency of the process and since it is relatively easy to conceal. Most equipment used in gas centrifuge facilities is under export control to prevent clandestine uranium enrichment. The Nuclear Suppliers' Group has compiled lists of nuclear related equipment and components that are of importance to export control. The control lists have also been included in the EU legislation.

  19. Rapid hydrogen gas generation using reactive thermal decomposition of uranium hydride.

    Energy Technology Data Exchange (ETDEWEB)

    Kanouff, Michael P.; Van Blarigan, Peter; Robinson, David B.; Shugard, Andrew D.; Gharagozloo, Patricia E.; Buffleben, George M.; James, Scott Carlton; Mills, Bernice E.

    2011-09-01

    Oxygen gas injection has been studied as one method for rapidly generating hydrogen gas from a uranium hydride storage system. Small scale reactors, 2.9 g UH{sub 3}, were used to study the process experimentally. Complimentary numerical simulations were used to better characterize and understand the strongly coupled chemical and thermal transport processes controlling hydrogen gas liberation. The results indicate that UH{sub 3} and O{sub 2} are sufficiently reactive to enable a well designed system to release gram quantities of hydrogen in {approx} 2 seconds over a broad temperature range. The major system-design challenge appears to be heat management. In addition to the oxidation tests, H/D isotope exchange experiments were performed. The rate limiting step in the overall gas-to-particle exchange process was found to be hydrogen diffusion in the {approx}0.5 {mu}m hydride particles. The experiments generated a set of high quality experimental data; from which effective intra-particle diffusion coefficients can be inferred.

  20. Fabrication of SnO2-Reduced Graphite Oxide Monolayer-Ordered Porous Film Gas Sensor with Tunable Sensitivity through Ultra-Violet Light Irradiation

    Science.gov (United States)

    Xu, Shipu; Sun, Fengqiang; Yang, Shumin; Pan, Zizhao; Long, Jinfeng; Gu, Fenglong

    2015-01-01

    A new graphene-based composite structure, monolayer-ordered macroporous film composed of a layer of orderly arranged macropores, was reported. As an example, SnO2-reduced graphite oxide monolayer-ordered macroporous film was fabricated on a ceramic tube substrate under the irradiation of ultra-violet light (UV), by taking the latex microsphere two-dimensional colloid crystal as a template. Graphite oxide sheets dispersed in SnSO4 aqueous solution exhibited excellent affinity with template microspheres and were in situ incorporated into the pore walls during UV-induced growth of SnO2. The growing and the as-formed SnO2, just like other photocatalytic semiconductor, could be excited to produce electrons and holes under UV irradiation. Electrons reduced GO and holes adsorbed corresponding negative ions, which changed the properties of the composite film. This film was directly used as gas-sensor and was able to display high sensitivity in detecting ethanol gas. More interestingly, on the basis of SnO2-induced photochemical behaviours, this sensor demonstrated tunable sensitivity when UV irradiation time was controlled during the fabrication process and post in water, respectively. This study provides efficient ways of conducting the in situ fabrication of a semiconductor-reduced graphite oxide film device with uniform surface structure and controllable properties. PMID:25758292

  1. The application of the pore population balance method to the calculation of the radiolytic weight loss and gas transport property changes of nuclear graphites

    International Nuclear Information System (INIS)

    Johnson, P.A.V.

    1982-01-01

    A pore population balance equation, previously used to describe the physical property changes of porous carbons during thermal oxidation in carbon dioxide, has been modified to treat the radiolytic oxidation of graphite in CO 2 /CO/CH 4 gas mixtures. Good agreement has been obtained between theory and experiment for the variation in the gas transport coefficients B, K and lambda of gilsonite graphite with absorbed radiation dose. Calculations indicate that the addition of blind pores to the transport porosity, and an allowance for the opening of closed pores with burn-off, do not account for the experimental fractional weight loss curve. An excellent fit is obtained, however, if a small volume of cylindrical pores of a mean radius approximately equal to the diffusion length of oxidising species in the coolant are present in the pore size spectrum. Gilsonite graphite therefore behaves as if the pore size distribution function is trimodal, with mean radii at about 0.5μm, 2.48μm and greater than or equal to 10.57μm. (author)

  2. Numerical simulation of bellows effect on flow and separation of uranium isotopes in a supercritical gas centrifuge

    International Nuclear Information System (INIS)

    Borisevich, V.D.; Morozov, O.E.; Godisov, O.N.

    2000-01-01

    Numerical solving of the Navier-Stokes and convection-diffusion equations by the finite difference technique has been applied to study the influence of bellows on the flow and separation of uranium isotopes in a single supercritical gas centrifuge. Dependence of the separative power of a gas centrifuge on geometric parameters and position of a bellows on a rotor wall as well as the effect of scoop drag and feed flow on isotope separation in a gas centrifuge with a bellows have been obtained in computing experiments. It was demonstrated that increase of the separative power with increase of the gas centrifuge length is less considerable than predicted by the Dirac's law

  3. State policies and requirements for management of uranium mining and milling in New Mexico. Volume IV. The supply of electric power and natural gas fuel as possible constraints on uranium production

    International Nuclear Information System (INIS)

    Page, G.B.

    1980-04-01

    The report contained in this volume considers the availability of electric power to supply uranium mines and mills. The report, submited to Sandia Laboratories by the New Mexico Department of Energy and Minerals (EMD), is reproduced without modification. The state concludes that the supply of power, including natural gas-fueled production, will not constrain uranium production

  4. Electrochemical characteristics of silver- and nickel-coated synthetic graphite prepared by a gas suspension spray coating method for the anode of lithium secondary batteries

    International Nuclear Information System (INIS)

    Choi, Won Chang; Byun, Dongjin; Lee, Joong Kee; Cho, Byung won

    2004-01-01

    Four kinds of synthetic graphite coated with silver and nickel for the anodes of lithium secondary batteries were prepared by a gas suspension spray coating method. The electrode coated with silver showed higher charge-discharge capacities due to a Ag-Li alloy, but rate capability decreased at higher charge-discharge rate. This result can be explained by the formation of an artificial Ag oxidation film with higher impedance, this lowered the rate capability at high charge-discharge rate due to its low electrical conductivity. Rate capability is improved, however, by coating nickel and silver together on the surface of synthetic graphite. The nickel which is inactive with oxidation reaction plays an important role as a conducting agent which enhanced the conductivity of the electrode

  5. Energy map of southwestern Wyoming, Part B: oil and gas, oil shale, uranium, and solar

    Science.gov (United States)

    Biewick, Laura R.H.; Wilson, Anna B.

    2014-01-01

    The U.S. Geological Survey (USGS) has compiled Part B of the Energy Map of Southwestern Wyoming for the Wyoming Landscape Conservation Initiative (WLCI). Part B consists of oil and gas, oil shale, uranium, and solar energy resource information in support of the WLCI. The WLCI represents the USGS partnership with other Department of the Interior Bureaus, State and local agencies, industry, academia, and private landowners, all of whom collaborate to maintain healthy landscapes, sustain wildlife, and preserve recreational and grazing uses while developing energy resources in southwestern Wyoming. This product is the second and final part of the Energy Map of Southwestern Wyoming series (also see USGS Data Series 683, http://pubs.usgs.gov/ds/683/), and encompasses all of Carbon, Lincoln, Sublette, Sweetwater, and Uinta Counties, as well as areas in Fremont County that are in the Great Divide and Green River Basins.

  6. Multi-detector system approach for unattended uranium enrichment monitoring at gas centrifuge enrichment plants

    Science.gov (United States)

    Favalli, A.; Lombardi, M.; MacArthur, D. W.; McCluskey, C.; Moss, C. E.; Paffett, M. T.; Ianakiev, K. D.

    2018-01-01

    Improving the quality of safeguards measurements at Gas Centrifuge Enrichment Plants while reducing the inspection effort is an important objective given the number of existing and new plants that need to be safeguarded. A useful tool in many safeguards approaches is the on-line monitoring of enrichment in process pipes. One requirement of such a monitor is a simple, reliable and precise passive measurement of the 186-keV line from 235U. The other information required is the amount of gas in the pipe, which can be obtained by a transmission or pressure measurement. We describe our research to develop such a passive measurement system. Unfortunately, a complication arises in the interpretation of the gamma measurements, from the contribution of uranium deposits on the wall of the pipe to the 186-keV peak. A multi-detector approach to address this complication is presented where two measurements, one with signal primarily from gas and one with signal primarily from deposits, are performed simultaneously with different detectors and geometries. This allows a correction to be made to the 186-keV peak for the contribution from the deposit. We present the design of the multi-detector system and the results of the experimental calibration of the proof-of-principle prototype built at LANL.

  7. Multi-detector system approach for unattended uranium enrichment monitoring at gas centrifuge enrichment plants

    International Nuclear Information System (INIS)

    Favalli, A.; Lombardi, M.; MacArthur, D. W.; McCluskey, C.; Moss, C. E.

    2017-01-01

    Improving the quality of safeguards measurements at Gas Centrifuge Enrichment Plants while reducing the inspection effort is an important objective given the number of existing and new plants that need to be safeguarded. A useful tool in many safeguards approaches is the on-line monitoring of enrichment in process pipes. One requirement of such a monitor is a simple, reliable and precise passive measurement of the 186-keV line from 235 U. The other information required is the amount of gas in the pipe, which can be obtained by a transmission or pressure measurement. Here, we describe our research to develop such a passive measurement system. Unfortunately, a complication arises in the interpretation of the gamma measurements, from the contribution of uranium deposits on the wall of the pipe to the 186-keV peak. A multi-detector approach to address this complication is presented where two measurements, one with signal primarily from gas and one with signal primarily from deposits, are performed simultaneously with different detectors and geometries. This allows a correction to be made to the 186-keV peak for the contribution from the deposit. Finally, we present the design of the multi-detector system and the results of the experimental calibration of the proof-of-principle prototype built at LANL.

  8. Advanced cost-effective surface geochemical techniques for oil/gas/uranium exploration, environmental assessments and pipeline monitoring - a template for India

    International Nuclear Information System (INIS)

    Lafleur, Paul; Chanrasekharan, G.Y.V.N.; Rajender Rao, S.

    2011-01-01

    Advanced geochemical soil gas methods have been successfully developed for the exploration of oil/gas/uranium and for environmental assessments. Application of these cost-effective technologies in India can substantially reduce exploration risk while accelerating the development of oil/gas/uranium onshore resources. A reliable and effective monitoring system using geochemical soil gas surveys ensures that CO 2 Enhanced Oil Recovery operations as well as CO 2 sequestration projects are safe and acceptable for the disposal of CO 2 , Soil gas surveys along with other technologies can also be applied for monitoring of oil/gas pipelines for leakage, especially those that are old or pass through populated regions

  9. Ground water conditions and the relation to uranium deposits in the Gas Hills area, Fremont and Natrona Counties, Wyoming

    International Nuclear Information System (INIS)

    Marks, L.Y.

    1978-03-01

    As ground water apparently leaches, transports, and deposits uranium in the Gas Hills area, central Wyoming, it is important to understand its distribution, movement, and relation to geology and ore bodies. Water table maps were prepared of the Wind River Basin; the most detailed work was in the Gas Hills area. The water table in the Gas Hills area slopes downward to the northwest, ranges in depth from near the ground surface to more than 200 feet, and has seasonal fluctuation of about five feet. Perched water tables and artesian conditions occur locally. The oxidized-unoxidized rock contact is probably roughly parallel to the water table, and averages about 25 feet above it; although locally the two surfaces are considerably farther apart and the oxidized-unoxidized contact may be below the water table. In many places the gradient of the water table changes near the contact between rocks of different permeability. It is conformable with the structure at some anticlines and its gradient changes abruptly near some faults. Most above-normal concentrations of uranium occur at local water table depressions or at water table terraces where the gradient of the water table flattens. At these places, the uraniferous ground water is slowed and is in contact with the reducing agents in the rocks for a relatively long time. This may allow reduction of soluble transported uranium (U +6 ) to insoluble U +4 ) so that uranium is precipitated

  10. Radiation damage of uranium

    International Nuclear Information System (INIS)

    Lazarevic, Dj.

    1966-11-01

    Study of radiation damage covered the following: Kinetics of electric resistance of uranium and uranium alloy with 1% of molybdenum dependent on the second phase and burnup rate; Study of gas precipitation and diffusion of bubbles by transmission electron microscopy; Numerical analysis of the influence of defects distribution and concentration on the rare gas precipitation in uranium; study of thermal sedimentation of uranium alloy with molybdenum; diffusion of rare gas in metal by gas chromatography method

  11. Eastern gas shales bibliography selected annotations: gas, oil, uranium, etc. Citations in bituminous shales worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Hall, V.S. (comp.)

    1980-06-01

    This bibliography contains 2702 citations, most of which are annotated. They are arranged by author in numerical order with a geographical index following the listing. The work is international in scope and covers the early geological literature, continuing through 1979 with a few 1980 citations in Addendum II. Addendum I contains a listing of the reports, well logs and symposiums of the Unconventional Gas Recovery Program (UGR) through August 1979. There is an author-subject index for these publications following the listing. The second part of Addendum I is a listing of the UGR maps which also has a subject-author index following the map listing. Addendum II includes several important new titles on the Devonian shale as well as a few older citations which were not found until after the bibliography had been numbered and essentially completed. A geographic index for these citations follows this listing.

  12. Studies of Reduced Graphene Oxide and Graphite Oxide in the Aspect of Their Possible Application in Gas Sensors

    Science.gov (United States)

    Drewniak, Sabina; Muzyka, Roksana; Stolarczyk, Agnieszka; Pustelny, Tadeusz; Kotyczka-Morańska, Michalina; Setkiewicz, Maciej

    2016-01-01

    The paper presents the results of investigations on resistance structures based on graphite oxide (GRO) and graphene oxide (rGO). The subject matter of the investigations was thaw the sensitivity of the tested structures was affected by hydrogen, nitrogen dioxide and carbon dioxide. The experiments were performed at a temperature range from 30 °C to 150 °C in two carrier gases: nitrogen and synthetic air. The measurements were also aimed at characterization of the graphite oxide and graphene oxide. In our measurements we used (among others) techniques such as: Atomic Force Microscopy (AFM); Scanning Electron Microscopy (SEM); Raman Spectroscopy (RS); Fourier Transform Infrared Spectroscopy (FT-IR) and X-ray Photoelectron Microscopy (XPS). The data resulting from the characterizations of graphite oxide and graphene oxide have made it possible to interpret the obtained results from the point of view of physicochemical changes occurring in these structures. PMID:26784198

  13. Studies of Reduced Graphene Oxide and Graphite Oxide in the Aspect of Their Possible Application in Gas Sensors.

    Science.gov (United States)

    Drewniak, Sabina; Muzyka, Roksana; Stolarczyk, Agnieszka; Pustelny, Tadeusz; Kotyczka-Morańska, Michalina; Setkiewicz, Maciej

    2016-01-15

    The paper presents the results of investigations on resistance structures based on graphite oxide (GRO) and graphene oxide (rGO). The subject matter of the investigations was thaw the sensitivity of the tested structures was affected by hydrogen, nitrogen dioxide and carbon dioxide. The experiments were performed at a temperature range from 30 °C to 150 °C in two carrier gases: nitrogen and synthetic air. The measurements were also aimed at characterization of the graphite oxide and graphene oxide. In our measurements we used (among others) techniques such as: Atomic Force Microscopy (AFM); Scanning Electron Microscopy (SEM); Raman Spectroscopy (RS); Fourier Transform Infrared Spectroscopy (FT-IR) and X-ray Photoelectron Microscopy (XPS). The data resulting from the characterizations of graphite oxide and graphene oxide have made it possible to interpret the obtained results from the point of view of physicochemical changes occurring in these structures.

  14. RADIO FREQUENCY IDENTIFICATION DEVICES: EFFECTIVENESS IN IMPROVING SAFEGUARDS AT GAS-CENTRIFUGE URANIUM-ENRICHMENT PLANTS

    International Nuclear Information System (INIS)

    JOE, J.

    2007-01-01

    Recent advances in radio frequency identification devices (RFIDs) have engendered a growing interest among international safeguards experts. Potentially, RFIDs could reduce inspection work, viz. the number of inspections, number of samples, and duration of the visits, and thus improve the efficiency and effectiveness of international safeguards. This study systematically examined the applications of RFIDs for IAEA safeguards at large gas-centrifuge enrichment plants (GCEPs). These analyses are expected to help identify the requirements and desirable properties for RFIDs, to provide insights into which vulnerabilities matter most, and help formulate the required assurance tests. This work, specifically assesses the application of RFIDs for the ''Option 4'' safeguards approach, proposed by Bruce Moran, U. S. Nuclear Regulatory Commission (NRC), for large gas-centrifuge uranium-enrichment plants. The features of ''Option 4'' safeguards include placing RFIDs on all feed, product and tails (F/P/T) cylinders, along with WID readers in all FP/T stations and accountability scales. Other features of Moran's ''Option 4'' are Mailbox declarations, monitoring of load-cell-based weighing systems at the F/P/T stations and accountability scales, and continuous enrichment monitors. Relevant diversion paths were explored to evaluate how RFIDs improve the efficiency and effectiveness of safeguards. Additionally, the analysis addresses the use of RFIDs in conjunction with video monitoring and neutron detectors in a perimeter-monitoring approach to show that RFIDs can help to detect unidentified cylinders

  15. Swelling and gas release of grain-boundary pores in uranium dioxide

    International Nuclear Information System (INIS)

    Schrire, D.I.

    1983-12-01

    The swelling and gas release of overpressured grain boundary pores is sintered unirradiated uranium dioxide were investigated under isothermal conditions. The pores became overpressured when the ambient pressure was reduced, and the excess pressure driving force caused growth and interconnection of the pores, leading to eventual gas release. Swelling was measured continuously by a linear variable differential transformer, and open and closed porosity fractions were determined after the tests by immersion density and quantitative microscopy measurements. The sinter porosity consisted of pores situated on grain faces, grain edges, and grain corners. Isolated pores maintained their equilibrium shape while growing, without any measurable change in dihedral angle. Interconnection occurred predominantly along grain edges, without any evidence of pore sharpening or crack propagation at low driving forces. Extensive open porosity occurred at a threshold density of about 85% TD. There was an almost linear dependence of the initial swelling rate on the driving force, with an activation energy of 200+- 8 kJ/mole, in good agreement with published values of the activation energy for grain boundary diffusion

  16. Global resources and energy trade. An overview for coal, natural gas, oil and uranium

    Energy Technology Data Exchange (ETDEWEB)

    Remme, U.; Blesl, M.; Fahl, U.

    2007-07-15

    Despite efforts to improve energy effi-ciency and increase the usage of renewable energy carriers, fossil fuels and nuclear energy will continue to be important sources of global energy supply for the coming decades. Present global oil and gas supply is characterized by a concentration of production in a few world areas, mainly the Middle East and the Former Soviet Union, and a transport from these regions to the industrialized countries. Depletion of conventional reserves, especially oil, in combination with a surge for energy in emerging economies, as China and India, how-ever, is expected to change this picture in the future: unconventional resources in other world regions may be exploited to cover the surge energy demand, infrastructure for energy transport along new routes may have to be established. To provide a data base for such ques-tions, this report gives an overview of the current global resource situation for coal, natural gas, oil and uranium. In the first part, an assessment of the con-ventional and unconventional reserves and resources as well as their supply costs is given for the different regions of the world. The second part describes the current energy trade infrastructure between world regions and estimates the costs for existing and new trade links between these regions. (orig.)

  17. Graphite selection for the PBMR reflector

    International Nuclear Information System (INIS)

    Marsden, B.J.; Preston, S.D.

    2000-01-01

    A high temperature, direct cycle gas turbine, graphite moderated, helium cooled, pebble-bed reactor (PBMR) is being designed and constructed in South Africa. One of the major components in the PBMR is the graphite reflector, which must be designed to last thirty-five full power years. Fast neutron irradiation changes the dimensions and material properties of reactor graphite, thus for design purposes a suitable graphite database is required. Data on the effect of irradiation on nuclear graphites has been gathered for many years, at considerable financial cost, but unfortunately these graphites are no longer available due to rationalization of the graphite industry and loss of key graphite coke supplies. However, it is possible, using un-irradiated graphite materials properties and knowledge of the particular graphite microstructure, to determine the probable irradiation behaviour. Three types of nuclear graphites are currently being considered for the PBMR reflector: an isostatically moulded, fine grained, high strength graphite and two extruded medium grained graphites of moderately high strength. Although there is some irradiation data available for these graphites, the data does not cover the temperature and dose range required for the PBMR. The available graphites have been examined to determine their microstructure and some of the key material properties are presented. (authors)

  18. Carbon-14 Graphitization Chemistry

    Science.gov (United States)

    Miller, James; Collon, Philippe; Laverne, Jay

    2014-09-01

    Accelerator Mass Spectrometry (AMS) is a process that allows for the analysis of mass of certain materials. It is a powerful process because it results in the ability to separate rare isotopes with very low abundances from a large background, which was previously impossible. Another advantage of AMS is that it only requires very small amounts of material for measurements. An important application of this process is radiocarbon dating because the rare 14C isotopes can be separated from the stable 14N background that is 10 to 13 orders of magnitude larger, and only small amounts of the old and fragile organic samples are necessary for measurement. Our group focuses on this radiocarbon dating through AMS. When performing AMS, the sample needs to be loaded into a cathode at the back of an ion source in order to produce a beam from the material to be analyzed. For carbon samples, the material must first be converted into graphite in order to be loaded into the cathode. My role in the group is to convert the organic substances into graphite. In order to graphitize the samples, a sample is first combusted to form carbon dioxide gas and then purified and reduced into the graphite form. After a couple weeks of research and with the help of various Physics professors, I developed a plan and began to construct the setup necessary to perform the graphitization. Once the apparatus is fully completed, the carbon samples will be graphitized and loaded into the AMS machine for analysis.

  19. Characterisation of Chlorine Behavior in French Graphite

    International Nuclear Information System (INIS)

    Blondel, A.; Moncoffre, N.; Toulhoat, N.; Bererd, N.; Petit, L.; Laurent, G.; Lamouroux, C.

    2016-01-01

    Chlorine 36 is one of the main radionuclides of concern for French graphite waste disposal. In order to help the understanding of its leaching behaviour under disposal conditions, the respective impact of temperature, irradiation and gas radiolysis on chlorine release in reactor has been studied. Chlorine 36 has been simulated through chlorine 37 ion implantation in virgin nuclear graphite samples. Results show that part of chlorine is highly mobile in graphite in the range of French reactors operating temperatures in relation with graphite structural recovering. Ballistic damage generated by irradiation also promotes chlorine release whereas no clear impact of the coolant gas radiolysis was observed in the absence of graphite radiolytic corrosion. (author)

  20. The effect of crystal structure stability on the mobility of gas bubbles in intermetallic uranium compounds

    International Nuclear Information System (INIS)

    Rest, J.; Hofman, G.L.; Birtcher, R.C.

    1988-01-01

    Irradiation experiments with certain low-enrichment, high-density, uranium-base intermetallic alloys that are candidate reactor fuel materials, such as U 3 Si and U 6 Fe, have revealed extraordinarily large voids at low and medium fuel burnup. This phenomenon of breakaway swelling does not occur in other fuel types, such as U 3 Si 2 and UAl 3 , where a distribution of relatively small and stable fission gas bubbles forms. In situ transmission electron microscope observations of ion radiation-induced rapid swelling of intermetallic materials are consistent with growth by plastic flow. Large radiation enhancement of plastic flow in amorphous materials has been observed in several independent experiments and is thought to be a general materials phenomenon. The basis for a microscopic theory of fission gas bubble behavior in irradiated amorphous compounds has been formulated. The assumption underlying the overall theory is that the evolution of the porosity from that observed in the crystalline material to that observed in irradiated amorphous U 3 Si as a function of fluence is due to a softening of the irradiated amorphous material. Bubble growth in the low-viscosity material has been approximated by an effective enhanced diffusivity. Mechanisms are included for the radiation-induced softening of the amorphous material, and for a relation between gas atom mobilities and radiation-induced (defect-generated) changes in the material. Results of the analysis indicate that the observed rapid swelling in U 3 Si arises directly from enhanced bubble migration and coalescence due to plastic flow. 34 refs., 11 figs

  1. Chloride pyrometallurgy of uranium ore. 1. Chlorination of phosphate ore using solid or gas chlorinating agent and carbon

    International Nuclear Information System (INIS)

    Taki, Tomihiro; Komoto, Shigetoshi; Otomura, Keiichiro; Takenaka, Toshihide; Sato, Nobuaki; Fujino, Takeo.

    1995-01-01

    A thermodynamical and pyrometallurgical study to recover uranium from the phosphate ores was undertaken using the chloride volatilization method. Iron was chlorinated with solid chlorinating agents such as NaCl and CaCl 2 in combination with activated carbon, which will be used for removing this element from the ore, but uranium was not. On the other hand, the chlorination using Cl 2 gas and activated carbon gave a good result at 1,223 K. Not only uranium but also iron, phosphorus, aluminum and silicon were found to form volatile chlorides which vaporized out of the ore, while calcium remained in the ore as non-volatile CaCl 2 . The chlorination condition was studied as functions of temperature, reaction time and carbon content. The volatilization ratio of uranium around 95% was obtained by heating the mixture of the ore and activated carbon (35 wt%) in a mixed gas flow of Cl 2 (200 ml/min) and N 2 (200 ml/min) at 1,223 K for 120 min. (author)

  2. Synchrotron X-Ray Study of Melting in Submonolayer Ar and other Rare-Gas Films on Graphite

    DEFF Research Database (Denmark)

    McTague, J. P.; Als-Nielsen, Jens Aage; Bohr, Jakob

    1982-01-01

    Synchrotron x-ray diffraction studies of the (10) peak of Ar on the (001) surface of ZYX graphite show a sharp but continuous broadening of the Bragg peak with increasing temperature. Below a coverage of ∼ 1 Ar atom per six surface carbon atoms (ρ=1) the onset of this transition occurs...

  3. Gas-phase thermal dissociation of uranium hexafluoride: Investigation by the technique of laser-powered homogeneous pyrolysis

    International Nuclear Information System (INIS)

    Bostick, W.D.; McCulla, W.H.; Trowbridge, L.D.

    1987-04-01

    In the gas-phase, uranium hexafluoride decomposes thermally in a quasi-unimolecular reaction to yield uranium pentafluoride and atomic fluorine. We have investigated this reaction using the relatively new technique of laser-powered homogeneous pyrolysis, in which a megawatt infrared laser is used to generate short pulses of high gas temperatures under strictly homogeneous conditions. In our investigation, SiF 4 is used as the sensitizer to absorb energy from a pulsed CO 2 laser and to transfer this energy by collisions with the reactant gas. Ethyl chloride is used as an external standard ''thermometer'' gas to permit estimation of the unimolecular reaction rate constants by a relative rate approach. When UF 6 is the reactant, CF 3 Cl is used as reagent to trap atomic fluorine reaction product, forming CF 4 as a stable indicator which is easily detected by infrared spectroscopy. Using these techniques, we estimate the UF 6 unimolecular reaction rate constant near the high-pressure limit. In the Appendix, we describe a computer program, written for the IBM PC, which predicts unimolecular rate constants based on the Rice-Ramsperger-Kassel theory. Parameterization of the theoretical model is discussed, and recommendations are made for ''appropriate'' input parameters for use in predicting the gas-phase unimolecular reaction rate for UF 6 as a function of temperature and gas composition and total pressure. 85 refs., 17 figs., 14 tabs

  4. Management of UKAEA graphite liabilities

    International Nuclear Information System (INIS)

    Wise, M.

    2001-01-01

    The UK Atomic Energy Authority (UKAEA) is responsible for managing its liabilities for redundant research reactors and other active facilities concerned with the development of the UK nuclear technology programme since 1947. These liabilities include irradiated graphite from a variety of different sources including low irradiation temperature reactor graphite (the Windscale Piles 1 and 2, British Energy Pile O and Graphite Low Energy Experimental Pile at Harwell and the Material Testing Reactors at Harwell and Dounreay), advanced gas-cooled reactor graphite (from the Windscale Advanced Gas-cooled Reactor) and graphite from fast reactor systems (neutron shield graphite from the Dounreay Prototype Fast Reactor and Dounreay Fast Reactor). The decommissioning and dismantling of these facilities will give rise to over 6,000 tonnes of graphite requiring disposal. The first graphite will be retrieved from the dismantling of Windscale Pile 1 and the Windscale Advanced Gas-cooled Reactor during the next five years. UKAEA has undertaken extensive studies to consider the best practicable options for disposing of these graphite liabilities in a manner that is safe whilst minimising the associated costs and technical risks. These options include (but are not limited to), disposal as Low Level Waste, incineration, or encapsulation and disposal as Intermediate Level Waste. There are a number of technical issues associated with each of these proposed disposal options; these include Wigner energy, radionuclide inventory determination, encapsulation of graphite dust, galvanic coupling interactions enhancing the corrosion of mild steel and public acceptability. UKAEA is currently developing packaging concepts and designing packaging plants for processing these graphite wastes in consultation with other holders of graphite wastes throughout Europe. 'Letters of Comfort' have been sought from both the Low Level Waste and the Intermediate Level Waste disposal organisations to support the

  5. Quantitative analysis of hydrogen gas formed by aqueous corrosion of metallic uranium

    International Nuclear Information System (INIS)

    Fonnesbeck, J.

    2000-01-01

    Three unirradiated EBR-II blanket fuel samples containing depleted uranium metal were corrosion tested in simulated J-13 well water at 90 C. The corrosion rate of the blanket uranium metal was then determined relative to H 2 formation. Corrosion of one of the samples was interrupted prior to complete oxidation of the uranium metal and the solid corrosion product was analyzed for UO 2 and UH 3

  6. Quantitative analysis of hydrogen gas formed by aqueous corrosion of metallic uranium

    Energy Technology Data Exchange (ETDEWEB)

    Fonnesbeck, J.

    2000-03-20

    Three unirradiated EBR-II blanket fuel samples containing depleted uranium metal were corrosion tested in simulated J-13 well water at 90 C. The corrosion rate of the blanket uranium metal was then determined relative to H{sub 2} formation. Corrosion of one of the samples was interrupted prior to complete oxidation of the uranium metal and the solid corrosion product was analyzed for UO{sub 2} and UH{sub 3}.

  7. Off gas processing device for degreasing furnace for uranium/plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ueda, Masaya; Akasaka, Takayuki; Noura, Takeshi.

    1996-01-01

    A low melting ingredient capturing-cooling trap connected to a degreasing sintering furnace by way of sealed pipelines, a burning/decomposing device for decomposing high melting ingredient gases discharged from the cooling trap by burning them and a gas sucking means for forming the flow of off gases are contained in a glovebox, the inside pressure of which is kept negative. Since the degreasing sintering furnace for uranium/plutonium mixed oxide fuels is disposed outside of the glovebox, operation can be performed safely without greatly increasing the scale of the device, and the back flow of gases is prevented easily by keeping the pressure in the inside of the glovebox negative. Further, a heater is disposed at the midway of the sealed pipelines from the degreasing sintering furnace to the cooling trap, the temperature is kept high to prevent deposition of low melting ingredients to prevent clogging of the sealed pipelines. Further, a portion of the pipelines is made extensible in the axial direction to eliminate thermal stresses caused by temperature change thereby enabling to extend the life of the sealed pipelines. (N.H.)

  8. Artificial graphites

    International Nuclear Information System (INIS)

    Maire, J.

    1984-01-01

    Artificial graphites are obtained by agglomeration of carbon powders with an organic binder, then by carbonisation at 1000 0 C and graphitization at 2800 0 C. After description of the processes and products, we show how the properties of the various materials lead to the various uses. Using graphite enables us to solve some problems, but it is not sufficient to satisfy all the need of the application. New carbonaceous material open application range. Finally, if some products are becoming obsolete, other ones are being developed in new applications [fr

  9. A three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition in graphite components of advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, D.O.; Robinson, A.T.; Allen, D.A.; Picton, D.J.; Thornton, D.A. [TCS, Serco, Rutherford House, Olympus Park, Quedgeley, Gloucester, Gloucestershire GL2 4NF (United Kingdom); Shaw, S.E. [EDF Energy, Barnet Way, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2011-07-01

    This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition. (authors)

  10. Experiments on graphite block gaps connected with leak flow in bottom-core structure of experimental very high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Kikuchi, Kenji; Futakawa, Masatoshi; Takizuka, Takakazu; Kaburaki, Hideo; Sanokawa, Konomo

    1984-01-01

    In order to minimize the leak flow rate of an experimental VHTR (a multi-purpose very high-temperature gas-cooled reactor), the graphite blocks are tightened to reduce the gap distance between blocks by core restrainers surrounded outside of the fixed reflectors of the bottom-core structure and seal elements are placed in the gaps. By using a 1/2.75-scale model of the bottom-core structure, the experiments on the following items have been carried out: a relationship between core restraint force and block gap, a relationship between core restraint force and inclined angle of the model, leak flow characteristics of seal elements etc. The conclusions derived from the experiments are as follows: (1) Core restraint force is significantly effective for decreasing the gap distance between hot plenum blocks, but ineffective for the gap between hot plenum block and fixed reflector. (2) Graphite seal element reduces the leak flow rate from the top surface of hot plenum block into plenum region to one-third. (author)

  11. Monitoring the mass of UF6 gas and uranium deposits in aluminium pipes using X-ray fluorescence and X-ray transmission gauges

    International Nuclear Information System (INIS)

    Packer, T.W.; Smith, S.M.

    1984-12-01

    In order to determine the enrichment of UF 6 gas in centrifuge plant pipework it is necessary to measure the mass of the gas (pressure) and the mass per unit area of any uranium deposited on the pipe. This paper shows that it is possible to determine the pressure of the UF 6 gas in pipes 120 mm in diameter using an energy-dispersive X-ray fluorescence spectrometer. Results are also given of transmission measurements made using a low power X-ray generator operated at two different applied voltages. A method of using the two measurements to determine the mass per unit area of deposited uranium is described. (author)

  12. 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

  13. About the elaboration of pure uranium dicarbide

    International Nuclear Information System (INIS)

    Besson, J.; Blum, P.; Guinet, Ph.; Spitz, J.

    1963-01-01

    In order to develop methods for the elaboration of as pure as possible uranium dicarbide, the authors report the study of different elaboration processes based on the reaction between uranium and carbon, or between uranium and hydrocarbon, or between uranium oxide and carbon. They finally choose a method which comprises an arc-induced fusion of a mixture of uranium dioxide and carbon. The fusion process is described. The influence of thermal treatments is discussed as well as the graphite electrode carburization

  14. Results from uranium deposition studies for development of a Limited Frequency-Unannounced Access (LFUA) inspection strategy for gas centrifuge enrichment plants

    International Nuclear Information System (INIS)

    Cooley, J.N.; Fields, L.W.; Swindle, D.W.

    1985-06-01

    Uranium deposition studies were performed on a test loop system designed to simulate process gas flow through the header piping of a gas centrifuge enrichment plant. The objectives of these studies were to investigate the effectiveness of an in-line gaseous cleaning agent in removing uranium in pipe deposits and to analyze long-term deposition growth and isotopic exchange under simulated centrifuge plant operating conditions. The test loop studies are described, the results are reported, and the implications for analyzing actual plant data are discussed. Results indicate that: 93% of the uranium deposit is removed within 15 min when a pipe is pressurized with gaseous ClF 3 ; the isotopic abundance of a highly enriched uranium deposit remains unchanged when UF 6 of a lower assay is introduced into the pipe; and air inleakage will be the cause of the largest deposits in centrifuge plant process header pipes. 3 refs., 3 figs., 3 tabs

  15. Reduction of uranium hexafluoride to uranium tetrafluoride

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  16. The effect of time-dependent ventilation and radon (thoron) gas emanation rates in underground uranium mines

    International Nuclear Information System (INIS)

    Bigu, J.

    1987-01-01

    A theoretical radiation mine model, suitable for underground uranium mines, has been investigated. In this model, the rate of ventilation and/or the radon (thoron) gas emanation from mine walls are time-dependent. Several cases of practical interest have been investigated including sinusoidal, linear, exponential, stepwise, or a combination of two or more of the above. Analytical solutions were obtained for the time-dependent radon (thoron) gas emanation rate. However, because of the extreme analytical complexity of the solutions corresponding to the time-dependent ventilation rate case, numerical solutions were found using a special Runge-Kutta procedure and the Hamming's modified predictor-corrector method for the solution of linear initial-value problems. The mine model makes provisions for losses of radioactivity, other than by ventilation and radioactive decay, by, say, plate-out on mine walls, and by other mechanisms. Radioactivity data, i.e., radon, thoron, and their progeny, obtained with the above mine model for a number of ventilation and emanation conditions, are presented. Experimental data obtained in an inactive stope of an underground uranium mine for a time-dependent air flow case are shown. Air flow conditions (ventilation rate) were determined by tracer gas techniques using SF 6

  17. Quantification of trace level of fluoride content in uranium oxide produced by deconversion of HEX gas by ion chromatography

    International Nuclear Information System (INIS)

    Unnikrishnan, E.K.; Padmakumar, P.R.; Shanmugavelu, P.; Sudhakar, T.M.; Bhowmik, A.

    2015-01-01

    Fluoride content in nuclear fuel is detrimental due to its corrosion behavior with cladding material. It is essential to monitor and control the fluoride concentration in nuclear material at various processing stages. Deconversion of upgraded HEX gas is carried out to produce uranium oxide. The performance of the deconversion process of HEX gas is evaluated for which trace level of fluoride concentration accompanying uranium oxide is considered as a marker. An analytical method has been developed for testing the uranium oxide produced from deconversion process of HEX gas. The method involves sample pretreatment followed by analysis using ion chromatography. The test method was validated for its performance using in house synthetic uranyl fluoride (UO 2 F 2 ) standard solutions prepared with different level of fluoride content. The results are in agreement with the expected values with the recovery in the range of 80-95%. This method has been successfully implemented for routine analysis of samples at our lab. Since UO 2 F 2 reference material is not available to validate this method, in house UO 2 F 2 standards were prepared from U 3 O 8 prepared from nuclear grade uranyl nitrate solution. UO 2 F 2 standards were prepared by converting U 3 O 8 to UO 2 F 2 by the addition of HF followed by H 2 O 2 at 200°C on a hot plate. The entire yellow colored UO 2 F 2 was dissolved in nano pure water and recrystallised several times to ensure that all free HF is removed. The crystals dried in air oven at 120° for three hours. Samples containing 1000 mg kg -1 fluoride prepared from this UO 2 F 2 , and subsequently from this sample containing 5 mg kg -1 to 35 mg kg -1 fluoride samples were prepared and analysed against fluoride CRM and the fluoride concentration obtained was analysed

  18. Heated uranium tetrafluoride target system to release non-rare gas fission products for the TRISTAN isotope separator

    International Nuclear Information System (INIS)

    Gill, R.L.

    1977-10-01

    Off-line experiments indicated that fluorides of As, Se, Br, Kr, Zr, Nb, Mo, Tc, Ru, Sb, Te, I and Xe could be volatilized, but except for Br, Kr, I and Xe, none of these elements were observed after mass separation in the on-line experiments. The results of the on-line experiments indicated a very low level of hydride contamination at ambient temperature and consequently, uranium tetrafluoride replaced uranyl stearate as the primary gaseous fission product target. Possible reasons for the failure of the heated target system to yield non-rare gas activities are discussed and suggestions for designing a new heated target system are presented

  19. Measurement of the enrichment of uranium-hexafluoride gas in product pipes in the centrifuge enrichment plant at Almelo

    International Nuclear Information System (INIS)

    Packer, T.W.; Lees, E.W.; Aaldijk, J.K.; Harry, R.J.S.

    1987-09-01

    One of the objectives of safeguarding centrifuge enrichment plants is to apply non-destructive measurements inside the cascade area to confirm that the enrichment level is in the low enriched uranium range. Research in the UK and USA has developed a NDA instrument which can confirm the presence of low enriched uranium on a rapid go/no go basis in cascade header pipework of their centrifuge enrichment plants. The instrument is based on a gamma spectroscopic measurement coupled with an X-ray fluorescence analysis. This report gives the results of measurements carried out at Almelo by the UKAEA Harwell, ECN Petten and KFA Juelich to determine if these techniques could be employed at Almelo and Gronau. The energy dispersive X-ray fluorescence analysis has been applied to determine the total mass of uranium in the gas phase, and the deposit correction technique and the two geometry technique have been applied at Almelo to correct the measured gamma intensities for those emitted by the deposit. After an executive summary the report discusses the principles of the two correction methods. A short description of the equipment precedes the presentation of the results of the measurements and the discussion. After the conclusions the report contains two appendices which contain the derivation of the formulae for the deposit correction technique and a discussion of the systematic errors of this technique. 8 figs.; 11 refs.; 6 tables

  20. A graphite foam reinforced by graphite particles

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

    2007-11-15

    Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

  1. Effects of gas phase impurities on the topochemical-kinetic behaviour of uranium hydride development

    International Nuclear Information System (INIS)

    Bloch, J.; Brami, D.; Kremner, A.; Mintz, M.H.; Ben-Gurion Univ. of the Negev, Beersheba

    1988-01-01

    The hydriding kinetics of bulk uranium and U-0.1 wt.% Cr, in the presence of oxidizing gaseous impurities (oxygen and CO), were studied by combined rate measurements and metallographic examinations of partially reacted samples. The effect of the gaseous impurity (type and concentration) was examined metallographically, and the kinetic data were discussed in relation to these examinations. Below about 100 0 C the reaction of uranium with pure hydrogen consists of the following sequence of steps: (i) Surface nucleation; (ii) homogeneous growth (pitting); (iii) relatively fast lateral growth leading to the formation of a reaction front which penetrates into the sample at a constant rate. The effects of oxygen and CO on the hydriding kinetics were related to their abilities to block hydrogen penetration into the uranium. Thus, it was found that oxygen affects only the penetration through the oxide layer, whereas CO affects the penetration through both the oxide and hydride layers. (orig.)

  2. The challenge of uranium exploration

    International Nuclear Information System (INIS)

    Fountain, D.K.

    1982-06-01

    The first uranium discoveries at Beaverlodge were made using simple radiometric methods: hand-held geiger counters. Since then techniques of uranium exploration have evolved through airborne radiometric surveys, tracking glacial boulder trains to their origins, and electromagnetic surveys to detect graphite associated with buried uranium deposits. Simple radiometric surveys can cost around $1 000. per day, while testing for deposits at depths of over 400 meters will cost more than $60 000. per drill hole

  3. Graphite oxidation in HTGR atmosphere

    International Nuclear Information System (INIS)

    Growcock, F.B.; Barry, J.J.; Finfrock, C.C.; Rivera, E.; Heiser, J.H. III

    1982-01-01

    On-going and recently completed studies of the effect of thermal oxidation on the structural integrity of HTGR candidate graphites are described, and some results are presented and discussed. This work includes the study of graphite properties which may play decisive roles in the graphites' resistance to oxidation and fracture: pore size distribution, specific surface area and impurity distribution. Studies of strength loss mechanisms in addition to normal oxidation are described. Emphasis is placed on investigations of the gas permeability of HTGR graphites and the surface burnoff phenomenon observed during recent density profile measurements. The recently completed studies of catalytic pitting and the effects of prestress and stress on reactivity and ultimate strength are also discussed

  4. Graphite materials for nuclear reactors

    International Nuclear Information System (INIS)

    Oku, Tatsuo

    1991-01-01

    Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

  5. Extraction and analysis of reducing alteration information of oil-gas in Bashibulake uranium ore district based on ASTER remote sensing data

    International Nuclear Information System (INIS)

    Ye Fawang; Liu Dechang; Zhao Yingjun; Yang Xu

    2008-01-01

    Beginning with the analysis of the spectral characteristics of sandstone with reducing alteration of oil-gas in Bashibulake ore district, the extract technology of reducing alteration information based on ASTER data is presented. Several remote sensing anomaly zones of reducing alteration information similar with that in uranium deposit are interpreted in study area. On the basis of above study, these alteration anomaly information are further classified by using the advantage of ASTER data with multi-band in SWIR, the geological significance for alteration anomaly information is respectively discussed. As a result, alteration anomalies good for uranium prospecting are really selected, which provides some important information for uranium exploration in outland of Bashibulake uranium ore area. (authors)

  6. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    International Nuclear Information System (INIS)

    Monado, F.; Permana, S.

    2013-01-01

    Full-text: A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8 % HM. From the neutronic point of view, this design is in compliance with good performance. (author)

  7. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    International Nuclear Information System (INIS)

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-01-01

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance

  8. Proposal of method for quantifying uranium weight in the interior of gas centrifuge, and performance evaluation of IF7 treatment reaction product separation and refinement

    International Nuclear Information System (INIS)

    Ema, Akira; Kado, Kazumi; Suzuki, Kazuhiko

    2011-01-01

    In uranium enrichment plant decommissioning, it is important to treat and dispose the gas centrifuge reasonably. At the Ningyo-Toge Environmental Engineering Center of JAEA, it has been planned to carry out IF 7 treatment for the gas centrifuge in the uranium enrichment plant for decontamination before dismantling. In order to carry out this IF 7 treatment, a nondestructive measurement technology to estimate the uranium weight in the interior of the gas centrifuge was needed. However, it was difficult to estimate exactly uranium weight using former methods. Therefore, in this study, we investigated a new nondestructive measurement. As a result, we succeeded in constructing theoretically the methodology using γ-ray intensity measured outside the gas centrifuge and MCNP code. Furthermore, the UF 6 and IF 5 mixture was recovered by IF 7 treatment. A technology to separate and refine IF 5 from this mixture was needed in order to use as a material of IF 7 gas. Therefore, in this study, we carried out IF 5 separation and refinement examination. This paper shows the investigation result of reasonable separation condition and the full-scale examination result. As a result, it could be proved that 86% of IF 5 was separated and refined from UF 6 and IF 5 mixture. (author)

  9. Uranium absorption study pile

    International Nuclear Information System (INIS)

    Raievski, V.; Sautiez, B.

    1959-01-01

    The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF 3 counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10 -6 boron per gr. of uranium) or better. (author) [fr

  10. Thermally exfoliated graphite oxide

    Science.gov (United States)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor); Abdala, Ahmed (Inventor)

    2011-01-01

    A modified graphite oxide material contains a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, wherein the thermally exfoliated graphite oxide displays no signature of the original graphite and/or graphite oxide, as determined by X-ray diffraction.

  11. Physics experiments in graphite lattices (1962); Experiences sur les reseaux a graphite (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    A review is made of the various experimental methods used to determine the physics of graphite, natural uranium lattices: integral lattice experiments; both absolute and differential, effective cross section measurements, both by activation methods and by analysis of irradiated fuels, fine structure measurements. A number of experimental results are also given. (authors) [French] On decrit les differentes methodes experimentales utilisees pour determiner les parametres physiques de reseaux a uranium-graphite. Il s'agit d'experiences globales: mesures absolues et relatives de laplaciens, mesures de sections efficaces effectives par activation et par analyses de combustibles irradies, mesures de structures fines. Un certain nombre de resultats experimentaux sont communiques. (auteurs)

  12. Principle design and data of graphite components

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo

    2004-01-01

    The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned

  13. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: fabrication of high-temperature gas-cooled reactor fuel containing uranium-233 and thorium

    International Nuclear Information System (INIS)

    Roddy, J.W.; Blanco, R.E.; Hill, G.S.; Moore, R.E.; Seagren, R.D.; Witherspoon, J.P.

    1976-06-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables

  14. Application of gas geochemistry in the search for different types of uranium deposits. Final report for the period 15 November 1991 - 15 March 1994

    International Nuclear Information System (INIS)

    Chen Guoliang

    1994-10-01

    This paper presents the study of primary and secondary gas halo and application in the search for different types of uranium deposits in China. Through measuring Rn, CO 2 , O 2 and Hg in various deposits of geographical and climatic conditions for the gas geochemical survey. Three comprehensive measuring methods of CO 2 , Rn and 0 2 were thought to be optimum for gas geochemical exploration because of its portability, rapidity, low cost, simple operation and limited interference. Our experimental data can be briefly summarized that higher concentration of CO 2 , Rn and lower concentration of secondary halo in soil show blinded deposits. Our project have finished the study of thermal emanating gas in soil and rock. The methods were confirmed by known deposits and can be used in uranium exploration or environment investigation. 4 refs, 23 figs, 26 tabs

  15. Thermodynamic analysis and experimental study on the chlorination of uranium oxide by gas-solid reaction

    International Nuclear Information System (INIS)

    Shin, Y.J.; Kim, I.S.; Shin, H.S.; Ro, S.G.; Park, H.S.

    1998-01-01

    In order to determine the operating condition of an uranium chlorination process with U 3 O 8 -C-Cl 2 system, the experimental conditions have been evaluated preliminarily by the thermochemical analysis and experimentally confirmed in this study. The dry-type chlorination of U 3 O 8 occurs as irreversible and exothermic reaction and produces many kinds of chloride compounds such as UCl 3 , UCl 4 , UCl 5 , and UCl 6 in the air and humidity controlled argon environment. Taking account of Gibbs free energy and vapor pressure for various chloride compounds, the proper temperature range of chlorination appears to be 863 to 953 K in aspects of increasing reaction rate and the yield of nonvolatile product. In the course of the experimental confirmation the powder of U 3 O 8 is perfectly converted into uranium chlorides within 4 hours above 863 K, and then the maximum fraction of uranium chloride remaining in the reactor is about 30% of total conversion mass. (author)

  16. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    Marsden, B.J.

    2001-01-01

    The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

  17. Progress in radioactive graphite waste management

    International Nuclear Information System (INIS)

    2010-07-01

    Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

  18. Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release

    International Nuclear Information System (INIS)

    Zacharie, I.

    1997-01-01

    In order to keep pressurised water reactors at a top level of safety, it is necessary to understand the chemical and mechanical interaction between the cladding and the fuel pellet due to a temperature increase during a rapid change in reactor. In this process, the swelling of uranium oxide plays an important role. It comes from a bubble precipitation of fission gases which are released when they are in contact with the outside. Therefore, the aim of this thesis consists in acquiring a better understanding of the mechanisms which come into play. Uranium oxide samples, from a two cycles irradiated fuel, first have been thermal treated between 1000 deg C and 1700 deg C for 5 minutes to ten hours. The gas release amount related to time has been measured for each treatment. The comparison of the experimental results with a numerical model has proved satisfactory: it seems that the gases release, after the formation of intergranular tunnels, is controlled by the diffusion phenomena. Afterwards, the swelling was measured on the samples. The microscopic examination shows that the bubbles are located in the grain boundaries and have a lenticular shape. The swelling can be explained by the bubbles coalescence and a model was developed based on this observation. An equation allows to calculate the intergranular swelling in function of time and temperature. The study gives the opportunity to predict the fission gases behaviour during a fuel temperature increase. (author)

  19. Standard test method for isotopic analysis of uranium hexafluoride by double standard single-collector gas mass spectrometer method

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This is a quantitative test method applicable to determining the mass percent of uranium isotopes in uranium hexafluoride (UF6) samples with 235U concentrations between 0.1 and 5.0 mass %. 1.2 This test method may be applicable for the entire range of 235U concentrations for which adequate standards are available. 1.3 This test method is for analysis by a gas magnetic sector mass spectrometer with a single collector using interpolation to determine the isotopic concentration of an unknown sample between two characterized UF6 standards. 1.4 This test method is to replace the existing test method currently published in Test Methods C761 and is used in the nuclear fuel cycle for UF6 isotopic analyses. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appro...

  20. Bridged graphite oxide materials

    Science.gov (United States)

    Herrera-Alonso, Margarita (Inventor); McAllister, Michael J. (Inventor); Aksay, Ilhan A. (Inventor); Prud'homme, Robert K. (Inventor)

    2010-01-01

    Bridged graphite oxide material comprising graphite sheets bridged by at least one diamine bridging group. The bridged graphite oxide material may be incorporated in polymer composites or used in adsorption media.

  1. Uranium oxide recovering method

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  2. On the search for uranium

    International Nuclear Information System (INIS)

    Forland, A.

    1987-01-01

    The research reactor JEEP, which was completed in 1951 at Institutt for atomenergi (IFA), Kjeller, Norway, became the first reactor in the world to be built outside the big-power states. Due to Norwegian production of heavy water, the reactor was constructed as a heavy water reactor using natural uranium as fuel. A graphite reflector surrounded the reactor tank. Both uranium and graphite had to be purchased abroad. Because of the Anglo-American monopoly of all sizable uranium sources in the Western part of the world, no uranium for the reactor was available on the free market. The present study analyses Norway's and IFA's foreign relations at the time of the reactor project, and focuses in particular on the choice of the future partner that IFA had to make in order to solve its uranium problem. Political considerations were among the factors behind the decision in 1951 to establish a joint Dutch-Norwegian atomic energy research institute

  3. Are oil and gas stocks from the Australian market riskier than coal and uranium stocks? Dependence risk analysis and portfolio optimization

    International Nuclear Information System (INIS)

    Arreola Hernandez, Jose

    2014-01-01

    This article models the dependence risk and resource allocation characteristics of two 20-stock coal–uranium and oil–gas sector portfolios from the Australian market in the context of the global financial crisis of 2008–2009. The modeling framework implemented consists of pair vine copulas and, linear and nonlinear portfolio optimization methods with respect to five risk measures. The paper's objectives are to find out if the oil and gas stocks are riskier than the coal and uranium stocks, to identify the optimization method and risk measure that produce the best risk-return trade-off, to recognize the stocks in which the optimal weight allocations converge on average, and to acknowledge the vine copula model that best accounts for the overall dependence of the energy portfolios. The research findings indicate that the oil stocks have higher dependence risk than the coal, uranium and gas stocks in financial crisis periods. The higher risk of the oil stocks is confirmed by the larger concentration of symmetric and asymmetric dependence they have in the negative tail. The canonical vine (c-vine) copula model is observed to better capture the overall dependence of the energy portfolios. The combination of a pair c-vine copula and nonlinear portfolio optimization produces the highest return relative to risk. The optimal weight allocations converge on average in some stocks. - Highlights: • Vine copula dependence modeling of coal, uranium, oil and gas stocks • Oil stocks are riskier than coal, uranium and gas stocks in financial crisis periods. • The c-vine model better captures the overall dependence of the energy portfolios. • Vine copulas and nonlinear optimization combined produce the best results. • The optimal weight allocations converge on average in some stocks

  4. Study on graphite samples for nuclear usage

    International Nuclear Information System (INIS)

    Suarez, J.C.M.; Silva Roseira, M. da

    1994-01-01

    Available as short communication only. The graphite, due to its properties (mechanical strength, thermal conductivity, high-temperature stability, machinability etc.) have many industrial applications, and consequently, an important strategic value. In the nuclear area, it has been used as moderator and reflector of neutrons in the fission process of uranium. The graphite can be produced from many types of carbonaceous materials by a variety of process dominated by the manufactures. This is the reason why there are in the world market a lot of graphite types with different physical and mechanical properties. The present investigation studies some physical characteristics of the graphite samples destined to use in a nuclear reactor. (author). 8 refs, 1 fig, 1 tab

  5. Methane generated from graphite--tritium interaction

    International Nuclear Information System (INIS)

    Coffin, D.O.; Walthers, C.R.

    1979-01-01

    When hydrogen isotopes are separated by cryogenic distillation, as little as 1 ppM of methane will eventually plug the still as frost accumulates on the column packings. Elemental carbon exposed to tritium generates methane spontaneously, and yet some dry transfer pumps, otherwise compatible with tritium, convey the gas with graphite rotors. This study was to determine the methane production rate for graphite in tritium. A pump manufacturer supplied graphite samples that we exposed to tritium gas at 0.8 atm. After 137 days we measured a methane synthesis rate of 6 ng/h per cm 2 of graphite exposed. At this rate methane might grow to a concentration of 0.01 ppM when pure tritium is transferred once through a typical graphite--rotor transfer pump. Such a low methane level will not cause column blockage, even if the cryogenic still is operated continuously for many years

  6. Process for purifying graphite

    International Nuclear Information System (INIS)

    Clausius, R.A.

    1985-01-01

    A process for purifying graphite comprising: comminuting graphite containing mineral matter to liberate at least a portion of the graphite particles from the mineral matter; mixing the comminuted graphite particles containing mineral matter with water and hydrocarbon oil to form a fluid slurry; separating a water phase containing mineral matter and a hydrocarbon oil phase containing grahite particles; and separating the graphite particles from the hydrocarbon oil to obtain graphite particles reduced in mineral matter. Depending upon the purity of the graphite desired, steps of the process can be repeated one or more times to provide a progressively purer graphite

  7. High temperature soldering of graphite

    International Nuclear Information System (INIS)

    Anikin, L.T.; Kravetskij, G.A.; Dergunova, V.S.

    1977-01-01

    The effect is studied of the brazing temperature on the strength of the brazed joint of graphite materials. In one case, iron and nickel are used as solder, and in another, molybdenum. The contact heating of the iron and nickel with the graphite has been studied in the temperature range of 1400-2400 ged C, and molybdenum, 2200-2600 deg C. The quality of the joints has been judged by the tensile strength at temperatures of 2500-2800 deg C and by the microstructure. An investigation into the kinetics of carbon dissolution in molten iron has shown that the failure of the graphite in contact with the iron melt is due to the incorporation of iron atoms in the interbase planes. The strength of a joint formed with the participation of the vapour-gas phase is 2.5 times higher than that of a joint obtained by graphite recrystallization through the carbon-containing metal melt. The critical temperatures are determined of graphite brazing with nickel, iron, and molybdenum interlayers, which sharply increase the strength of the brazed joint as a result of the formation of a vapour-gas phase and deposition of fine-crystal carbon

  8. Steady-state fission gas behavior in uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Steele, W.G.; Wazzan, A.R.; Okrent, D.

    1989-01-01

    An analysis of fission gas release and induced swelling in steady state irradiated U-Pu-Zr metal fuels is developed and computer coded. The code is used to simulate, with fair success, some gas release and induced swelling data obtained under the IFR program. It is determined that fuel microstructural changes resulting from zirconium migration, anisotropic swelling, and thermal variations are major factors affecting swelling and gas release behavior. (orig.)

  9. Heat Transfer During Evaporation of Cesium From Graphite Surface in an Argon Environment

    Directory of Open Access Journals (Sweden)

    Bespala Evgeny

    2016-01-01

    Full Text Available The article focuses on discussion of problem of graphite radioactive waste formation and accumulation. It is shown that irradiated nuclear graphite being inalienable part of uranium-graphite reactor may contain fission and activation products. Much attention is given to the process of formation of radioactive cesium on the graphite element surface. It is described a process of plasma decontamination of irradiated graphite in inert argon atmosphere. Quasi-one mathematical model is offered, it describes heat transfer process in graphite-cesium-argon system. Article shows results of calculation of temperature field inside the unit cell. Authors determined the factors which influence on temperature change.

  10. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

    2015-06-15

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

  11. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

    Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

    1995-01-01

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

  12. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

    Hayden, H.W. Jr.; Horton, J.A.; Elliott, G.R.B.

    1995-06-06

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

  13. Effects of various tailings covers on radon gas emanation from pyritic uranium tailings

    International Nuclear Information System (INIS)

    Dave, N.K.; Lim, T.P.

    1987-01-01

    Radon emanation studies were carried out at an inactive pyritic uranium tailings site in Elliot Lake, Ontario, Canada, to evaluate the effects of various existing dry and wet covers on radon flux rates. Measurements were taken using activated charcoal cartridges for various surface covers consisting of bare, vegetated, acidophilic moss with high degree of water saturation, compacted crushed rock and gravel, and winter snow. The results showed that at a given site, there was no significant difference in radon emanation rates between various tailings covers and bare tailings. In particular, no increase In radon emanation rates from vegetated areas compared to bare tailings was observed. Radon emanation rates varied spatially depending on tailings grain size, porosity, moisture content and on pressure and water table variations. The emanation rates were higher for tailings with low water contents compared to those for wet and moss covered tailings

  14. Development of the chemical decontamination process of uranium enrichment gas centrifuges

    International Nuclear Information System (INIS)

    Mita, Yutaka; Endo, Yuji; Yamanaka, Toshihiro; Oohashi, Yusuke

    2002-01-01

    In Ningyo-Toge Environmental Engineering Center, many of the centrifuges that were tested for uranium enrichment are kept in storage. In the future, it will be necessary to dispose of them properly. By categorizing these centrifuges as 'items that are not required to be treated as radioactive waste', chemical decontamination tests were conducted with the wet process (diluted sulfuric acid) to reduce the amount of such radioactive waste. As a result, concerning the rotors, the assumed radioactive level was attained as items that are not required to be treated as radioactive waste', but the effectiveness of the casings varied. As a future subject, in order to find the optimal decontamination process, the basic test shall be conducted continuously. By taking economical efficiency and the processing time into consideration, the decontamination process will be evaluated and a rational method examined. (author)

  15. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Lai Shigang; Sun Libin; Zhang Zhengming

    2013-01-01

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  16. An improved model of fission gas atom transport in irradiated uranium dioxide

    Science.gov (United States)

    Shea, J. H.

    2018-04-01

    The hitherto standard approach to predicting fission gas release has been a pure diffusion gas atom transport model based upon Fick's law. An additional mechanism has subsequently been identified from experimental data at high burnup and has been summarised in an empirical model that is considered to embody a so-called fuel matrix 'saturation' phenomenon whereby the fuel matrix has become saturated with fission gas so that the continued addition of extra fission gas atoms results in their expulsion from the fuel matrix into the fuel rod plenum. The present paper proposes a different approach by constructing an enhanced fission gas transport law consisting of two components: 1) Fick's law and 2) a so-called drift term. The new transport law can be shown to be effectively identical in its predictions to the 'saturation' approach and is more readily physically justifiable. The method introduces a generalisation of the standard diffusion equation which is dubbed the Drift Diffusion Equation. According to the magnitude of a dimensionless Péclet number, P, the new equation can vary from pure diffusion to pure drift, which latter represents a collective motion of the fission gas atoms through the fuel matrix at a translational velocity. Comparison is made between the saturation and enhanced transport approaches. Because of its dependence on P, the Drift Diffusion Equation is shown to be more effective at managing the transition from one type of limiting transport phenomenon to the other. Thus it can adapt appropriately according to the reactor operation.

  17. French activities on gas cooled reactors

    International Nuclear Information System (INIS)

    Bastien, D.

    1996-01-01

    The gas cooled reactor programme in France originally consisted of eight Natural Uranium Graphite Gas Cooled Reactors (UNGG). These eight units, which are now permanently shutdown, represented a combined net electrical power of 2,375 MW and a total operational history of 163 years. Studies related to these reactors concern monitoring and dismantling of decommissioned facilities, including the development of methods for dismantling. France has been monitoring the development of HTRs throughout the world since 1979, when it halted its own HTR R and D programme. France actively participates in three CRPs set up by the IAEA. (author). 1 tab

  18. Nondestructive determination of uranium-235 enrichment in gas ultracentrifugation enrichment plants

    International Nuclear Information System (INIS)

    Lauppe, W.D.; Richter, B.; Stein, G.

    1990-02-01

    Based on similar studies in the USA and the UK, two γ spectroscopic techniques were propagated by the IAEA, after they had been sucesssfully tried at Capenhurst (GB) centrifuge plant. It was assumed by the IAEA that these methods would be transferable to the Almelo and Gronau plants, even without knowledge of the specific plant parameters there. The scope of this research project covered the study and further development of the proposed γ-spectroscopic measuring methods for applicability in the GUC plant in Gronau. The research came within the terms of reference of Task C.14.7 of the IAEA and BMFT (German Federal Ministry of Research and Technology)'s joint research and development programme. When the funded project first started, the Gronau plant was not yet on stream. The initial measurements were therefore, taken in Almelo, where conditions were similar to those in Gronau. The report describes the underlying problems which occur in relation to the non-destructive measuring of 235 U enrichment under the specific boundary conditions in Almelo and Gronau. The results illustrate the limitations of application of the measuring techniques, even after adjustment to actual conditions in Almelo and Gronau. It is not always possible to obtain a reliable yes/no answer in favour of slightly enriched uranium, particularly in product lines with extreme boundary conditions, irrespective of the technique applied. (orig.) [de

  19. Removal of 14C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction

    International Nuclear Information System (INIS)

    Dunzik-Gougar, Mary Lou; Windes, Will; Marsden, Barry

    2014-01-01

    The aim of the research presented here was to identify the chemical form of 14 C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14 C, with a half-life of 5730 years.

  20. Radiation damage of uranium; Radijaciono ostecenje urana

    Energy Technology Data Exchange (ETDEWEB)

    Lazarevic, Dj [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    Study of radiation damage covered the following: Kinetics of electric resistance of uranium and uranium alloy with 1% of molybdenum dependent on the second phase and burnup rate; Study of gas precipitation and diffusion of bubbles by transmission electron microscopy; Numerical analysis of the influence of defects distribution and concentration on the rare gas precipitation in uranium; study of thermal sedimentation of uranium alloy with molybdenum; diffusion of rare gas in metal by gas chromatography method.

  1. Radiolytic graphite oxidation revisited

    International Nuclear Information System (INIS)

    Minshall, P.C.; Sadler, I.A.; Wickham, A.J.

    1996-01-01

    The importance of radiolytic oxidation in graphite-moderated CO 2 -cooled reactors has long been recognised, especially in the Advanced Gas-Cooled Reactors where potential rates are higher because of the higher gas pressure and ratings than the earlier Magnox designs. In all such reactors, the rate of oxidation is partly inhibited by the CO produced in the reaction and, in the AGR, further reduced by the deliberate addition of CH 4 . Significant roles are also played by H 2 and H 2 O. This paper reviews briefly the mechanisms of these processes and the data on which they are based. However, operational experience has demonstrated that these basic principles are unsatisfactory in a number of respects. Gilsocarbon graphites produced by different manufacturers have demonstrated a significant difference in oxidation rate despite a similar specification and apparent equivalence in their pore size and distribution, considered to be the dominant influence on oxidation rate for a given coolant-gas composition. Separately, the inhibiting influence of CH 4 , which for many years had been considered to arise from the formation of a sacrificial deposit on the pore walls, cannot adequately be explained by the actual quantities of such deposits found in monitoring samples which frequently contain far less deposited carbon than do samples from Magnox reactors where the only source of such deposits is the CO. The paper also describes the current status of moderator weight-loss predictions for Magnox and AGR Moderators and the validation of the POGO and DIFFUSE6 codes respectively. 2 refs, 5 figs

  2. Neutronic Analysis and Radiological Safety of RSG-GAS Reactor on 300 Grams Uranium Silicide Core

    International Nuclear Information System (INIS)

    Pande Made Udiyani; Lily Suparlina; Rokhmadi

    2007-01-01

    As starting of usage silicide U 250 g fuel element in the core of RSG-GAS and will be continued with usage of silicide U 300 g fuel element, hence done beforehand neutronic analyse and radiological safety of RSG-GAS. Calculation done by ORIGEN2.1 code to calculate source term, and also by PC-COSYMA code to calculate radiological safety of radioactive dispersion from RSG-GAS. Calculation of radioactive dispersion done at condition of reactor is postulated be happened an accident of LOCA causing one fuel element to melt. Neutronic analysis indicate that silicide U 250 g full core shall to be operated beforehand during 625 MWD before converted to silicide U 300 g core. During operation of transition core with mixture of silicide U 250 g and 300 g, all parameter fulfill criterion of safety Designed Balance core of silicide U 300 g will be reached at the time of fifth full core. Result of calculation indicate that through mixture core of silicide U 250 and 300 g proposed can form silicide U 300 g balance core of reactor RSG-GAS safely. Calculation of radiology safety by deterministic for silicide U 300 g balance core, and accident postulation which is equal to core of silicide U 250 g yield output in the form of radiation activity (radionuclide concentration in the air and deposition on the ground), radiation dose (collective and individual), radiation effect (short- and long-range), which accepted by society in each perceived sector. Result of calculation indicated that dose accepted by society is not pass permitted boundary for public society if happened accident. (author)

  3. Inhibition of oxidation in nuclear graphite

    International Nuclear Information System (INIS)

    Winston, Philip L.; Sterbentz, James W.; Windes, William E.

    2015-01-01

    Graphite is a fundamental material of high-temperature gas-cooled nuclear reactors, providing both structure and neutron moderation. Its high thermal conductivity, chemical inertness, thermal heat capacity, and high thermal structural stability under normal and off-normal conditions contribute to the inherent safety of these reactor designs. One of the primary safety issues for a high-temperature graphite reactor core is the possibility of rapid oxidation of the carbon structure during an off-normal design basis event where an oxidising atmosphere (air ingress) can be introduced to the hot core. Although the current Generation IV high-temperature reactor designs attempt to mitigate any damage caused by a postulated air ingress event, the use of graphite components that inhibit oxidation is a logical step to increase the safety of these reactors. Recent experimental studies of graphite containing between 5.5 and 7 wt% boron carbide (B 4 C) indicate that oxidation is dramatically reduced even at prolonged exposures at temperatures up to 900 deg. C. The proposed addition of B 4 C to graphite components in the nuclear core would necessarily be enriched in B-11 isotope in order to minimise B-10 neutron absorption and graphite swelling. The enriched boron can be added to the graphite during billet fabrication. Experimental oxidation rate results and potential applications for borated graphite in nuclear reactor components will be discussed. (authors)

  4. Energy evaluations, graphite corrosion in Bugey I

    International Nuclear Information System (INIS)

    Brisbois, J.; Fiche, C.

    1967-01-01

    Bugey I presents a problem of radiolytic corrosion of the graphite by the CO 2 under pressure at high temperature. This report aims to evaluate the energy transferred to the gas by a Bugey I core cell, in normal operating conditions. The water, the carbon oxides and the hydrogen formed quantities are deduced as the consumed graphite and methane. Experimental studies are realized in parallel to validate the presented results. (A.L.B.)

  5. Hypervelocity impacts into graphite

    Science.gov (United States)

    Latunde-Dada, S.; Cheesman, C.; Day, D.; Harrison, W.; Price, S.

    2011-03-01

    Studies have been conducted into the characterisation of the behaviour of commercial graphite (brittle) when subjected to hypervelocity impacts by a range of projectiles. The experiments were conducted with a two-stage gas gun capable of launching projectiles of differing density and strength to speeds of about 6kms-1 at right angles into target plates. The damage caused is quantified by measurements of the crater depth and diameters. From the experimental data collected, scaling laws were derived which correlate the crater dimensions to the velocity and the density of the projectile. It was found that for moderate projectile densities the crater dimensions obey the '2/3 power law' which applies to ductile materials.

  6. Hypervelocity impacts into graphite

    International Nuclear Information System (INIS)

    Latunde-Dada, S; Cheesman, C; Day, D; Harrison, W; Price, S

    2011-01-01

    Studies have been conducted into the characterisation of the behaviour of commercial graphite (brittle) when subjected to hypervelocity impacts by a range of projectiles. The experiments were conducted with a two-stage gas gun capable of launching projectiles of differing density and strength to speeds of about 6kms -1 at right angles into target plates. The damage caused is quantified by measurements of the crater depth and diameters. From the experimental data collected, scaling laws were derived which correlate the crater dimensions to the velocity and the density of the projectile. It was found that for moderate projectile densities the crater dimensions obey the '2/3 power law' which applies to ductile materials.

  7. Uranium industry in the USSR

    International Nuclear Information System (INIS)

    Nikipelov, B.V.; Chernov, A.G.

    1990-01-01

    A brief historical account of the Soviet production of natural and enriched uranium is given. The geological and geographical location of major uranium deposits are mentioned. The processing of natural ores including in-situ leaching (ISL) is also briefly described. Gas centrifuges play a large part in uranium enrichment. The role of Techsnabexport for the export of nuclear materials is explained

  8. Trace metal distribution and mobility in drill cuttings and produced waters from Marcellus Shale gas extraction: Uranium, arsenic, barium

    International Nuclear Information System (INIS)

    Phan, Thai T.; Capo, Rosemary C.; Stewart, Brian W.; Graney, Joseph R.; Johnson, Jason D.; Sharma, Shikha; Toro, Jaime

    2015-01-01

    Highlights: • Distributions of U, As, and Ba in Marcellus Shale were determined. • As is primarily associated with sulfide minerals, Ba with exchange sites. • Most U is in the silicate minerals, but up to 20% is partitioned into carbonate. • Low [U] and [As] in produced water are consistent with reducing downhole conditions. • Proper waste management should account for potential mobilization of U and As. - Abstract: Development of unconventional shale gas wells can generate significant quantities of drilling waste, including trace metal-rich black shale from the lateral portion of the drillhole. We carried out sequential extractions on 15 samples of dry-drilled cuttings and core material from the gas-producing Middle Devonian Marcellus Shale and surrounding units to identify the host phases and evaluate the mobility of selected trace elements during cuttings disposal. Maximum whole rock concentrations of uranium (U), arsenic (As), and barium (Ba) were 47, 90, and 3333 mg kg −1 , respectively. Sequential chemical extractions suggest that although silicate minerals are the primary host for U, as much as 20% can be present in carbonate minerals. Up to 74% of the Ba in shale was extracted from exchangeable sites in the shale, while As is primarily associated with organic matter and sulfide minerals that could be mobilized by oxidation. For comparison, U and As concentrations were also measured in 43 produced water samples returned from Marcellus Shale gas wells. Low U concentrations in produced water (<0.084–3.26 μg L −1 ) are consistent with low-oxygen conditions in the wellbore, in which U would be in its reduced, immobile form. Arsenic was below detection in all produced water samples, which is also consistent with reducing conditions in the wellbore minimizing oxidation of As-bearing sulfide minerals. Geochemical modeling to determine mobility under surface storage and disposal conditions indicates that oxidation and/or dissolution of U

  9. Phonon scattering in graphite

    International Nuclear Information System (INIS)

    Wagner, P.

    1976-04-01

    Effects on graphite thermal conductivities due to controlled alterations of the graphite structure by impurity addition, porosity, and neutron irradiation are shown to be consistent with the phonon-scattering formulation 1/l = Σ/sub i equals 1/sup/n/ 1/l/sub i/. Observed temperature effects on these doped and irradiated graphites are also explained by this mechanism

  10. Method of manufacturing a graphite coated fuel can

    International Nuclear Information System (INIS)

    Saito, Koichi; Uchida, Shunsuke.

    1984-01-01

    Purpose: To improve the close bondability and homogeneity of a graphite coating formed at the inner surface of a fuel can. Method: A coating containing graphite dispersed in a volatile organic solvent is used and a graphite coating is formed to the inner surface of a fuel can by way of a plunger method. After applying graphite coating, an inert gas is caused to flow at a certain flow rate to the inside of the fuel can horizontally rotaged so that gassification and evaporation of the volatile organic solvent contained in the graphite coating may be promoted. Since drying of the graphite coating coated to the inner surface of the fuel can thus be controlled, a graphite coating with satisfactory close bondability and homogeneity can be formed. (Kawakami, Y.)

  11. Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-08

    Prior laboratory testing identified sodium nitrate and nitrite to be the most promising agents to minimize hydrogen generation from uranium metal aqueous corrosion in Hanford Site K Basin sludge. Of the two, nitrate was determined to be better because of higher chemical capacity, lower toxicity, more reliable efficacy, and fewer side reactions than nitrite. The present lab tests were run to determine if nitrate’s beneficial effects to lower H2 generation in simulated and genuine sludge continued for simulated sludge mixed with agents to immobilize water to help meet the Waste Isolation Pilot Plant (WIPP) waste acceptance drainable liquid criterion. Tests were run at ~60°C, 80°C, and 95°C using near spherical high-purity uranium metal beads and simulated sludge to emulate uranium-rich KW containerized sludge currently residing in engineered containers KW-210 and KW-220. Immobilization agents tested were Portland cement (PC), a commercial blend of PC with sepiolite clay (Aquaset II H), granulated sepiolite clay (Aquaset II G), and sepiolite clay powder (Aquaset II). In all cases except tests with Aquaset II G, the simulated sludge was mixed intimately with the immobilization agent before testing commenced. For the granulated Aquaset II G clay was added to the top of the settled sludge/solution mixture according to manufacturer application directions. The gas volumes and compositions, uranium metal corrosion mass losses, and nitrite, ammonia, and hydroxide concentrations in the interstitial solutions were measured. Uranium metal corrosion rates were compared with rates forecast from the known uranium metal anoxic water corrosion rate law. The ratios of the forecast to the observed rates were calculated to find the corrosion rate attenuation factors. Hydrogen quantities also were measured and compared with quantities expected based on non-attenuated H2 generation at the full forecast anoxic corrosion rate to arrive at H2 attenuation factors. The uranium metal

  12. Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

    International Nuclear Information System (INIS)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-01-01

    Prior laboratory testing identified sodium nitrate and nitrite to be the most promising agents to minimize hydrogen generation from uranium metal aqueous corrosion in Hanford Site K Basin sludge. Of the two, nitrate was determined to be better because of higher chemical capacity, lower toxicity, more reliable efficacy, and fewer side reactions than nitrite. The present lab tests were run to determine if nitrate's beneficial effects to lower H2 generation in simulated and genuine sludge continued for simulated sludge mixed with agents to immobilize water to help meet the Waste Isolation Pilot Plant (WIPP) waste acceptance drainable liquid criterion. Tests were run at ∼60 C, 80 C, and 95 C using near spherical high-purity uranium metal beads and simulated sludge to emulate uranium-rich KW containerized sludge currently residing in engineered containers KW-210 and KW-220. Immobilization agents tested were Portland cement (PC), a commercial blend of PC with sepiolite clay (Aquaset II H), granulated sepiolite clay (Aquaset II G), and sepiolite clay powder (Aquaset II). In all cases except tests with Aquaset II G, the simulated sludge was mixed intimately with the immobilization agent before testing commenced. For the granulated Aquaset II G clay was added to the top of the settled sludge/solution mixture according to manufacturer application directions. The gas volumes and compositions, uranium metal corrosion mass losses, and nitrite, ammonia, and hydroxide concentrations in the interstitial solutions were measured. Uranium metal corrosion rates were compared with rates forecast from the known uranium metal anoxic water corrosion rate law. The ratios of the forecast to the observed rates were calculated to find the corrosion rate attenuation factors. Hydrogen quantities also were measured and compared with quantities expected based on non-attenuated H2 generation at the full forecast anoxic corrosion rate to arrive at H2 attenuation factors. The uranium metal

  13. Scale-Up Information for Gas-Phase Ammonia Treatment of Uranium in the Vadose Zone at the Hanford Site Central Plateau

    Energy Technology Data Exchange (ETDEWEB)

    Truex, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Szecsody, James E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhong, Lirong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thomle, Jonathan N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Johnson, Timothy C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-01

    Uranium is present in the vadose zone at the Hanford Central Plateau and is of concern for protection of groundwater. The Deep Vadose Zone Treatability Test Plan for the Hanford Central Plateau identified gas-phase treatment and geochemical manipulation as potentially effective treatment approaches for uranium and technetium in the Hanford Central Plateau vadose zone. Based on laboratory evaluation, use of ammonia vapor was selected as the most promising uranium treatment candidate for further development and field testing. While laboratory tests have shown that ammonia treatment effectively reduces the mobility of uranium, additional information is needed to enable deployment of this technology for remediation. Of importance for field applications are aspects of the technology associated with effective distribution of ammonia to a targeted treatment zone, understanding the fate of injected ammonia and its impact on subsurface conditions, and identifying effective monitoring approaches. In addition, information is needed to select equipment and operational parameters for a field design. As part of development efforts for the ammonia technology for remediation of vadose zone uranium contamination, field scale-up issues were identified and have been addressed through a series of laboratory and modeling efforts. This report presents a conceptual description for field application of the ammonia treatment process, engineering calculations to support treatment design, ammonia transport information, field application monitoring approaches, and a discussion of processes affecting the fate of ammonia in the subsurface. The report compiles this information from previous publications and from recent research and development activities. The intent of this report is to provide technical information about these scale-up elements to support the design and operation of a field test for the ammonia treatment technology.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  15. Graphite moderated reactor for thermoelectric generation

    International Nuclear Information System (INIS)

    Akazawa, Issei; Yamada, Akira; Mizogami, Yorikata

    1998-01-01

    Fuel rods filled with cladded fuel particles distributed and filled are buried each at a predetermined distance in graphite blocks situated in a reactor core. Perforation channels for helium gas as coolants are formed to the periphery thereof passing through vertically. An alkali metal thermoelectric power generation module is disposed to the upper lid of a reactor container while being supported by a securing receptacle. Helium gas in the coolant channels in the graphite blocks in the reactor core absorbs nuclear reaction heat, to be heated to a high temperature, rises upwardly by the reduction of the specific gravity, and then flows into an upper space above the laminated graphite block layer. Then the gas collides against a ceiling and turns, and flows down in a circular gap around the circumference of the alkali metal thermoelectric generation module. In this case, it transfers heat to the alkali metal thermoelectric generation module. (I.N.)

  16. Acceptance test for graphite components and construction status of HTTR

    International Nuclear Information System (INIS)

    Iyoku, T.; Ishihara, M.; Maruyama, S.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs

  17. Starting up a programme of atomic piles using compressed gas; Le demarrage d'un programme de piles atomiques a gaz comprime

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    1) An examination of the intellectual and material resources which have directed the French programme towards: a) the natural uranium and plutonium system, b) the use of compressed gas as heat transfer fluid (primary fluid). 2) The parts played in exploring the field by the pile EL2 and G1, EL2 a natural uranium, heavy water and compressed gas pile, G1 a natural uranium, graphite and atmospheric air pile. 3) Development of the neutronics of graphite piles: physical study of G1. 4) The examination of certain problem posed by centres equipped with natural uranium, graphite and compressed carbon dioxide piles: structure, special materials, fluid circuits, maximum efficiency. Economic aspects. 5) Aids to progress: a) piles for testing materials and for tests on canned fuel elements, b) laboratory and calculation facilities. 6) Possible new orientations of compressed gas piles: a) raising of the pressure, b) enriched fuel, c) higher temperatures, d) use of heavy water. (author) [French] 1) Examen des ressources - intellectuelles et materielles - qui ont oriente le programme fran is vers: a) la voie de l'uranium naturel et du plutonium; b) l'emploi comme fluide pour le transfert de la chaleur (fluide primaire) d'un gaz comprime. 2) Le role d'exploration des piles EL2 et G1, EL2 pile a uranium naturel, eau lourde et gaz comprime, G1 pile a uranium naturel, graphite et air atmospherique. 3) Developpement de la neutronique des piles a graphite: l'etude physique de G1. 4) Examen de certains problemes poses par les centrales equipees de piles a uranium naturel, graphite et gaz carbonique comprime: structure, materiaux speciaux, circuits de fluides, optimisation. Aspects economiques. 5) Les auxiliaires du progres: a) piles pour essai de materiaux et pour essais de cartouches, b) moyens de laboratoire et moyens de calcul. 6) Orientations nouvelles possibles des piles a gaz comprime: a) elevation de la pression, b) combustible enrichi, c) temperatures elevees, d) emploi de l

  18. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1976-01-01

    The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)

  19. Study on the properties of the fuel compact for High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho; Lee, Young-woo; Cho, Moon-sung

    2015-01-01

    High Temperature Gas-cooled Reactors (HTGR), one of the Gen-IV reactors, have been using the fuel element which is manufactured by the graphite matrix, surrounding Tristructural-isotropic (TRISO)-coated Uranium particles. Factors with these characteristics effecting on the matrix of fuel compact are chosen and their impacts on the properties are studied. The fuel elements are considered with two types of concepts for HTGR, which are the block type reactor and the pebble bed reactor. In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength with the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and the two kinds of candidate binder (Phenol and Polyvinyl butyral) were chosen and mixed with each other, formed and heated to measure mechanical properties. The objective of this research is to optimize the materials and composition of the mixture and the forming process by evaluating the mechanical properties before/after carbonization and heat treatment. From the mechanical test results, the mechanical properties of graphite pellets was related to the various conditions such as the contents and kinds of binder, the kinds of graphite and the heat treatments. In the result of the compressive strength and Vicker's hardness, the 10 wt% phenol binder added R+S graphite pellet was relatively higher mechanical properties than other pellets. The contents of Phenol binder, the kinds of graphite powder and the temperature of carbonization and heat treatment are considered important factors for the properties. To optimize the mechanical properties of fuel elements, the role of binders and the properties of graphites will be investigated as

  20. Characterization of graphite dust produced by pneumatic lift

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Kang, Feiyu [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Yang, Xiaoyong; Li, Weihua [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 100084 (China)

    2016-08-15

    Highlights: • Generation of graphite dust by pneumatic lift. • Determination of morphology and particle size distribution of graphite dust. • The size of graphite dust in this study is compared to AVR and THTR-300 results. • Graphite dust originates from both filler and binder of the matrix graphite. - Abstract: Graphite dust is an important safety concern of high-temperature gas-cooled reactor (HTR). The graphite dust could adsorb fission products, and the radioactive dust is transported by the coolant gas and deposited on the surface of the primary loop. The simulation of coagulation, aggregation, deposition, and resuspension behavior of graphite dust requires parameters such as particle size distribution and particle shape, but currently very limited data on graphite dust is available. The only data we have are from AVR and THTR-300, however, the AVR result is likely to be prejudiced by the oil ingress. In pebble-bed HTR, graphite dust is generally produced by mechanical abrasion, in particular, by the abrasion of graphite pebbles in the lifting pipe of the fuel handling system. Here we demonstrate the generation and characterization of graphite dust that were produced by pneumatic lift. This graphite dust could substitute the real dust in HTR for characterization. The dust, exhibiting a lamellar morphology, showed a number-weighted average particle size of 2.38 μm and a volume-weighted average size of 14.62 μm. These two sizes were larger than the AVR and THTR results. The discrepancy is possibly due to the irradiation effect and prejudice caused by the oil ingress accident. It is also confirmed by the Raman spectrum that both the filler particle and binder contribute to the dust generation.

  1. Development of a radiochemical method for analyzing radon gas in uranium mine atmospheres: covering the period February 3, 1975--March 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Stein, L.; Shearer, J.A.; Hohorst, F.A.; Markun, F.

    1977-01-14

    A simplified radiochemical method has been developed for quantitatively analyzing radon gas in underground uranium mines. In this method, a measured volume of air is drawn by a pump through a drying tube and a cartridge containing dioxygenyl hexafluoroantimonate reagent. Radon is captured as a nonvolatile product. After radioactive equilibrium has been established between radon and its short-lived daughters (approximately 4 hours), the gamma-emission of the cartridge is measured with a scintillation counter. The amount of radon is then calculated from the gamma-emission rate. The effect of cartridge geometry, reagent load, and air flow rate upon collection efficiency and counting efficiency is reported.

  2. Development of a radiochemical method for analyzing radon gas in uranium mine atmospheres: covering the period February 3, 1975--March 31, 1976

    International Nuclear Information System (INIS)

    Stein, L.; Shearer, J.A.; Hohorst, F.A.; Markun, F.

    1977-01-01

    A simplified radiochemical method has been developed for quantitatively analyzing radon gas in underground uranium mines. In this method, a measured volume of air is drawn by a pump through a drying tube and a cartridge containing dioxygenyl hexafluoroantimonate reagent. Radon is captured as a nonvolatile product. After radioactive equilibrium has been established between radon and its short-lived daughters (approximately 4 hours), the gamma-emission of the cartridge is measured with a scintillation counter. The amount of radon is then calculated from the gamma-emission rate. The effect of cartridge geometry, reagent load, and air flow rate upon collection efficiency and counting efficiency is reported

  3. Technical development of graphite waste treatment in NUPEC

    International Nuclear Information System (INIS)

    Saishu, S.; Inoue, T.

    2001-01-01

    In Japan, Tokai Power Station, which is a Gas Cooled Reactor and uses graphite as moderator, ceased operation at the end of March in 1998 and it is planned to transfer to decommissioning stage. In this decommissioning stage it is very important to be able to treat and dispose the graphite waste in order to carry out the decommissioning safely and economically. NUPEC has been developing the graphite treatment and disposal technology since 1997 and we introduce the outline of the technical development. For the technology on high density packing into disposal container, the high density packing method and the assessment method on nuclide leaching characteristics were developed, and the cementing test for graphite powder by using Tokai spare graphite was performed and the hydrophobic characteristics between graphite and cement was grasped and the accelerator candidature for affinity was selected. From the view point of economical treatment, the incinerating technology was selected as candidature, and the methods for incinerating graphite and treating off gas are developed. The method of collecting C-14 in off gas was selected for reducing the off gas radiation level. The applicability of actual graphite treatment technology was considered from the view point of safety, economics and preparation of technical standard; the technical theme appeared, the developing planning items were established, and the detailed and actual scale tests will be carried out according to the planning. (author)

  4. Laboratory evaluation of the hydrogen sulfide gas treatment approach for remediation of chromate-, uranium(VI)-, and nitrate-contaminated soils

    International Nuclear Information System (INIS)

    Thornton, E.C.; Baechler, M.A.; Beck, M.A.; Amonette, J.E.

    1994-08-01

    Bench-scale soil treatment tests were conducted as part of an effort to develop and implement an in situ chemical treatment approach to the remediation of metal and radionuclide contaminated soils through the use of reactive gases. In general, > 90% immobilization of chromium and > 50% immobilization of uranium was achieved. Leach test results indicate that the treatment process is irreversible for chromium but partially reversible for uranium indicates that immobilization for this contaminant is more readily achieved in organic rich soils. This observation is ascribed to the reducing nature of organic matter. Additional tests were also conducted with soils contaminated to the 5,000 ppm level with nitrate. Nitrate was not found to interfere significantly with treatment of the contaminants. Nitrite was observed in the leachate samples obtained from tests with an organic-rich soil containing clay, however. Leachate chemistries suggested that no other significantly hazardous byproducts were generated by the treatment process and that soil alteration effects were minimal. Test results also suggest that treatment effectiveness is somewhat lower in very dry soils but still able to immobilize chromium and uranium to an acceptable degree. Results of these testing activities indicate that the concentration of hydrogen sulfide in the gas mixture is not a limited factor in treatment as long as a sufficient volume of the mixture is delivered to the soil to achieve a mole ratio of hydrogen sulfide to contaminant of at least 10

  5. Nuclear graphite wear properties and estimation of graphite dust production in HTR-10

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Xiaowei, E-mail: xwluo@tsinghua.edu.cn; Wang, Xiaoxin; Shi, Li; Yu, Xiaoyu; Yu, Suyuan

    2017-04-15

    Highlights: • Graphite dust. • The wear properties of graphite. • Pebble bed. • High Temperature Gas-cooled Reactor. • Fuel element. - Abstract: The issue of the graphite dust has been a research focus for the safety of High Temperature Gas-cooled Reactors (HTGRs), especially for the pebble bed reactors. Most of the graphite dust is produced from the wear of fuel elements during cycling of fuel elements. However, due to the complexity of the motion of the fuel elements in the pebble bed, there is no systematic method developed to predict the amount the graphite dust in a pebble bed reactor. In this paper, the study of the flow of the fuel elements in the pebble bed was carried out. Both theoretical calculation and numerical analysis by Discrete Element Method (DEM) software PFC3D were conducted to obtain the normal forces and sliding distances of the fuel elements in pebble bed. The wearing theory was then integrated with PFC3D to estimate the amount of the graphite dust in a pebble bed reactor, 10 MW High Temperature gas-cooled test Reactor (HTR-10).

  6. A graphite nanoeraser

    DEFF Research Database (Denmark)

    Liu, Ze; Bøggild, Peter; Yang, Jia-rui

    2011-01-01

    We present here a method for cleaning intermediate-size (up to 50 nm) contamination from highly oriented pyrolytic graphite and graphene. Electron-beam-induced deposition of carbonaceous material on graphene and graphite surfaces inside a scanning electron microscope, which is difficult to remove...... by conventional techniques, can be removed by direct mechanical wiping using a graphite nanoeraser, thus drastically reducing the amount of contamination. We discuss potential applications of this cleaning procedure....

  7. Uranium enrichment

    International Nuclear Information System (INIS)

    Mohrhauer, H.

    1982-01-01

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

  8. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Energy Technology Data Exchange (ETDEWEB)

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  9. Status of Chronic Oxidation Studies of Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Contescu, Cristian I. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mee, Robert W. [Univ. of Tennessee, Knoxville, TN (United States)

    2016-05-01

    Graphite will undergo extremely slow, but continuous oxidation by traces of moisture that will be present, albeit at very low levels, in the helium coolant of HTGR. This chronic oxidation may cause degradation of mechanical strength and thermal properties of graphite components if a porous oxidation layer penetrates deep enough in the bulk of graphite components during the lifetime of the reactor. The current research on graphite chronic oxidation is motivated by the acute need to understand the behavior of each graphite grade during prolonged exposure to high temperature chemical attack by moisture. The goal is to provide the elements needed to develop predictive models for long-time oxidation behavior of graphite components in the cooling helium of HTGR. The tasks derived from this goal are: (1) Oxidation rate measurements in order to determine and validate a comprehensive kinetic model suitable for prediction of intrinsic oxidation rates as a function of temperature and oxidant gas composition; (2) Characterization of effective diffusivity of water vapor in the graphite pore system in order to account for the in-pore transport of moisture; and (3) Development and validation of a predictive model for the penetration depth of the oxidized layer, in order to assess the risk of oxidation caused damage of particular graphite grades after prolonged exposure to the environment of helium coolant in HTGR. The most important and most time consuming of these tasks is the measurement of oxidation rates in accelerated oxidation tests (but still under kinetic control) and the development of a reliable kinetic model. This report summarizes the status of chronic oxidation studies on graphite, and then focuses on model development activities, progress of kinetic measurements, validation of results, and improvement of the kinetic models. Analysis of current and past results obtained with three grades of showed that the classical Langmuir-Hinshelwood model cannot reproduce all

  10. Oxidation Resistant Graphite Studies

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; R. Smith

    2014-07-01

    The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

  11. Progress in radioactive graphite waste management. Additional information

    International Nuclear Information System (INIS)

    2010-06-01

    Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

  12. Thermal conductivity of a graphite bipolar plate (BPP) and its thermal contact resistance with fuel cell gas diffusion layers: Effect of compression, PTFE, micro porous layer (MPL), BPP out-of-flatness and cyclic load

    Science.gov (United States)

    Sadeghifar, Hamidreza; Djilali, Ned; Bahrami, Majid

    2015-01-01

    This paper reports on measurements of thermal conductivity of a graphite bipolar plate (BPP) as a function of temperature and its thermal contact resistance (TCR) with treated and untreated gas diffusion layers (GDLs). The thermal conductivity of the BPP decreases with temperature and its thermal contact resistance with GDLs, which has been overlooked in the literature, is found to be dominant over a relatively wide range of compression. The effects of PTFE loading, micro porous layer (MPL), compression, and BPP out-of-flatness are also investigated experimentally. It is found that high PTFE loadings, MPL and even small BPP out-of-flatness increase the BPP-GDL thermal contact resistance dramatically. The paper also presents the effect of cyclic load on the total resistance of a GDL-BPP assembly, which sheds light on the behavior of these materials under operating conditions in polymer electrolyte membrane fuel cells.

  13. Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

    Science.gov (United States)

    Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

    2017-10-01

    Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

  14. Process for producing uranium oxide rich compositions from uranium hexafluoride

    International Nuclear Information System (INIS)

    DeHollander, W.R.; Fenimore, C.P.

    1978-01-01

    Conversion of gaseous uranium hexafluoride to a uranium dioxide rich composition in the presence of an active flame in a reactor defining a reaction zone is achieved by separately introducing a first gaseous reactant comprising a mixture of uranium hexafluoride and a reducing carrier gas, and a second gaseous reactant comprising an oxygen-containing gas. The reactants are separated by a shielding gas as they are introduced to the reaction zone. The shielding gas temporarily separates the gaseous reactants and temporarily prevents substantial mixing and reacting of the gaseous reactants. The flame occurring in the reaction zone is maintained away from contact with the inlet introducing the mixture to the reaction zone. After suitable treatment, the uranium dioxide rich composition is capable of being fabricated into bodies of desired configuration for loading into nuclear fuel rods. Alternatively, an oxygen-containing gas as a third gaseous reactant is introduced when the uranium hexafluoride conversion to the uranium dioxide rich composition is substantially complete. This results in oxidizing the uranium dioxide rich composition to a higher oxide of uranium with conversion of any residual reducing gas to its oxidized form

  15. Depleted uranium

    International Nuclear Information System (INIS)

    Huffer, E.; Nifenecker, H.

    2001-02-01

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

  16. Carbon nanoencapsulation of uranium dicarbide

    International Nuclear Information System (INIS)

    Pasqualini, E.

    1996-01-01

    Nanoparticles of uranium dicarbide encapsulated in carbon smaller than 100 nm have been obtained by chemical reactions at high temperature. Two types of nanocapsules were identified and characterized. The majority of them had small diffuse kernel surfaces, with dimensions between 5 and 15 nm, surrounded by thick spherical carbon cover. Others, in minor quantity and ranging from 15 to 40 nm, were polyhedrical and surrounded with several perfect graphite layers oriented parallel to their external surface. The nanocapsules are as chemically inert as graphite. (orig.)

  17. Carbon nanoencapsulation of uranium dicarbide

    Energy Technology Data Exchange (ETDEWEB)

    Pasqualini, E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. Combustibles Nucleares; Adelfang, P. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. Combustibles Nucleares; Regueiro, M.N. [EPM-Matformag, CNRS, Grenoble (France)

    1996-07-01

    Nanoparticles of uranium dicarbide encapsulated in carbon smaller than 100 nm have been obtained by chemical reactions at high temperature. Two types of nanocapsules were identified and characterized. The majority of them had small diffuse kernel surfaces, with dimensions between 5 and 15 nm, surrounded by thick spherical carbon cover. Others, in minor quantity and ranging from 15 to 40 nm, were polyhedrical and surrounded with several perfect graphite layers oriented parallel to their external surface. The nanocapsules are as chemically inert as graphite. (orig.).

  18. Method for producing dustless graphite spheres from waste graphite fines

    Science.gov (United States)

    Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

    2012-05-08

    A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

  19. Hydrogen adsorption on and solubility in graphites

    International Nuclear Information System (INIS)

    Kanashenko, S.L.; Wampler, W.R.

    1996-01-01

    The experimental data on adsorption and solubility of hydrogen isotopes in graphite over a wide range of temperatures and pressures are reviewed. Langmuir adsorption isotherms are proposed for the hydrogen-graphite interaction. The entropy and enthalpy of adsorption are estimated, allowing for effects of relaxation of dangling sp 2 bonds. Three kinds of traps are proposed: edge carbon atoms of interstitial loops with an adsorption enthalpy relative to H 2 gas of -4.4 eV/H 2 (unrelaxed, Trap 1), edge carbon atoms at grain surfaces with an adsorption enthalpy of -2.3 eV/H 2 (relaxed, Trap 2), and basal plane adsorption sites with an enthalpy of +2.43 eV/H 2 (Trap 3). The adsorption capacity of different types of graphite depends on the concentration of traps which depends on the crystalline microstructure of the material. The number of potential sites for the 'true solubility' (Trap 3) is assumed to be about one site per carbon atom in all types of graphite, but the endothermic character of this solubility leads to a negligible H inventory compared to the concentration of hydrogen in type 1 and type 2 traps for temperatures and gas pressures used in the experiments. Irradiation with neutrons or carbon atoms increases the concentration of type 1 and type 2 traps from about 20 and 200 appm respectively for unirradiated (POCO AXF-5Q) graphite to about 1500 and 5000 appm, respectively, at damage levels above 1 dpa. (orig.)

  20. Graphite targets at LAMPF

    International Nuclear Information System (INIS)

    Brown, R.D.; Grisham, D.L.

    1983-01-01

    Rotating polycrystalline and stationary pyrolytic graphite target designs for the LAMPF experimental area are described. Examples of finite element calculations of temperatures and stresses are presented. Some results of a metallographic investigation of irradiated pyrolytic graphite target plates are included, together with a brief description of high temperature bearings for the rotating targets

  1. Electrochemical treatment of graphite

    Energy Technology Data Exchange (ETDEWEB)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electrochemical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment, ECT of graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones with respect to the treatment rate and purity (ronghness) of the surface. A small quantity of sludge (6-8%) under ECT is in highly alkali electrolytes.

  2. Electrochemical treatment of graphite

    International Nuclear Information System (INIS)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electroche-- mical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment ECT graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones this is treatment rate and purity (ronghness) of the surface. A sMall quantity of sludge (6-8%) under ECT is in highly alkali electrolytes

  3. Graphite oxidation and structural strength of graphite support column in VHTR

    International Nuclear Information System (INIS)

    Park, Byung Ha; No, Hee Cheno; Kim, Eung Soo; Oh, Chang H.

    2009-01-01

    The air-ingress event by a large pipe break is an important accident considered in design of very high-temperature gas-cooled reactors (VHTR). Core-collapse prediction is a main safety issue. Structural failure model are technically required. The objective of this study is to develop structural failure model for the supporting graphite material in the lower plenum of the GT-MHR (gas-turbine-modular high temperature reactor). Graphite support column is important for VHTR structural integrity. Graphite support columns are under the axial load. Critical strength of graphite column is related to slenderness ratio and bulk density. Through compression tests for fresh and oxidized graphite columns we show that compressive strength of IG-110 was 79.46 MPa. And, the buckling strength of IG-110 column was expressed by the empirical formula: σ 0 =σ straight-line - C L/r, σ straight-line =91.31 MPa, C=1.01. The results of uniform and non-uniform oxidation tests show that the strength degradation of oxidized graphite column is expressed in the following non-dimensional form: σ/σ 0 =exp(-kd), k=0.111. Also, from the results of the uniform oxidation test with a complicated-shape column, we found out that the above non-dimensional equation obtained from the uniform oxidation test is applicable to a uniform oxidation case with a complicated-shape column. (author)

  4. Integration of antibody by surface functionalization of graphite-encapsulated magnetic beads using ammonia gas plasma technology for capturing influenza A virus.

    Science.gov (United States)

    Sakudo, Akikazu; Chou, Han; Ikuta, Kazuyoshi; Nagatsu, Masaaki

    2015-05-01

    Antibody-integrated magnetic beads have been functionalized for influenza A virus capture. First, ammonia plasma produced by a radio frequency power source was reacted with the surface of graphite-encapsulated magnetic beads to introduce amino groups. Anti-influenza A virus hemagglutinin antibody was then anchored by its surface sulfide groups to the amino groups on the beads via N-succinimidyl 3-(2-pyridyldithio) propionate. After incubation with influenza A virus, adsorption of the virus to the beads was confirmed by immunochromatography, polymerase chain reaction (PCR), enzyme-linked immunosorbent assay (ELISA), and inoculation of chicken embryonated eggs, indicating that virus infectivity is maintained and that the proposed method is useful for the enhanced detection and isolation of influenza A virus. Copyright © 2015 Elsevier Ltd. All rights reserved.

  5. Trace Analysis of Boron in Nuclear Graphite by Means of Gas Chromatography; Analyse de traces de bore dans le graphite nucleaire au moyen de la chromatographie gazeuse; Analiz metodom gazovoj khromatografii sledov bora v yadernom grafite; Analisis de vestigios de boro en el grafito nuclear por cromatografia de gases

    Energy Technology Data Exchange (ETDEWEB)

    Zado, F. [Rudjer Boskovlc Institute, Zagreb, Yugoslavia (Croatia)

    1963-11-15

    No literature is available about the application of gas chromatography in trace analysis of boron in graphite. The following methods of transformation of boron into its volatile compounds are discussed: (a) Ignition of graphite in a stream of oxygen and subsequent transformation of boron oxide into volatile methyl borate which is then analysed on a Dilkens Aerograph H Model 96 gas chromatograph with silicone column and hydrogen as carrier (concentration method). (b) Extraction of boron from the graphite by means of sodium fluoride at 2800{sup o}C with simultaneous chlorination and trapping of boron trichloride, which is then analysed (direct method). A home-made gas chromatograph with a thermal conductivity detector and nitrogen as a carrier was used. The column was made of glass with a 20% (wt./wt.) fluorocarbon oil on kieselguhr. Special precautions were taken on account of the sensitivity of boron trichloride to moisture. (author) [French] Rien n'a ete publie sur l'application de la chromatographie gazeuse a l'analyse de traces de bore dans le graphite. L'auteur examine les methodes suivantes de transformation du bore en composes volatils: a) Brulage du graphite dans un courant d'oxygene, puis transformation de l'oxyde de bore en borate de methyle volatil, que l'on analyse ensuite a l'aide d'un appareil a chromatographie gazeuse Dilkens Aerograph H, Modele 96, utilisant une colonne de silicium et de rhydrogene comme entrafheur (methode de concentration). b) Extraction du bore contenu dans le graphite par chauffage avec du fluorure de sodium a 2800{sup o}C, avec chloruration simultanee et captage du trichlorure de bore qui est ensuite analyse (methode directe). On s' est servi d'un appareil a chromatographie gazeuse construit sur place, dote d'un detecteur a conductibilite thermique et utilisant l'azote comme entrafheur. La colonne etait faite de verre contenant 20% en poids de fluorocarbure huileux sur du kieselguhr. Des precautions particulieres ont ete prises

  6. Determination of uranium in coated fuel particle compact by potassium fluoride fusion-gravimetric method

    International Nuclear Information System (INIS)

    Ito, Mitsuo; Iso, Shuichi; Hoshino, Akira; Suzuki, Shuichi.

    1992-03-01

    Potassium fluoride-gravimetric method has been developed for the determination of uranium in TRISO type-coated fuel particle compact. Graphite matrix in the fuel compact is burned off by heating it in a platinum crucible at 850degC. The coated fuel particles thus obtained are decomposed by fusion with potassium fluoride at 900degC. The melt was dissolved with sulfuric acid. Uranium is precipitated as ammonium diuranate, by passing ammonia gas through the solution. The resulting precipitate is heated in a muffle furnace at 850degC, to convert uranium into triuranium octoxide. Uranium in the triuranium octoxide was determined gravimetrically. Ten grams of caoted fuel particles were completely decomposed by fusion with 50 g of potassium fluoride at 900degC for 3 hrs. Analytical result for uranium in the fuel compact by the proposed method was 21.04 ± 0.05 g (n = 3), and was in good agreement with that obtained by non-destructive γ-ray measurement method : 21.01 ± 0.07 g (n = 3). (author)

  7. URANIUM SEPARATION PROCESS

    Science.gov (United States)

    Lyon, W.L.

    1962-04-17

    A method of separating uranium oxides from PuO/sub 2/, ThO/sub 2/, and other actinide oxides is described. The oxide mixture is suspended in a fused salt melt and a chlorinating agent such as chlorine gas or phosgene is sparged through the suspension. Uranium oxides are selectively chlorinated and dissolve in the melt, which may then be filtered to remove the unchlorinated oxides of the other actinides. (AEC)

  8. Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

    OpenAIRE

    Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

    2012-01-01

    Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

  9. Studies on yttrium oxide coatings for corrosion protection against molten uranium

    International Nuclear Information System (INIS)

    Chakravarthy, Y.; Bhandari, Subhankar; Pragatheeswaran; Thiyagarajan, T.K.; Ananthapadmanabhan, P.V.; Das, A.K.; Kumar, Jay; Kutty, T.R.G.

    2012-01-01

    Yttrium oxide is resistant to corrosion by molten uranium and its alloys. Yttrium oxide is recommended as a protective oxide layer on graphite and metal components used for melting and processing uranium and its alloys. This paper presents studies on the efficacy of plasma sprayed yttrium oxide coatings for barrier applications against molten uranium

  10. Thermal Properties of G-348 Graphite

    Energy Technology Data Exchange (ETDEWEB)

    McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Valentin, Francisco I. [City Univ. (CUNY), NY (United States)

    2017-04-01

    Fundamental measurements have been obtained in the INL Graphite Characterization Laboratory to deduce the temperature dependence of thermal conductivity for G-348 isotropic graphite, which has been used by City College of New York in thermal experiments related to gas-cooled nuclear reactors. Measurements of thermal diffusivity, mass, volume and thermal expansion were converted to thermal conductivity in accordance with ASTM Standard Practice C781-08 (R-2014). Data are tabulated and a preliminary correlation for the thermal conductivity is presented as a function of temperature from laboratory temperature to 1000C.

  11. Fission Product Sorptivity in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Tompson, Jr., Robert V. [Univ. of Missouri, Columbia, MO (United States); Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States); Ghosh, Tushar [Univ. of Missouri, Columbia, MO (United States); Viswanath, Dabir [Univ. of Missouri, Columbia, MO (United States); Walton, Kyle [Univ. of Missouri, Columbia, MO (United States); Haffner, Robert [Univ. of Missouri, Columbia, MO (United States)

    2015-04-01

    graduate student meant that data acquisition with the packed bed systems ended up competing for the graduate student’s available time with the electrodynamic balance redesign and assembly portions of the project. This competition for available time was eventually mitigated to some extent by the later recruitment of an undergraduate student to help with data collection using the packed bed system. It was only the recruitment of the second student that allowed the single particle balance design and construction efforts to proceed as far as they did during the project period. It should be added that some significant time was also spent by the graduate student cataloging previous work involving graphite. This eventually resulted in a review paper being submitted and accepted (“Adsorption of Iodine on Graphite in High Temperature Gas-Cooled Reactor Systems: A Review,” Kyle L. Walton, Tushar K. Ghosh, Dabir S. Viswanath, Sudarshan K. Loyalka, Robert V. Tompson). Our specific revised objectives in this project were as follows: Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using an EDB and a temperature controlled EDB; Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using a packed column bed apparatus; Explore the effect that charge has on the adsorption isotherms of iodine by varying the charges on and the voltages used to suspend the microscopic particles in the EDB; and To interpret these results in terms of the existing models (Langmuir, BET, Freundlich, and others) which we will modify as necessary to include charge related effects.

  12. Characterization of single crystal uranium-oxide thin films grown via reactive-gas magnetron sputtering on yttria-stabilized zirconia and sapphire

    Energy Technology Data Exchange (ETDEWEB)

    Strehle, Melissa M.; Heuser, Brent J., E-mail: bheuser@illinois.edu; Elbakhshwan, Mohamed S.; Han Xiaochun; Gennardo, David J.; Pappas, Harrison K.; Ju, Hyunsu

    2012-06-30

    The microstructure and valence states of three single crystal thin film systems, UO{sub 2} on (11{sup Macron }02) r-plane sapphire, UO{sub 2} on (001) yttria-stabilized zirconia, and U{sub 3}O{sub 8} on (11{sup Macron }02) r-plane sapphire, grown via reactive-gas magnetron sputtering are analyzed primarily with X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), and ultraviolet photoelectron spectroscopy (UPS). XRD analysis indicates the growth of single crystal domains with varying degrees of mosaicity. XPS and UPS analyses yield U-4f, U-5f, O-1s, and O-2p electron binding energies consistent with reported bulk values. A change from p-type to n-type semiconductor behavior induced by preferential sputtering of oxygen during depth profile analysis was observed with both XPS and UPS. Trivalent cation impurities (Nd and Al) in UO{sub 2} lower the Fermi level, shifting the XPS spectral weight. This observation is consistent with hole-doping of a Mott-Hubbard insulator. The uranium oxide-(11{sup Macron }02) sapphire system is unstable with respect to Al interdiffusion across the film-substrate interface at elevated temperature. - Highlights: Black-Right-Pointing-Pointer Single crystal uranium-oxides grown on sapphire and yttria-stabilized zirconia. Black-Right-Pointing-Pointer Anion and cation valence states studied by photoelectron emission spectroscopy. Black-Right-Pointing-Pointer Trivalent Nd and Al impurities lower the Fermi level. Black-Right-Pointing-Pointer Uranium-oxide films on sapphire found to be unstable with respect to Al interdiffusion.

  13. Abrasion behavior of graphite pebble in lifting pipe of pebble-bed HTR

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke; Su, Jiageng [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Zhou, Hongbo [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Chinergy Co., LTD., Beijing 100193 (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Yu, Suyun, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 10084 (China)

    2015-11-15

    Highlights: • Quantitative determination of abrasion rate of graphite pebbles in different lifting velocities. • Abrasion behavior of graphite pebble in helium, air and nitrogen. • In helium, intensive collisions caused by oscillatory motion result in more graphite dust production. - Abstract: A pebble-bed high-temperature gas-cooled reactor (pebble-bed HTR) uses a helium coolant, graphite core structure, and spherical fuel elements. The pebble-bed design enables on-line refueling, avoiding refueling shutdowns. During circulation process, the pebbles are lifted pneumatically via a stainless steel lifting pipe and reinserted into the reactor. Inevitably, the movement of the fuel elements as they recirculate in the reactor produces graphite dust. Mechanical wear is the primary source of graphite dust production. Specifically, the sources are mechanisms of pebble–pebble contact, pebble–wall (structural graphite) contact, and fuel handling (pebble–metal abrasion). The key contribution to graphite dust production is from the fuel handling system, particularly from the lifting pipe. During pneumatic lift, graphite pebbles undergo multiple collisions with the stainless steel lifting pipe, thereby causing abrasion of the graphite pebbles and producing graphite dust. The present work explored the abrasion behavior of graphite pebble in the lifting pipe by measuring the abrasion rate at different lifting velocities. The abrasion rate of the graphite pebble in helium was found much higher than those in air and nitrogen. This gas environment effect could be explained by either tribology behavior or dynamic behavior. Friction testing excluded the possibility of tribology reason. The dynamic behavior of the graphite pebble was captured by analysis of the audio waveforms during pneumatic lift. The analysis results revealed unique dynamic behavior of the graphite pebble in helium. Oscillation and consequently intensive collisions occur during pneumatic lift, causing

  14. Surface coating of graphite pebbles for Korean HCCR TBM

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of); Yun, Young-Hoon, E-mail: yunh2@dsu.ac.kr [Dongshin University, Naju (Korea, Republic of); Park, Yi-Hyun; Ahn, Mu-Young; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Highlights: • A CVR-SiC coating was successfully formed on graphite pebbles for neutron reflector. • Dense and fine-grained surface morphologies of the SiC coatings were observed. • Oxidation resistance of the CVR-SiC-coated graphite pebbles was improved. - Abstract: The new concept of the recently modified Helium-Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) is to adopt a graphite reflector in the form of a pebble bed. A protective SiC coating is applied to the graphite pebbles to prohibit their reaction with steam or air as well as dust generation during TBM operation. In this research, the chemical vapor reaction (CVR) method was applied to fabricate SiC-coated graphite pebbles in a silica source. Relatively dense CVR-SiC coating was successfully formed on the graphite pebbles through the reduction of the graphite phase with SiO gas that was simply created from the silica source at 1850 °C (2 h). The microstructural features, XRD patterns, pore-size distribution and oxidation behavior of the SiC-coated graphite pebbles were investigated. To develop the practical process, which will be applied for mass production hereafter, a novel alternative method was applied to form the layer of SiC coating on the graphite pebbles over the silica source.

  15. Surface coating of graphite pebbles for Korean HCCR TBM

    International Nuclear Information System (INIS)

    Lee, Youngmin; Yun, Young-Hoon; Park, Yi-Hyun; Ahn, Mu-Young; Cho, Seungyon

    2014-01-01

    Highlights: • A CVR-SiC coating was successfully formed on graphite pebbles for neutron reflector. • Dense and fine-grained surface morphologies of the SiC coatings were observed. • Oxidation resistance of the CVR-SiC-coated graphite pebbles was improved. - Abstract: The new concept of the recently modified Helium-Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) is to adopt a graphite reflector in the form of a pebble bed. A protective SiC coating is applied to the graphite pebbles to prohibit their reaction with steam or air as well as dust generation during TBM operation. In this research, the chemical vapor reaction (CVR) method was applied to fabricate SiC-coated graphite pebbles in a silica source. Relatively dense CVR-SiC coating was successfully formed on the graphite pebbles through the reduction of the graphite phase with SiO gas that was simply created from the silica source at 1850 °C (2 h). The microstructural features, XRD patterns, pore-size distribution and oxidation behavior of the SiC-coated graphite pebbles were investigated. To develop the practical process, which will be applied for mass production hereafter, a novel alternative method was applied to form the layer of SiC coating on the graphite pebbles over the silica source

  16. Uranium conversion

    International Nuclear Information System (INIS)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

    2006-03-01

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

  17. Recent developments in graphite

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

  18. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

    1987-03-01

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source

  19. Field measurements of mixed exposure of operators to radioactive aerosol, gas and quartz in confinement of mining equipment cabs during open-pit mining of high-grade uranium ores

    International Nuclear Information System (INIS)

    Yoshida, K.; Atiemo, M.A.; Markham, J.W.

    1982-07-01

    A series of field measurements of miners mixed exposure to radon and daughters, uranium ore dust and respirable quartz, was conducted in an open-pit mine in Northern Saskatchewan during 1980-81. Control of radon gas levels in the mining equipment cabs is required. Dust may be reduced by minimizing the resuspension of dust from contaminated surfaces within the cabs

  20. Uranium exploration

    International Nuclear Information System (INIS)

    De Voto, R.H.

    1984-01-01

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

  1. Emanation of /sup 232/U daughter products from submicrometer particles of uranium oxide and thorium dioxide by nuclear recoil and inert gas diffusion

    Energy Technology Data Exchange (ETDEWEB)

    Coombs, M.A.; Cuddihy, R.G. (Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM (USA). Inhalation Toxicology Research Inst.)

    1983-01-01

    Emanation of /sup 232/U daughter products by nuclear recoil and inert gas diffusion from spherical, submicrometer particles of uranium oxide and thorium dioxide was studied. Monodisperse samples of particles containing 1% /sup 232/U and having physical diameters between 0.1 and 1 ..mu..m were used for the emanation measurements. Thorium-228 ions recoiling from the particles after alpha-decay of /sup 232/U were collected electrostatically on a recoil cathode. Radon-220 diffusing from the particles was swept by an airstream into a 4 l. chamber where the /sup 220/Rn daughters were collected on a second cathode. Mathematical models of radionuclide emanation from spherical particles were used to calculate the recoil range of /sup 228/Th and the diffusion coefficient of /sup 220/Rn in the particle matrix. A /sup 228/Th recoil range of 0.02 ..mu..m and a /sup 220/Rn diffusion coefficient of 3 x 10/sup -14/ cm/sup 2//sec were obtained in both uranium oxide and thorium dioxide particles.

  2. Graphite core design in UK reactors

    International Nuclear Information System (INIS)

    Davies, M.W.

    1996-01-01

    The cores in the first power producing Magnox reactors in the UK were designed with only a limited amount of information available regarding the anisotropic dimensional change behaviour of Pile Grade graphite. As more information was gained it was necessary to make modifications to the design, some minor, some major. As the cores being built became larger, and with the switch to the Advanced Gas-cooled Reactor (AGR) with its much higher power density, additional problems had to be overcome such as increased dimensional change and radiolytic oxidation by the carbon dioxide coolant. For the AGRs a more isotropic graphite was required, with a lower initial open pore volume and higher strength. Gilsocarbon graphite was developed and was selected for all the AGRs built in the UK. Methane bearing coolants are used to limit radiolytic oxidation. (author). 5 figs

  3. Nondestructive evaluation of nuclear-grade graphite

    Science.gov (United States)

    Kunerth, D. C.; McJunkin, T. R.

    2012-05-01

    The material of choice for the core of the high-temperature gas-cooled reactors being developed by the U.S. Department of Energy's Next Generation Nuclear Plant Program is graphite. Graphite is a composite material whose properties are highly dependent on the base material and manufacturing methods. In addition to the material variations intrinsic to the manufacturing process, graphite will also undergo changes in material properties resulting from radiation damage and possible oxidation within the reactor. Idaho National Laboratory is presently evaluating the viability of conventional nondestructive evaluation techniques to characterize the material variations inherent to manufacturing and in-service degradation. Approaches of interest include x-ray radiography, eddy currents, and ultrasonics.

  4. The utilization of a pressurized-graphite/water/oxygen mixture for irradiated graphite incineration

    International Nuclear Information System (INIS)

    Antonini, G.; Perotin, J.P.; Charlot, P.

    1992-01-01

    The authors demonstrate the interest of the utilization of a pressurized-graphite/water/oxygen mixture in the incineration of irradiated graphite. The aqueous phase comes in the form of a three-dimensional system that traps pressurized oxygen, the pulverulent solid being dispersed at the liquid/gas interfaces. These three-phasic formulations give the following advantages: reduction of the apparent viscosity of the mixture in comparison with a solid/liquid mixture at the same solid concentration; reduction of the solid/liquid interactions; self-pulverizability. thus promoting reduction of the flame length utilization of conventional burners; reduction of the flue gas flow rate; complete thermal destruction of graphite. (author)

  5. Exploitation Contradictions Concerning Multi-Energy Resources among Coal, Gas, Oil, and Uranium: A Case Study in the Ordos Basin (Western North China Craton and Southern Side of Yinshan Mountains

    Directory of Open Access Journals (Sweden)

    Xiaowei Feng

    2016-02-01

    Full Text Available The particular “rich coal, meager oil, and deficient gas” energy structure of China determines its high degree of dependence on coal resources. After over 100 years of high-intensity mining activities in Northeast China, East Region, and the Southern Region, coal mining in these areas is facing a series of serious problems, which force China’s energy exploitation map to be rewritten. New energy bases will move to the western and northern regions in the next few years. However, overlapping phenomena of multiple resources are frequently encountered. Previous exploitation mainly focused on coal mining, which destroys many mutualistic and accompanying resources, such as uranium, gas, and oil. Aiming at solving this unscientific development mode, this research presents a case study in the Ordos Basin, where uranium, coal, and gas/oil show a three-dimensional overlapping phenomenon along the vertical downward direction. The upper uranium and lower coal situation in this basin is remarkable; specifically, coal mining disturbs the overlaying aquifer, thus requiring the uranium to be leached first. The technical approach must be sufficiently reliable to avoid the leakage of radioactive elements in subsequent coal mining procedures. Hence, the unbalanced injection and extraction of uranium mining is used to completely eradicate the discharged emissions to the environment. The gas and oil are typically not extracted because of their deep occurrence strata and their overlapping phenomenon with coal seams. Use of the integrated coal and gas production method is recommended, and relevant fracturing methods to increase the gas migrating degree in the strata are also introduced. The results and recommendations in this study are applicable in some other areas with similarities.

  6. Fossile fuel and uranium resources

    International Nuclear Information System (INIS)

    Gorkum, A.A. van.

    1975-01-01

    The world's resources of coal, lignite, oil, natural gas, shale oil and uranium are reviewed. These quantities depend on the prices which make new resources exploitable. Uranium resources are given exclusively for the USSR, Eastern Europe and China. Their value in terms of energy depends heavily on the reactor type used. All figures given are estimated to be conservative

  7. Melting temperature of graphite

    International Nuclear Information System (INIS)

    Korobenko, V.N.; Savvatimskiy, A.I.

    2001-01-01

    Full Text: Pulse of electrical current is used for fast heating (∼ 1 μs) of metal and graphite specimens placed in dielectric solid media. Specimen consists of two strips (90 μm in thick) placed together with small gap so they form a black body model. Quasy-monocrystal graphite specimens were used for uniform heating of graphite. Temperature measurements were fulfilled with fast pyrometer and with composite 2-strip black body model up to melting temperature. There were fulfilled experiments with zirconium and tungsten of the same black body construction. Additional temperature measurements of liquid zirconium and liquid tungsten are made. Specific heat capacity (c P ) of liquid zirconium and of liquid tungsten has a common feature in c P diminishing just after melting. It reveals c P diminishing after melting in both cases over the narrow temperature range up to usual values known from steady state measurements. Over the next wide temperature range heat capacity for W (up to 5000 K) and Zr (up to 4100 K) show different dependencies of heat capacity on temperature in liquid state. The experiments confirmed a high quality of 2-strip black body model used for graphite temperature measurements. Melting temperature plateau of tungsten (3690 K) was used for pyrometer calibration area for graphite temperature measurement. As a result, a preliminary value of graphite melting temperature of 4800 K was obtained. (author)

  8. 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

  9. Evaluation of the significance of inverse oxidation for HTGR graphites

    International Nuclear Information System (INIS)

    Lee, B.S.; Heiser, J. III; Sastre, C.

    1983-01-01

    The inverse oxidation refers to a higher mass loss inside the graphite than the outside. In 1980, Wichner et al reported this phenomenon (referred to as inside/out corrosion) observed in some H451 graphites, and offered an explanation that a catalyst (almost certainly Fe) is activated by the progressively increasing reducing conditions found in the graphite interior. Recently, Morgan and Thomas (1982) investigated this phenomenon is PGX graphites, and agreed on the existing mechanism to explain this pheomenon. They also called for attention to the possibility that this phenomenon may occur under HTGR (High Temperature Gas-Cooled Reactor) operating conditions. The purpose of this paper is to confirm the above mentioned explanation for this phenomenon and to evaluate the significance of this effect for HTGR graphites under realistic reactor conditions

  10. 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

  11. Graphite reactor physics; Physique des piles a graphite

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Noc, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm{sup 2}, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors) [French] Entreprise il y a dix ans a l'occasion de la construction des piles de Marcoule, l'etude de la

  12. Nuclear graphite ageing and turnaround

    International Nuclear Information System (INIS)

    Marsden, B.J.; Hall, G.N.; Smart, J.

    2001-01-01

    Graphite moderated reactors are being operated in many countries including, the UK, Russia, Lithuania, Ukraine and Japan. Many of these reactors will operate well into the next century. New designs of High Temperature Graphite Moderated Reactors (HTRS) are being built in China and Japan. The design life of these graphite-moderated reactors is governed by the ageing of the graphite core due to fast neutron damage, and also, in the case of carbon dioxide cooled reactors by the rate of oxidation of the graphite. Nuclear graphites are polycrystalline in nature and it is the irradiation-induced damage to the individual graphite crystals that determines the material property changes with age. The life of a graphite component in a nuclear reactor can be related to the graphite irradiation induced dimensional changes. Graphites typically shrink with age, until a point is reached where the shrinkage stops and the graphite starts to swell. This change from shrinkage to swelling is known as ''turnaround''. It is well known that pre-oxidising graphite specimens caused ''turnaround'' to be delayed, thus extending the life of the graphite, and hence the life of the reactor. However, there was no satisfactory explanation of this behaviour. This paper presents a numerical crystal based model of dimensional change in graphite, which explains the delay in ''turnaround'' in the pre-oxidised specimens irradiated in a fast neutron flux, in terms of crystal accommodation and orientation and change in compliance due to radiolytic oxidation. (author)

  13. Gas explosions and thermal runaways during external heating abuse of commercial lithium-ion graphite-LiCoO2 cells at different levels of ageing

    Science.gov (United States)

    Larsson, Fredrik; Bertilsson, Simon; Furlani, Maurizio; Albinsson, Ingvar; Mellander, Bengt-Erik

    2018-01-01

    Commercial 6.8 Ah lithium-ion cells with different ageing/status have been abused by external heating in an oven. Prior to the abuse test, selected cells were aged either by C/2 cycling up to 300 cycles or stored at 60 °C. Gas emissions were measured by FTIR and three separate vents were identified, two well before the thermal runaway while the third occurred simultaneously with the thermal runaway releasing heavy smoke and gas. Emissions of toxic carbon monoxide (CO), hydrogen fluoride (HF) and phosphorous oxyfluoride (POF3) were detected in the third vent, regardless if there was a fire or not. All abused cells went into thermal runaway and emitted smoke and gas, the working cells also released flames as well as sparks. The dead cells were however less reactive but still underwent thermal runaway. For about half of the working cells, for all levels of cycle ageing, ignition of the accumulated battery released gases occurred about 15 s after the thermal runaway resulting in a gas explosion. The thermal runaway temperature, about 190 °C, varied somewhat for the different cell ageing/status where a weak local minimum was found for cells cycled between 100 and 200 times.

  14. Contribution to the study of can deformations in the fuel elements of gas-graphite reactors during thermal cycling; Contribution a l'etude des deformations des gaines des elements combustibles de reacteur graphite-gaz au cours du cyclage thermique

    Energy Technology Data Exchange (ETDEWEB)

    Gauthron, M; Boudouresques, B; Delpeyroux, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The cans of fuel cartridges used in reactors of the gas-graphite type have either longitudinal fins of variable thickness, short herring-bone fins, or else a mixture of the two. An important test of the strength of these cartridges is their behaviour during thermal cycling carried out in cells reproducing in-pile conditions. It has been observed during with rapid cooling that there occurs a shortening at the base of the fins which can be accompanied in particular by a compression effect at the fin type, which has a tendency to curl, and by a tractive force acting on the body of the can at the ends of the longitudinal fins; this last phenomenon can result in a fracturing of the welds at the extremities or of the ends of the cartridge. This report presents first of all the way in which the stress diagram can be drawn for a can touching the fuel, and then the effect of the ratchet along a fin fixed to a bar with or without grooves. Finally the importance is shown of the test cycling variables (temperature, heating and cooling rates). (authors) [French] Les gaines des cartouches combustibles des reacteurs de la filiere graphite-gaz comportent soit des ailettes longitudinales plus ou moins epaisses, soit de courtes ailettes a chevrons, soit un ensemble des deux. Un test important de la tenue des cartouches, est la tenue au cyclage thermique en cellule pour reproduire le comportement en pile. On a observe au cours des cyclages a refroidissement rapide, un raccourcissement a la base des ailettes qui peut s'accompagner notamment d'une mise en compression du sommet de l'ailette qui a tendance a friser, et d'une traction exercee sur le corps des gaines au bout des ailettes longitudinales; ce dernier phenomene peut se traduire par des ruptures de soudures d'extremites ou des parties terminales de la cartouche. Ce rapport presente d'abord la maniere dont peut etre trace le diagramme des contraintes dans une gaine liee au combustible, puis l'effet du rochet le long d

  15. Yellowcake processing in uranium recovery

    International Nuclear Information System (INIS)

    Paul, J.M.

    1981-01-01

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

  16. Yellowcake processing in uranium recovery

    Energy Technology Data Exchange (ETDEWEB)

    Paul, J.M.

    1981-10-06

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

  17. Medium-size high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760 0 C (1400 0 F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

  18. The use of graphite for the reduction of void reactivity in CANDU reactors

    International Nuclear Information System (INIS)

    Min, B.J.; Kim, B.G.; Sim, K-S.

    1995-01-01

    Coolant void reactivity can be reduced by using burnable poison in CANDU reactors. The use of graphite in the fuel bundle is introduced to reduce coolant void reactivity by adding an appropriate amount of burnable poison in the central rod. This study shows that sufficiently low void reactivity which in controllable by Reactor Regulating System (RRS) can be achieved by using graphite used fuel with slightly enriched uranium. Zero void reactivity can be also obtained by using graphite used fuel with a large central rod. A new fuel bundle with graphite rods can substantially reduce the void reactivity with less burnup penalty compared to previously proposed low void reactivity fuel with depleted uranium. (author)

  19. Development of fracture toughness test method for nuclear grade graphite

    Energy Technology Data Exchange (ETDEWEB)

    Chang, C. H.; Lee, J. S.; Cho, H. C.; Kim, D. J.; Lee, D. J. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    Because of its high strength and stability at very high temperature, as well as very low thermal neutron absorption cross-section, graphite has been widely used as a structural material in Gas Cooled Reactors (GCR). Recently, many countries are developing the Very High Temperature gas cooled Reactor (VHTR) because of the potentials of hydrogen production, as well as its safety and viable economics. In VHTR, helium gas serves as the primary coolant. Graphite will be used as a reflector, moderator and core structural materials. The life time of graphite is determined from dimensional changes due to neutron irradiation, which closely relates to the changes of crystal structure. The changes of both lattice parameter and crystallite size can be easily measured by X-ray diffraction method. However, due to high cost and long time of neutron irradiation test, ion irradiation test is being performed instead in KAERI. Therefore, it is essential to develop the technique for measurement of ion irradiation damage of nuclear graphite. Fracture toughness of nuclear grade graphite is one of the key properties in the design and development of VHTR. It is important not only to evaluate the various properties of candidate graphite but also to assess the integrity of nuclear grade graphite during operation. Although fracture toughness tests on graphite have been performed in many laboratories, there have been wide variations in values of the calculated fracture toughness, due to the differences in the geometry of specimens and test conditions. Hence, standard test method for nuclear graphite is required to obtain the reliable fracture toughness values. Crack growth behavior of nuclear grade graphite shows rising R-curve which means the increase in crack growth resistance as the crack length increases. Crack bridging and microcracking have been proposed to be the dominant mechanisms of rising R-curve behavior. In this paper, the technique to measure the changes of crystallite size and

  20. Hydrogen storage in graphite nanofibers

    Energy Technology Data Exchange (ETDEWEB)

    Park, C.; Tan, C.D.; Hidalgo, R.; Baker, R.T.K.; Rodriguez, N.M. [Northeastern Univ., Boston, MA (United States). Chemistry Dept.

    1998-08-01

    Graphite nanofibers (GNF) are a type of material that is produced by the decomposition of carbon containing gases over metal catalyst particles at temperatures around 600 C. These molecularly engineered structures consist of graphene sheets perfectly arranged in a parallel, perpendicular or at angle orientation with respect to the fiber axis. The most important feature of the material is that only edges are exposed. Such an arrangement imparts the material with unique properties for gas adsorption because the evenly separated layers constitute the most ordered set of nanopores that can accommodate an adsorbate in the most efficient manner. In addition, the non-rigid pore walls can also expand so as to accommodate hydrogen in a multilayer conformation. Of the many varieties of structures that can be produced the authors have discovered that when gram quantities of a selected number of GNF are exposed to hydrogen at pressures of {approximately} 2,000 psi, they are capable of adsorbing and storing up to 40 wt% of hydrogen. It is believed that a strong interaction is established between hydrogen and the delocalized p-electrons present in the graphite layers and therefore a new type of chemistry is occurring within these confined structures.

  1. Recompressed exfoliated graphite articles

    Science.gov (United States)

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2013-08-06

    This invention provides an electrically conductive, less anisotropic, recompressed exfoliated graphite article comprising a mixture of (a) expanded or exfoliated graphite flakes; and (b) particles of non-expandable graphite or carbon, wherein the non-expandable graphite or carbon particles are in the amount of between about 3% and about 70% by weight based on the total weight of the particles and the expanded graphite flakes combined; wherein the mixture is compressed to form the article having an apparent bulk density of from about 0.1 g/cm.sup.3 to about 2.0 g/cm.sup.3. The article exhibits a thickness-direction conductivity typically greater than 50 S/cm, more typically greater than 100 S/cm, and most typically greater than 200 S/cm. The article, when used in a thin foil or sheet form, can be a useful component in a sheet molding compound plate used as a fuel cell separator or flow field plate. The article may also be used as a current collector for a battery, supercapacitor, or any other electrochemical cell.

  2. Activation calculations for dismantling - The feedback of a 7 years experience in activation calculations for graphite gas cooled reactors in France

    International Nuclear Information System (INIS)

    Eid, M.; Nimal, J.C.; Gerat, L.M.

    1994-01-01

    This is a revision of the past seven years experience in activation calculations for dismantling. It aims at evaluating the experience and at making better understanding to help in decision making during the following phases. Five gas cooled reactors are shutdown and are waiting for the EDF (Electricite De France) dismantling decision. The sixth (BUGEY1) will be shutdown by 1994 and will be waiting a dismantling decision as well. (authors). 3 figs., 3 tabs

  3. The status of graphite development for gas cooled reactors. Proceedings of a specialists` meeting held in Tokai, Japan, 9-12 September 1991

    Energy Technology Data Exchange (ETDEWEB)

    1993-02-01

    The meeting was convened by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors. It was attended by 61 participants from 6 countries. The meeting covered the following subjects: overview of national programs; design criteria, fracture mechanisms and component test; materials development and properties; non-destructive examination, inspection and surveillance. The participants presented 33 papers on behalf of their countries. A separate abstract was prepared for each of these papers. Refs, figs, tabs, photos and diagrams.

  4. Single nanocrystals of uranium dicarbide encapsulated in carbon

    International Nuclear Information System (INIS)

    Pasqualini, Enrique; Adelfang, Pablo

    1996-01-01

    The simultaneous volatilization of carbon and uranium compounds in an inert atmosphere produces the encapsulation of uranium dicarbide in graphitic material. These composite particles are of nanoscopic dimensions (smaller than 100 nanometers) and have the appearance of a black soot. Nanocapsules are as chemically inert as graphite and providers a more secure handling and processing of nuclear toxic materials. They preserve uranium dicarbide from environmental decomposition. The kernel of the nanoparticles is a single crystal of uranium dicarbide. The covers of the crystal are of graphitic structures of two types: parallel graphite layers and randomly oriented graphite crystallites. Both type of covers surround the UC 2 nanocrystal kernel. Density separation and measurements of specific area (BET technique) of purified soot were done. Particular adsorption and desorption kinetics of nitrogen monolayers indicate the presence of nanoparticles agglomeration. This is confirmed by direct observation by transmission electron microscopy (TEM). From density measurement and high resolution transmission electron microscopy (HRTEM) observation it can be inferred that there are gaps between kernel and cover and probably closed porosity inside clusters. Raman spectroscopy indicates only the presence of graphitic type carbon. Rietveld refinement of the nanoencapsulated material helps in the interpretation of X-Ray diffraction spectra. (author)

  5. Single nanocrystals of uranium dicarbide encapsulated in carbon

    Energy Technology Data Exchange (ETDEWEB)

    Pasqualini, Enrique; Adelfang, Pablo [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. de Combustibles Nucleares

    1996-07-01

    The simultaneous volatilization of carbon and uranium compounds in an inert atmosphere produces the encapsulation of uranium dicarbide in graphitic material. These composite particles are of nanoscopic dimensions (smaller than 100 nanometers) and have the appearance of a black soot. Nanocapsules are as chemically inert as graphite and providers a more secure handling and processing of nuclear toxic materials. They preserve uranium dicarbide from environmental decomposition. The kernel of the nanoparticles is a single crystal of uranium dicarbide. The covers of the crystal are of graphitic structures of two types: parallel graphite layers and randomly oriented graphite crystallites. Both type of covers surround the UC{sub 2} nanocrystal kernel. Density separation and measurements of specific area (BET technique) of purified soot were done. Particular adsorption and desorption kinetics of nitrogen monolayers indicate the presence of nanoparticles agglomeration. This is confirmed by direct observation by transmission electron microscopy (TEM). From density measurement and high resolution transmission electron microscopy (HRTEM) observation it can be inferred that there are gaps between kernel and cover and probably closed porosity inside clusters. Raman spectroscopy indicates only the presence of graphitic type carbon. Rietveld refinement of the nanoencapsulated material helps in the interpretation of X-Ray diffraction spectra. (author)

  6. Bromine intercalated graphite for lightweight composite conductors

    KAUST Repository

    Amassian, Aram; Patole, Archana

    2017-01-01

    A method of fabricating a bromine-graphite/metal composite includes intercalating bromine within layers of graphite via liquid-phase bromination to create brominated-graphite and consolidating the brominated-graphite with a metal nanopowder via a

  7. Effect of total pressure on graphite oxidation

    International Nuclear Information System (INIS)

    Burnette, R.D.; Hoot, C.G.

    1983-04-01

    Graphite corrosion in the high-temperature gas-cooled reactor (HTGR) is calculated using two key assumptions: (1) the kinetic, catalysis, and transport characteristics of graphite determined by bench-scale tests apply to large components at reactor conditions and (2) the effects of high pressure and turbulent flow are predictable. To better understand the differences between laboratory tests and reactor conditions, a high-pressure test loop (HPTL) has been constructed and used to perform tests at reactor temperature, pressure, and flow conditions. The HPTL is intended to determine the functional dependence of oxidation rate and characteristics on total pressure and gas velocity and to compare the oxidation results with calculations using models and codes developed for the reactor

  8. Cesium diffusion in graphite

    International Nuclear Information System (INIS)

    Evans, R.B. III; Davis, W. Jr.; Sutton, A.L. Jr.

    1980-05-01

    Experiments on diffusion of 137 Cs in five types of graphite were performed. The document provides a completion of the report that was started and includes a presentation of all of the diffusion data, previously unpublished. Except for data on mass transfer of 137 Cs in the Hawker-Siddeley graphite, analyses of experimental results were initiated but not completed. The mass transfer process of cesium in HS-1-1 graphite at 600 to 1000 0 C in a helium atmosphere is essentially pure diffusion wherein values of (E/epsilon) and ΔE of the equation D/epsilon = (D/epsilon) 0 exp [-ΔE/RT] are about 4 x 10 -2 cm 2 /s and 30 kcal/mole, respectively

  9. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  10. 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)

  11. Pyrolysis and its potential use in nuclear graphite disposal

    International Nuclear Information System (INIS)

    Mason, J.B.; Bradbury, D.

    2001-01-01

    Graphite is used as a moderator material in a number of nuclear reactor designs, such as MAGNOX and AGR gas cooled reactors in the United Kingdom and the RBMK design in Russia. During construction the moderator of the reactor is usually installed as an interlocking structure of graphite bricks. At the end of reactor life the graphite moderator, weighing typically 2,000 tonnes, is a radioactive waste which requires eventual management. Radioactive graphite disposal options conventionally include: In-situ SAFESTORE for extended periods to permit manual disassembly of the graphite moderator through decay of short-lived radionuclides. Robotic or manual disassembly of the reactor core followed by disposal of the graphite blocks. Robotic or manual disassembly of the reactor core followed by incineration of the graphite and release of the resulting carbon dioxide Studsvik, Inc. is a nuclear waste management and waste processing company organised to serve the US nuclear utility and government facilities. Studsvik's management and technical staff have a wealth of experience in processing liquid, slurry and solid low level radioactive waste using (amongst others) pyrolysis and steam reforming techniques. Bradtec is a UK company specialising in decontamination and waste management. This paper describes the use of pyrolysis and steam reforming techniques to gasify graphite leading to a low volume off-gas product. This allows the following options/advantages. Safe release of any stored Wigner energy in the graphite. The process can accept small pieces or a water-slurry of graphite, which enables the graphite to be removed from the reactor core by mechanical machining or water cutting techniques, applied remotely in the reactor fuel channels. In certain situations the process could be used to gasify the reactor moderator in-situ. The low volume of the off-gas product enables non-carbon radioactive impurities to be efficiently separated from the off-gas. The off-gas product can

  12. Interaction of the Helium, Hydrogen, Air, Argon, and Nitrogen Bubbles with Graphite Surface in Water.

    Science.gov (United States)

    Bartali, Ruben; Otyepka, Michal; Pykal, Martin; Lazar, Petr; Micheli, Victor; Gottardi, Gloria; Laidani, Nadhira

    2017-05-24

    The interaction of the confined gas with solid surface immersed in water is a common theme of many important fields such as self-cleaning surface, gas storage, and sensing. For that reason, we investigated the gas-graphite interaction in the water medium. The graphite surface was prepared by mechanical exfoliation of highly oriented pyrolytic graphite (HOPG). The surface chemistry and morphology were studied by X-ray photoelectron spectroscopy, profilometry, and atomic force microscopy. The surface energy of HOPG was estimated by contact angle measurements using the Owens-Wendt method. The interaction of gases (Ar, He, H 2 , N 2 , and air) with graphite was studied by a captive bubble method, in which the gas bubble was in contact with the exfoliated graphite surface in water media. The experimental data were corroborated by molecular dynamics simulations and density functional theory calculations. The surface energy of HOPG equaled to 52.8 mJ/m 2 and more of 95% of the surface energy was attributed to dispersion interactions. The results on gas-surface interaction indicated that HOPG surface had gasphilic behavior for helium and hydrogen, while gasphobic behavior for argon and nitrogen. The results showed that the variation of the gas contact angle was related to the balance between the gas-surface and gas-gas interaction potentials. For helium and hydrogen the gas-surface interaction was particularly high compared to gas-gas interaction and this promoted the favorable interaction with graphite surface.

  13. In-Situ Measurements of Low Enrichment Uranium Holdup Process Gas Piping at K-25 - Paper for Waste Management Symposia 2010 East Tennessee Technology Park Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Rasmussen, B.

    2010-01-01

    This document is the final version of a paper submitted to the Waste Management Symposia, Phoenix, 2010, abstract BJC/OR-3280. The primary document from which this paper was condensed is In-Situ Measurement of Low Enrichment Uranium Holdup in Process Gas Piping at K-25 Using NaI/HMS4 Gamma Detection Systems, BJC/OR-3355. This work explores the sufficiency and limitations of the Holdup Measurement System 4 (HJVIS4) software algorithms applied to measurements of low enriched uranium holdup in gaseous diffusion process gas piping. HMS4 has been used extensively during the decommissioning and demolition project of the K-25 building for U-235 holdup quantification. The HMS4 software is an integral part of one of the primary nondestructive assay (NDA) systems which was successfully tested and qualified for holdup deposit quantification in the process gas piping of the K-25 building. The initial qualification focused on the measurement of highly enriched UO 2 F 2 deposits. The purpose of this work was to determine if that qualification could be extended to include the quantification of holdup in UO 2 F 2 deposits of lower enrichment. Sample field data are presented to provide evidence in support of the theoretical foundation. The HMS4 algorithms were investigated in detail and found to sufficiently compensate for UO 2 F 2 source self-attenuation effects, over the range of expected enrichment (4-40%), in the North and East Wings of the K-25 building. The limitations of the HMS4 algorithms were explored for a described set of conditions with respect to area source measurements of low enriched UO 2 F 2 deposits when used in conjunction with a 1 inch by 1/2 inch sodium iodide (NaI) scintillation detector. The theoretical limitations of HMS4, based on the expected conditions in the process gas system of the K-25 building, are related back to the required data quality objectives (DQO) for the NBA measurement system established for the K-25 demolition project. The combined

  14. Graphite Oxidation Simulation in HTR Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  15. Graphite waste incineration in a fluidized bed

    International Nuclear Information System (INIS)

    Guiroy, J.J.

    1996-01-01

    French gas-cooled reactors belonging to the Atomic Energy Commission (CEA), Electricite de France (EDF), Hifrensa (Spain), etc., commissioned between the 1950s and 1970s, have generated large quantities of graphite wastes, mainly in the form of spent fuel sleeves. Furthermore, some of these reactors scheduled for dismantling in the near future (such as the G2 and G3 reactors at Marcoule) have cores consisting of graphite blocks. Consequently, a fraction of the contaminated graphite, amounting to 6000 t in France for example, must be processed in the coming years. For this processing, incineration using a circulating fluidized bed combustor has been selected as a possible solution and validated. However, the first operation to be performed involves recovering this graphite waste, and particularly, first of all, the spent fuel sleeves that were stored in silos during the years of reactor operation. Subsequent to the final shutdown of the Spanish gas-cooled reactor unit, Vandellos 1, the operating utility Hifrensa awarded contracts to a Framatome Iberica SA/ENSA consortium for removing, sorting, and prepackaging of the waste stored in three silos on the Vandellos site, essentially graphite sleeves. On the other hand, a program to validate the Framatome fluidized bed incineration process was carried out using a prototype incinerator installed at Le Creusot, France. The validation program included 22 twelve-hour tests and one 120-hour test. Particular attention was paid to the safety aspects of this project. During the performance of the validation program, a preliminary safety assessment was carried out. An impact assessment was performed with the help of the French Institute for Protection and Nuclear Safety, taking into account the preliminary spectra supplied by the CEA and EDF, and the activities of the radionuclides susceptible of being released into the atmosphere during the incineration. (author). 4 refs, 11 figs, 1 tab

  16. Intercomparison of graphite irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Hering, H; Perio, P; Seguin, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    While fast neutrons only are effective in damaging graphite, results of irradiations are more or less universally expressed in terms of thermal neutron fluxes. This paper attempts to correlate irradiations made in different reactors, i.e., in fluxes of different spectral compositions. Those attempts are based on comparison of 1) bulk length change and volume expansion, and 2) crystalline properties (e.g., lattice parameter C, magnetic susceptibility, stored energy, etc.). The methods used by various authors for determining the lattice constants of irradiated graphite are discussed. (author)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. Mineralogical study of uraniferous graphitic ore from Deogpyeong, Mogso and southern part of Daejeon area

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D J; Nam, S K [Korean Inst. of Energy and Resources, Seoul (Republic of Korea)

    1981-11-01

    Uranium minerals of torbernite, metatorbernite, metatyuyamunite and autunite have been identified from the uraniferous ores in graphitic beds of Ogcheon Group in Deogpyeong, Mogso and southern part of Daejeon area. Polarizing and ore microscopic studies, and chemical and X-ray powder diffraction analyses were carried out on the uraniferous graphite and associated materials. Main component minerals of uraniferous samples are graphite and quartz. Minor minerals are calcite, muscovite, sericite, andalusite, barite, kaolinite, hyaline opal, uranium minerals, sulfides such as pyrite, chalcopyrite, zincblende, and pyrrhotite, limonite, zeolite minerals such as laumontite and heulandite. Metatyuyamunite, torbernite, metatorbernite and autunite generally occur together with secondary minerals such as kaolinite, hyaline opal, calcite and limonite. They were found along the minor fissures or on the surface. Secondary uranium minerals described above were formed by supergenetic origin from primary uranium mineral. Uraniferous phosphatic nodule from Deogpyeong area are mainly composed of graphite and fluorapatite. And minor minerals are barite, quartz, muscovite and pyrite. Autoradiograph from uraniferous nodule shows that uranium enrichment in outer part of nodules is much higher than in inner part. This feature coincides with chemical analyses data of this uraniferous nodule.

  19. Graphite-based photovoltaic cells

    Science.gov (United States)

    Lagally, Max; Liu, Feng

    2010-12-28

    The present invention uses lithographically patterned graphite stacks as the basic building elements of an efficient and economical photovoltaic cell. The basic design of the graphite-based photovoltaic cells includes a plurality of spatially separated graphite stacks, each comprising a plurality of vertically stacked, semiconducting graphene sheets (carbon nanoribbons) bridging electrically conductive contacts.

  20. Laser and gas centrifuge enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Heinonen, Olli [Senior Fellow, Belfer Center for Science and International Affairs, Harvard Kennedy School, Cambridge, Massachusetts (United States)

    2014-05-09

    Principles of uranium isotope enrichment using various laser and gas centrifuge techniques are briefly discussed. Examples on production of high enriched uranium are given. Concerns regarding the possibility of using low end technologies to produce weapons grade uranium are explained. Based on current assessments commercial enrichment services are able to cover the global needs of enriched uranium in the foreseeable future.

  1. Evaluating the effectiveness of dilution of the recovered uranium with depleted uranium and low-enriched uranium to obtain fuel for VVER reactors

    International Nuclear Information System (INIS)

    Smirnov, A Yu; Sulaberidze, G A; Dudnikov, A A; Nevinitsa, V A

    2016-01-01

    The possibility of the recovered uranium enrichment in a cascade of gas centrifuges with three feed flows (depleted uranium, low-enriched uranium, recovered uranium) with simultaneous dilution of U-232,234,236 isotopes was shown. A series of numerical experiments were performed for different content of U-235 in low-enriched uranium. It has been demonstrated that the selected combination of diluents can simultaneously reduce the cost of separative work and the consumption of natural uranium, not only with respect to the previously used multi-flow cascade schemes, but also in comparison to the standard cascade for uranium enrichment. (paper)

  2. Polyphase diffusion of fission products in graphite

    International Nuclear Information System (INIS)

    Dannert, V.

    1989-05-01

    The report attempts to give an introduction into the subject of fission product transport in nuclear graphite and results in an extended proposal of a transport-model. Beginning with a rough description of the graphite in question, an idea about the physical transport-phenomena in graphite is developed. Some of the basic experimental methods, especially techniques of porosimetry, determination of sorption-isotherms and of course several transport-experiments, are briefly described and their results are discussed. Some of the most frequent transport models are introduced and assessed with the criteria emphasized in this report. An extended model is proposed including the following main ideas: The transport of the fission-products is regarded as a two-phase-diffusion process through the open pores of the graphite. The two phases are: surface-diffusion and gas-diffusion. A time-dependent coupling of the two diffusion-phases by sorption-isotherms and a concentration-dependence of the surface diffusion coefficient, also related to the physical behaviour of the sorption-isotherms, are the basic properties of the proposed model. (orig./HP) [de

  3. Model equations for Calculating Rn-gas Concentrations in Air of Uranium Exploratory Tunnels, Allouga, West-Central Sinai , Egypt

    International Nuclear Information System (INIS)

    Abdel-Monem, A.A.; Soliman, S.F.H.; Abd El-Kader, F.H.; El-Naggar, A.M.; Eissa, H.M.; Abd El-Hafez, A.A.

    2001-01-01

    Gabal Allouga area is located some 40 km due east from Abu Zenima town on the east coast of the Gulf of Suez, West-Central Sinai, Egypt. A network of exploratory tunnels totaling 670m in length and approximately 2x2 m in cross section, were excavated within a paleosol clayey bed. They host (Fe, Mn)-, Cu-, and U-mineralizations. Portions of the tunnels are naturally ventilated and others portions are non-ventilated and show ground water seepage through fractures. Model equations were developed for calculating the Rn-gas concentrations in the air of the tunnels under dry conditions where Rn-gas transport is mainly by air flow through porous media as well as for wet conditions where Rn-gas transport is mainly by ground water flow into the tunnels. Under dry conditions the model calculated Rn-gas concentrations(15.2-60.6 PCi/1) are consistent with measured values by active techniques (3.26-22.85 pCi/1) and by SSNTD techniques (19-69.1 pCi/1) when the Rn-emanation coefficient (alpha= 0.05-0.2), the emanating rock thickness (X=10 cm) and U-concentration averages 30 ppm. Under wet and non-ventilated conditions the model calculated Rn-gas concentrations (159-1248 pCi/1) are consistent with the measured values by active techniques (231-1348 pCi/1) and by SSNTD techniques (144-999pCi/1), when the Rn-emanation coefficient (alpha=0.1-0.25), the ground water flow (F=0.04-0.10 ml/s -1 cm -2 ) and U-concertrations (100-250ppm)

  4. Electronic properties of graphite

    International Nuclear Information System (INIS)

    Schneider, J.

    2010-10-01

    In this thesis, low-temperature magneto-transport (T ∼ 10 mK) and the de Haas-van Alphen effect of both natural graphite and highly oriented pyrolytic graphite (HOPG) are examined. In the first part, low field magneto-transport up to B = 11 T is discussed. A Fourier analysis of the background removed signal shows that the electric transport in graphite is governed by two types of charge carriers, electrons and holes. Their phase and frequency values are in agreement with the predictions of the SWM-model. The SWM-model is confirmed by detailed band structure calculations using the magnetic field Hamiltonian of graphite. The movement of the Fermi at B > 2 T is calculated self-consistently assuming that the sum of the electron and hole concentrations is constant. The second part of the thesis deals with high field magneto-transport of natural graphite in the magnetic field range 0 ≤ B ≤ 28 T. Both spin splitting of magneto-transport features in tilted field configuration and the onset of the charge density wave (CDW) phase for different temperatures with the magnetic field applied normal to the sample plane are discussed. Concerning the Zeeman effect, the SWM calculations including the Fermi energy movement require a g-factor of g* equal to 2.5 ± 0.1 to reproduce the spin spilt features. The measurements of the charge density wave state confirm that its onset magnetic field can be described by a Bardeen-Cooper-Schrieffer (BCS)-type formula. The measurements of the de Haas-van Alphen effect are in agreement with the results of the magneto-transport measurements at low field. (author)

  5. Device for the chlorinating volatization of thorium and/or uranium

    International Nuclear Information System (INIS)

    Laser, M.; Bohnenstingl, J.; Fischer, E.

    1976-01-01

    The invention relates to a device for the chlorinating volatization of nuclear fuel or breeding elements, preferably of thorium or uranium carbide and/or thorium or uranium oxide. A cupola furnace serves as reaction vessel, being connected to the condenser via a reactiongas pipe. In the interior of the cupola furnace, a tube is arranged consisting of porous graphite and heatable by direct current transfer, the tube being surrounded by a casing of ceramic material (sillimanite or rotosile). The chlorine gas is introduced into the cupola furnace via an additional socket. The gaseous reaction products penetrate the porous wall of the inner tube and enter the condenser first via the space between casing and inner tube and then via the reaction-gas pipe. The condenser consists of a gas-permeable tube and a surrounding tube, the space between being passed by the coolant gas. The maximum grain size of the ground pellets should be smaller than 300 μm. (HPH) [de

  6. Analytical standards for accountability of uranium hexafluoride - 1972

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    An analytical standard for the accountability of uranium hexafluoride is presented that includes procedures for subsampling, determination of uranium, determination of metallic impurities and isotopic analysis by gas and thermal ionization mass spectrometry

  7. Thermodynamic data for uranium fluorides

    International Nuclear Information System (INIS)

    Leitnaker, J.M.

    1983-03-01

    Self-consistent thermodynamic data have been tabulated for uranium fluorides between UF 4 and UF 6 , including UF 4 (solid and gas), U 4 F 17 (solid), U 2 F 9 (solid), UF 5 (solid and gas), U 2 F 10 (gas), and UF 6 (solid, liquid, and gas). Included are thermal function - the heat capacity, enthalpy, and free energy function, heats of formation, and vaporization behavior

  8. Uranium in Canada

    International Nuclear Information System (INIS)

    1989-01-01

    In 1988 Canada's five uranium producers reported output of concentrate containing a record 12,470 metric tons of uranium (tU), or about one third of total Western world production. Shipments exceeded 13,200 tU, valued at $Cdn 1.1 billion. Most of Canada's uranium output is available for export for peaceful purposes, as domestic requirements represent about 15 percent of production. The six uranium marketers signed new sales contracts for over 11,000 tU, mostly destined for the United States. Annual exports peaked in 1987 at 12,790 tU, falling back to 10,430 tU in 1988. Forward domestic and export contract commitments were more than 70,000 tU and 60,000 tU, respectively, as of early 1989. The uranium industry in Canada was restructured and consolidated by merger and acquisition, including the formation of Cameco. Three uranium projects were also advanced. The Athabasca Basin is the primary target for the discovery of high-grade low-cost uranium deposits. Discovery of new reserves in 1987 and 1988 did not fully replace the record output over the two-year period. The estimate of overall resources as of January 1989 was down by 4 percent from January 1987 to a total (measured, indicated and inferred) of 544,000 tU. Exploration expenditures reached $Cdn 37 million in 1987 and $59 million in 1988, due largely to the test mining programs at the Cigar Lake and Midwest projects in Saskatchewan. Spot market prices fell to all-time lows from 1987 to mid-1989, and there is little sign of relief. Canadian uranium production capability could fall below 12,000 tU before the late 1990s; however, should market conditions warrant output could be increased beyond 15,000 tU. Canada's known uranium resources are more than sufficient to meet the 30-year fuel requirements of those reactors in Canada that are now or are expected to be in service by the late 1990s. There is significant potential for discovering additional uranium resources. Canada's uranium production is equivalent, in

  9. Rn-Gas Concentration and Working Level Measurements Using SSNTD in Uranium Exploration Galleries Allouga Mine, Sinai, Egypt

    International Nuclear Information System (INIS)

    Abdel-Monem, A.A.; Hassan, S.F.; Abdel-Kader, F.H.; El-Naggar, A.M.; Essia, H.M.; Abdel-Hafez, A.A.

    2008-01-01

    Measurements of Rn-gas concentrations and Working Level (WL), were carried out in the U-exploration galleries at El- Allouga Mine, Sinai, Egypt by passive techniques (SSNTD) during the four seasons ( Summer, Fall, Winter and Spring) using four different detector types: CR-39, MK, CN-85 and LR-115.Twenty eight (28) stations were chosen for this purpose reflecting different environmental conditions as measurement sites within the galleries. The Rn-gas concentrations , in the summer period ranged from 25.86 to 44.2 pCi/l in the ventilated stations and from 488.98 to 611.16 pCi/l in the non-ventilated stations. In the fall period , the average Rn-gas concentrations in the ventilated stations ranged from 31.61 to 56.36 pCi/l while in the non-ventilated stations from 457.61 to 621.52 pCi/l. In the winter period, the average Rn-gas concentrations in the ventilated stations ranged from 27.59 to 66.45 pCi/l while in the non- ventilated stations from 499.09 to 603.81 pCi/l. In the spring the Rn-gas concentrations ranged from 30.8 to 46.77 pCi/l in the ventilated stations, whereas, in the non-ventilated stations from 404.06 to 445.18 pCi/l. The (WL), in the summer period, ranged from 0.143 to 0.247 in the ventilated sector and from 4.408 to 5.497 in the non-ventilated stations .In fall, the( WL) ranged from 0.166 to 0.295 in the ventilated stations and from 4.123 to 5.624 in the non-ventilated stations. In the winter, the (WL) ranged from 0.105 to 0.37 in the ventilated stations and from 4.138 to 5.26 in the non-ventilated stations. In the spring, the (WL) in the ventilated stations ranged from 0.152 to 0.241 and from 3.696 to 4.087 in the non-ventilated stations. These results indicate that: i)The low measured Rn gas and (WL) values in the ventilated stations reflect the effect of variations in meteorological conditions on (WL) determination where the air flow carries the Rn-gas before it decays and the daughters are plated onto the SSNTD . ii) The larger ranges for Rn-gas

  10. Hydrogeological investigations in the Harwell region: the use of environmental isotopes, inert gas contents, and the uranium decay series

    International Nuclear Information System (INIS)

    Alexander, J.; Andrews, J.N.

    1984-12-01

    A comprehensive range of environmental isotopes, radioelement and dissolved gas contents have been measured in groundwaters from the high permeability formations of the Harwell area. These analyses were undertaken as part of a hydrochemical validation of groundwater circulation patterns derived from potentiometric data. These investigations have focused upon the Corallian and Great Oolite formations since these sandwich the Oxford Clay. Geochemical, isotopic, radioelement and inert gas studies have demonstrated consistent trends which substantiate fluid migration patterns derived from hydraulic considerations. Groundwaters at downdip localities in both the Corallian and Great Oolite formations are the oldest waters sampled from the region. Variations in trends in parameters can be attributed to cross-formational flow and subsequent mixing of groundwaters. Individually these techniques can only provide limited information, but the combination of methods used have provided corroborative evidence concerning the direction of fluid circulation in the Harwell region. (author)

  11. Czechoslovak uranium

    International Nuclear Information System (INIS)

    Pluskal, O.

    1992-01-01

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

  12. Coupled reaction-diffusion equations to model the fission gas release in the irradiation of the uranium dioxide

    International Nuclear Information System (INIS)

    Moyano, Edgardo A.; Scarpettini, Alberto F.

    2003-01-01

    A semi linear model of weakly coupled parabolic p.d.e. with reaction-diffusion is investigated. The system describes fission gas transfer from grain interior of UO 2 to grain boundaries. The problem is studied in a bounded domain. Using the upper-lower solutions method, two monotone sequences for the finite differences equations are constructed. Reasons are mentioned that allow to affirm that in the proposed functional sector the algorithm converges to the unique solution of the differential system. (author)

  13. Analysis of electrochemical disintegration process of graphite matrix

    International Nuclear Information System (INIS)

    Tian Lifang; Wen Mingfen; Chen Jing

    2010-01-01

    The electrochemical method with ammonium nitrate as electrolyte was studied to disintegrate the graphite matrix from the simulative fuel elements for high temperature gas-cooled reactor. The influences of process parameters, including salt concentration, system temperature and current density, on the disintegration rate of graphite fragments were investigated in the present work. The experimental results showed that the disintegration rate depended slightly on the temperature and salt concentration. The current density strongly affected the disintegration rate of graphite fragments. Furthermore, the content of introduced oxygen in final graphite fragments was independent of the current density and the concentration of electrolyte. Moreover, the structural evolution of graphite was analyzed based on the microstructural parameters determined by X-ray diffraction profile fitting analysis using MAUD (material analysis using diffraction) before and after the disintegration process. It may safely be concluded that the graphite disintegration can be ascribed to the influences of the intercalation of foreign molecules in between crystal planes and the partial oxidation involved. The disintegration process was described deeply composed of intercalate part and further oxidation part of carbon which effected together to lead to the collapse of graphite crystals.

  14. Assessment of management modes for graphite from reactor decommissioning

    International Nuclear Information System (INIS)

    White, I.F.; Smith, G.M.; Saunders, L.J.; Kaye, C.J.; Martin, T.J.; Clarke, G.H.; Wakerley, M.W.

    1984-01-01

    A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts

  15. Characterization of nuclear graphite elastic properties using laser ultrasonic methods

    Science.gov (United States)

    Zeng, Fan W.; Han, Karen; Olasov, Lauren R.; Gallego, Nidia C.; Contescu, Cristian I.; Spicer, James B.

    2015-05-01

    Laser ultrasonic methods have been used to characterize the elastic behaviors of commercially-available and legacy nuclear graphites. Since ultrasonic techniques are sensitive to various aspects of graphite microstructure including preferred grain orientation, microcrack orientation and porosity, laser ultrasonics is a candidate technique for monitoring graphite degradation and structural integrity in environments expected in high-temperature, gas-cooled nuclear reactors. Aspects of materials texture can be assessed by studying ultrasonic wavespeeds as a function of propagation direction and polarization. Shear wave birefringence measurements, in particular, can be used to evaluate elastic anisotropy. In this work, laser ultrasonic measurements of graphite moduli have been made to provide insight into the relationship between the microstructures and the macroscopic stiffnesses of these materials. In particular, laser ultrasonic measurements have been made using laser line sources to produce shear waves with specific polarizations. By varying the line orientation relative to the sample, shear wave birefringence measurements have been recorded. Results from shear wave birefringence measurements show that an isostatically molded graphite, such as PCIB, behaves isotropically, while an extruded graphite, such as H-451, displays significant ultrasonic texture. Graphites have complicated microstructures that depend on the manufacturing processes used, and ultrasonic texture in these materials could originate from grain orientation and preferred microcrack alignment. Effects on material isotropy due to service related microstructural changes are possible and the ultimate aim of this work is to determine the degree to which these changes can be assessed nondestructively using laser ultrasonics measurements.

  16. Discussion of metallogenic substance source of Xiangshan uranium orefield

    International Nuclear Information System (INIS)

    Shao Fei; Tang Xiangsheng; Zou Maoqin; Hu Maomei; He Xiaomei; Chen Xiaoming; Xu Hengli

    2008-01-01

    Analysis of uranium source is a key problem for study on uranium deposit genesis. Based on analysis of general implication for determination of uranium source on distribution characteristics of regional uranium abundance, according to temporal and spatial evolution of regional metallogenic substances in process of geological history, and combining with indication for analysis of uranium source by Pb isotopic composition of ores and REE geochemistry of both rocks and ores in Xiangshan orefield, Lower Cambrian strata are determined as regional uranium source bed, Xiangshan volcanic basin is the accumulation area for regional metallogenic substances, magma and hydrothermal solution of post magmatism are media for uranium. Magmatism realizes uranium migration from 'source' to 'accumulation'. In process of magmatic evolution, uranium transformed into gas phase to provide substance base for uranium mineralization. Fluid-rock interaction of post magmatism also promoted some uranium from schist of the basement and rhyodacite into metallogenic solution. (authors)

  17. Corrosion behavior of a positive graphite electrode in vanadium redox flow battery

    International Nuclear Information System (INIS)

    Liu Huijun; Xu Qian; Yan Chuanwei; Qiao Yonglian

    2011-01-01

    Graphical abstract: The overpotential for gas evolution on positive graphite electrode decreases due to the functional groups of COOH and C=O introduced on the surface of graphite electrode during corrosion process, which can self-catalyze the oxidation of carbon atoms therefore, accelerates corrosion process. Highlights: → Initial potential for gas evolution is higher than 1.60 V vs SCE. → Factors affecting the graphite corrosion are investigated. → Functional groups of COOH and C=O introduced during corrosion process. → The groups can self-catalyze the oxidation of carbon atoms. - Abstract: The graphite plate is easily suffered from corosion because of CO 2 evolution when it acts as the positive electrode for vanadium redox flow battery. The aim is to obtain the initial potential for gas evolution on a positive graphite electrode in 2 mol dm -3 H 2 SO 4 + 2 mol dm -3 VOSO 4 solution. The effects of polarization potential, operating temperature and polarization time on extent of graphite corrosion are investigated by potentiodynamic and potentiostatic techniques. The surface characteristics of graphite electrode before and after corrosion are examined by scanning electron microscopy, atomic force microscopy, and X-ray photoelectron spectroscopy. The results show that the gas begins to evolve on the graphite electrode when the anodic polarization potential is higher than 1.60 V vs saturated calomel electrode at 20 deg. C. The CO 2 evolution on the graphite electrode can lead to intergranular corrosion of the graphite when the polarization potential reaches 1.75 V. In addition, the functional groups of COOH and C=O introduced on the surface of graphite electrode during corrosion can catalyze the formation of CO 2 , therefore, accelerates the corrosion rate of graphite electrode.

  18. Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

    2002-12-01

    The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

  19. Graphitization in Carbon MEMS and Carbon NEMS

    Science.gov (United States)

    Sharma, Swati

    Carbon MEMS (CMEMS) and Carbon NEMS (CNEMS) are an emerging class of miniaturized devices. Due to the numerous advantages such as scalable manufacturing processes, inexpensive and readily available precursor polymer materials, tunable surface properties and biocompatibility, carbon has become a preferred material for a wide variety of future sensing applications. Single suspended carbon nanowires (CNWs) integrated on CMEMS structures fabricated by electrospinning of SU8 photoresist on photolithographially patterned SU8 followed by pyrolysis are utilized for understanding the graphitization process in micro and nano carbon materials. These monolithic CNW-CMEMS structures enable the fabrication of very high aspect ratio CNWs of predefined length. The CNWs thus fabricated display core---shell structures having a graphitic shell with a glassy carbon core. The electrical conductivity of these CNWs is increased by about 100% compared to glassy carbon as a result of enhanced graphitization. We explore various tunable fabrication and pyrolysis parameters to improve graphitization in the resulting CNWs. We also suggest gas-sensing application of the thus fabricated single suspended CNW-CMEMS devices by using the CNW as a nano-hotplate for local chemical vapor deposition. In this thesis we also report on results from an optimization study of SU8 photoresist derived carbon electrodes. These electrodes were applied to the simultaneous detection of traces of Cd(II) and Pb(II) through anodic stripping voltammetry and detection limits as low as 0.7 and 0.8 microgL-1 were achieved. To further improve upon the electrochemical behavior of the carbon electrodes we elucidate a modified pyrolysis technique featuring an ultra-fast temperature ramp for obtaining bubbled porous carbon from lithographically patterned SU8. We conclude this dissertation by suggesting the possible future works on enhancing graphitization as well as on electrochemical applications

  20. On estimating the fracture probability of nuclear graphite components

    International Nuclear Information System (INIS)

    Srinivasan, Makuteswara

    2008-01-01

    The properties of nuclear grade graphites exhibit anisotropy and could vary considerably within a manufactured block. Graphite strength is affected by the direction of alignment of the constituent coke particles, which is dictated by the forming method, coke particle size, and the size, shape, and orientation distribution of pores in the structure. In this paper, a Weibull failure probability analysis for components is presented using the American Society of Testing Materials strength specification for nuclear grade graphites for core components in advanced high-temperature gas-cooled reactors. The risk of rupture (probability of fracture) and survival probability (reliability) of large graphite blocks are calculated for varying and discrete values of service tensile stresses. The limitations in these calculations are discussed from considerations of actual reactor environmental conditions that could potentially degrade the specification properties because of damage due to complex interactions between irradiation, temperature, stress, and variability in reactor operation

  1. Impact-Contact Analysis of Prismatic Graphite Blocks Using Abaqus

    International Nuclear Information System (INIS)

    Kang, Ji Ho; Kim, Gyeong Ho; Choi, Woo Seok

    2010-12-01

    Graphite blocks are the important core components of the high temperature gas-cooled reactor. As these blocks are simply stacked in array, collisions among neighboring components may occur during earthquakes or accidents. The final objective of the research project is to develop a reliable seismic model of the stacked graphite blocks from which their behavior can be predicted and, thus, they are designed to have sufficient strength to maintain their structural integrity during the anticipated occurrences. The work summarized in this report is a first step toward the big picture and is dedicated to build a realistic impact-contact dynamics model of the graphite block using a commercial FEM package, Abaqus. The developed model will be further used to assist building a reliable lumped dynamics model of these stacked graphite components

  2. TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

    International Nuclear Information System (INIS)

    Weicht, U.; Mueller, W.

    1983-01-01

    1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

  3. Uranium dioxide calcining apparatus

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  4. Uranium ores

    International Nuclear Information System (INIS)

    Poty, B.; Roux, J.

    1998-01-01

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

  5. Uranium market

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. Uranium enrichment

    International Nuclear Information System (INIS)

    1990-01-01

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

  7. Harwell Graphite Calorimeter

    International Nuclear Information System (INIS)

    Linacre, J.K.

    1970-01-01

    The calorimeter is of the steady state temperature difference type. It contains a graphite sample supported axially in a graphite outer jacket, the assembly being contained in a thin stainless steel outer can. The temperature of the jacket and the temperature difference between sample and jacket are measured by chromel-alumel thermocouples. The instrument is calibrated by means of an electric heater of low mass positioned on the axis of the sample. The resistance of the heater is known and both current through the heater and the potential across it may be measured. The instrument is filled with nitrogen at a pressure of one half atmosphere at room temperature. The calorimeter has been designed for prolonged operation at temperatures up to 600°C, and dose rates up to 1 Wg -1 , and instruments have been in use for periods in excess of one year

  8. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2009-04-15

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than {approx}7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  9. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Science.gov (United States)

    Rest, J.; Hofman, G. L.; Kim, Yeon Soo

    2009-04-01

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  10. 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

  11. A standard graphite block

    Energy Technology Data Exchange (ETDEWEB)

    Ivkovic, M; Zdravkovic, Z; Sotic, O [Department of Reactor Physics and Dynamics, Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1966-04-15

    A graphite block was calibrated for the thermal neutron flux of the Ra-Be source using indium foils as detectors. Experimental values of the thermal neutron flux along the central vertical axis of the system were corrected for the self-shielding effect and depression of flux in the detector. The experimental values obtained were compared with the values calculated on the basis of solving the conservation neutron equation by the continuous slowing-down theory. In this theoretical calculation of the flux the Ra-Be source was divided into three resonance energy regions. The measurement of the thermal neutron diffusion length in the standard graphite block is described. The measurements were performed in the thermal neutron region of the system. The experimental results were interpreted by the diffusion theory for point thermal neutron source in the finite system. The thermal neutron diffusion length was calculated to be L= 50.9 {+-}3.1 cm for the following graphite characteristics: density = 1.7 g/cm{sup 3}; boron content = 0.1 ppm; absorption cross section = 3.7 mb.

  12. A standard graphite block

    International Nuclear Information System (INIS)

    Ivkovic, M.; Zdravkovic, Z.; Sotic, O.

    1966-04-01

    A graphite block was calibrated for the thermal neutron flux of the Ra-Be source using indium foils as detectors. Experimental values of the thermal neutron flux along the central vertical axis of the system were corrected for the self-shielding effect and depression of flux in the detector. The experimental values obtained were compared with the values calculated on the basis of solving the conservation neutron equation by the continuous slowing-down theory. In this theoretical calculation of the flux the Ra-Be source was divided into three resonance energy regions. The measurement of the thermal neutron diffusion length in the standard graphite block is described. The measurements were performed in the thermal neutron region of the system. The experimental results were interpreted by the diffusion theory for point thermal neutron source in the finite system. The thermal neutron diffusion length was calculated to be L= 50.9 ±3.1 cm for the following graphite characteristics: density = 1.7 g/cm 3 ; boron content = 0.1 ppm; absorption cross section = 3.7 mb

  13. 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

  14. 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

  15. Structural disorder of graphite and implications for graphite thermometry

    Science.gov (United States)

    Kirilova, Martina; Toy, Virginia; Rooney, Jeremy S.; Giorgetti, Carolina; Gordon, Keith C.; Collettini, Cristiano; Takeshita, Toru

    2018-02-01

    Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25 megapascal (MPa) and aseismic velocities of 1, 10 and 100 µm s-1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

  16. Structural disorder of graphite and implications for graphite thermometry

    Directory of Open Access Journals (Sweden)

    M. Kirilova

    2018-02-01

    Full Text Available Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25  megapascal (MPa and aseismic velocities of 1, 10 and 100 µm s−1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

  17. Bromine intercalated graphite for lightweight composite conductors

    KAUST Repository

    Amassian, Aram

    2017-07-20

    A method of fabricating a bromine-graphite/metal composite includes intercalating bromine within layers of graphite via liquid-phase bromination to create brominated-graphite and consolidating the brominated-graphite with a metal nanopowder via a mechanical pressing operation to generate a bromine-graphite/metal composite material.

  18. Chemical stabilization of graphite surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Bistrika, Alexander A.; Lerner, Michael M.

    2018-04-03

    Embodiments of a device, or a component of a device, including a stabilized graphite surface, methods of stabilizing graphite surfaces, and uses for the devices or components are disclosed. The device or component includes a surface comprising graphite, and a plurality of haloaryl ions and/or haloalkyl ions bound to at least a portion of the graphite. The ions may be perhaloaryl ions and/or perhaloalkyl ions. In certain embodiments, the ions are perfluorobenzenesulfonate anions. Embodiments of the device or component including stabilized graphite surfaces may maintain a steady-state oxidation or reduction surface current density after being exposed to continuous oxidation conditions for a period of at least 1-100 hours. The device or component is prepared by exposing a graphite-containing surface to an acidic aqueous solution of the ions under oxidizing conditions. The device or component can be exposed in situ to the solution.

  19. Uranium mining

    International Nuclear Information System (INIS)

    Lange, G.

    1975-01-01

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

  20. Uranium enrichment

    International Nuclear Information System (INIS)

    1989-01-01

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

  1. Uranium resources

    International Nuclear Information System (INIS)

    1976-01-01

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

  2. Heat exchanger using graphite foam

    Science.gov (United States)

    Campagna, Michael Joseph; Callas, James John

    2012-09-25

    A heat exchanger is disclosed. The heat exchanger may have an inlet configured to receive a first fluid and an outlet configured to discharge the first fluid. The heat exchanger may further have at least one passageway configured to conduct the first fluid from the inlet to the outlet. The at least one passageway may be composed of a graphite foam and a layer of graphite material on the exterior of the graphite foam. The layer of graphite material may form at least a partial barrier between the first fluid and a second fluid external to the at least one passageway.

  3. Uranium absorption study pile; Empilement pour le controle de l'absorption de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Raievski, V; Sautiez, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF{sub 3} counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10{sup -6} boron per gr. of uranium) or better. (author) [French] Nous decrivons un dispositif permettant de mesurer l'absorption des elements combustibles d'une pile. Ce dispositif est constitue par un empilement de graphite dont la region centrale est formee par un reseau regulier de barres d'uranium. Des sources de RaBe et des compteurs a BF{sub 3} sont places de part et d'autre de cette region. En comparant un chargement d'uranium a un chargement etalon d'environ 560 kg, on peut determiner la difference d'absorption entre ces deux chargements. La sensibilite permettrait de deceler une variation d'absorption de l'ordre du ppm de bore (10{sup -6} g de bore par gramme d'uranium) et peut-etre mieux. (auteur)

  4. Spectrographic determination of impurities in uranium tetrafluoride

    International Nuclear Information System (INIS)

    Capdevila Perez, C.; Roca Adell, M.; Alvarez Gonzalez, F.

    1967-01-01

    A carrier distillation method for the determination of Ag, Al, As, B, Cd, Cr, Cu, Fe, Mg, Mn, Ni, Pb, and Si in uranium tetrafluoride was develop ped. the previous addition of 25% Y 2 3 prevents the excitation of uranium by conversion of the volatile UF 4 into U 3 0 8 during the arc discharge. NaCl or Ga 2 0 3 , containing Ge and V as internal standards, are used as carriers, and samples are arced in 10 Amp. d.c. arc in a graphite anode cup. 7 mm diameter, 10 mm deep, being the weight of charge 300 mg. (Author) 14 refs

  5. Uranium supply and demand

    Energy Technology Data Exchange (ETDEWEB)

    Spriggs, M J

    1976-01-01

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

  6. Role of nuclear grade graphite in controlling oxidation in modular HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Windes, Willaim [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kane, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of core environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.

  7. Kinetics of Chronic Oxidation of NBG-17 Nuclear Graphite by Water Vapor

    Energy Technology Data Exchange (ETDEWEB)

    Contescu, Cristian I [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burchell, Timothy D [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mee, Robert [Univ. of Tennessee, Knoxville, TN (United States)

    2015-05-01

    This report presents the results of kinetic measurements during accelerated oxidation tests of NBG-17 nuclear graphite by low concentration of water vapor and hydrogen in ultra-high purity helium. The objective is to determine the parameters in the Langmuir-Hinshelwood (L-H) equation describing the oxidation kinetics of nuclear graphite in the helium coolant of high temperature gas-cooled reactors (HTGR). Although the helium coolant chemistry is strictly controlled during normal operating conditions, trace amounts of moisture (predictably < 0.2 ppm) cannot be avoided. Prolonged exposure of graphite components to water vapor at high temperature will cause very slow (chronic) oxidation over the lifetime of graphite components. This behavior must be understood and predicted for the design and safe operation of gas-cooled nuclear reactors. The results reported here show that, in general, oxidation by water of graphite NBG-17 obeys the L-H mechanism, previously documented for other graphite grades. However, the characteristic kinetic parameters that best describe oxidation rates measured for graphite NBG-17 are different than those reported previously for grades H-451 (General Atomics, 1978) and PCEA (ORNL, 2013). In some specific conditions, certain deviations from the generally accepted L-H model were observed for graphite NBG-17. This graphite is manufactured in Germany by SGL Carbon Group and is a possible candidate for the fuel elements and reflector blocks of HTGR.

  8. Studies of the role of molten materials in interactions with UO2 and graphite

    International Nuclear Information System (INIS)

    Fink, J.K.; Heiberger, J.J.; Leibowitz, L.

    1979-01-01

    Graphite, which is being considered as a lower reactor shield in gas-cooled fast reactors, would be contacted by core debris during a core disruptive accident. Information on the interaction of graphite, UO 2 , and stainless steel is needed in assessing the safety of the GCFR. In an ongoing study of the interaction of graphite, UO 2 , and stainless steel, the effects of the steel components have been investigated by electron microprobe scans, x-ray diffraction, and reaction-rate measurements. Experiments to study the role of the reaction product, FeUC 2 , in the interaction suggested that FeUC 2 promotes the interaction by acting as a carrier to bring graphite to the reaction site. Additional experiments using pyrolytic graphite show that while the reaction rate is decreased at 2400 K, at higher temperatures the rate is similar to that using other grades of graphite

  9. Thermogravimetric and Differential Scanning Calorimetric Behavior of Ball-Milled Nuclear Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eung Seon; Kim, Min Hwan; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi Hyun; Cho, Seung Yon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    An examination was made to characterize the oxidation behavior of ball-milled nuclear graphite powder through a TG-DSC analysis. With the ball milling time, the BET surface area increased with the reduction of particle size, but decreased with the chemisorptions of O{sub 2} on the activated surface. The enhancement of the oxidation after the ball milling is attributed to both increases in the specific surface area and atomic scale defects in the graphite structure. In a high temperature gas-cooled reactor, nuclear graphite has been widely used as fuel elements, moderator or reflector blocks, and core support structures owing to its excellent moderating power, mechanical properties and machinability. For the same reason, it will be used in a helium cooled ceramic reflector test blanket module for the ITER. Each submodule has a seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebbles packed tritium breeder layers, and a reflector layer packed with 1 mm diameter graphite pebbles to reduce the volume of beryllium. The abrasion of graphite structures owing to relative motion or thermal cycle during operation may produce graphite dust. It is expected that graphite dust will be more oxidative than bulk graphite, and thus the oxidation behavior of graphite dust must be examined to analyze the safety of the reactors during an air ingress accident. In this study, the thermal stability of ball-milled graphite powder was investigated using a simultaneous thermogravimeter-differential scanning calorimeter.

  10. Surface area-burnoff correlation for the steam--graphite reaction

    International Nuclear Information System (INIS)

    Stark, W.A. Jr.; Malinauskas, A.P.

    1977-01-01

    The oxidation of core graphite by steam of air represents a problem area of significant concern in safety analyses for the high temperature gas cooled reactor (HTGR). Core and core-support graphite integrity and strength deteriorate with oxidation of the graphite, and oxidation furthermore could affect the rate of fission product release under upset conditions. Consequently, modeling of core response during steam or air ingress conditions requires an expression for the rate of graphite interaction with those impurities. The steam--graphite reaction in particular is a complex interaction of mass transport within the graphite with chemi-sorption and reaction on accessible surfaces; experimental results from graphite to graphite are highly variable, and the description of the reaction is not yet completely consistent. A simple etch pit model relating surface area to burnoff has been proposed and shown to provide reasonable correlation with experimental data obtained from steam oxidation studies of nuclear grade H-327 graphite. Unaccounted differences between theory and experiment arise at burnoffs exceeding 3 to 5 percent. The model, while not complete nor comprehensive, is consistent with experimental observations of graphite oxidation by O 2 (air), CO 2 , or H 2 O, and could have some utility in safety analysis

  11. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  12. A METHOD OF PREPARING URANIUM DIOXIDE

    Science.gov (United States)

    Scott, F.A.; Mudge, L.K.

    1963-12-17

    A process of purifying raw, in particular plutonium- and fission- products-containing, uranium dioxide is described. The uranium dioxide is dissolved in a molten chloride mixture containing potassium chloride plus sodium, lithium, magnesium, or lead chloride under anhydrous conditions; an electric current and a chlorinating gas are passed through the mixture whereby pure uranium dioxide is deposited on and at the same time partially redissolved from the cathode. (AEC)

  13. Purification and preparation of graphite oxide from natural graphite

    Energy Technology Data Exchange (ETDEWEB)

    Panatarani, C., E-mail: c.panatarani@phys.unpad.ac.id; Muthahhari, N.; Joni, I. Made [Instrumentation Systems and Functional Material Processing Laboratory, Department of Physics, Faculty of Mathematics and Natural Sciences, Universitas Padjadjaran, Padjadjaran University, Jl. Raya Bandung-Sumedang KM 21, Jatinangor, 45363, Jawa Barat (Indonesia); Rianto, Anton [Grafindo Nusantara Ltd., Belagio Mall Lantai 2, Unit 0 L3-19, Kawasan Mega Kuningan, Kav. B4 No.3, Jakarta Selatan (Indonesia)

    2016-03-11

    Graphite oxide has attracted much interest as a possible route for preparation of natural graphite in the large-scale production and manipulation of graphene as a material with extraordinary electronic properties. Graphite oxide was prepared by modified Hummers method from purified natural graphite sample from West Kalimantan. We demonstrated that natural graphite is well-purified by acid leaching method. The purified graphite was proceed for intercalating process by modifying Hummers method. The modification is on the reaction time and temperature of the intercalation process. The materials used in the intercalating process are H{sub 2}SO{sub 4} and KMNO{sub 4}. The purified natural graphite is analyzed by carbon content based on Loss on Ignition test. The thermo gravimetricanalysis and the Fouriertransform infrared spectroscopy are performed to investigate the oxidation results of the obtained GO which is indicated by the existence of functional groups. In addition, the X-ray diffraction and energy dispersive X-ray spectroscopy are also applied to characterize respectively for the crystal structure and elemental analysis. The results confirmed that natural graphite samples with 68% carbon content was purified into 97.68 % carbon content. While the intercalation process formed a formation of functional groups in the obtained GO. The results show that the temperature and reaction times have improved the efficiency of the oxidation process. It is concluded that these method could be considered as an important route for large-scale production of graphene.

  14. Strength degradation of oxidized graphite support column in VHTR

    International Nuclear Information System (INIS)

    Park, Byung Ha; No, Hee Cheon

    2010-01-01

    Air-ingress events caused by large pipe breaks are important accidents considered in the design of Very High Temperature Gas-Cooled Reactors (VHTRs). A main safety concern for this type of event is the possibility of core collapse following the failure of the graphite support column, which can be oxidized by ingressed air. In this study, the main target is to predict the strength of the oxidized graphite support column. Through compression tests for fresh and oxidized graphite columns, the compressive strength of IG-110 was obtained. The buckling strength of the IG-110 column is expressed using the following empirical straight-line formula: σ cr,buckling =91.34-1.01(L/r). Graphite oxidation in Zone 1 is volume reaction and that in Zone 3 is surface reaction. We notice that the ultimate strength of the graphite column oxidized in Zones 1 and 3 only depends on the slenderness ratio and bulk density. Its strength degradation oxidized in Zone 1 is expressed in the following nondimensional form: σ/σ 0 =exp(-kd), k=0.114. We found that the strength degradation of a graphite column, oxidized in Zone 3, follows the above buckling empirical formula as the slenderness of the column changes. (author)

  15. Graphite in Science and Nuclear Technique

    OpenAIRE

    Zhmurikov, E. I.; Bubnenkov, I. A.; Dremov, V. V.; Samarin, S. I.; Pokrovsky, A. S.; Harkov, D. V.

    2013-01-01

    The monograph is devoted to the application of graphite and graphite composites in science and technology. The structure and electrical properties, the technological aspects of production of high-strength synthetic graphites, the dynamics of the graphite destruction, traditionally used in the nuclear industry are discussed. It is focuses on the characteristics of graphitization and properties of graphite composites based on carbon isotope 13C. The book is based, generally, on the original res...

  16. Absolute analysis of uranium isotopic concentrations with a gas ion source mass spectrometer; Analyses absolues des concentrations isotopiques de l'uranium par spectrometre de masse equipe d'une source a gaz

    Energy Technology Data Exchange (ETDEWEB)

    Chaussy, L.; Boyer, R. [Commissariat a l' Energie Atomique, Pierrelatte (France)

    1969-07-01

    Mass spectrometer with electronic bombardment ions source for routine uranium isotopic analysis are used like relative measurements apparatus. We show that such mass spectrometers can be used for absolute measurements with a very high sensitivity and precision which are ten times better than theses of thermo-ionic ions source mass spectrometer. We examine the causes of systematic errors and we give experimental data. In particular natural uranium sample used as reference give: U{sub 5} = 0.7202 {+-} 0.0005 atoms per cent; U{sub 4} = 0.00552 {+-} 0.0003 atoms per cent. The use of this method is justified for standards control. (authors) [French] Les spectrometres de masse a source par bombardement electronique pour l'analyse de l'uranium sous forme d'hexafluorure, sont utilises en routine comme des appareils de mesure relative. On montre que l'on peut utiliser de tels appareils pour effectuer des mesures absolues avec une excellente sensibilite et reproductibilite, dix fois superieure a celle des spectrometres a source thermoionique. On examine en detail les causes d'erreurs systematiques et on donne des resultats experimentaux. En particulier, l'analyse d'un echantillon d'uranium naturel donne: U{sub 5} = 0.7202 {+-} 0.0005 atomes pour cent; U{sub 4} = 0.00552 {+-} 0.0003 atomes pour cent. La technique de mesure est utile pour le controle d'etalons isotopiques. (auteurs)

  17. EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE.

    Science.gov (United States)

    Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

  19. Uranium toxicology

    International Nuclear Information System (INIS)

    Ferreyra, Mariana D.; Suarez Mendez, Sebastian

    1997-01-01

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

  20. Gas

    International Nuclear Information System (INIS)

    1996-01-01

    The French government has decided to modify the conditions of extension of local natural gas authorities to neighbouring districts. The European Union is studying the conditions of internal gas market with the objective of more open markets although considering public service requirements

  1. Derivation of a radionuclide inventory for irradiated graphite-chlorine-36 inventory determination

    International Nuclear Information System (INIS)

    Brown, F.J.; Palmer, J.D.; Wood, P.

    2001-01-01

    The irradiation of materials in nuclear reactors results in neutron activation of component elements. Irradiated graphite wastes arise from their use in UK gas-cooled research and commercial reactor cores, and in fuel element components, where the graphite has acted as the neutron moderator. During irradiation the residual chlorine, which was used to purify the graphite during manufacture, is activated to chlorine-36. This isotope is long-lived and poorly retarded by geological barriers, and may therefore be a key radionuclide with respect to post-closure disposal facilities performance. United Kingdom Nirex Limited, currently responsible for the development of a disposal route for intermediate-level radioactive wastes in the UK, carried out a major research programme to support an overall assessment of the chlorine-36 activity of all wastes including graphite reactor components. The various UK gas cooled reactors reactors have used a range of graphite components made from diverse graphite types; this has necessitated a systematic programme to cover the wide range of graphite and production processes. The programme consisted of: precursor measurements - on the surface and/or bulk of representative samples of relevant materials, using specially developed methods; transfer studies - to quantify the potential for transfer of Cl-36 into and between waste streams during irradiation of graphite; theoretical assessments - to support the calculational methodology; actual measurements - to confirm the modelling. For graphite, a total of 458 measurements on samples from 57 batches were performed, to provide a detailed understanding of the composition of nuclear graphite. The work has resulted in the generation of probability density functions (PDF) for the mean chlorine concentration of three classes of graphite: fuel element graphite; Magnox moderator and reflector graphite and AGR reflector graphite; AGR moderator graphite. Transfer studies have shown that a significant

  2. Gravity Effects on the Free Vibration of a Graphite Column

    International Nuclear Information System (INIS)

    Ki, Dong-Ok; Kim, Jong-Bum; Park, Keun-Bae; Lee, Won-Jae

    2006-01-01

    The gravity effects on the free vibration of a graphite column are studied. Graphite block is a key component of a HTGR (High Temperature Gas Cooled Reactor). The major core elements, such as the fuel blocks and neutron reflector blocks, of HTTR (High Temperature Test Reactor, Japan) and GT-MHR (Gas Turbine- Modular Helium Reactor, USA) consist of stacked hexagonal graphite blocks forming a group of columns. The vibration of the columns induced by earthquakes may lead to solid impacts between graphite blocks and structural integrity problems. The study of free vibration characteristics of the graphite block column is the first step in the core internal structure dynamic analysis. Gravity force bring a negative stiffness term to the transverse vibration analysis of heavy long column structures, and results in natural frequency reductions. Generally it is not considered in the not so tall structure cases, because the gravity term makes the analysis and design complicated. Therefore it is important to check whether the gravity effect is severe or not

  3. Modification of structural graphite machining

    International Nuclear Information System (INIS)

    Lavrenev, M.M.

    1979-01-01

    Studied are machining procedures for structural graphites (GMZ, MG, MG-1, PPG) most widely used in industry, of the article mass being about 50 kg. Presented are dependences necessary for the calculation of cross sections of chip suction tappers and duster pipelines in machine shops for structural graphite machining

  4. Electron oxidation of graphite by fluorospecies

    International Nuclear Information System (INIS)

    Rosenthal, G.L.

    1984-09-01

    The fluoride-ion affinity (A/sub F - /) of phosphorus pentafluoride was determined to be 100 kcal/mole from the heats of reaction of the Lewis bases SF 4 and ClO 2 F with PF 5 near room temperature. The fluoride-ion affinity of boron trifluoride was determined to be 92 kcal/mole from the heat of reaction of ClO 2 F with BF 3 . The crystal structure of ClO 2 BF 4 was determined and a precise lattice energy was calculated from this structure and used to determined A/sub F - /. Both PF 5 and BF 3 were found to react with graphite in the presence of fluorine gas to yield a variety of non-stoichiometric compounds. The fluoride-ion affinity of silicon tetrafluoride is not known, but it does not react with graphite and F 2 except at high pressures. These and previous results suggested a threshold in oxidizing power of intercalating species below which the oxidative intercalation reaction would not occur. The reduction of C/sub x/PF 6 by PF 3 proved that the reaction is thermodynamically controlled to some extent. The displacement of PF 5 in C/sub x/PF 6 by BF 3 (with a smaller A/sub F - /) suggested that two BF 3 molecules may have a larger fluoride-ion affinity than one PF 5 and that B 2 F 7 - may be a stable anion in graphite. Conductivity studies of PF/sub x/ and BF/sub y/ salts showed that a large drop in conductivity when the reaction reaches first stage is due in the most part to direct fluorination of carbon in graphite

  5. Glass-Graphite Composite Materials

    International Nuclear Information System (INIS)

    Mayzan, M.Z.H.; Lloyd, J.W.; Heath, P.G.; Stennett, M.C.; Hyatt, N.C.; Hand, R.J.

    2016-01-01

    A summary is presented of investigations into the potential of producing glass-composite materials for the immobilisation of graphite or other carbonaceous materials arising from nuclear power generation. The methods are primarily based on the production of base glasses which are subsequently sintered with powdered graphite or simulant TRISO particles. Consideration is also given to the direct preparation of glass-graphite composite materials using microwave technology. Production of dense composite wasteforms with TRISO particles was more successful than with powdered graphite, as wasteforms containing larger amounts of graphite were resistant to densification and the glasses tried did not penetrate the pores under the pressureless conditions used. Based on the results obtained it is concluded that the production of dense glassgraphite composite wasteforms will require the application of pressure. (author)

  6. Pilot experiments with relativistic uranium projectile and fission fragments thermalized in a cryogenic gas-filled stopping cell

    Energy Technology Data Exchange (ETDEWEB)

    Reiter, Moritz Pascal

    2015-07-01

    online experiments, which is about three times larger than any stopping cell, using RF structures for the extraction of ions, has demonstrated. The area density and therefore the stopping power of the CSC is limited by the differential pumping. To overcome this limitation the CSC was tested with neon as a stopping gas with area densities of up to 11.3 mg/cm{sup 2} helium equivalent, demonstrating a unprecedented area density for stopping cells based on RF structures. The RF carpet performed reliably and its potential for the future FAIR stopping cell was shown. During the experiments at GSI the mean extraction time of {sup 221}Ac ions from the CSC to a silicon surface detector was measured, it amounts to 24 ms. This value is well in agreement with offline measurements using a pulsed {sup 223}Ra recoil ion source. The combination of a high density stopping cell with high total efficiencies and a non-scanning high-resolution mass spectrometer can be used as an independent identification detector for exotic nuclei by their mass, allowing a recalibration of the in-flight detectors of any fragment separator. As a proof-of-principal experiment the CSC and a MR-TOF-MS have been used as a mass tagger for the FRS at GSI. 134-I ions were produced by in-flight fission from an {sup 238}U primary beam at 1000 MeV/u and identified by the mass tagger. The new method does not rely on specific decay properties and therefore allows a recalibration of the fragment separator independent of the fragment and can also be used with stable nuclides. The usage of the CSC and a MR-TOF-MS will allow fast recalibration and a more effective usage of the limited amount of beam time for all experiments with exotic nuclei even in the case the nuclide of interest is not clearly identified by the in-flight detection scheme. With the CSC low energy experiments such as high-precision mass measurements and decay spectroscopy were made possible, the half lifes of {sup 221}Ac and {sup 223}Th have been measured

  7. Pilot experiments with relativistic uranium projectile and fission fragments thermalized in a cryogenic gas-filled stopping cell

    International Nuclear Information System (INIS)

    Reiter, Moritz Pascal

    2015-01-01

    experiments, which is about three times larger than any stopping cell, using RF structures for the extraction of ions, has demonstrated. The area density and therefore the stopping power of the CSC is limited by the differential pumping. To overcome this limitation the CSC was tested with neon as a stopping gas with area densities of up to 11.3 mg/cm"2 helium equivalent, demonstrating a unprecedented area density for stopping cells based on RF structures. The RF carpet performed reliably and its potential for the future FAIR stopping cell was shown. During the experiments at GSI the mean extraction time of "2"2"1Ac ions from the CSC to a silicon surface detector was measured, it amounts to 24 ms. This value is well in agreement with offline measurements using a pulsed "2"2"3Ra recoil ion source. The combination of a high density stopping cell with high total efficiencies and a non-scanning high-resolution mass spectrometer can be used as an independent identification detector for exotic nuclei by their mass, allowing a recalibration of the in-flight detectors of any fragment separator. As a proof-of-principal experiment the CSC and a MR-TOF-MS have been used as a mass tagger for the FRS at GSI. 134-I ions were produced by in-flight fission from an "2"3"8U primary beam at 1000 MeV/u and identified by the mass tagger. The new method does not rely on specific decay properties and therefore allows a recalibration of the fragment separator independent of the fragment and can also be used with stable nuclides. The usage of the CSC and a MR-TOF-MS will allow fast recalibration and a more effective usage of the limited amount of beam time for all experiments with exotic nuclei even in the case the nuclide of interest is not clearly identified by the in-flight detection scheme. With the CSC low energy experiments such as high-precision mass measurements and decay spectroscopy were made possible, the half lifes of "2"2"1Ac and "2"2"3Th have been measured, alpha spectroscopy of short

  8. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

    1964-01-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

  9. Rossing uranium

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

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

  10. Effect of reacting surface density on the overall graphite oxidation rate

    International Nuclear Information System (INIS)

    Oh, Chang; Kim, Eung; Lim, Jong; Schultz, Richard; Petti, David

    2009-01-01

    Graphite oxidation in an air-ingress accident is presently a very important issue for the reactor safety of the very high temperature gas cooled-reactor (VHTR), the concept of the next generation nuclear plant (NGNP) because of its potential problems such as mechanical degradation of the supporting graphite in the lower plenum of the VHTR might lead to core collapse if the countermeasure is taken carefully. The oxidation process of graphite has known to be affected by various factors, including temperature, pressure, oxygen concentration, types of graphite, graphite shape and size, flow distribution, etc. However, our recent study reveals that the internal pore characteristics play very important roles in the overall graphite oxidation rate. One of the main issues regarding graphite oxidation is the potential core collapse problem that may occur following the degradation of graphite mechanical strength. In analyzing this phenomenon, it is very important to understand the relationship between the degree of oxidization and strength degradation. In addition, the change of oxidation rate by graphite oxidation degree characterization by burn-off (ratio of the oxidized graphite density to the original density) should be quantified because graphite strength degradation is followed by graphite density decrease, which highly affects oxidation rates and patterns. Because the density change is proportional to the internal pore surface area, they should be quantified in advance. In order to understand the above issues, the following experiments were performed: (1) Experiment on the fracture of the oxidized graphite and validation of the previous correlations, (2) Experiment on the change of oxidation rate using graphite density and data collection, (3) Measure the BET surface area of the graphite. The experiments were performed using H451 (Great Lakes Carbon Corporation) and IG-110 (Toyo Tanso Co., Ltd) graphite. The reason for the use of those graphite materials is because

  11. Uranium, depleted uranium, biological effects

    International Nuclear Information System (INIS)

    2001-01-01

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

  12. Uranium loans

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

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

  13. Reactivity of hydrogen with uranium in the presence of Pt

    International Nuclear Information System (INIS)

    Balooch, M.; Siekhaus, W.J.

    1997-07-01

    The surface-reaction of di-hydrogen with uranium in the presence of Pt clusters has been studied using scanning tunneling microscopy (STM). Uranium was deposited on highly oriented pyrolytic graphite (HOPG) and annealed at temperatures up to 1200 degrees C to obtain atomically pyrolytic flat surfaces. Pt clusters were then formed using evaporation from a Pt source onto the surface and subsequent annealing. Hydrogen mainly attacked uranium in the vicinity of Pt clusters and formed hydride. The hydride formation probability is almost constant at 2.3x10 -4 over the range of exposures studied

  14. AEC determines uranium enrichment policy

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The Advisory Committee on Uranium Enrichment of the Atomic Energy Commission (AEC) has submitted a report to AEC chairman concerning the promotion of the introduction of advanced material, high performance centrifuges to replace conventional metallic drum centrifuges, and the development of next generation advanced centrifuges. The report also called for the postponement until around 1997 of the decision whether the development should be continued or not on atomic vapor laser isotope separation (AVLIS) and molecular laser isotope separation (MLIS) processes, as well as the virtual freezing of the construction of a chemical process demonstration plant. The report was approved by the AEC chairman in August. The uranium enrichment service market in the world will continue to be characterized by oversupply. The domestic situation of uranium enrichment supply-demand trend, progress of the expansion of Rokkasho enrichment plant, the trend in the development of gas centrifuge process and the basic philosophy of commercializing domestic uranium enrichment are reported. (K.I.)

  15. Instrumentation for the uranium mining industry

    International Nuclear Information System (INIS)

    Gray, R.A.

    1978-01-01

    Brief descriptions are presented concerning instruments used in uranium mining, including R meter, radon daughter working level counter, radon gas detectors, alpha contamination monitors, air samplers, ore grade evaluators and gamma energy analyzers

  16. Differences in the irradiation effects of IG-110 and IG-430 nuclear graphites : effects of coke difference

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Kim, Gen Chan; Kim, Eung Seon; Hong, Jin Ki; Chang, Jong Hwa

    2005-01-01

    In the high temperature gas cooled reactors (HTGRs), graphite acts as a moderator and reflector as well as a major structural component that may provide channels for the fuel and coolant gas, channels for control and shut down, and thermal and neutron shielding. During a reactor operation, many of the physical, chemical and mechanical properties of these graphite components are significantly modified as a function of the temperature, environment, and an irradiation. On the other hand, currently, all the nuclear graphites are being manufactured from two types of cokes, i.e., petroleum and coal-tar pitch coke, and it has been understood that the type of coke plays the most critical role determining the properties of a specific graphite grade. To investigate the effects of coke types on the irradiation response of a graphite, two graphites of different cokes were irradiated by 3 MeV C+ ions and the differences in the response of ion-irradiation were investigated

  17. Some equipment for graphite research in swimming pool reactors

    International Nuclear Information System (INIS)

    Seguin, M.; Arragon, Ph.; Dupont, G.; Gentil, J.; Tanis, G.

    1964-01-01

    The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [fr

  18. Acoustic emission from polycrystalline graphites

    International Nuclear Information System (INIS)

    Ioka, I.; Yoda, S.; Oku, T.; Miyamoto, Y.

    1987-01-01

    Acoustic emission was monitored from polycrystalline graphites with different microstructure (pore size and pore volume) subjected to compressive loading. The graphites used in this study comprised five brands, that is, PGX, ISEM-1, IG-11, IG-15, and ISO-88. A root mean square (RMS) voltage and event counts of acoustic emission for graphites were measured during compressive loading. The acoustic emission was measured using a computed-based data acquisition and analysis system. The graphites were first deformed up to 80 % of the average fracture stress, then unloaded and reloaded again until the fracture occured. During the first loading, the change in RMS voltage for acoustic emission was detected from the initial stage. During the unloading, the RMS voltage became zero level as soon as the applied stress was released and then gradually rose to a peak and declined. The behavior indicated that the reversed plastic deformation occured in graphites. During the second loading, the RMS voltage gently increased until the applied stress exceeded the maximum stress of the first loading; there is no Kaiser effect in the graphites. A bicrystal model could give a reasonable explanation of this results. The empirical equation between the ratio of σ AE to σ f and σ f was obtained. It is considered that the detection of microfracture by the acoustic emission technique is effective in macrofracture prediction of polycrystalline graphites. (author)

  19. Uranium extraction from gold-uranium ores

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

  20. Uranium mining

    International Nuclear Information System (INIS)

    2008-01-01

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

  1. Uranium enrichment in the United States

    International Nuclear Information System (INIS)

    Hill, J.H.; Parks, J.W.

    1975-01-01

    History, improvement programs, status of electrical power availability, demands for uranium enrichment, operating plan for the U. S. enriching facilities, working inventory of enriched uranium, possible factors affecting deviations in the operating plan, status of gaseous diffusion technology, status of U. S. gas centrifuge advances, transfer of enrichment technology, gaseous diffusion--gas centrifuge comparison, new enrichment capacity, U. S. separative work pricing, and investment in nuclear energy are discussed. (LK)

  2. From Core to Capture: Graphite Management by Gasification and Carbon Capture & Storage (CCS)

    International Nuclear Information System (INIS)

    Goodwin, J.; Bradbury, D.; Black, S.; Tomlinson, T.; Livesey, B.; Robinson, J.; Lindberg, M.; Newton, C.; Jones, A.; Wickham, A.

    2016-01-01

    Radioactive graphite waste arises principally from the moderators of graphite/gas-cooled reactors at the end of life of the reactors. Commercial power producing reactors (for example, Magnox, AGR and RBMK) have graphite moderators, each containing several thousand tonnes of graphite, with the UK having the largest inventory of over 90,000 tonnes. Additionally, there are smaller quantities of graphite arising from other sources such as fuel element components. The current long term strategy for management of reactor graphite in the UK is for these wastes to be conditioned for disposal followed by transfer to a geological disposal facility (GDF). With this baseline position, these wastes will account for about 30% of the ILW inventory in a GDF. As the volume of the graphite waste is so large, it is not currently economic to retrieve and process the graphite in advance of the availability of a geological disposal facility. Recent work by the NDA has ascribed a much smaller “incremental” volume of 2% due to graphite, calculated on the basis that the GDF has to be a certain size anyway in order to dissipate the decay heat from high level waste

  3. Chemisputtering of interstellar graphite grains

    International Nuclear Information System (INIS)

    Draine, B.T.

    1979-01-01

    The rate of erosion of interstellar graphite grains as a result of chemical reaction with H, N, and O is estimated using the available experiment evidence. It is argued that ''chemical sputtering'' yields for interstellar graphite grains will be much less than unity, contrary to earlier estimates by Barlow and Silk. Chemical sputtering of graphite grains in evolving H II regions is found to be unimportant, except in extremely compact (n/sub H/> or approx. =10 5 cm -3 ) H II regions. Alternative explanations are considered for the apparent weakness of the lambda=2175 A extinction ''bump'' in the direction of several early type stars

  4. Obtention of nuclear grade graphite

    International Nuclear Information System (INIS)

    Ferreira, M.L.

    1984-01-01

    The impurity level of natural graphite found in some of the most important mines of the State of Minas Gerais - Brasil is determined. It is also concerned with the development and use of natural graphite in nuclear reactors. Standard methods for chemical and instrumentsal analysis such as Spectrografic Determination by Emission, Spectrografic Determination by X-Rays, Spectrografic Determination by Atomic Asorption, Photometric Determination, and also chemical and physical methods for separation of impurities as well standard method for Estimating the Thermal Neutron Absorption Cross Section of graphite were employed. Some aditionals methods of purification to the ordinary treatment such as the use of metanol and halogens are also described. (Author) [pt

  5. 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

  6. Natural graphite demand and supply - Implications for electric vehicle battery requirements

    Science.gov (United States)

    Olson, Donald W.; Virta, Robert L.; Mahdavi, Mahbood; Sangine, Elizabeth S.; Fortier, Steven M.

    2016-01-01

    Electric vehicles have been promoted to reduce greenhouse gas emissions and lessen U.S. dependence on petroleum for transportation. Growth in U.S. sales of electric vehicles has been hindered by technical difficulties and the high cost of the lithium-ion batteries used to power many electric vehicles (more than 50% of the vehicle cost). Groundbreaking has begun for a lithium-ion battery factory in Nevada that, at capacity, could manufacture enough batteries to power 500,000 electric vehicles of various types and provide economies of scale to reduce the cost of batteries. Currently, primary synthetic graphite derived from petroleum coke is used in the anode of most lithium-ion batteries. An alternate may be the use of natural flake graphite, which would result in estimated graphite cost reductions of more than US$400 per vehicle at 2013 prices. Most natural flake graphite is sourced from China, the world's leading graphite producer. Sourcing natural flake graphite from deposits in North America could reduce raw material transportation costs and, given China's growing internal demand for flake graphite for its industries and ongoing environmental, labor, and mining issues, may ensure a more reliable and environmentally conscious supply of graphite. North America has flake graphite resources, and Canada is currently a producer, but most new mining projects in the United States require more than 10 yr to reach production, and demand could exceed supplies of flake graphite. Natural flake graphite may serve only to supplement synthetic graphite, at least for the short-term outlook.

  7. Enriched uranium recovery at Los Alamos

    International Nuclear Information System (INIS)

    Herrick, C.C.

    1984-01-01

    Graphite casting scrap, fuel elements and nongraphite combustibles are calcined to impure oxides. These materials along with zircaloy fuel elements and refractory solids are leach-dissolved separately in HF-HNO 3 acid to solubilize the contained enriched uranium. The resulting slurry is filtered and the clear filtrate (to which mineral acid solutions bearing enriched uranium may be added) are passed through solvent extraction. The solvent extraction product is filtered, precipitated with H 2 O 2 and the precipitate calcined to U 3 O 8 . Metal is made from U 3 O 8 by conversion to UO 2 , hydrofluorination and reduction to metal. Throughput is 150 to 900 kg uranium per year depending on the type of scrap

  8. A microstructural study of dynamic crack propagation in nuclear graphites

    International Nuclear Information System (INIS)

    Burchell, T.D.; McEnaney, B.; Tucker, M.O.; Rose, A.P.G.

    1986-01-01

    This paper reports a new microstructural study of dynamic crack propagation in three nuclear graphites: (i) PGA, the moderator material in UK Magnox reactors; (ii) IMl-24, the moderator material in UK Advanced gas cooled reactors (AGR); and (iii) a pitch coke graphite, which is used in the fabrication of AGR fuel sleeves. The fracture mechanisms in nuclear graphites are initiated by microcrack formation at low stresses. Typically, microcracks form in regions of well-aligned binder or at favourably-oriented pores, where stress is concentrated. With increasing applied loads, microcracks propagate taking advantage of easy cleavage paths or linking with pores. Eventually, coalescence of such cracks and inherent porosity produces a crack of critical length for fast fracture. (orig./MM)

  9. Uranium enrichment

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  10. Hydrogen pumping and release by graphite under high flux plasma bombardment

    International Nuclear Information System (INIS)

    Hirooka, Y.; Leung, W.K.; Conn, R.W.; Goebel, D.M.; Labombard, B.; Nygren, R.; Wilson, K.L.

    1988-01-01

    Inert gas (helium or argon) plasma bombardment has been found to increase the surface gas adsorptivity of isotropic graphite (POCO-graphite), which can then getter residual gases in a high vacuum system. The inert gas plasma bombardment was carried out at a flux ∼ 1 x 10 18 ions s -1 cm -2 to a fluence of the order of 10 21 ions/cm 2 and at temperatures around 800 degree C. The plasma bombarding energy was varied between 100 and 200 eV. The gettering speed of the activated graphite surface is estimated to be as large as 25 liters s -1 cm -2 at total pressures between 10 -6 and 10 -7 torr. The gettering capacity estimated is 0.025 torr-liter/cm 2 at room temperature. The gettering capability of graphite can be easily recovered by repeating inert gas plasma bombardment. The activated graphite surface exhibits a smooth, sponge-like morphology with significantly increased pore openings, which correlates with the observed increase in the surface gas adsorptivity. The activated graphite surface has been observed to pump hydrogen plasma particles as well. From calibrated H-alpha measurements, the dynamic hydrogen retention capacity is evaluated to be as large as 2 x 10 18 H/cm 2 at temperatures below 100 degree C and at a plasma bombarding energy of 300 eV

  11. Analysis of beryllium and depleted uranium: An overview of detection methods in aerosols and soils

    International Nuclear Information System (INIS)

    Camins, I.; Shinn, J.H.

    1988-06-01

    We conducted a survey of commercially available methods for analysis of beryllium and depleted uranium in aerosols and soils to find a reliable, cost-effective, and sufficiently precise method for researchers involved in environmental testing at the Yuma Proving Ground, Yuma, Arizona. Criteria used for evaluation include cost, method of analysis, specificity, sensitivity, reproducibility, applicability, and commercial availability. We found that atomic absorption spectrometry with graphite furnace meets these criteria for testing samples for beryllium. We found that this method can also be used to test samples for depleted uranium. However, atomic absorption with graphite furnace is not as sensitive a measurement method for depleted uranium as it is for beryllium, so we recommend that quality control of depleted uranium analysis be maintained by testing 10 of every 1000 samples by neutron activation analysis. We also evaluated 45 companies and institutions that provide analyses of beryllium and depleted uranium. 5 refs., 1 tab

  12. Deposition of tantalum carbide coatings on graphite by laser interactions

    Science.gov (United States)

    Veligdan, James; Branch, D.; Vanier, P. E.; Barietta, R. E.

    1994-01-01

    Graphite surfaces can be hardened and protected from erosion by hydrogen at high temperatures by refractory metal carbide coatings, which are usually prepared by chemical vapor deposition (CVD) or chemical vapor reaction (CVR) methods. These techniques rely on heating the substrate to a temperature where a volatile metal halide decomposes and reacts with either a hydrocarbon gas or with carbon from the substrate. For CVR techniques, deposition temperatures must be in excess of 2000 C in order to achieve favorable deposition kinetics. In an effort to lower the bulk substrate deposition temperature, the use of laser interactions with both the substrate and the metal halide deposition gas has been employed. Initial testing involved the use of a CO2 laser to heat the surface of a graphite substrate and a KrF excimer laser to accomplish a photodecomposition of TaCl5 gas near the substrate. The results of preliminary experiments using these techniques are described.

  13. Investigation on wear behavior of graphite baII under different pneumatic conveying environments

    International Nuclear Information System (INIS)

    Chen Zhipeng; Zheng Yanhua; Shi Lei; Yu Suyuan

    2014-01-01

    An experimental platform was built in the Institute of Nuclear and New Energy Technology (INET) to investigate the wear behavior of the graphite ball under the operational condition of the high temperature gas-cooled reactor (HTGR) fuel handling system. In this experimental platform, a series of experiments were carried out under different pneumatic conveying environments with the graphite balls, which were made of the material same as the fuel element matrix graphite (A3) of the 10 MW high temperature gas cooled reactor (HTR-10). The effect of the pneumatic conveying condition on the wear rate of graphite ball has been investigated, and the results include: (1) There is an obvious linear relationship between the wear rate and the feeding velocity of graphite ball elevated in the stainless steel elevating tube, and the wear rate will increase with the increase of the feeding velocity. (2) The wear rate of graphite ball under helium environment is significantly greater than that under air and nitrogen environments, which is caused by the different effects of various gas environments on mechanical properties of graphite. (author)

  14. Graphite in Science and Nuclear Technology

    OpenAIRE

    Zhmurikov, Evgenij

    2015-01-01

    This review is devoted to the application of graphite and graphite composites in the science and technology. Structure and electrical properties, technological aspects of producing of high-strength artificial graphite and dynamics of its destruction are considered. These type of graphite are traditionally used in the nuclear industry, so author concentrates on actual problems of application and testing of graphite materials in modern science and technology. Translated from chapters 1 of monog...

  15. Mesostructure of graphite composite and its lifetime

    OpenAIRE

    Zhmurikov, Evgenij

    2015-01-01

    This review is devoted to the application of graphite and graphite composites in science and technology. Structure and electrical properties, as so technological aspects of producing of high strength artificial graphite and dynamics of its destruction are considered. These type of graphite are traditionally used in the nuclear industry. Generally, the review relies, on the original results and concentrates on actual problems of application and testing of graphite materials in modern nuclear p...

  16. Thermodynamic data for uranium fluorides

    Energy Technology Data Exchange (ETDEWEB)

    Leitnaker, J.M.

    1983-03-01

    Self-consistent thermodynamic data have been tabulated for uranium fluorides between UF/sub 4/ and UF/sub 6/, including UF/sub 4/ (solid and gas), U/sub 4/F/sub 17/ (solid), U/sub 2/F/sub 9/ (solid), UF/sub 5/ (solid and gas), U/sub 2/F/sub 10/ (gas), and UF/sub 6/ (solid, liquid, and gas). Included are thermal function - the heat capacity, enthalpy, and free energy function, heats of formation, and vaporization behavior.

  17. Graphite surveillance in N Reactor

    International Nuclear Information System (INIS)

    Woodruff, E.M.

    1991-09-01

    Graphite dimensional changes in N Reactor during its 24 yr operating history are reviewed. Test irradiation results, block measurements, stack profiles, top of reflector motion monitors, and visual observations of distortion are described. 18 refs., 14 figs., 1 tab

  18. The French emirs of uranium

    International Nuclear Information System (INIS)

    Lucas, N.

    1978-01-01

    The French energy plan is described under the following heads: structure of supply/use; nuclear energy; oil consumption; coal consumption; gas consumption; energy conservation; renewable resources. The section on nuclear energy covers: nuclear programmes, uranium supplies (mining and enrichment), fuel fabrication and reprocessing, finance and economics, political and administrative aspects, nuclear trade, development of fast breeder reactor. (U.K.)

  19. Airborne uranium, its concentration and toxicity in uranium enrichment facilities

    International Nuclear Information System (INIS)

    Thomas, J.; Mauro, J.; Ryniker, J.; Fellman, R.

    1979-02-01

    The release of uranium hexafluoride and its hydrolysis products into the work environment of a plant for enriching uranium by means of gas centrifuges is discussed. The maximum permissible mass and curie concentration of airborne uranium (U) is identified as a function of the enrichment level (i.e., U-235/total U), and chemical and physical form. A discussion of the chemical and radiological toxicity of uranium as a function of enrichment and chemical form is included. The toxicity of products of UF 6 hydrolysis in the atmosphere, namely, UO 2 F 2 and HF, the particle size of toxic particulate material produced from this hydrolysis, and the toxic effects of HF and other potential fluoride compounds are also discussed. Results of an investigation of known effects of humidity and temperature on particle size of UO 2 F 2 produced by the reaction of UF 6 with water vapor in the air are reported. The relationship of the solubility of uranium compounds to their toxic effects was studied. Identification and discussion of the standards potentially applicable to airborne uranium compounds in the working environment are presented. The effectiveness of High Efficiency Particulate (HEPA) filters subjected to the corrosive environment imposed by the presence of hydrogen fluoride is discussed

  20. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

    The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  1. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Grodzki Marcin

    2017-12-01

    Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  2. Uranium update

    International Nuclear Information System (INIS)

    Steane, R.

    1997-01-01

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

  3. A reverse method for the determination of the radiological inventory of irradiated graphite at reactor scale

    Energy Technology Data Exchange (ETDEWEB)

    Nicaise, Gregory [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-roses (France); Poncet, Bernard [EDF-DP2D, Lyon (France)

    2016-11-15

    Irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France (EDF). Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Because of graphite high purity, radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. Therefore the assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A reverse method using an identification calculation-measurement process is proposed in order to assess the radionuclide inventory as precisely as possible.

  4. Machining of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Morris, T.O.

    1981-01-01

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

  5. Geology and exploration of the Rum Jungle Uranium Field

    International Nuclear Information System (INIS)

    Fraser, W.J.

    1980-01-01

    The Rum Jungle Uranium Field was discovered by a private prospector in 1949. A total of 3530 tonnes of uranium oxide was mined and treated from four ore-bodies by Territory Enterprises Pty. Limited who managed the Rum Jungle Project on behalf of the Australian Atomic Energy Commission until the closure of operations in 1971. One small low grade uranium orebody remains to be developed. Lead, zinc, copper, cobalt and nickel were found zoned sub vertically with uranium at one deposit. One medium sized lead, zinc, copper, cobalt and nickel deposit remains to be developed and one small copper deposit with minor uranium was mined. The basemetal deposits show a regional zoning relationship with the known uranium mineralization. Uranium and basemetal mineralization is hosted by graphitic or chloritic, pyritic shales at the contact with a magnesite. These rocks are in the lower part of a sequence of Lower Proterozoic sediments which unconformably overlie Archaean basement complexes. The sediments and complexes are displaced by Giants Reef Fault and sub-parallel shears and linears may further control mineralization. Nearly 50km of the prospective shale/magnesite contact was tested by total count radiometric surveys, various electrical methods, auger, rotary percussion and diamond drilling. The source for the uranium mineralization was probably the Archaean basement complexes from which uranium was initially deposited as protore by either chemical precipitation or clay adsorption in the shale units or as detrital placers in quartz pebble conglomerates immediately overlying the basement complexes. (author)

  6. Hydrogen pumping and release by graphite under high flux plasma bombardment

    International Nuclear Information System (INIS)

    Hirooka, Y.; Leung, W.K.; Conn, R.W.; Goebel, D.M.; LaBombard, B.; Nygren, R.; Wilson, K.L.

    1988-01-01

    Inert gas (helium or argon) plasma bombardment has been found to increase the surface gas adsorptivity of isotropic graphite (POCO-graphite), which can then getter residual gases in a high vacuum system. The inert gas plasma bombardment was carried out at a flux ≅ 1 x 10 18 ions s -1 cm -2 to a fluence of the order of 10 21 ions/cm 2 and at temperatures around 800 0 C. The gettering capability of graphite can be easily recovered by repeating inert gas plasma bombardment. The activated graphite surface exhibits a smooth, sponge-like morphology with significantly increased pore openings, which correlates with the observed increase in the surface gas adsorptivity. The activated graphite surface has been observed to pump hydrogen plasma particles as well. From calibrated H-alpha measurements, the dynamic hydrogen retention capacity is evaluated to be as large as 2 x 10 18 H/cm 2 at temperatures below 100 0 C and at a plasma bombarding energy of 300 eV. The graphite temperature was varied between 15 and 480 0 C. Due to the plasma particle pumping capability, hydrogen recycling from the activated graphite surface is significantly reduced, relative to that from a pre-saturated surface. A pre-saturated surface was also observed to reproducibly pump a hydrogen plasma to a concentration of 9.5 x 10 17 H/cm 2 . The hydrogen retention capacity of graphite is found to decrease with increasing temperature. A transient pumping mechanism associated with the sponge-like surface morphology is conjectured to explain the large hydrogen retention capacity. Hydrogen release behavior under helium and argon plasma bombardment was also investigated, and the result indicated the possibility of some in-pore retrapping effect. 43 refs., 11 figs

  7. Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

    International Nuclear Information System (INIS)

    Košťál, Michal; Rypar, Vojtěch; Milčák, Ján; Juříček, Vlastimil; Losa, Evžen; Forget, Benoit; Harper, Sterling

    2016-01-01

    Highlights: • A light water critical facility for graphite reactivity worth measurements. • Comparison of calculated and measured k eff . • Effect of graphite description on k eff . - Abstract: Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations.

  8. Fluid inclusion and oxygen isotope studies of the Nabarlek and Jabiluka uranium deposits, Northern Territory, Australia

    International Nuclear Information System (INIS)

    Ypma, P.J.M.; Fuzikawa, K.

    1980-01-01

    We lack a basic understanding of the solutions producing the uranium deposits of the Alligator Rivers Uranium Field (ARUF). Several theories have been proposed ranging from syngenetic, epigenetic hydrothermal, epigenetic metamorphogenic, surficial origin (Ferguson et al., this volume), and mobilization by evaporite deposits. As for a precipitation mechanism, we do not seem to find much beyond the presence of graphite in some ore-bearing and intra-formational strata, and pre-uranium sulphides, none of which reducing factors are common throughout all ore bodies. This study was initiated with the aim of obtaining direct fluid inclusion evidence of the solution transport and precipitation of uranium

  9. Uranium mining

    International Nuclear Information System (INIS)

    Cheeseman, E.W.

    1980-01-01

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

  10. Uranium metallogenic model related to CO2 and hydrocarbon in granite type uranium deposits

    International Nuclear Information System (INIS)

    Ou Guangxi; Chen Anfu; Cui Jianyong; Xu Yinhuan; Wang Chunhua; Xu Yan

    2001-01-01

    The report is concerned with the inseparable connections between the uranium migration, enrichment rule and the geochemical characteristics of CO 2 and hydrocarbon gas, as well as the relations between the deposit locations and the gas abnormal distribution in rocky body, which are based on the analysis of some data and phenomena in 11 typical deposits in 2 granite type uranium ore fields, including the observations of 250 rocky fluid inclusion sections and the analyzed data of which 2470 are in gas composition, 200 in uranium content, 50 in thermometry. All the conclusions are drawn from different angles for the first time and this new exploration and advancement fills up the blank of gas geochemistry study in uranium deposits or other metal deposits

  11. Uranium industry annual 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

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

  12. Uranium industry annual 1996

    International Nuclear Information System (INIS)

    1997-04-01

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

  13. Uranium industry annual, 1991

    International Nuclear Information System (INIS)

    1992-10-01

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

  14. Detection of uranium enrichment activities using environmental monitoring techniques

    International Nuclear Information System (INIS)

    Belew, W.L.; Carter, J.A.; Smith, D.H.; Walker, R.L.

    1993-01-01

    Uranium enrichment processes have the capability of producing weapons-grade material in the form of highly enriched uranium. Thus, detection of undeclared uranium enrichment activities is an international safeguards concern. The uranium separation technologies currently in use employ UF 6 gas as a separation medium, and trace quantities of enriched uranium are inevitably released to the environment from these facilities. The isotopic content of uranium in the vegetation, soil, and water near the plant site will be altered by these releases and can provide a signature for detecting the presence of enriched uranium activities. This paper discusses environmental sampling and analytical procedures that have been used for the detection of uranium enrichment facilities and possible safeguards applications of these techniques

  15. Verification and validation of the THYTAN code for the graphite oxidation analysis in the HTGR systems

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi

    2014-12-01

    The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V and V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement. (author)

  16. Some equipment for graphite research in swimming pool reactors; Quelques dispositifs d'etude du graphite dans les piles piscines

    Energy Technology Data Exchange (ETDEWEB)

    Seguin, M; Arragon, Ph; Dupont, G; Gentil, J; Tanis, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [French] Les dispositifs d'irradiation decrits servent aux etudes relatives a la filiere des reacteurs a uranium naturel, moderes au graphite et refroidis par le gaz carbonique. Ils sont generalement concus pour etre utilises dans des piles piscines. L'accent a ete mis sur: - l'utilisation au maximum du volume d'irradiation, - le recours aux solutions technologiques les plus simples, - la standardisation de certaines parties constitutives. Cette standardisation impose un usinage precis et un montage soigne, lesquels sont egalement necessaires lorsqu'on doit obtenir une temperature d'irradiation relativement basse alors que l'echauffement nucleaire est important. Enfin, la conception de ces dispositifs est valable pour irradier d'autres materiaux non fissiles ou fissiles. (auteurs)

  17. Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.

  18. The latest figures on uranium

    International Nuclear Information System (INIS)

    Vance, R.

    2010-01-01

    According to the latest figures on uranium, soon to be published by the NEA, uranium resources, production and demand are all on the rise. Exploration efforts have increased recently in line with the expected expansion of nuclear energy in the coming years. Total identified resources have grown and are now sufficient to cover 100 years of supply at 2008 rates of consumption. Costs of production have, however, also increased. This article is based on the latest edition of the 'Red Book', Uranium 2009: Resources, Production and Demand, which presents the results of the most recent biennial review of world uranium market fundamentals and a statistical profile of the world uranium industry as of 1 January 2009. It contains official data provided by OECD Nuclear Energy Agency (NEA) and International Atomic Energy Agency (IAEA) member countries on uranium exploration, resources, production and reactor-related requirements. Projections of nuclear generating capacity and reactor-related uranium requirements through 2035 are also provided as well as a discussion of long-term uranium supply and demand issues. Despite recent declines stemming from the global financial crisis, world demand for electricity is expected to continue to grow significantly over the next several decades to meet the needs of an increasing population and economic growth. The recognition by an increasing number of governments that nuclear power can produce competitively priced, base-load electricity that is essentially free of greenhouse gas emissions, coupled with the role that nuclear can play in enhancing security of energy supply, increases the prospects for growth in nuclear generating capacity, although the magnitude of that growth remains to be determined. Regardless of the role that nuclear energy ultimately plays in meeting rising electricity demand, the uranium resource base is more than adequate to meet projected requirements. Meeting even high-case requirements to 2035 would consume less

  19. Uranium and thorium migration under dislocative metamorphism

    International Nuclear Information System (INIS)

    Titov, V.K.; Bilibina, T.V.; Dashkova, A.D.; Il'in, V.K.; Makarova, L.I.; Shmuraeva, L.Ya.

    1978-01-01

    Investigated were peculiarities of uranium and thorium behaviour in the process of dislocation metamorphism on the basis of regional fracture zones of early-proterozoic embedding of Ukrainian, Aldan and Baltic shields. The studied zones correspond to tectonite of green-shale and almandin-amphibolite facies of regional metamorphism according to mineral associations. The most peculiar feature of the tectonites of green-shale facies is uranium presence in migrationally able forms, which can be involved afterwards into the ore process by hydrothermal solutions. Adsorved forms of uranium on the crystal surface or separate grains and in the cracks, as well as microinclusions of uranium minerals, selectively timed to mineral structure defects prevail among easily mobile uranium compounds. Dissolved uranium is present, evidently in gas-liquid inclusions in minerals and pore waters. There forms of uranium presence are peculiar for epidote-chlorite mylonites, as well as cataclasites and diaphthorites related to them by blastomylonites of almandin-amphibolite facies. Wide range of manifestation of this process, caused by multikilometer extension of deep fracture zones permit to consider the formations of green-shale facies of dislocation metamorphism as one of the main uranium sources in deposit formation in different uranium-ore associations different age

  20. Gasket for uranium enrichment plant

    Energy Technology Data Exchange (ETDEWEB)

    Kishi, S; Aiyoshi, H

    1977-02-08

    A gasket to be inserted between flange joints in the equipments and pipe lines of an uranium enrichment plant having neither permeability nor adsorptivity to water while maintaining mechanical, physical and chemical properties of an elastomer gasket is described. A gasket made of an elastomeric material such as a polymer is integratedly formed at its surface with anti-slip projections. The gasket is further surrounded at its upper and lower peripheral sides, as well as outer circumferential portion with a U-sectioned cover (enclosure) made of fluoro-plastics. In this arrangement, the gasket main body shows a gas-tightness for uranium hexafluoride gas and the cover exhibits a gas-tightness for other component gases such as moisture to thereby prevent degradation of the gasket due to absorption and permeation of the moisture.

  1. Computer simulation of the steam--graphite reaction under isothermal and steady-state conditions

    International Nuclear Information System (INIS)

    Joy, D.S.; Stem, S.C.

    1975-05-01

    A mathematical model was formulated to describe the isothermal, steady-state diffusion and reaction of steam in a graphite matrix. A generalized Langmuir-Hinshelwood equation is used to represent the steam-graphite reaction rate. The model also includes diffusion in the gas phase adjacent to the graphite matrix. A computer program, written to numerically integrate the resulting differential equations, is described. The coupled nonlinear differential equations in the graphite phase are solved using the IBM Continuous System Modeling Program. Classical finite difference techniques are used for the gas-phase calculations. An iterative procedure is required to couple the two sets of calculations. Several sample problems are presented to demonstrate the utility of the model. (U.S.)

  2. Proserpine - plutonium 239 - Proserpine - uranium 235 - comparison of experimental results; Proserpine - plutonium 239 - proserpine - uranium 235 - comparaison de resultats experimentaux

    Energy Technology Data Exchange (ETDEWEB)

    Brunet, J P; Caizergues, R; Clouet D' Orval, Ch; Kremser, J; Moret-Bailly, J; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The Proserpine homogeneous reactor is constituted by a tank, 25 cm dia, 30 cm high, surrounded by a composite reflector made of beryllium oxide and graphite. In this tank can be made critical plutonium or 90 per cent enriched uranium solutions, the fissile substances being in the form of a dissolved salt. In varying the concentration of the solution, critical masses were studied as a function of the level of the liquid in the tank. The minimum critical mass is 256 {+-} 2 grs for plutonium and 409 {+-} 3 grs for uranium 235. In the range of the critical concentrations which were studied, the neutronic properties of fissionable solutions of plutonium and enriched uranium were compared for identical geometries. (authors) [French] Proserpine est un reacteur homogene comportant une cuve de diametre 25 cm, de hauteur 30 cm, entouree d'un reflecteur composite d'oxyde de beryllium et de graphite. On y a rendu critiques des solutions de plutonium ou d'uranium enrichi a 90 pour cent, le produit fissile se trouvant sous la forme d'un sel dissous. En faisant varier la concentration de la solution, on a etudie les masses critiques en fonction de la hauteur du liquide dans la cuve. La masse- critique minimum est, pour le plutonium de 256 {+-} 2 g, pour l'uranium 235 de 409 {+-} 3 g. Dans la gamme des concentrations critiques etudiees, on a compare, dans des conditions de geometrie identique, les proprietes neutroniques des solutions fissiles de plutonium et d'uranium enrichi. (auteurs)

  3. Oxidation of PCEA nuclear graphite by low water concentrations in helium

    Energy Technology Data Exchange (ETDEWEB)

    Contescu, Cristian I., E-mail: ContescuCI@ornl.gov [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6087 (United States); Mee, Robert W. [Department of Business Analytics and Statistics, University of Tennessee, Knoxville, TN 37996-0525 (United States); Wang, Peng [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6087 (United States); Romanova, Anna V.; Burchell, Timothy D. [Department of Business Analytics and Statistics, University of Tennessee, Knoxville, TN 37996-0525 (United States)

    2014-10-15

    Accelerated oxidation tests were performed to determine kinetic parameters of the chronic oxidation reaction (i.e. slow, continuous, and persistent) of PCEA graphite in contact with helium coolant containing low moisture concentrations in high temperature gas-cooled reactors. To the authors’ knowledge such a study has not been done since the detailed analysis of reaction of H-451 graphite with steam (Velasquez, Hightower, Burnette, 1978). Since that H-451 graphite is now unavailable, it is urgently needed to characterize chronic oxidation behavior of new graphite grades that are being considered for use in gas-cooled reactors. The Langmuir–Hinshelwood mechanism of carbon oxidation by water results in a non-linear reaction rate expression, with at least six different parameters. They were determined in accelerated oxidation experiments that covered a large range of temperatures (800–1100 °C), and partial pressures of water (15–850 Pa) and hydrogen (30–150 Pa) and used graphite specimens thin enough (4 mm) in order to avoid diffusion effects. Data analysis employed a statistical method based on multiple likelihood estimation of parameters and simultaneous fitting of non-linear equations. The results show significant material-specific differences between graphite grades PCEA and H-451 which were attributed to microstructural dissimilarity between the two materials. It is concluded that kinetic data cannot be transferred from one graphite grade to another.

  4. Argon/UF6 plasma exhaust gas reconstitution experiments using preheated fluorine and on-line diagnostics. [fissioning uranium plasma core reactor design

    Science.gov (United States)

    Roman, W. C.

    1979-01-01

    The feasibility of employing a flowing, high-temperature, pure fluorine/UF6 regeneration system to efficiently convert a large fraction of the effluent plasma exhaust back to pure UF6 was demonstrated. The custom built T.O.F. mass spectrometer sampling system permitted on-line measurements of the UF6 concentration at different locations in the exhaust system. Negligible amounts ( 100 ppm) of UF6 were detected in the axial bypass exhaust duct and the exhaust ducts downstream of the cryogenic trap system used to collect the UF6, thus verifying the overall system efficiency over a range of operating conditions. Use of a porous Monel duct as part of the exhaust duct system, including provision for injection of pure fluorine, provided a viable technique to eliminate uranium compound residue on the inside surface of the exhaust ducts. Typical uranium compound mass deposition per unit area of duct was 2 micron g/sq cm. This porous duct technique is directly applicable to future uranium compound transfer exhaust systems. Throughout these experiments, additional basic data on the corrosion aspects of hot, pressurized UF6/fluorine were also accumulated.

  5. Experience with graphite in JET

    International Nuclear Information System (INIS)

    Pick, M.A.; Celentano, G.; Deksnis, E.; Dietz, K.J.; Shaw, R.; Sonnenberg, K.; Walravens, M.

    1987-01-01

    During the current operational period of JET more than 50% of the internal area of the machine is covered in graphite tiles. This includes the 15 m 2 of carbon tiles installed in the new toroidal limiter, the 40 poloidal belts of graphite tiles covering the U-joints and bellows as well as a two metre high ring (-- 20 m 2 ) or carbon tiles on the inner wall of the Torus. A ring of tiles in the equatorial plane (3 tiles high) consists of carbon-carbon fibre tiles. Test bed results indicated that the fine grained graphite tiles cracked at ∼ 1 kW/cm 2 for 2s of irradiation whereas the carbon-carbon fibre tiles were able to sustain a flux, limited by the irradiation facility, of 3.5 kW for 3s without any damage. The authors report on the generally positive experience they have had had with the installed graphite during the present and previous in-vessel configurations. This includes the physical integrity of the tiles under severe conditions such as high energy run-away electron beams, plasma disruptions and high heat fluxes. They report on the importance of the precise positioning of the inner wall and x-point tiles at the very high power fluxes of JET and the effect of deviations on both graphite and carbon-fibre tiles

  6. Porous (Swiss-Cheese Graphite

    Directory of Open Access Journals (Sweden)

    Joseph P. Abrahamson

    2018-05-01

    Full Text Available Porous graphite was prepared without the use of template by rapidly heating the carbonization products from mixtures of anthracene, fluorene, and pyrene with a CO2 laser. Rapid CO2 laser heating at a rate of 1.8 × 106 °C/s vaporizes out the fluorene-pyrene derived pitch while annealing the anthracene coke. The resulting structure is that of graphite with 100 nm spherical pores. The graphitizablity of the porous material is the same as pure anthracene coke. Transmission electron microscopy revealed that the interfaces between graphitic layers and the pore walls are unimpeded. Traditional furnace annealing does not result in the porous structure as the heating rates are too slow to vaporize out the pitch, thereby illustrating the advantage of fast thermal processing. The resultant porous graphite was prelithiated and used as an anode in lithium ion capacitors. The porous graphite when lithiated had a specific capacity of 200 mAh/g at 100 mA/g. The assembled lithium ion capacitor demonstrated an energy density as high as 75 Wh/kg when cycled between 2.2 V and 4.2 V.

  7. Experiments with HEU (93.14 wt.%) metal annuli with internal graphite cylinder

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiaobo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wehmann, Udo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, only three experimental configurations are described here. They are internal graphite reflected metal uranium assemblies with three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches). These experiments can be found in Reference 1 and in their associated logbook

  8. Strategy and perspective for uranium exploration in Egypt

    International Nuclear Information System (INIS)

    Hassan, M.A.; Salman, A.B.; Assaf, H.S.; Mahdy, M.M.

    1995-01-01

    Uranium exploration started in Egypt about three decades ago. This was performed by applying integrated airborne and ground radiometric prospecting. The latter was conducted upon selected areas having rather favorable geological criteria. These activities resulted in the discovery of great numbers of radiometric anomalies, with several uranium occurrences in various geologic environments in granitic and sedimentary rocks. Some of these uranium occurrences show good potential for developing into workable uranium deposits. Small-scale exploratory tunnelling and drilling works have been carried out at some of these occurrences. Leaching studies and pilot experiments were carried out on technological samples to evaluate ore's suitability for uranium extraction. However, no assured reserves of uranium have been reached yet. The demands for uranium to satisfy the near future Egyptian nuclear power generation necessitates some development in the national strategy for uranium exploration. This will be achieved through intense programmes for ground geophysics and drilling from surface and underground mining works, in addition to radon emanometry and logging of oil and gas wells. Moreover, non conventional procedures for uranium extraction such as heap-leaching may be followed to exploit small-scale uranium deposits. In this developed strategy, the present uranium occurrences are modellized and categorized following the IAEA classification. The characteristics of the present uranium occurrences will be utilized in prospecting new areas. Subsidiary resources in phosphorites, black sands and rare metal deposits could supply additional quantities of uranium, in addition to thorium and rare earth elements. (author). 34 refs, 4 figs, 1 tab

  9. The characteristics of TiC and oxidation resistance and mechanical properties of TiC coated graphite under corrosive environment

    International Nuclear Information System (INIS)

    Yoda, Shinichi; Oku, Tatsuo; Ioka, Ikuo; Umekawa, Shokichi.

    1982-07-01

    Core region of the Very High Temperature Gas Cooled Reactor (VHTR) consists mainly of polycrystalline graphite whose mechanical properties degradated by corrosion resulting from such impurities as O 2 , H 2 O, and CO 2 in coolant He gas. Mechanical properties and oxidation resistance of TiC coated graphite under corrosive condition were examined in order to evaluate the effects of TiC coating on preventing the graphite from its degradation in service condition of the VHTR. Characteristics of TiC coating was also examined using EPMA. Holding the specimen at 1373 K for 6 hr produced strong interface between TiC coating and the graphite, however, microcracks on TiC coating was observed, the origin of which is ascribed to mismatch in thermal expansion between TiC coating and the graphite. Oxidation rate of TiC coated graphite was one-thirds of that of uncoated graphite, which demonstrated that TiC coating on the graphite improved the oxidation resistance of the graphite. However, debonding of TiC coating layer at the interface was observed after heating for 3 to 4 hr in the oxidation condition. Changes in Young's modulus of TiC coated graphite were a half of that of uncoated graphite. Flexural strength of TiC coated graphite remained at the original value up to about 4 hr oxidation, therafter it decreased abruptly as was the trend of uncoated graphite. It is concluded that TiC coating on graphite materials is very effective in improving oxidation resistance and suppressing degradation of mechanical properties of the graphite. (author)

  10. The chemical industry of uranium in France

    International Nuclear Information System (INIS)

    Goldschmidt, B.

    1955-01-01

    The actual CEA program is concerned with the construction of two large graphite reactors, each of those containing at least one hundred tons of uranium metal with nuclear purity. The uranium for these two reactors will be regularly supplied by new resources discovered in France and Madagascar in the last five years. The working and treatment of such ore have led to the creation of an important french industry of which the general outline and principle are described. The operated ores have got different natures and concentration, individual characteristics are described for the main ores.The most high-grade ore are transported to a central plant in Bouchet near Paris; the low-grade ore are concentrated by physical methods or chemical processes of which principles and economy are studied with constancy. The acid processes are the only used until now, although the carbonated alkaline processes has been studied in France. The next following steps after the acid process until the obtention of uranium rich concentrate are described. The purification steps of uranium compounds to nuclear purity material are described as well as the steps to elaborate metal of which the purity grade will be specify. Finally, the economic aspects of uranium production difficulty will be considered in relation with technical progresses which we can expect to achieve in the future. (M.P.)

  11. Production of nanodiamonds by high-energy ion irradiation of graphite at room temperature

    International Nuclear Information System (INIS)

    Daulton, T.L.; Kirk, M.A.; Lewis, R.S.; Rehn, L.E.

    2001-01-01

    It has previously been shown that graphite can be transformed into diamond by MeV electron and ion irradiation at temperatures above approximately 600 deg. C. However, there exists geological evidence suggesting that carbonaceous materials can be transformed to diamond by irradiation at substantially lower temperatures. For example, submicron-size diamond aggregates have been found in uranium-rich, Precambrian carbonaceous deposits that never experienced high temperature or pressure. To test if diamonds can be formed at lower irradiation temperatures, sheets of fine-grain polycrystalline graphite were bombarded at 20 deg. C with 350±50 MeV Kr ions to fluences of 6x10 12 cm -2 using the Argonne tandem linear accelerator system (ATLAS). Ion-irradiated (and unirradiated control) graphite specimens were then subjected to acid dissolution treatments to remove untransformed graphite and isolate diamonds that were produced; these acid residues were subsequently characterized by high-resolution and analytical electron microscopy. The acid residue of the ion-irradiated graphite was found to contain nanodiamonds, demonstrating that ion irradiation of graphite at ambient temperature can produce diamond. The diamond yield under our irradiation conditions is low, ∼0.01 diamonds/ion. An important observation that emerges from comparing the present result with previous observations of diamond formation during irradiation is that nanodiamonds form under a surprisingly wide range of irradiation conditions. This propensity may be related to the very small difference in the graphite and diamond free-energies coupled with surface-energy considerations that may alter the relative stability of diamond and graphite at nanometer sizes

  12. Uranium - what role

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  13. Brazilian uranium exploration program

    International Nuclear Information System (INIS)

    Marques, J.P.M.

    1981-01-01

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

  14. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-11-01

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

  15. Uranium production

    International Nuclear Information System (INIS)

    Spriggs, M.

    1980-01-01

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

  16. Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release; Traitements thermiques de l`oxyde d`uranium irradie dans un reacteur a eau pressurisee (R.E.P.): gonflement et relachement des gaz de fission

    Energy Technology Data Exchange (ETDEWEB)

    Zacharie, I

    1997-03-27

    In order to keep pressurised water reactors at a top level of safety, it is necessary to understand the chemical and mechanical interaction between the cladding and the fuel pellet due to a temperature increase during a rapid change in reactor. In this process, the swelling of uranium oxide plays an important role. It comes from a bubble precipitation of fission gases which are released when they are in contact with the outside. Therefore, the aim of this thesis consists in acquiring a better understanding of the mechanisms which come into play. Uranium oxide samples, from a two cycles irradiated fuel, first have been thermal treated between 1000 deg C and 1700 deg C for 5 minutes to ten hours. The gas release amount related to time has been measured for each treatment. The comparison of the experimental results with a numerical model has proved satisfactory: it seems that the gases release, after the formation of intergranular tunnels, is controlled by the diffusion phenomena. Afterwards, the swelling was measured on the samples. The microscopic examination shows that the bubbles are located in the grain boundaries and have a lenticular shape. The swelling can be explained by the bubbles coalescence and a model was developed based on this observation. An equation allows to calculate the intergranular swelling in function of time and temperature. The study gives the opportunity to predict the fission gases behaviour during a fuel temperature increase. (author) 56 refs.

  17. Thermal Pyrolytic Graphite Enhanced Components

    Science.gov (United States)

    Hardesty, Robert E. (Inventor)

    2015-01-01

    A thermally conductive composite material, a thermal transfer device made of the material, and a method for making the material are disclosed. Apertures or depressions are formed in aluminum or aluminum alloy. Plugs are formed of thermal pyrolytic graphite. An amount of silicon sufficient for liquid interface diffusion bonding is applied, for example by vapor deposition or use of aluminum silicon alloy foil. The plugs are inserted in the apertures or depressions. Bonding energy is applied, for example by applying pressure and heat using a hot isostatic press. The thermal pyrolytic graphite, aluminum or aluminum alloy and silicon form a eutectic alloy. As a result, the plugs are bonded into the apertures or depressions. The composite material can be machined to produce finished devices such as the thermal transfer device. Thermally conductive planes of the thermal pyrolytic graphite plugs may be aligned in parallel to present a thermal conduction path.

  18. Description of the magnox type of gas cooled reactor (MAGNOX)

    International Nuclear Information System (INIS)

    Jensen, S.E.; Nonboel, E.

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO 2 ) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  19. Friction anisotropy in boronated graphite

    International Nuclear Information System (INIS)

    Kumar, N.; Radhika, R.; Kozakov, A.T.; Pandian, R.; Chakravarty, S.; Ravindran, T.R.; Dash, S.; Tyagi, A.K.

    2015-01-01

    Graphical abstract: - Highlights: • Friction anisotropy in boronated graphite is observed in macroscopic sliding condition. • Low friction coefficient is observed in basal plane and becomes high in prismatic direction. • 3D phase of boronated graphite transformed into 2D structure after friction test. • Chemical activity is high in prismatic plane forming strong bonds between the sliding interfaces. - Abstract: Anisotropic friction behavior in macroscopic scale was observed in boronated graphite. Depending upon sliding speed and normal loads, this value was found to be in the range 0.1–0.35 in the direction of basal plane and becomes high 0.2–0.8 in prismatic face. Grazing-incidence X-ray diffraction analysis shows prominent reflection of (0 0 2) plane at basal and prismatic directions of boronated graphite. However, in both the wear tracks (1 1 0) plane become prominent and this transformation is induced by frictional energy. The structural transformation in wear tracks is supported by micro-Raman analysis which revealed that 3D phase of boronated graphite converted into a disordered 2D lattice structure. Thus, the structural aspect of disorder is similar in both the wear tracks and graphite transfer layers. Therefore, the crystallographic aspect is not adequate to explain anisotropic friction behavior. Results of X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy shows weak signature of oxygen complexes and functional groups in wear track of basal plane while these species dominate in prismatic direction. Abundance of these functional groups in prismatic plane indicates availability of chemically active sites tends to forming strong bonds between the sliding interfaces which eventually increases friction coefficient

  20. Friction anisotropy in boronated graphite

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, N., E-mail: niranjan@igcar.gov.in [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Radhika, R. [Crystal Growth Centre, Anna University, Chennai (India); Kozakov, A.T. [Research Institute of Physics, Southern Federal University, Rostov-on-Don (Russian Federation); Pandian, R. [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Chakravarty, S. [UGC-DAE CSR, Kalpakkam (India); Ravindran, T.R.; Dash, S.; Tyagi, A.K. [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2015-01-01

    Graphical abstract: - Highlights: • Friction anisotropy in boronated graphite is observed in macroscopic sliding condition. • Low friction coefficient is observed in basal plane and becomes high in prismatic direction. • 3D phase of boronated graphite transformed into 2D structure after friction test. • Chemical activity is high in prismatic plane forming strong bonds between the sliding interfaces. - Abstract: Anisotropic friction behavior in macroscopic scale was observed in boronated graphite. Depending upon sliding speed and normal loads, this value was found to be in the range 0.1–0.35 in the direction of basal plane and becomes high 0.2–0.8 in prismatic face. Grazing-incidence X-ray diffraction analysis shows prominent reflection of (0 0 2) plane at basal and prismatic directions of boronated graphite. However, in both the wear tracks (1 1 0) plane become prominent and this transformation is induced by frictional energy. The structural transformation in wear tracks is supported by micro-Raman analysis which revealed that 3D phase of boronated graphite converted into a disordered 2D lattice structure. Thus, the structural aspect of disorder is similar in both the wear tracks and graphite transfer layers. Therefore, the crystallographic aspect is not adequate to explain anisotropic friction behavior. Results of X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy shows weak signature of oxygen complexes and functional groups in wear track of basal plane while these species dominate in prismatic direction. Abundance of these functional groups in prismatic plane indicates availability of chemically active sites tends to forming strong bonds between the sliding interfaces which eventually increases friction coefficient.